WorldWideScience

Sample records for include direct reactor

  1. Directions for improved fusion reactors

    International Nuclear Information System (INIS)

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

    1986-01-01

    Conceptual fusion reactor studies over the past 10 to 15 years have projected systems that may be too large, complex, and costly to be of commercial interest. One main direction for improved fusion reactors points towards smaller, higher-power-density approaches. First-order economic issues (i.e., unit direct cost and cost of electricity) are used to support the need for more compact fusion reactors. A generic fusion physics/engineering/costing model is used to provide a quantiative basis for these arguments for specific fusion concepts

  2. Catalyst support structure, catalyst including the structure, reactor including a catalyst, and methods of forming same

    Science.gov (United States)

    Van Norman, Staci A.; Aston, Victoria J.; Weimer, Alan W.

    2017-05-09

    Structures, catalysts, and reactors suitable for use for a variety of applications, including gas-to-liquid and coal-to-liquid processes and methods of forming the structures, catalysts, and reactors are disclosed. The catalyst material can be deposited onto an inner wall of a microtubular reactor and/or onto porous tungsten support structures using atomic layer deposition techniques.

  3. Direct conversion nuclear reactor space power systems

    International Nuclear Information System (INIS)

    Britt, E.J.; Fitzpatrick, G.O.

    1982-01-01

    This paper presents the results of a study of space nuclear reactor power systems using either thermoelectric or thermionic energy converters. An in-core reactor design and two heat pipe cooled out-of-core reactor designs were considered. One of the out-of-core cases utilized, long heat pipes (LHP) directly coupled to the energy converter. The second utilized a larger number of smaller heat pipes (mini-pipe) radiatively coupled to the energy converter. In all cases the entire system, including power conditioning, was constrained to be launched in a single shuttle flight. Assuming presently available performance, both the LHP thermoelectric system and minipipe thermionic system, designed to produce 100 kWe for seven years, would have a specific mass near 22kg/kWe. The specific mass of the thermionic minipipe system designed for a one year mission is 165 kg/kWe due to less fuel swelling. Shuttle imposed growth limits are near 300 kWe and 1.2 MWe for the thermoelectric and thermionic systems, respectively. Converter performance improvements could double this potential, and over 10 MWe may be possible for very short missions

  4. Direct Energy Conversion for Fast Reactors

    International Nuclear Information System (INIS)

    Brown, N.; Cooper, J.; Vogt, D.; Chapline, G.; Turchi, P.; Barbee Jr., T.; Farmer, J.

    2000-01-01

    Strategic Computing Initiative (ASCI), should improve the speed and decrease the cost of developing new TEGs. The system concept to be evaluated is shown in Figure 1. Liquid metal is used to transport heat away from the nuclear heat source and to the TEG. Air or liquid (water or a liquid metal) is used to transport heat away from the cold side of the TEG. Typical reactor coolants include sodium or eutectic mixtures of lead-bismuth. These are coolants that have been used to cool fast neutron reactors. Heat from the liquid metal coolant is rejected through the thermal electric materials, thereby producing electrical power directly. The temperature gradient could extend from as high as 1300 K to 300 K, although fast reactor structural materials (including those used to clad the fuel) currently used limit the high temperature to about 825K

  5. Development Directions For CANDU and Slowpoke Reactors

    International Nuclear Information System (INIS)

    Brooks, Gordon L.

    1990-01-01

    This paper provides a broader-based discussion of overall development directions foreseen for CANDU reactors, particularly those which have further evolved sine the earlier paper. The paper then discusses development directions for the Slowpokes Energy System which is a small nuclear heat source intended to meet local heating needs for building complexes and municipal heating systems. In evolving a sound development direction, it is necessary, firstly, to address the question of requirements, viz., what are the requirements which future nuclear power plants must satisfy if they are to be successful? Today, some in the nuclear industry believe that the most important of such requirements relates to the need for 'safer' reactors. Indeed, some proponents of this view would seem to suggest that if only we could develop such 'safer' reactors, suddenly all of our problem s with public acceptance would disappear and utilities would form long lines waiting to purchase such marvellous machines. I do not share such a simplistic view nor, indeed, do many of my colleagues in the international nuclear power industry

  6. I. Reactor safety (including comments on criticisms of WASH-1400)

    International Nuclear Information System (INIS)

    1976-01-01

    A major concern in any nuclear power programme is a reactor accident resulting in a large release of radioactivity to the environment. Serious reactor accidents are possible and the risk of such accidents cannot be reduced to zero i.e. absolute safety cannot be assured. All that can be expected is that the measures used to ensure safety in the design and operation of a reactor are such that the risk of accident is reduced to acceptably low levels. No member of the general public is known to have died or been injured as a result of an accident in over 1000 commercial nuclear power reactor-years. Some accidents in power reactors in operation today have come close enough to an environmental release of radioactivity to cause serious public concern about future safety. Apparent inadequacies in safety practices disclosed by former members of the nuclear power industry have added to this concern. To obtain an objective appraisal of the reactor safety issue this report examines the measures taken in the design and operation of nuclear reactors to reduce the probability of accident to acceptably low levels

  7. Evaluation of fatigue data including reactor water environmental effects

    International Nuclear Information System (INIS)

    Rosinski, S.T.; Nickell, R.E.; Van Der Sluys, W.A.; Yukawa, S.

    2002-01-01

    Laboratory data have been gathered in the past decade indicating a significant reduction in component fatigue life when reactor water environmental effects are experimentally simulated. However, these laboratory data have not been supported by nuclear power plant component operating experience. The laboratory data under simulated operating conditions are being used to support arguments for revising the design-basis fatigue curves in the ASME Code Section III, Division 1, for Class 1 components. A thorough review of available laboratory fatigue data and their applicability to actual component operating conditions was performed. The evaluation divided the assembly, review and assessment of existing laboratory fatigue data and its applicability to plant operating conditions into four principal tasks: (1) review of available laboratory data relative to thresholds for environmental parameters, such as temperature, reactor water oxidation potential, strain rate, strain amplitude, reactor water flow rate, and component metal sulfur content; (2) determination of the relevance of the laboratory data to actual plant operating conditions; (3) review of laboratory S-N data curve-fitting models; and (4) assessment of existing ASME Code Section III Class 1 margins This paper summarizes the results of the data review. In addition, recommendations are made for additional laboratory testing intended to improve the applicability of laboratory test results under simulated reactor water environmental conditions. (authors)

  8. Reactor protection system including engineered features actuation system

    International Nuclear Information System (INIS)

    Palmaers, W.

    1982-01-01

    The safety concept requires to ensure that - the reactor protection system - the active engineered safeguard - and the necessary auxiliary systems are so designed and interfaced in respect of design and mode of action that, in the event of single component failure reliable control of the consequences of accidents remains ensured at all times and that the availability of the power plant is not limited unnecessarily. In order to satisfy these requirements due, importance was attached to a consistent spacial separation of the mutually redundant subsystems of the active safety equipment. The design and layout of the reactor protection system, of the power supply (emergency power supply), and of the auxiliary systems important from the safety engineering point of view, are such that their subsystems also largely satisfy the requirements of independence and spacial separation. (orig./RW)

  9. Future directions in boiling water reactor design

    International Nuclear Information System (INIS)

    Wilkins, D.R.; Hucik, S.A.; Duncan, J.D.; Sweeney, J.I.

    1987-01-01

    The Advanced Boiling Water Reactor (ABWR) is being developed by an international team of BWR manufacturers to respond to worldwide utility needs in the 1990's. Major objectives of the ABWR program are design simplification; improved safety and reliability; reduced construction, fuel and operating costs; improved maneuver-ability; and reduced occupational exposure and radwaste. The ABWR incorporates the best proven features from BWR designs in Europe, Japan and the United States and application of leading edge technology. Key features of the ABWR are internal recirculation pumps; fine-motion, electrohydraulic control rod drives; digital control and instrumentation; multiplexed, fiber optic cabling netwoek; pressure suppression containment with horizontal vents; cylindrical reinforced concrete containment; structural integration of the containment and reactor building; severe accident capability; state-of-the-art fuel; advanced trubine/generator with 52'' last stage buckets; and advanced radwaste technology. The ABWR is ready for lead plant application in Japan, where it is being developed as the next generation Japan standard BWR under the guidance and leadership of The Tokyo Electric Power Company, Inc. and a group of Japanese BWR utilities. In the United States it is being adapted to the needs of US utilities through the Electric Power Research Institute's Advanced LWR Requirements Program, and is being reviewed by the US Nuclear Regulatory Commission for certification as a preapproved US standard BWR under the US Department of Energy's ALWR Design Verification Program. These cooperative Japanese and US programs are expected to establish the ABWR as a world class BWR for the 1990's...... (author)

  10. Nuclear reactor melt-retention structure to mitigate direct containment heating

    International Nuclear Information System (INIS)

    Tutu, N.K.; Ginsberg, T.; Klages, J.R.

    1991-01-01

    This patent describes a nuclear reactor melt-retention structure that functions to retain molten core material within a melt retention chamber to mitigate the extent of direct containment heating. The structure being adapted to be positioned within or adjacent to a pressurized or boiling water nuclear reactor containment building at a location such that at least a portion of the melt retention structure is lower than and to one side of the nuclear reactor pressure vessel, and such that the structure is adjacent to a gas escape channel means that communicates between the reactor cavity and the containment building of the reactor. It comprises a melt-retention chamber, wall means defining a passageway extending between the reactor cavity underneath the reactor pressure vessel and one side of the chamber, the passageway including vent means extending through an upper wall portion thereof. The vent means being in communication with the upper region of the reactor containment building, whereby gas and steam discharged from the reactor pressure vessel are vented through the passageway and vent means into the gas-escape channel means and the reactor containment building

  11. Direct energy conversion for fusion reactors

    International Nuclear Information System (INIS)

    Barr, W.L.

    1977-01-01

    Complex multistage plasma converters were tested at efficiencies approaching 90% at low energies and powers, and simpler, more cost-effective versions at 65% efficiency. Laboratory tests of neutral-beam direct converters at 15 keV and 2 kW gave 70% efficiency. A 120-keV, 1.5-MW version is being tested

  12. High power density reactors based on direct cooled particle beds

    Science.gov (United States)

    Powell, J. R.; Horn, F. L.

    Reactors based on direct cooled High Temperature Gas Cooled Reactor (HTGR) type particle fuel are described. The small diameter particle fuel is packed between concentric porous cylinders to make annular fuel elements, with the inlet coolant gas flowing inwards. Hot exit gas flows out along the central channel of each element. Because of the very large heat transfer area in the packed beds, power densities in particle bed reactors (PBRs) are extremely high resulting in compact, lightweight systems. Coolant exit temperatures are high, because of the ceramic fuel temperature capabilities, and the reactors can be ramped to full power and temperature very rapidly. PBR systems can generate very high burst power levels using open cycle hydrogen coolant, or high continuous powers using closed cycle helium coolant. PBR technology is described and development requirements assessed.

  13. High power density reactors based on direct cooled particle beds

    International Nuclear Information System (INIS)

    Powell, J.R.; Horn, F.L.

    1985-01-01

    Reactors based on direct cooled HTGR type particle fuel are described. The small diameter particle fuel is packed between concentric porous cylinders to make annular fuel elements, with the inlet coolant gas flowing inwards. Hot exit gas flows out long the central channel of each element. Because of the very large heat transfer area in the packed beds, power densities in particle bed reactors (PBR's) are extremely high resulting in compact, lightweight systems. Coolant exit temperatures are high, because of the ceramic fuel temperature capabilities, and the reactors can be ramped to full power and temperature very rapidly. PBR systems can generate very high burst power levels using open cycle hydrogen coolant, or high continuous powers using closed cycle helium coolant. PBR technology is described and development requirements assessed. 12 figs

  14. Joining and fabrication techniques for high temperature structures including the first wall in fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Ho Jin; Lee, B. S.; Kim, K. B

    2003-09-01

    The materials for PFC's (Plasma Facing Components) in a fusion reactor are severely irradiated with fusion products in facing the high temperature plasma during the operation. The refractory materials can be maintained their excellent properties in severe operating condition by lowering surface temperature by bonding them to the high thermal conducting materials of heat sink. Hence, the joining and bonding techniques between dissimilar materials is considered to be important in case of the fusion reactor or nuclear reactor which is operated at high temperature. The first wall in the fusion reactor is heated to approximately 1000 .deg. C and irradiated severely by the plasma. In ITER, beryllium is expected as the primary armour candidate for the PFC's; other candidates including W, Mo, SiC, B4C, C/C and Si{sub 3}N{sub 4}. Since the heat affected zones in the PFC's processed by conventional welding are reported to have embrittlement and degradation in the sever operation condition, both brazing and diffusion bonding are being considered as prime candidates for the joining technique. In this report, both the materials including ceramics and the fabrication techniques including joining technique between dissimilar materials for PFC's are described. The described joining technique between the refractory materials and the dissimilar materials may be applicable for the fusion reactor and Generation-4 future nuclear reactor which are operated at high temperature and high irradiation.

  15. Joining and fabrication techniques for high temperature structures including the first wall in fusion reactor

    International Nuclear Information System (INIS)

    Lee, Ho Jin; Lee, B. S.; Kim, K. B.

    2003-09-01

    The materials for PFC's (Plasma Facing Components) in a fusion reactor are severely irradiated with fusion products in facing the high temperature plasma during the operation. The refractory materials can be maintained their excellent properties in severe operating condition by lowering surface temperature by bonding them to the high thermal conducting materials of heat sink. Hence, the joining and bonding techniques between dissimilar materials is considered to be important in case of the fusion reactor or nuclear reactor which is operated at high temperature. The first wall in the fusion reactor is heated to approximately 1000 .deg. C and irradiated severely by the plasma. In ITER, beryllium is expected as the primary armour candidate for the PFC's; other candidates including W, Mo, SiC, B4C, C/C and Si 3 N 4 . Since the heat affected zones in the PFC's processed by conventional welding are reported to have embrittlement and degradation in the sever operation condition, both brazing and diffusion bonding are being considered as prime candidates for the joining technique. In this report, both the materials including ceramics and the fabrication techniques including joining technique between dissimilar materials for PFC's are described. The described joining technique between the refractory materials and the dissimilar materials may be applicable for the fusion reactor and Generation-4 future nuclear reactor which are operated at high temperature and high irradiation

  16. The direct conversion of heat into electricity in reactors

    International Nuclear Information System (INIS)

    Devin, B.; Bliaux, J.; Lesueur, R.

    1964-01-01

    The direct conversion of heat into electricity by thermionic emission in an atomic reactor has been studied with the triple aim of its utilisation: as an energy source for a space device, at the head of a conventional conversion system in power installations, or finally in association with the thermoelectric conversion in very low power installations. The laboratory experiments were mainly orientated towards the electron extraction of metals and compounds and their behaviour at high temperatures. Converters furnishing up to 50 amps at 0. 4 volts with an efficiency close to 10 p. 100 have been constructed in the laboratory; the emitters were heated by electron bombardment and were composed of tungsten covered with an uranium carbide deposit or molybdenum covered with cesium. The main aspects of the coupling between the converter and the reactor have been covered from the point of view of electronics: the influence of the mismatching of the load on the temperature of the emitter and the influence of thermal flux density on the temperature of the emitter and the stability of the converter. Converters using uranium carbide as the electron emitter have been tested in reactors. Tests have been made under dynamic conditions in order to determine the dynamic characteristics. The load matching curves have been constructed and the overall performances of several cells coupled in such a way as to form a reactor rod have been deduced. This information is fundamental to the design of a control system for a thermionic conversion reactor. The problems associated with the reliability of thermionic converters connected in series in the same reactor rod have been examined theoretically. Finally, the absorption isotherms have been drawn at the ambient temperatures for krypton and xenon on activated carbon with the aim of investigating the escape of fission products in a converter. (author) [fr

  17. Thermoelectric direct energy conversion system for fast reactors

    International Nuclear Information System (INIS)

    Kambe, Mitsuru

    1995-01-01

    A concept of direct energy conversion system for fast reactors has been presented by using FGM thermoelectric (TE) cell elements combined with FGM compliant pads based on the assumption that energy conversion efficiency of 20% could be achieved. The design involves TE energy conversion modules in which 360 TE cell elements are assembled. These energy conversion modules are connected to sodium and water circuits by cesium heat pipes and water heat pipes, respectively. The following design approach has been demonstrated. 1) Approximately 4100 energy conversion modules installed in a 150 MWt fast reactor can affords 27 MW of electricity. 2) An energy conversion building (single floor, 15 m x 15 m) enables to eliminate intermediate heat exchangers, steam generators and sodium-water reaction counter measures. (author)

  18. Review of direct energy conversion for fusion reactors

    International Nuclear Information System (INIS)

    Barr, W.L.; Moir, R.W.

    1976-01-01

    The direct conversion to electrical energy of the energy carried by the leakage plasma from a fusion reactor and by the ions that are not converted to neutrals in a neutral-beam injector is discussed. The conversion process is electrostatic deceleration and direct particle collection as distinct from plasma expansion against a time-varying magnetic field or conversion in an EXB duct (both MHD). Relatively simple 1-stage plasma direct converters are discussed which can have efficiencies of about 50 percent. More complex and costly (measured in $/kW) 2-, 3-, 4-, and 22-stage concepts have been tested at efficiencies approaching 90 percent. Beam direct converters have been tested at 15 keV and 2 kW of power at 70 +- 2 percent efficiency, and a test of a 120-keV, 1-MW version is being prepared. Designs for a 120-keV, 4-MW unit are presented. The beam direct converter, besides saving on power supplies and on beam dumps, should raise the efficiency of creating a neutral beam from 40 percent without direct conversion to 70 percent with direct conversion for a 120-keV deuterium beam. The technological limits determining power handling and lifetime such as space-charge effects, heat removal, electrode material, sputtering, blistering, voltage holding, and insulation design, are discussed. The application of plasma direct converters to toroidal plasma confinement concepts is also discussed

  19. Direct vessel inclined injection system for reduction of emergency core coolant direct bypass in advanced reactors

    International Nuclear Information System (INIS)

    Yoon, Sang H.; Lee, Jong G.; Suh, Kune Y.

    2006-01-01

    Multidimensional thermal hydraulics in the APR1400 (Advanced Power Reactor 1400 MWe) downcomer during a large-break loss-of-coolant accident (LBLOCA) plays a pivotal role in determining the capability of the safety injection system. APR1400 adopts the direct vessel injection (DVI) method for more effective core penetration of the emergency core cooling (ECC) water than the cold leg injection (CLI) method in the OPR1000 (Optimized Power Reactor 1000 MWe). The DVI method turned out to be prone to occasionally lack in efficacious delivery of ECC to the reactor core during the reflood phase of a LBLOCA, however. This study intends to demonstrate a direct vessel inclined injection (DVII) method, one of various ideas with which to maximize the ECC core penetration and to minimize the direct bypass through the break during the reflood phase of a LBLOCA. The 1/7 scaled down THETA (Transient Hydrodynamics Engineering Test Apparatus) tests show that a vertical inclined nozzle angle of the DVII system increases the downward momentum of the injected ECC water by reducing the degree of impingement on the reactor downcomer, whereby lessening the extent of the direct bypass through the break. The proposed method may be combined with other innovative measures with which to ensure an enough thermal margin in the core during the course of a LBLOCA in APR1400

  20. DIRECT-CYCLE, BOILING-WATER NUCLEAR REACTOR

    Science.gov (United States)

    Harrer, J.M.; Fromm, L.W. Jr.; Kolba, V.M.

    1962-08-14

    A direct-cycle boiling-water nuclear reactor is described that employs a closed vessel and a plurality of fuel assemblies, each comprising an outer tube closed at its lower end, an inner tube, fuel rods in the space between the tubes and within the inner tube. A body of water lying within the pressure vessel and outside the fuel assemblies is converted to saturated steam, which enters each fuel assembly at the top and is converted to superheated steam in the fuel assembly while it is passing therethrough first downward through the space between the inner and outer tubes of the fuel assembly and then upward through the inner tube. (AEC)

  1. Improved best estimate plus uncertainty methodology, including advanced validation concepts, to license evolving nuclear reactors

    International Nuclear Information System (INIS)

    Unal, C.; Williams, B.; Hemez, F.; Atamturktur, S.H.; McClure, P.

    2011-01-01

    Research highlights: → The best estimate plus uncertainty methodology (BEPU) is one option in the licensing of nuclear reactors. → The challenges for extending the BEPU method for fuel qualification for an advanced reactor fuel are primarily driven by schedule, the need for data, and the sufficiency of the data. → In this paper we develop an extended BEPU methodology that can potentially be used to address these new challenges in the design and licensing of advanced nuclear reactors. → The main components of the proposed methodology are verification, validation, calibration, and uncertainty quantification. → The methodology includes a formalism to quantify an adequate level of validation (predictive maturity) with respect to existing data, so that required new testing can be minimized, saving cost by demonstrating that further testing will not enhance the quality of the predictive tools. - Abstract: Many evolving nuclear energy technologies use advanced predictive multiscale, multiphysics modeling and simulation (M and S) capabilities to reduce the cost and schedule of design and licensing. Historically, the role of experiments has been as a primary tool for the design and understanding of nuclear system behavior, while M and S played the subordinate role of supporting experiments. In the new era of multiscale, multiphysics computational-based technology development, this role has been reversed. The experiments will still be needed, but they will be performed at different scales to calibrate and validate the models leading to predictive simulations for design and licensing. Minimizing the required number of validation experiments produces cost and time savings. The use of multiscale, multiphysics models introduces challenges in validating these predictive tools - traditional methodologies will have to be modified to address these challenges. This paper gives the basic aspects of a methodology that can potentially be used to address these new challenges in

  2. Direct energy conversion of radiation energy in fusion reactor

    International Nuclear Information System (INIS)

    Yamaguchi, S.; Iiyoshi, A.; Motojima, O.; Okamoto, M.; Sudo, S.; Ohnishi, M.; Onozuka, M.; Uenosono, C.

    1993-11-01

    Direct energy conversion from plasma heat flux has been studied. Since major parts of fusion energy in the advanced fusion reactor are radiation and charged particle energies, the flexible design of the blanket is possible. We discuss the potentiality of the thermoelectric element that generates electricity by temperature gradient in conductors. A strong magnetic field is used to confine the fusion plasma, therefore, it is appropriate to consider the effect of the magnetic field. We propose a new element which is called Nernst element. The new element needs the magnetic field and the temperature gradient. We compare the efficiency of these two elements in a semiconductor model. Finally, a direct energy conversion are mentioned. (author)

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

    International Nuclear Information System (INIS)

    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

  4. Future directions of the USNRC reactor safety research program

    International Nuclear Information System (INIS)

    Morrison, D.L.; King, T.L.

    1997-01-01

    Changes in technical issues and available budgets have caused shifts in the reactor safety research program of the U.S. Nuclear Regulatory Commission (NRC). Issues related to maintenance of expertise and facilities, increased cooperation (domestically and internationally) and priority of programs are influencing the program's future directions. To provide the technical basis to ensure that there is adequate protection of public health and safety in the operation of nuclear power plants, the priorities within NRC's research program in the near future are to provide the technical basis for identifying and addressing safety issues related to plant aging, license renewal and operations, and to support the transition to risk-informed, performance-based regulation

  5. MOTHER MK II: An advanced direct cycle high temperature gas reactor

    International Nuclear Information System (INIS)

    Hart, R.S.; Kendall, J.M.; Marsden, B.J.

    2003-01-01

    The MOTHER (MOdular Thermal HElium Reactor) power plant concepts employ high temperature gas reactors utilizing TRISO fuel, graphite moderator, and helium coolant, in combination with a direct Brayton cycle for electricity generation. The helium coolant from the reactor vessel passes through a Power Conversion Unit (PCU), which includes a turbine-generator, recuperator, precooler, intercooler and turbine-compressors, before being returned to the reactor vessel. The PCU substitutes for the reactor coolant system pumps and steam generators and most of the Balance Of Plant (BOP), including the steam turbines and condensers, employed by conventional nuclear power plants utilizing water cooled reactors. This provides a compact, efficient, and relatively simple plant configuration. The MOTHER MK I conceptual design, completed in the 1987 - 1989 time frame, was developed to economically meet the energy demands for extracting and processing heavy oil from the tar sands of western Canada. However, considerable effort was made to maximize the market potential beyond this application. Consistent with the remote and very high labour rate environment in the tar sands region, simplification of maintenance procedures and facilitation of 'change-out' in lieu of in situ repair was a design focus. MOTHER MK I had a thermal output of 288 MW and produced 120 MW electrical when operated in the electricity only production mode. An annular Prismatic reactor core was utilized, largely to minimize day-to-day operations activities. Key features of the power conversion system included two Power Conversion Units (144 MW th each), the horizontal orientation of all rotating machinery and major heat exchangers axes, high speed rotating machinery (17,030 rpm for the turbine-compressors and 10,200 rpm for the power turbine-generator), gas (helium) bearings for all rotating machinery, and solid state frequency conversion from 170 cps (at full power) to the grid frequency. Recognizing that the on

  6. Annexes to the lecture on reactor protection system including engineered features actuation system

    International Nuclear Information System (INIS)

    Palmaers, W.

    1982-01-01

    The present paper deals with the fundamentals for a reactor protection system and discusses the following topics: - System lay-out - Analog measured data acquisition - Analog measured data processing - Limit value generation and logical gating - Procesing of the reactor protection actuation signals - Decoupling of the reactor protection system - Mechanical lay-out - Monitoring system and - Emergency control station. (orig./RW)

  7. A carbon dioxide partial condensation direct cycle for advanced gas cooled fast and thermal reactors

    International Nuclear Information System (INIS)

    Yasuyoshi, Kato; Takeshi, NItawaki; Yoshio, Yoshizawa

    2001-01-01

    A carbon dioxide partial condensation direct cycle concept has been proposed for gas cooled fast and thermal reactors. The fast reactor with the concept are evaluated to be a potential alternative option to liquid metal cooled fast reactors, providing comparable cycle efficiency at the same core outlet temperature, eliminating the safety problems, simplifying the heat transport system and making easier plant maintenance. The thermal reactor with the concept is expected to be an alternative solution to current high temperature gas cooled reactors (HTGRs) with helium gas turbines, allowing comparable cycle efficiency at the moderate temperature of 650 C instead of 800 C in HTGRs. (author)

  8. Distribution and abundance of submerged macrophytes, including Myriophyllum spicatum L. (Angiospermae), in a reactor cooling reservoir

    International Nuclear Information System (INIS)

    Grace, J.B.; Tilly, L.J.

    1974-01-01

    The distribution of aquatic macrophytes, including Myriophyllum spicatum L., in a reactor cooling reservoir was analyzed by biomass removal. Samples were collected by use of a corer from hot, warm, and cold stations at progressive distances from a thermal effluent. Standing crop was doubled at the warm station as compared to the cold station due to the enhancement of Myriophyllum spicatum L. However, M. spicatum was least abundant at the hot station where Najas guadalupensis (Spreng.) Morong. predominated. Possible causes for the restriction of M. spicatum from the hot station are given. Maximum rooting depth of vegetation was reduced near the effluent apparently due to a combination of decreased light penetration and a higher photosynthesis-respiration compensation level at higher temperatures. (U.S.)

  9. Modular assembly for supporting, straining, and directing flow to a core in a nuclear reactor

    Science.gov (United States)

    Pennell, William E.

    1977-01-01

    A reactor core support arrangement for supporting, straining, and providing fluid flow to the core and periphery of a nuclear reactor during normal operation. A plurality of removable inlet modular units are contained within permanent liners in the lower supporting plate of the reactor vessel lower internals. During normal operation (1) each inlet modular unit directs main coolant flow to a plurality of core assemblies, the latter being removably supported in receptacles in the upper portion of the modular unit and (2) each inlet modular unit may direct bypass flow to a low pressure annular region of the reactor vessel. Each inlet modular unit may include special fluid seals interposed between mating surfaces of the inlet modular units and the core assemblies and between the inlet modular units and the liners, to minimize leakage and achieve an hydraulic balance. Utilizing the hydraulic balance, the modular units are held in the liners and the assemblies are held in the modular unit receptacles by their own respective weight. Included as part of the permanent liners below the horizontal support plate are generally hexagonal axial debris barriers. The axial debris barriers collectively form a bottom boundary of a secondary high pressure plenum, the upper boundary of which is the bottom surface of the horizontal support plate. Peripheral liners include radial debris barriers which collectively form a barrier against debris entry radially. During normal operation primary coolant inlet openings in the liner, below the axial debris barriers, pass a large amount of coolant into the inlet modular units, and secondary coolant inlet openings in the portion of the liners within the secondary plenum pass a small amount of coolant into the inlet modular units. The secondary coolant inlet openings also provide alternative coolant inlet flow paths in the unlikely event of blockage of the primary inlet openings. The primary inlet openings have characteristics which limit the

  10. Issues and future direction of thermal-hydraulics research and development in nuclear power reactors

    International Nuclear Information System (INIS)

    Saha, P.; Aksan, N.; Andersen, J.; Yan, J.; Simoneau, J.P.; Leung, L.; Bertrand, F.; Aoto, K.; Kamide, H.

    2013-01-01

    The paper archives the proceedings of an expert panel discussion on the issues and future direction of thermal-hydraulic research and development in nuclear power reactors held at the NURETH-14 conference in Toronto, Canada, in September 2011. Thermal-hydraulic issues related to both operating and advanced reactors are presented. Advances in thermal-hydraulics have significantly improved the performance of operating reactors. Further thermal-hydraulics research and development is continuing in both experimental and computational areas for operating reactors, reactors under construction or ready for near-term deployment, and advanced Generation-IV reactors. As the computing power increases, the fine-scale multi-physics computational models, coupled with the systems analysis code, are expected to provide answers to many challenging problems in both operating and advanced reactor designs

  11. Direct numerical simulation of reactor two-phase flows enabled by high-performance computing

    Energy Technology Data Exchange (ETDEWEB)

    Fang, Jun; Cambareri, Joseph J.; Brown, Cameron S.; Feng, Jinyong; Gouws, Andre; Li, Mengnan; Bolotnov, Igor A.

    2018-04-01

    Nuclear reactor two-phase flows remain a great engineering challenge, where the high-resolution two-phase flow database which can inform practical model development is still sparse due to the extreme reactor operation conditions and measurement difficulties. Owing to the rapid growth of computing power, the direct numerical simulation (DNS) is enjoying a renewed interest in investigating the related flow problems. A combination between DNS and an interface tracking method can provide a unique opportunity to study two-phase flows based on first principles calculations. More importantly, state-of-the-art high-performance computing (HPC) facilities are helping unlock this great potential. This paper reviews the recent research progress of two-phase flow DNS related to reactor applications. The progress in large-scale bubbly flow DNS has been focused not only on the sheer size of those simulations in terms of resolved Reynolds number, but also on the associated advanced modeling and analysis techniques. Specifically, the current areas of active research include modeling of sub-cooled boiling, bubble coalescence, as well as the advanced post-processing toolkit for bubbly flow simulations in reactor geometries. A novel bubble tracking method has been developed to track the evolution of bubbles in two-phase bubbly flow. Also, spectral analysis of DNS database in different geometries has been performed to investigate the modulation of the energy spectrum slope due to bubble-induced turbulence. In addition, the single-and two-phase analysis results are presented for turbulent flows within the pressurized water reactor (PWR) core geometries. The related simulations are possible to carry out only with the world leading HPC platforms. These simulations are allowing more complex turbulence model development and validation for use in 3D multiphase computational fluid dynamics (M-CFD) codes.

  12. Measuring of fissile isotopes partial antineutrino spectra in direct experiment at nuclear reactor

    OpenAIRE

    Sinev, V. V.

    2009-01-01

    The direct measuring method is considered to get nuclear reactor antineutrino spectrum. We suppose to isolate partial spectra of the fissile isotopes by using the method of antineutrino spectrum extraction from the inverse beta decay positron spectrum applied at Rovno experiment. This admits to increase the accuracy of partial antineutrino spectra forming the total nuclear reactor spectrum. It is important for the analysis of the reactor core fuel composition and could be applied for non-prol...

  13. Zirconium-based alloys, nuclear fuel rods and nuclear reactors including such alloys, and related methods

    Science.gov (United States)

    Mariani, Robert Dominick

    2014-09-09

    Zirconium-based metal alloy compositions comprise zirconium, a first additive in which the permeability of hydrogen decreases with increasing temperatures at least over a temperature range extending from 350.degree. C. to 750.degree. C., and a second additive having a solubility in zirconium over the temperature range extending from 350.degree. C. to 750.degree. C. At least one of a solubility of the first additive in the second additive over the temperature range extending from 350.degree. C. to 750.degree. C. and a solubility of the second additive in the first additive over the temperature range extending from 350.degree. C. to 750.degree. C. is higher than the solubility of the second additive in zirconium over the temperature range extending from 350.degree. C. to 750.degree. C. Nuclear fuel rods include a cladding material comprising such metal alloy compositions, and nuclear reactors include such fuel rods. Methods are used to fabricate such zirconium-based metal alloy compositions.

  14. Test Results from a Direct Drive Gas Reactor Simulator Coupled to a Brayton Power Conversion Unit

    Science.gov (United States)

    Hervol, David S.; Briggs, Maxwell H.; Owen, Albert K.; Bragg-Sitton, Shannon M.; Godfroy, Thomas J.

    2010-01-01

    Component level testing of power conversion units proposed for use in fission surface power systems has typically been done using relatively simple electric heaters for thermal input. These heaters do not adequately represent the geometry or response of proposed reactors. As testing of fission surface power systems transitions from the component level to the system level it becomes necessary to more accurately replicate these reactors using reactor simulators. The Direct Drive Gas-Brayton Power Conversion Unit test activity at the NASA Glenn Research Center integrates a reactor simulator with an existing Brayton test rig. The response of the reactor simulator to a change in Brayton shaft speed is shown as well as the response of the Brayton to an insertion of reactivity, corresponding to a drum reconfiguration. The lessons learned from these tests can be used to improve the design of future reactor simulators which can be used in system level fission surface power tests.

  15. Electromagnetic field nonuniformities in large area, high-frequency capacitive plasma reactors, including electrode asymmetry effects

    International Nuclear Information System (INIS)

    Sansonnens, L; Howling, A A; Hollenstein, Ch

    2006-01-01

    Electromagnetic wave propagation effects can give rise to important limitations for processing uniformity in large area, radio-frequency (rf) capacitive plasma reactors. The electromagnetic wavefield solution is derived for a capacitive, high-frequency, cylindrical reactor with symmetric or asymmetric electrode areas containing a uniform plasma slab. It is shown that only two distinct electromagnetic modes are necessary and sufficient to determine the electromagnetic fields everywhere within the reactor except close to the sidewalls. The first mode gives rise to the interelectrode rf voltage standing wave effect associated with high frequencies in large area reactors, and the second mode gives rise to the telegraph effect associated with asymmetric electrode areas, which necessitates the redistribution of rf current along the plasma to maintain rf current continuity. This work gives a unified treatment of both effects which have previously been studied separately, experimentally and theoretically, in the literature. The equivalent circuit of each mode is also derived from its respective dispersion relation. Examples of this electromagnetic wavefield solution show that both modes can cause nonuniformity of the plasma rf potential, depending on the reactor geometry, excitation frequency and plasma permittivity and sheath width, which has consequences for large-area plasma processing

  16. Study on the seismic response of reactor vessel of pool type LMFBR including fluid-structure interaction

    International Nuclear Information System (INIS)

    Tanimoto, K.; Ito, T.; Fujita, K.; Kurihara, C.; Sawada, Y.; Sakurai, A.

    1988-01-01

    The paper presents the seismic response of reactor vessel of pool type LMFBR with fluid-structure interaction. The reactor vessel has bottom support arrangement, the same core support system as Super-Phenix in France. Due to the bottom support arrangement, the level of core support is lower than that of the side support arrangement. So, in this reactor vessel, the displacement of the core top tends to increase because of the core's rocking. In this study, we investigated the vibration and seismic response characteristics of the reactor vessel. Therefore, the seismic experiments were carried out using one-eighth scale model and the seismic response including FSI and sloshing were investigated. From this study, the effect of liquid on the vibration characteristics and the seismic response characteristics of reactor vessel were clarified and sloshing characteristics were also clarified. It was confirmed that FEM analysis with FSI can reproduce the seismic behavior of the reactor vessel and is applicable to seismic design of the pool type LMFBR with bottom support arrangement. (author). 5 refs, 14 figs, 2 tabs

  17. Effects of embedment including slip and separation on seismic SSI response of a nuclear reactor building

    International Nuclear Information System (INIS)

    Saxena, Navjeev; Paul, D.K.

    2012-01-01

    Highlights: ► Both the slip and separation of reactor base reduce with increase in embedment. ► The slip and separation become insignificant beyond 1/4 and 1/2 embedment respectively. ► The stresses in reactor reduce significantly upto 1/4 embedment. ► The stress reduction with embedment is more pronounced in case of tensile stresses. ► The modeling of interface is important beyond 1/8 embedment as stresses are underestimated otherwise. - Abstract: The seismic response of nuclear reactor containment building considering the effects of embedment, slip and separation at soil–structure interface requires modeling of the soil, structure and interface altogether. Slip and separation at the interface causes stress redistribution in the soil and the structure around the interface. The embedment changes the dynamic characteristics of the soil–structure system. Consideration of these aspects allows capturing the realistic response of the structure, which has been a research gap and presented here individually as well as taken together. Finite element analysis has been carried out in time domain to attempt the highly nonlinear problem. The study draws important conclusions useful for design of nuclear reactor containment building.

  18. An add-on system including a micro-reactor for an atr-ir spectrometer

    DEFF Research Database (Denmark)

    2014-01-01

    The invention relates to an add-on system for an attenuated total reflectance infrared (ATR-IR) spectrometer, the add-on system allowing for time-resolved in situ IR measurements of heterogeneous mixtures. The add-on device comprises a micro-reactor (300A) forming a sample cavity (305) when...

  19. Small Reactor Designs Suitable for Direct Nuclear Thermal Propulsion: Interim Report

    Energy Technology Data Exchange (ETDEWEB)

    Bruce G. Schnitzler

    2012-01-01

    Advancement of U.S. scientific, security, and economic interests requires high performance propulsion systems to support missions beyond low Earth orbit. A robust space exploration program will include robotic outer planet and crewed missions to a variety of destinations including the moon, near Earth objects, and eventually Mars. Past studies, in particular those in support of both the Strategic Defense Initiative (SDI) and the Space Exploration Initiative (SEI), have shown nuclear thermal propulsion systems provide superior performance for high mass high propulsive delta-V missions. In NASA's recent Mars Design Reference Architecture (DRA) 5.0 study, nuclear thermal propulsion (NTP) was again selected over chemical propulsion as the preferred in-space transportation system option for the human exploration of Mars because of its high thrust and high specific impulse ({approx}900 s) capability, increased tolerance to payload mass growth and architecture changes, and lower total initial mass in low Earth orbit. The recently announced national space policy2 supports the development and use of space nuclear power systems where such systems safely enable or significantly enhance space exploration or operational capabilities. An extensive nuclear thermal rocket technology development effort was conducted under the Rover/NERVA, GE-710 and ANL nuclear rocket programs (1955-1973). Both graphite and refractory metal alloy fuel types were pursued. The primary and significantly larger Rover/NERVA program focused on graphite type fuels. Research, development, and testing of high temperature graphite fuels was conducted. Reactors and engines employing these fuels were designed, built, and ground tested. The GE-710 and ANL programs focused on an alternative ceramic-metallic 'cermet' fuel type consisting of UO2 (or UN) fuel embedded in a refractory metal matrix such as tungsten. The General Electric program examined closed loop concepts for space or terrestrial

  20. Economic Analysis on Direct Use of Spent Pressurized Water Reactor Fuel in CANDU Reactors - I: DUPIC Fuel Fabrication Cost

    International Nuclear Information System (INIS)

    Choi, Hangbok; Ko, Won Il; Yang, Myung Seung

    2001-01-01

    A preliminary conceptual design of a Direct Use of spent Pressurized water reactor (PWR) fuel In Canada deuterium uranium (CANDU) reactors (DUPIC) fuel fabrication plant was studied, which annually converts spent PWR fuel of 400 tonnes heavy element (HE) into CANDU fuel. The capital and operating costs were estimated from the viewpoint of conceptual design. Assuming that the annual discount rate is 5% during the construction (5 yr) and operation period (40 yr) and contingency is 25% of the capital cost, the levelized unit cost (LUC) of DUPIC fuel fabrication was estimated to be 616 $/kg HE, which is mostly governed by annual operation and maintenance costs that correspond to 63% of LUC. Among the operation and maintenance cost components being considered, the waste disposal cost has the dominant effect on LUC (∼49%). From sensitivity analyses of production capacity, discount rate, and contingency, it was found that the production capacity of the plant is the major parameter that affects the LUC

  1. A Carbon Dioxide Gas Turbine Direct Cycle with Partial Condensation for Nuclear Reactors

    International Nuclear Information System (INIS)

    Yasuyoshi Kato; Takeshi Nitawaki; Yoshio Yoshizawa

    2002-01-01

    A carbon dioxide gas turbine power generation system with a partial condensation cycle has been proposed for thermal and fast nuclear reactors, in which compression is done partly in the liquid phase and partly in the gas phase. This cycle achieves higher cycle efficiency than a He direct cycle mainly due to reduced compressor work of the liquid phase and of the carbon dioxide real gas effect, especially in the vicinity of the critical point. If this cycle is applied to a thermal reactor, efficiency of this cycle is about 55% at a reactor outlet temperature of 900 deg. C and pressure of 12.5 MPa, which is higher by about 10% than a typical helium direct gas turbine cycle plant (PBMR) at 900 deg. C and 8.4 MPa; this cycle also provides comparable cycle efficiency at the moderate core outlet temperature of 600 deg. C with that of the helium cycle at 900 deg. C. If this cycle is applied to a fast reactor, it is anticipated to be an alternative to liquid metal cooled fast reactors that can provide slightly higher cycle efficiency at the same core outlet temperature; it would eliminate safety problems, simplify the heat transport system and simplify plant maintenance. A passive decay heat removal system is realized by connecting a liquid carbon dioxide storage tank with the reactor vessel and by supplying carbon dioxide gasified from the tank to the core in case of depressurization event. (authors)

  2. A study on direct energy conversion systems for fast reactors

    International Nuclear Information System (INIS)

    Brown, Neil; Stoner, Joseph; Sayre, Edwin

    2001-11-01

    An innovative conceptual design of a thermoelectric generator for a terrestrial power system has been developed and evaluated. The concept is based on experience gained during the development of the SP-100 lithium cooled nuclear power system for space application. In place of the SP-100 planner multi-couple cell a radial arrangement of the thermocouples is used to build the thermoelectric cell used in the generator. A conceptual approach for integrating the cell into a tubular element that can be fabricated into a power converter heat exchanger module was evaluated. A method for fabricating the cell, tubular assemblies of ten cells and the converter modules has also been developed. Using the module conceptual design, the thermal-electrical performance of several systems was evaluated. The systems evaluated were assumed to operate with primary coolant at 900degC, 550degC and 600degC. System arrangements and size and thermal efficiencies were established for the first two operating temperatures at power level of 1.7 and 20MWe. It was assumed they used a thermoelectric material with a projected value for the figure of merit, ZT, of 3.0. The systems were also evaluated assuming a ZT of 0.95 for thermoelectric components operating at the lower temperature. The 600degC system was sized for 1MWe assuming a thermoelectric material figure of merit of 0.95. The ZT value of 3.0 is very speculative while the value of 0.95 may be achieved with currently available materials. System efficiencies between 8.2 and 24.8% were calculated and it estimated that it would require arrays of 86 to 3560 of the standard modules for power levels from 1MWe to 20MWe respectively. These systems would cover areas of 6000 to 62,500 square feet The key element in the conceptual design is the innovative thermoelectric cell. A logical next step in development of terrestrial thermoelectric converters for application to high temperature reactors is the detailed design and testing of this cell. A useful

  3. Reactor

    International Nuclear Information System (INIS)

    Fujibayashi, Toru.

    1976-01-01

    Object: To provide a boiling water reactor which can enhance a quake resisting strength and flatten power distribution. Structure: At least more than four fuel bundles, in which a plurality of fuel rods are arranged in lattice fashion which upper and lower portions are supported by tie-plates, are bundled and then covered by a square channel box. The control rod is movably arranged within a space formed by adjoining channel boxes. A spacer of trapezoidal section is disposed in the central portion on the side of the channel box over substantially full length in height direction, and a neutron instrumented tube is disposed in the central portion inside the channel box. Thus, where a horizontal load is exerted due to earthquake or the like, the spacers come into contact with each other to support the channel box and prevent it from abnormal vibrations. (Furukawa, Y.)

  4. Improved best estimate plus uncertainty methodology including advanced validation concepts to license evolving nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Unal, Cetin [Los Alamos National Laboratory; Williams, Brian [Los Alamos National Laboratory; Mc Clure, Patrick [Los Alamos National Laboratory; Nelson, Ralph A [IDAHO NATIONAL LAB

    2010-01-01

    Many evolving nuclear energy programs plan to use advanced predictive multi-scale multi-physics simulation and modeling capabilities to reduce cost and time from design through licensing. Historically, the role of experiments was primary tool for design and understanding of nuclear system behavior while modeling and simulation played the subordinate role of supporting experiments. In the new era of multi-scale multi-physics computational based technology development, the experiments will still be needed but they will be performed at different scales to calibrate and validate models leading predictive simulations. Cost saving goals of programs will require us to minimize the required number of validation experiments. Utilization of more multi-scale multi-physics models introduces complexities in the validation of predictive tools. Traditional methodologies will have to be modified to address these arising issues. This paper lays out the basic aspects of a methodology that can be potentially used to address these new challenges in design and licensing of evolving nuclear technology programs. The main components of the proposed methodology are verification, validation, calibration, and uncertainty quantification. An enhanced calibration concept is introduced and is accomplished through data assimilation. The goal is to enable best-estimate prediction of system behaviors in both normal and safety related environments. To achieve this goal requires the additional steps of estimating the domain of validation and quantification of uncertainties that allow for extension of results to areas of the validation domain that are not directly tested with experiments, which might include extension of the modeling and simulation (M&S) capabilities for application to full-scale systems. The new methodology suggests a formalism to quantify an adequate level of validation (predictive maturity) with respect to required selective data so that required testing can be minimized for cost

  5. Improved best estimate plus uncertainty methodology including advanced validation concepts to license evolving nuclear reactors

    International Nuclear Information System (INIS)

    Unal, Cetin; Williams, Brian; McClure, Patrick; Nelson, Ralph A.

    2010-01-01

    Many evolving nuclear energy programs plan to use advanced predictive multi-scale multi-physics simulation and modeling capabilities to reduce cost and time from design through licensing. Historically, the role of experiments was primary tool for design and understanding of nuclear system behavior while modeling and simulation played the subordinate role of supporting experiments. In the new era of multi-scale multi-physics computational based technology development, the experiments will still be needed but they will be performed at different scales to calibrate and validate models leading predictive simulations. Cost saving goals of programs will require us to minimize the required number of validation experiments. Utilization of more multi-scale multi-physics models introduces complexities in the validation of predictive tools. Traditional methodologies will have to be modified to address these arising issues. This paper lays out the basic aspects of a methodology that can be potentially used to address these new challenges in design and licensing of evolving nuclear technology programs. The main components of the proposed methodology are verification, validation, calibration, and uncertainty quantification. An enhanced calibration concept is introduced and is accomplished through data assimilation. The goal is to enable best-estimate prediction of system behaviors in both normal and safety related environments. To achieve this goal requires the additional steps of estimating the domain of validation and quantification of uncertainties that allow for extension of results to areas of the validation domain that are not directly tested with experiments, which might include extension of the modeling and simulation (M and S) capabilities for application to full-scale systems. The new methodology suggests a formalism to quantify an adequate level of validation (predictive maturity) with respect to required selective data so that required testing can be minimized for

  6. Recent advances in severe accident technology - direct containment heating in advanced light water reactors

    International Nuclear Information System (INIS)

    Fontana, M.H.

    1993-01-01

    The issues affecting high-pressure melt ejection (HPME) and the consequential containment pressurization from direct containment heating (DCH), as they affect advanced light water reactors (ALWRs), specifically advanced pressurized water reactors (APWRs), were reviewed by the U.S. Department of Energy Advanced Reactor Severe Accident Program (ARSAP). Recommendations from ARSAP regarding the design of APWRs to minimize DCH are embodied within the Electric Power Research Institute ALWR Utility Requirements Document, which specifies (a) a large, strong containment; (b) an in-containment refueling water storage tank; (c) a reactor cavity configuration that minimizes energy transport to the containment atmosphere; and (d) a reactor coolant system depressurization system. Experimental and analytical efforts, which have focused on current-generation plants, and analyses for APWRs were reviewed. Although DCH is a subject of continuous research and considerable uncertainties remain, it is the judgment of the ARSAP that reactors complying with the recommended design requirements would have a low probability of early containment failure due to HPME and DCH

  7. 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

  8. What is past is prologue: future directions in tokamak power reactor design research

    International Nuclear Information System (INIS)

    Conn, R.W.

    1976-01-01

    Conceptual tokamak power reactor designs over the last five years have provided us with many fundamental insights regarding tokamaks as fusion reactors. This first generation of studies has helped lay the groundwork upon which to build improvements in reactor design and begin a process of optimization. After reviewing the first generation of studies and the primary conclusions they produced, we discuss four current designs that are representative of present trends in this area of research. In particular, we discuss the trends towards reduced reactor size and higher neutron wall loadings. Moving in this direction requires new approaches to many subsystem designs. We describe new approaches and future directions in first wall and blanket designs that can achieve reliable operation and reasonable lifetime, the use of cryogenic but normal aluminum magnets for the pulsed coils in a tokamak, blanket designs that allow elimination of the intermediate loop, and low activity shields and toroidal field magnets. We close with a discussion of the future role of conceptual reactor design research and the need for close interaction with ongoing experiments in fusion technology

  9. Reactor

    International Nuclear Information System (INIS)

    Evans, R.M.

    1976-01-01

    Disclosed is 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. 1 claim, 16 figures

  10. Solar reforming of methane in a direct absorption catalytic reactor on a parabolic dish. 2: Modeling and analysis

    Science.gov (United States)

    Skocypec, Russell D.; Hogan, Roy E., Jr.; Muir, James F.

    1991-01-01

    The catalytically enhanced solar absorption receiver (CAESAR) experiment was conducted to determine the thermal, chemical, and mechanical performance of a commercial-scale, dish-mounted, direct catalytic absorption receiver (DCAR) reactor over a range of steady state and transient (cloud) operating conditions. The focus of the experiment is on global performance such as receiver efficiencies and overall methane conversion; it was not intended to provide data for code validation. A numerical model was previously developed to provide guidance in the design of the absorber. The one-dimensional, planar and steady-state model incorporates, the following energy transfer mechanisms: solar and infrared radiation, heterogeneous chemical reaction, conduction in the solid phase, and convection between the fluid and solid phases. A number of upgrades to the model and improved property values are presented here. Model predictions are shown to bound the experimental axial thermocouple data when experimental uncertainties are included. Global predictions are made using a technique in which the incident solar flux distribution is subdivided into flux contour bands. Model predictions for each band are then spatially integrated to provide global predictions such as reactor efficiencies and methane conversions. Global predictions are shown to compare well with experimental data. Reactor predictions for anticipated operating conditions suggest a further decrease in optical density at the front of the absorber inner disk may be beneficial. The need to conduct code-validation experiments is identified as being essential in improving the confidence in the capability to predict large-scale reactor operation.

  11. Reactor

    International Nuclear Information System (INIS)

    Toyama, Masahiro; Kasai, Shigeo.

    1978-01-01

    Purpose: To provide a lmfbr type reactor wherein effusion of coolants through a loop contact portion is reduced even when fuel assemblies float up, and misloading of reactor core constituting elements is prevented thereby improving the reactor safety. Constitution: The reactor core constituents are secured in the reactor by utilizing the differential pressure between the high-pressure cooling chamber and low-pressure cooling chamber. A resistance port is formed at the upper part of a connecting pipe, and which is connect the low-pressure cooling chamber and the lower surface of the reactor core constituent. This resistance part is formed such that the internal sectional area of the connecting pipe is made larger stepwise toward the upper part, and the cylinder is formed larger so that it profiles the inner surface of the connecting pipe. (Aizawa, K.)

  12. Reactor

    International Nuclear Information System (INIS)

    Ikeda, Masaomi; Kashimura, Kazuo; Inoue, Kazuyuki; Nishioka, Kazuya.

    1979-01-01

    Purpose: To facilitate the construction of a reactor containment building, whereby the inspections of the outer wall of a reactor container after the completion of the construction of the reactor building can be easily carried out. Constitution: In a reactor accommodated in a container encircled by a building wall, a space is provided between the container and the building wall encircling the container, and a metal wall is provided in the space so that it is fitted in the building wall in an attachable or detatchable manner. (Aizawa, K.)

  13. Enhancing VHTR Passive Safety and Economy with Thermal Radiation Based Direct Reactor Auxiliary Cooling System

    Energy Technology Data Exchange (ETDEWEB)

    Haihua Zhao; Hongbin Zhang; Ling Zou; Xiaodong Sun

    2012-06-01

    One of the most important requirements for Gen. IV Very High Temperature Reactor (VHTR) is passive safety. Currently all the gas cooled version of VHTR designs use Reactor Vessel Auxiliary Cooling System (RVACS) for passive decay heat removal. The decay heat first is transferred to the core barrel by conduction and radiation, and then to the reactor vessel by thermal radiation and convection; finally the decay heat is transferred to natural circulated air or water systems. RVACS can be characterized as a surface based decay heat removal system. The RVACS is especially suitable for smaller power reactors since small systems have relatively larger surface area to volume ratio. However, RVACS limits the maximum achievable power level for modular VHTRs due to the mismatch between the reactor power (proportional to volume) and decay heat removal capability (proportional to surface area). When the relative decay heat removal capability decreases, the peak fuel temperature increases, even close to the design limit. Annular core designs with inner graphite reflector can mitigate this effect; therefore can further increase the reactor power. Another way to increase the reactor power is to increase power density. However, the reactor power is also limited by the decay heat removal capability. Besides the safety considerations, VHTRs also need to be economical in order to compete with other reactor concepts and other types of energy sources. The limit of decay heat removal capability set by using RVACS has affected the economy of VHTRs. A potential alternative solution is to use a volume-based passive decay heat removal system, called Direct Reactor Auxiliary Cooling Systems (DRACS), to remove or mitigate the limitation on decay heat removal capability. DRACS composes of natural circulation loops with two sets of heat exchangers, one on the reactor side and another on the environment side. For the reactor side, cooling pipes will be inserted into holes made in the outer or

  14. What is past is prologue: future directions in Tokamak Power Reactor Design Research

    International Nuclear Information System (INIS)

    Conn, R.W.

    1976-01-01

    After reviewing the first generation of studies and the primary conclusions they produced, four current designs are discussed that are representative of present trends in this area of research. In particular, the trends towards reduced reactor size and higher neutron wall loadings are discussed. Moving in this direction requires new approaches to many subsystem designs. New approaches and future directions in first wall and blanket designs that can achieve reliable operation and reasonable lifetime, the use of cryogenic but normal aluminum magnets for the pulsed coils in a tokamak, blanket designs that allow elimination of the intermediate loop, and low activity shields and toroidal field magnets are described. A discussion is given of the future role of conceptual reactor design research and the need for close interactions with ongoing experiments in fusion technology

  15. Effect of reactor temperature on direct growth of carbon nanomaterials on stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Edzatty, A. N., E-mail: nuredzatty@gmail.com; Syazwan, S. M., E-mail: mdsyazwan.sanusi@gmail.com; Norzilah, A. H., E-mail: norzilah@unimap.edu.my; Jamaludin, S. B., E-mail: sbaharin@unimap.edu.my [Centre of Excellence for Frontier Materials Research, School of Materials Engineering, University Malaysia Perlis (Malaysia)

    2016-07-19

    Currently, carbon nanomaterials (CNMs) are widely used for various applications due to their extraordinary electrical, thermal and mechanical properties. In this work, CNMs were directly grown on the stainless steel (SS316) via chemical vapor deposition (CVD). Acetone was used as a carbon source and argon was used as carrier gas, to transport the acetone vapor into the reactor when the reaction occurred. Different reactor temperature such as 700, 750, 800, 850 and 900 °C were used to study their effect on CNMs growth. The growth time and argon flow rate were fixed at 30 minutes and 200 ml/min, respectively. Characterization of the morphology of the SS316 surface after CNMs growth using Scanning Electron Microscopy (SEM) showed that the diameter of grown-CNMs increased with the reactor temperature. Energy Dispersive X-ray (EDX) was used to analyze the chemical composition of the SS316 before and after CNMs growth, where the results showed that reduction of catalyst elements such as iron (Fe) and nickel (Ni) at high temperature (700 – 900 °C). Atomic Force Microscopy (AFM) analysis showed that the nano-sized hills were in the range from 21 to 80 nm. The best reactor temperature to produce CNMs was at 800 °C.

  16. Economic Analysis on Direct Use of Spent Pressurized Water Reactor Fuel in CANDU Reactors - II: DUPIC Fuel-Handling Cost

    International Nuclear Information System (INIS)

    Choi, Hangbok; Ko, Won Il; Yang, Myung Seung; Namgung, Ihn; Na, Bok-Gyun

    2001-01-01

    The Direct Use of spent Pressurized water reactor fuel In Canada deuterium uranium (CANDU) reactors (DUPIC) fuel-handling technique has been investigated through a conceptual design study to estimate the unit cost that can be used for the DUPIC fuel cycle cost calculation. The conceptual design study has shown that fresh DUPIC fuel can be transferred to the core following the existing spent-fuel discharge route, provided that new fuel-handling equipment, such as the manipulator, opening/sealing tool of shipping casks, new fuel magazine, new fuel ram, dryer, gamma-ray detector, etc., are installed. The reverse path loading option is known to minimize the number of additional pieces of equipment for fuel handling, because it utilizes the existing spent-fuel handling equipment, and the discharge of spent DUPIC fuel can be done through the existing spent-fuel handling system without any modification. However, because the decay heat of spent DUPIC fuel is much higher than that of spent natural uranium fuel, the extra cooling capacity should be supplemented in the spent-fuel storage bay. Based on the conceptual design study, the capital cost for DUPIC fuel handling and extra storage cooling capacity was estimated to be $3 750 000 (as of December 1999) per CANDU plant. The levelized unit cost of DUPIC fuel handling was then obtained by considering the amount of fuel that will be required during the lifetime of a plant, which is 5.13 $/kg heavy metal. Compared with the other unit costs of the fuel cycle components, it is expected that DUPIC fuel handling has only a minor effect on the overall fuel cycle cost

  17. Pipe elbow stiffness coefficients including shear and bend flexibility factors for use in direct stiffness codes

    International Nuclear Information System (INIS)

    Perry, R.F.

    1977-01-01

    Historically, developments of computer codes used for piping analysis were based upon the flexibility method of structural analysis. Because of the specialized techniques employed in this method, the codes handled systems composed of only piping elements. Over the past ten years, the direct stiffness method has gained great popularity because of its systematic solution procedure regardless of the type of structural elements composing the system. A great advantage is realized with a direct stiffness code that combines piping elements along with other structural elements such as beams, plates, and shells, in a single model. One common problem, however, has been the lack of an accurate pipe elbow element that would adequately represent the effects of transverse shear and bend flexibility factors. The purpose of the present paper is to present a systematic derivation of the required 12x12 stiffness matrix and load vectors for a three dimensional pipe elbow element which includes the effects of transverse shear and pipe bend flexibility according to the ASME Boiler and Pressure Vessel Code, Section III. The results are presented analytically and as FORTRAN subroutines to be directly incorporated into existing direct stiffness codes. (Auth.)

  18. Modeling of the Direct Current Generator Including the Magnetic Saturation and Temperature Effects

    Directory of Open Access Journals (Sweden)

    Alfonso J. Mercado-Samur

    2013-11-01

    Full Text Available In this paper the inclusion of temperature effect on the field resistance on the direct current generator model DC1A, which is valid to stability studies is proposed. First, the linear generator model is presented, after the effect of magnetic saturation and the change in the resistance value due to temperature produced by the field current are included. The comparison of experimental results and model simulations to validate the model is used. A direct current generator model which is a better representation of the generator is obtained. Visual comparison between simulations and experimental results shows the success of the proposed model, because it presents the lowest error of the compared models. The accuracy of the proposed model is observed via Modified Normalized Sum of Squared Errors index equal to 3.8979%.

  19. Design, Testing and Modeling of the Direct Reactor Auxiliary Cooling System for AHTRs

    International Nuclear Information System (INIS)

    Lv, Quiping; Sun, Xiaodong; Chtistensen, Richard; Blue, Thomas; Yoder, Graydon; Wilson, Dane

    2015-01-01

    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.

  20. 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.

  1. Technology of direct conversion for mirror reactor end-loss plasma

    International Nuclear Information System (INIS)

    Barr, W.L.; Moir, R.W.

    1980-01-01

    Design concepts are presented for plasma direct convertors (PDC) intended primarily for use on the end-loss plasma from tandem-mirror reactors. Recent experimental results confirm most of these design concepts. Both a one-stage and a two-stage PDC were tested in reactor-like conditions using a 100-kV, 6-kW ion beam. In a separate test on the end of the TMX machine, a single stage PDC recovered 79 W for a net efficiency of 50%. Tandem mirror devices are well suited to PDC. The high minimum energy of the end-loss ions, the magnetic expansion outside the mirrors, and the vacuum conditions in the end tanks required by the confined plasma, all preexist. The inclusion of a PDC is therefore a rather small addition. These facts and the scale parameters for a PDC are discussed

  2. Direct-phase-variable model of a synchronous reluctance motor including all slot and winding harmonics

    International Nuclear Information System (INIS)

    Obe, Emeka S.; Binder, A.

    2011-01-01

    A detailed model in direct-phase variables of a synchronous reluctance motor operating at mains voltage and frequency is presented. The model includes the stator and rotor slot openings, the actual winding layout and the reluctance rotor geometry. Hence, all mmf and permeance harmonics are taken into account. It is seen that non-negligible harmonics introduced by slots are present in the inductances computed by the winding function procedure. These harmonics are usually ignored in d-q models. The machine performance is simulated in the stator reference frame to depict the difference between this new direct-phase model including all harmonics and the conventional rotor reference frame d-q model. Saturation is included by using a polynomial fitting the variation of d-axis inductance with stator current obtained by finite-element software FEMAG DC (registered) . The detailed phase-variable model can yield torque pulsations comparable to those obtained from finite elements while the d-q model cannot.

  3. Optimal conditions in direct dimethyl ether synthesis from syngas utilizing a dual-type fluidized bed reactor

    International Nuclear Information System (INIS)

    Yousefi, Ahmad; Eslamloueyan, Reza; Kazerooni, Nooshin Moradi

    2017-01-01

    Concerns over environmental pollution and ever-increasing energy demand have urged the global community to tap clean-burning fuels among which dimethyl ether is a promising candidate for contribution in the transportation sector. Direct dimethyl ether synthesis from syngas, in which methanol production and dehydration take place simultaneously, is arguably the preferred route for large scale production. In this study, direct dimethyl ether synthesis is proposed in an industrial dual-type fluidized bed reactor. This configuration involves two fluidized bed reactors operating in different conditions. In the first catalytic reactor (water-cooled reactor), the synthesis gas is partly converted to methanol after being preheated by the reaction heat in the second reactor (gas-cooled reactor). A two-phase generalized comprehensive reactor model, comprised of the flow in three different regimes is applied and a smooth transition between flow regimes is provided based on the probabilistic averaging approach. The optimal operating conditions are sought by employing differential evolution algorithm as a robust optimization strategy. The dimethyl ether mole fraction is considered as the objective function during the optimization. The results show considerable dimethyl ether enhancement by 16% and 14% compared to the conventional direct dimethyl ether synthesis reactor and dual-type fixed bed dimethyl ether reactor arrangements, respectively. - Highlights: • Dual-type catalytic fluidized bed reactors for dimethyl ether synthesis is studied. • A two-phase comprehensive model comprised of flow in three regimes is used. • Probabilistic averaging approach is applied for smooth transitions between regimes. • Differential evolution method is employed to determine optimal operating conditions. • Production capacity is remarkably enhanced compared to conventional reactor.

  4. A market based active/reactive dispatch including transformer taps and reactor and capacitor banks using Simulated Annealing

    International Nuclear Information System (INIS)

    Gomes, Mario Helder; Saraiva, Joao Tome

    2009-01-01

    This paper describes an optimization model to be used by System Operators in order to validate the economic schedules obtained by Market Operators together with the injections from Bilateral Contracts. These studies will be performed off-line in the day before operation and the developed model is based on adjustment bids submitted by generators and loads and it is used by System Operators if that is necessary to enforce technical or security constraints. This model corresponds to an enhancement of an approach described in a previous paper and it now includes discrete components as transformer taps and reactor and capacitor banks. The resulting mixed integer formulation is solved using Simulated Annealing, a well known metaheuristic specially suited for combinatorial problems. Once the Simulated Annealing converges and the values of the discrete variables are fixed, the resulting non-linear continuous problem is solved using Sequential Linear Programming to get the final solution. The developed model corresponds to an AC version, it includes constraints related with the capability diagram of synchronous generators and variables allowing the computation of the active power required to balance active losses. Finally, the paper includes a Case Study based on the IEEE 118 bus system to illustrate the results that it is possible to obtain and their interest. (author)

  5. A market based active/reactive dispatch including transformer taps and reactor and capacitor banks using Simulated Annealing

    Energy Technology Data Exchange (ETDEWEB)

    Gomes, Mario Helder [Departamento de Engenharia Electrotecnica, Instituto Politecnico de Tomar, Quinta do Contador, Estrada da Serra, 2300 Tomar (Portugal); Saraiva, Joao Tome [INESC Porto, Faculdade de Engenharia, Universidade do Porto, Campus da FEUP, Rua Dr. Roberto Frias, 4200-465 Porto (Portugal)

    2009-06-15

    This paper describes an optimization model to be used by System Operators in order to validate the economic schedules obtained by Market Operators together with the injections from Bilateral Contracts. These studies will be performed off-line in the day before operation and the developed model is based on adjustment bids submitted by generators and loads and it is used by System Operators if that is necessary to enforce technical or security constraints. This model corresponds to an enhancement of an approach described in a previous paper and it now includes discrete components as transformer taps and reactor and capacitor banks. The resulting mixed integer formulation is solved using Simulated Annealing, a well known metaheuristic specially suited for combinatorial problems. Once the Simulated Annealing converges and the values of the discrete variables are fixed, the resulting non-linear continuous problem is solved using Sequential Linear Programming to get the final solution. The developed model corresponds to an AC version, it includes constraints related with the capability diagram of synchronous generators and variables allowing the computation of the active power required to balance active losses. Finally, the paper includes a Case Study based on the IEEE 118 bus system to illustrate the results that it is possible to obtain and their interest. (author)

  6. Modeling the distribution of ammonia across Europe including bi-directional surface–atmosphere exchange

    Directory of Open Access Journals (Sweden)

    R. J. Wichink Kruit

    2012-12-01

    Full Text Available A large shortcoming of current chemistry transport models (CTM for simulating the fate of ammonia in the atmosphere is the lack of a description of the bi-directional surface–atmosphere exchange. In this paper, results of an update of the surface–atmosphere exchange module DEPAC, i.e. DEPosition of Acidifying Compounds, in the chemistry transport model LOTOS-EUROS are discussed. It is shown that with the new description, which includes bi-directional surface–atmosphere exchange, the modeled ammonia concentrations increase almost everywhere, in particular in agricultural source areas. The reason is that by using a compensation point the ammonia lifetime and transport distance is increased. As a consequence, deposition of ammonia and ammonium decreases in agricultural source areas, while it increases in large nature areas and remote regions especially in southern Scandinavia. The inclusion of a compensation point for water reduces the dry deposition over sea and allows reproducing the observed marine background concentrations at coastal locations to a better extent. A comparison with measurements shows that the model results better represent the measured ammonia concentrations. The concentrations in nature areas are slightly overestimated, while the concentrations in agricultural source areas are still underestimated. Although the introduction of the compensation point improves the model performance, the modeling of ammonia remains challenging. Important aspects are emission patterns in space and time as well as a proper approach to deal with the high concentration gradients in relation to model resolution. In short, the inclusion of a bi-directional surface–atmosphere exchange is a significant step forward for modeling ammonia.

  7. Direct deterministic method for neutronics analysis and computation of asymptotic burnup distribution in a recirculating pebble-bed reactor

    International Nuclear Information System (INIS)

    Terry, W.K.; Gougar, H.D.; Ougouag, A.M.

    2002-01-01

    A new deterministic method has been developed for the neutronics analysis of a pebble-bed reactor (PBR). The method accounts for the flow of pebbles explicitly and couples the flow to the neutronics. The method allows modeling of once-through cycles as well as cycles in which pebbles are recirculated through the core an arbitrary number of times. This new work is distinguished from older methods by the systematically semi-analytical approach it takes. In particular, whereas older methods use the finite-difference approach (or an equivalent one) for the discretization and the solution of the burnup equation, the present work integrates the relevant differential equation analytically in discrete and complementary sub-domains of the reactor. Like some of the finite-difference codes, the new method obtains the asymptotic fuel-loading pattern directly, without modeling any intermediate loading pattern. This is a significant advantage for the design and optimization of the asymptotic fuel-loading pattern. The new method is capable of modeling directly both the once-through-then-out fuel cycle and the pebble recirculating fuel cycle. Although it currently includes a finite-difference neutronics solver, the new method has been implemented into a modular code that incorporates the framework for the future coupling to an efficient solver such as a nodal method and to modern cross section preparation capabilities. In its current state, the deterministic method presented here is capable of quick and efficient design and optimization calculations for the in-core PBR fuel cycle. The method can also be used as a practical 'scoping' tool. It could, for example, be applied to determine the potential of the PBR for resisting nuclear-weapons proliferation and to optimize proliferation-resistant features. However, the purpose of this paper is to show that the method itself is viable. Refinements to the code are under way, with the objective of producing a powerful reactor physics

  8. Small Fast Spectrum Reactor Designs Suitable for Direct Nuclear Thermal Propulsion

    Energy Technology Data Exchange (ETDEWEB)

    Bruce G. Schnitzler; Stanley K. Borowski

    2012-07-01

    Advancement of U.S. scientific, security, and economic interests through a robust space exploration program requires high performance propulsion systems to support a variety of robotic and crewed missions beyond low Earth orbit. Past studies, in particular those in support of both the Strategic Defense Initiative (SDI) and Space Exploration Initiative (SEI), have shown nuclear thermal propulsion systems provide superior performance for high mass high propulsive delta-V missions. The recent NASA Design Reference Architecture (DRA) 5.0 Study re-examined mission, payload, and transportation system requirements for a human Mars landing mission in the post-2030 timeframe. Nuclear thermal propulsion was again identified as the preferred in-space transportation system. A common nuclear thermal propulsion stage with three 25,000-lbf thrust engines was used for all primary mission maneuvers. Moderately lower thrust engines may also have important roles. In particular, lower thrust engine designs demonstrating the critical technologies that are directly extensible to other thrust levels are attractive from a ground testing perspective. An extensive nuclear thermal rocket technology development effort was conducted from 1955-1973 under the Rover/NERVA Program. Both graphite and refractory metal alloy fuel types were pursued. Reactors and engines employing graphite based fuels were designed, built and ground tested. A number of fast spectrum reactor and engine designs employing refractory metal alloy fuel types were proposed and designed, but none were built. The Small Nuclear Rocket Engine (SNRE) was the last engine design studied by the Los Alamos National Laboratory during the program. At the time, this engine was a state-of-the-art graphite based fuel design incorporating lessons learned from the very successful technology development program. The SNRE was a nominal 16,000-lbf thrust engine originally intended for unmanned applications with relatively short engine

  9. Small Fast Spectrum Reactor Designs Suitable for Direct Nuclear Thermal Propulsion

    Science.gov (United States)

    Schnitzler, Bruce G.; Borowski, Stanley K.

    2012-01-01

    Advancement of U.S. scientific, security, and economic interests through a robust space exploration program requires high performance propulsion systems to support a variety of robotic and crewed missions beyond low Earth orbit. Past studies, in particular those in support of the Space Exploration Initiative (SEI), have shown nuclear thermal propulsion systems provide superior performance for high mass high propulsive delta-V missions. The recent NASA Design Reference Architecture (DRA) 5.0 Study re-examined mission, payload, and transportation system requirements for a human Mars landing mission in the post-2030 timeframe. Nuclear thermal propulsion was again identified as the preferred in-space transportation system. A common nuclear thermal propulsion stage with three 25,000-lbf thrust engines was used for all primary mission maneuvers. Moderately lower thrust engines may also have important roles. In particular, lower thrust engine designs demonstrating the critical technologies that are directly extensible to other thrust levels are attractive from a ground testing perspective. An extensive nuclear thermal rocket technology development effort was conducted from 1955-1973 under the Rover/NERVA Program. Both graphite and refractory metal alloy fuel types were pursued. Reactors and engines employing graphite based fuels were designed, built and ground tested. A number of fast spectrum reactor and engine designs employing refractory metal alloy fuel types were proposed and designed, but none were built. The Small Nuclear Rocket Engine (SNRE) was the last engine design studied by the Los Alamos National Laboratory during the program. At the time, this engine was a state-of-the-art graphite based fuel design incorporating lessons learned from the very successful technology development program. The SNRE was a nominal 16,000-lbf thrust engine originally intended for unmanned applications with relatively short engine operations and the engine and stage design were

  10. 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.)

  11. Experimental direct digital control of the power plant A1 reactor based on a modern control theory approach

    International Nuclear Information System (INIS)

    Karpeta, C.

    1979-01-01

    The objective of the project was to accumulate technical experience with application of modern control theory in nuclear power by carrying out a case study of an experimental direct digital control at the A1 reactor about its nominal steady state. The research has proved that slightly modified method of solution of the linear stochastic regulator problem can be successfully applied in design of digital control system of nuclear power reactors

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

    Science.gov (United States)

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

    1975-01-01

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

  13. Test Results From a Direct Drive Gas Reactor Simulator Coupled to a Brayton Power Conversion Unit

    Science.gov (United States)

    Hervol, David S.; Briggs, Maxwell H.; Owen, Albert K.; Bragg-Sitton, Shannon M.

    2009-01-01

    The Brayton Power Conversion Unit (BPCU) located at NASA Glenn Research Center (GRC) in Cleveland, OH is a closed cycle system incorporating a turboaltemator, recuperator, and gas cooler connected by gas ducts to an external gas heater. For this series of tests, the BPCU was modified by replacing the gas heater with the Direct Drive Gas heater or DOG. The DOG uses electric resistance heaters to simulate a fast spectrum nuclear reactor similar to those proposed for space power applications. The combined system thermal transient behavior was the focus of these tests. The BPCU was operated at various steady state points. At each point it was subjected to transient changes involving shaft rotational speed or DOG electrical input. This paper outlines the changes made to the test unit and describes the testing that took place along with the test results.

  14. Design of water rod cores of a direct cycle supercritical-pressure light water reactor

    International Nuclear Information System (INIS)

    Okano, Yasushi; Koshizuka, Sei-Ichi; Oka, Yoshiaki

    1994-01-01

    A conceptual design of a direct-cycle supercritical-pressure light water reactor with water rods is presented. Three types of water rods are analyzed: single, semi-double and full-double tubes. A water rod replaces seven fuel rods in a triangular lattice. The coolant density change in the water rods and the fuel channel is calculated using a code developed in the present study. The full double tube is most superior in terms of the distribution of the moderator. The number of fuel rods to water rods is 198:19, which makes optimum moderation. The average enrichment becomes 4.13%. The axial power flattening is finally achieved by partial length fuel rods and enrichment split of 0.25%. The discharge burnup is 45 GWd/t. (author)

  15. Preparation of Biodiesel from Microalgae and Palm Oil by Direct Transesterification in a Batch Microwave Reactor

    International Nuclear Information System (INIS)

    Marwan; Suhendrayatna; Indarti, E

    2015-01-01

    The present work was aimed to study the so-called direct transesterification of microalgae lipids to biodiesel in a batch microwave reactor. As a comparison, preparation of palm oil to biodiesel by alkaline catalyzed ethanolysis was also carried out. Palm oil biodiesel was recovered close to an equilibrium conversion (94-96% yield) under microwave heating for at least 6 min, while the conventional method required more than 45 minutes reaching the same yield. A very short reaction time suggests the benefit of microwave effect over conventional heating method in making biodiesel. FTIR analysis revealed the presence of fatty acid ethyl esters with no undesired chemical groups or compounds formed due to local heat generated by microwave effect, thus the conversion only followed transesterification route. Oil containing microalgae of Chlorella sp. isolated from the local brackish water pond was used as a potential source of biodiesel. High yield of biodiesel (above 0.6 g/g of dried algae) was also attainable for the direct transesterification of microalgae in the microwave reactor. Effect of water content of the algae biomass became insignificant at 11.9%(w/w) or less, related to the algae biomass dried for longer than 6 h. Fast transesterification of the algal oil towards equilibrium conversion was obtained at reaction time of 6 min, and at longer times the biodiesel yield remains unchanged. FAME profile indicates unsaturated fatty acids as major constituents. It was shown that microwave irradiation contributes not only to enhance the transeseterification, but also to assist effective release of fatty acid containing molecules (e.g. triacylglycerol, free fatty acids and phospholipids) from algal cells. (paper)

  16. Development of thin PdeAg supported membranes for fluidized bed membrane reactors including WGS related gases

    OpenAIRE

    Fernandez, E Ekain; Helmi Siasi Farimani, A Arash; Coenen, KT Kai; Meléndez, J; Viviente, JL; Pacheco Tanaka, DA; Sint Annaland, van, M Martin; Gallucci, F Fausto

    2015-01-01

    This paper reports the preparation, characterization and stability tests of Pd-based thin membranes for fluidized bed membrane reactor applications. Various thin membranes have been prepared by simultaneous Pd-Ag electroless plating. A simple technique for sealing of the produced membranes is reported and discussed. The membranes have been characterized for single gas permeation, and afterwards used for permeation of mixtures of gases and under fluidization conditions. The membranes have show...

  17. Sensorless Direct Flux Vector Control of Synchronous Reluctance Motors Including Standstill, MTPA and Flux Weakening

    OpenAIRE

    Yousefi-Talouki, Arzhang; Pescetto, Paolo; Pellegrino, Gian-Mario Luigi

    2017-01-01

    This paper proposes a sensorless direct flux vector control scheme for synchronous reluctance motor drives. Torque is controlled at constant switching frequency, via the closed loop regulation of the stator flux linkage vector and of the current component in quadrature with it, using the stator flux oriented reference frame. A hybrid flux and position observer combines back-electromotive force integration with pulsating voltage injection around zero speed. Around zero speed, the position obse...

  18. A multimethod investigation including direct observation of 3751 patient visits to 120 dental offices

    Directory of Open Access Journals (Sweden)

    Stephen Wotman

    2010-05-01

    Full Text Available Stephen Wotman1, Catherine A Demko1, Kristin Victoroff1, Joseph J Sudano2, James A Lalumandier11Department of Community Dentistry, Case Western Reserve University, School of Dental Medicine, Cleveland, OH, USA; 2Center of Health Care Research and Policy, Case Western Reserve University, School of Medicine, Cleveland, OH, USAAbstract: This report defines verbal interactions between practitioners and patients as core activities of dental practice. Trained teams spent four days in 120 Ohio dental practices observing 3751 patient encounters with dentists and hygienists. Direct observation of practice characteristics, procedures performed, and how procedure and nonprocedure time was utilized during patient visits was recorded using a modified Davis Observation Code that classified patient contact time into 24 behavioral categories. Dentist, hygienist, and patient characteristics were gathered by questionnaire. The most common nonprocedure behaviors observed for dentists were chatting, evaluation feedback, history taking, and answering patient questions. Hygienists added preventive counseling. We distinguish between preventive procedures and counseling in actual dental offices that are members of a practice-based research network. Almost a third of the dentist’s and half of the hygienist’s patient contact time is utilized for nonprocedure behaviors during patient encounters. These interactions may be linked to patient and practitioner satisfaction and effectiveness of self-care instruction.Keywords: dental practice, dental practice core activities, direct observation of dental practice, Dental Davis Observation Code, dentist, hygienist patient behaviors

  19. Direct measurement of the cross section of neutron-neutron scattering at the YAGUAR reactor. Substantiation of the experiment technique

    International Nuclear Information System (INIS)

    Chernukhin, Yu.G.; Kandiev, Ya.Z.; Lartsev, V.D.; Levakov, B.G.; Modestov, D.G.; Simonenko, V.A.; Streltsov, S.I.; Khmel'nitskij, D.V.

    2006-01-01

    The main stage of experiment for direct measurement of cross section of neutron-neutron scattering σ nn at low energies (E nn determination. It was shown, that for achieving the criterion ε ∼ 4% it will be necessary to have 40-50 pulses of a reactor [ru

  20. A new force field including charge directionality for TMAO in aqueous solution

    Energy Technology Data Exchange (ETDEWEB)

    Usui, Kota; Nagata, Yuki, E-mail: sulpizi@uni-mainz.de, E-mail: nagata@mpip-mainz.mpg.de; Hunger, Johannes; Bonn, Mischa [Max Planck Institute for Polymer Research, Ackermannweg 10, 55128 Mainz (Germany); Sulpizi, Marialore, E-mail: sulpizi@uni-mainz.de, E-mail: nagata@mpip-mainz.mpg.de [Johannes Gutenberg University Mainz, Staudingerweg 7, 55099 Mainz (Germany)

    2016-08-14

    We propose a new force field for trimethylamine N-oxide (TMAO), which is designed to reproduce the long-lived and highly directional hydrogen bond between the TMAO oxygen (O{sub TMAO}) atom and surrounding water molecules. Based on the data obtained by ab initio molecular dynamics simulations, we introduce three dummy sites around O{sub TMAO} to mimic the O{sub TMAO} lone pairs and we migrate the negative charge on the O{sub TMAO} to the dummy sites. The force field model developed here improves both structural and dynamical properties of aqueous TMAO solutions. Moreover, it reproduces the experimentally observed dependence of viscosity upon increasing TMAO concentration quantitatively. The simple procedure of the force field construction makes it easy to implement in molecular dynamics simulation packages and makes it compatible with the existing biomolecular force fields. This paves the path for further investigation of protein-TMAO interaction in aqueous solutions.

  1. A new force field including charge directionality for TMAO in aqueous solution

    International Nuclear Information System (INIS)

    Usui, Kota; Nagata, Yuki; Hunger, Johannes; Bonn, Mischa; Sulpizi, Marialore

    2016-01-01

    We propose a new force field for trimethylamine N-oxide (TMAO), which is designed to reproduce the long-lived and highly directional hydrogen bond between the TMAO oxygen (O TMAO ) atom and surrounding water molecules. Based on the data obtained by ab initio molecular dynamics simulations, we introduce three dummy sites around O TMAO to mimic the O TMAO lone pairs and we migrate the negative charge on the O TMAO to the dummy sites. The force field model developed here improves both structural and dynamical properties of aqueous TMAO solutions. Moreover, it reproduces the experimentally observed dependence of viscosity upon increasing TMAO concentration quantitatively. The simple procedure of the force field construction makes it easy to implement in molecular dynamics simulation packages and makes it compatible with the existing biomolecular force fields. This paves the path for further investigation of protein-TMAO interaction in aqueous solutions.

  2. An integrated mathematical model for chemical oxygen demand (COD) removal in moving bed biofilm reactors (MBBR) including predation and hydrolysis.

    Science.gov (United States)

    Revilla, Marta; Galán, Berta; Viguri, Javier R

    2016-07-01

    An integrated mathematical model is proposed for modelling a moving bed biofilm reactor (MBBR) for removal of chemical oxygen demand (COD) under aerobic conditions. The composite model combines the following: (i) a one-dimensional biofilm model, (ii) a bulk liquid model, and (iii) biological processes in the bulk liquid and biofilm considering the interactions among autotrophic, heterotrophic and predator microorganisms. Depending on the values for the soluble biodegradable COD loading rate (SCLR), the model takes into account a) the hydrolysis of slowly biodegradable compounds in the bulk liquid, and b) the growth of predator microorganisms in the bulk liquid and in the biofilm. The integration of the model and the SCLR allows a general description of the behaviour of COD removal by the MBBR under various conditions. The model is applied for two in-series MBBR wastewater plant from an integrated cellulose and viscose production and accurately describes the experimental concentrations of COD, total suspended solids (TSS), nitrogen and phosphorous obtained during 14 months working at different SCLRs and nutrient dosages. The representation of the microorganism group distribution in the biofilm and in the bulk liquid allow for verification of the presence of predator microorganisms in the second reactor under some operational conditions. Copyright © 2016 Elsevier Ltd. All rights reserved.

  3. Basic aspects of the concept of reactor compartment (including damaged compartments) management during utilization of nuclear powered submarines -- High priority R and D

    International Nuclear Information System (INIS)

    Mazokin, V.A.; Borisov, V.V.; Netecha, M.E.; Orlov, Yu.V.; Stanislavsky, G.A.; Ulasevich, V.K.

    1996-01-01

    Large-scale decommissioning of Russian nuclear-powered submarines (NPS) and their utilization prospects gave rise to numerous complicated scientific and technical, as well as economic, problems. Problems of handling of radioactive equipment from the reactor compartments (RC) are among the vital ones, arousing a growing concern with the public. Without solution of the problems the processes of NPS utilization can not be considered completed. It involves potential hazard, for the environment both from NPS being paid up (temporal on-float storage) with unloaded spent nuclear fuel (SNF), and RC, cut from submarine hull, containing highly radioactive equipment and materials but no SNF. Diverse variations of the concept of reactor compartment handling of NPS subject to, utilization are possible, but, in principle, there are essentially two variants: (1) RC utilization directly in the course of NPS utilization, envisaging removal of radioactive equipment from the reactor compartment and its cutting; (2) RC utilization after preliminary long-term storage of radioactive equipment of nuclear steam supply system (NSSS) in the standard sites within RC, when radioactivity of SGF equipment and materials as a result of natural decay of radionuclides is reduced to the values permitting RC cutting without employment of special means. When analyzing this or that variant of the concept, various factors are to be allowed for. First and foremost, the technology of RC handling is to provide: (1) minimum amount of radioactive wastes; (2) minimum radiation burden to the environment and personnel; (3) minimum material and financial expenditures; and (4) maximum radiation and ecological safety. RDIPE in cooperation with the leading enterprises of RF Minatom, Goskomoboronprom and Ministry of Defense has developed a concept of handling radioactive equipment in NPS reactor compartments, and its principle concepts are presented here

  4. Co-operation of the CMEA member countries in the development of different reactor types, including certain aspects of their nuclear fuel cycle

    International Nuclear Information System (INIS)

    Panasenkov, A.; Barbur, I.; Barchenkov, A.; Molnar, L.; Tolpygo, V.; Khake, V.; Shcherbinin, B.

    1977-01-01

    The report gives an account of the problems of the projected development of atomic power and evaluates its role in the fuel and power complex and long-range development of interconnected power systems of the CMEA member countries. The report emphasizes the importance of scientific and technical co-operation in the creation of thermal and fast-neutron power reactors with 1000-1500MW(e) units, and in the elaboration of nuclear plants for heating services. The positive experience of the international scientific and research group of scientists of the CMEA member countries carrying out reactor-physics studies on critical assemblies is mentioned. The report contains basic conclusions from the forecasts for nuclear power in the CMEA member countries up to 1990, including forecasting methodology; the role of nuclear power plants in saving natural and enriched uranium for a projected period; and the impact of nuclear power development rates on its structure (thermal and fast reactor ratio). It lists the impacts of scientific and technical co-operation of the CMEA member countries on the fuel cycle, including the transport of spent nuclear fuel, its recovery, reprocessing and radioactive waste disposal. Particular effects of co-operation of the CMEA member countries on the radiation safety of nuclear power plants and environmental protection are analysed. The report notes the role of the international economic associations Interatomenergo and Interatominstrument in the accelerated development of nuclear power. (author)

  5. Economic Analysis on Direct Use of Spent Pressurized Water Reactor Fuel in CANDU Reactors - IV: DUPIC Fuel Cycle Cost

    International Nuclear Information System (INIS)

    Ko, Won Il; Choi, Hangbok; Yang, Myung Seung

    2001-01-01

    This study examines the economics of the DUPIC fuel cycle using unit costs of fuel cycle components estimated based on conceptual designs. The fuel cycle cost (FCC) was calculated by a deterministic method in which reference values of fuel cycle components are used. The FCC was then analyzed by a Monte Carlo simulation to get the uncertainty of the FCC associated with the unit costs of the fuel cycle components. From the deterministic analysis on the equilibrium fuel cycle model, the DUPIC FCC was estimated to be 6.21 to 6.34 mills/kW.h for DUPIC fuel options, which is a little smaller than that of the once-through FCC by 0.07 to 0.27 mills/kW.h. Considering the uncertainty (0.40 to 0.44 mills/kW.h) of the FCC estimated by the Monte Carlo simulation method, the cost difference between the DUPIC and once-through fuel cycle is negligible. On the other hand, the material balance calculation has shown that the DUPIC fuel cycle can save natural uranium resources by ∼20% and reduce the spent fuel arising by ∼65% compared with the once-through fuel cycle. In conclusion, the DUPIC fuel cycle is comparable with the once-through fuel cycle from the viewpoint of FCC. In the future, it should be important to consider factors such as the environmental benefit owing to natural uranium savings, the capability of reusing spent pressurized water reactor fuel, and the safeguardability of the fuel cycle when deciding on an advanced nuclear fuel cycle option

  6. Reactor safeguards

    CERN Document Server

    Russell, Charles R

    1962-01-01

    Reactor Safeguards provides information for all who are interested in the subject of reactor safeguards. Much of the material is descriptive although some sections are written for the engineer or physicist directly concerned with hazards analysis or site selection problems. The book opens with an introductory chapter on radiation hazards, the construction of nuclear reactors, safety issues, and the operation of nuclear reactors. This is followed by separate chapters that discuss radioactive materials, reactor kinetics, control and safety systems, containment, safety features for water reactor

  7. Reactor vessel

    NARCIS (Netherlands)

    Makkee, M.; Kapteijn, F.; Moulijn, J.A.

    1999-01-01

    A reactor vessel (1) comprises a reactor body (2) through which channels (3) are provided whose surface comprises longitudinal inwardly directed parts (4) and is provided with a catalyst (6), as well as buffer bodies (8, 12) connected to the channels (3) on both sides of the reactor body (2) and

  8. Research reactors - an overview

    International Nuclear Information System (INIS)

    West, C.D.

    1997-01-01

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

  9. Isotopic and molecular fractionation in combustion; three routes to molecular marker validation, including direct molecular 'dating' (GC/AMS)

    Science.gov (United States)

    Currie, L. A.; Klouda, G. A.; Benner, B. A.; Garrity, K.; Eglinton, T. I.

    The identification of unique isotopic, elemental, and molecular markers for sources of combustion aerosol has growing practical importance because of the potential effects of fine particle aerosol on health, visibility and global climate. It is urgent, therefore, that substantial efforts be directed toward the validation of assumptions involving the use of such tracers for source apportionment. We describe here three independent routes toward carbonaceous aerosol molecular marker identification and validation: (1) tracer regression and multivariate statistical techniques applied to field measurements of mixed source, carbonaceous aerosols; (2) a new development in aerosol 14C metrology: direct, pure compound accelerator mass spectrometry (AMS) by off-line GC/AMS ('molecular dating'); and (3) direct observation of isotopic and molecular source emissions during controlled laboratory combustion of specific fuels. Findings from the combined studies include: independent support for benzo( ghi)perylene as a motor vehicle tracer from the first (statistical) and second (direct 'dating') studies; a new indication, from the third (controlled combustion) study, of a relation between 13C isotopic fractionation and PAH molecular fractionation, also linked with fuel and stage of combustion; and quantitative data showing the influence of both fuel type and combustion conditions on the yields of such species as elemental carbon and PAH, reinforcing the importance of exercising caution when applying presumed conservative elemental or organic tracers to fossil or biomass burning field data as in the first study.

  10. Degradation of nicotine in water solutions using a water falling film DBD plasma reactor: direct and indirect treatment

    Science.gov (United States)

    Krupež, Jelena; Kovačević, Vesna V.; Jović, Milica; Roglić, Goran M.; Natić, Maja M.; Kuraica, Milorad M.; Obradović, Bratislav M.; Dojčinović, Biljana P.

    2018-05-01

    Nicotine degradation efficiency in water solutions was studied using a water falling film dielectric barrier discharge (DBD) reactor. Two different treatments were applied: direct treatment, the recirculation of the solution through a DBD reactor, and indirect treatment, the bubbling of the gas from the DBD through the porous filter into the solution. In a separate experiment, samples spiked with nicotine in double distilled water (ddH2O) and tap water were studied and compared after both treatments. Furthermore, the effects of the homogeneous catalysts, namely, Fe2+ and H2O2, were tested in the direct treatment. Nicotine degradation efficiency was determined using high-performance liquid chromatography. A degradation efficiency of 90% was achieved after the direct treatment catalyzed with Fe2+. In order to analyze the biodegradability, mineralization level, and toxicity of the obtained solutions, after all degradation procedures the values of the following parameters were determined: total organic carbon, chemical oxygen demand, biochemical oxygen demand, and the Artemia salina toxicity test. The results showed that an increase in biodegradability was obtained, after all treatments. A partial nicotine mineralization was achieved and the mortality of the A. salina organism decreased in the treated samples, all of which indicating the effective removal of nicotine and the creation of less toxic solutions. Nicotine degradation products were identified using ultrahigh-performance liquid chromatography coupled with a linear ion trap Orbitrap hybrid mass spectrometer and a simple mechanism for oxidative degradation of nicotine in non-thermal plasma systems is proposed.

  11. Basic Experimental Simulation of a Traveling Wave Direct Energy Converter for a D-3He Fusion Reactor

    Science.gov (United States)

    Takeno, Hiromasa; Ikeda, Youichi; Yamada, Takashi; Noda, Kazuyuki; Yasaka, Yasuyoshi

    2000-09-01

    We simulated a traveling wave direct energy converter for application in a D-3He fusion reactor. The deceleration of a simulated ion beam was studied, and the dependence of the efficiency of the decelerator on its length was examined. The effectiveness of spatial variation in the phase velocity of the traveling wave was demonstrated experimentally. Optimization of the structure of the decelerator is discussed with the aid of numerical simulations. By employing the proper decelerator structure, the efficiency per unit length of about 0.2 with the normalized decelerator length of about 2.5 is expected.

  12. Basic experimental stimulation of a traveling wave direct energy converter for a D-3He fusion reactor

    International Nuclear Information System (INIS)

    Takeno, Hiromasa; Ikeda, Youichi; Yamada, Takashi; Noda, Kazuyuki; Yasaka, Yasuyoshi

    2000-01-01

    We simulated a traveling wave direct energy converter for application in a D- 3 He fusion reactor. The deceleration of a simulated ion beam was studied, and the dependence of the efficiency of the decelerator on its length was examined. The effectiveness of spatial variation in the phase velocity of the traveling wave was demonstrated experimentally. Optimization of the structure of the decelerator is discussed with the aid of numerical simulations. By employing the proper decelerator structure, the efficiency per unit length of about 0.2 with the normalized decelerator length of about 2.5 is expected. (author)

  13. Application of directional solidification ingot (LSD) in forging of PWR reactor vessel heads

    International Nuclear Information System (INIS)

    Benhamou, C.; Poitrault, I.

    1985-09-01

    Creusot-Loire Industrie uses this type of ingot for manufacture of Framatome 1300 and 1450 MW 4-loop PWR reactor vessel heads. This type of ingot offers a number advantages: improved internal soundness; greater chemical, structural and mechanical homogeneity of the finished part; simplified forging process. After a brief description of the pouring and solidification processes, this paper presents an analysis of the results of examinations performed on the prototype forging, as well as review of results obtained during industrial fabrication of dished heads from LSD ingots. The advantages of the LSD ingot over conventional ingots are discussed in conclusion

  14. Evaluation on ultrasonic examination methods applied to Ni-base alloy weld including cracks due to stress corrosion cracking found in BWR reactor internal

    International Nuclear Information System (INIS)

    Aoki, Takayuki; Kobayashi, Hiroyuki; Higuchi, Shinichi; Shimizu, Sadato

    2005-01-01

    A Ni-base alloy weld, including cracks due to stress corrosion cracking found in the reactor internal of the oldest BWR in Japan, Tsuruga unit 1, in 1999, was examined by three (3) types of UT method. After this examination, a depth of each crack was confirmed by carrying out a little excavation with a grinder and PT examination by turns until each crack disappeared. Then, the depth measured by the former method was compared with the one measured by the latter method. In this fashion, performances of the UT methods were verified. As a result, a combination of the three types of UT method was found to meet the acceptance criteria given by ASME Sec.XI Appendix VIII, Performance Demonstration for Ultrasonic Examination Systems-Supplement 6. In this paper, the results of the UT examination described above and their evaluation are discussed. (author)

  15. BOILING REACTORS

    Science.gov (United States)

    Untermyer, S.

    1962-04-10

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

  16. Particle Swarm Optimization-Based Direct Inverse Control for Controlling the Power Level of the Indonesian Multipurpose Reactor

    Directory of Open Access Journals (Sweden)

    Yoyok Dwi Setyo Pambudi

    2016-01-01

    Full Text Available A neural network-direct inverse control (NN-DIC has been simulated to automatically control the power level of nuclear reactors. This method has been tested on an Indonesian pool type multipurpose reactor, namely, Reaktor Serba Guna-GA Siwabessy (RSG-GAS. The result confirmed that this method still cannot minimize errors and shorten the learning process time. A new method is therefore needed which will improve the performance of the DIC. The objective of this study is to develop a particle swarm optimization-based direct inverse control (PSO-DIC to overcome the weaknesses of the NN-DIC. In the proposed PSO-DIC, the PSO algorithm is integrated into the DIC technique to train the weights of the DIC controller. This integration is able to accelerate the learning process. To improve the performance of the system identification, a backpropagation (BP algorithm is introduced into the PSO algorithm. To show the feasibility and effectiveness of this proposed PSO-DIC technique, a case study on power level control of RSG-GAS is performed. The simulation results confirm that the PSO-DIC has better performance than NN-DIC. The new developed PSO-DIC has smaller steady-state error and less overshoot and oscillation.

  17. Directing filtration to narrow molecular weight distribution of oligodextran in an enzymatic membrane reactor

    DEFF Research Database (Denmark)

    Su, Ziran; Luo, Jianquan; Pinelo, Manuel

    2018-01-01

    product, hypersaline wastewater discharge and potential safety hazards. In this work, a novel enzymatic membrane reactor (EMR) system to produce oligodextran is proposed, whereby in-situ product recovery can be manipulated to control the Mw distribution of the resulting products. Results showed......Oligodextrans with molecular weight (Mw) within the range of 5.0–8.0kDa have great commercial potential as precursors of iron-dextran for anemia treatment. Traditional oligodextran production consists of sucrose fermentation, acid hydrolysis and ethanol precipitation, which results in an uneven Mw...... narrow distribution and high productivity. Moreover, high transmembrane pressures (3 bars) and low stirring rates (160rpm) promoted yields beyond 50% in 120min. Higher permeate fluxes prevented further product hydrolysis and enhanced the yield. However, the resulting concentration polarization (CP...

  18. A service laboratory's view of the status and direction of reactor vessel surveillance

    International Nuclear Information System (INIS)

    Norris, E.B.

    1981-01-01

    Advances in testing techniques and analysis procedures have had a minor impact to date on the conduct of reactor vessel material surveillance programs. However, major thrusts in the near future will be associated with the development of elastic-plastic fracture toughness data on irradiated materials and improvements in analysis techniques for projecting surveillance results to the pressure vessel wall. In this regard, increased emphasis will be placed on the development of R-curves from the results of J-integral tests. Also, efforts will be increased to develop a better understanding of neutron irradiation embrittlement mechanisms, to determine if a time dependency of damage can lead to saturation and to evaluate the significance of small variations in irradiation temperature on the embrittlement response

  19. Carbon Footprint of Inbound Tourism to Iceland: A Consumption-Based Life-Cycle Assessment including Direct and Indirect Emissions

    Directory of Open Access Journals (Sweden)

    Hannah Sharp

    2016-11-01

    Full Text Available The greenhouse gas (GHG emissions caused by tourism have been studied from several perspectives, but few studies exist that include all direct and indirect emissions, particularly those from aviation. In this study, an input/output-based hybrid life-cycle assessment (LCA method is developed to assess the consumption-based carbon footprint of the average tourist including direct and indirect emissions. The total inbound tourism-related GHG emissions are also calculated within a certain region. As a demonstration of the method, the full carbon footprint of an average tourist is assessed as well as the total GHG emissions induced by tourism to Iceland over the period of 2010–2015, with the presented approach applicable in other contexts as well. Iceland provides an interesting case due to three features: (1 the tourism sector in Iceland is the fastest-growing industry in the country with an annual growth rate of over 20% over the past five years; (2 almost all tourists arrive by air; and (3 the country has an almost emissions-free energy industry and an import-dominated economy, which emphasise the role of the indirect emissions. According to the assessment, the carbon footprint for the average tourist is 1.35 tons of CO2-eq, but ranges from 1.1 to 3.2 tons of CO2-eq depending on the distance travelled by air. Furthermore, this footprint is increasing due to the rise in average flight distances travelled to reach the country. The total GHG emissions caused by tourism in Iceland have tripled from approximately 600,000 tons of CO2-eq in 2010 to 1,800,000 tons in 2015. Aviation accounts for 50%–82% of this impact (depending on the flight distance underlining the importance of air travel, especially as tourism-related aviation is forecasted to grow significantly in the near future. From a method perspective, the carbon footprinting application presented in the study would seem to provide an efficient way to study both the direct and indirect

  20. 3-dimensional magnetotelluric inversion including topography using deformed hexahedral edge finite elements and direct solvers parallelized on symmetric multiprocessor computers - Part II: direct data-space inverse solution

    Science.gov (United States)

    Kordy, M.; Wannamaker, P.; Maris, V.; Cherkaev, E.; Hill, G.

    2016-01-01

    Following the creation described in Part I of a deformable edge finite-element simulator for 3-D magnetotelluric (MT) responses using direct solvers, in Part II we develop an algorithm named HexMT for 3-D regularized inversion of MT data including topography. Direct solvers parallelized on large-RAM, symmetric multiprocessor (SMP) workstations are used also for the Gauss-Newton model update. By exploiting the data-space approach, the computational cost of the model update becomes much less in both time and computer memory than the cost of the forward simulation. In order to regularize using the second norm of the gradient, we factor the matrix related to the regularization term and apply its inverse to the Jacobian, which is done using the MKL PARDISO library. For dense matrix multiplication and factorization related to the model update, we use the PLASMA library which shows very good scalability across processor cores. A synthetic test inversion using a simple hill model shows that including topography can be important; in this case depression of the electric field by the hill can cause false conductors at depth or mask the presence of resistive structure. With a simple model of two buried bricks, a uniform spatial weighting for the norm of model smoothing recovered more accurate locations for the tomographic images compared to weightings which were a function of parameter Jacobians. We implement joint inversion for static distortion matrices tested using the Dublin secret model 2, for which we are able to reduce nRMS to ˜1.1 while avoiding oscillatory convergence. Finally we test the code on field data by inverting full impedance and tipper MT responses collected around Mount St Helens in the Cascade volcanic chain. Among several prominent structures, the north-south trending, eruption-controlling shear zone is clearly imaged in the inversion.

  1. Mirror reactor studies

    International Nuclear Information System (INIS)

    Moir, R.W.; Barr, W.L.; Bender, D.J.

    1977-01-01

    Design studies of a fusion mirror reactor, a fusion-fission mirror reactor, and two small mirror reactors are summarized. The fusion reactor uses 150-keV neutral-beam injectors based on the acceleration of negative ions. The injectors provide over 1 GW of continuous power at an efficiency greater than 80%. The fusion reactor has three-stage, modularized, Venetian blind, plasma direct converter with a predicted efficiency of 59% and a new concept for removal of the lune-shaped blanket: a crane is brought between the two halves of the Yin-Yang magnet, which are separated by a float. The design has desirable features such as steady-state operation, minimal impurity problems, and low first-wall thermal stress. The major disadvantage is low Q resulting in high re-circulating power and hence high cost of electrical power. However, the direct capital cost per unit of gross electrical power is reasonable [$1000/kW(e)]. By contrast, the fusion-fission reactor design is not penalized by re-circulating power and uses relatively near-term fusion technology being developed for the fusion power program. New results are presented on the Th- 233 U and the U- 239 Pu fuel cycles. The purpose of this hybrid is fuel production, with projected costs at $55/g of Pu or $127/g of 233 U. Blanket and cooling system designs, including an emergency cooling system, by General Atomic Company, lead us to the opinion that the reactor can meet expected safety standards for licensing. The smallest mirror reactor having only a shield between the plasma and the coil is the 4.2-m long fusion engineering research facility (FERF) designed for material irradiation. The smallest mirror reactor having both a blanket and shield is the 7.5-m long experimental power reactor (EPR), which has both a fusion and a fusion-fission version. (author)

  2. Antioxidant effects of coumarins include direct radical scavenging, metal chelation and inhibition of ROS-producing enzymes.

    Science.gov (United States)

    Filipský, Tomáš; Říha, Michal; Macáková, Kateřina; Anzenbacherová, Eva; Karlíčková, Jana; Mladěnka, Přemysl

    2015-01-01

    Coumarins represent a large group of 1,2-benzopyrone derivatives which have been identified in many natural sources and synthetized as well. Several studies have shown that their antioxidant capacity is not based only on direct scavenging of reactive oxygen and nitrogen species (RONS) but other mechanisms are also involved. These include: a) the chelation of transient metals iron and copper, which are known to catalyse the Fenton reaction; and b) the inhibition of RONS-producing enzymes (e.g. xanthine oxidase, myeloperoxidase and lipoxygenase), suggesting that mechanism(s) involved on cellular level are complex and synergistic. Moreover, many factors must be taken into account when analysing structure-antioxidant capacity relationships of coumarins due to different in vitro/in vivo methodological approaches. The structural features necessary for the direct RONS scavenging and metal chelation are apparently similar and the ideal structures are 6,7-dihydroxy- or 7,8-dihydroxycoumarins. However, the clinical outcome is unknown, because these coumarins are able to reduce copper and iron, and may thus paradoxically potentiate the Fenton chemistry. The similar structural features appear to be associated with inhibition of lipoxygenase, probably due to interference with iron in its active site. Contrarily, 6,7-dihydroxycoumarin seems to be the most active coumarin in the inhibition of xanthine oxidase while its derivative bearing the 4-methyl group or 7,8-dihydroxycoumarin are less active or inactive. In addition, coumarins may hinder the induction of inducible NO-synthase and cyclooxygenase- 2. Sparse data on inhibition of myeloperoxidase do not enable any clear conclusion, but some coumarins may block it.

  3. Integrity of the reactor vessel support system for a postulated reactor vessel closure head drop event

    International Nuclear Information System (INIS)

    Kim, Tae Wan; Lee, Ki Young; Lee, Dae Hee; Kim, Kang Soo

    1996-01-01

    The integrity of reactor vessel support system of the Korean Standard Nuclear Power Plant (KSNPP) is investigated for a postulated reactor vessel closure head drop event. The closure head is disassembled from the reactor vessel during refueling process or general inspection of reactor vessel and internal structures, and carried to proposed location by the head lift rig. A postulated closure head drop event could be anticipated during closure head handling process. The drop event may cause an impact load on the reactor vessel and supporting system. The integrity of the supporting system is directly relevant to that of reactor vessel and reactor internals including fuels. Results derived by elastic impact analysis, linear and non-linear buckling analysis and elasto-plastic stress analysis of the supporting system implied that the integrity of the reactor vessel supporting system is intact for a postulated reactor closure head drop event. 9 refs., 10 figs., 3 tabs. (author)

  4. The management of metastatic radioiodine-refractory differentiated thyroid cancer requires an integrated approach including both directed and systemic therapies.

    Science.gov (United States)

    Cooray, Shamil D; Topliss, Duncan J

    2017-01-01

    A 58-year-old man with metastatic radioiodine-refractory differentiated thyroid cancer (DTC) presented with left thigh and right flank numbness. He had known progressive and widespread bony metastases, for which he received palliative radiotherapy, and multiple bilateral asymptomatic pulmonary metastases. CT scan and MRI of the spine revealed metastases at right T10-L1 vertebrae with extension into the central canal and epidural disease at T10 and T11 causing cord displacement and canal stenosis but retention of spinal cord signal. Spinal surgery was followed by palliative radiotherapy resulting in symptom resolution. Two months later, sorafenib received approval for use in Australia and was commenced and up-titrated with symptomatic management of mild adverse effects. Follow-up CT scan three months after commencement of sorafenib revealed regression of pulmonary metastases but no evident change in most bone metastases except for an advancing lesion eroding into the right acetabulum. The patient underwent a right total hip replacement, intra-lesional curettage and cementing. After six months of sorafenib therapy, CT scanning showed enlarging liver lesions with marked elevation of serum thyroglobulin. Lenvatinib was commenced and sorafenib was ceased. He now has stable disease with a falling thyroglobulin more than 5 years after metastatic radioiodine-refractory DTC was diagnosed. In DTC, 5% of distant metastases become radioiodine-refractory, resulting in a median overall survival of 2.5-3.5 years. Tyrosine kinase inhibitor (TKI) therapy has recently been demonstrated to increase progression-free survival in these patients but poses some unique management issues and is best used as part of an integrated approach with directed therapy. Directed therapies may have greater potential to control localised disease and related symptoms when compared to systemic therapies.Consider TKI therapy in progressive disease where benefits outweigh risks.Active surveillance and

  5. In-vessel Retention Strategy for High Power Reactors - K-INERI Final Report (includes SBLB Test Results for Task 3 on External Reactor Vessel Cooling (ERVC) Boiling Data and CHF Enhancement Correlations)

    Energy Technology Data Exchange (ETDEWEB)

    F. B. Cheung; J. Yang; M. B. Dizon; J. Rempe

    2005-01-01

    In-vessel retention (IVR) of core melt is a key severe accident management strategy adopted by some operating nuclear power plants and proposed for some advanced light water reactors (ALWRs). If there were inadequate cooling during a reactor accident, a significant amount of core material could become molten and relocate to the lower head of the reactor vessel, as happened in the Three Mile Island Unit 2 (TMI-2) accident. If it is possible to ensure that the vessel head remains intact so that relocated core materials are retained within the vessel, the enhanced safety associated with these plants can reduce concerns about containment failure and associated risk. For example, the enhanced safety of the Westinghouse Advanced 600 MWe PWR (AP600), which relied upon External Reactor Vessel Cooling (ERVC) for IVR, resulted in the U.S. Nuclear Regulatory Commission (US NRC) approving the design without requiring certain conventional features common to existing LWRs. However, it is not clear that currently proposed external reactor vessel cooling (ERVC) without additional enhancements could provide sufficient heat removal for higher-power reactors (up to 1500 MWe). Hence, a collaborative, three-year, U.S. - Korean International Nuclear Energy Research Initiative (INERI) project was completed in which the Idaho National Engineering and Environmental Laboratory (INEEL), Seoul National University (SNU), Pennsylvania State University (PSU), and the Korea Atomic Energy Research Institute (KAERI) investigated the performance of ERVC and an in-vessel core catcher (IVCC) to determine if IVR is feasible for reactors up to 1500 MWe.

  6. Use of computational fluid dynamics codes for safety analysis of nuclear reactor systems, including containment. Summary report of a technical meeting

    International Nuclear Information System (INIS)

    2003-11-01

    Safety analysis is an important tool for justifying the safety of nuclear power plants. Typically, this type of analysis is performed by means of system computer codes with one dimensional approximation for modelling real plant systems. However, in the nuclear area there are issues for which traditional treatment using one dimensional system codes is considered inadequate for modelling local flow and heat transfer phenomena. There is therefore increasing interest in the application of three dimensional computational fluid dynamics (CFD) codes as a supplement to or in combination with system codes. There are a number of both commercial (general purpose) CFD codes as well as special codes for nuclear safety applications available. With further progress in safety analysis techniques, the increasing use of CFD codes for nuclear applications is expected. At present, the main objective with respect to CFD codes is generally to improve confidence in the available analysis tools and to achieve a more reliable approach to safety relevant issues. An exchange of views and experience can facilitate and speed up progress in the implementation of this objective. Both the International Atomic Energy Agency (IAEA) and the Nuclear Energy Agency of the Organisation for Economic Co-operation and Development (OECD/NEA) believed that it would be advantageous to provide a forum for such an exchange. Therefore, within the framework of the Working Group on the Analysis and Management of Accidents of the NEA's Committee on the Safety of Nuclear Installations, the IAEA and the NEA agreed to jointly organize the Technical Meeting on the Use of Computational Fluid Dynamics Codes for Safety Analysis of Reactor Systems, including Containment. The meeting was held in Pisa, Italy, from 11 to 14 November 2002. The publication constitutes the report of the Technical Meeting. It includes short summaries of the presentations that were made and of the discussions as well as conclusions and

  7. A traveling wave direct energy converter for a D-{sup 3}He fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sato, K.; Katayama, H.; Miyawaki, F. [Himeji Inst. of Tech., Hyogo (Japan); Tajima, T.

    1994-12-31

    A concept of a traveling wave direct energy converter (TWDEC) is developed for 14.7-MeV fusion protons based on the principle of a backward wave oscillator. Separation of fusion protons from thermal ions is accomplished by using ExB ion drift. Energy conversion rate up to 0.87 is attained by applying three-stage modulation of the proton beam. A one-dimensional particle-circuit code is developed to examine self-excitation of the traveling wave and its stability under loading. Electrostatic wave with a fixed frequency is excited spontaneously, and stability of the wave is ensured under loading. (author).

  8. Tratamiento del Colorante Azo Rojo Directo 23 mediante Reactores Discontinuos Secuenciados Anaerobios/Aerobios Treatment of Direct Red 23 Azo Dye by Anaerobic/Aerobic Sequencing Batch Reactors

    OpenAIRE

    Fernanda Morales-Guzmán; Rosa M Melgoza-Alemán

    2009-01-01

    Se evaluó el tratamiento del colorante azo Rojo Directo 23 mediante un proceso anaerobio/aerobio integrado en un sólo reactor. Se utilizaron dos reactores: uno de lecho expandido con carbón activado granular como mediador redox y otro conteniendo la biomasa suspendida, ambos operados en discontinuo. Se evaluó la eficiencia del proceso mediante la reducción y transformación del colorante a aminas aromáticas totales en la fase anaerobia y la degradación de éstas en la fase aerobia. La eliminaci...

  9. Research reactors

    International Nuclear Information System (INIS)

    Kowarski, L.

    1955-01-01

    It brings together the techniques data which are involved in the discussion about the utility for a research institute to acquire an atomic reactor for research purposes. This type of decision are often taken by non-specialist people who can need a brief presentation of a research reactor and its possibilities in term of research before asking advises to experts. In a first part, it draws up a list of the different research programs which can be studied by getting a research reactor. First of all is the reactor behaviour and kinetics studies (reproducibility factor, exploration of neutron density, effect of reactor structure, effect of material irradiation...). Physical studies includes study of the behaviour of the control system, studies of neutron resonance phenomena and study of the fission process for example. Chemical studies involves the study of manipulation and control of hot material, characterisation of nuclear species produced in the reactor and chemical effects of irradiation on chemical properties and reactions. Biology and medicine research involves studies of irradiation on man and animals, genetics research, food or medical tools sterilization and neutron beams effect on tumour for example. A large number of other subjects can be studied in a reactor research as reactor construction material research, fabrication of radioactive sources for radiographic techniques or applied research as in agriculture or electronic. The second part discussed the technological considerations when choosing the reactor type. The technological factors, which are considered for its choice, are the power of the reactor, the nature of the fuel which is used, the type of moderator (water, heavy water, graphite or BeO) and the reflector, the type of coolants, the protection shield and the control systems. In the third part, it described the characteristics (place of installation, type of combustible and comments) and performance (power, neutron flux ) of already existing

  10. Advanced-fueled fusion reactors suitable for direct energy conversion. Project note: temperature-gradient enhancement of electrical fields in insulators

    International Nuclear Information System (INIS)

    Blum, A.S.; Mancebo, L.

    1976-01-01

    Direct energy converters for use on controlled fusion reactors utilize electrodes operated at elevated voltages and temperatures. The insulating elements that position these electrodes must support large voltages and under some circumstances large thermal gradients. It is shown that even modest thermal gradients can cause major alterations of the electric-field distribution within the insulating element

  11. Gas cooled reactor assessment. Volume II. Final report, February 9, 1976--June 30, 1976

    International Nuclear Information System (INIS)

    1976-08-01

    This report was prepared to document the estimated power plant capital and operating costs, and the safety and environmental assessments used in support of the Gas Cooled Reactor Assessment performed by Arthur D. Little, Inc. (ADL), for the U.S. Energy Research and Development Administration. The gas-cooled reactor technologies investigated include: the High Temperature Gas Reactor Steam Cycle (HTGR-SC), the HTGR Direct Cycle (HTGR-DC), the Very High Temperature Reactor (VHTR) and the Gas Cooled Fast Reactor (GCFR). Reference technologies used for comparison include: Light Water Reactors (LWR), the Liquid Metal Fast Breeder Reactor (LMFBR), conventional coal-fired steam plants, and coal combustion for process heat

  12. Nuclear reactor vessel fuel thermal insulating barrier

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-03-19

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

  13. Evaluation of the process performance of a down-flow hanging sponge reactor for direct treatment of domestic wastewater in Bangkok, Thailand.

    Science.gov (United States)

    Miyaoka, Yuma; Yoochatchaval, Wilasinee; Sumino, Haruhiko; Banjongproo, Pathan; Yamaguchi, Takashi; Onodera, Takashi; Okadera, Tomohiro; Syutsubo, Kazuaki

    2017-08-24

    This study assesses the performance of an aerobic trickling filter, down-flow hanging sponge (DHS) reactor, as a decentralized domestic wastewater treatment technology. Also, the characteristic eukaryotic community structure in DHS reactor was investigated. Long-term operation of a DHS reactor for direct treatment of domestic wastewater (COD = 150-170 mg/L and BOD = 60-90 mg/L) was performed under the average ambient temperature ranged from 28°C to 31°C in Bangkok, Thailand. Throughout the evaluation period of 550 days, the DHS reactor at a hydraulic retention time of 3 h showed better performance than the existing oxidation ditch process in the removal of organic carbon (COD removal rate = 80-83% and BOD removal rate = 91%), nitrogen compounds (total nitrogen removal rate = 45-51% and NH 4 + -N removal rate = 95-98%), and low excess sludge production (0.04 gTS/gCOD removed). The clone library based on the 18S ribosomal ribonucleic acid gene sequence revealed that phylogenetic diversity of 18S rRNA gene in the DHS reactor was higher than that of the present oxidation ditch process. Furthermore, the DHS reactor also demonstrated sufficient COD and NH 4 + -N removal efficiency under flow rate fluctuation conditions that simulates a small-scale treatment facility. The results show that a DHS reactor could be applied as a decentralized domestic wastewater treatment technology in tropical regions such as Bangkok, Thailand.

  14. CO{sub 2} direct cycles suitable for AGR type reactors; Cycles directs de gaz carbonique applicables aux reacteurs du genre AGR

    Energy Technology Data Exchange (ETDEWEB)

    Maillet, E. [Commissariat a l' Energie Atomique. Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France)

    1967-10-01

    The perspectives given by the gas turbines under pressure, to build simple nuclear power plants and acieving significantly high yield, are specified. The CO{sub 2} is characterised by by good efficiency under moderate temperature (500 to 750 Celsius degrees), compactness and the simpleness of machines and the safe exploitation (supply, storage, relief cooling, thermosyphon). The revision of thermal properties of the CO{sub 2} and loss elements show that several direct cycles would fit in particular to the AGR type reactors. Cycles that would diverge a little from classical models and able to lead to power and heat generation can lead by simple means to the best results. Several satisfying solutions present for the starting up, the power regulation and the stopping. The nuclear power plant components and the functioning safety are equally considered in the present report. The conclusions stimulate the studies and realizations of carbon dioxide gas turbines in when approprite. [French] Les perspectives offertes par la turbine a gaz sous pression, pour construire des centrales nucleaires simples et de rendement progressivement eleve, se precisent actuellement. le CO{sub 2} se distingue par sa bonne efficacite a temperature moderee (500 a 750 degres celsius), la compacite et la simplicite des machines, et la surete qu'il apporte a l'exploitation ( approvisionnement, stockage, refroidissement de secours, thermosiphon). La revision des proprietes thermophysiques du CO{sub 2} et des elements de pertes montre que divers cycles directs conviendraient en particulier aux reacteurs agr ou derives. Des cycles s'ecartant peu des modeles classiques, et se pretant ulterieurement a la production simultanee d'electricite et de chaleur, peuvent conduire par des moyens simples aux meilleurs resultats d'ensemble. Plusieurs solutions satisfaisantes se presentent pour le demarrage, le reglage de la puissance et l'arret. Les composants de la centrale et la

  15. Nuclear reactor

    International Nuclear Information System (INIS)

    Rau, P.

    1980-01-01

    The reactor core of nuclear reactors usually is composed of individual elongated fuel elements that may be vertically arranged and through which coolant flows in axial direction, preferably from bottom to top. With their lower end the fuel elements gear in an opening of a lower support grid forming part of the core structure. According to the invention a locking is provided there, part of which is a control element that is movable along the fuel element axis. The corresponding locking element is engaged behind a lateral projection in the opening of the support grid. The invention is particularly suitable for breeder or converter reactors. (orig.) [de

  16. The management of metastatic radioiodine-refractory differentiated thyroid cancer requires an integrated approach including both directed and systemic therapies

    Directory of Open Access Journals (Sweden)

    Shamil D Cooray

    2017-02-01

    Full Text Available A 58-year-old man with metastatic radioiodine-refractory differentiated thyroid cancer (DTC presented with left thigh and right flank numbness. He had known progressive and widespread bony metastases, for which he received palliative radiotherapy, and multiple bilateral asymptomatic pulmonary metastases. CT scan and MRI of the spine revealed metastases at right T10–L1 vertebrae with extension into the central canal and epidural disease at T10 and T11 causing cord displacement and canal stenosis but retention of spinal cord signal. Spinal surgery was followed by palliative radiotherapy resulting in symptom resolution. Two months later, sorafenib received approval for use in Australia and was commenced and up-titrated with symptomatic management of mild adverse effects. Follow-up CT scan three months after commencement of sorafenib revealed regression of pulmonary metastases but no evident change in most bone metastases except for an advancing lesion eroding into the right acetabulum. The patient underwent a right total hip replacement, intra-lesional curettage and cementing. After six months of sorafenib therapy, CT scanning showed enlarging liver lesions with marked elevation of serum thyroglobulin. Lenvatinib was commenced and sorafenib was ceased. He now has stable disease with a falling thyroglobulin more than 5 years after metastatic radioiodine-refractory DTC was diagnosed. In DTC, 5% of distant metastases become radioiodine-refractory, resulting in a median overall survival of 2.5–3.5 years. Tyrosine kinase inhibitor (TKI therapy has recently been demonstrated to increase progression-free survival in these patients but poses some unique management issues and is best used as part of an integrated approach with directed therapy.

  17. SREB, a GATA transcription factor that directs disparate fates in Blastomyces dermatitidis including morphogenesis and siderophore biosynthesis.

    Directory of Open Access Journals (Sweden)

    Gregory M Gauthier

    2010-04-01

    Full Text Available Blastomyces dermatitidis belongs to a group of human pathogenic fungi that exhibit thermal dimorphism. At 22 degrees C, these fungi grow as mold that produce conidia or infectious particles, whereas at 37 degrees C they convert to budding yeast. The ability to switch between these forms is essential for virulence in mammals and may enable these organisms to survive in the soil. To identify genes that regulate this phase transition, we used Agrobacterium tumefaciens to mutagenize B. dermatitidis conidia and screened transformants for defects in morphogenesis. We found that the GATA transcription factor SREB governs multiple fates in B. dermatitidis: phase transition from yeast to mold, cell growth at 22 degrees C, and biosynthesis of siderophores under iron-replete conditions. Insertional and null mutants fail to convert to mold, do not accumulate significant biomass at 22 degrees C, and are unable to suppress siderophore biosynthesis under iron-replete conditions. The defect in morphogenesis in the SREB mutant was independent of exogenous iron concentration, suggesting that SREB promotes the phase transition by altering the expression of genes that are unrelated to siderophore biosynthesis. Using bioinformatic and gene expression analyses, we identified candidate genes with upstream GATA sites whose expression is altered in the null mutant that may be direct or indirect targets of SREB and promote the phase transition. We conclude that SREB functions as a transcription factor that promotes morphogenesis and regulates siderophore biosynthesis. To our knowledge, this is the first gene identified that promotes the conversion from yeast to mold in the dimorphic fungi, and may shed light on environmental persistence of these pathogens.

  18. Nuclear reactor physics course for reactor operators

    International Nuclear Information System (INIS)

    Baeten, P.

    2006-01-01

    The education and training of nuclear reactor operators is important to guarantee the safe operation of present and future nuclear reactors. Therefore, a course on basic 'Nuclear reactor physics' in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The aim of the basic course on 'Nuclear Reactor Physics for reactor operators' is to provide the reactor operators with a basic understanding of the main concepts relevant to nuclear reactors. Seen the education level of the participants, mathematical derivations are simplified and reduced to a minimum, but not completely eliminated

  19. Tracking the direct impact of rainfall on groundwater at Mt. Fuji by multiple analyses including microbial DNA

    Science.gov (United States)

    Sugiyama, Ayumi; Masuda, Suguru; Nagaosa, Kazuyo; Tsujimura, Maki; Kato, Kenji

    2018-02-01

    A total of 2 to 3 million tons of spring water flushes out from the foot of Mt. Fuji, the largest volcanic mountain in Japan. Based on the concept of piston flow transport, residence time of stored groundwater at Mt. Fuji was estimated at ˜ 15-30 years by the 36Cl / Cl ratio (Tosaki et al., 2011). This range, however, represents the average residence time of groundwater that was mixed before it flushed out. To elucidate the route of groundwater in a given system, we determined signatures of direct impacts of rainfall on groundwater, using microbial, stable isotopic (δ18O), and chemical analyses (concentration of silica). Chemical analysis of the groundwater gave an average value of the water, which was already mixed with waters from various sources and routes in the subsurface environment. The microbial analysis suggested locations of water origin and paths. In situ observation during four rainfall events revealed that the stable oxygen isotopic signature obtained from spring water (at 726 m a.s.l., site SP-0 m) and shallow groundwater (at 150 m a.s.l., site GW-42 m), where the average recharge height from rainfall was 1700-1800 m, became greater than values observed prior to a torrential rain producing more than 300 mm of precipitation. The concentration of silica decreased after this event. In addition, the abundance of Bacteria in spring water increased, suggesting the influence of heavy rain. Such changes did not appear when rainfall was less than 100 mm per event. The above findings indicate a rapid flow of rain through the shallow part of the aquifer, which appeared within a few weeks of torrential rain extracting abundant microbes from soil in the studied geologic setting. Interestingly, we found that after the torrential rain, the abundance of Archaea increased in the deep groundwater at site GW-550 m, ˜ 12 km downstream of SP-0 m. However, chemical parameters did not show any change after the event. This suggests that strengthened piston flow caused by

  20. Development of seismic isolation system in vertical direction. Coupled response analysis of reactor building and common deck isolation system

    International Nuclear Information System (INIS)

    Morishita, Masaki

    1997-04-01

    Study was made on the effect of dynamic interaction between reactor building and common deck isolation structure. Dynamic response analyses were made with a coupled and a decoupled model of the reactor building and the isolation structure. The effect of the dynamic interaction was evaluated by comparing these analysis results. The results of the study can be summarized as below; (1) Dynamic coupling effect between reactor building and the common deck isolation structure is not significant. This can be attributed to the fact that the isolation frequency is sufficiently lower than the fundamental frequency of the soil-structure system. (2) The decoupled analysis model was found to give response results with practically sufficient accuracy and slight conservatism. (3) The effect of the response of common deck isolation structure on the response of the reactor building is also negligible. (author)

  1. Structure, Stability and Emissions of Lean Direct Injection Combustion, including a Novel Multi-Point LDI System for NOx Reduction

    Science.gov (United States)

    Villalva Gomez, Rodrigo

    Experimental research on Lean Direct Injection (LDI) combustors for gas turbine applications is presented. LDI combustion is an alternative to lean premixed combustion which has the potential of equivalent reduction of oxides of nitrogen (NOx) emissions and of peak combustor exit temperatures, but without some drawbacks of premixed combustors, such as flashback and autoignition. Simultaneous observations of the velocity field and reaction zone of an LDI swirl-stabilized combustor with a mixing tube at atmospheric conditions, with the goal of studying the flame stabilization mechanism, are shown. The flame was consistently anchored at the shear layer formed by the high-speed reactants exiting the mixing tube and the low speed recirculation region. Individual image analysis of the location of the tip of the recirculation zone and tip of the reaction region confirmed previously observed trends, but showed that calculation of the distance between these two points for corresponding image pairs yields results no different than when calculated from random image pairs. This most likely indicates a lag in the anchoring of the flame to changes in the recirculation zone, coupled with significant stochastic variation. An alternate LDI approach, multi-point LDI (MLDI), is also tested experimentally. A single large fuel nozzle is replaced by multiple small fuel nozzles to improve atomization and reduce the total volume of the high-temperature, low velocity recirculation zones, reducing NOx formation. The combustor researched employs a novel staged approach to allow good performance across a wide range of conditions by using a combination of nozzle types optimized to various power settings. The combustor has three independent fuel circuits referenced as pilot, intermediate, and outer. Emissions measurements, OH* chemiluminescence imaging, and thermoacoustic instability studies were run in a pressurized combustion facility at pressures from 2.0 to 5.3 bar. Combustor performance

  2. Tracking the direct impact of rainfall on groundwater at Mt. Fuji by multiple analyses including microbial DNA

    Directory of Open Access Journals (Sweden)

    A. Sugiyama

    2018-02-01

    Full Text Available A total of 2 to 3 million tons of spring water flushes out from the foot of Mt. Fuji, the largest volcanic mountain in Japan. Based on the concept of piston flow transport, residence time of stored groundwater at Mt. Fuji was estimated at  ∼  15–30 years by the 36Cl ∕ Cl ratio (Tosaki et al., 2011. This range, however, represents the average residence time of groundwater that was mixed before it flushed out. To elucidate the route of groundwater in a given system, we determined signatures of direct impacts of rainfall on groundwater, using microbial, stable isotopic (δ18O, and chemical analyses (concentration of silica. Chemical analysis of the groundwater gave an average value of the water, which was already mixed with waters from various sources and routes in the subsurface environment. The microbial analysis suggested locations of water origin and paths. In situ observation during four rainfall events revealed that the stable oxygen isotopic signature obtained from spring water (at 726 m a.s.l., site SP-0 m and shallow groundwater (at 150 m a.s.l., site GW-42 m, where the average recharge height from rainfall was 1700–1800 m, became greater than values observed prior to a torrential rain producing more than 300 mm of precipitation. The concentration of silica decreased after this event. In addition, the abundance of Bacteria in spring water increased, suggesting the influence of heavy rain. Such changes did not appear when rainfall was less than 100 mm per event. The above findings indicate a rapid flow of rain through the shallow part of the aquifer, which appeared within a few weeks of torrential rain extracting abundant microbes from soil in the studied geologic setting. Interestingly, we found that after the torrential rain, the abundance of Archaea increased in the deep groundwater at site GW-550 m,  ∼  12 km downstream of SP-0 m. However, chemical parameters did not show any change

  3. American National Standard: fire-protection program criteria for research reactors

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    This standard provides criteria for a fire protection program for research reactor facilities and for the reactor safety-related systems included in those facilities. It stresses preservation of the capability to achieve and maintain safe shutdown of the reactor, and includes consideration of both direct fire hazards and indirect or consequential hazards

  4. Chemical Reactors.

    Science.gov (United States)

    Kenney, C. N.

    1980-01-01

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

  5. Direct digital control of furnaces irradiated in nuclear reactors; Surveillance et regulation multiplexee par calculateur numerique de fours irradies

    Energy Technology Data Exchange (ETDEWEB)

    Joumard, R. [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

    1969-07-01

    An experimental direct digital control system has been realised in the 'C.E.N.G.', in order to verify that a computer makes easier the control of the experiments done in the nuclear reactors and to solve the theoretical and technical difficulties. The regulation is applied to thermal processes. The sampled data systems theory permits to choose the type of an efficient and simple digital compensator, and to establish a diagram which gives the values of the correcting parameters (obtained by minimizing the difference between the output and the input when perturbations occur). The programme execute, in simultaneity, supervision and regulation. Complex possibilities of printing out measures and alarms existed. The computer works out an incremental correction which makes step motors to turn. These motors act on the heating organs. The theoretical values and answers have been confirmed. The accuracy was limited essentially by the input quantification (1/1000 th). The comfort of such a system has been noticeable. (author) [French] Une installation de controle numerique direct fut realisee a titre experimental au C.E.N.G pour verifier qu'un ordinateur rendait plus aisee l'exploitation des experiences faites en pile nucleaire et pour degager les difficultes theoriques et techniques. La regulation s'applique a des processus thermiques. La theorie des systemes echantillonnes a permis de choisir un type de correcteur numerique simple et efficace et d'etablir un abaque qui donne les valeurs des parametres correcteurs minimisant les ecarts enregistres entre la reponse et la consigne en presence de perturbations. Le programme effectuait simultanement de la surveillance et de la regulation. Une restitution complexe des informations et des alarmes sur machine a ecrire etait possible. Le calculateur elaborait une correction incrementielle qui faisait tourner des moteurs pas a pas, lesquels commandaient les organes de puissance de chauffage. Les valeurs

  6. Report of the Federal Ministry for the Environment, Protection of Nature and Reactor Safety, on the reactor accident of Chernobyl, its repercussions, and precautions taken or to be taken - including addenda

    International Nuclear Information System (INIS)

    Petroll, M.

    1986-01-01

    Apart from the report of the Federal Ministry for the Environment, the publication contains the following chapters: 1) Monitoring of environmental radioactivity; 2) analysis of propagation processes; 3) control and measuring points of the Federal Laender to monitor environmental radioactivity; 4) determination of the local dose rate; 5) concentration of radioactivity in air and soil (graphs); 6) up-to-date knowledge of events, measures; 7) nuclear power plants in the Federal Republic of Germany (review of technical safety); 8) interim report of the Committee on Reactor Safety - Reaktorsicherheitskommission - for preliminary evaluation; 9) interim report of the Committee on Radiation Protection - Strahlenschutzkommission - for assessment and evaluation of the effects. (HP) [de

  7. Nuclear reactor apparatus

    International Nuclear Information System (INIS)

    Braun, H.E.; Bonnet, H.P.

    1978-01-01

    The reactor and its containment, instead of being supported on a solid concrete pad, are supported on a truss formed of upper and lower reinforced horizontal plates and vertical walls integrated into a rigid structure. The plates and walls from chambers within which the auxiliary components of the reactor, such as valves, pumping equipment and various tanks, are disposed. Certain of the chambers are also access passages for personnel, pipe chases, valve chambers and the like. In particular the truss includes an annular chamber. This chamber is lined and sealed by a corrosion-resistant liner and contains coolant and serves as a refueling cooling storage tank. This tank is directly below the primary-coolant conductor loops which extend from the reactor above the upper plate. The upper plate includes a sump connected to the tank through which coolant flows into the tank in the event of the occurrence of a loss-of-coolant accident. The truss extends beyond the containment and has chambers in the extending annulus. Pumps for circulating the coolant between the refueling coolant storage tank and the reactor are provided in certain of these chambers. The pumps are connected to the reactor by relatively short coolant conductors. Access to these pumps is readily afforded through hatches in the extending annulus

  8. 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.

  9. Department of reactor technology

    International Nuclear Information System (INIS)

    1980-01-01

    The activities of the Department of Reactor Technology at Risoe during 1979 are described. The work is presented in five chapters: Reactor Engineering, Reactor Physics and Dynamics, Heat Transfer and Hydraulics, The DR 1 Reactor, and Non-Nuclear Activities. A list of the staff and of publications is included. (author)

  10. Ozone generation by negative direct current corona discharges in dry air fed coaxial wire-cylinder reactors

    Science.gov (United States)

    Yehia, Ashraf; Mizuno, Akira

    2013-05-01

    An analytical study was made in this paper for calculating the ozone generation by negative dc corona discharges. The corona discharges were formed in a coaxial wire-cylinder reactor. The reactor was fed by dry air flowing with constant rates at atmospheric pressure and room temperature, and stressed by a negative dc voltage. The current-voltage characteristics of the negative dc corona discharges formed inside the reactor were measured in parallel with concentration of the generated ozone under different operating conditions. An empirical equation was derived from the experimental results for calculating the ozone concentration generated inside the reactor. The results, that have been recalculated by using the derived equation, have agreed with the experimental results over the whole range of the investigated parameters, except in the saturation range for the ozone concentration. Therefore, the derived equation represents a suitable criterion for expecting the ozone concentration generated by negative dc corona discharges in dry air fed coaxial wire-cylinder reactors under any operating conditions in range of the investigated parameters.

  11. Nuclear reactor neutron shielding

    Science.gov (United States)

    Speaker, Daniel P; Neeley, Gary W; Inman, James B

    2017-09-12

    A nuclear reactor includes a reactor pressure vessel and a nuclear reactor core comprising fissile material disposed in a lower portion of the reactor pressure vessel. The lower portion of the reactor pressure vessel is disposed in a reactor cavity. An annular neutron stop is located at an elevation above the uppermost elevation of the nuclear reactor core. The annular neutron stop comprises neutron absorbing material filling an annular gap between the reactor pressure vessel and the wall of the reactor cavity. The annular neutron stop may comprise an outer neutron stop ring attached to the wall of the reactor cavity, and an inner neutron stop ring attached to the reactor pressure vessel. An excore instrument guide tube penetrates through the annular neutron stop, and a neutron plug comprising neutron absorbing material is disposed in the tube at the penetration through the neutron stop.

  12. A green approach to ethyl acetate: Quantitative conversion of ethanol through direct dehydrogenation in a Pd-Ag membrane reactor

    KAUST Repository

    Zeng, Gaofeng

    2012-11-07

    Pincers do the trick: The conversion of ethanol to ethyl acetate and hydrogen was achieved using a pincer-Ru catalyst in a Pd-Ag membrane reactor. Near quantitative conversions and yields could be achieved without the need for acid or base promoters or hydrogen acceptors (see scheme). © 2012 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

  13. Applying the model of Goal-Directed Behavior, including descriptive norms, to physical activity intentions: A contribution to improving the Theory of Planned Behavior

    Science.gov (United States)

    The theory of planned behavior (TPB) has received its fair share of criticism lately, including calls for it to retire. We contributed to improving the theory by testing extensions such as the model of goal-directed behavior (MGDB, which adds desire and anticipated positive and negative emotions) ap...

  14. Test reactor: basic to U.S. breeder reactor development

    International Nuclear Information System (INIS)

    Miller, B.J.; Harness, A.J.

    1975-01-01

    Long-range energy planning in the U. S. includes development of a national commercial breeder reactor program. U. S. development of the LMFBR is following a conservative sequence of extensive technology development through use of test reactors and demonstration plants prior to construction of commercial plants. Because materials and fuel technology development is considered the first vital step in this sequence, initial U. S. efforts have been directed to the design and construction of a unique test reactor. The Fast Flux Test Facility, FFTF, is a 400 MW(t) reactor with driver fuel locations, open test locations, and closed loops for higher risk experiments. The FFTF will provide a prototypic LMFBR core environment with sufficient instrumentation for detailed core environmental characterization and a testing capability substituted for breeder capability. The unique comprehensive fuel and materials testing capability of the FFTF will be key to achieving long-range objectives of increased power density, improved breeding gain and shorter doubling times. (auth)

  15. Seismic attenuation system for a nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Liszkai, Tamas; Cadell, Seth

    2018-01-30

    A system for attenuating seismic forces includes a reactor pressure vessel containing nuclear fuel and a containment vessel that houses the reactor pressure vessel. Both the reactor pressure vessel and the containment vessel include a bottom head. Additionally, the system includes a base support to contact a support surface on which the containment vessel is positioned in a substantially vertical orientation. An attenuation device is located between the bottom head of the reactor pressure vessel and the bottom head of the containment vessel. Seismic forces that travel from the base support to the reactor pressure vessel via the containment vessel are attenuated by the attenuation device in a direction that is substantially lateral to the vertical orientation of the containment vessel.

  16. Reactor Neutrinos

    Directory of Open Access Journals (Sweden)

    Soo-Bong Kim

    2013-01-01

    Full Text Available 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 recently the most precise determination of the neutrino mixing angle θ13. This paper provides an overview of the upcoming experiments and of the projects under development, including the determination of the neutrino mass hierarchy and the possible use of neutrinos for society, for nonproliferation of nuclear materials, and geophysics.

  17. Survey of research reactors

    International Nuclear Information System (INIS)

    Boek, H.; Villa, M.

    2004-06-01

    A survey of reasearch reactors based on the IAEA Nuclear Research Reactor Data Base (RRDB) was done. This database includes information on 273 operating research reactors ranging in power from zero to several hundred MW. From these 273 operating research reactors 205 reactors have a power level below 5 MW, the remaining 68 reactors range from 5 MW up to several 100 MW thermal power. The major reactor types with common design are: Siemens Unterrichtsreaktors, 1.2 Argonaut reactors, Slowpoke reactors, the miniature neutron source reactors, TRIGA reactors, material testing reactors and high flux reactors. Technical data such as: power, fuel material, fuel type, enrichment, maximum neutron flux density and experimental facilities for each reactor type as well as a description of their utilization in physics and chemistry, medicine and biology, academic research and teaching, training purposes (students and physicists, operating personnel), industrial application (neutron radiography, silicon neutron transmutation doping facilities) are provided. The geographically distribution of these reactors is also shown. As conclusions the author discussed the advantages (low capital cost, low operating cost, low burn up, simple to operate, safe, less restrictive containment and sitting requirements, versatility) and disadvantages (lower sensitivity for NAA, limited radioisotope production, limited use of neutron beams, limited access to the core, licensing) of low power research reactors. 24 figs., refs. 15, Tab. 1 (nevyjel)

  18. A direct method for numerical solution of a class of nonlinear Volterra integro-differential equations and its application to the nonlinear fission and fusion reactor kinetics

    International Nuclear Information System (INIS)

    Nakahara, Yasuaki; Ise, Takeharu; Kobayashi, Kensuke; Itoh, Yasuyuki

    1975-12-01

    A new method has been developed for numerical solution of a class of nonlinear Volterra integro-differential equations with quadratic nonlinearity. After dividing the domain of the variable into subintervals, piecewise approximations are applied in the subintervals. The equation is first integrated over a subinterval to obtain the piecewise equation, to which six approximate treatments are applied, i.e. fully explicit, fully implicit, Crank-Nicolson, linear interpolation, quadratic and cubic spline. The numerical solution at each time step is obtained directly as a positive root of the resulting algebraic quadratic equation. The point reactor kinetics with a ramp reactivity insertion, linear temperature feedback and delayed neutrons can be described by one of this type of nonlinear Volterra integro-differential equations. The algorithm is applied to the Argonne benchmark problem and a model problem for a fast reactor without delayed neutrons. The fully implicit method has been found to be unconditionally stable in the sense that it always gives the positive real roots. The cubic spline method is divergent, and the other four methods are intermediate in between. From the estimation of the stability, convergency, accuracy and CPU time, it is concluded that the Crank-Nicolson method is best, then the linear interpolation method comes closely next to it. Discussions are also made on the possibility of applying the algorithm to the fusion reactor kinetics in the form of a nonlinear partial differential equation. (auth.)

  19. Reactor Physics Training

    International Nuclear Information System (INIS)

    Baeten, P.

    2007-01-01

    University courses in nuclear reactor physics at the universities consist of a theoretical description of the physics and technology of nuclear reactors. In order to demonstrate the basic concepts in reactor physics, training exercises in nuclear reactor installations are also desirable. Since the number of reactor facilities is however strongly decreasing in Europe, it becomes difficult to offer to students a means for demonstrating the basic concepts in reactor physics by performing training exercises in nuclear installations. Universities do not generally possess the capabilities for performing training exercises. Therefore, SCK-CEN offers universities the possibility to perform (on a commercial basis) training exercises at its infrastructure consisting of two research reactors (BR1 and VENUS). Besides the organisation of training exercises in the framework of university courses, SCK-CEN also organizes theoretical courses in reactor physics for the education and training of nuclear reactor operators. It is indeed a very important subject to guarantee the safe operation of present and future nuclear reactors. In this framework, an understanding of the fundamental principles of nuclear reactor physics is also necessary for reactor operators. Therefore, the organisation of a basic Nuclear reactor physics course at the level of reactor operators in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The objectives this activity are: (1) to provide training and education activities in reactor physics for university students and (2) to organise courses in nuclear reactor physics for reactor operators

  20. Safeguarding research reactors

    International Nuclear Information System (INIS)

    Powers, J.A.

    1983-03-01

    The report is organized in four sections, including the introduction. The second section contains a discussion of the characteristics and attributes of research reactors important to safeguards. In this section, research reactors are described according to their power level, if greater than 25 thermal megawatts, or according to each fuel type. This descriptive discussion includes both reactor and reactor fuel information of a generic nature, according to the following categories. 1. Research reactors with more than 25 megawatts thermal power, 2. Plate fuelled reactors, 3. Assembly fuelled reactors. 4. Research reactors fuelled with individual rods. 5. Disk fuelled reactors, and 6. Research reactors fuelled with aqueous homogeneous fuel. The third section consists of a brief discussion of general IAEA safeguards as they apply to research reactors. This section is based on IAEA safeguards implementation documents and technical reports that are used to establish Agency-State agreements and facility attachments. The fourth and last section describes inspection activities at research reactors necessary to meet Agency objectives. The scope of the activities extends to both pre and post inspection as well as the on-site inspection and includes the examination of records and reports relative to reactor operation and to receipts, shipments and certain internal transfers, periodic verification of fresh fuel, spent fuel and core fuel, activities related to containment and surveillance, and other selected activities, depending on the reactor

  1. Towards nuclear fusion reactors

    International Nuclear Information System (INIS)

    1993-11-01

    The results of nuclear fusion researches in JAERI are summarized. In this report, following themes are collected: the concept of fusion reactor (including ITER), fusion reactor safety, plasma confinement, fusion reactor equipment, and so on. Includes glossary. (J.P.N.)

  2. Neutronic reactor

    International Nuclear Information System (INIS)

    Carleton, J.T.

    1977-01-01

    A graphite-moderated nuclear reactor includes channels between blocks of graphite and also includes spacer blocks between adjacent channeled blocks with an axis of extension normal to that of the axis of elongation of the channeled blocks to minimize changes in the physical properties of the graphite as a result of prolonged neutron bombardment. 3 claims, 6 figures

  3. On the definition of a DEMO (demonstration) reactor

    International Nuclear Information System (INIS)

    Cole, H.C.; Challender, R.S.

    1987-01-01

    The paper is an appendix to a study of the reactor relevance of the NET design concept. The latter study examines whether the technologies and design principles proposed for NET can be directly extrapolated to a demonstration (DEMO) reactor. The authors have suggested a definition of a DEMO, and listed what they considered to be the most important implications of this definition. A table of parameters is included comparing published DEMO's with typical commercial reactor and 'pre-DEMO' studies. (U.K.)

  4. Survey of the laser-solenoid fusion reactor

    International Nuclear Information System (INIS)

    Amherd, N.A.

    1975-09-01

    This report surveys the prospects for a laser-solenoid fusion reactor. A sample reactor and scaling laws are used to describe the concept's characteristics. Experimental results are reviewed, and the laser and magnet technologies that undergird the laser-solenoid concept are analyzed. Finally, a systems analysis of fusion power reactors is given, including a discussion of direct conversion and fusion-fission effects, to ascertain the system attributes of the laser-solenoid configuration

  5. Future uses of research reactors

    International Nuclear Information System (INIS)

    Fleming, R.F.

    1990-01-01

    The use of research reactors in the next decade will be driven by two key forces: the retirement of obsolete facilities and the development of new capabilities, both in the United States and worldwide. Most research reactors in use in the United States began operation in the period from the late 1950s to the middle of the 1960s, both at universities and at the national laboratories. New design concepts include the 10-MW MAPLE by Atomic Energy of Canada Limited intended to provide an order-of-magnitude upgrade for the 2-MW university swimming-pool facilities and the advanced neutron source by the U.S. Department of Energy intended as the ultimate steady-state neutron facility. A rational program for U.S. research reactors might include the following components: (1) an inexpensive and very reliable reactor to replace the underutilized or underfunded university facilities of 2 MW or less; (2) a simple design in the 10- to 20-MW range to provide full research capability for some universities and other laboratories (the MAPLE design is a step in this direction); and (3) one or more of a high-end reactor with a power level up to 250 MW. Recent experience has shown that an important design criterion for any such facilities must be reliability of operation. All of these reactors would provide high-quality neutron beams as well as a variety of irradiation facilities

  6. Water cooled nuclear reactor

    International Nuclear Information System (INIS)

    1975-01-01

    A description is given of a cooling water intake collector for a nuclear reactor. It includes multiple sub-collectors extending out in a generally parallel manner to each other, each one having a first end and a second one separated along their length, and multiple water outlets for connecting each one to a corresponding pressure tube of the reactor. A first end tube and a second one connect the sub-collector tubes together to their first and second ends respectively. It also includes multiple collector tubes extending transversely by crossing over the sub-collector tubes and separated from each other in the direction of these tubes. Each collector tubes has a water intake for connecting to a water pump and multiple connecting tubes separated over its length and connecting each one to the corresponding sub-collector [fr

  7. Fuel arrangement for high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Tobin, J.M.

    1978-01-01

    Disclosed is a fuel arrangement for a high temperature gas cooled reactor including fuel assemblies with separate directly cooled fissile and fertile fuel elements removably inserted in an elongated moderator block also having a passageway for control elements

  8. Potential of direct metal deposition technology for manufacturing thick functionally graded coatings and parts for reactors components

    Science.gov (United States)

    Thivillon, L.; Bertrand, Ph.; Laget, B.; Smurov, I.

    2009-03-01

    Direct metal deposition (DMD) is an automated 3D deposition process arising from laser cladding technology with co-axial powder injection to refine or refurbish parts. Recently DMD has been extended to manufacture large-size near-net-shape components. When applied for manufacturing new parts (or their refinement), DMD can provide tailored thermal properties, high corrosion resistance, tailored tribology, multifunctional performance and cost savings due to smart material combinations. In repair (refurbishment) operations, DMD can be applied for parts with a wide variety of geometries and sizes. In contrast to the current tool repair techniques such as tungsten inert gas (TIG), metal inert gas (MIG) and plasma welding, laser cladding technology by DMD offers a well-controlled heat-treated zone due to the high energy density of the laser beam. In addition, this technology may be used for preventative maintenance and design changes/up-grading. One of the advantages of DMD is the possibility to build functionally graded coatings (from 1 mm thickness and higher) and 3D multi-material objects (for example, 100 mm-sized monolithic rectangular) in a single-step manufacturing cycle by using up to 4-channel powder feeder. Approved materials are: Fe (including stainless steel), Ni and Co alloys, (Cu,Ni 10%), WC compounds, TiC compounds. The developed coatings/parts are characterized by low porosity (mechanical support structures, hermetic joints, tubes with complex geometry, and tailored inside and outside surface properties, etc.

  9. Technical specifications, Hanford production reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gilbert, W.D. [comp.

    1962-06-25

    These technical specifications are applicable to the eight operating production reactor facilities, B, C, D, DR, F, H, KE, and KW. Covered are operating and performance restrictions and administrative procedures. Areas covered by the operating and performance restrictions are reactivity, reactor control and safety elements, power level, temperature and heat flux, reactor fuel loadings, reactor coolant systems, reactor confinement, test facilities, code compliance, and reactor scram set points. Administrative procedures include process control procedures, training programs, audits and inspections, and reports and records.

  10. Pyro-electrochemical reprocessing of irradiated MOX fast reactor fuel, testing of the reprocessing process with direct MOX fuel production

    International Nuclear Information System (INIS)

    Kormilitzyn, M.V.; Vavilov, S.K.; Bychkov, A.V.; Skiba, O.V.; Chistyakov, V.M.; Tselichshev, I.V.

    2000-01-01

    One of the advanced technologies for fast reactor fuel recycle is pyro-electrochemical molten salt technology. In 1998 we began to study the next phase of the irradiated oxide fuel reprocessing new process MOX → MOX. This process involves the following steps: - Dissolution of irradiated fuel in molten alkaline metal chlorides, - Purification of melt from fission products that are co-deposited with uranium and plutonium oxides, - Electrochemical co-deposition of uranium and plutonium oxides under the controlled cathode potential, - Production of granulated MOX (crushing,salt separation and sizing), and - Purification of melt from fission products by phosphate precipitation. In 1998 a series of experiments were prepared and carried out in order to validate this process. It was shown that the proposed reprocessing flowsheet of irradiated MOX fuel verified the feasibility of its decontamination from most of its fission products (rare earths, cesium) and minor-actinides (americium, curium)

  11. Loss mechanisms of travelling wave direct energy converters for D-{sup 3}He FRC fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ishikawa, M. E-mail: misikawa@kz.tsukuba.ac.jp; Yamane, T.; Shimizu, Y.; Momota, H.; Tomita, Y.; Miley, G.H

    2000-11-01

    Loss mechanisms of TWDEC for a D-{sup 3}He fuelled FRC fusion reactor are studied with numerical analysis. (1) The self-excitation of a travelling wave has been attained with external electric circuits, which shows the conversion efficiency is about 70-73%. (2) Three-dimensional behavior has been studied by axisymmetric approximation, showing that the loss related to the radial non-uniformity effect without collision with grids is about 5%. (3) The loss related to the collision with grids is the most important, being about 11%. (4) Effects of secondary electrons produced by the collision with the grids are rather small and subrings can suppress the effects of secondary electrons. (5) The loss related to leaked 3.6 MeV {alpha} particles is about 45-60% of the total energy of leaked {alpha} particles. (6) The gross efficiency of TWDEC is estimated to be over 60%.

  12. Fluidized-bed nuclear reactor

    International Nuclear Information System (INIS)

    Grimmett, E.S.; Kunze, J.F.

    1975-01-01

    A reactor vessel containing a fluidized-bed region of particulate material including both a neutron-moderating and a fertile substance is described. A gas flow including fissile material passes through the vessel at a sufficient rate to fluidize the particulate material and at a sufficient density to support a thermal fission reaction within the fluidized-bed region. The high-temperature portion of a heat transfer system is located within the fluidized-bed region of the reactor vessel in direct contact with the fluidized particles. Heat released by fission is thereby transferred at an enhanced rate to a coolant circulating within the heat transfer system. Fission products are continuously removed from the gas flow and supplemental fissile material added during the reactor operation. (U.S.)

  13. Reactor licensing

    International Nuclear Information System (INIS)

    Harvie, J.D.

    2002-01-01

    This presentation discusses reactor licensing and includes the legislative basis for licensing, other relevant legislation , the purpose of the Nuclear Safety and Control Act, important regulations, regulatory document, policies, and standards. It also discusses the role of the CNSC, its mandate and safety philosophy

  14. Licensing of nuclear reactor operators

    International Nuclear Information System (INIS)

    1979-09-01

    Recommendations are presented for the licensing of nuclear reactor operators in units licensed according to the legislation in effect. They apply to all physical persons designated by the Operating Organization of the nuclear reactor or reactors to execute any of the following functional activities: a) to manipulate the controls of a definite reactor b) to direct the authorized activities of the reactor operators licesed according to the present recommendations. (F.E.) [pt

  15. Reactor power measuring device

    International Nuclear Information System (INIS)

    Izumi, Mikio; Sano, Yuji; Seki, Eiji; Yoshida, Toshifumi; Ito, Toshiaki.

    1993-01-01

    The present invention provides a self-powered long detector having a sensitivity over the entire length of a reactor core as an entire control rod withdrawal range of a BWR type reactor, and a reactor power measuring device using a gamma ray thermometer which scarcely causes sensitivity degradation. That is, a hollow protection pipe is disposed passing through the reactor core from the outside of a reactor pressure vessel. The self-powered long detectors and the gamma ray thermometers are inserted and installed in the protection pipe. An average reactor power in an axial direction of the reactor relative to a certain position in the horizontal cross section of the reactor core is determined based on the power of the self-powered long detector over the entire length of the reactor core. Since the response of the self-powered detector relative to a local power change is rapid, the output is used as an input signal to a safety protection device of the reactor core. Further, a gamma ray thermometer secured in the reactor and having scarce sensitivity degradation is used instead of an incore travelling neutron monitor used for relative calibration of an existent neutron monitor secured in the reactor. (I.S.)

  16. An innovative way of thinking nuclear waste management - Neutron physics of a reactor directly operating on SNF.

    Science.gov (United States)

    Merk, Bruno; Litskevich, Dzianis; Bankhead, Mark; Taylor, Richard J

    2017-01-01

    A solution for the nuclear waste problem is the key challenge for an extensive use of nuclear reactors as a major carbon free, sustainable, and applied highly reliable energy source. Partitioning and Transmutation (P&T) promises a solution for improved waste management. Current strategies rely on systems designed in the 60's for the massive production of plutonium. We propose an innovative strategic development plan based on invention and innovation described with the concept of developments in s-curves identifying the current boundary conditions, and the evolvable objectives. This leads to the ultimate, universal vision for energy production characterized by minimal use of resources and production of waste, while being economically affordable and safe, secure and reliable in operation. This vision is transformed into a mission for a disruptive development of the future nuclear energy system operated by burning of existing spent nuclear fuel (SNF) without prior reprocessing. This highly innovative approach fulfils the sustainability goals and creates new options for P&T. A proof on the feasibility from neutronic point of view is given demonstrating sufficient breeding of fissile material from the inserted SNF. The system does neither require new resources nor produce additional waste, thus it provides a highly sustainable option for a future nuclear system fulfilling the requests of P&T as side effect. In addition, this nuclear system provides enhanced resistance against misuse of Pu and a significantly reduced fuel cycle. However, the new system requires a demand driven rethinking of the separation process to be efficient.

  17. Reactor System Design

    International Nuclear Information System (INIS)

    Chi, S. K.; Kim, G. K.; Yeo, J. W.

    2006-08-01

    SMART NPP(Nuclear Power Plant) has been developed for duel purpose, electricity generation and energy supply for seawater desalination. The objective of this project IS to design the reactor system of SMART pilot plant(SMART-P) which will be built and operated for the integrated technology verification of SMART. SMART-P is an integral reactor in which primary components of reactor coolant system are enclosed in single pressure vessel without connecting pipes. The major components installed within a vessel includes a core, twelve steam generator cassettes, a low-temperature self pressurizer, twelve control rod drives, and two main coolant pumps. SMART-P reactor system design was categorized to the reactor coe design, fluid system design, reactor mechanical design, major component design and MMIS design. Reactor safety -analysis and performance analysis were performed for developed SMART=P reactor system. Also, the preparation of safety analysis report, and the technical support for licensing acquisition are performed

  18. Applying the Model of Goal-Directed Behavior, Including Descriptive Norms, to Physical Activity Intentions: A Contribution to Improving the Theory of Planned Behavior.

    Science.gov (United States)

    Esposito, Gabriele; van Bavel, René; Baranowski, Tom; Duch-Brown, Néstor

    2016-08-01

    The theory of planned behavior (TPB) has received its fair share of criticism lately, including calls for it to retire. We contribute to improving the theory by testing extensions such as the model of goal-directed behavior (MGDB, which adds desire and anticipated positive and negative emotions) applied to physical activity (PA) intention. We also test the inclusion of a descriptive norms construct as an addition to the subjective norms construct, also applied to PA, resulting in two additional models: TPB including descriptive norms (TPB + DN) and MGDB including descriptive norms (MGDB + DN). The study is based on an online survey of 400 young adult Internet users, previously enrolled in a subject pool. Confirmatory factor analysis (CFA) showed that TPB and TPB + DN were not fit for purpose, while MGDB and MGDB + DN were. Structural equation modelling (SEM) conducted on MGDB and MGDB + DN showed that the inclusion of descriptive norms took over the significance of injunctive norms, and increased the model's account of total variance in intention to be physically active. © The Author(s) 2016.

  19. A STUDY OF CHINESE YUAN (RMB APPRECIATION ACCOMPANYING WITH OTHERS FACTORS INCLUDING FOREIGN DIRECT INVESTMENT (FDI AND THEIR EFFECT ON CHINA ECONOMY

    Directory of Open Access Journals (Sweden)

    Ping-fu (Brian LAI

    2014-07-01

    Full Text Available The Chinese Yuan (RMB has been on the trend of appreciation over the last decade, and such a trend will likely be continuing for some years over the next decade. According to some scholars in their published literatures, the appreciation of RMB, the influx of Foreign Direct Investment (FDI has been ongoing accompanying the sustained growing economy in mainland China over the past decade. It is believed that the China economy has an implication from some significant factors including appreciation of RMB, interest rate of RMB, inflation and continuous increase of FDI for the next several years. The present study aims to provide an emphasis on investigation into effect on China economy as a result of appreciation of RMB and FDI together with some other factors, and to provide an outlook on the economy in China for the coming decades. First, a review was carried on relevant background information and development history of RMB and FDI. There are many reasons and factors behind leading to the sustained growth in the economy in China in the last decade and such effects were in coverage in the literature review. An overview of the development of RMB exchange mechanism, and other variables including (1 RMB exchange rate, (2 China interest rate, (3 Foreign Direct Investment (FDI, (4 Trade Balance of China, (5 Annual Inflation rate in China, (6 Energy Consumption in China, (7 Foreign Exchange Reserve in China, (8 China wages, (9 China External Debt and (10 China Consumer Price Index, which may have effect on the growth of the economy in China is covered in the literature review conducted in Chapter 2.

  20. Reactor core of nuclear reactor

    International Nuclear Information System (INIS)

    Sasagawa, Masaru; Masuda, Hiroyuki; Mogi, Toshihiko; Kanazawa, Nobuhiro.

    1994-01-01

    In a reactor core, a fuel inventory at an outer peripheral region is made smaller than that at a central region. Fuel assemblies comprising a small number of large-diameter fuel rods are used at the central region and fuel assemblies comprising a great number of smalldiameter fuel rods are used at the outer peripheral region. Since a burning degradation rate of the fuels at the outer peripheral region can be increased, the burning degradation rate at the infinite multiplication factor of fuels at the outer region can substantially be made identical with that of the fuels in the inner region. As a result, the power distribution in the direction of the reactor core can be flattened throughout the entire period of the burning cycle. Further, it is also possible to make the degradation rate of fuels at the outer region substantially identical with that of fuels at the inner side. A power peak formed at the outer circumferential portion of the reactor core of advanced burning can be lowered to improve the fuel integrity, and also improve the reactor safety and operation efficiency. (N.H.)

  1. An innovative way of thinking nuclear waste management – Neutron physics of a reactor directly operating on SNF

    Science.gov (United States)

    Litskevich, Dzianis; Bankhead, Mark; Taylor, Richard J.

    2017-01-01

    A solution for the nuclear waste problem is the key challenge for an extensive use of nuclear reactors as a major carbon free, sustainable, and applied highly reliable energy source. Partitioning and Transmutation (P&T) promises a solution for improved waste management. Current strategies rely on systems designed in the 60’s for the massive production of plutonium. We propose an innovative strategic development plan based on invention and innovation described with the concept of developments in s-curves identifying the current boundary conditions, and the evolvable objectives. This leads to the ultimate, universal vision for energy production characterized by minimal use of resources and production of waste, while being economically affordable and safe, secure and reliable in operation. This vision is transformed into a mission for a disruptive development of the future nuclear energy system operated by burning of existing spent nuclear fuel (SNF) without prior reprocessing. This highly innovative approach fulfils the sustainability goals and creates new options for P&T. A proof on the feasibility from neutronic point of view is given demonstrating sufficient breeding of fissile material from the inserted SNF. The system does neither require new resources nor produce additional waste, thus it provides a highly sustainable option for a future nuclear system fulfilling the requests of P&T as side effect. In addition, this nuclear system provides enhanced resistance against misuse of Pu and a significantly reduced fuel cycle. However, the new system requires a demand driven rethinking of the separation process to be efficient. PMID:28749952

  2. An innovative way of thinking nuclear waste management - Neutron physics of a reactor directly operating on SNF.

    Directory of Open Access Journals (Sweden)

    Bruno Merk

    Full Text Available A solution for the nuclear waste problem is the key challenge for an extensive use of nuclear reactors as a major carbon free, sustainable, and applied highly reliable energy source. Partitioning and Transmutation (P&T promises a solution for improved waste management. Current strategies rely on systems designed in the 60's for the massive production of plutonium. We propose an innovative strategic development plan based on invention and innovation described with the concept of developments in s-curves identifying the current boundary conditions, and the evolvable objectives. This leads to the ultimate, universal vision for energy production characterized by minimal use of resources and production of waste, while being economically affordable and safe, secure and reliable in operation. This vision is transformed into a mission for a disruptive development of the future nuclear energy system operated by burning of existing spent nuclear fuel (SNF without prior reprocessing. This highly innovative approach fulfils the sustainability goals and creates new options for P&T. A proof on the feasibility from neutronic point of view is given demonstrating sufficient breeding of fissile material from the inserted SNF. The system does neither require new resources nor produce additional waste, thus it provides a highly sustainable option for a future nuclear system fulfilling the requests of P&T as side effect. In addition, this nuclear system provides enhanced resistance against misuse of Pu and a significantly reduced fuel cycle. However, the new system requires a demand driven rethinking of the separation process to be efficient.

  3. Nuclear reactors

    International Nuclear Information System (INIS)

    Barre, Bertrand

    2015-10-01

    After some remarks on the nuclear fuel, on the chain reaction control, on fuel loading and unloading, this article proposes descriptions of the design, principles and operations of different types of nuclear reactors as well as comments on their presence and use in different countries: pressurized water reactors (design of the primary and secondary circuits, volume and chemistry control, backup injection circuits), boiling water reactors, heavy water reactors, graphite and boiling water reactors, graphite-gas reactors, fast breeder reactors, and fourth generation reactors (definition, fast breeding). For these last ones, six concepts are presented: sodium-cooled fast reactor, lead-cooled fast reactor, gas-cooled fast reactor, high temperature gas-cooled reactor, supercritical water-cooled reactor, and molten salt reactor

  4. Reactor core and passive safety systems descriptions of a next generation pressure tube reactor - mechanical aspects

    Energy Technology Data Exchange (ETDEWEB)

    Yetisir, M.; Gaudet, M.; Rhodes, D.; Hamilton, H.; Pencer, J. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2014-07-01

    Canada has been developing a channel-type supercritical water-cooled nuclear reactor concept, often called the Canadian SCWR. The objective of this reactor concept is to meet the technology goals of the Generation IV International Forum (GIF) for the next generation nuclear reactor development, which include enhanced safety features (inherent safe operation and deploying passive safety features), improved resource utilization, sustainable fuel cycle, and greater proliferation resistance than Generation III nuclear reactors. The Canadian SCWR core concept consists of a high-pressure inlet plenum, a separate low-pressure heavy water moderator contained in a calandria vessel, and 336 pressure tubes surrounded by the moderator. The reactor uses supercritical water as a coolant, and a direct steam power cycle to generate electricity. The reactor concept incorporates advanced safety features such as passive core cooling, long-term decay heat rejection to the environment and fuel melt prevention via passive moderator cooling. These features significantly reduce core damage frequency relative to existing nuclear reactors. This paper presents a description of the design concepts for the Canadian SCWR core, reactor building layout and the plant layout. Passive safety concepts are also described that address containment and core cooling following a loss-of coolant accident, as well as long term reactor heat removal at station blackout conditions. (author)

  5. Power Reactor Embrittlement Data Base

    International Nuclear Information System (INIS)

    Kam, F.B.K.; Stallmann, F.W.; Wang, J.A.

    1990-01-01

    Regulatory and research evaluations of embrittlement predication models and of pressure vessel integrity can be greatly expedited by the use of a well-designed, computerized data base. The Power Reactor Embrittlement Data Base (PR-EDB) is such a comprehensive collection of data for US commercial nuclear reactors. The Nuclear Regulatory Commission (NRC) has provided financial support, and the Electric Power Research Institute (EPRI) has provided technical assistance in the quality assurance (QA) of the data to establish an industry-wide data base that will be maintained and updated on a long-term basis. Successful applications of the data base to several of NRC's evaluations have received favorable response and support for its continuation. The future direction of the data base has been designed to include the test reactor and other types of data of interest to the regulators and the researchers. 1 ref

  6. Reactor container

    International Nuclear Information System (INIS)

    Oyamada, Osamu; Furukawa, Hideyasu; Uozumi, Hiroto.

    1979-01-01

    Purpose: To lower the position of an intermediate slab within a reactor container and fitting a heat insulating material to the inner wall of said intermediate slab, whereby a space for a control rod exchanging device and thermal stresses of the inner peripheral wall are lowered. Constitution: In the pedestal at the lower part of a reactor pressure vessel there is formed an intermediate slab at a position lower than diaphragm floor slab of the outer periphery of the pedestal thereby to secure a space for providing automatic exchanging device of a control rod driving device. Futhermore, a heat insulating material is fitted to the inner peripheral wall at the upper side of the intermediate slab part, and the temperature gradient in the wall thickness direction at the time of a piping rupture trouble is made gentle, and thermal stresses at the inner peripheral wall are lowered. (Sekiya, K.)

  7. Investigating effects of near fault rupture directivity on seismic hazard assessment of the site of Tehran Research Reactor (TRR)

    Energy Technology Data Exchange (ETDEWEB)

    Sepanloo, Kamran; Saberi, Reza [Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of); Alinejad, Majid [Atomic Energy Organization of Iran, Tehran (Iran, Islamic Republic of); Bazarchi, Ehsan [Tabriz Univ. (Iran, Islamic Republic of)

    2017-10-15

    It is estimated that the occurrence of a major-earthquake in Tehran, Iran, which is not far-fetched, would face the country with a huge amount of collapsed structures, economical losses and fatalities. The issue becomes more important while the site of interest is attributed to the nuclear facilities and any under-estimation in predicting the design ground motion may cause a real disaster. In this study, using calculations coded in MATLAB, PSHA was conducted for the site of TRR. It was concluded that most of the hazard for considered site in a 10000-year period comes from distances lower than 20 km and considering rupture directivity effects of the North Tehran fault, as the nearest seismicity source to considered site, using narrowband method affected the response spectrum significantly. Therefore, it is necessary to incorporate the near fault rupture directivity effects into the higher levels of seismic hazard assessment attributed to important sites.

  8. Effects of secondary electrons in traveling wave direct energy converter for D-{sup 3}He fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ishikawa, Motoo [Tsukuba Univ., Ibaraki (Japan); Shimizu, Yuuki [Kyoto Univ. (Japan); Momota, Hiromu; Tomita, Yukihiro [National Inst. for Fusion Science, Toki, Gifu (Japan)

    2000-06-01

    Nuclear fusion is expected as a main energy source in the future. D-{sup 3}He nuclear fusion has some better characteristics such as low neutron emission and capability of direct energy conversion than D-T or D-D nuclear fusion. The present paper studies effects of secondary electrons produced at grids of TWDEC (Traveling Wave Direct Energy Conversion) by the high-energy protons. It is found that the loss caused by the acceleration of electrons is about 1% of the input kinetic energy of protons, and effects of the electrical charge and the radiation are small. Therefore, all the effects are much smaller than the input power of protons. (author)

  9. Reactor Engineering Department annual report

    International Nuclear Information System (INIS)

    1993-09-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1992 (April 1, 1992-March 31, 1993). The major Department's programs promoted in the year are the assessment of the high conversion light water reactor, the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor physics facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project were also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

  10. Reactor engineering department annual report

    International Nuclear Information System (INIS)

    1990-09-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1989 (April 1, 1989 - March 31, 1990). One of major Department's programs is the assessment of the high conversion light water reactor and the design activities of advanced reactor system. Development of a high energy proton linear accelerator for the nuclear engineering including is also TRU incineration promoted. Other major tasks of the Department are various basic researches on nuclear data and group constants, theoretical methods and code development, on reactor physics experiments and analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics, technology assessment of nuclear energy and technology developments related to the reactor physics facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

  11. Cooperation in reactor design evaluation and licensing

    Energy Technology Data Exchange (ETDEWEB)

    Kaufer, B.; Wasylyk, A. [World Nuclear Association, London (United Kingdom)

    2014-07-01

    In January 2007 the World Nuclear Association (WNA) established the Cooperation in Reactor Design Evaluation and Licensing (CORDEL) Working Group with the aim of stimulating a dialogue between the nuclear industry (including reactor vendors, operators and utilities) and nuclear regulators (national and international organisations) on the benefits and means of achieving a worldwide convergence of reactor safety standards for reactor designs. From the time of its inception to the present, CORDEL has evolved from a group of experts discussing how to achieve international standardisation in nuclear safety design to an established and recognised working group dedicated to analysing and forging common understandings in key areas as input to major decisions on nuclear energy policy. This paper will review the general directions and activities CORDEL plans to undertake during the next five-year period, including its general strategy, activities, priorities and interactions with its customers in order to meet its objectives. (author)

  12. Nuclear reactor physics

    CERN Document Server

    Stacey, Weston M

    2010-01-01

    Nuclear reactor physics is the core discipline of nuclear engineering. Nuclear reactors now account for a significant portion of the electrical power generated worldwide, and new power reactors with improved fuel cycles are being developed. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. The second edition of this successful comprehensive textbook and reference on basic and advanced nuclear reactor physics has been completely updated, revised and enlarged to include the latest developme

  13. Persistence of mixed cryoglobulinemia despite cure of hepatitis C with new oral antiviral therapy including direct-acting antiviral sofosbuvir: A case series.

    Science.gov (United States)

    Cornella, Scott L; Stine, Jonathan G; Kelly, Virginia; Caldwell, Stephen H; Shah, Neeral L

    2015-05-01

    Obtaining a sustained virologic response (SVR) in patients with chronic hepatitis C virus (HCV) can decrease hepatic complications and be curative, however, extrahepatic manifestations including mixed cryoglobulinemia (MCN) may persist with interferon-based therapy. Our objective was to review our experience in treating patients with new oral antiviral agents and to assess common factors associated with MCN persistence despite SVR. We analyzed a case series of five patients with genotype one chronic HCV complicated by MCN who had persistence of cryoglobulins despite completion of triple therapy with oral antiviral agents (boceprivir, telaprivir or sofosbuvir). Patients with cirrhosis appear to have a decreased ability to clear immune complexes. We observed that early viral response by week 8 of therapy and longer periods of undetectable virus on treatment correlated with eventual clearance of serum cryoglobulins in patients without cirrhosis. Two patients were treated with anti-B-cell agent rituximab prior to starting therapy for HCV; this did not lead to a more effective clearance of cryoglobulins. We suggest that a longer treatment course than the standard 24 weeks with triple therapy could aid in the clearance of these immune complexes and cryoglobulins in cirrhotics. More studies to determine the ideal duration of treatment for chronic HCV and coincident MCN are needed, especially in light of the new all oral direct-acting antiviral regimens that are now recommended for HCV treatment.

  14. Applications of plasma core reactors to terrestrial energy systems

    Science.gov (United States)

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

    1974-01-01

    Plasma core reactors offer several new options for future energy needs in addition to space power and propulsion applications. Power extraction from plasma core reactors with gaseous nuclear fuel allows operation at temperatures higher than conventional reactors. Highly efficient thermodynamic cycles and applications employing direct coupling of radiant energy are possible. Conceptual configurations of plasma core reactors for terrestrial applications are described. Closed-cycle gas turbines, MHD systems, photo- and thermo-chemical hydrogen production processes, and laser systems using plasma core reactors as prime energy sources are considered. Cycle efficiencies in the range of 50 to 65 percent are calculated for closed-cycle gas turbine and MHD electrical generators. Reactor advantages include continuous fuel reprocessing which limits inventory of radioactive by-products and thorium-U-233 breeder configurations with about 5-year doubling times.-

  15. Autonomous Control of Space Nuclear Reactors

    Science.gov (United States)

    Merk, John

    2013-01-01

    Nuclear reactors to support future robotic and manned missions impose new and innovative technological requirements for their control and protection instrumentation. Long-duration surface missions necessitate reliable autonomous operation, and manned missions impose added requirements for failsafe reactor protection. There is a need for an advanced instrumentation and control system for space-nuclear reactors that addresses both aspects of autonomous operation and safety. The Reactor Instrumentation and Control System (RICS) consists of two functionally independent systems: the Reactor Protection System (RPS) and the Supervision and Control System (SCS). Through these two systems, the RICS both supervises and controls a nuclear reactor during normal operational states, as well as monitors the operation of the reactor and, upon sensing a system anomaly, automatically takes the appropriate actions to prevent an unsafe or potentially unsafe condition from occurring. The RPS encompasses all electrical and mechanical devices and circuitry, from sensors to actuation device output terminals. The SCS contains a comprehensive data acquisition system to measure continuously different groups of variables consisting of primary measurement elements, transmitters, or conditioning modules. These reactor control variables can be categorized into two groups: those directly related to the behavior of the core (known as nuclear variables) and those related to secondary systems (known as process variables). Reliable closed-loop reactor control is achieved by processing the acquired variables and actuating the appropriate device drivers to maintain the reactor in a safe operating state. The SCS must prevent a deviation from the reactor nominal conditions by managing limitation functions in order to avoid RPS actions. The RICS has four identical redundancies that comply with physical separation, electrical isolation, and functional independence. This architecture complies with the

  16. Nuclear reactor internals arrangement

    International Nuclear Information System (INIS)

    Frisch, E.; Andrews, H.N.

    1976-01-01

    A nuclear reactor internals arrangement is disclosed which facilitates reactor refueling. A reactor vessel and a nuclear core is utilized in conjunction with an upper core support arrangement having means for storing withdrawn control rods therein. The upper core support is mounted to the underside of the reactor vessel closure head so that upon withdrawal of the control rods into the upper core support, the closure head, the upper core support and the control rods are removed as a single unit thereby directly exposing the core for purposes of refueling

  17. Neutronic reactor

    International Nuclear Information System (INIS)

    Lewis, W.R.

    1978-01-01

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

  18. H Reactor

    Data.gov (United States)

    Federal Laboratory Consortium — The H Reactor was the first reactor to be built at Hanford after World War II.It became operational in October of 1949, and represented the fourth nuclear reactor on...

  19. Research reactor fuel - an update

    International Nuclear Information System (INIS)

    Finlay, M.R.; Ripley, M.I.

    2003-01-01

    In the two years since the last ANA conference there have been marked changes in the research reactor fuel scene. A new low-enriched uranium (LEU) fuel, 'monolithic' uranium molybdenum, has shown such promise in initial trials that it may be suitable to meet the objectives of the Joint Declaration signed by Presidents Bush and Putin to commit to converting all US and Russian research reactors to LEU by 2012. Development of more conventional aluminium dispersion UMo LEU fuel has continued in the meantime and is entering the final qualification stage of multiple full sized element irradiations. Despite this progress, the original 2005 timetable for UMo fuel qualification has slipped and research reactors, including the RRR, may not convert from silicide to UMo fuel before 2007. The operators of the Swedish R2 reactor have been forced to pursue the direct route of qualifying a UMo lead test assembly (LTA) in order to meet spent fuel disposal requirements of the Swedish law. The LTA has recently been fabricated and is expected to be loaded shortly into the R2 reactor. We present an update of our previous ANA paper and details of the qualification process for UMo fuel

  20. Spinning fluids reactor

    Science.gov (United States)

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

    2012-11-20

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

  1. One piece reactor removal

    International Nuclear Information System (INIS)

    Chia, Wei-Min; Wang, Song-Feng

    1993-01-01

    The strategy of Taiwan Research Reactor Renewal plan is to remove the old reactor block with One Piece Reactor Removal (OPRR) method for installing a new research reactor in original building. In this paper, the engineering design of each transportation works including the work method, the major equipments, the design policy and design criteria is described and discussed. In addition, to ensure the reactor block is safety transported for storage and to guarantee the integrity of reactor base mat is maintained for new reactor, operation safety is drawn special attention, particularly under seismic condition, to warrant safe operation of OPRR. ALARA principle and Below Regulatory Concern (BRC) practice were also incorporated in the planning to minimize the collective dose and the total amount of radioactive wastes. All these activities are introduced in this paper. (J.P.N.)

  2. Regulations for RA reactor operation

    International Nuclear Information System (INIS)

    1980-09-01

    Regulations for RA reactor operation are written in accordance with the legal regulations defined by the Law about radiation protection and related legal acts, as well as technical standards according to the IAEA recommendations. The contents of this book include: fundamental data about the reactor; legal regulations for reactor operation; organizational scheme for reactor operation; general and detailed instructions for operation, behaviour in the reactor building, performing experiments; operating rules for operation under steady state and accidental conditions [sr

  3. Generation III+ Reactor Portfolio

    International Nuclear Information System (INIS)

    2010-03-01

    While the power generation needs of utilities are unique and diverse, they are all faced with the double challenge of meeting growing electricity needs while curbing CO 2 emissions. To answer these diverse needs and help tackle this challenge, AREVA has developed several reactor models which are briefly described in this document: The EPR TM Reactor: designed on the basis of the Konvoi (Germany) and N4 (France) reactors, the EPRTM reactor is an evolutionary model designed to achieve best-in-class safety and operational performance levels. The ATMEA1 TM reactor: jointly designed by Mitsubishi Heavy Industries and AREVA through ATMEA, their common company. This reactor design benefits from the competencies and expertise of the two mother companies, which have commissioned close to 130 reactor units. The KERENA TM reactor: Designed on the basis of the most recent German BWR reactors (Gundremmingen) the KERENA TM reactor relies on proven technology while also including innovative, yet thoroughly tested, features. The optimal combination of active and passive safety systems for a boiling water reactor achieves a very low probability of severe accident

  4. A comparison of stresses in molar teeth restored with inlays and direct restorations, including polymerization shrinkage of composite resin and tooth loading during mastication.

    Science.gov (United States)

    Dejak, Beata; Młotkowski, Andrzej

    2015-03-01

    Polymerization shrinkage of composites is one of the main causes of leakage around dental restorations. Despite the large numbers of studies there is no consensus, what kind of teeth reconstruction--direct or indirect composite restorations are the most beneficial and the most durable. The aim was to compare equivalent stresses and contact adhesive stresses in molar teeth with class II MOD cavities, which were restored with inlays and direct restorations (taking into account polymerization shrinkage of composite resin) during simulated mastication. The study was conducted using the finite elements method with the application of contact elements. Three 3D models of first molars were created: model A was an intact tooth; model B--a tooth with a composite inlay, and model C--a tooth with a direct composite restoration. Polymerization linear shrinkage 0.7% of a direct composite restoration and resin luting cement was simulated (load 1). A computer simulation of mastication was performed (load 2). In these 2 situations, equivalent stresses according to the modified von Mises criterion (mvM) in the materials of mandibular first molar models with different restorations were calculated and compared. Contact stresses in the luting cement-tooth tissue adhesive interface around the restorations were also assessed and analyzed. Equivalent stresses in a tooth with a direct composite restoration (the entire volume of which was affected by polymerization shrinkage) were many times higher than in the tooth restored with a composite inlay (where shrinkage was present only in a thin layer of the luting cement). In dentin and enamel the stress values were 8-14 times higher, and were 13 times higher in the direct restoration than in the inlay. Likewise, contact stresses in the adhesive bond around the direct restoration were 6.5-7.7 times higher compared to an extraorally cured restoration. In the masticatory simulation, shear contact stresses in the adhesive bond around the direct

  5. Unconventional liquid metal cooled fast reactors

    International Nuclear Information System (INIS)

    Spinrad, B.I.; Rohach, A.F.; Razzaque, M.M.

    1989-06-01

    This report describes the rationale for, design of and analytical studies on an unconventional sodium-cooled power reactor, called the Trench Reactor. It derives its name from the long, narrow sodium pool in which the reactor is placed. Unconventional features include: pool shape; reactor shape (also long and narrow); reflector control; low power density; hot-leg primary pumping; absence of a cold sodium pool; large core boxes rather than a large number of subassemblies; large diameter metal fuel; vessel suspension from cables; and vessel cooling by natural circulation of building atmosphere (nitrogen) at all times. These features all seem feasible. They result in a system that is capable of at least a ten year reload interval and shows good safety through direct physical response to loss-of-heat-sink, loss-of-flow and limited-reactivity nuclear transients. 43 figs., 43 tabs

  6. Visual examination in nuclear reactor inservice inspection

    International Nuclear Information System (INIS)

    Kornvik, L.A.

    1985-01-01

    Visual examination is an important inspection method for nuclear reactors. New developments in TV technology give new possibilities for inspections that contribute to the economic and safe operation of nuclear power plants. As a supplier of nuclear boiling water reactors, ASEA-ATOM is constantly following-up its delivered plants by inspecting different parts of the reactor system after various periods of service. Inspections include both standard NDT-methods and different visual examination methods. TV inspection often offers advantages over other methods. Special developments such as the use of color TV, miniature TV cameras, development of stereo TV and the use of miniature remote controlled vehicles greatly enhance the usefulness and applicability of visual examination. This will for example, make it possible to make more adequate evaluation of indications and to take direct in-picture measurements. It will also give added possibilities to inspect reactor internals with respect to possible cracking or other defects

  7. Mirror fusion reactors

    International Nuclear Information System (INIS)

    Anon.

    1978-01-01

    Conceptual design studies were made of fusion reactors based on the three current mirror-confinement concepts: the standard mirror, the tandem mirror, and the field-reversed mirror. Recent studies of the standard mirror have emphasized its potential as a fusion-fission hybrid reactor, designed to produce fuel for fission reactors. We have designed a large commercial hybrid and a small pilot-plant hybrid based on standard mirror confinement. Tandem mirror designs include a commercial 1000-MWe fusion power plant and a nearer term tandem mirror hybrid. Field-reversed mirror designs include a multicell commercial reactor producing 75 MWe and a single-cell pilot plant

  8. Fusion Reactor Materials

    International Nuclear Information System (INIS)

    Decreton, M.

    2002-01-01

    The objective of SCK-CEN's programme on fusion reactor materials is to contribute to the knowledge on the radiation-induced behaviour of fusion reactor materials and components as well as to help the international community in building the scientific and technical basis needed for the construction of the future reactor. Ongoing projects include: the study of the mechanical and chemical (corrosion) behaviour of structural materials under neutron irradiation and water coolant environment; the investigation of the characteristics of irradiated first wall material such as beryllium; investigations on the management of materials resulting from the dismantling of fusion reactors including waste disposal. Progress and achievements in these areas in 2001 are discussed

  9. Mirror fusion reactors

    International Nuclear Information System (INIS)

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

    1978-01-01

    We have carried out conceptual design studies of fusion reactors based on the three current mirror confinement concepts: the standard mirror, the tandem mirror, and the field-reversed mirror. Recent studies of the standard mirror have emphasized its potential as a fusion-fission hybrid reactor, designed to produce fission fuel for fission reactors. We have designed a large commercial hybrid based on standard mirror confinement, and also a small pilot plant hybrid. Tandem mirror designs include a commercial 1000 MWe fusion power plant and a nearer term tandem mirror hybrid. Field-reversed mirror designs include a multicell commercial reactor producing 75 MWe and a single cell pilot plant

  10. Fusion reactor development: A review

    International Nuclear Information System (INIS)

    Anon.

    1983-01-01

    This paper is a review of the current prospects for fusion reactor development based upon the present status in plasma physics research, fusion technology development and reactor conceptual design for the tokamak magnetic confinement concept. Recent advances in tokamak plasma research and fusion technology development are summarized. The direction and conclusions of tokamak reactor conceptual design are discussed. The status of alternate magnetic confinement concept research is reviewed briefly. A feasible timetable for the development of fusion reactors is presented

  11. Power reactors in Member States. 1978 edition

    International Nuclear Information System (INIS)

    1978-01-01

    The computer-based reactor listing gives information on reactor core characteristics and plant systems for all power reactors in operation under construction and planned. The following two tables are included to give a general picture of the overall situation: Reactor types and net electrical power; Reactor units and net electrical power by country and cumulated by year

  12. Three-dimensional magnetotelluric inversion including topography using deformed hexahedral edge finite elements, direct solvers and data space Gauss-Newton, parallelized on SMP computers

    Science.gov (United States)

    Kordy, M. A.; Wannamaker, P. E.; Maris, V.; Cherkaev, E.; Hill, G. J.

    2014-12-01

    We have developed an algorithm for 3D simulation and inversion of magnetotelluric (MT) responses using deformable hexahedral finite elements that permits incorporation of topography. Direct solvers parallelized on symmetric multiprocessor (SMP), single-chassis workstations with large RAM are used for the forward solution, parameter jacobians, and model update. The forward simulator, jacobians calculations, as well as synthetic and real data inversion are presented. We use first-order edge elements to represent the secondary electric field (E), yielding accuracy O(h) for E and its curl (magnetic field). For very low frequency or small material admittivity, the E-field requires divergence correction. Using Hodge decomposition, correction may be applied after the forward solution is calculated. It allows accurate E-field solutions in dielectric air. The system matrix factorization is computed using the MUMPS library, which shows moderately good scalability through 12 processor cores but limited gains beyond that. The factored matrix is used to calculate the forward response as well as the jacobians of field and MT responses using the reciprocity theorem. Comparison with other codes demonstrates accuracy of our forward calculations. We consider a popular conductive/resistive double brick structure and several topographic models. In particular, the ability of finite elements to represent smooth topographic slopes permits accurate simulation of refraction of electromagnetic waves normal to the slopes at high frequencies. Run time tests indicate that for meshes as large as 150x150x60 elements, MT forward response and jacobians can be calculated in ~2.5 hours per frequency. For inversion, we implemented data space Gauss-Newton method, which offers reduction in memory requirement and a significant speedup of the parameter step versus model space approach. For dense matrix operations we use tiling approach of PLASMA library, which shows very good scalability. In synthetic

  13. Degradation and mineralization of Direct Blue 71 in a circulating upflow reactor by UV/TiO2 process and employing a new method in kinetic study

    International Nuclear Information System (INIS)

    Saien, J.; Soleymani, A.R.

    2007-01-01

    Direct Blue 71 (C 40 H 23 N 7 Na 4 O 13 S 4 ), an azo dye with a high worldwide consumption and providing toxic effluents, can be highly degraded using TiO 2 catalyst suspension and irradiation with a UV-C lamp in a circulating upflow photo-reactor with no dead zone. An initial concentration of 50 mg L -1 of dye, within the range of typical concentration in textile wastewaters, was used. The influence of catalyst concentration, pH and temperature were investigated. The results showed that degradation of this dye can be conducted in the both processes of only UV irradiation and UV/TiO 2 ; but with the aim of mineralization, the later process provides significantly better results. Accordingly, a degradation of more than 97% of dye was achieved by applying the optimal operational parameters with 40 mg L -1 of catalyst, natural pH and 45 deg. C, during 120 min irradiation. A removal of about 50% of COD could also be obtained at the same time. In kinetic investigations, the effect of catalyst particles' turbidity was taken into account and the rate of degradation of the dye, under mild conditions, was expressed as the sum of the rates of individual photolysis and photocatalysis process branches, with mainly influence of the bulk hydroxyl radicals

  14. Degradation and mineralization of Direct Blue 71 in a circulating upflow reactor by UV/TiO{sub 2} process and employing a new method in kinetic study

    Energy Technology Data Exchange (ETDEWEB)

    Saien, J. [Department of Applied Chemistry, University of Bu-Ali Sina, Hamedan 65174 (Iran, Islamic Republic of)]. E-mail: saien@basu.ac.ir; Soleymani, A.R. [Department of Applied Chemistry, University of Bu-Ali Sina, Hamedan 65174 (Iran, Islamic Republic of)

    2007-06-01

    Direct Blue 71 (C{sub 40}H{sub 23}N{sub 7}Na{sub 4}O{sub 13}S{sub 4}), an azo dye with a high worldwide consumption and providing toxic effluents, can be highly degraded using TiO{sub 2} catalyst suspension and irradiation with a UV-C lamp in a circulating upflow photo-reactor with no dead zone. An initial concentration of 50 mg L{sup -1} of dye, within the range of typical concentration in textile wastewaters, was used. The influence of catalyst concentration, pH and temperature were investigated. The results showed that degradation of this dye can be conducted in the both processes of only UV irradiation and UV/TiO{sub 2}; but with the aim of mineralization, the later process provides significantly better results. Accordingly, a degradation of more than 97% of dye was achieved by applying the optimal operational parameters with 40 mg L{sup -1} of catalyst, natural pH and 45 deg. C, during 120 min irradiation. A removal of about 50% of COD could also be obtained at the same time. In kinetic investigations, the effect of catalyst particles' turbidity was taken into account and the rate of degradation of the dye, under mild conditions, was expressed as the sum of the rates of individual photolysis and photocatalysis process branches, with mainly influence of the bulk hydroxyl radicals.

  15. Degradation and mineralization of Direct Blue 71 in a circulating upflow reactor by UV/TiO2 process and employing a new method in kinetic study.

    Science.gov (United States)

    Saien, J; Soleymani, A R

    2007-06-01

    Direct Blue 71 (C(40)H(23)N(7)Na(4)O(13)S(4)), an azo dye with a high worldwide consumption and providing toxic effluents, can be highly degraded using TiO(2) catalyst suspension and irradiation with a UV-C lamp in a circulating upflow photo-reactor with no dead zone. An initial concentration of 50 mgL(-1) of dye, within the range of typical concentration in textile wastewaters, was used. The influence of catalyst concentration, pH and temperature were investigated. The results showed that degradation of this dye can be conducted in the both processes of only UV irradiation and UV/TiO(2); but with the aim of mineralization, the later process provides significantly better results. Accordingly, a degradation of more than 97% of dye was achieved by applying the optimal operational parameters with 40 mgL(-1) of catalyst, natural pH and 45 degrees C, during 120 min irradiation. A removal of about 50% of COD could also be obtained at the same time. In kinetic investigations, the effect of catalyst particles' turbidity was taken into account and the rate of degradation of the dye, under mild conditions, was expressed as the sum of the rates of individual photolysis and photocatalysis process branches, with mainly influence of the bulk hydroxyl radicals.

  16. Mimic of OSU research reactor

    International Nuclear Information System (INIS)

    Lu, Hong; Miller, D.W.

    1991-01-01

    The Ohio State University research reactor (OSURR) is undergoing improvements in its research and educational capabilities. A computer-based digital data acquisition system, including a reactor system mimic, will be installed as part of these improvements. The system will monitor the reactor system parameters available to the reactor operator either in digital parameters available to the reactor operator either in digital or analog form. The system includes two computers. All the signals are sent to computer 1, which processes the data and sends the data through a serial port to computer 2 with a video graphics array VGA monitor, which is utilized to display the mimic system of the reactor

  17. The replacement research reactor

    International Nuclear Information System (INIS)

    Cameron, R.

    1999-01-01

    As a consequences of the government decision in September 1997. ANSTO established a replacement research reactor project to manage the procurement of the replacement reactor through the necessary approval, tendering and contract management stages This paper provides an update of the status of the project including the completion of the Environmental Impact Statement. Prequalification and Public Works Committee processes. The aims of the project, management organisation, reactor type and expected capabilities are also described

  18. Fuel leak detection apparatus for gas cooled nuclear reactors

    Science.gov (United States)

    Burnette, Richard D.

    1977-01-01

    Apparatus is disclosed for detecting nuclear fuel leaks within nuclear power system reactors, such as high temperature gas cooled reactors. The apparatus includes a probe assembly that is inserted into the high temperature reactor coolant gaseous stream. The probe has an aperture adapted to communicate gaseous fluid between its inside and outside surfaces and also contains an inner tube for sampling gaseous fluid present near the aperture. A high pressure supply of noncontaminated gas is provided to selectively balance the pressure of the stream being sampled to prevent gas from entering the probe through the aperture. The apparatus includes valves that are operable to cause various directional flows and pressures, which valves are located outside of the reactor walls to permit maintenance work and the like to be performed without shutting down the reactor.

  19. Design Considerations for Economically Competitive Sodium Cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hongbin Zhang; Haihua Zhao

    2009-05-01

    The technological viability of sodium cooled fast reactors (SFR) has been established by various experimental and prototype (demonstration) reactors such as EBR-II, FFTF, Phénix, JOYO, BN-600 etc. However, the economic competitiveness of SFR has not been proven yet. The perceived high cost premium of SFRs over LWRs has been the primary impediment to the commercial expansion of SFR technologies. In this paper, cost reduction options are discussed for advanced SFR designs. These include a hybrid loop-pool design to optimize the primary system, multiple reheat and intercooling helium Brayton cycle for the power conversion system and the potential for suppression of intermediate heat transport system. The design options for the fully passive decay heat removal systems are also thoroughly examined. These include direct reactor auxiliary cooling system (DRACS), reactor vessel auxiliary cooling system (RVACS) and the newly proposed pool reactor auxiliary cooling system (PRACS) in the context of the hybrid loop-pool design.

  20. Design of generic coal conversion facilities: Process release---Direct coal liquefaction

    Energy Technology Data Exchange (ETDEWEB)

    1991-09-01

    The direct liquefaction portion of the PETC generic direct coal liquefaction process development unit (PDU) is being designed to provide maximum operating flexibility. The PDU design will permit catalytic and non-catalytic liquefaction concepts to be investigated at their proof-of-the-concept stages before any larger scale operations are attempted. The principal variations from concept to concept are reactor configurations and types. These include thermal reactor, ebullating bed reactor, slurry phase reactor and fixed bed reactor, as well as different types of catalyst. All of these operating modes are necessary to define and identify the optimum process conditions and configurations for determining improved economical liquefaction technology.

  1. RB research nuclear reactor - Annual report for 1985, I - III

    International Nuclear Information System (INIS)

    Markovic, H.; Pesic, M.; Vranic, S.; Petronijevic, M.; Jevremovic, M.; Ilic, I.

    1985-01-01

    This paper describes the operation of the RB reactor during 1985. The report includes 3 parts: Engineering description and operation of the RB reactor, reactor components and maintenance, and RB reactor utilization in 1985 [sr

  2. Research reactor modernization and refurbishment

    International Nuclear Information System (INIS)

    2009-08-01

    Many recent, high profile research reactor unplanned shutdowns can be directly linked to different challenges which have evolved over time. The concept of ageing management is certainly nothing new to nuclear facilities, however, these events are highlighting the direct impact unplanned shutdowns at research reactors have on various stakeholders who depend on research reactor goods and services. Provided the demand for these goods and services remains strong, large capital projects are anticipated to continue in order to sustain future operation of many research reactors. It is within this context that the IAEA organized a Technical Workshop to launch a broader Agency activity on research reactor modernization and refurbishment (M and R). The workshop was hosted by the operating organization of the HOR Research Reactor in Delft, the Netherlands, in October 2006. Forty participants from twenty-three countries participated in the meeting: with representation from Africa, Asia Pacific, Eastern Europe, North America, South America and Western Europe. The specific objectives of this workshop were to present facility reports on completed, existing and planned M and R projects, including the project objectives, scope and main characteristics; and to specifically report on: - the project impact (planned or actual) on the primary and key supporting motivation for the M and R project; - the project impact (planned or actual) on the design basis, safety, and/or regulatory-related reports; - the project impact (planned or actual) on facility utilization; - significant lessons learned during or following the completion of M and R work. Contributions from this workshop were reviewed by experts during a consultancy meeting held in Vienna in December 2007. The experts selected final contributions for inclusion in this report. Requests were also distributed to some authors for additional detail as well as new authors for known projects not submitted during the initial 2006 workshop

  3. Reactor operation safety information document

    Energy Technology Data Exchange (ETDEWEB)

    1990-01-01

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

  4. Fast Spectrum Reactors

    CERN Document Server

    Todd, Donald; Tsvetkov, Pavel

    2012-01-01

    Fast Spectrum Reactors presents a detailed overview of world-wide technology contributing to the development of fast spectrum reactors. With a unique focus on the capabilities of fast spectrum reactors to address nuclear waste transmutation issues, in addition to the well-known capabilities of breeding new fuel, this volume describes how fast spectrum reactors contribute to the wide application of nuclear power systems to serve the global nuclear renaissance while minimizing nuclear proliferation concerns. Readers will find an introduction to the sustainable development of nuclear energy and the role of fast reactors, in addition to an economic analysis of nuclear reactors. A section devoted to neutronics offers the current trends in nuclear design, such as performance parameters and the optimization of advanced power systems. The latest findings on fuel management, partitioning and transmutation include the physics, efficiency and strategies of transmutation, homogeneous and heterogeneous recycling, in addit...

  5. Evaluation of Alternate Materials for Coated Particle Fuels for the Gas-Cooled Fast Reactor. Laboratory Directed Research and Development Program FY 2006 Final Report

    International Nuclear Information System (INIS)

    Paul A. Demkowicz; Karen Wright; Jian Gan; David Petti; Todd Allen; Jake Blanchard

    2006-01-01

    Candidate ceramic materials were studied to determine their suitability as Gas-Cooled Fast Reactor particle fuel coatings. The ceramics examined in this work were: TiC, TiN, ZrC, ZrN, AlN, and SiC. The studies focused on (1) chemical reactivity of the ceramics with fission products palladium and rhodium, (2) the thermomechanical stresses that develop in the fuel coatings from a variety of causes during burnup, and (3) the radiation resiliency of the materials. The chemical reactivity of TiC, TiN, ZrC, and ZrN with Pd and Rh were all found to be much lower than that of SiC. A number of important chemical behaviors were observed at the ceramic-metal interfaces, including the formation of specific intermetallic phases and a variation in reaction rates for the different ceramics investigated. Based on the data collected in this work, the nitride ceramics (TiN and ZrN) exhibit chemical behavior that is characterized by lower reaction rates with Pd and Rh than the carbides TiC and ZrC. The thermomechanical stresses in spherical fuel particle ceramic coatings were modeled using finite element analysis, and included contributions from differential thermal expansion, fission gas pressure, fuel kernel swelling, and thermal creep. In general the tangential stresses in the coatings during full reactor operation are tensile, with ZrC showing the lowest values among TiC, ZrC, and SiC (TiN and ZrN were excluded from the comprehensive calculations due to a lack of available materials data). The work has highlighted the fact that thermal creep plays a critical role in the development of the stress state of the coatings by relaxing many of the stresses at high temperatures. To perform ion irradiations of sample materials, an irradiation beamline and high-temperature sample irradiation stage was constructed at the University of Wisconsin's 1.7MV Tandem Accelerator Facility. This facility is now capable of irradiating of materials to high dose while controlling sample temperature up

  6. Evaluation of Alternate Materials for Coated Particle Fuels for the Gas-Cooled Fast Reactor. Laboratory Directed Research and Development Program FY 2006 Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Paul A. Demkowicz; Karen Wright; Jian Gan; David Petti; Todd Allen; Jake Blanchard

    2006-09-01

    Candidate ceramic materials were studied to determine their suitability as Gas-Cooled Fast Reactor particle fuel coatings. The ceramics examined in this work were: TiC, TiN, ZrC, ZrN, AlN, and SiC. The studies focused on (i) chemical reactivity of the ceramics with fission products palladium and rhodium, (ii) the thermomechanical stresses that develop in the fuel coatings from a variety of causes during burnup, and (iii) the radiation resiliency of the materials. The chemical reactivity of TiC, TiN, ZrC, and ZrN with Pd and Rh were all found to be much lower than that of SiC. A number of important chemical behaviors were observed at the ceramic-metal interfaces, including the formation of specific intermetallic phases and a variation in reaction rates for the different ceramics investigated. Based on the data collected in this work, the nitride ceramics (TiN and ZrN) exhibit chemical behavior that is characterized by lower reaction rates with Pd and Rh than the carbides TiC and ZrC. The thermomechanical stresses in spherical fuel particle ceramic coatings were modeled using finite element analysis, and included contributions from differential thermal expansion, fission gas pressure, fuel kernel swelling, and thermal creep. In general the tangential stresses in the coatings during full reactor operation are tensile, with ZrC showing the lowest values among TiC, ZrC, and SiC (TiN and ZrN were excluded from the comprehensive calculations due to a lack of available materials data). The work has highlighted the fact that thermal creep plays a critical role in the development of the stress state of the coatings by relaxing many of the stresses at high temperatures. To perform ion irradiations of sample materials, an irradiation beamline and high-temperature sample irradiation stage was constructed at the University of Wisconsin’s 1.7MV Tandem Accelerator Facility. This facility is now capable of irradiating of materials to high dose while controlling sample temperature

  7. Nuclear reactor safety systems

    International Nuclear Information System (INIS)

    Ball, R.M.; Roberts, R.C.

    1980-01-01

    A safety system for shutting down a nuclear reactor under overload conditions is described. The system includes a series of parallel-connected computer memory type look-up tables each of which receives data on a particular reactor parameter and in each of which a precalculated functional value for that parameter is stored indicative of the percentage of maximum reactor load that the parameter contributes. The various functional values corresponding to the actual measured parameters are added together to provide a control signal used to shut down the reactor under overload conditions. (U.K.)

  8. Future fuel cycle and reactor strategies

    International Nuclear Information System (INIS)

    Meneley, D.A.

    1999-01-01

    Within the framework of the 1997 IAEA Symposium 'Future Fuel Cycle and Reactor Strategies Adjusting to New Realities', Working Group No.3 produced a Key Issues paper addressing the title of the symposium. The scope of the Key Issues paper included those factors that are expected to remain or become important in the time period from 2015 to 2050, considering all facets of nuclear energy utilization from ore extraction to final disposal of waste products. The paper addressed the factors influencing the choice of reactor and fuel cycle. It then addressed the quantitatively largest category of reactor types expected to be important during the period; that is, thermal reactors burning uranium and plutonium fuel. The fast reactor then was discussed both as a stand-alone technology and as might be used in combination with thermal reactors. Thorium fuel use was discussed briefly. The present paper includes of a digest of the Key Issues Paper. Some comparisons arc made between the directions suggested in that paper and those indicated by the Abstracts of this Technical Committee Meeting- Recommendations are made for work which might be undertaken in the short and medium time frames, to ensure that fuel cycle technologies and processes established by the year 2050 will support the continuation of nuclear energy applications in the long term. (author)

  9. Advanced Demonstration and Test Reactor Options Study

    Energy Technology Data Exchange (ETDEWEB)

    Petti, David Andrew [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Gehin, J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (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); Heidet, F. [Argonne National Lab. (ANL), Argonne, IL (United States); Kinsey, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Grandy, Christopher [Argonne National Lab. (ANL), Argonne, IL (United States); Qualls, A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hoffman, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Croson, D. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-01-01

    Global efforts to address climate change will require large-scale decarbonization of energy production in the United States and elsewhere. Nuclear power already provides 20% of electricity production in the United States (U.S.) and is increasing in countries undergoing rapid growth around the world. Because reliable, grid-stabilizing, low emission electricity generation, energy security, and energy resource diversity will be increasingly valued, nuclear power’s share of electricity production has a potential to grow. In addition, there are non electricity applications (e.g., process heat, desalination, hydrogen production) that could be better served by advanced nuclear systems. Thus, the timely development, demonstration, and commercialization of advanced nuclear reactors could diversify the nuclear technologies available and offer attractive technology options to expand the impact of nuclear energy for electricity generation and non-electricity missions. The purpose of this planning study is to provide transparent and defensible technology options for a test and/or demonstration reactor(s) to be built to support public policy, innovation and long term commercialization within the context of the Department of Energy’s (DOE’s) broader commitment to pursuing an “all of the above” clean energy strategy and associated time lines. This planning study includes identification of the key features and timing needed for advanced test or demonstration reactors to support research, development, and technology demonstration leading to the commercialization of power plants built upon these advanced reactor platforms. This planning study is consistent with the Congressional language contained within the fiscal year 2015 appropriation that directed the DOE to conduct a planning study to evaluate “advanced reactor technology options, capabilities, and requirements within the context of national needs and public policy to support innovation in nuclear energy

  10. 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.

  11. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2001-01-01

    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

  12. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2002-04-01

    SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. 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 materials testing 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 2001 are summarised.

  13. Annual progress report FY 1977. [Computer calculations of light water reactor dynamics and safety

    Energy Technology Data Exchange (ETDEWEB)

    Hansen, K.F.; Henry, A.F.

    1977-07-01

    Progress is summarized in a project directed toward development of numerical methods suitable for the computer solution of problems in reactor dynamics and safety. Specific areas of research include methods of integration of the time-dependent diffusion equations by finite difference and finite element methods; representation of reactor properties by various homogenization procedures; application of synthesis methods; and development of response matrix techniques.

  14. Photocatalytic reactor

    Science.gov (United States)

    Bischoff, Brian L.; Fain, Douglas E.; Stockdale, John A. D.

    1999-01-01

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

  15. RB research reactor Safety Report

    International Nuclear Information System (INIS)

    Sotic, O.; Pesic, M.; Vranic, S.

    1979-04-01

    This RB reactor safety report is a revised and improved version of the Safety report written in 1962. It contains descriptions of: reactor building, reactor hall, control room, laboratories, reactor components, reactor control system, heavy water loop, neutron source, safety system, dosimetry system, alarm system, neutron converter, experimental channels. Safety aspects of the reactor operation include analyses of accident causes, errors during operation, measures for preventing uncontrolled activity changes, analysis of the maximum possible accident in case of different core configurations with natural uranium, slightly and highly enriched fuel; influence of possible seismic events

  16. The fusion reactor

    International Nuclear Information System (INIS)

    Brennan, M.H.

    1974-01-01

    Basic principles of the fusion reactor are outlined. Plasma heating and confinement schemes are described. These confinement systems include the linear Z pinch, magnetic mirrors and Tokamaks. A fusion reactor is described and a discussion is given of its environmental impact and its fuel situation. (R.L.)

  17. A feasibility study of a linear laser heated solenoid fusion reactor. Final report

    International Nuclear Information System (INIS)

    Steinhauer, L.C.

    1976-02-01

    This report examines the feasibility of a laser heated solenoid as a fusion or fusion-fission reactor system. The objective of this study, was an assessment of the laser heated solenoid reactor concept in terms of its plasma physics, engineering design, and commercial feasibility. Within the study many pertinent reactor aspects were treated including: physics of the laser-plasma interaction; thermonuclear behavior of a slender plasma column; end-losses under reactor conditions; design of a modular first wall, a hybrid (both superconducting and normal) magnet, a large CO 2 laser system; reactor blanket; electrical storage elements; neutronics; radiation damage, and tritium processing. Self-consistent reactor configurations were developed for both pure fusion and fusion-fission designs, with the latter designed both to produce power and/or fissile fuels for conventional fission reactors. Appendix A is a bibliography with commentary of theoretical and experimental studies that have been directed at the laser heated solenoid

  18. Study of direct Cp violation in B decay into vector mesons including rho zero-omega mixing in the framework of the LHCb experiment

    International Nuclear Information System (INIS)

    Rimbault, C.

    2004-02-01

    ρ 0 - ω mixing effect on direct CP violation in B decay into vector mesons is one of the main studies of this work. The first part is dedicated to the calculation of the decay amplitudes of the channels B → Vρ 0 (ω) which have been represented by a model. We have used the helicity formalism. In such a way branching ratios and asymmetries depending on form factor models and other parameters are predicted. Direct CP violation appears at several levels: in branching ratios, in angular distributions and in differential asymmetry as a function of ρ 0 - ω mass. The dominance of the longitudinal polarization in the studied channels is confirmed by Babar and Belle experimental results. We calculated too the strong phase and the ratio of Penguin to Tree amplitudes for each channel. In a second part, was developed an analysis of the channel B 0 → K *0 ρ 0 (ω) in the framework of LHCb experiment. It will start in 2007 and is dedicated to b flavor study and CP violation. The realistic analysis which has been performed shows that this channel is not appropriate to observe ρ 0 - ω mixing effect on asymmetry in LHCb, while the ρ + ρ 0 (ω) channel, for which we have predicted a branching ratio value confirmed by Babar and Belle, is much more promising. (author)

  19. 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

  20. Directly catalytic upgrading bio-oil vapor produced by prairie cordgrass pyrolysis over Ni/HZSM-5 using a two stage reactor

    Directory of Open Access Journals (Sweden)

    Shouyun Cheng

    2015-06-01

    Full Text Available Catalytic cracking is one of the most promising processes for thermochemical conversion of biomass to advanced biofuels in recent years. However, current effectiveness of catalysts and conversion efficiency still remain challenges. An investigation of directly catalytic upgrading bio-oil vapors produced in prairie cordgrass (PCG pyrolysis over Ni/HZSM-5 and HZSM-5 in a two stage packed-bed reactor was carried out. The Ni/HZSM-5 catalyst was synthesized using an impregnation method. Fresh and used catalysts were characterized by BET and XRD. The effects of catalysts on pyrolysis products yields and quality were examined. Both catalysts improved bio-oil product distribution compared to non-catalytic treatment. When PCG pyrolysis vapor was treated with absence of catalyst, the produced bio-oils contained higher alcohols (10.97% and furans (10.14%. In contrast, the bio-oils contained the second highest hydrocarbons (34.97%)and the highest phenols (46.97% when PCG pyrolysis vapor was treated with Ni/HZSM-5. Bio-oils containing less ketones and aldehydes were produced by both Ni/HZSM-5 and HZSM-5, but no ketones were found in Ni/HZSM-5 treatment compared to HZSM-5 (2.94%. The pyrolysis gas compositions were also affected by the presenting of HZSM-5 or Ni/HZSM-5 during the catalytic upgrading process. However, higher heating values and elemental compositions (C, H and N of bio-chars produced in all treatments had no significant difference.

  1. IVA2 - a computer code for modelling of transient 3D-three phase three component flows using three velocity fields in cylindrical geometry with arbitrary internals including nuclear reactor PWR/BWR-core

    International Nuclear Information System (INIS)

    Kolev, N.I.

    1986-06-01

    This report contains a formal code description (description of the input data, contents of the COMMON blocks, functions of the IVA2/001 routines). In addition the nonformal description of the current IVA2/001 constitutive package and the reactor core model are given. (orig.) [de

  2. Nuclear reactor

    International Nuclear Information System (INIS)

    Schweiger, F.; Glahe, E.

    1976-01-01

    In a nuclear reactor of the kind which is charged with spherical reaction elements and in which control rods are arranged to be thrust directly into the charge, each control rod has at least one screw thread on its external surface so that as the rod is thrust into the charge it is caused to rotate and thus make penetration easier. The length of each control rod may have two distinct portions, a latter portion which carries a screw thread and a lead-in portion which is shorter than the latter portion and which may carry a thread of greater pitch than that on the latter portion or may have a number of axially extending ribs instead of a thread

  3. Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Martens, Frederick H. [Argonne National Laboratory; Jacobson, Norman H.

    1968-09-01

    This booklet discusses research reactors - reactors designed to provide a source of neutrons and/or gamma radiation for research, or to aid in the investigation of the effects of radiation on any type of material.

  4. Nuclear reactors

    International Nuclear Information System (INIS)

    Middleton, J.E.

    1977-01-01

    Reference is made to water cooled reactors and in particular to the cooling system of steam generating heavy water reactors (SGHWR). A two-coolant circuit is described for the latter. Full constructural details are given. (U.K.)

  5. Mirror reactor surface study

    Energy Technology Data Exchange (ETDEWEB)

    Hunt, A. L.; Damm, C. C.; Futch, A. H.; Hiskes, J. R.; Meisenheimer, R. G.; Moir, R. W.; Simonen, T. C.; Stallard, B. W.; Taylor, C. E.

    1976-09-01

    A general survey is presented of surface-related phenomena associated with the following mirror reactor elements: plasma first wall, ion sources, neutral beams, director converters, vacuum systems, and plasma diagnostics. A discussion of surface phenomena in possible abnormal reactor operation is included. Several studies which appear to merit immediate attention and which are essential to the development of mirror reactors are abstracted from the list of recommended areas for surface work. The appendix contains a discussion of the fundamentals of particle/surface interactions. The interactions surveyed are backscattering, thermal desorption, sputtering, diffusion, particle ranges in solids, and surface spectroscopic methods. A bibliography lists references in a number of categories pertinent to mirror reactors. Several complete published and unpublished reports on surface aspects of current mirror plasma experiments and reactor developments are also included.

  6. Mirror reactor surface study

    International Nuclear Information System (INIS)

    Hunt, A.L.; Damm, C.C.; Futch, A.H.; Hiskes, J.R.; Meisenheimer, R.G.; Moir, R.W.; Simonen, T.C.; Stallard, B.W.; Taylor, C.E.

    1976-01-01

    A general survey is presented of surface-related phenomena associated with the following mirror reactor elements: plasma first wall, ion sources, neutral beams, director converters, vacuum systems, and plasma diagnostics. A discussion of surface phenomena in possible abnormal reactor operation is included. Several studies which appear to merit immediate attention and which are essential to the development of mirror reactors are abstracted from the list of recommended areas for surface work. The appendix contains a discussion of the fundamentals of particle/surface interactions. The interactions surveyed are backscattering, thermal desorption, sputtering, diffusion, particle ranges in solids, and surface spectroscopic methods. A bibliography lists references in a number of categories pertinent to mirror reactors. Several complete published and unpublished reports on surface aspects of current mirror plasma experiments and reactor developments are also included

  7. Fast Reactor Knowledge Management at IGCAR, India

    International Nuclear Information System (INIS)

    Kuriakose, K.K.

    2013-01-01

    The Process Architecture: → Acquire: Solicitation; Voluntary submission; Mandatory requirements; Interview/Observation; → Quality Control: Review/Editing; Certification; Quality index; → Disseminate: Publish through the Technology architecture; Formal/Informal Meetings; COPs; → Utilize: Projects; Day-to-day activities; → Maintenance; → Retirement. Mission: To conduct a broad based multidisciplinary programme of scientific research and advanced engineering development, directed towards the establishment of the technology of Sodium Cooled Fast Breeder Reactors (FBR) and associated fuel cycle facilities in the Country. The mission includes the development and applications of new and improved materials, techniques, equipment and systems for FBRs, pursue basic research to achieve breakthroughs in Fast Reactor technology

  8. Fuel handling system of nuclear reactor plants

    International Nuclear Information System (INIS)

    Faulstich, D.L.

    1991-01-01

    This patent describes a fuel handing system for nuclear reactor plants comprising a reactor vessel having an openable top and removable cover for refueling and containing therein, submerged in coolant water substantially filling the reactor vessel, a fuel core including a multiplicity of fuel bundles formed of groups of sealed tube elements enclosing fissionable fuel assembled into units. It comprises a fuel bundle handing platform moveable over the open top of the reactor vessel; a fuel bundle handing mast extendable downward from the platform with a lower end projecting into the open top reactor vessel to the fuel core submerged in water; a grapple head mounted on the lower end of the mast provided with grappling hook means for attaching to and transporting fuel bundles into and out from the fuel core; and a camera with a prismatic viewing head surrounded by a radioactive resisting quartz cylinder and enclosed within the grapple head which is provided with at least three windows with at least two windows provided with an angled surface for aiming the camera prismatic viewing head in different directions and thereby viewing the fuel bundles of the fuel core from different perspectives, and having a cable connecting the camera with a viewing monitor located above the reactor vessel for observing the fuel bundles of the fuel core and for enabling aiming of the camera prismatic viewing head through the windows by an operator

  9. Safe operation of research reactors and critical assemblies. Code of practice and annexes. 1984 ed

    International Nuclear Information System (INIS)

    1984-01-01

    The safe operation of research reactors and critical assemblies (hereafter termed 'reactors') requires proper design, construction, management and supervision. This Code of Practice deals mainly with management and supervision. The provisions of the Code apply to the whole life of the reactor, including modification, updating and upgrading. The Code may be subject to revision in the light of experience and the state of technology. The Code is aimed at defining minimum requirements for the safe operation of reactors. Emphasis is placed on which safety requirements should be met rather than on specifying how these requirements may be met. The Code also provides guidance and information to persons and authorities responsible for the operation of reactors. The Code recommends that documents dealing with the operation of reactors and including safety analyses be prepared and submitted for review and approval to a regulatory body. Operation would be authorized on the understanding that it would comply with limits and conditions designed to ensure safety. The Code covers a wide range of reactor types, which gives rise to a variety of safety issues. Safety issues applicable to specific reactor types only (e.g. fast reactors) are not necessarily covered in this Code. Some of the recommendations in the Code are not directly applicable to critical assemblies. A recommendation may therefore be interpreted according to the type of reactor concerned. In such cases the words 'adequate' and 'appropriate' are used to mean 'adequate' or 'appropriate' for the type of reactor under consideration.

  10. Reactor Neutrinos

    OpenAIRE

    Kim, Soo-Bong; Lasserre, Thierry; Wang, Yifang

    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 ...

  11. Safety and licensing for small and medium power reactors

    International Nuclear Information System (INIS)

    Trauger, D.B.

    1987-01-01

    Proposed new concepts for small and medium power reactors differ substantially from traditional Light Water Reactors (LWRs). Although designers have a large base of experience in safety and licensing, much of it is not relevant to new concepts. It can be a disadvantage if regulators apply LWR rules directly. A fresh start is appropriate. The extensive interactions between industry, regulators, and the public complicates but may enhance safety. It is basic to recognize the features that distinguish nuclear energy safety from that for other industries. These features include: nuclear reactivity, fission product radiation, and radioactive decay heat. Small and medium power reactors offer potential advantages over LWRs, particularly for reactivity and decay heat

  12. The organization of research reactor safety in the UKAEA

    International Nuclear Information System (INIS)

    Redpath, W.

    1983-01-01

    The present state of organization and development of research reactor safety in the UKAEA are outlined by addressing the fundamental safety principles which have been adopted in keeping with national health and safety requirement. The organisation, assessment and monitoring of research reactor safety on complex multi-discipline and multi-activity nuclear research and development site are discussed. Methods of safety assessment, such as probabilistic risk assessment and risk acceptance criteria, which have been developed and applied in practice are explained, and some indication of the directions in which some of the current developments in the safety of UKAEA research reactors is also included. (A.J.)

  13. Endoscopy in patients on antiplatelet or anticoagulant therapy, including direct oral anticoagulants: British Society of Gastroenterology (BSG) and European Society of Gastrointestinal Endoscopy (ESGE) guidelines

    Science.gov (United States)

    Veitch, Andrew M; Vanbiervliet, Geoffroy; Gershlick, Anthony H; Boustiere, Christian; Baglin, Trevor P; Smith, Lesley-Ann; Radaelli, Franco; Knight, Evelyn; Gralnek, Ian M; Hassan, Cesare; Dumonceau, Jean-Marc

    2016-01-01

    The risk of endoscopy in patients on antithrombotics depends on the risks of procedural haemorrhage versus thrombosis due to discontinuation of therapy. P2Y12 receptor antagonists (clopidogrel, prasugrel, ticagrelor) For low-risk endoscopic procedures we recommend continuing P2Y12 receptor antagonists as single or dual antiplatelet therapy (low quality evidence, strong recommendation); For high-risk endoscopic procedures in patients at low thrombotic risk, we recommend discontinuing P2Y12 receptor antagonists five days before the procedure (moderate quality evidence, strong recommendation). In patients on dual antiplatelet therapy, we suggest continuing aspirin (low quality evidence, weak recommendation). For high-risk endoscopic procedures in patients at high thrombotic risk, we recommend continuing aspirin and liaising with a cardiologist about the risk/benefit of discontinuation of P2Y12 receptor antagonists (high quality evidence, strong recommendation). Warfarin The advice for warfarin is fundamentally unchanged from British Society of Gastroenterology (BSG) 2008 guidance. Direct Oral Anticoagulants (DOAC) For low-risk endoscopic procedures we suggest omitting the morning dose of DOAC on the day of the procedure (very low quality evidence, weak recommendation); For high-risk endoscopic procedures, we recommend that the last dose of DOAC be taken ≥48 h before the procedure (very low quality evidence, strong recommendation). For patients on dabigatran with CrCl (or estimated glomerular filtration rate, eGFR) of 30–50 mL/min we recommend that the last dose of DOAC be taken 72 h before the procedure (very low quality evidence, strong recommendation). In any patient with rapidly deteriorating renal function a haematologist should be consulted (low quality evidence, strong recommendation). PMID:26873868

  14. Endoscopy in patients on antiplatelet or anticoagulant therapy, including direct oral anticoagulants: British Society of Gastroenterology (BSG) and European Society of Gastrointestinal Endoscopy (ESGE) guidelines.

    Science.gov (United States)

    Veitch, Andrew M; Vanbiervliet, Geoffroy; Gershlick, Anthony H; Boustiere, Christian; Baglin, Trevor P; Smith, Lesley-Ann; Radaelli, Franco; Knight, Evelyn; Gralnek, Ian M; Hassan, Cesare; Dumonceau, Jean-Marc

    2016-04-01

    The risk of endoscopy in patients on antithrombotics depends on the risks of procedural haemorrhage vs. thrombosis due to discontinuation of therapy. P2Y12 receptor antagonists (clopidogrel, prasugrel, ticagrelor): For low-risk endoscopic procedures we recommend continuing P2Y12 receptor antagonists as single or dual antiplatelet therapy (low quality evidence, strong recommendation);For high-risk endoscopic procedures in patients at low thrombotic risk, we recommend discontinuing P2Y12 receptor antagonists five days before the procedure (moderate quality evidence, strong recommendation). In patients on dual antiplatelet therapy, we suggest continuing aspirin (low quality evidence, weak recommendation).For high-risk endoscopic procedures in patients at high thrombotic risk, we recommend continuing aspirin and liaising with a cardiologist about the risk/benefit of discontinuation of P2Y12 receptor antagonists (high quality evidence, strong recommendation). Warfarin: The advice for warfarin is fundamentally unchanged from BSG 2008 guidance. Direct Oral Anticoagulants (DOAC): For low-risk endoscopic procedures we suggest omitting the morning dose of DOAC on the day of the procedure (very low quality evidence, weak recommendation). For high-risk endoscopic procedures, we recommend that the last dose of DOAC be taken ≥ 48 hours before the procedure (very low quality evidence, strong recommendation). For patients on dabigatran with CrCl (or estimated glomerular filtration rate, eGFR) of 30 - 50 mL/min we recommend that the last dose of DOAC be taken 72 hours before the procedure (very low quality evidence, strong recommendation). In any patient with rapidly deteriorating renal function a haematologist should be consulted (low quality evidence, strong recommendation). © Georg Thieme Verlag KG Stuttgart · New York.

  15. EspC, an Autotransporter Protein Secreted by Enteropathogenic Escherichia coli, Causes Apoptosis and Necrosis through Caspase and Calpain Activation, Including Direct Procaspase-3 Cleavage

    Directory of Open Access Journals (Sweden)

    Antonio Serapio-Palacios

    2016-06-01

    Full Text Available Enteropathogenic Escherichia coli (EPEC has the ability to antagonize host apoptosis during infection through promotion and inhibition of effectors injected by the type III secretion system (T3SS, but the total number of these effectors and the overall functional relationships between these effectors during infection are poorly understood. EspC produced by EPEC cleaves fodrin, paxillin, and focal adhesion kinase (FAK, which are also cleaved by caspases and calpains during apoptosis. Here we show the role of EspC in cell death induced by EPEC. EspC is involved in EPEC-mediated cell death and induces both apoptosis and necrosis in epithelial cells. EspC induces apoptosis through the mitochondrial apoptotic pathway by provoking (i a decrease in the expression levels of antiapoptotic protein Bcl-2, (ii translocation of the proapoptotic protein Bax from cytosol to mitochondria, (iii cytochrome c release from mitochondria to the cytoplasm, (iv loss of mitochondrial membrane potential, (v caspase-9 activation, (vi cleavage of procaspase-3 and (vii an increase in caspase-3 activity, (viii PARP proteolysis, and (ix nuclear fragmentation and an increase in the sub-G1 population. Interestingly, EspC-induced apoptosis was triggered through a dual mechanism involving both independent and dependent functions of its EspC serine protease motif, the direct cleavage of procaspase-3 being dependent on this motif. This is the first report showing a shortcut for induction of apoptosis by the catalytic activity of an EPEC protein. Furthermore, this atypical intrinsic apoptosis appeared to induce necrosis through the activation of calpain and through the increase of intracellular calcium induced by EspC. Our data indicate that EspC plays a relevant role in cell death induced by EPEC.

  16. Mirror Advanced Reactor Study (MARS)

    International Nuclear Information System (INIS)

    Logan, B.G.

    1983-01-01

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

  17. Nuclear power reactors

    International Nuclear Information System (INIS)

    1982-11-01

    After an introduction and general explanation of nuclear power the following reactor types are described: magnox thermal reactor; advanced gas-cooled reactor (AGR); pressurised water reactor (PWR); fast reactors (sodium cooled); boiling water reactor (BWR); CANDU thermal reactor; steam generating heavy water reactor (SGHWR); high temperature reactor (HTR); Leningrad (RMBK) type water-cooled graphite moderated reactor. (U.K.)

  18. Reactor physics and reactor computations

    International Nuclear Information System (INIS)

    Ronen, Y.; Elias, E.

    1994-01-01

    Mathematical methods and computer calculations for nuclear and thermonuclear reactor kinetics, reactor physics, neutron transport theory, core lattice parameters, waste treatment by transmutation, breeding, nuclear and thermonuclear fuels are the main interests of the conference

  19. Research reactors

    International Nuclear Information System (INIS)

    Merchie, Francois

    2015-10-01

    This article proposes an overview of research reactors, i.e. nuclear reactors of less than 100 MW. Generally, these reactors are used as neutron generators for basic research in matter sciences and for technological research as a support to power reactors. The author proposes an overview of the general design of research reactors in terms of core size, of number of fissions, of neutron flow, of neutron space distribution. He outlines that this design is a compromise between a compact enough core, a sufficient experiment volume, and high enough power densities without affecting neutron performance or its experimental use. The author evokes the safety framework (same regulations as for power reactors, more constraining measures after Fukushima, international bodies). He presents the main characteristics and operation of the two families which represent almost all research reactors; firstly, heavy water reactors (photos, drawings and figures illustrate different examples); and secondly light water moderated and cooled reactors with a distinction between open core pool reactors like Melusine and Triton, pool reactors with containment, experimental fast breeder reactors (Rapsodie, the Russian BOR 60, the Chinese CEFR). The author describes the main uses of research reactors: basic research, applied and technological research, safety tests, production of radio-isotopes for medicine and industry, analysis of elements present under the form of traces at very low concentrations, non destructive testing, doping of silicon mono-crystalline ingots. The author then discusses the relationship between research reactors and non proliferation, and finally evokes perspectives (decrease of the number of research reactors in the world, the Jules Horowitz project)

  20. Nuclear characteristic simulation device for reactor core

    International Nuclear Information System (INIS)

    Arakawa, Akio; Kobayashi, Yuji.

    1994-01-01

    In a simulation device for nuclear characteristic of a PWR type reactor, there are provided a one-dimensional reactor core dynamic characteristic model for simulating one-dimensional neutron flux distribution in the axial direction of the reactor core and average reactor power based on each of inputted signals of control rod pattern, a reactor core flow rate, reactor core pressure and reactor core inlet enthalphy, and a three-dimensional reactor core dynamic characteristic mode for simulating three-dimensional power distribution of the reactor core, and a nuclear instrumentation model for calculating read value of the nuclear instrumentation disposed in the reactor based on the average reactor core power and the reactor core three-dimensional power distribution. A one-dimensional neutron flux distribution in the axial direction of the reactor core, a reactor core average power, a reactor core three-dimensional power distribution and a nuclear instrumentation read value are calculated. As a result, the three-dimensional power distribution and the power level are continuously calculated. Further, since the transient change of the three-dimensional neutron flux distribution is calculated accurately on real time, more actual response relative to a power monitoring device of the reactor core and operation performance can be simulated. (N.H.)

  1. Multi-purpose reactor

    International Nuclear Information System (INIS)

    1991-05-01

    The Multi-Purpose-Reactor (MPR), is a pool-type reactor with an open water surface and variable core arrangement. Its main feature is plant safety and reliability. Its power is 22MW t h, cooled by light water and moderated by beryllium. It has platetype fuel elements (MTR type, approx. 20%. enriched uranium) clad in aluminium. Its cobalt (Co 60 ) production capacity is 50000 Ci/yr, 200 Ci/gr. The distribution of the reactor core and associated control and safety systems is essentially based on the following design criteria: - upwards cooling flow, to waive the need for cooling flow inversion in case the reactor is cooled by natural convection if confronted with a loss of pumping power, and in order to establish a superior heat transfer potential (a higher coolant saturation temperature); - easy access to the reactor core from top of pool level with the reactor operating at full power, in order to facilitate actual implementation of experiments. Consequently, mechanisms associated to control and safety rods s,re located underneath the reactor tank; - free access of reactor personnel to top of pool level with the reactor operating at full power. This aids in the training of personnel and the actual carrying out of experiments, hence: - a vast water column was placed over the core to act as radiation shielding; - the core's external area is cooled by a downwards flow which leads to a decay tank beyond the pool (for N 16 to decay); - a small downwards flow was directed to stream downwards from above the reactor core in order to drag along any possibly active element; and - a stagnant hot layer system was placed at top of pool level so as to minimize the upwards coolant flow rising towards pool level

  2. Reactor container

    International Nuclear Information System (INIS)

    Ichiki, Tadaharu; Nagatomi, Shozo.

    1976-01-01

    Object: To provide a jet and missile protective wall of a configuration being inflated toward the center of a reactor container on the inside of a body of the reactor container disposed within a biological shield wall to thereby increase safety of the reactor container. Structure: A jet and missile protective wall comprised of curved surfaces internally formed with a plurality of arch inflations filled with concrete between inner and outer iron plates and shape steel beam is provided between a reactor container surrounded by a biological shield wall and a thermal shield wall surrounding the reactor pressure vessel, and an adiabatic heat insulating material is filled in space therebetween. (Yoshino, Y.)

  3. Research reactor standards and their impact on the TRIGA reactor community

    International Nuclear Information System (INIS)

    Richards, W.J.

    1980-01-01

    The American Nuclear Society has established a standards committee devoted to writing standards for research reactors. This committee was formed in 1971 and has since that time written over 15 standards that cover all aspects of research reactor operation. The committee has representation from virtually every group concerned with research reactors and their operation. This organization includes University reactors, National laboratory reactors, Nuclear Regulatory commission, Department of Energy and private nuclear companies and insurers. Since its beginning the committee has developed standards in the following areas: Standard for the development of technical specifications for research reactors; Quality control for plate-type uranium-aluminium fuel elements; Records and reports for research reactors; Selection and training of personnel for research reactors; Review of experiments for research reactors; Research reactor site evaluation; Quality assurance program requirements for research reactors; Decommissioning of research reactors; Radiological control at research reactor facilities; Design objectives for and monitoring of systems controlling research reactor effluents; Physical security for research reactor facilities; Criteria for the reactor safety systems of research reactors; Emergency planning for research reactors; Fire protection program requirements for research reactors; Standard for administrative controls for research reactors. Besides writing the above standards, the committee is very active in using communications with the nuclear regulatory commission on proposed rules or positions which will affect the research reactor community

  4. ULTRA SCWR+: Practical advanced water reactor concepts

    International Nuclear Information System (INIS)

    Duffey, Romney; Khartabil, Hussam; Kuran, Sermet; Zhou, Tracy; Pioro, Igor

    2008-01-01

    Modern thermal power plants now utilize supercritical steam cycles with thermal efficiencies of over 45%. Recent developments have lead to Ultra-SuperCritical (USC) systems, which adopt reheat turbines that can attain efficiencies of over 50%. Because these turbines are already developed, demonstrated and deployed worldwide, and use existing and traditional steam cycle technology, the simplest nuclear advance is to utilize these proven thermal cycle conditions by coupling this turbine type to a reactor. This development direction is fundamentally counter to the usual approach of adopting high-temperature gas-cooled (helium-cooled) reactor cycles, for which turbines have yet to be demonstrated on commercial scale unlike the supercritical steam turbines. The ULTRA (Ultra-supercritical Light water Thermal ReActor) SCWR+ concept adopts the fundamental design approach of matching a water and steam-cooled reactor to the ultra-supercritical steam cycle, adopting the existing and planned thermal power plant turbines. The HP and IP sections are fed with conditions of 25 MPa/625degC and 7 MPa/700degC, respectively, to achieve operating plant thermal efficiencies in excess of 50%, with a direct turbine cycle. By using such low-pressure reheated steam, this concept also adopts technology that was explored and used many years ago in existing water reactors, with the potential to produce large quantities of low cost heat, which can be used for other industrial and district processes. Pressure-Tube (PT) reactors are suitable for adoption of this design approach and, in addition, have other advantages that will significantly improve water-cooled reactor technology. These additional advantages include enhanced safety and improved resource utilization and proliferation resistance. This paper describes the PT-SCWR+ concept and its potential enhancements. (author)

  5. Systematic evaluation program review of NRC Safety Topic VI-10.A associated with the electrical, instrumentation and control portions of the testing of reactor trip system and engineered safety features, including response time for the Dresden station, Unit II nuclear power plant

    International Nuclear Information System (INIS)

    St Leger-Barter, G.

    1980-11-01

    This report documents the technical evaluation and review of NRC Safety Topic VI-10.A, associated with the electrical, instrumentation, and control portions of the testing of reactor trip systems and engineered safety features including response time for the Dresden II nuclear power plant, using current licensing criteria

  6. Department of Reactor Technology

    DEFF Research Database (Denmark)

    Risø National Laboratory, Roskilde

    The general development of the Department of Reactor Technology at Risø during 1981 is presented, and the activities within the major subject fields are described in some detail. Lists of staff, publications, and computer programs are included.......The general development of the Department of Reactor Technology at Risø during 1981 is presented, and the activities within the major subject fields are described in some detail. Lists of staff, publications, and computer programs are included....

  7. Reactor power control device

    International Nuclear Information System (INIS)

    Imaruoka, Hiromitsu.

    1994-01-01

    A high pressure water injection recycling system comprising injection pipelines of a high pressure water injection system and a flow rate control means in communication with a pool of a pressure control chamber is disposed to a feedwater system of a BWR type reactor. In addition, the flow rate control means is controlled by a power control device comprising a scram impossible transient event judging section, a required injection flow rate calculation section for high pressure water injection system and a control signal calculation section. Feed water flow rate to be supplied to the reactor is controlled upon occurrence of a scram impossible transient event of the reactor. The scram impossible transient event is judged based on reactor output signals and scram operation demand signals and injection flow rate is calculated based on a predetermined reactor water level, and condensate storage tank water or pressure control chamber pool water is injected to the reactor. With such procedures, water level can be ensured and power can be suppressed. Further, condensate storage tank water of low enthalpy is introduced to the pressure suppression chamber pool to directly control elevation of water temperature and ensure integrity of the pressure vessel and the reactor container. (N.H.)

  8. Water cooled nuclear reactors

    International Nuclear Information System (INIS)

    Donaldson, A.J.

    1989-01-01

    In order to reduce any loss of primary water coolant from around a reactor core of a water cooled nuclear reactor caused by any failure of a pressure vessel, an inner vessel is positioned within and spaced from the pressure vessel. The reactor core and main portion of the primary water coolant circuit and a heat exchanger are positioned within the inner vessel to maintain some primary water coolant around the reactor core and to allow residual decay heat to be removed from the reactor core by the heat exchanger. In the embodiment shown an aperture at the upper region of the inner vessel is dimensioned configured and arranged to prevent steam from a steam space of an integral pressurised water cooled nuclear reactor for a ship entering the main portion of the primary water coolant circuit in the inner vessel if the longitudinal axis of the nuclear reactor is displaced from its normal substantially vertical position to an abnormal position at an angle to the vertical direction. Shields are integral with the inner vessel. (author)

  9. 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.

  10. RB research nuclear reactor - Annual report for 1986, I - III

    International Nuclear Information System (INIS)

    Markovic, H.; Pesic, M.; Vranic, S.; Petronijevic, M.; Jevremovic, M.; Ilic, I.

    1987-01-01

    This report includes data concerning the RB reactor operation in 1986, state of the reactor components, data about the employed personnel and the database of experimental and other reactor related devices. It is made of 3 parts: Engineering description and operation of the RB reactor including dosimetry, reactor staff data and financial report; Reactor facility components and maintenance; RB reactor operation and utilization in 1986 [sr

  11. Fuel Development For Gas-Cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    M. K. Meyer

    2006-06-01

    The Generation IV Gas-cooled Fast Reactor (GFR) concept is proposed to combine the advantages of high-temperature gas-cooled reactors (such as efficient direct conversion with a gas turbine and the potential for application of high-temperature process heat), with the sustainability advantages that are possible with a fast-spectrum reactor. The latter include the ability to fission all transuranics and the potential for breeding. The GFR is part of a consistent set of gas-cooled reactors that includes a medium-term Pebble Bed Modular Reactor (PBMR)-like concept, or concepts based on the Gas Turbine Modular Helium Reactor (GT-MHR), and specialized concepts such as the Very High Temperature Reactor (VHTR), as well as actinide burning concepts [ ]. To achieve the necessary high power density and the ability to retain fission gas at high temperature, the primary fuel concept proposed for testing in the United States is a dispersion coated fuel particles in a ceramic matrix. Alternative fuel concepts considered in the U.S. and internationally include coated particle beds, ceramic clad fuel pins, and novel ceramic ‘honeycomb’ structures. Both mixed carbide and mixed nitride-based solid solutions are considered as fuel phases.

  12. The development of fast reactors in France

    International Nuclear Information System (INIS)

    Vautrey, L.

    1982-01-01

    Only minor changes were introduced in the French nuclear programme by the new government in 1981. The operating conditions of Rapsodie were very satisfactory up to January 1982. After a leak in the double primary jacket (nitrogen circuit) the reactor was shut down for investigations. Phenix is continuing to operate smoothly. Construction of Super Phenix (Creys Malville power plant) is proceeding normally though with some delay. The studies for the future (after Creys Malville) are following their way both for the Project 1500 (Super Phenix 2) and for the specific plants of the fuel cycle. Research and development are largely directed toward Super Phenix 1 needs and the prospects of Super Phenix 2. International cooperation remains very intensive. The financial resources devoted to the development of fast reactors are globally stable. Including fuel cycle and safety (but excluding the Phenix operation) about 1300 millions of francs will be devoted to fast reactors by the C.E.A. in 1982. (author)

  13. Project for the ASN opinion related to the generic directions of the periodic safety re-examination associated to the fourth decennial inspections of 900 MWe reactors. Contribution of the ANCCLI to the public consultation

    International Nuclear Information System (INIS)

    2016-01-01

    As the first exploited nuclear reactors are about to reach their forty years of operation, and as EDF wishes to extend their exploitation duration beyond 40 years, EDF proposed new directions for the periodic safety examination associated with the fourth decennial inspections of the concerned reactors (900 MWe). The ANCCLI herein proposes its contribution to this issue. This document first describes elements of context, discusses the evolution of margins, discusses how uncertainties are to be managed, discusses the definition of limits and the impact of ageing on the in-depth defence strategy, and addresses the issue of the number of files to be examined. It also comments and discusses various issues related to safety re-examination: existence of a global view, required levels, unacceptable deviations, assessment criteria for reasonableness, identified points of attention, specific case of de-activation pools, and articulation with past or planned reinforcements. The next parts address operator means, and the decision process

  14. Utilization of research reactors - A global perspective

    International Nuclear Information System (INIS)

    Muranaka, R.G.

    1988-01-01

    This paper presents 1) a worldwide picture of research reactors, operable, shutdown, under construction and planned, 2) statistics on utilization of research reactors including TRIGA reactors, and 3) some results of a survey conducted during 1988 on the utilization of research reactors in developing Member States in the Asia-Pacific Region

  15. Nuclear problems of power reactors safety

    International Nuclear Information System (INIS)

    Kocic, A.

    1980-01-01

    The objective of this presentation is to emphasize the conditions that would be of high importance in safety analyses concerned first of all with reactor core. It describes reactor kinetics processes in the core and build up of fission products, classification of reactor accidents related to the core, risk estimation and includes a list of importance reactor accidents

  16. Integrated nuclear reactor

    International Nuclear Information System (INIS)

    Pales, I.; Hasko, V.

    1984-01-01

    The reactor is provided with an integrated circuit of primary medium circulation with hydraulic pump drive. The pump drive which is a blade hydraulic facility is placed in the reactor vessel together with the pump. The primary medium flows through the core and enters the inter-tube space of the secondary circuit heat exchanger. The secondary circuit medium is supplied under the bottom tube plate with a supply pipe. From it the flow of secondary medium is directed to the blades of the hydraulic facility, e.g. the turbine. The turbine drives the pump which transports the primary medium to the reactor core. The secondary medium enters the heat exchanger tubes and through their walls receives the heat from the primary medium. This design reduces capital costs of the reactor and increases its safety. (E.S.)

  17. Mirror hybrid reactor optimization studies

    International Nuclear Information System (INIS)

    Bender, D.J.

    1976-01-01

    A system model of the mirror hybrid reactor has been developed. The major components of the model include (1) the reactor description, (2) a capital cost analysis, (3) various fuel management schemes, and (4) an economic analysis that includes the hybrid plus its associated fission burner reactors. The results presented describe the optimization of the mirror hybrid reactor, the objective being to minimize the cost of electricity from the hybrid fission-burner reactor complex. We have examined hybrid reactors with two types of blankets, one containing natural uranium, the other thorium. The major difference between the two optimized reactors is that the uranium hybrid is a significant net electrical power producer, whereas the thorium hybrid just about breaks even on electrical power. Our projected costs for fissile fuel production are approximately 50 $/g for 239 Pu and approximately 125 $/g for 233 U

  18. Reactor Safety Research Programs

    Energy Technology Data Exchange (ETDEWEB)

    Edler, S. K.

    1981-07-01

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

  19. Hybrid reactors

    International Nuclear Information System (INIS)

    Moir, R.W.

    1980-01-01

    The rationale for hybrid fusion-fission reactors is the production of fissile fuel for fission reactors. A new class of reactor, the fission-suppressed hybrid promises unusually good safety features as well as the ability to support 25 light-water reactors of the same nuclear power rating, or even more high-conversion-ratio reactors such as the heavy-water type. One 4000-MW nuclear hybrid can produce 7200 kg of 233 U per year. To obtain good economics, injector efficiency times plasma gain (eta/sub i/Q) should be greater than 2, the wall load should be greater than 1 MW.m -2 , and the hybrid should cost less than 6 times the cost of a light-water reactor. Introduction rates for the fission-suppressed hybrid are usually rapid

  20. Fusion Reactor Materials

    International Nuclear Information System (INIS)

    Decreton, M.

    2001-01-01

    The objective of SCK-CEN's programme on fusion reactor materials is to contribute to the knowledge on the behaviour of fusion reactor materials and components during and after irradiation. Ongoing projects include: the study of the mechanical behaviour of structural materials under neutron irradiation; the investigation of the characteristics of irradiated first wall material such as beryllium; the detection of abrupt electrical degradation of insulating ceramics under high temperature and neutron irradiation; and the study of dismantling and waste disposal strategy for fusion reactors. Progress and achievements in these areas in 2000 are discussed

  1. Fast Breeder Reactor studies

    International Nuclear Information System (INIS)

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

    1980-07-01

    This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts

  2. Fast Breeder Reactor studies

    Energy Technology Data Exchange (ETDEWEB)

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

    1980-07-01

    This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts.

  3. Heterogeneous reactors

    International Nuclear Information System (INIS)

    Moura Neto, C. de; Nair, R.P.K.

    1979-08-01

    The microscopic study of a cell is meant for the determination of the infinite multiplication factor of the cell, which is given by the four factor formula: K(infinite) = n(epsilon)pf. The analysis of an homogeneous reactor is similar to that of an heterogeneous reactor, but each factor of the four factor formula can not be calculated by the formulas developed in the case of an homogeneous reactor. A great number of methods was developed for the calculation of heterogeneous reactors and some of them are discussed. (Author) [pt

  4. Chernobyl reactor accident

    International Nuclear Information System (INIS)

    1986-05-01

    Following the accident at Chernobyl nuclear reactor, WHO organized on 6 May 1986 in Copenhagen a one day consultation of experts with knowledge in the fields of meteorology, radiation protection, biological effects, reactor technology, emergency procedures, public health and psychology in order to analyse the development of events and their consequences and to provide guidance as to the needs for immediate public health action. The present report provides detailed information on the transportation and dispersion of the radioactive material in the atmosphere, especially volatile elements, during the release period 26 April - 5 May. Presented are the calculated directions and locations of the radioactive plume over Europe in the first 5 days after the accident, submitted by the Swedish Meteorological and Hydrological Institute. The calculations have been made for two heights, 1500m and 750m and the plume directions are grouped into five periods, covering five European areas. The consequences of the accident inside the USSR and the radiological consequences outside the USSR are presented including the exposure routes and the biological effects, paying particular attention to iodine-131 effects. Summarized are the first reported measured exposure rates above background, iodine-131 deposition and concentrations in milk and the remedial actions taken in various European countries. Concerning the cesium-137 problem, based on the UNSCEAR assessment of the consequences of the nuclear fallout, one concludes that the cesium contamination outside the USSR is not likely to cause any serious problems. Finally, the conclusions and the recommendations of the meeting, taking into account both the short-term and longer term considerations are presented

  5. Fast reactor programme in India

    Indian Academy of Sciences (India)

    2015-09-04

    Sep 4, 2015 ... 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 ...

  6. Fusion Reactor Materials

    Energy Technology Data Exchange (ETDEWEB)

    Decreton, M

    2002-04-01

    The objective of SCK-CEN's programme on fusion reactor materials is to contribute to the knowledge on the radiation-induced behaviour of fusion reactor materials and components as well as to help the international community in building the scientific and technical basis needed for the construction of the future reactor. Ongoing projects include: the study of the mechanical and chemical (corrosion) behaviour of structural materials under neutron irradiation and water coolant environment; the investigation of the characteristics of irradiated first wall material such as beryllium; investigations on the management of materials resulting from the dismantling of fusion reactors including waste disposal. Progress and achievements in these areas in 2001 are discussed.

  7. Reactor vessel sealing plug

    International Nuclear Information System (INIS)

    Dooley, R.A.

    1986-01-01

    This invention relates to an apparatus and method for sealing the cold leg nozzles of a nuclear reactor pressure vessel from a remote location during maintenance and inspection of associated steam generators and pumps while the pressure vessel and refueling canal are filled with water. The apparatus includes a sealing plug for mechanically sealing the cold leg nozzle from the inside of a reactor pressure vessel. The sealing plugs include a primary and a secondary O-ring. An installation tool is suspended within the reactor vessel and carries the sealing plug. The tool telescopes to insert the sealing plug within the cold leg nozzle, and to subsequently remove the plug. Hydraulic means are used to activate the sealing plug, and support means serve to suspend the installation tool within the reactor vessel during installation and removal of the sealing plug

  8. Gas cooled high temperature reactor with a heap of pebble shaped fuel elements and absorber rods which can be driven directly into the heap of pebbles

    International Nuclear Information System (INIS)

    Elter, C.; Schmitt, H.; Schoening, J.; Weicht, U.

    1980-01-01

    The absorber rod for the graphite moderated, helium cooled reactor is cylindrical and has a tip in the shape of the frustrum of a cone. It consists of three coaxially arranged sleeve tubes made of steel, the inner and centre sleeve tubes surrounding the absorber part (B4C) so as to be gastight. The inner sleeve tube represents the supporting tube and is cooled by cold gas, as is the annular gap between the centre and outer sleeve tube. (RW) [de

  9. Reactor technology: power conversion systems and reactor operation and maintenance

    International Nuclear Information System (INIS)

    Powell, J.R.

    1977-01-01

    The use of advanced fuels permits the use of coolants (organic, high pressure helium) that result in power conversion systems with good thermal efficiency and relatively low cost. Water coolant would significantly reduce thermal efficiency, while lithium and salt coolants, which have been proposed for DT reactors, will have comparable power conversion efficiencies, but will probably be significantly more expensive. Helium cooled blankets with direct gas turbine power conversion cycles can also be used with DT reactors, but activation problems will be more severe, and the portion of blanket power in the metallic structure will probably not be available for the direct cycle, because of temperature limitations. A very important potential advantage of advanced fuel reactors over DT fusion reactors is the possibility of easier blanket maintenance and reduced down time for replacement. If unexpected leaks occur, in most cases the leaking circuit can be shut off and a redundant cooling curcuit will take over the thermal load. With the D-He 3 reactor, it appears practical to do this while the reactor is operating, as long as the leak is small enough not to shut down the reactor. Redundancy for Cat-D reactors has not been explored in detail, but appears feasible in principle. The idea of mobile units operating in the reactor chamber for service and maintenance of radioactive elements is explored

  10. Research reactor support

    International Nuclear Information System (INIS)

    2005-01-01

    Research reactors (RRs) have been used in a wide range of applications including nuclear power development, basic physics research, education and training, medical isotope production, geology, industry and other fields. However, many research reactors are fuelled with High Enriched Uranium (HEU), are underutilized and aging, and have significant quantities of spent fuel. HEU inventories (fresh and spent) pose security risks Unavailability of a high-density-reprocessable fuel hinders conversion and limits back-end options and represents a survival dilemma for many RRs. Improvement of interim spent fuel storage is required at some RRs. Many RRs are under-utilized and/or inadequately funded and need to find users for their services, or permanently shut down and eventually decommission. Reluctance to decommission affect both cost and safety (loss of experienced staff ) and many shut down but not decommissioned RR with fresh and/or spent fuel at the sites invoke serious concern. The IAEA's research reactor support helps to ensure that research reactors can be operated efficiently with fuels and targets of lower proliferation and security concern and that operators have appropriate technology and options to manage RR fuel cycle issues, especially on long term interim storage of spent research reactor fuel. Availability of a high-density-reprocessable fuel would expand and improve back end options. The International Atomic Energy Agency provides assistance to Member States to convert research reactors from High Enriched Uranium fuel and targets (for medical isotope production) to qualified Low Enriched Uranium fuel and targets while maintaining reactor performance levels. The assistance includes provision of handbooks and training in the performance of core conversion studies, advice for the procurement of LEU fuel, and expert services for LEU fuel acceptance. The IAEA further provides technical and administrative support for countries considering repatriation of its

  11. Spiral-shaped disinfection reactors

    KAUST Repository

    Ghaffour, Noreddine

    2015-08-20

    This disclosure includes disinfection reactors and processes for the disinfection of water. Some disinfection reactors include a body that defines an inlet, an outlet, and a spiral flow path between the inlet and the outlet, in which the body is configured to receive water and a disinfectant at the inlet such that the water is exposed to the disinfectant as the water flows through the spiral flow path. Also disclosed are processes for disinfecting water in such disinfection reactors.

  12. Reactor pressure vessel nozzle

    Science.gov (United States)

    Challberg, Roy C.; Upton, Hubert A.

    1994-01-01

    A nozzle for joining a pool of water to a nuclear reactor pressure vessel includes a tubular body having a proximal end joinable to the pressure vessel and a distal end joinable in flow communication with the pool. The body includes a flow passage therethrough having in serial flow communication a first port at the distal end, a throat spaced axially from the first port, a conical channel extending axially from the throat, and a second port at the proximal end which is joinable in flow communication with the pressure vessel. The inner diameter of the flow passage decreases from the first port to the throat and then increases along the conical channel to the second port. In this way, the conical channel acts as a diverging channel or diffuser in the forward flow direction from the first port to the second port for recovering pressure due to the flow restriction provided by the throat. In the backflow direction from the second port to the first port, the conical channel is a converging channel and with the abrupt increase in flow area from the throat to the first port collectively increase resistance to flow therethrough.

  13. KS-150 reactor control

    International Nuclear Information System (INIS)

    Wagner, K.

    1974-01-01

    A thorough description is presented of the control and protection system of the Bohunice A-1 reactor. The system including auxiliary facilities was developed, manufactured and installed at the reactor by the SKODA Works, Plzen. The system parameters are listed and a brief account is also given of the development efforts and of the physical and power start-up of the A-1 nuclear power plant. (L.O.)

  14. Lateral restraint assembly for reactor core

    Science.gov (United States)

    Gorholt, Wilhelm; Luci, Raymond K.

    1986-01-01

    A restraint assembly for use in restraining lateral movement of a reactor core relative to a reactor vessel wherein a plurality of restraint assemblies are interposed between the reactor core and the reactor vessel in circumferentially spaced relation about the core. Each lateral restraint assembly includes a face plate urged against the outer periphery of the core by a plurality of compression springs which enable radial preloading of outer reflector blocks about the core and resist low-level lateral motion of the core. A fixed radial key member cooperates with each face plate in a manner enabling vertical movement of the face plate relative to the key member but restraining movement of the face plate transverse to the key member in a plane transverse to the center axis of the core. In this manner, the key members which have their axes transverse to or subtending acute angles with the direction of a high energy force tending to move the core laterally relative to the reactor vessel restrain such lateral movement.

  15. Reactor Sharing at Rensselaer Critical Facility

    International Nuclear Information System (INIS)

    D. Steiner, D. Harris, T. Trumbull

    2006-01-01

    This final report summarizes the reactor sharing activities at the Rensselaer Critical Facility. An example of a typical tour is also included. Reactor sharing at the RCF brings outside groups into the facility for a tour, an explanation of reactor matters, and a reactor measurement. It has involved groups ranging from high school classes to advanced college groups and in size from a few to about 50 visitors. The RCF differs from other university reactors in that its fuel is like that of large power reactors, and its research and curriculum are dedicated to power reactor matters

  16. Nuclear reactor

    International Nuclear Information System (INIS)

    Irion, L.; Tautz, J.; Ulrych, G.

    1976-01-01

    This additional patent complements the arrangement of non-return valves to prevent loss of cooling water on fracture of external tubes in the main coolant circuit (according to PS 24 24 427.7) by ensuring that the easily movable valves only operate in case of a fault, but do not flutter in operation, because the direction of flow is not the same at each location where they are installed. The remedy for this undesirable effect consists of allocating 1 non-return valve unit with 5 to 10 valves to each (of several) ducts for the cooling water intake. These units are installed in the annular space between the reactor vessel and the pressure vessel below the inlet of the ducts. Due to flow guidance surfaces in the same space, the incoming cooling water is deflected downwards and as the guiding surfaces are closed at the sides, must pass parallel to the valves of the non-return valve unit. On fracture of the external cooling water inlet pipe concerned, all valves of this unit close due to reversal of flow on the outlet side. (TK) [de

  17. Nuclear reactor

    International Nuclear Information System (INIS)

    Mysels, K.J.; Shenoy, A.S.

    1976-01-01

    A nuclear reactor is described in which the core consists of a number of fuel regions through each of which regulated coolant flows. The coolant from neighbouring fuel regions is combined in a manner which results in an averaging of the coolant temperature at the outlet of the core. By this method the presence of hot streaks in the reactor is reduced. (UK)

  18. OECD Halden reactor project

    International Nuclear Information System (INIS)

    1979-01-01

    This is the nineteenth annual Report on the OECD Halden Reactor Project, describing activities at the Project during 1978, the last year of the 1976-1978 Halden Agreement. Work continued in two main fields: test fuel irradiation and fuel research, and computer-based process supervision and control. Project research on water reactor fuel focusses on various aspects of fuel behavior under normal, and off-normal transient conditions. In 1978, participating organisations continued to submit test fuel for irradiation in the Halden boiling heavy-water reactor, in instrumented test assemblies designed and manufactured by the Project. Work included analysis of the impact of fuel design and reactor operating conditions on fuel cladding behavior. Fuel performance modelling included characterization of thermal and mechanical behavior at high burn-up, of fuel failure modes, and improvement of data qualification procedures to reduce and quantify error bands on in-reactor measurements. Instrument development yielded new or improved designs for measuring rod temperature, internal pressure, axial neutron flux shape determination, and for detecting cladding defects. Work on computer-based methods of reactor supervision and control included continued development of a system for predictive core surveillance, and of special mathematical methods for core power distribution control

  19. Nuclear reactor

    International Nuclear Information System (INIS)

    Tilliette, Z.

    1975-01-01

    A description is given of a nuclear reactor and especially a high-temperature reactor in which provision is made within a pressure vessel for a main cavity containing the reactor core and a series of vertical cylindrical pods arranged in spaced relation around the main cavity and each adapted to communicate with the cavity through two collector ducts or headers for the primary fluid which flows downwards through the reactor core. Each pod contains two superposed steam-generator and circulator sets disposed in substantially symmetrical relation on each side of the hot primary-fluid header which conveys the primary fluid from the reactor cavity to the pod, the circulators of both sets being mounted respectively at the bottom and top ends of the pod

  20. Mirror reactor studies

    International Nuclear Information System (INIS)

    Moir, R.W.; Barr, W.L.; Bender, D.J.

    1976-01-01

    Design studies of a fusion mirror reactor, a fusion-fission mirror reactor, and two small mirror reactors are summarized. The fusion reactor uses 150-keV neutral-beam injectors based on the acceleration of negative ions. The injectors provide over 1 GW of continuous power at an efficiency greater than 80 percent. The fusion reactor has three-stage, modularized, Venetian blind, plasma direct converter with a predicted efficiency of 59 percent and a new concept for removal of the lune-shaped blanket: a crane is brought between the two halves of the Yin-Yang magnet, which are separated by a float. The design has desirable features such as steady-state operation, minimal impurity problems, and low first-wall thermal stress. The major disadvantage is low Q resulting in high recirculating power and hence high cost of electrical power. However, the direct capital cost per unit of gross electrical power is reasonable [$1000/kW(e)

  1. Proceedings of the international symposium on research reactor safety operations and modifications

    International Nuclear Information System (INIS)

    1990-03-01

    The International Symposium on Research Reactor Safety, Operations and Modifications was organized by the International Atomic Energy Agency in cooperation with Atomic Energy of Canada Limited-Research Company. The main objectives of this Symposium were: (1) to exchange information and to discuss current perspectives and concerns relating to all aspects to research reactor safety, operations, and modifications; and, (2) to present views and to discuss future initiatives and directions for research reactor design, operations, utilization, and safety. The symposium topics included: research reactor programmes and experience; research reactor design safety and analysis; research reactor modifications and decommissioning; research reactor licensing; and new research reactors. These topics were covered during eight oral sessions and three poster sessions. These Proceedings include the full text of the 93 papers presented. The subject of Symposium was quite wide-ranging in that it covered essentially all aspects of research reactor safety, operations, and modifications. This was considered to be appropriate and timely given the 326 research reactors currently in operation in some 56 countries; given the degree of their utilization which ranges from pure and applied research to radioisotopes production to basic training and manpower development; and given that many of these reactors are undergoing extensive modifications, core conversions, power upratings, and are becoming the subject of safety reassessment and regulatory reviews. Although the Symposium covered many topics, the majority of papers and discussions tended to focus mainly on research reactor safety. This was seen as a clear sign of the continuing recognition of the fundamental importance of identifying and addressing, particularly through international cooperation, issues and concerns associated with research reactor safety

  2. RADIATION FACILITY FOR NUCLEAR REACTORS

    Science.gov (United States)

    Currier, E.L. Jr.; Nicklas, J.H.

    1961-12-12

    A radiation facility is designed for irradiating samples in close proximity to the core of a nuclear reactor. The facility comprises essentially a tubular member extending through the biological shield of the reactor and containing a manipulatable rod having the sample carrier at its inner end, the carrier being longitudinally movable from a position in close proximity to the reactor core to a position between the inner and outer faces of the shield. Shield plugs are provided within the tubular member to prevent direct radiation from the core emanating therethrough. In this device, samples may be inserted or removed during normal operation of the reactor without exposing personnel to direct radiation from the reactor core. A storage chamber is also provided within the radiation facility to contain an irradiated sample during the period of time required to reduce the radioactivity enough to permit removal of the sample for external handling. (AEC)

  3. Method of producing gaseous products using a downflow reactor

    Science.gov (United States)

    Cortright, Randy D; Rozmiarek, Robert T; Hornemann, Charles C

    2014-09-16

    Reactor systems and methods are provided for the catalytic conversion of liquid feedstocks to synthesis gases and other noncondensable gaseous products. The reactor systems include a heat exchange reactor configured to allow the liquid feedstock and gas product to flow concurrently in a downflow direction. The reactor systems and methods are particularly useful for producing hydrogen and light hydrocarbons from biomass-derived oxygenated hydrocarbons using aqueous phase reforming. The generated gases may find used as a fuel source for energy generation via PEM fuel cells, solid-oxide fuel cells, internal combustion engines, or gas turbine gensets, or used in other chemical processes to produce additional products. The gaseous products may also be collected for later use or distribution.

  4. Reactor BR2. Introduction

    Energy Technology Data Exchange (ETDEWEB)

    Gubel, P

    2001-04-01

    The BR2 is a materials testing reactor and is still one of SCK-CEN's important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. During the last three years, the availability of the installation was maintained at an average level of 97.6 percent. In the year 2000, the reactor was operated for a total of 104 days at a mean power of 56 MW. In 2000, most irradiation experiments were performed in the CALLISTO PWR loop. The report describes irradiations achieved or under preparation in 2000, including the development of advanced facilities and concept studies for new programmes. An overview of the scientific irradiation programmes as well as of the R and D programme of the BR2 reactor in 2000 is given.

  5. Reactor BR2. Introduction

    International Nuclear Information System (INIS)

    Gubel, P.

    2002-01-01

    The BR2 materials testing reactor is one of SCK-CEN's most important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. In 2001, the reactor was operated for a total of 123 days at a mean power of 59 MW in order to satisfy the irradiation conditions of the internal and external programmes using mainly the CALLISTO PWR loop. The mean consumption of fresh fuel elements was 5.26 per 1000 MWd. Main achievements in 2001 included the development of a three-dimensional full-scale model of the BR2 reactor for simulation and prediction of irradiation conditions for various experiments; the construction of the FUTURE-MT device designed for the irradiation of fuel plates under representative conditions of geometry, neutron spectrum, heat flux and thermal-hydraulic conditions and the development of in-pile instrumentation and a data acquisition system

  6. Reactor BR2. Introduction

    Energy Technology Data Exchange (ETDEWEB)

    Gubel, P

    2002-04-01

    The BR2 materials testing reactor is one of SCK-CEN's most important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. In 2001, the reactor was operated for a total of 123 days at a mean power of 59 MW in order to satisfy the irradiation conditions of the internal and external programmes using mainly the CALLISTO PWR loop. The mean consumption of fresh fuel elements was 5.26 per 1000 MWd. Main achievements in 2001 included the development of a three-dimensional full-scale model of the BR2 reactor for simulation and prediction of irradiation conditions for various experiments; the construction of the FUTURE-MT device designed for the irradiation of fuel plates under representative conditions of geometry, neutron spectrum, heat flux and thermal-hydraulic conditions and the development of in-pile instrumentation and a data acquisition system.

  7. Nuclear research reactors activities in INVAP

    International Nuclear Information System (INIS)

    Ordonez, Juan Pablo

    2013-01-01

    This presentation describes the different activities in the research reactor field that are being carried out by INVAP. INVAP is presently involved in the design of three new research reactors in three different countries. The RA-10 is a multipurpose reactor, in Argentina, planned as a replacement for the RA-3 reactor. INVAP was contracted by CNEA for carrying out the preliminary engineering for this reactor, and has recently been contracted by CNEA for the detailed engineering. CNEA groups are strongly involved in the design of this reactor. The RMB is a multipurpose reactor, planned by CNEN from Brazil. CNEN, through REDETEC, has contracted INVAP to carry out the preliminary engineering for this reactor. As the user requirements for RA-10 and RMB are very similar, an agreement was signed between Argentina and Brasil governments to cooperate in these two projects. The agreement included that both reactors would use the OPAL reactor in Australia, design and built by INVAP, as a reference reactor. INVAP has also designed the LPRR reactor for KACST in Saudi Arabia. The LPRR is a 30 kw reactor for educational purposes. KACST initially contracted INVAP for the engineering for this reactor and has recently signed the contract with INVAP for building the reactor. General details of these three reactors will be presented

  8. Official announcement of the directive on protection of nuclear power plant equipped with LWR-type reactors from human intrusion or other interference by third parties. Announcement of BMU (German Federal Ministry Environment), of 6 Dec. 1995 - RS I 3 13151 - 6/14

    International Nuclear Information System (INIS)

    1996-01-01

    An operating permit for a nuclear power plant is to be granted only if the applicant and facility operator presents evidence guaranteeing the legally required physical protection and other security measures for protection from human instrusion and other type of interference. As a basis for review and licensing, the competent authorities in 1987 have issued a directive specifying the requirements to be met for physical protection of nuclear power plant equipped with PWR-type reactors, and in 1994 followed a second, analogous directive relating to nuclear power plant with BWR-type reactors. The directive now announced for physical protection of nuclear power plant equipped with LWR-type reactors combines and replaces the two former ones, and from the date of the announcement is the only applicable directive. The text of the directive is not reproduced for reasons of secrecy protection. (orig./CB) [de

  9. Reactor neutrons in nuclear astrophysics

    OpenAIRE

    Reifarth, R.; Glorius, J.; Gobel, K.; Heftrich, T.; Jentschel, M.; Jurado, B.; Käppeler, F.; Köster, U.; Langer, C.; Litvinov, Y.A.; Weigand, M.

    2017-01-01

    The huge neutron fluxes offer the possibility to use research reactors to produce isotopes of interest, which can be investigated afterwards. An example is the half-lives of long-lived isotopes like 129I. A direct usage of reactor neutrons in the astrophysical energy regime is only possible, if the corresponding ions are not at rest in the laboratory frame. The combination of an ion storage ring with a reactor and a neutron guide could open the path to direct measurements of neutron-induced c...

  10. Nuclear reactor kinetics and control

    International Nuclear Information System (INIS)

    Lewins, J.

    1978-01-01

    A consistent, integrated account of modern developments in the study of nuclear reactor kinetics and the problem of their efficient and safe control. It aims to prepare the student for advanced study and research or practical work in the field. Special features include treatments of noise theory, reliability theory and safety related studies. It covers all aspects of the operation and control of nuclear reactors, power and research and is complete in providing physical data methods of calculation and solution including questions of equipment reliability. The work uses illustrations of the main types of reactors in use in the UK, USA and Europe. Each chapter contains problems and worked examples suitable for course work and study. The subject is covered in chapters, entitled: introductory review; neutron and precursor equations; elementary solutions at low power; linear reactor process dynamics with feedback; power reactor control systems; fluctuations and reactor noise; safety and reliability; nonlinear systems (safety and control); analogue computing. (author)

  11. Power reactors in member states

    International Nuclear Information System (INIS)

    1975-01-01

    This is the first issue of a periodical computer-based listing of civilian nuclear power reactors in the Member States of the IAEA, presenting the situation as of 1 April 1975. It is intended as a replacement for the Agency's previous annual publication of ''Power and Research Reactors in Member States''. In the new format, the listing contains more information about power reactors in operation, under construction, planned and shut down. As far as possible all the basic design data relating to reactors in operation have been included. In future these data will be included also for other power reactors, so that the publication will serve to give a clear picture of the technical progress achieved. Test and research reactors and critical facilities are no longer listed. Of interest to nuclear power planners, nuclear system designers, nuclear plant operators and interested professional engineers and scientists

  12. Reactor simulator development. Workshop material

    International Nuclear Information System (INIS)

    2001-01-01

    process, including numerical methods, and an introduction to the CASSIM TM development system. Techniques and practice are presented, with exercises, for development and application of reactor simulators

  13. SLOWPOKE reactor

    International Nuclear Information System (INIS)

    Evans, D.J.R.; Downs, W.E.

    1974-01-01

    The SLOWPOKE reactor is described, which is a small pool type with thermal neutron fluxes ranging from 10 11 -10 12 n cm -2 sec -1 . It differs in many ways from conventional pool type, namely small critical mass, beryllium reflector and a closed reactor container. The reactor is designed as small and simply as possible, and consistently with safety and good operating practice. Access to the present model is via pneumatic irradiation tubes only, which limits the use of the facility to activation analysis, tracer production and training. (Mori, K.)

  14. Reactor pressure vessel materials

    International Nuclear Information System (INIS)

    Suzuki, K.

    1998-01-01

    As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. Chapter 3 offers a detailed treatment of the selection criteria and properties of reactor pressure vessel materials. The main attention is directed towards steel and ingot making and the subsequent material processing

  15. Inorganic membranes and catalytic reactors

    OpenAIRE

    Rangel, Maria do Carmo

    1997-01-01

    Membrane reactors are reviewed with emphasis in their applications in catalysis field. The basic principles of these systems are presented as well as a historical development. The several kinds of catalytic membranes and their preparations are discussed including the problems, needs and challenges to be solved in order to use these reactors in commercial processes. Some applications of inorganic membrane reactors are also shown. It was concluded that these systems have a great potential for i...

  16. Decree n. 2007-534 of the 4. april 2007 allowing the creation of the base nuclear installation named Flamanville 3, including a EPR type reactor, on the site of Flamanville (Manche); Decret no 2007-534 du 10 avril 2007 autorisant la creation de l'installation nucleaire de base denommee Flamanville 3, comportant un reacteur nucleaire de type EPR, sur le site de Flamanville (Manche)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2007-04-15

    This decree gives the authorization to EDF to create on the site of Flamanville a nuclear installation including a PWR type reactor for a power of 4500 MW and devoted to the electric production. This reactor will can use uranium oxide or a mixture of uranium oxide and plutonium oxide. Considerations concerning the safety are given, as well as the control of the impact of this exploitation on the populations and the environment. (N.C.)

  17. NUCLEAR REACTOR

    Science.gov (United States)

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

    1958-01-21

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

  18. REACTOR SHIELD

    Science.gov (United States)

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

    1959-02-17

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

  19. NEUTRONIC REACTOR

    Science.gov (United States)

    Anderson, H.L.

    1960-09-20

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

  20. RB research nuclear reactor, Annual report for 1989, I - III

    International Nuclear Information System (INIS)

    Stefanovic, D.; Pesic, M.; Hadimahmutovic, N.; Vranic, S.; Petronijevic, M.; Jevremovic, M.; Ilic, I.

    1989-12-01

    This report is made of three parts. Part one contains a short description of the reactor, reactor operation, incidents, status of reactor equipment and components (nuclear fuel, heavy water, reactor vessel, heavy water circulation system, electronic, electric and mechanical equipment, auxiliary systems and Vax-8250 computer). It includes dosimetry and radiation protection data, personnel and financial data. Second part of this report in concerned with maintenance of reactor components and instrumentation. Part three includes data about reactor utilization during 1989

  1. Reactor container

    International Nuclear Information System (INIS)

    Kagawa, Tatsuo; Yanai, Ryoichi.

    1976-01-01

    Object: To provide a reactor container which is free from water shock action or condensing vibrations and cannot be readily broken by a missile from a pump impeller, pipe whipping, steam jet reaction, etc., and which also quickly condenses issuing steam and possesses a large vibration-proof strength. Structure: A high pressure containment vessel accommodating a pressure container includes a plurality of pressurized water tanks arranged along its inner periphery, and a pneumatic valve is provided in a lower portion of each of these pressurized water tanks. If an accident occurs, vapor is caused to issue from the pressure container into the vessel. When a certain value is reached, the pneumatic valves are opened, whereby the gas within the pressurized water tanks causes pressurized water to flow through the pipe and be ejected from spray nozzles to cause condensation of water within the vessel. Further, water of a pool within the container is circulated to allow heat release to the outside. (Horiuchi, T.)

  2. Reactor technology

    International Nuclear Information System (INIS)

    Erdoes, P.

    1977-01-01

    This is one of a series of articles discussing aspects of nuclear engineering ranging from a survey of various reactor types for static and mobile use to mention of atomic thermo-electric batteries of atomic thermo-electric batteries for cardiac pacemakers. Various statistics are presented on power generation in Europe and U.S.A. and economics are discussed in some detail. Molten salt reactors and research machines are also described. (G.M.E.)

  3. TU Electric reactor physics model verification: Power reactor benchmark

    International Nuclear Information System (INIS)

    Willingham, C.E.; Killgore, M.R.

    1988-01-01

    Power reactor benchmark calculations using the advanced code package CASMO-3/SIMULATE-3 have been performed for six cycles of Prairie Island Unit 1. The reload fuel designs for the selected cycles included gadolinia as a burnable absorber, natural uranium axial blankets and increased water-to-fuel ratio. The calculated results for both startup reactor physics tests (boron endpoints, control rod worths, and isothermal temperature coefficients) and full power depletion results were compared to measured plant data. These comparisons show that the TU Electric reactor physics models accurately predict important measured parameters for power reactors

  4. Oregon State University TRIGA Reactor annual report

    International Nuclear Information System (INIS)

    Anderson, T.V.; Johnson, A.G.; Bennett, S.L.; Ringle, J.C.

    1979-01-01

    The use of the Oregon State University TRIGA Reactor during the year ending June 30, 1979, is summarized. Environmental and radiation protection data related to reactor operation and effluents are included

  5. Materials for nuclear reactors

    International Nuclear Information System (INIS)

    Banerjee, S.; Kamath, H.S.

    2005-01-01

    The improved performance of present generation nuclear reactors and the realization of advanced reactor concepts, both, require development of better materials. Physical metallurgy/materials science principles which have been exploited in meeting the exacting requirements of nuclear reactor materials (fuels and structural materials), are outlined citing a few specific examples. While the incentive for improvement of traditional fuels (e.g., UO 2 fuel) is primarily for increasing the average core burn up, the development of advanced fuels (e.g., MOX, mixed carbide, nitride, silicide and dispersion fuels) are directed towards better utilization of fissile and fertile inventories through adaptation of innovative fuel cycles. As the burn up of UO 2 fuel reaches higher levels, a more detailed and quantitative understanding of the phenomena such as fission gas release, fuel restructuring induced by radiation and thermal gradients and pellet-clad interaction is being achieved. Development of zirconium based alloys for both cladding and pressure tube applications is discussed with reference to their physical metallurgy, fabrication techniques and in-reactor degradation mechanisms. The issue of radiation embrittlement of reactor pressure vessels (RPVs) is covered drawing a comparison between the western and eastern specifications of RPV steels. The search for new materials which can stand higher rates of atomic displacement due to radiation has led to the development of swelling resistant austenitic and ferritic stainless steels for fast reactor applications as exemplified by the development of the D-9 steel for Indian fast breeder reactor. The presentation will conclude by listing various materials related phenomena, which have a strong bearing on the successful development of future nuclear energy systems. (author)

  6. Introduction to the neutron kinetics of nuclear power reactors

    CERN Document Server

    Tyror, J G; Grant, P J

    2013-01-01

    An Introduction to the Neutron Kinetics of Nuclear Power Reactors introduces the reader to the neutron kinetics of nuclear power reactors. Topics covered include the neutron physics of reactor kinetics, feedback effects, water-moderated reactors, fast reactors, and methods of plant control. The reactor transients following faults are also discussed, along with the use of computers in the study of power reactor kinetics. This book is comprised of eight chapters and begins with an overview of the reactor physics characteristics of a nuclear power reactor and their influence on system design and

  7. Propulsion reactors

    International Nuclear Information System (INIS)

    Anon.

    1999-01-01

    A nuclear reactor equips the recently constructed French aircraft- carrier Charles-De-Gaulle, in a few months the second nuclear submarine (SNLE) of new generation will be operational. In last october the government launched the program Barracuda which consists of 6 submarines (SNA) whose series head will be operational in 2010. The main asset of nuclear propulsion is to allow an almost unlimited autonomy: soft water, air are produced inside the submarine and the maximum time spent underwater is only limited by human capacity to cope with confinement. CEA has 3 missions concerning country defence. First the designing, the fabrication and the maintenance of weapons, secondly the supplying of fissile materials and thirdly the nuclear propulsion. A new generation of propulsion reactors is being studied and a ground installation involving a test reactor equivalent to that on board is being built. This test reactor (RES) will simulate any type of on-board reactors by adjusting temperature, pressure, flowrate and even equipment such as steam generator. This reactor will validate the technological choices for the Barracuda program. (A.C.)

  8. Advanced Demonstration and Test Reactor Options Study

    International Nuclear Information System (INIS)

    Petti, David Andrew; Hill, R.; Gehin, J.; Gougar, Hans David; Strydom, Gerhard; Heidet, F.; Kinsey, J.; Grandy, Christopher; Qualls, A.; Brown, Nicholas; Powers, J.; Hoffman, E.; Croson, D.

    2017-01-01

    Global efforts to address climate change will require large-scale decarbonization of energy production in the United States and elsewhere. Nuclear power already provides 20% of electricity production in the United States (U.S.) and is increasing in countries undergoing rapid growth around the world. Because reliable, grid-stabilizing, low emission electricity generation, energy security, and energy resource diversity will be increasingly valued, nuclear power's share of electricity production has a potential to grow. In addition, there are non electricity applications (e.g., process heat, desalination, hydrogen production) that could be better served by advanced nuclear systems. Thus, the timely development, demonstration, and commercialization of advanced nuclear reactors could diversify the nuclear technologies available and offer attractive technology options to expand the impact of nuclear energy for electricity generation and non-electricity missions. The purpose of this planning study is to provide transparent and defensible technology options for a test and/or demonstration reactor(s) to be built to support public policy, innovation and long term commercialization within the context of the Department of Energy's (DOE's) broader commitment to pursuing an 'all of the above' clean energy strategy and associated time lines. This planning study includes identification of the key features and timing needed for advanced test or demonstration reactors to support research, development, and technology demonstration leading to the commercialization of power plants built upon these advanced reactor platforms. This planning study is consistent with the Congressional language contained within the fiscal year 2015 appropriation that directed the DOE to conduct a planning study to evaluate 'advanced reactor technology options, capabilities, and requirements within the context of national needs and public policy to support innovation in nuclear

  9. Reactors PMT backlog

    International Nuclear Information System (INIS)

    Olson, H.P.

    1986-01-01

    The overall backlog of action items within the Reactors Priority Maintenance Tracking Systems (PMT) is similar to that resulting from Reactor Incident Reports. At least 1000 open action items are being tracked (excluding procedure revisions); the exact number is not obvious because some items are being tracked in more than one tracking document. About 20 to 25% of the incomplete items are directly related to reactor safety. About 70% of these were initiated within the last two years. A few of the remaining 30% date back to the 1977-1980 time frame. The tracking systems that are in place serve particular needs but they are unconnected and do not provide a means to understand and manage the overall Reactors PMT backlog. Shortcomings are summarized. Several initiatives are in progress or planned by Reactors PMT to facilitate improved tracking of backlog items with computer software systems. Under consideration is the Integrated Living Schedule (ILS) approach to prioritize work items and schedule resources to manage the backlog in the most effective way. ILS systems are being used effectively by a number of nuclear power utilities. 6 refs., 6 figs., 1 tab

  10. Tandem mirror reactors

    International Nuclear Information System (INIS)

    Logan, B.G.; Barr, W.L.; Bender, D.J.

    1978-01-01

    We have made preliminary designs of tandem mirror fusion reactors burning D-T fuel and of fusion-fission (hybrid) tandem mirrors producing both fissile fuel and electricity. For the hybrid reactor, we find that by using stream-stabilized, 2XIIB-like plugs and by injecting 200-keV deuterium beams into a tritium-plasma target confined electrostatically in the solenoid (two-component operation), we obtain a useful Q (fusion power/injection power) near unity. The D-T tandem reactor parameters are optimized to obtain the minimum capital cost per kW(e) net. For $200/kW(e) of 1200-keV neutral beam injection power in the plugs and a solenoid cost of about $3 million per metre length, the optimum Q is near 5. To allow for more expensive injector costs, a higher D-T reactor Q of 10 is obtainable with either increased power output or decreased neutron wall loading. Fokker--Planck calculations show steady-state Q approximately 5 for D-D tandem reactors burning only deuterium fuel and its reaction products, with most of the charged-particle fusion power recovered in a direct converter

  11. Power reactors in Member States. 1979 edition

    International Nuclear Information System (INIS)

    1979-01-01

    This is the fifth issue of a periodic computer-based listing of nuclear power reactors, presenting the situation as of 1 May 1979. The basic design data for all reactors in operation, under construction, planned and shut down have been included. The following two tables are included to give a general picture of the overall situation: Table I: Reactor types and net electrical power. Table II: Reactor units and net electrical powered by country cummulated by year

  12. Fusion Reactor Materials

    International Nuclear Information System (INIS)

    Decreton, M.

    2000-01-01

    SCK-CEN's research and development programme on fusion reactor materials includes: (1) the study of the mechanical behaviour of structural materials under neutron irradiation (including steels, inconel, molybdenum, chromium); (2) the determination and modelling of the characteristics of irradiated first wall materials such as beryllium; (3) the detection of abrupt electrical degradation of insulating ceramics under high temperature and neutron irradiation; (4) the study of the dismantling and waste disposal strategy for fusion reactors.; (5) a feasibility study for the testing of blanket modules under neutron radiation. Main achievements in these topical areas in the year 1999 are summarised

  13. Fusion Reactor Materials

    Energy Technology Data Exchange (ETDEWEB)

    Decreton, M

    2000-07-01

    SCK-CEN's research and development programme on fusion reactor materials includes: (1) the study of the mechanical behaviour of structural materials under neutron irradiation (including steels, inconel, molybdenum, chromium); (2) the determination and modelling of the characteristics of irradiated first wall materials such as beryllium; (3) the detection of abrupt electrical degradation of insulating ceramics under high temperature and neutron irradiation; (4) the study of the dismantling and waste disposal strategy for fusion reactors.; (5) a feasibility study for the testing of blanket modules under neutron radiation. Main achievements in these topical areas in the year 1999 are summarised.

  14. Fuel handling grapple for nuclear reactor plants

    International Nuclear Information System (INIS)

    Rousar, D.L.

    1992-01-01

    This patent describes a fuel handling system for nuclear reactor plants. It comprises: a reactor vessel having an openable top and removable cover and containing therein, submerged in water substantially filling the reactor vessel, a fuel core including a multiplicity of fuel bundles formed of groups of sealed tube elements enclosing fissionable fuel assembled into units, the fuel handling system consisting essentially of the combination of: a fuel bundle handling platform movable over the open top of the reactor vessel; a fuel bundle handling mast extendable downward from the platform with a lower end projecting into the open top reactor vessel to the fuel core submerged in water; a grapple head mounted on the lower end of the mast provided with grapple means comprising complementary hooks which pivot inward toward each other to securely grasp a bail handle of a nuclear reactor fuel bundle and pivot backward away from each other to release a bail handle; the grapple means having a hollow cylindrical support shaft fixed within the grapple head with hollow cylindrical sleeves rotatably mounted and fixed in longitudinal axial position on the support shaft and each sleeve having complementary hooks secured thereto whereby each hook pivots with the rotation of the sleeve secured thereto; and the hollow cylindrical support shaft being provided with complementary orifices on opposite sides of its hollow cylindrical and intermediate to the sleeves mounted thereon whereby the orifices on both sides of the hollow cylindrical support shaft are vertically aligned providing a direct in-line optical viewing path downward there-through and a remote operator positioned above the grapple means can observe from overhead the area immediately below the grapple hooks

  15. Restart of R reactor at SRP

    International Nuclear Information System (INIS)

    McDonell, W.R.

    1983-01-01

    Restart of the Savannah River R-Reactor is an alternative to L-Reactor operation for increased production of defense nuclear material. R-Reactor was shut down in 1964 after 11-years operation and has been on standby for 19 years. This report presents a description of R-Reactor operation to serve as a basis for analysis of environmental impacts after restoration to meet current SRP performance standards. R-Reactor operation would differ from L-Reactor operation principally in discharge and recycle of effluent cooling water to Par Pond, rather than direct discharge to the Savannah River by way of Steel Creek. Significant differences in environmental effects could result. A costly renovation program would be required to restore R-Reactor to operability, and the reactor could not contribute to material production before about 1989

  16. Research reactor utilization, safety, decommissioning, fuel and waste management. Posters of an international conference

    International Nuclear Information System (INIS)

    2005-01-01

    For more than 50 years research reactors have played an important role in the development of nuclear science and technology. They have made significant contributions to a large number of disciplines as well as to the educational and research programmes of about 70 countries world wide. About 675 research reactors have been built to date, of which some 278 are now operating in 59 countries (86 of them in 38 developing Member States). Altogether over 13,000 reactor-years of cumulative operational experience has been gained during this remarkable period. The objective of this conference was to foster the exchange of information on current research reactor concerns related to safety, operation, utilization, decommissioning and to provide a forum for reactor operators, designers, managers, users and regulators to share experience, exchange opinions and to discuss options and priorities. The topical areas covered were: a) Utilization, including new trends and directions for utilization of research reactors. Effective management of research reactors and associated facilities. Engineering considerations and experience related to refurbishment and modifications. Strategic planning and marketing. Classical applications (nuclear activation analysis, isotope production, neutron beam applications, industrial irradiations, medical applications). Training for operators. Educational programmes using a reactor. Current developments in design and fabrication of experimental facilities. Irradiation facilities. Projects for regional uses of facilities. Core management and calculation tools. Future trends for reactors. Use of simulators for training and educational programmes. b) Safety, including experience with the preparation and review of safety analysis reports. Human factors in safety analysis. Management of extended shutdown periods. Modifications: safety analysis, regulatory aspects, commissioning programmes. Engineering safety features. Safety culture. Safety peer reviews and

  17. Recent advances in the utilization and the irradiation technology of the refurbished BR2 reactor

    International Nuclear Information System (INIS)

    Dekeyser, J.; Benoit, P.; Decloedt, C.; Pouleur, Y.; Verwimp, A.; Weber, M.; Vankeerberghen, M.; Ponsard, B.

    1999-01-01

    Operation and utilization of the materials testing reactor BR2 at the Belgian Nuclear Research Centre (SCK·CEN) has since its start in 1963 always followed closely the needs and developments of nuclear technology. In particular, a multitude of irradiation experiments have been carried out for most types of nuclear power reactors, existing or under design. Since the early 1990s and increased focus was directed towards more specific irradiation testing needs for light water reactor fuels and materials, although other areas of utilization continued as well (e.g. fusion reactor materials, safety research, ...), including also the growing activities of radioisotope production and silicon doping. An important milestone was the decision in 1994 to implement a comprehensive refurbishment programme for the BR2 reactor and plant installations. The scope of this programme comprised very substantial studies and hardware interventions, which have been completed in early 1997 within planning and budget. Directly connected to this strategic decision for reactor refurbishment was the reinforcement of our efforts to requalify and upgrade the existing irradiation facilities and to develop advanced devices in BR2 to support emerging programs in the following fields: - LWR pressure vessel steel, - LWR irradiation assisted stress corrosion cracking (IASCC), - reliability and safety of high-burnup LWR fuel, - fusion reactor materials and blanket components, - fast neutron reactor fuels and actinide burning, - extension and diversification of radioisotope production. The paper highlights these advances in the areas of BR2 utilisation and the ongoing development activities for the required new generation of irradiations devices. (author)

  18. Test reactor risk assessment methodology

    International Nuclear Information System (INIS)

    Jennings, R.H.; Rawlins, J.K.; Stewart, M.E.

    1976-04-01

    A methodology has been developed for the identification of accident initiating events and the fault modeling of systems, including common mode identification, as these methods are applied in overall test reactor risk assessment. The methods are exemplified by a determination of risks to a loss of primary coolant flow in the Engineering Test Reactor

  19. Conceptual innovations in hybrid reactors

    International Nuclear Information System (INIS)

    Greenspan, E.; Miley, G.H.

    1980-01-01

    A number of innovations in the conception of fusion-fission hybrid reactors, including the blanket, the fusion driver, the coupling of the fusion and the fission components as well as the application of hybrid reactors are described, and their feasibility assessed

  20. Test reactor risk assessment methodology

    Energy Technology Data Exchange (ETDEWEB)

    Jennings, R.H.; Rawlins, J.K.; Stewart, M.E.

    1976-04-01

    A methodology has been developed for the identification of accident initiating events and the fault modeling of systems, including common mode identification, as these methods are applied in overall test reactor risk assessment. The methods are exemplified by a determination of risks to a loss of primary coolant flow in the Engineering Test Reactor.

  1. Radioactive waste management for reactors

    International Nuclear Information System (INIS)

    Rodger, W.A.

    1974-01-01

    Radioactive waste management practices at nuclear power plants are summarized. The types of waste produced and methods for treating various types of wastes are described. The waste management systems, including simplified flow diagrams, for typical boiling water reactors and pressurized water reactors are discussed. (U.S.)

  2. Nuclear Reactor RA Safety Report, Vol. 14, Safety protection measures

    International Nuclear Information System (INIS)

    1986-11-01

    Nuclear reactor accidents can be caused by three type of errors: failure of reactor components including (1) control and measuring instrumentation, (2) errors in operation procedure, (3) natural disasters. Safety during reactor operation are secured during its design and construction and later during operation. Both construction and administrative procedures are applied to attain safe operation. Technical safety features include fission product barriers, fuel elements cladding, primary reactor components (reactor vessel, primary cooling pipes, heat exchanger in the pump), reactor building. Safety system is the system for safe reactor shutdown and auxiliary safety system. RA reactor operating regulations and instructions are administrative acts applied to avoid possible human error caused accidents [sr

  3. Nuclear Reactor RA Safety Report, Vol. 4, Reactor

    International Nuclear Information System (INIS)

    1986-11-01

    RA research reactor is thermal heavy water moderated and cooled reactor. Metal uranium 2% enriched fuel elements were used at the beginning of its operation. Since 1976, 80% enriched uranium oxide dispersed in aluminium fuel elements were gradually introduced into the core and are the only ones presently used. Reactor core is cylindrical, having diameter 40 cm and 123 cm high. Reaktor core is made up of 82 fuel elements in aluminium channels, lattice is square, lattice pitch 13 cm. Reactor vessel is cylindrical made of 8 mm thick aluminium, inside diameter 140 cm and 5.5 m high surrounded with neutron reflector and biological shield. There is no containment, the reactor building is playing the shielding role. Three pumps enable circulation of heavy water in the primary cooling circuit. Degradation of heavy water is prevented by helium cover gas. Control rods with cadmium regulate the reactor operation. There are eleven absorption rods, seven are used for long term reactivity compensation, two for automatic power regulation and two for safety shutdown. Total anti reactivity of the rods amounts to 24%. RA reactor is equipped with a number of experimental channels, 45 vertical (9 in the core), 34 in the graphite reflector and two in the water biological shield; and six horizontal channels regularly distributed in the core. This volume include detailed description of systems and components of the RA reactor, reactor core parameters, thermal hydraulics of the core, fuel elements, fuel elements handling equipment, fuel management, and experimental devices [sr

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

    International Nuclear Information System (INIS)

    Hunsbedt, A.N.

    1988-01-01

    This patent describes a sodium cooled reactor of the type having a reactor hot pool, a slightly lower pressure reactor cold pool and a reactor vessel liner defining a reactor vessel liner flow gap separating the hot pool and the cold pool along the reactor vessel sidewalls and wherein the normal sodium circuit in the reactor includes main sodium reactor coolant pumps having a suction on the lower pressure sodium cold pool and an outlet to a reactor core; the reactor core for heating the sodium and discharging the sodium to the reactor hot pool; a heat exchanger for receiving sodium from the hot pool, and removing heat from the sodium and discharging the sodium to the lower pressure cold pool; the improvement across the reactor vessel liner comprising: a jet pump having a venturi installed across the reactor vessel liner, the jet pump having a lower inlet from the reactor vessel cold pool across the reactor vessel liner and an upper outlet to the reactor vessel hot pool

  5. Fuel transfer system for a nuclear reactor

    International Nuclear Information System (INIS)

    Katz, L.R.; Marshall, J.R.; Desmarchais, W.E.

    1977-01-01

    Disclosed is a fuel transfer system for moving nuclear reactor fuel assemblies from a new fuel storage pit to a containment area containing the nuclear reactor, and for transferring spent fuel assemblies under water from the reactor to a spent fuel storage area. The system includes an underwater track which extends through a wall dividing the fuel building from the reactor containment and a car on the track serves as the vehicle for moving fuel assemblies between these two areas. The car is driven by a motor and linkage extending from an operating deck to a chain belt drive on the car. A housing pivotally mounted at its center on the car is hydraulically actuated to vertically receive a fuel assembly which then is rotated to a horizontal position to permit movement through the wall between the containment and fuel building areas. Return to the vertical position provides for fuel assembly removal and the reverse process is repeated when transferring an assembly in the opposite direction. Limit switches used in controlling operation of the system are designed to be replaced from the operating deck when necessary by tools designed for this purpose. 5 claims, 8 figures

  6. Nuclear reactors

    International Nuclear Information System (INIS)

    Prescott, R.F.; George, B.V.; Baglin, C.J.

    1978-01-01

    Reference is made to thermal insulation on the inner surfaces of containment vessels of fluid cooled nuclear reactors and particularly in situations where the thermal insulation must also serve a structural function and transmit substantial load forces to the surface which it covers. An arrangement is described that meets this requirement and also provides for core support means that favourably influences the flow of hot coolant from the lower end of the core into a plenum space in the hearth of the reactor. The arrangement comprises a course of thermally insulating bricks arranged as a mosaic covering a wall of the reactor and a course of thermally insulating tiles arranged as a mosaic covering the course of bricks. Full constructional details are given. (UK)

  7. Reactor utilization

    International Nuclear Information System (INIS)

    Zecevic, V.

    1963-01-01

    In 1962, the RA reactor was operated almost three times more than in 1961, producing total of 25 555 MWh. Diagram containing comparative data about reactor operation for 1960, 1961, and 1962, percent of fuel used and U-235 burnup shows increase in reactor operation. Number of samples irradiated was 659, number of experiments done was 16. mean powered level was 5.93 MW. Fuel was added into the core twice during the reporting year. In fact the core was increased from 56 to 68 fuel channels and later to 84 fuel channels. Fuel was added to the core when the reactivity worth decreased to the minimum operation level due to burnup. In addition to this 5 central fuel channels were exchanged with fresh fuel in february for the purpose of irradiation in the VISA-2 channel

  8. Bioconversion reactor

    Science.gov (United States)

    McCarty, Perry L.; Bachmann, Andre

    1992-01-01

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

  9. Thermonuclear reactor

    International Nuclear Information System (INIS)

    Araki, Takao; Saito, Yasushi.

    1987-01-01

    Purpose: To reduce the seismic wave responsivity of an exhaust duct shields thereby preventing the release of tritium in an evacuating device due to failures upon earthquakes. Constitution: The ends on the cutting side of upper outer exhaust duct shields of a thermonuclear reactor are connected with a plurality of support beams. In a case where seismic vibrations are exerted to such a thermonuclear reactor, since the ends on the cutting side are coupled with the support beams, vibrations of the upper outer exhaust duct shields are greatly restricted. Thus, since there is no more such a possibility, for example, that an exhaust duct connected to the upper portion of a reactor main body is greatly distorted due to the seismic response of the upper outside exhaust duct shields to result in the failure of the connection portion with a vacuum pump, the release of tritium due to failure of the evacuating device can be prevented. (Yoshino, Y.)

  10. Thermal and flow design of helium-cooled reactors

    International Nuclear Information System (INIS)

    Melese, G.; Katz, R.

    1984-01-01

    This book continues the American Nuclear Society's series of monographs on nuclear science and technology. Chapters of the book include information on the first-generation gas-cooled reactors; HTGR reactor developments; reactor core heat transfer; mechanical problems related to the primary coolant circuit; HTGR design bases; core thermal design; gas turbines; process heat HTGR reactors; GCFR reactor thermal hydraulics; and gas cooling of fusion reactors

  11. Method and apparatus for reducing the power level in a nuclear reactor during temperature transient

    International Nuclear Information System (INIS)

    1980-01-01

    Devices for moving control rods with respect to reactor cores are described. Objects of the present invention include prevention of damage to a reactor and its surroundings during a temperature excursion and especially during a loss of flow accident and reduction of the power level in a nuclear reactor in a manner that is directly proportional to the increase in temperature of the primary coolant. Preferably the present invention provides a system that produces a negative temperature coefficient of reactivity in a liquid metal fast breeder reactor and is able to reduce the power level in a nuclear reactor after a loss of flow accident to less than 15 percent of full power without the intervention of any sensing devices or control rod actuators. (U.K.)

  12. Sustainability management for operating organizations of research reactors

    International Nuclear Information System (INIS)

    Kibrit, Eduardo; Aquino, Afonso Rodrigues de

    2017-01-01

    Sustainable development is development that meets the needs of the present without compromising the ability of future generations to meet their own needs. In a country like Brazil, where nuclear activity is geared towards peaceful purposes, any operating organization of research reactor should emphasize its commitment to social, environmental, economic and institutional aspects. Social aspects include research and development, production and supply of radiopharmaceuticals, radiation safety and special training for the nuclear sector. Environmental aspects include control of the surroundings and knowledge directed towards environment preservation. Economic aspects include import substitution and diversification of production. Institutional aspects include technology, innovation and knowledge. These aspects, if considered in the management system of an operating organization of research reactor, will help with its long-term maintenance and success in an increasingly competitive market scenario. About this, we propose a sustainability management system approach for operating organizations of research reactors. A bibliographical review on the theme is made. A methodology for identifying indicators for measuring sustainability in nuclear research reactors processes is also described. Finally, we propose a methodology for sustainability perception assessment to be applied at operating organizations of research reactors. (author)

  13. Oscillatory flow chemical reactors

    Directory of Open Access Journals (Sweden)

    Slavnić Danijela S.

    2014-01-01

    Full Text Available Global market competition, increase in energy and other production costs, demands for high quality products and reduction of waste are forcing pharmaceutical, fine chemicals and biochemical industries, to search for radical solutions. One of the most effective ways to improve the overall production (cost reduction and better control of reactions is a transition from batch to continuous processes. However, the reactions of interests for the mentioned industry sectors are often slow, thus continuous tubular reactors would be impractically long for flow regimes which provide sufficient heat and mass transfer and narrow residence time distribution. The oscillatory flow reactors (OFR are newer type of tube reactors which can offer solution by providing continuous operation with approximately plug flow pattern, low shear stress rates and enhanced mass and heat transfer. These benefits are the result of very good mixing in OFR achieved by vortex generation. OFR consists of cylindrical tube containing equally spaced orifice baffles. Fluid oscillations are superimposed on a net (laminar flow. Eddies are generated when oscillating fluid collides with baffles and passes through orifices. Generation and propagation of vortices create uniform mixing in each reactor cavity (between baffles, providing an overall flow pattern which is close to plug flow. Oscillations can be created by direct action of a piston or a diaphragm on fluid (or alternatively on baffles. This article provides an overview of oscillatory flow reactor technology, its operating principles and basic design and scale - up characteristics. Further, the article reviews the key research findings in heat and mass transfer, shear stress, residence time distribution in OFR, presenting their advantages over the conventional reactors. Finally, relevant process intensification examples from pharmaceutical, polymer and biofuels industries are presented.

  14. Fusion reactor wastes

    International Nuclear Information System (INIS)

    Young, J.R.

    1976-01-01

    The fusion reactor currently is being developed as a clean source of electricity with an essentially infinite source of fuel. These reactors are visualized as using a fusion reaction to generate large quantities of high temperature energy which can be used as process heat or for the generation of electricity. The energy would be created primarily as the kinetic energy of neutrons or other reaction products. Neutron energy could be converted to high-temperature heat by moderation and capture of the neutrons. The energy of other reaction products could be converted to high-temperature heat by capture, or directly to electricity by direct conversion electrostatic equipment. An analysis to determine the wastes released as a result of operation of fusion power plants is presented

  15. Advanced boiling water reactor

    International Nuclear Information System (INIS)

    Nishimura, N.; Nakai, H.; Ross, M.A.

    1999-01-01

    In the Boiling Water Reactor (BWR) system, steam generated within the nuclear boiler is sent directly to the main turbine. This direct cycle steam delivery system enables the BWR to have a compact power generation building design. Another feature of the BWR is the inherent safety that results from the negative reactivity coefficient of the steam void in the core. Based on the significant construction and operation experience accumulated on the BWR throughout the world, the ABWR was developed to further improve the BWR characteristics and to achieve higher performance goals. The ABWR adopted 'First of a Kind' type technologies to achieve the desired performance improvements. The Reactor Internal Pump (RIP), Fine Motion Control Rod Drive (FMCRD), Reinforced Concrete Containment Vessel (RCCV), three full divisions of Emergency Core Cooling System (ECCS), integrated digital Instrumentation and Control (I and C), and a high thermal efficiency main steam turbine system were developed and introduced into the ABWR. (author)

  16. Auxiliary reactor for a hydrocarbon reforming system

    Science.gov (United States)

    Clawson, Lawrence G.; Dorson, Matthew H.; Mitchell, William L.; Nowicki, Brian J.; Bentley, Jeffrey M.; Davis, Robert; Rumsey, Jennifer W.

    2006-01-17

    An auxiliary reactor for use with a reformer reactor having at least one reaction zone, and including a burner for burning fuel and creating a heated auxiliary reactor gas stream, and heat exchanger for transferring heat from auxiliary reactor gas stream and heat transfer medium, preferably two-phase water, to reformer reaction zone. Auxiliary reactor may include first cylindrical wall defining a chamber for burning fuel and creating a heated auxiliary reactor gas stream, the chamber having an inlet end, an outlet end, a second cylindrical wall surrounding first wall and a second annular chamber there between. The reactor being configured so heated auxiliary reactor gas flows out the outlet end and into and through second annular chamber and conduit which is disposed in second annular chamber, the conduit adapted to carry heat transfer medium and being connectable to reformer reaction zone for additional heat exchange.

  17. Gas Reactor International Cooperative Program. Interim report. Construction and operating experience of selected European Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    1978-09-01

    The construction and operating experience of selected European Gas-Cooled Reactors is summarized along with technical descriptions of the plants. Included in the report are the AVR Experimental Pebble Bed Reactor, the Dragon Reactor, AGR Reactors, and the Thorium High Temperature Reactor (THTR). The study demonstrates that the European experience has been favorable and forms a good foundation for the development of Advanced High Temperature Reactors

  18. Advances by the Integral Fast Reactor Program

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  19. Lateral restraint assembly in a nuclear reactor

    International Nuclear Information System (INIS)

    Brown, S.J.; Gorholt, W.

    1977-01-01

    A lateral restraint assembly is described for a reactor of, for example, the high temperature gas-cooled type which commonly includes a reactor core of relatively complex construction supported within a shell or vessel providing a shielded cavity for containing the reactor core. (U.K.)

  20. Reactor Safety Research: Semiannual report, July-December 1986

    International Nuclear Information System (INIS)

    1987-11-01

    Sandia National Laboratories is conducting, under USNRC sponsorship, phenomenological research related to the safety of commercial nuclear power reactors. The research includes experiments to simulate the phenomenology of the accident conditions and the development of analytical models, verified by experiment, which can be used to predict reactor and safety systems performance and behavior under abnormal conditions. The objective of this work is to provide NRC requisite data bases and analytical methods to (1) identify and define safety issues, (2) understand the progression of risk-significant accident sequences, and (3) conduct safety assessments. The collective NRC-sponsored effort at Sandia National Laboratories is directed at enhancing the tehcnology base supporting licensing decisions

  1. Reactor Safety Research: Semiannual report, July-December 1986

    Energy Technology Data Exchange (ETDEWEB)

    1987-11-01

    Sandia National Laboratories is conducting, under USNRC sponsorship, phenomenological research related to the safety of commercial nuclear power reactors. The research includes experiments to simulate the phenomenology of the accident conditions and the development of analytical models, verified by experiment, which can be used to predict reactor and safety systems performance and behavior under abnormal conditions. The objective of this work is to provide NRC requisite data bases and analytical methods to (1) identify and define safety issues, (2) understand the progression of risk-significant accident sequences, and (3) conduct safety assessments. The collective NRC-sponsored effort at Sandia National Laboratories is directed at enhancing the tehcnology base supporting licensing decisions.

  2. Axial and Radial Gas Holdup in Bubble Column Reactor

    International Nuclear Information System (INIS)

    Wagh, Sameer M.; Ansari, Mohashin E Alan; Kene, Pragati T.

    2014-01-01

    Bubble column reactors are considered the reactor of choice for numerous applications including oxidation, hydrogenation, waste water treatment, and Fischer-Tropsch (FT) synthesis. They are widely used in a variety of industrial applications for carrying out gas-liquid and gas-liquid-solid reactions. In this paper, the computational fluid dynamics (CFD) model is used for predicting the gas holdup and its distribution along radial and axial direction are presented. Gas holdup increases linearly with increase in gas velocity. Gas bubbles tends to concentrate more towards the center of the column and follows a wavy path

  3. ARIES tokamak reactor study

    International Nuclear Information System (INIS)

    Steiner, D.; Embrechts, M.

    1990-07-01

    This is a status report on technical progress relative to the tasks identified for the fifth year of Grant No. FG02-85-ER52118. The ARIES tokamak reactor study is a multi-institutional effort to develop several visions of the tokamak as an attractive fusion reactor with enhanced economic, safety, and environmental features. The ARIES study is being coordinated by UCLA and involves a number of institutions, including RPI. The RPI group has been pursuing the following areas of research in the context of the ARIES-I design effort: MHD equilibrium and stability analyses; plasma-edge modeling and blanket materials issues. Progress in these areas is summarized herein

  4. Particle bed reactor modeling

    Science.gov (United States)

    Sapyta, Joe; Reid, Hank; Walton, Lew

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

  5. Fast quench reactor method

    Science.gov (United States)

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

    1999-08-10

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

  6. Decommissioning activities for Salaspils research reactor - 59055

    International Nuclear Information System (INIS)

    Abramenkovs, A.; Malnacs, J.

    2012-01-01

    In May 1995, the Latvian government decided to shut down the Salaspils Research Reactor (SRR). The reactor is out of operation since July 1998. A conceptual study for the decommissioning of SRR has been carried out by Noell-KRC-Energie- und Umwelttechnik GmbH at 1998-1999. The Latvian government decided to start the direct dismantling to 'green field' in October 26, 1999. The upgrade of decommissioning and dismantling plan was performed in 2003-2004 years, which change the main goal of decommissioning to the 'brown field'. The paper deals with the SRR decommissioning experience during 1999-2010. The main decommissioning stages are discussed including spent fuel and radioactive wastes management. The legal aspects and procedures for decommissioning of SRR are described in the paper. It was found, that the involvement of stakeholders at the early stages significantly promotes the decommissioning of nuclear facility. Radioactive waste management's main efforts were devoted to collecting and conditioning of 'historical' radioactive wastes from different storages outside and inside of reactor hall. All radioactive materials (more than 96 tons) were conditioned in concrete containers for disposal in the radioactive wastes repository 'Radons' at Baldone site. The dismantling of contaminated and activated components of SRR systems is discussed in paper. The cementation of dismantled radioactive wastes in concrete containers is discussed. Infrastructure of SRR, including personal protective and radiation measurement equipment, for decommissioning purposes was upgraded significantly. Additional attention was devoted to the free release measurement's technique. The certified laboratory was installed for supporting of all decommissioning activities. All non-radioactive equipments and materials outside of reactor buildings were released for clearance and dismantled for reusing or conventional disposing. Weakly contaminated materials from reactor hall were collected

  7. Radiation detectors for reactors

    International Nuclear Information System (INIS)

    Balagi, V.

    2005-01-01

    Detection and measurement of radiation plays a vital role in nuclear reactors from the point of view of control and safety, personnel protection and process control applications. Various types of radiation are measured over a wide range of intensity. Consequently a variety of detectors find use in nuclear reactors. Some of these devices have been developed in Electronics Division. They include gas-filled detectors such as 10 B-lined proportional counters and chambers, fission detectors and BF 3 counters are used for the measurement of neutron flux both for reactor control and safety, process control as well as health physics instrumentation. In-core neutron flux instrumentation employs the use detectors such as miniature fission detectors and self-powered detectors. In this development effort, several indigenous materials, technologies and innovations have been employed to suit the specific requirement of nuclear reactor applications. This has particular significance in view of the fact that several new types of reactors such as P-4, PWR and AHWR critical facilities, FBTR, PFBR as well as the refurbishment of old units like CIRUS are being developed. The development work has sought to overcome some difficulties associated with the non-availability of isotopically enriched neutron-sensing materials, achieving all-welded construction etc. The present paper describes some of these innovations and performance results. (author)

  8. Study of advanced fission power reactor development for the United States. Volume II

    International Nuclear Information System (INIS)

    1976-01-01

    This report presents the results of a multi-phase research study which had as its objective the comparative study of various advanced fission reactors and evaluation of alternate strategies for their development in the USA through the year 2020. By direction from NSF, ''advanced'' reactors were defined as those which met the dual requirements of (1) offering a significant improvement in fissile fuel utilization as compared to light-water reactors and (2) currently receiving U.S. Government funding. (A detailed study of the LMFBR was specifically excluded, but cursory baseline data were obtained from ERDA sources.) Included initially were the High-Temperature Gas-Cooled Reactor (HTGR), Gas-Cooled Fast Reactor (GCFR), Molten Salt Reactor (MSR), and Light-Water Breeder Reactor (LWBR). Subsequently, the CANDU Heavy Water Reactor (HWR) was included for comparison due to increased interest in its potential. This volume presents the reasoning process and analytical methods utilized to arrive at the conclusions for the overall study

  9. Neutronic/Thermalhydraulic Coupling Technigues for Sodium Cooled Fast Reactor Simulations

    Energy Technology Data Exchange (ETDEWEB)

    Jean Ragusa; Andrew Siegel; Jean-Michel Ruggieri

    2010-09-28

    The objective of this project was to test new coupling algorithms and enable efficient and scalable multi-physics simulations of advanced nuclear reactors, with considerations regarding the implementation of such algorithms in massively parallel environments. Numerical tests were carried out to verify the proposed approach and the examples included some reactor transients. The project was directly related to the Sodium Fast Reactor program element of the Generation IV Nuclear Energy Systems Initiative and the Advanced Fuel cycle Initiative, and, supported the requirement of high-fidelity simulation as a mean of achieving the goals of the presidential Global Nuclear Energy Partnership (GNEP) vision.

  10. Neutronic/Thermal-hydraulic Coupling Technigues for Sodium Cooled Fast Reactor Simulations

    International Nuclear Information System (INIS)

    Ragusa, Jean; Siegel, Andrew; Ruggieri, Jean-Michel

    2010-01-01

    The objective of this project was to test new coupling algorithms and enable efficient and scalable multi-physics simulations of advanced nuclear reactors, with considerations regarding the implementation of such algorithms in massively parallel environments. Numerical tests were carried out to verify the proposed approach and the examples included some reactor transients. The project was directly related to the Sodium Fast Reactor program element of the Generation IV Nuclear Energy Systems Initiative and the Advanced Fuel cycle Initiative, and, supported the requirement of high-fidelity simulation as a mean of achieving the goals of the presidential Global Nuclear Energy Partnership (GNEP) vision.

  11. Refueling of nuclear reactors

    International Nuclear Information System (INIS)

    Meuschke, R.E.

    1987-01-01

    This patent describes the unrodded refueling of a nuclear reactor having fuel assemblies and upper internals with apparatus including a lifting rig and a lift plate. The upper internals of the reactor are secured to the lifting rig. A method is given of reinserting in the fuel assemblies of the reactor the rods which penetrate into the fuel assemblies, such as control rods and/or coolant-displacement rods. The penetrating rods are connected to drive rods, the drive rods and penetrating rods being suspended from the lift plate, the lift plate and the drive rods and penetrating rods suspended therefrom being supported on a removable support in an upper position on the lifting rig

  12. BWR type reactor system

    International Nuclear Information System (INIS)

    Morooka, Shin-ichi.

    1980-01-01

    Purpose: To reduce the internal structure in a reactor by rapidly and efficiently transferring heat generated in a reactor core out of the reactor and eliminating the danger of radiation exposure. Constitution: Steam generated in a pressure vessel is introduced into heat pipe group by inserting the heat pipe group into the steam dome of the pressure vessel. The introduced steam is condensed in the heat pipes to transfer the heat of the steam to the heat pipe group. The transferred heat is transmitted to a heat exchanger provided out of a containment vessel to generate steam to operate a turbine. Thus, it is not necessary to introduce the steam including radioactive substance externally and can remove only the heat so as to carry out the desired purpose. (Kamimura, M.)

  13. IAEA activities related to research reactor fuel conversion and spent fuel return programmes

    International Nuclear Information System (INIS)

    Ritchie, I.G.; Adelfang, P.; Goldman, I.N.

    2004-01-01

    Full text: The IAEA has been involved for more than twenty years in supporting international nuclear non-proliferation efforts associated with reducing the amount of highly enriched uranium (HEU) in international commerce. IAEA projects and activities have directly supported the Reduced Enrichment for Research and Test Reactors (RERTR) programme, as well as directly associated efforts to return research reactor fuel to the country of origin where it was originally enriched. IAEA efforts have included the development and maintenance of several data bases with information related to research reactors and research reactor spent fuel inventories that have been essential in planning and managing both RERTR and spent fuel return programmes. Other IAEA regular budget programmes have supported research reactor fuel conversion from HEU to low enriched uranium, and in addressing issues common to many member states with spent fuel management problems and concerns. The paper briefly describes IAEA involvement since the early 1980's in these areas, including regular budget and Technical Co-operation programme activities, and focuses on efforts in the past five years to continue to support and accelerate U.S. and Russian research reactor spent fuel return programmes. It is hoped that an announcement of the extension of the U.S. Acceptance Programme, which is expected in the very near future, will facilitate the life extensions of many productive TRIGA reactors around the world. (author)

  14. Nuclear reactor

    International Nuclear Information System (INIS)

    Gilroy, J.E.

    1980-01-01

    An improved cover structure for liquid metal cooled fast breeder type reactors is described which it is claimed reduces the temperature differential across the intermediate grid plate of the core cover structure and thereby reduces its subjection to thermal stresses. (UK)

  15. Neutronic reactor

    International Nuclear Information System (INIS)

    Wende, C.W.J.

    1976-01-01

    The method of operating a water-cooled neutronic reactor having a graphite moderator is described which comprises flowing a gaseous mixture of carbon dioxide and helium, in which the helium comprises 40--60 volume percent of the mixture, in contact with the graphite moderator. 2 claims, 4 figures

  16. Neutronic reactor

    International Nuclear Information System (INIS)

    Wende, C.W.J.

    1976-01-01

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

  17. Comparison between TRU burning reactors and commercial fast reactor

    International Nuclear Information System (INIS)

    Fujimura, Koji; Sanda, Toshio; Ogawa, Takashi

    2001-03-01

    Research and development for stabilizing or shortening the radioactive wastes including in spent nuclear fuel are widely conducted in view point of reducing the environmental impact. Especially it is effective way to irradiate and transmute long-lived TRU by fast reactors. Two types of loading way were previously proposed. The former is loading relatively small amount of TRU in all commercial fast reactors and the latter is loading large amount of TRU in a few TRU burning reactors. This study has been intended to contribute to the feasibility studies on commercialized fast reactor cycle system. The transmutation and nuclear characteristics of TRU burning reactors were evaluated and compared with those of conventional transmutation system using commercial type fast reactor based upon the investigation of technical information about TRU burning reactors. Major results are summarized as follows. (1) Investigation of technical information about TRU burning reactors. Based on published reports and papers, technical information about TRU burning reactor concepts transmutation system using convectional commercial type fast reactors were investigated. Transmutation and nuclear characteristics or R and D issue were investigated based on these results. Homogeneously loading of about 5 wt% MAs on core fuels in the conventional commercial type fast reactor may not cause significant impact on the nuclear core characteristics. Transmutation of MAs being produced in about five fast reactors generating the same output is feasible. The helium cooled MA burning fast reactor core concept propose by JAERI attains criticality using particle type nitride fuels which contain more than 60 wt% MA. This reactor could transmute MAs being produced in more than ten 1000 MWe-LWRs. Ultra-long life core concepts attaining more than 30 years operation without refueling by utilizing MA's nuclear characteristics as burnable absorber and fertile nuclides were proposed. Those were pointed out that

  18. Guidance for the application of an assessment methodology for innovative nuclear energy systems. INPRO manual - Overview of the methodology. Vol. 1 of 9 of the final report of phase 1 of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) including a CD-ROM comprising all volumes

    International Nuclear Information System (INIS)

    2008-11-01

    The International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) was initiated in the year 2000, based on a resolution of the IAEA General Conference (GC(44)/RES/21). The main objectives of INPRO are (1) to help to ensure that nuclear energy is available to contribute in fulfilling energy needs in the 21st century in a sustainable manner, (2) to bring together both technology holders and technology users to consider jointly the international and national actions required to achieve desired innovations in nuclear reactors and fuel cycles; and (3) to create a forum to involve all relevant stakeholders that will have an impact on, draw from, and complement the activities of existing institutions, as well as ongoing initiatives at the national and international level. This document follows the guidelines of the INPRO report 'Methodology for the assessment of innovative nuclear reactors and fuel cycles, Report of Phase 1B (first part) of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO)', IAEA-TECDOC-1434 (2004), together with its previous report Guidance for the evaluation for innovative nuclear reactors and fuel cycles, Report of Phase 1A of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO), IAEA-TECDOC-1362 (2003). This INPRO manual is comprised of an overview volume (laid out in this report), and eight additional volumes (available on a CD-ROM attached to the inside back cover of this report) covering the areas of economics (Volume 2), infrastructure (Volume 3), waste management (Volume 4), proliferation resistance (Volume 5), physical protection (Volume 6), environment (Volume 7), safety of reactors (Volume 8), and safety of nuclear fuel cycle facilities (Volume 9). The overview volume sets out the philosophy of INPRO and a general discussion of the INPRO methodology. This overview volume discusses the relationship of INPRO with the UN concept of sustainability to demonstrate how the

  19. BR2 Reactor: Introduction

    International Nuclear Information System (INIS)

    Moons, F.

    2007-01-01

    The irradiations in the BR2 reactor are in collaboration with or at the request of third parties such as the European Commission, the IAEA, research centres and utilities, reactor vendors or fuel manufacturers. The reactor also contributes significantly to the production of radioisotopes for medical and industrial applications, to neutron silicon doping for the semiconductor industry and to scientific irradiations for universities. Along the ongoing programmes on fuel and materials development, several new irradiation devices are in use or in design. Amongst others a loop providing enhanced cooling for novel materials testing reactor fuel, a device for high temperature gas cooled fuel as well as a rig for the irradiation of metallurgical samples in a Pb-Bi environment. A full scale 3-D heterogeneous model of BR2 is available. The model describes the real hyperbolic arrangement of the reactor and includes the detailed 3-D space dependent distribution of the isotopic fuel depletion in the fuel elements. The model is validated on the reactivity measurements of several tens of BR2 operation cycles. The accurate calculations of the axial and radial distributions of the poisoning of the beryllium matrix by 3 He, 6 Li and 3T are verified on the measured reactivity losses used to predict the reactivity behavior for the coming decades. The model calculates the main functionals in reactor physics like: conventional thermal and equivalent fission neutron fluxes, number of displacements per atom, fission rate, thermal power characteristics as heat flux and linear power density, neutron/gamma heating, determination of the fission energy deposited in fuel plates/rods, neutron multiplication factor and fuel burn-up. For each reactor irradiation project, a detailed geometry model of the experimental device and of its neighborhood is developed. Neutron fluxes are predicted within approximately 10 percent in comparison with the dosimetry measurements. Fission rate, heat flux and

  20. LMFBR type reactor

    International Nuclear Information System (INIS)

    Kumaoka, Yoshio; Kawamura, Yutaka.

    1990-01-01

    A main vessel support skirt is supported by a base mat of reactor buildings and the base mat is supported by means of an earthquake-proof device on concretes of a lower raft disposed on ground rocks. The earthquake-proof device is constituted by alternately stacking, laminating and press-bonding thin steel plates together with thin rubber layers between an upper flange and a lower raft. Thus, for the horizontal seismic vibrations, the period of the swinging in the horizontal direction of the buildings is made greater than the swinging period of earthquakes by the earthquake-proof device to reduce the impact shocks of earthquakes. Further, for the vertical seismic vibrations, the input seismic movements are not amplified during transmission from the base mat of the buildings to the reactor structure by way of the support skirt of the main vessel, due to the shortened load transmission path and the seismic power design to the reactor structure can be moderated sufficiently. A safety LMFBR type reactor with reduced construction cost and improved reliability can be attained. (N.H.)

  1. Nuclear power reactor physics

    International Nuclear Information System (INIS)

    Barjon, Robert

    1975-01-01

    The purpose of this book is to explain the physical working conditions of nuclear reactors for the benefit of non-specialized engineers and engineering students. One of the leading ideas of this course is to distinguish between two fundamentally different concepts: - a science which could be called neutrodynamics (as distinct from neutron physics which covers the knowledge of the neutron considered as an elementary particle and the study of its interactions with nuclei); the aim of this science is to study the interaction of the neutron gas with real material media; the introduction will however be restricted to its simplified expression, the theory and equation of diffusion; - a special application: reactor physics, which is introduced when the diffusing and absorbing material medium is also multiplying. For this reason the chapter on fission is used to introduce this section. In practice the section on reactor physics is much longer than that devoted to neutrodynamics and it is developed in what seemed to be the most relevant direction: nuclear power reactors. Every effort was made to meet the following three requirements: to define the physical bases of neutron interaction with different materials, to give a correct mathematical treatment within the limit of necessary simplifying hypotheses clearly explained; to propose, whenever possible, numerical applications in order to fix orders of magnitude [fr

  2. Research Reactors Coalitions

    International Nuclear Information System (INIS)

    Borio, Andrea; Bradley, E.; Vyshniauskas, Judy; Ridikas, Danas

    2013-01-01

    When considering the potential role of an existing RR or possibly the construction of a new RR, it is clear that a nuclear science and technology programme (including nuclear power) could benefit provided the RR is safely and competently managed, well utilised and adequately funded. Based on MSs experience, a domestic RR may not be required to develop a nuclear power programme, provided the decision takes advantage of foreign expertise, including access to foreign RRs facilities and RRs regional/international networks. If a country decides to gain access to a foreign research reactor, it may need considering the potential risk of change in the political relationship with the host country that could compromise the achievement of its national relevant objectives. This risk may be offset by availability of many options within one or more regional/international RRs networks and coalitions. Examples include the use of existing RRs in vendor, non-vendor countries and, in some cases non-nuclear power countries, to develop human resources in support of the introduction of nuclear power elsewhere. International RR networking trends are most evident with high flux, higher capability, and more complex fuel and material testing RRs being shared through international partnerships. However, networks involving low-medium power RRs for education and training purposes are also gaining a more prominent role to support nuclear capacity building in newcomer MSs. Networking through the internet seems also to be a promising way to support, as complementary offer to direct access to RRs facilities, MSs nuclear capacity building objectives (e.g. the IRL project)

  3. High Temperature reactors status 1977

    International Nuclear Information System (INIS)

    Ekholm, R.; Bosaeus, J.; Carleson, G.; Gelin, R.; Jirlow, K.; Linder, S.; Menon, S.; Runfors, U.; Vieider, G.

    1978-03-01

    The objective of this report is to summarize the current state-of-the-art of HTR technology as part of follow-up studies of the development of advanced fission reactor systems. These studies have been performed at AB Atomenergi since fiscal year 1975/76 and are financed by governmental funds for energy R and D. In this report emphasis is given to the following main aspects of the HTR development: - a survey of the major HTR - R and D programmes; - the description of HTR technology including remaining development problems and uncertainties; - the analysis of the safety and environmental characteristics of the HTR systems; - the analysis of the incentives for the introduction of various HTR types. The report contains also information kindly provided directly by experts from several organisations developing the HTR-systems

  4. Nuclear reactor with control rods

    International Nuclear Information System (INIS)

    Obermeyer, F.D.; Berringer, R.T.

    1979-01-01

    A liquid-cooled nuclear reactor including fuel assemblies mounted within a reactor vessel having linearly movable control rods passing through control rod guide tubes into respective aligned fuel assemblies is described. Reactor coolant circulates through the assemblies. Guide tubes and other vessel internals structures located above the assemblies and is discharged through an outlet nozzle positioned above the elevation of primary flow openings in the guide tube walls. The guide tube includes internal horizontal supports and a length limited continuous control rod guide which, in conjunction with the flow openings, alleviate detrimental coolant cross flows and frictional restraints imposed upon the control rods

  5. Development of safety analysis technology for integral reactor; evaluation on safety concerns of integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hee Chul; Kim, Woong Sik; Lee, J. H. [Korea Institute of Nuclear Safety, Taejeon (Korea)

    2002-03-01

    The Nuclear Desalination Plant (NDP) is being developed to produce electricity and fresh water, and is expected to locate near population zone. In the aspect of safety, it is required to protect the public and environment from the possible releases of fission products and to prevent the fresh water from the contamination of radioactivity. Thus, in this study, the safety characteristics of the integral reactor adopting passive and inherent safety features significantly different from existing nuclear power plants were investigated. Also, safety requirements applicable to the NDP were analyzed based on the regulatory requirements for current light water reactor and advanced reactor designs, and user requirements for small-medium size reactors. Based on these analyses, some safety concerns to be considered in the design stage have been identified and discussed. They include the use of proven technology for new safety features, systematic event classification and selection, strengthening containment function, and the safety impacts on desalination-related systems. The study presents the general safety requirements applicable to licensing of an integral reactor and suggests additional regulatory requirements, which need to be developed, based on the direction to resolution of the safety concerns. The efforts to identify and technically resolve the safety concerns in the design stage will provide the early confidence of SMART safety and the technical basis to evaluate the safety to designers and reviewers in the future. Suggestion on the development of additional regulatory requirements will contribute for the regulator to taking actions for licensing of an integral reactor. 66 refs., 5 figs., 24 tabs. (Author)

  6. Conceptual design of RFC reactor

    International Nuclear Information System (INIS)

    Kumazawa, R.; Adati, K.; Hatori, T.; Ichimura, M.; Obayashi, H.; Okamura, S.; Sato, T.; Watari, T.; Emmert, G.A.

    1982-01-01

    A parametic analysis and a preliminary conceptual design for RFC reactor (including cusp field) with and without alpha particle heating are described. Steady state operations can be obtained for various RF ponderomotive potential in cases of alpha particle heating. (author)

  7. Reactor operation environmental information document

    Energy Technology Data Exchange (ETDEWEB)

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

    1989-12-01

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

  8. RA reactor operation and maintenance

    International Nuclear Information System (INIS)

    Zecevic, V.

    1963-02-01

    This volume includes the final report on RA reactor operation and utilization of the experimental facilities in 1962, detailed analysis of the system for heavy water distillation and calibration of the system for measuring the activity of the air

  9. Broad-Application Test Reactor

    International Nuclear Information System (INIS)

    Motloch, C.G.

    1992-05-01

    This report is about a new, safe, and operationally efficient DOE reactor of nuclear research and testing proposed for the early to mid- 21st Century. Dubbed the Broad-Application Test Reactor (BATR), the proposed facility incorporates a multiple-application, multiple-mission design to support DOE programs such as naval reactors and space power and propulsion, as well as research in medical, science, isotope, and electronics arenas. DOE research reactors are aging, and implementing major replacement projects requires long lead times. Primary design drivers include safety, low risk, minimum operation cost, mission flexibility, waste minimization, and long life. Scientists and engineers at the Idaho National Engineering Laboratory are evaluating possible fuel forms, structural materials, reactor geometries, coolants, and moderators

  10. AREVA's nuclear reactors portfolio

    International Nuclear Information System (INIS)

    Marincic, A.

    2009-01-01

    A reasonable assumption for the estimated new build market for the next 25 years is over 340 GWe net. The number of prospect countries is growing almost each day. To address this new build market, AREVA is developing a comprehensive portfolio of reactors intended to meet a wide range of power requirements and of technology choices. The EPR reactor is the flagship of the fleet. Intended for large power requirements, the four first EPRs are being built in Finland, France and China. Other countries and customers are in view, citing just two examples: the Usa where the U.S. EPR has been selected as the technology of choice by several U.S utilities; and the United Kingdom where the Generic Design Acceptance process of the EPR design submitted by AREVA and EDF is well under way, and where there is a strong will to have a plant on line in 2017. For medium power ranges, the AREVA portfolio includes a boiling water reactor and a pressurized water reactor which both offer all of the advantages of an advanced plant design, with excellent safety performance and competitive power generation cost: -) KERENA (1250+ MWe), developed in collaboration with several European utilities, and in particular with Eon; -) ATMEA 1 (1100+ MWe), a 3-loop evolutionary PWR which is being developed by AREVA and Mitsubishi. AREVA is also preparing the future and is deeply involved into Gen IV concepts. It has developed the ANTARES modular HTR reactor (pre-conceptual design completed) and is building upon its vast Sodium Fast Reactor experience to take part into the development of the next prototype. (author)

  11. Nuclear Reactors and Technology

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-01-01

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

  12. Nuclear reactor

    International Nuclear Information System (INIS)

    Gibbons, J.F.; McLaughlin, D.J.

    1978-01-01

    In the pressure vessel of the water-cooled nuclear reactor there is provided an internal flange on which the one- or two-part core barrel is hanging by means of an external flange. A cylinder is extending from the reactor vessel closure downwards to a seat on the core cupport structure and serves as compression element for the transmission of the clamping load from the closure head to the core barrel (upper guide structure). With the core barrel, subject to tensile stress, between the vessel internal flange and its seat on one hand and the compression of the cylinder resp. hold-down element between the closure head and the seat on the other a very strong, elastic sprung structure is obtained. (DG) [de

  13. Nuclear reactor

    International Nuclear Information System (INIS)

    Aleite, W.; Bock, H.W.; Struensee, S.

    1976-01-01

    The invention concerns the use of burnable poisons in a nuclear reactor, especially in PWRs, in order to improve the controllability of the reactor. An unsymmetrical arrangement in the lattice is provided, if necessary also by insertion of special rods for these additions. It is proposed to arrange the burnable poisons in fuel elements taken over from a previous burn-up cycle and to distribute them, going out from the side facing the control rods, over not more than 20% of the lenth of the fuel elements. It seems sufficient, for the burnable poisons to bind an initial reactivity of only 0.1% and to become ineffective after normal operation of 3 to 4 months. (ORU) [de

  14. Reactor container

    International Nuclear Information System (INIS)

    Otsuka, Hiroaki; Yoshida, Takashi.

    1979-01-01

    Purpose: To prevent rain water falling along the outer wall of the container during the construction work of an atomic power plant from making ingress into the inner part of a reactor container through a large size material carry-in port. Constitution: A weir for preventing the ingress of rain water is provided on the border between the foot floor of a large material carry-in port provided on the side surface at the bottom part of the reactor container and the floor surface of the building. This weir is of a semi-circular plate shape, and formed so that the lower semi-circular part of the carry-in port is tightly closed. (Kamimura, M.)

  15. Reactor container

    International Nuclear Information System (INIS)

    Ichiki, Tadaharu; Saba, Kazuhisa.

    1979-01-01

    Purpose: To improve the earthquake resistance as well as reduce the size of a container for a nuclear reactor with no adverse effects on the decrease of impact shock to the container and shortening of construction step. Constitution: Reinforcing profile steel materials are welded longitudinally and transversely to the inner surface of a container, and inner steel plates are secured to the above profile steel materials while keeping a gap between the materials and the container. Reactor shielding wall planted to the base concrete of the container is mounted to the pressure vessel, and main steam pipeways secured by the transverse beams and led to the outside of container is connected. This can improve the rigidity earthquake strength and the safetiness against the increase in the inside pressure upon failures of the container. (Yoshino, Y.)

  16. The flow measurement methods for the primary system of integral reactors

    International Nuclear Information System (INIS)

    Lee, J.; Seo, J. K.; Lee, D. J.

    2001-01-01

    It is the common features of the integral reactors that the main components of the primary system are installed within the reactor vessel, and so there are no any flow pipes connecting the reactor coolant pumps or steam generators. Due to no any flow pipes, it is impossible to measure the differential pressure at the primary system of the integral reactors, and it also makes impossible measure the primary coolant flow rate. The objective of the study is to draw up the flow measurement methods for the primary system of integral reactors. As a result of the review, we have made a selection of the flow measurement method by pump speed, bt HBM, and by pump motor power as the flow measurement methods for the primary system of integral reactors. Peculiarly, we did not found out a precedent which the direct pump motor power-flow rate curve is used as the flow measurement method in the existing commercial nuclear power reactors. Therefore, to use this method for integral reactors, it is needed to bear the follow-up measures in mind. The follow-up measures is included in this report

  17. An Innovative Hybrid Loop-Pool Design for Sodium Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Haihua Zhao; Hongbin Zhang

    2007-01-01

    The existing sodium cooled fast reactors (SFR) have two types of designs--loop type and pool type. In the loop type design, such as JOYO (Japan) [1] and MONJU (Japan), the primary coolant is circulated through intermediate heat exchangers (IHX) external to the reactor tank. The major advantages of loop design include compactness and easy maintenance. The disadvantage is higher possibility of sodium leakage. In the pool type design such as EBR-II (USA), BN-600M(Russia), Superphenix (France) and European Fast Reactor [2], the reactor core, primary pumps, IHXs and direct reactor auxiliary cooling system (DRACS) heat exchangers (DHX) all are immersed in a pool of sodium coolant within the reactor vessel, making a loss of primary coolant extremely unlikely. However, the pool type design makes primary system large. In the latest ANL's Advanced Burner Test Reactor (ABTR) design [3], the primary system is configured in a pool-type arrangement. The hot sodium at core outlet temperature in hot pool is separated from the cold sodium at core inlet temperature in cold pool by a single integrated structure called Redan. Redan provides the exchange of the hot sodium from hot pool to cold pool through IHXs. The IHXs were chosen as the traditional tube-shell design. This type of IHXs is large in size and hence large reactor vessel is needed

  18. BWR type reactor core

    International Nuclear Information System (INIS)

    Tatemichi, Shin-ichiro.

    1981-01-01

    Purpose: To eliminate the variation in the power distribution of a BWR type reactor core in the axial direction even if the flow rate is increased or decreased by providing a difference in the void coefficient between the upper part and the lower parts of the reactor core, and increasing the void coefficient at the lower part of the reactor core. Constitution: The void coefficient of the lower region from the center to the lower part along the axial direction of a nuclear fuel assembly is increased to decrease the dependence on the flow rate of the axial power distribution of the nuclear fuel assembly. That is, a water/fuel ratio is varied, the water in non-boiled region is increased or the neutron spectrum is varied so as to vary the void coefficient. In order to exemplify it, the rate of the internal pellets of the fuel rod of the nuclear fuel assembly or the shape of the channel box is varied. Accordingly, the power does not considerably vary even if the flow rate is altered since the power is varied in the power operation. (Yoshihara, H.)

  19. Nuclear reactor

    International Nuclear Information System (INIS)

    Shirakawa, Toshihisa.

    1979-01-01

    Purpose: To prevent cladding tube injuries due to thermal expansion of each of the pellets by successively extracting each of the control rods loaded in the reactor core from those having less number of notches, as well as facilitate the handling work for the control rods. Constitution: A recycle flow control device is provided to a circulation pump for forcibly circulating coolants in the reactor container and an operational device is provided for receiving each of the signals concerning number of notches for each of the control rods and flow control depending on the xenon poisoning effect obtained from the signals derived from the in-core instrument system connected to the reactor core. The operational device is connected with a control rod drive for moving each of the control rods up and down and a recycle flow control device. The operational device is set with a pattern for the aimed control rod power and the sequence of extraction. Upon extraction of the control rods, they are extracted successively from those having less notch numbers. (Moriyama, K.)

  20. Reactor building

    International Nuclear Information System (INIS)

    Ebata, Sakae.

    1990-01-01

    At least one valve rack is disposed in a reactor building, on which pipeways to a main closure valve, valves and bypasses of turbines are placed and contained. The valve rack is fixed to the main body of the building or to a base mat. Since the reactor building is designed as class A earthquake-proofness and for maintaining the S 1 function, the valve rack can be fixed to the building main body or to the base mat. With such a constitution, the portions for maintaining the S 1 function are concentrated to the reactor building. As a result, the dispersion of structures of earthquake-proof portion corresponding to the reference earthquake vibration S 1 can be prevented. Accordingly, the conditions for the earthquake-proof design of the turbine building and the turbine/electric generator supporting rack are defined as only the class B earthquake-proof design conditions. In view of the above, the amount of building materials can be saved and the time for construction can be shortened. (I.S.)

  1. Heat-pipe thermionic reactor concept

    DEFF Research Database (Denmark)

    Storm Pedersen, E.

    1967-01-01

    Main components are reactor core, heat pipe, thermionic converter, secondary cooling system, and waste heat radiator; thermal power generated in reactor core is transported by heat pipes to thermionic converters located outside reactor core behind radiation shield; thermionic emitters are in direct...... contact with outside envelope of heat pipes and collectors are in contact with liquid metal secondary cooling system that transfers waste heat to radiator....

  2. Safety and licensing for small and medium power reactors

    International Nuclear Information System (INIS)

    Trauger, D.B.

    1988-01-01

    Proposed new concepts for small and medium power reactors differ substantially from traditional Light Water Reactors (LWRs). Although designers have a large base of experience in safety and licensing, much of it is not relevant to new concepts. It can be a disadvantage if regulators apply LWR rules directly. A fresh start is appropriate. The extensive interactions between industry, regulators, and the public complicate but may enhance safety. It is basic to recognize the features that distinguish nuclear energy safety from that for other industries. These features include: Nuclear reactivity, fission product radiation, and radioactive decay heat. Small and medium power reactors offer potential advantages over LWRs, particularly for reactivity and decay heat. (orig.)

  3. Safety and licensing for small and medium power reactors

    International Nuclear Information System (INIS)

    Trauger, D.B.

    1987-01-01

    Proposed new concepts for small and medium power reactors differ substantially from traditional Light Water Reactors (LWRs). Although designers have a large base of experience in safety and licensing, much of it is not relevant to new concepts. It can be a disadvantage if regulators apply LWR rules directly. A fresh start is appropriate. The extensive interactions between industry, regulators, and the public complicates but may enhance safety. It is basic to recognize the features that distinguish nuclear energy safety from that for other industries. These features include: nuclear reactivity, fission product radiation, and radioactive decay heat: small and medium power reactors offer potential advantages over LWRs, particularly for reactivity and decay heat. (author)

  4. Hybrid Reactor Simulation of Boiling Water Reactor Power Oscillations

    International Nuclear Information System (INIS)

    Huang Zhengyu; Edwards, Robert M.

    2003-01-01

    Hybrid reactor simulation (HRS) of boiling water reactor (BWR) instabilities, including in-phase and out-of-phase (OOP) oscillations, has been implemented on The Pennsylvania State University TRIGA reactor. The TRIGA reactor's power response is used to simulate reactor neutron dynamics for in-phase oscillation or the fundamental mode of the reactor modal kinetics for OOP oscillations. The reactor power signal drives a real-time boiling channel simulation, and the calculated reactivity feedback is in turn fed into the TRIGA reactor via an experimental changeable reactivity device. The thermal-hydraulic dynamics, together with first harmonic mode power dynamics, is digitally simulated in the real-time environment. The real-time digital simulation of boiling channel thermal hydraulics is performed by solving constitutive equations for different regions in the channel and is realized by a high-performance personal computer. The nonlinearity of the thermal-hydraulic model ensures the capability to simulate the oscillation phenomena, limit cycle and OOP oscillation, in BWR nuclear power plants. By adjusting reactivity feedback gains for both modes, various oscillation combinations can be realized in the experiment. The dynamics of axially lumped power distribution over the core is displayed in three-dimensional graphs. The HRS reactor power response mimics the BWR core-wide power stability phenomena. In the OOP oscillation HRS, the combination of reactor response and the simulated first harmonic power using shaping functions mimics BWR regional power oscillations. With this HRS testbed, a monitoring and/or control system designed for BWR power oscillations can be experimentally tested and verified

  5. Operation and utilizations of Dalat nuclear research reactor

    International Nuclear Information System (INIS)

    Hien, P.Z.

    1988-01-01

    The reconstructed Dalat nuclear research reactor was commissioned in March 1984 and up to September 1988 more than 6200 hours of operation at nominal power have been recorded. The major utilizations of the reactor include radioisotope production, activation analysis, nuclear data research and training. A brief review of the utilizations of the reactor is presented. Some aspects of reactor safety are also discussed. (author)

  6. N Reactor Deactivation Program Plan

    International Nuclear Information System (INIS)

    Walsh, J.L.

    1993-12-01

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

  7. Economic analysis of EBT reactor

    International Nuclear Information System (INIS)

    Woo, J.T.; Uckan, N.A.; Lidsky, L.M.

    1977-01-01

    In order to establish the economic potential of the Elmo Bumpy Torus (EBT) reactor, two independent system-costing models have been developed. Both models predict capital costs of approximately $400/kW(th). These relatively low costs reflect the simplicity of the EBTR design. In particular, the modular nature of the individual blanket-shield segments, the low costs ''accelerator style'' containment building, high beta, and steady-state operation lead to relatively low reactor costs. A detailed cost breakdown for subsystems is analyzed. High cost and high uncertainty subsystems are identified to direct further emphasis into those areas. The calculated capital costs for the EBT reactor are compared with those costs quoted for tokamak reactors

  8. Preliminary Study of Potential Market for Small Reactors

    International Nuclear Information System (INIS)

    Minato, A.; Brown, N. W.

    2008-01-01

    Small reactors are an energy supply for a specific purpose and oriented for a different market than large reactors. Small reactors will provide a local solution for developed and developing countries, such as, in remote areas, on small grids, or for non-electricity applications such as, district heating, seawater desalination and process heat. Single or medium sized power stations with small reactors should be compared with single fissile or renewable energy source and not be compared with large reactors. CRIEPI and LLNL have studied the business opportunities for small reactors. The small reactor concept is planned for initial use in small remote communities and in developing countries with small power distribution grid. Rapid installation and simple operation of the small plants is intended to support use in these communities without requiring development of a substantial nuclear technology infrastructure. In this study, two approaches were used in the assessment of the potential market. The first approach took a global look at the need for small nuclear plants. Then selected countries and sites were identified based on countries expressing interest in small reactors to the IAEA and consideration of sites in the US and Japan. (1) Tunisia, Mexico, Indonesia, Uruguay, Egypt and Argentina have demonstrated interest in nuclear power. Selecting one of these is dependent on political and socioeconomic factors, some of which have been identified, that require direct interaction with the countries to establish if they represent real opportunities. (2) The states of Hawaii and Alaska in the United States have high power cost and remote or island communities that may benefit from small nuclear plants. Alaska has shown greater interest in power alternatives including small than Hawaii and there is clearly less public resistance to nuclear power in Alaska. (3) The countries in Central America are actively expanding their power grids but have not demonstrated great interest

  9. Actinide recycling in reactors

    International Nuclear Information System (INIS)

    Kuesters, H.; Wiese, H.W.; Krieg, B.

    1995-01-01

    The objective is an assessment of the transmutation of long-lived actinides and fission products and the incineration of plutonium for reducing the risk potential of radioactive waste from reactors in comparison to direct waste disposal. The contribution gives an interim account on homogeneous and heterogeneous recycling of 'risk nuclides' in thermal and fast reactors. Important results: - A homogeneous 5 percent admixture of minor actinides (MA) from N4-PWRs to EFR fuel would allow a transmutation not only of the EFR MA, but in addition of the MA from 5 or 6 PWRs of equal power. However, the incineration is restricted by safety considerations. - LWR have only a very low MA incineration potential, due to their disadvantageous neutron capture/fission ratio. - In order to keep the Cm inventory at a low level, it is advantageous to concentrate the Am heterogeneously in particular fuel elements or rods. (orig./HP)

  10. Research Program of a Super Fast Reactor

    International Nuclear Information System (INIS)

    Oka, Yoshiaki; Ishiwatari, Yuki; Liu, Jie; Terai, Takayuki; Nagasaki, Shinya; Muroya, Yusa; Abe, Hiroaki; Mori, Hideo; Akiba, Masato; Akimoto, Hajime; Okumura, Keisuke; Akasaka, Naoaki; GOTO, Shoji

    2006-01-01

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

  11. Reliability analysis of reactor pressure vessel intensity

    International Nuclear Information System (INIS)

    Zheng Liangang; Lu Yongbo

    2012-01-01

    This paper performs the reliability analysis of reactor pressure vessel (RPV) with ANSYS. The analysis method include direct Monte Carlo Simulation method, Latin Hypercube Sampling, central composite design and Box-Behnken Matrix design. The RPV integrity reliability under given input condition is proposed. The result shows that the effects on the RPV base material reliability are internal press, allowable basic stress and elasticity modulus of base material in descending order, and the effects on the bolt reliability are allowable basic stress of bolt material, preload of bolt and internal press in descending order. (authors)

  12. Nuclear reactor built, being built, or planned

    International Nuclear Information System (INIS)

    1991-06-01

    This document contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1990. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE, from the US Nuclear Regulatory Commission, from the US reactor manufacturers who are the principal nuclear contractors for foreign reactor locations, from US and foreign embassies, and from foreign governmental nuclear departments. The book is divided into three major sections: Section 1 consists of a reactor locator map and reactor tables; Section 2 includes nuclear reactors that are operating, being built, or planned; and Section 3 includes reactors that have been shut down permanently or dismantled. Sections 2 and 3 contain the following classification of reactors: Civilian, Production, Military, Export, and Critical Assembly

  13. CEA programme on gas cooled reactors

    International Nuclear Information System (INIS)

    Carre, F.; Fiorini, G.L.; Chapelot, Ph.; Gauthier, J.C.

    2002-01-01

    Future nuclear energy systems studies conducted by the CEA aim at investigating and developing promising technologies for future reactors, fuels and fuel cycles, for nuclear power to play a major part in sustainable energy policies. Reactors and fuel cycles are considered as integral parts of a nuclear system to be optimised as a whole. Major goals assigned to future nuclear energy systems are the following: reinforced economic competitiveness with other electricity generation means, with a special emphasis on reducing the investment cost; enhanced reliability and safety, through an improved management of reactor operation in normal and abnormal plant conditions; minimum production of long lived radioactive waste; resource saving through an effective and flexible use of the available resources of fissile and fertile materials; enhanced resistance to proliferation risks. The three latter goals are essential for the sustainability of nuclear energy in the long term. Additional considerations such as the potentialities for other applications than electricity generation (co-generation, production of hydrogen, sea water desalination) take on an increasing importance. Sustainability goals call for fast neutron spectra (to transmute nuclear waste and to breed fertile fuel) and for recycling actinides from the spent fuel (plutonium and minor actinides). New applications and economic competitiveness call for high temperature technologies (850 deg C), that afford high conversion efficiencies and hence less radioactive waste production and discharged heat. These orientations call for breakthroughs beyond light water reactors. Therefore, as a result of a screening review of candidate technologies, the CEA has selected an innovative concept of high temperature gas cooled reactor with a fast neutron spectrum, robust refractory fuel, direct conversion with a gas turbine, and integrated on-site fuel cycle as a promising system for a sustainable energy development. This objective

  14. Generic magnetic fusion reactor cost assessment

    International Nuclear Information System (INIS)

    Sheffield, J.

    1985-01-01

    The Fusion Energy Division of the Oak Ridge National Laboratory discusses ''generic'' magnetic fusion reactors. The author comments on DT burning magnetic fusion reactor models being possibly operational in the 21st century. Representative parameters from D-T reactor studies are given, as well as a shematic diagram of a generic fusion reactor. Values are given for winding pack current density for existing and future superconducting coils. Topics included are the variation of the cost of electricity (COE), the dependence of the COE on the net electric power of the reactor, and COE formula definitions

  15. Artificial intelligence applications to nuclear reactor diagnostics

    International Nuclear Information System (INIS)

    Lee, J.C.; Hassberger, J.A.; Wehe, D.K.

    1987-01-01

    The authors research into applications of artificial intelligence to nuclear reactor diagnostics involves three main areas. In the first area, the authors combine reactor simulation models and expert systems to diagnose the state of the plant. The second area examines ways in which the rule or knowledge base of an intelligent controller can be generated systematically from either fault trees or acquired plant data. Third, efforts are described to develop the capabilities to validate these techniques in a realistic reactor setting. The techniques are applicable to all reactor types, including fast reactors

  16. Safety of research reactors (Design and Operation)

    International Nuclear Information System (INIS)

    Dirar, H. M.

    2012-06-01

    The primary objective of this thesis is to conduct a comprehensive up-to-date literature review on the current status of safety of research reactor both in design and operation providing the future trends in safety of research reactors. Data and technical information of variety selected historical research reactors were thoroughly reviewed and evaluated, furthermore illustrations of the material of fuel, control rods, shielding, moderators and coolants used were discussed. Insight study of some historical research reactors was carried with considering sample cases such as Chicago Pile-1, F-1 reactor, Chalk River Laboratories,. The National Research Experimental Reactor and others. The current status of research reactors and their geographical distribution, reactor category and utilization is also covered. Examples of some recent advanced reactors were studied like safety barriers of HANARO of Korea including safety doors of the hall and building entrance and finger print identification which prevent the reactor from sabotage. On the basis of the results of this research, it is apparent that a high quality of safety of nuclear reactors can be attained by achieving enough robust construction, designing components of high levels of efficiency, replacing the compounds of the reactor in order to avoid corrosion and degradation with age, coupled with experienced scientists and technical staffs to operate nuclear research facilities.(Author)

  17. Irradiation Facilities at the Advanced Test Reactor

    International Nuclear Information System (INIS)

    S. Blaine Grover

    2005-01-01

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

  18. Reaction behaviour of reactor materials, including concrete during core meltdown

    International Nuclear Information System (INIS)

    Holleck, H.; Hofmann, P.; Skokan, A.

    1976-01-01

    The problems dealt with in the studies on the material behaviour of core meltdowns are sketched, and an outline is given on the experimental investigations now in progress. These studies aim at establishing a model which is as realistic as possible, and the paper tries to show that knowledge on the reaction behaviour is a basic prerequisite for this. (orig./TK) [de

  19. Nuclear Reactor Sharing Program

    International Nuclear Information System (INIS)

    1994-01-01

    The Ohio State University Research Reactor (OSURR) is licensed to operate at a maximum power level of 500 kW. A pool-type reactor using flat-plate, low enriched fuel elements, the OSURR provides several experimental facilities including two 6-inch i.d. beam ports, a graphite thermal column, several graphite-isotope-irradiation elements, a pneumatic transfer system (Rabbit), various dry tubes, and a Central Irradiation Facility (CIF). The core arrangement and accessibility facilitates research programs involving material activation or core parameter studies. The OSURR control room is large enough to accommodate laboratory groups which can use control instrumentation for monitoring of experiments. The control instrumentation is relatively simple, without a large amount of duplication. This facilitates opportunities for hands-on experience in reactor operation by nuclear engineering students making reactor parameter measurements. For neutron activation analysis and analyses of natural environmental radioactivity, the NRL maintains the gamma ray spectroscopy system (GRSS). It is comprised of two PC-based 8192-channel multichannel analyzers (MCAs) with all the required software for quantitative analysis. A 3 double-prime x 3 double-prime NaI(Tl), a 14 percent Ge(Li), and a High Purity Germanium detector are currently available for use with the spectroscopy system

  20. Nuclear research reactors

    International Nuclear Information System (INIS)

    1985-01-01

    It's presented data about nuclear research reactors in the world, retrieved from the Sien (Nuclear and Energetic Information System) data bank. The information are organized in table forms as follows: research reactors by countries; research reactors by type; research reactors by fuel and research reactors by purpose. (E.G.) [pt

  1. Core clamping device for a nuclear reactor

    International Nuclear Information System (INIS)

    Guenther, R.W.

    1974-01-01

    The core clamping device for a fast neutron reactor includes clamps to support the fuel zone against the pressure vessel. The clamps are arranged around the circumference of the core. They consist of torsion bars arranged parallel at some distance around the core with lever arms attached to the ends whose force is directed in the opposite direction, pressing against the wall of the pressure vessel. The lever arms and pressure plates also actuated by the ends of the torsion bars transfer the stress, the pressure plates acting upon the fuel elements or fuel assemblies. Coupling between the ends of the torsion bars and the pressure plates is achieved by end carrier plates directly attached to the torsion bars and radially movable. This clamping device follows the thermal expansions of the core, allows specific elements to be disengaged in sections and saves space between the core and the neutron reflectors. (DG) [de

  2. TU electric reactor model verification

    International Nuclear Information System (INIS)

    Willingham, C.E.; Killgore, M.R.

    1989-01-01

    Power reactor benchmark calculations using the code package CASMO-3/SIMULATE-3 have been performed for six cycles of Prairie Island Unit 1. The reload fuel designs for the selected cycles include gadolinia as a burnable absorber, natural uranium axial blankets, and increased water-to-fuel ratio. The calculated results for both low-power physics tests (boron end points, control rod worths, and isothermal temperature coefficients) and full-power operation (power distributions and boron letdown) are compared to measured plant data. These comparisons show that the TU Electric reactor physics models accurately predict important physics parameters for power reactors

  3. Final progress report on Grant No. DE-FG02-81ER10229, U.S. Department of Energy Reactor Sharing Program at the University of Wisconsin, July 15, 2000 - May 31, 2001

    Energy Technology Data Exchange (ETDEWEB)

    Agasie, Robert J.

    2001-12-27

    The Reactor Sharing Program makes the facilities of the University of Wisconsin Nuclear Reactor Laboratory available to other educational institutions. Uses include direct instruction, student theses projects, and staff research projects. A list of using institutions and a brief description of use is given.

  4. Nuclear heating reactor, an advanced and passive reactor

    International Nuclear Information System (INIS)

    Wang Dazhong; Zheng Wenxiang

    1994-01-01

    The nuclear heating reactor (NHR) is designed with a number of the advanced and innovative features, including integrated arrangement, natural circulation, self-pressurized performance, dual vessel structure, hydraulic control rod drive and passive safety systems. Being an advanced and passive reactor, the NHR can serve as a clean, safe and economic energy source. This paper describes the development status, main design and safety features of the NHR. 3 refs., 2 tabs., 5 figs

  5. Nuclear reactor

    International Nuclear Information System (INIS)

    Anthony, A.J.; Gruber, E.A.

    1979-01-01

    A nuclear reactor with control rods in channels between fuel assemblies wherein the fuel assemblies incorporate guide rods which protrude outwardly into the control rod channels to prevent the control rods from engaging the fuel elements. The guide rods also extend back into the fuel assembly such that they are relatively rigid members. The guide rods are tied to the fuel assembly end or support plates and serve as structural members which are supported independently of the fuel element. Fuel element spacing and support means may be attached to the guide rods. 9 claims

  6. Plant Control of the High Performance Light Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Schlagenhaufer, Marc; Starflinger, J.; Schulenberg, T. [Institute for Nuclear and Energy Technologies, Forschungszentrum Karlsruhe GmbH, Hermann-von-Helmholtz-Platz 1, Eggenstein-Leopoldshafen, Baden-Wuertemberg 76344 (Germany)

    2009-06-15

    The latest design concept of the High Performance Light Water Reactor (HPLWR) includes a thermal core in which supercritical water at 25 MPa inlet pressure is heated up from 280 deg. C reactor inlet temperature to 500 deg. C core exit temperature in three steps with intermediate coolant mixing to minimize peak cladding temperatures of the fuel rods. A direct supercritical steam cycle of the HPLWR has been designed with high, intermediate and low pressure turbines with a single reheat to 441 deg. C at 4.04 MPa pressure. Three low pressure pre-heaters and four high pressure pre-heaters are foreseen to achieve the envisaged reactor inlet temperature of 280 deg. C at full load. A feedwater tank of 603 m{sup 3} at 0.55 MPa pressure serves as an accumulator for normal and accidental conditions. The steam cycle has been modelled with APROS, developed by VTT Finland, to provide thermodynamic data and cycle efficiency values under full load and part load operation conditions as well as the transient response to load changes. A plant control system has been designed in which the reactor inlet pressure is controlled by the turbine valve, the reactor power is controlled by the feedwater pumps while the life steam temperature is controlled by control rods, and the reheat temperature is controlled by the reheater valve. Neglecting the reactivity control, the core power can also be treated as input parameter such that the life steam temperature is directly controlled by the feedwater mass flow. The plant control can handle all loading and de-loading cycles including complete shut down. A constant pressure at reactor inlet is foreseen for all load cases. Peak temperatures of the fuel pins are checked with a simplified core model. Two shut down procedures starting at 50% load are presented. A reactor scram with turbine states the safe shut down of the whole plant. To avoid hard material temperature changes, a controlled shut down procedure is designed. The rotational speed of the

  7. Development of technologies for nuclear reactors of small and medium sized

    International Nuclear Information System (INIS)

    2011-08-01

    This meeting include: countries presentations, themes and objectives of the training course, reactor types, design, EPR, APR1400, A P 1000, A PWR, ATMEA 1, VVER-1000, A PWR, ATMEA 1, VVER 1200, Boiling Water Reactor, A BWR, A BWR -II, ESBUR, Ke ren, AREVA, Heavy Water Reactor, Candu 6, Acr-1000, HWR, Bw, Iris, CAREM NuCcale, Smart, KLT-HOS, Westinghouse small modular Reactor, Gas Cooled Reactors, PBMR, React ores enfriados con metales liquidos, Hs, Prism,Terra Power, Hyper ion, appliance's no electric as de energia, Generation IV Reactors,VHTR, Gas Fast Reactor, Sodium Fast Reactor, Molten salt Reactor, Lfr, Water Cooled Reactor, Technology Assessment Process, Fukushima accident.

  8. Reactor Engineering Department annual report (April 1, 1986 - March 31, 1987)

    International Nuclear Information System (INIS)

    1987-08-01

    Research and development activities in the Department of Reactor Engineering in the fiscal year 1986 are described. The major activities of the Department are closely related to the reactor physics of very high temperature gas-cooled reactor, high conversion light water reactor and liquid metal fast breeder reactor and to blanket neutronics of fusion reactor. Contents of this report are divided into the activities on nuclear data and group constants, theoretical methods and code development, reactor physics experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control, diagnosis and robotics. The activity of the Research Committee on Reactor Physics is also included. (author)

  9. Diagnosis of electric equipment at the Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    Nguyen Truong Sinh

    1999-01-01

    The Dalat Nuclear Research Reactor (DNRR) is a pool type of its kind in the world: Soviet-designed core and control system harmoniously integrated into the left-over infrastructure of the former American-made TRIGA MARK II reactor, which includes the reactor tank and shielding, graphite reflector, beam tubes and thermal column. The reactor is mainly used for radioisotope and radiopharmaceutical production, elemental analysis using neutron activation techniques, neutron beam exploitation, silicon doping, and reactor physics experimentation. For safe operation of the reactor maintenance work has been carried out for the reactor control and instrumentation, reactor cooling, ventilation, radiomonitoring, mechanical, normal electric supply systems as well as emergency electric diesel generators and the water treatment station. Technical management of the reactor includes periodical maintenance as required by technical specifications, training, re-training and control of knowledge for reactor staff. During recent years, periodic preventive maintenance (PPM) has been carried out for the electric machines of the technological systems. (author)

  10. Reactor vessel sealing plug

    International Nuclear Information System (INIS)

    Dooley, R.A.

    1986-01-01

    An apparatus is described for sealing a cold leg nozzle of a nuclear reactor pressure vessel from a remote location comprising: at least one sealing plug for mechanically sealing the nozzle from the inside of the reactor pressure vessel. The sealing plug includes a plate and a cone assembly having an end part receptive in the nozzle, the plate being axially moveable relative to the cone assembly. The plate and cone assembly have confronting bevelled edges defining an opening therebetween. A primary O-ring is disposed about the opening and is supported on the bevelled edges, the plate being guidably mounted to the cone assembly for movement toward the cone assembly to radially expand the primary O-ring into sealing engagement with the nozzle. A means is included for providing relative movement between the outer plate and the cone assembly

  11. Nuclear reactor cavity floor passive heat removal system

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, Tyler A.; Neeley, Gary W.; Inman, James B.

    2018-03-06

    A nuclear reactor includes a reactor core disposed in a reactor pressure vessel. A radiological containment contains the nuclear reactor and includes a concrete floor located underneath the nuclear reactor. An ex vessel corium retention system includes flow channels embedded in the concrete floor located underneath the nuclear reactor, an inlet in fluid communication with first ends of the flow channels, and an outlet in fluid communication with second ends of the flow channels. In some embodiments the inlet is in fluid communication with the interior of the radiological containment at a first elevation and the outlet is in fluid communication with the interior of the radiological containment at a second elevation higher than the first elevation. The radiological containment may include a reactor cavity containing a lower portion of the pressure vessel, wherein the concrete floor located underneath the nuclear reactor is the reactor cavity floor.

  12. Improvement to reactor vessel

    International Nuclear Information System (INIS)

    1974-01-01

    The vessel described includes a prestressed concrete vessel containing a chamber and a removable cover closing this chamber. The cover is in concrete and is kept in its closed position by main and auxiliary retainers, comprising fittings integral with the concrete of the vessel. The auxiliary retainers pass through the concrete of the cover. This improvement may be applied to BWR, PWR and LMFBR type reactor vessel [fr

  13. Fusion Reactor Materials

    International Nuclear Information System (INIS)

    Moons, F.

    1998-01-01

    SCK-CEN's programme on fusion reactor materials includes studies (1) to investigate fracture mechanics of neutron-irradiated beryllium; (2) to describe the helium behaviour in irradiated beryllium at atomic scale; (3) to define the kinetics of beryllium reacting with air or steam; (3) to perform a feasibility study for the testing of integrated blanket modules under neutron irradiation. Progress and achievements in 1997 are reported

  14. IAEA activities related to research reactor fuel conversion and spent fuel return programs

    International Nuclear Information System (INIS)

    Goldman, Ira N.; Adelfang, Pablo; Ritchie, Iain G.

    2005-01-01

    The IAEA has been involved for more than twenty years in supporting international nuclear non-proliferation efforts associated with reducing the amount of highly enriched uranium (HEU) in international commerce. IAEA projects and activities have directly supported the Reduced Enrichment for Research and Test Reactors (RERTR) programme, as well as directly associated efforts to return research reactor fuel to the country where it was originally enriched. IAEA efforts have included the development and maintenance of several data bases with information related to research reactors and research reactor spent fuel inventories that have been essential in planning and managing both RERTR and spent fuel return programmes. Other IAEA regular budget programs have supported research reactor fuel conversion from HEU to low enriched uranium (LEU), and in addressing issues common to many member states with spent fuel management problems and concerns. The paper briefly describes IAEA involvement since the early 1980's in these areas, including regular budget and Technical Co-operation programme activities, and focuses on efforts in the past five years to continue to support and accelerate U.S. and Russian research reactor spent fuel return programmes. (author)

  15. D and DR Reactors

    Data.gov (United States)

    Federal Laboratory Consortium — The world's second full-scale nuclear reactor was the D Reactor at Hanford which was built in the early 1940's and went operational in December of 1944.D Reactor ran...

  16. Reactor core fuel management

    International Nuclear Information System (INIS)

    Silvennoinen, P.

    1976-01-01

    The subject is covered in chapters, entitled: concepts of reactor physics; neutron diffusion; core heat transfer; reactivity; reactor operation; variables of core management; computer code modules; alternative reactor concepts; methods of optimization; general system aspects. (U.K.)

  17. Hybrid adsorptive membrane reactor

    Science.gov (United States)

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

    2011-03-01

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

  18. Brazilian multipurpose reactor

    International Nuclear Information System (INIS)

    2014-01-01

    The Brazilian Multipurpose Reactor (RMB) Project is an action of the Federal Government, through the Ministry of Science Technology and Innovation (MCTI) and has its execution under the responsibility of the Brazilian National Nuclear Energy Commission (CNEN). Within the CNEN, the project is coordinated by the Research and Development Directorate (DPD) and developed through research units of this board: Institute of Nuclear Energy Research (IPEN); Nuclear Engineering Institute (IEN); Centre for Development of Nuclear Technology (CDTN); Regional Center of Nuclear Sciences (CRCN-NE); and Institute of Radiation Protection and Dosimetry (IRD). The Navy Technological Center in Sao Paulo (CTMSP) and also the participation of other research centers, universities, laboratories and companies in the nuclear sector are important and strategic partnerships. The conceptual design and the safety analysis of the reactor and main facilities, related to nuclear and environmental licensing, are performed by technicians of the research units of DPD / CNEN. The basic design was contracted to engineering companies as INTERTHECNE from Brazil and INVAP from Argentine. The research units from DPD/CNEN are also responsible for the design verification on all engineering documents developed by the contracted companies. The construction and installation should be performed by specific national companies and international partnerships. The Nuclear Reactor RMB will be a open pool type reactor with maximum power of 30 MW and have the OPAL nuclear reactor of 20 MW, built in Australia and designed by INVAP, as reference. The RMB reactor core will have a 5x5 configuration, consisting of 23 elements fuels (EC) of U 3 Si 2 dispersion-type Al having a density of up to 3.5 gU/cm 3 and enrichment of 19.75% by weight of 23 5 U. Two positions will be available in the core for materials irradiation devices. The main objectives of the RMB Reactor and the other nuclear and radioactive facilities are

  19. Mechanical design of a light water breeder reactor

    Science.gov (United States)

    Fauth, Jr., William L.; Jones, Daniel S.; Kolsun, George J.; Erbes, John G.; Brennan, John J.; Weissburg, James A.; Sharbaugh, John E.

    1976-01-01

    In a light water reactor system using the thorium-232 -- uranium-233 fuel system in a seed-blanket modular core configuration having the modules arranged in a symmetrical array surrounded by a reflector blanket region, the seed regions are disposed for a longitudinal movement between the fixed or stationary blanket region which surrounds each seed region. Control of the reactor is obtained by moving the inner seed region thus changing the geometry of the reactor, and thereby changing the leakage of neutrons from the relatively small seed region into the blanket region. The mechanical design of the Light Water Breeder Reactor (LWBR) core includes means for axially positioning of movable fuel assemblies to achieve the neutron economy required of a breeder reactor, a structure necessary to adequately support the fuel modules without imposing penalties on the breeding capability, a structure necessary to support fuel rods in a closely packed array and a structure necessary to direct and control the flow of coolant to regions in the core in accordance with the heat transfer requirements.

  20. Foundational development of an advanced nuclear reactor integrated safety code.

    Energy Technology Data Exchange (ETDEWEB)

    Clarno, Kevin (Oak Ridge National Laboratory, Oak Ridge, TN); Lorber, Alfred Abraham; Pryor, Richard J.; Spotz, William F.; Schmidt, Rodney Cannon; Belcourt, Kenneth (Ktech Corporation, Albuquerque, NM); Hooper, Russell Warren; Humphries, Larry LaRon

    2010-02-01

    This report describes the activities and results of a Sandia LDRD project whose objective was to develop and demonstrate foundational aspects of a next-generation nuclear reactor safety code that leverages advanced computational technology. The project scope was directed towards the systems-level modeling and simulation of an advanced, sodium cooled fast reactor, but the approach developed has a more general applicability. The major accomplishments of the LDRD are centered around the following two activities. (1) The development and testing of LIME, a Lightweight Integrating Multi-physics Environment for coupling codes that is designed to enable both 'legacy' and 'new' physics codes to be combined and strongly coupled using advanced nonlinear solution methods. (2) The development and initial demonstration of BRISC, a prototype next-generation nuclear reactor integrated safety code. BRISC leverages LIME to tightly couple the physics models in several different codes (written in a variety of languages) into one integrated package for simulating accident scenarios in a liquid sodium cooled 'burner' nuclear reactor. Other activities and accomplishments of the LDRD include (a) further development, application and demonstration of the 'non-linear elimination' strategy to enable physics codes that do not provide residuals to be incorporated into LIME, (b) significant extensions of the RIO CFD code capabilities, (c) complex 3D solid modeling and meshing of major fast reactor components and regions, and (d) an approach for multi-physics coupling across non-conformal mesh interfaces.

  1. Development of advanced strain diagnostic techniques for reactor environments.

    Energy Technology Data Exchange (ETDEWEB)

    Fleming, Darryn D.; Holschuh, Thomas Vernon,; Miller, Timothy J.; Hall, Aaron Christopher; Urrea, David Anthony,; Parma, Edward J.,

    2013-02-01

    The following research is operated as a Laboratory Directed Research and Development (LDRD) initiative at Sandia National Laboratories. The long-term goals of the program include sophisticated diagnostics of advanced fuels testing for nuclear reactors for the Department of Energy (DOE) Gen IV program, with the future capability to provide real-time measurement of strain in fuel rod cladding during operation in situ at any research or power reactor in the United States. By quantifying the stress and strain in fuel rods, it is possible to significantly improve fuel rod design, and consequently, to improve the performance and lifetime of the cladding. During the past year of this program, two sets of experiments were performed: small-scale tests to ensure reliability of the gages, and reactor pulse experiments involving the most viable samples in the Annulated Core Research Reactor (ACRR), located onsite at Sandia. Strain measurement techniques that can provide useful data in the extreme environment of a nuclear reactor core are needed to characterize nuclear fuel rods. This report documents the progression of solutions to this issue that were explored for feasibility in FY12 at Sandia National Laboratories, Albuquerque, NM.

  2. A comparison of radioactive waste from first generation fusion reactors and fast fission reactors with actinide recycling

    International Nuclear Information System (INIS)

    Koch, M.; Kazimi, M.S.

    1991-04-01

    Limitations of the fission fuel resources will presumably mandate the replacement of thermal fission reactors by fast fission reactors that operate on a self-sufficient closed fuel cycle. This replacement might take place within the next one hundred years, so the direct competitors of fusion reactors will be fission reactors of the latter rather than the former type. Also, fast fission reactors, in contrast to thermal fission reactors, have the potential for transmuting long-lived actinides into short-lived fission products. The associated reduction of the long-term activation of radioactive waste due to actinides makes the comparison of radioactive waste from fast fission reactors to that from fusion reactors more rewarding than the comparison of radioactive waste from thermal fission reactors to that from fusion reactors. Radioactive waste from an experimental and a commercial fast fission reactor and an experimental and a commercial fusion reactor has been characterized. The fast fission reactors chosen for this study were the Experimental Breeder Reactor 2 and the Integral Fast Reactor. The fusion reactors chosen for this study were the International Thermonuclear Experimental Reactor and a Reduced Activation Ferrite Helium Tokamak. The comparison of radioactive waste parameters shows that radioactive waste from the experimental fast fission reactor may be less hazardous than that from the experimental fusion reactor. Inclusion of the actinides would reverse this conclusion only in the long-term. Radioactive waste from the commercial fusion reactor may always be less hazardous than that from the commercial fast fission reactor, irrespective of the inclusion or exclusion of the actinides. The fusion waste would even be far less hazardous, if advanced structural materials, like silicon carbide or vanadium alloy, were employed

  3. A comparison of radioactive waste from first generation fusion reactors and fast fission reactors with actinide recycling

    Energy Technology Data Exchange (ETDEWEB)

    Koch, M.; Kazimi, M.S.

    1991-04-01

    Limitations of the fission fuel resources will presumably mandate the replacement of thermal fission reactors by fast fission reactors that operate on a self-sufficient closed fuel cycle. This replacement might take place within the next one hundred years, so the direct competitors of fusion reactors will be fission reactors of the latter rather than the former type. Also, fast fission reactors, in contrast to thermal fission reactors, have the potential for transmuting long-lived actinides into short-lived fission products. The associated reduction of the long-term activation of radioactive waste due to actinides makes the comparison of radioactive waste from fast fission reactors to that from fusion reactors more rewarding than the comparison of radioactive waste from thermal fission reactors to that from fusion reactors. Radioactive waste from an experimental and a commercial fast fission reactor and an experimental and a commercial fusion reactor has been characterized. The fast fission reactors chosen for this study were the Experimental Breeder Reactor 2 and the Integral Fast Reactor. The fusion reactors chosen for this study were the International Thermonuclear Experimental Reactor and a Reduced Activation Ferrite Helium Tokamak. The comparison of radioactive waste parameters shows that radioactive waste from the experimental fast fission reactor may be less hazardous than that from the experimental fusion reactor. Inclusion of the actinides would reverse this conclusion only in the long-term. Radioactive waste from the commercial fusion reactor may always be less hazardous than that from the commercial fast fission reactor, irrespective of the inclusion or exclusion of the actinides. The fusion waste would even be far less hazardous, if advanced structural materials, like silicon carbide or vanadium alloy, were employed.

  4. Reactor core design of Gas Turbine High Temperature Reactor 300

    International Nuclear Information System (INIS)

    Kunitomi, Kazuhiko; Katanishi, Shoji; Takada, Shoji; Yan Xing; Tsuji, Nobumasa

    2004-01-01

    Japan Atomic Energy Research Institute (JAERI) has been designing Japan's original gas turbine high temperature reactor, Gas Turbine High Temperature Reactor 300 (GTHTR300). The greatly simplified design based on salient features of the High Temperature Gas-cooled Reactor (HTGR) with a closed helium gas turbine enables the GTHTR300 a highly efficient and economically competitive reactor to be deployed in early 2010s. Also, the GTHTR300 fully taking advantage of various experiences accumulated in design, construction and operation of the High Temperature Engineering Test Reactor (HTTR) and existing fossil fired gas turbine systems reduces technological development concerning a reactor system and electric generation system. Original design features of this system are the reactor core design based on a newly proposed refueling scheme named sandwich shuffling, conventional steel material usage for a reactor pressure vessel (RPV), an innovative coolant flow scheme and a horizontally installed gas turbine unit. The GTHTR300 can be continuously operated without the refueling for 2 years. Due to these salient features, the capital cost of the GTHTR300 is less than a target cost of 200,000 yen (1667 US$)/kW e, and the electric generation cost is close to a target cost of 4 yen (3.3 US cents)/kW h. This paper describes the original design features focusing on the reactor core design and the in-core structure design, including the innovative coolant flow scheme for cooling the RPV. The present study is entrusted from the Ministry of Education, Culture, Sports, Science and Technology of Japan

  5. Reactor container

    International Nuclear Information System (INIS)

    Oikawa, Hirohide; Otonari, Jun-ichiro; Tozaki, Yuka.

    1993-01-01

    Partition walls are disposed between a reactor pressure vessel and a suppression chamber to separate a dry well to an upper portion and a lower portion. A communication pipe is disposed to the partition walls. One end of the communication pipe is opened in an upper portion of the dry well at a position higher than a hole disposed to a bent tube of the suppression chamber. When coolants overflow from a depressurization valve by an erroneous operation of an emergency reactor core cooling device, the coolants accumulate in the upper portion of the dry well. When the pipeline is ruptured at the upper portion of the pressure vessel, only the inside of the pressure vessel and the upper portion of the dry well are submerged in water. In this case, the water level of the coolants does not elevate to the opening of the commuication pipe but they flow into the suppression chamber from the hole disposed to the bent tube. Since the coolants do not flow out to the lower portion of the dry well, important equipments such as control rod drives disposed at the lower portion of the dry wall can be prevented from submerging in water. (I.N.)

  6. Reactor container

    Energy Technology Data Exchange (ETDEWEB)

    Oikawa, Hirohide; Otonari, Jun-ichiro; Tozaki, Yuka.

    1993-09-07

    Partition walls are disposed between a reactor pressure vessel and a suppression chamber to separate a dry well to an upper portion and a lower portion. A communication pipe is disposed to the partition walls. One end of the communication pipe is opened in an upper portion of the dry well at a position higher than a hole disposed to a bent tube of the suppression chamber. When coolants overflow from a depressurization valve by an erroneous operation of an emergency reactor core cooling device, the coolants accumulate in the upper portion of the dry well. When the pipeline is ruptured at the upper portion of the pressure vessel, only the inside of the pressure vessel and the upper portion of the dry well are submerged in water. In this case, the water level of the coolants does not elevate to the opening of the commuication pipe but they flow into the suppression chamber from the hole disposed to the bent tube. Since the coolants do not flow out to the lower portion of the dry well, important equipments such as control rod drives disposed at the lower portion of the dry wall can be prevented from submerging in water. (I.N.).

  7. Reactor monitor

    International Nuclear Information System (INIS)

    Takada, Tamotsu.

    1992-01-01

    The device of the present invention monitors a reactor so that each of the operations for the relocation of fuel assemblies and the withdrawal and the insertion of control rods upon exchange of fuel assemblies and control rods in the reactor. That is, when an operator conducts relocating operation by way of a fuel assembly operation section, the device of the present invention judges whether the operation indication is adequate or not, based on the information of control rod arrangement in a control rod memory section. When the operation indication is wrong, a stop signal is sent to a fuel assembly relocating device. Further, when the operator conducts control rod operation by way of a control rod operation section, the device of the present invention judges in the control rod withdrawal judging section, as to whether the operation indication given by the operator is adequate or not by comparing it with fuel assembly arrangement information. When the operation indication is wrong, a stop signal is sent to control rod drives. With such procedures, increase of nuclear heating upon occurrence of erroneous operation can be prevented. (I.S.)

  8. Tests for validation of fast neutron reactors safety

    International Nuclear Information System (INIS)

    Nagata, T.; Yamashita, H.

    2001-01-01

    Japanese scientific research and design enterprises in cooperation with industrial and power generating corporations implement a project on creating a fast neutron reactor of the ultimate safety. One of the basic expected results from such a development is creation of a reactor core structure that is able to eliminate recriticality occurrence in the course of reactor accident involving fuel melting. One of the possible ways to solve this problem is to include pipes (meant for specifying directed (controlled) molten fuel relocation) into fuel assembly structure. In the course of conduction and subsequent implementation of such a design the basic issue is to experimentally confirm the operating capacity of FA having such a structure and that is called FAIDUS. Within EAGLE Project on experimental basis of IAE NNC RK an activity has been started on preparation and conduction of out-of-pile and in-pile tests. During tests a sodium coolant will be used. Studies are conducted by NNC RK in cooperation with the Japanese corporations JAPC and JNC. Basic objective of out-of-pile tests was to obtain preliminary information on fuel relocation behavior under conditions simulating accident involving melting of core consisting of FAIDUS FA, which will help to clarify simulation criteria and to develop the most optimum structure of the experimental channel for reactor experiments conduction. The basic objective of in-pile tests was the experimental confirmation of operating capacity of FAIDUS FA model under reactor conditions. According to the program two tests are planned to be performed at IGR reactor: tests for validation of fast neutron reactor safety, and out-of-pile tests at EAGLE experimental facility without sodium coolant

  9. The siting of UK nuclear reactors.

    Science.gov (United States)

    Grimston, Malcolm; Nuttall, William J; Vaughan, Geoff

    2014-06-01

    Choosing a suitable site for a nuclear power station requires the consideration and balancing of several factors. Some 'physical' site characteristics, such as the local climate and the potential for seismic activity, will be generic to all reactors designs, while others, such as the availability of cooling water, the area of land required and geological conditions capable of sustaining the weight of the reactor and other buildings will to an extent be dependent on the particular design of reactor chosen (or alternatively the reactor design chosen may to an extent be dependent on the characteristics of an available site). However, one particularly interesting tension is a human and demographic one. On the one hand it is beneficial to place nuclear stations close to centres of population, to reduce transmission losses and other costs (including to the local environment) of transporting electricity over large distances from generator to consumer. On the other it is advantageous to place nuclear stations some distance away from such population centres in order to minimise the potential human consequences of a major release of radioactive materials in the (extremely unlikely) event of a major nuclear accident, not only in terms of direct exposure but also concerning the management of emergency planning, notably evacuation.This paper considers the emergence of policies aimed at managing this tension in the UK. In the first phase of nuclear development (roughly speaking 1945-1965) there was a highly cautious attitude, with installations being placed in remote rural locations with very low population density. The second phase (1965-1985) saw a more relaxed approach, allowing the development of AGR nuclear power stations (which with concrete pressure vessels were regarded as significantly safer) closer to population centres (in 'semi-urban' locations, notably at Hartlepool and Heysham). In the third phase (1985-2005) there was very little new nuclear development, Sizewell

  10. Prospect of realizing nuclear fusion reactors

    International Nuclear Information System (INIS)

    1989-01-01

    This Report describes the results of the research work on nuclear fusion, which CRIEPI has carried out for about ten years from the standpoint of electric power utilities, potential user of its energy. The principal points are; (a) economic analysis (calculation of costs) based on Japanese analysis procedures and database of commercial fusion reactors, including fusion-fission hybrid reactors, and (b) conceptual design of two types of hybrid reactors, that is, fission-fuel producing DMHR (Demonstration Molten-Salt Hybrid Reactor) and electric-power producing THPR (Tokamak Hybrid Power Reactor). The Report consists of the following chapters: 1. Introduction. 2. Conceptual Design of Hybrid Reactors. 3. Economic Analysis of Commercial Fusion Reactors. 4. Basic Studies Applicable Also to Nuclear Fusion Technology. 5. List of Published Reports and Papers; 6. Conclusion. Appendices. (author)

  11. Pneumatic transport systems for TRIGA reactors

    International Nuclear Information System (INIS)

    Bolton, John A.

    1970-01-01

    Main parameters and advantages of pneumatically operated systems, primarily those operated by gas pressure are discussed. The special irradiation ends for the TRIGA reactor are described. To give some idea of the complexity of some modern systems, the author presents the large system currently operating at the National Bureau of Standards in Washington. In this system, 13 stations are located throughout the radiochemistry laboratories and three irradiation ends are located in the reactor, which is a 14-megawatt unit. The system incorporates practically every fail-safe device possible, including ball valves located on all capsule lines entering the reactor area, designed to close automatically in the event of a reactor scram, and at that time capsules within the reactor would be diverted by means of switches located on the inside of the reactor wall. The whole system is under final control of a permission control panel located in the reactor control room. Many other safety accessories of the system are described

  12. Seismic research on graphite reactor core

    International Nuclear Information System (INIS)

    Lai Shigang; Sun Libin; Zhang Zhengming

    2013-01-01

    Background: Reactors with graphite core structure include production reactor, water-cooled graphite reactor, gas-cooled reactor, high-temperature gas-cooled reactor and so on. Multi-body graphite core structure has nonlinear response under seismic excitation, which is different from the response of general civil structure, metal connection structure or bolted structure. Purpose: In order to provide references for the designing and construction of HTR-PM. This paper reviews the history of reactor seismic research evaluation from certain countries, and summarizes the research methods and research results. Methods: By comparing the methods adopted in different gas-cooled reactor cores, inspiration for our own HTR seismic research was achieved. Results and Conclusions: In this paper, the research ideas of graphite core seismic during the process of designing, constructing and operating HTR-10 are expounded. Also the project progress of HTR-PM and the research on side reflection with the theory of similarity is introduced. (authors)

  13. Nuclear reactor coolant and cover gas system

    International Nuclear Information System (INIS)

    George, J.A.; Redding, A.H.; Tower, S.N.

    1976-01-01

    A core cooling system is disclosed for a nuclear reactor of the type utilizing a liquid coolant with a cover gas above free surfaces of the coolant. The disclosed system provides for a large inventory of reactor coolant and a balanced low pressure cover gas arrangement. A flow restricting device disposed within a reactor vessel achieves a pressure of the cover gas in the reactor vessel lower than the pressure of the reactor coolant in the vessel. The low gas pressure is maintained over all free surfaces of the coolant in the cooling system including a coolant reservoir tank. Reactor coolant stored in the reservoir tank allows for the large reactor coolant inventory provided by the invention

  14. Measurements of neutron flux in the RA reactor

    International Nuclear Information System (INIS)

    Raisic, N.

    1961-12-01

    This report includes the following separate parts: Thermal neutron flux in the experimental channels od RA reactor; Epithermal neutron flux in the experimental channels od RA reactor; Fast neutron flux in the experimental channels od RA reactor; Thermal neutron flux in the thermal column and biological experimental channel; Neutronic measurements in the RA reactor cell; Temperature reactivity coefficient of the RA reactor; design of the device for measuring the activity of wire [sr

  15. Nuclear reactor building

    Energy Technology Data Exchange (ETDEWEB)

    Oshima, Nobuaki.

    1991-08-09

    The secondary container in a nuclear reactor building is made of a transparent structure having a shielding performance such as lead glass, by which the inside of the secondary container can be seen without undergoing radiation exposure. In addition, an operator transportation facility capable of carrying about 5 to 10 operators at one time is disposed, and the side of the facility on the secondary container is constituted with a transparent material such as glass, to provide a structure capable of observing the inside of the secondary container. The ventilation and air conditioning in the operator's transportation facility is in communication with the atmosphere of a not-controlled area. Accordingly, operators at the outside of the reactor building can reach the operator's transportation facility without taking and procedures for entering the controlled area and without undergoing radiation exposure. The inside of the secondary container in the reactor building can be seen from various directions through the transparent structure having the shielding performance. (N.H.).

  16. Nuclear reactor sealing system

    International Nuclear Information System (INIS)

    McEdwards, J.A.

    1983-01-01

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

  17. SP-100 reactor design

    International Nuclear Information System (INIS)

    Armijo, J.S.; Atwell, J.; Pluta, P.R.; Smith, M.A.; Solorzano, E.R.

    1987-01-01

    The SP-100 space reactor power system is being designed and developed as part of the Ground Engineering System (GES) contract between General Electric Company as the system developer and the Department of Energy. Other key participants in the GES program include Westinghouse Hanford Company (site operator), Los Alamos National Laboratory (fuel development and production), Oak Ridge National Laboratory (materials), and Westinghouse, Advanced Energy Systems Division (shield, HTS equipment). The GES Program includes two major elements. First, the development of a Reference Flight System design at 100 kWe output to the user, and second the validation of the Reference Flight System design by analysis and by testing. Development of key technologies along with component and system testing is an essential part of the validation program. The nuclear subsystem validation includes the design, manufacture, assembly and operational testing of a Ground Reactor Test Assembly. The subject of this paper is the reactor design for the Reference Flight System. The reference flight design is in the preliminary design stage and will evolve over the next year

  18. Concerning modification of plan for installation of reactors (addition of No.4 reactor and modification of No.1, No.2 and No.3 reactor facilities) at Hamaoka nuclear power plant of Chubu Electric Power Co., Ltd. (reply to inquiry)

    International Nuclear Information System (INIS)

    1988-01-01

    The Nuclear Safety Commission on Oct. 22, 1987 directed the Nuclear Reactor Safety Expert Group to carry out a study on it, made deliberations after receiving a report from the Group on Jun. 28, and submitted the findings to the Minister of International Trade and Industry on Jul. 14. The study and deliberations were intended to determine the conformity of the modifications to the applicable laws relating to control of nuclear material, nuclear fuel and nuclear reactor. The investigation on the location covered the site conditions, geology, seismic environment, meteorology, hydrology, and social environment. The investigations on the safety design of the reactor facilities addressed the design of the overall reactor facilities, anti-earthquake design, and design of other equipment including reactor core, fuel equipment, monitors, reactivity controllers, safety protection equipment, and emergency core cooler. Other investigations included exposure evaluation and accident analysis. It was concluded that the modifications would not have adverse effects on the safety of the reactor facilities. (Nogami, K.)

  19. Gas Reactor International Cooperative Program. Interim report. Safety and licensing evaluaion of German Pebble Bed Reactor concepts

    International Nuclear Information System (INIS)

    1978-09-01

    The Pebble Bed Gas Cooled Reactor, as developed in the Federal Republic of Germany, was reviewed from a United States Safety and Licensing perspective. The primary concepts considered were the steam cycle electric generating pebble bed (HTR-K) and the process heat pebble bed (PNP), although generic consideration of the direct cycle gas turbine pebble bed (HHT) was included. The study examines potential U.S. licensing issues and offers some suggestions as to required development areas

  20. Gas Reactor International Cooperative Program. Interim report. Safety and licensing evaluaion of German Pebble Bed Reactor concepts

    Energy Technology Data Exchange (ETDEWEB)

    1978-09-01

    The Pebble Bed Gas Cooled Reactor, as developed in the Federal Republic of Germany, was reviewed from a United States Safety and Licensing perspective. The primary concepts considered were the steam cycle electric generating pebble bed (HTR-K) and the process heat pebble bed (PNP), although generic consideration of the direct cycle gas turbine pebble bed (HHT) was included. The study examines potential U.S. licensing issues and offers some suggestions as to required development areas.

  1. Manual for the operation of research reactors

    International Nuclear Information System (INIS)

    1965-01-01

    The great majority of the research reactors in newly established centres are light-water cooled and are often also light-water moderated. Consequently, the IAEA has decided to publish in its Technical Reports Series a manual dealing with the technical and practical problems associated with the safe and efficient operation of this type of reactor. Even though this manual is limited to light-water reactors in its direct application and presents the practices and experience at one specific reactor centre, it may also be useful for other reactor types because of the general relevance of the problems discussed and the long experience upon which it is based. It has, naturally, no regulatory character but it is hoped that it will be found helpful by staff occupied in all phases of the practical operation of research reactors, and also by those responsible for planning their experimental use. 23 refs, tabs

  2. Recent progress in stellarator reactor conceptual design

    International Nuclear Information System (INIS)

    Miller, R.L.

    1985-01-01

    The Stellarator/Torsatron/Heliotron (S/T/H) class of toroidal magnetic fusion reactor designs continues to offer a distinct and in several ways superior approach to eventual commercial competitiveness. Although no major, integrated conceptual reactor design activity is presently underway, a number of international research efforts suggest avenues for the substantial improvement of the S/T/H reactor embodiment, which derive from recent experimental and theoretical progress and are responsive to current trends in fusion-reactor projection to set the stage for a third generation of designs. Recent S/T/H reactor design activity is reviewed and the impact of the changing technical and programmatic context on the direction of future S/T/H reactor design studies is outlined

  3. MELCOR development for existing and advanced reactors

    International Nuclear Information System (INIS)

    Summers, R.M.

    1993-01-01

    Recent efforts in MELCOR development to address previously identified deficiencies have resulted in release of MELCOR 1.8.2, a much-improved version of the code. Major new models have been implemented for direct containment heating, ice condensers, debris quenching, lower plenum debris behavior, core materials interactions' and radial relocation of debris. Significant improvements have also been made in the modeling of interfacial momentum exchange and in the modeling of fission product release, condensation/evaporation, and aerosol behavior. Efforts are underway to address two-phase hydrodynamics difficulties, to improve modeling of water condensation on structures and fine-scale natural circulation within the reactor vessel, and to implement CORCON-Mod3. Improvements are also being made to MELCOR's capability to handle new features of the advanced light water reactor designs, including drainage of water films on connected heat structures, heat transfer from the external surface of the reactor vessel to a flooded cavity, and creep rupture failure of the lower head. Additional development needs in other areas are discussed

  4. Needs and Requirements for Future Research Reactors (ORNL Perspectives)

    International Nuclear Information System (INIS)

    Ilas, Germina; Bryan, Chris; Gehin, Jess C.

    2016-01-01

    The High Flux Isotope Reactor (HFIR) is a vital national and international resource for neutron science research, production of radioisotopes, and materials irradiation. While HFIR is expected to continue operation for the foreseeable future, interest is growing in understanding future research reactors features, needs, and requirements. To clarify, discuss, and compile these needs from the perspective of Oak Ridge National Laboratory (ORNL) research and development (R&D) missions, a workshop, titled ''Needs and Requirements for Future Research Reactors'', was held at ORNL on May 12, 2015. The workshop engaged ORNL staff that is directly involved in research using HFIR to collect valuable input on the reactor's current and future missions. The workshop provided an interactive forum for a fruitful exchange of opinions, and included a mix of short presentations and open discussions. ORNL staff members made 15 technical presentations based on their experience and areas of expertise, and discussed those capabilities of the HFIR and future research reactors that are essential for their current and future R&D needs. The workshop was attended by approximately 60 participants from three ORNL directorates. The agenda is included in Appendix A. This document summarizes the feedback provided by workshop contributors and participants. It also includes information and insights addressing key points that originated from the dialogue started at the workshop. A general overview is provided on the design features and capabilities of high performance research reactors currently in use or under construction worldwide. Recent and ongoing design efforts in the US and internationally are briefly summarized, followed by conclusions and recommendations.

  5. Pseudo-harmonics method: an application to thermal reactors

    International Nuclear Information System (INIS)

    Silva, F.C. da; Rotenberg, S.; Thome Filho, Z.D.

    1985-10-01

    Several applications of the Pseudo-Harmonics method are presented, aiming to calculate the neutron flux and the perturbed eigenvalue of a nuclear reactor, like PWR, with three enrichment regions as Angra-1 reactor. In the reference reactor, perturbations of several types as global as local were simulated. The results were compared with those from the direct calculation. (E.G.) [pt

  6. US DOE Idaho national laboratory reactor decommissioning

    International Nuclear Information System (INIS)

    Szilagyi, Andrew

    2012-01-01

    The United States Department of Energy (DOE) primary contractor, CH2M-WG Idaho was awarded the cleanup and deactivation and decommissioning contract in May 2005 for the Idaho National Lab (INL). The scope of this work included dispositioning over 200 Facilities and 3 Reactors Complexes (Engineering Test Reactor (ETR), Materials Test Reactor (MTR) and Power Burst Facility (PBF) Reactor). Two additional reactors were added to the scope of the contract during the period of performance. The Zero Power Physics Reactor (ZPPR) disposition was added under a separate subcontractor with the INL lab contractor and the Experimental Breeder Reactor II (EBR-II) disposition was added through American Recovery and Reinvestment Act (ARRA) Funding. All of the reactors have been removed and disposed of with the exception of EBR-II which is scheduled for disposition approximately March of 2012. A brief synopsis of the 5 reactors is provided. For the purpose of this paper the ZPPR reactor due to its unique design as compared to the other four reactors, and the fact that is was relatively lightly contaminated and irradiated will not be discussed with the other four reactors. The ZPPR reactor was readily accessible and was a relatively non-complex removal as compared to the other reactors. Additionally the EBR-II reactor is currently undergoing D and D and will have limited mention in this paper. Prior to decommissioning the reactors, a risk based closure model was applied. This model exercised through the Comprehensive Environmental Response, Compensation and Liability Act (CERCLA), Non-Time Critical Removal Action (NTCRA) Process which evaluated several options. The options included; No further action - maintain as is, long term stewardship and monitoring (mothball), entombment in place and reactor removal. Prior to commencing full scale D and D, hazardous constituents were removed including cadmium, beryllium, sodium (passivated and elemental), PCB oils and electrical components, lead

  7. Man Machine Interface Analysis For The Equipment And Instrumentation In MCR Of The TRIGA Mark II Modified Reactor

    International Nuclear Information System (INIS)

    Itjeu K; Piping S; Suharyo W; Kussigit S; Darlis; Bambang S; Sasongko Y

    2000-01-01

    Procedure and function of installation devices have been reviewed for the equipment and instrumentation in Main Control Room of TRIGA MARK II Modified Reactor. Analysis reviews carried out from the operation data/performances of the equipment and instrumentation of TRIGA II modified reactor and then evaluated. Instrumentation system and reactor control of TRIGA II is the reactor control system controlled by microprocessor installed at TM-II reactor to change analog system. These system consists of 4 parts that in such are control system console (CSC), data acquisition control unit (DAC), power monitor channel and temperature monitor. Beside that the system provided by supporting devices, for instances screen monitor, relay panel, analog and recorder includes analog input terminals and relay. Based on analysis reviewed from the procedures and function of installation devices at the obtaining system have been found that the display equipment enable to analyze reactor safety, enable to show operation condition directly and to perform information target to the each functions. Control equipment are consistent to the role of colour codes, shape and size, and also control classification fit to the reactor safety requirements priority. Alarm devices enable to transfer information promptly from the abnormal accidents and includes its anticipation. The early warning system (alarm, screen features) will show texts warning or warning signs if the reactor system is in the abnormal condition and then the operators will do proper needed immediately

  8. U.S. Domestic Reactor Conversion Programs

    International Nuclear Information System (INIS)

    Woolstenhulme, Eric

    2008-01-01

    The Conversion Projects Include: the revision of the facilities safety basis documents and supporting analysis, the fabrication of new LEU fuel, the change-out of the reactor core, and the removal of the used HEU fuel (by INL University Fuels Program or DOE-NE). The major entities involved are: the U.S. Nuclear Regulatory Commission, the University reactor department, the fuel and hardware fabricators, the Spent fuel receipt facilities, the Spent fuel shipping services, and the U.S. Department of Energy and their subcontractors. Three major Reactor Conversion Program milestones have been accomplished since 2006: the conversion of the TRIGA reactor at Texas A and M University Nuclear Science Center, the conversion of the University of Florida Training Reactor, and the conversion of the Purdue University Reactor. Four Reactor Conversion Program milestones yet to be accomplished in 2008 and 2009: the Washington State University Nuclear Radiation Center reactor, the Oregon State University TRIGA Reactor, the University of Wisconsin Nuclear Reactor, and the Neutron Radiography Reactor Facility. NNSA is committed to doing things cheaper, better, smarter, safer through a 'Lessons Learned' process. The conversion team assessed each major activity grouping: Project Initiation, Conversion Proposal Development, Fuel Fabrication and Hardware, Core Conversion, and Spent Nuclear Fuel Removal. Issues were identified and recommendations were given

  9. Review of Operation and Maintenance Support Systems for Research Reactors

    International Nuclear Information System (INIS)

    Jin, Kyungho; Heo, Gyunyoung; Park, Jaekwan

    2014-01-01

    Operation support systems do not directly control the plant but it can aid decision making itself by obtaining and analyzing large amounts of data. Recently, the demand of research reactor is growing and the need for operation support systems is increasing, but it has not been applied for research reactors. This study analyzes operation and maintenance support systems of NPPs and suggests appropriate systems for research reactors based on analysis. In this paper, operation support systems for research reactors are suggested by comparing with those of power reactors. Currently, research reactors do not cover special systems in order to improve safety and operability in comparison with power reactors. Therefore we expect to improve worth to use by introducing appropriate systems for research reactors. In further research, we will develop an appropriate system such as applications or tools that can be applied to the research reactor

  10. NOx reduction by ozone injection and direct plasma treatment

    DEFF Research Database (Denmark)

    Stamate, Eugen; Salewski, Mirko

    2012-01-01

    NOx reduction by ozone injection and direct plasma treatment is investigated for different process parameters in a 6 m long serpentine reactor. Several aspects including the role of mixing scheme, water vapours, steep temperature gradient and time dependet NOx levels are taken into consideration...

  11. Status of fast breeder reactor development in the United States of America - April 1985

    International Nuclear Information System (INIS)

    Horton, K.E.

    1986-01-01

    In order to provide a continuum of development for liquid metal reactors, the U.S. program has been reshaped into two portions -- advanced converter reactor technology including the Liquid Metal Reactor for the near and intermediate term, and the Liquid Metal Fast Breeder Reactor for deployment in the twenty-first century. The focused research and development program is directed at innovative means to improve economics, provide inherent safety, and to meet the needs of the ultimate user, the utilities. Work in the fuels and materials, and reprocessing areas is being continued to support eventual deployment. A major factor in successful deployment will be the effectiveness of international collaboration in reducing costs and duplication of efforts

  12. Nuclear reactor

    International Nuclear Information System (INIS)

    Schabert, H.P.; Weber, R.; Bauer, A.

    1975-01-01

    The refuelling of a PWR power reactor of about 1,200 MWe is performed by a transport pipe in the containment leading from an external to an internal fuel pit. A wagon to transport the fuel elements can go from a vertical loading position to an also vertical deloading position in the inner fuel pit via guide rollers. The necessary horizontal movement is effected by means of a cable line through the transport pipe which is inclined at least 10 0 . Gravity thus helps in the movement to the deloading position. The cable line with winch is fastened outside the containment. Swivelling devices tip the wagon from the horizontal to the vertical position or vice versa. Loading and deloading are done laterally. (TK/LH) [de

  13. NEUTRONIC REACTOR

    Science.gov (United States)

    McGarry, R.J.

    1958-04-22

    Fluid-cooled nuclear reactors of the type that utilize finned uranium fuel elements disposed in coolant channels in a moderater are described. The coolant channels are provided with removable bushings composed of a non- fissionable material. The interior walls of the bushings have a plurality of spaced, longtudinal ribs separated by grooves which receive the fins on the fuel elements. The lands between the grooves are spaced from the fuel elements to form flow passages, and the size of the now passages progressively decreases as the dlstance from the center of the core increases for the purpose of producing a greater cooling effect at the center to maintain a uniform temperature throughout the core.

  14. Fast reactor database. 2006 update

    International Nuclear Information System (INIS)

    2006-12-01

    plants. The recurring themes are the selection and summary of the data associated with the choice of coolant, fuel and structural materials, reduction of the steel weight, simplification of the plant design/layout, other important fast reactor design issues, and how to solve these problems. In the field of fast reactor design and operational data, the last reference document published by the IAEA was the 1996 Fast Reactor Database (IAEA-TECDOC-866). Since its publication, quite a lot has happened: the construction of two new reactors has been launched, and conceptual/design studies were initiated for various fast reactors, e.g. the Japanese JSFR-1500 and the Russian BN-1800 (both cooled by sodium), as well as for a wholly new line of LMFR concepts - modular reactors cooled by sodium and by lead-bismuth alloy, and prototype and demonstration commercial size fast reactors cooled by lead. The data were produced by the IAEA's Technical Working Group on Fast Reactors (TWG-FR). For many of the TWG-FR Member States there is a significant history of fast reactor development, often extending over a period of 40+ years. The new and updated information on LMFR, which are in operation, under construction or development, has been prepared with contributions from China, India, Japan, Republic of Korea and the Russian Federation. The information contained in IAEA-TECDOC-866, produced by France, Germany, Italy, the UK and the USA, was included in the present report with some modification taking into account last events

  15. Particle Bed Reactor scaling relationships

    International Nuclear Information System (INIS)

    Slovik, G.; Araj, K.; Horn, F.L.; Ludewig, H.; Benenati, R.

    1987-01-01

    Scaling relationships for Particle Bed Reactors (PBRs) are discussed. The particular applications are short duration systems, i.e., for propulsion or burst power. Particle Bed Reactors can use a wide selection of different moderators and reflectors and be designed for such a wide range of power and bed power densities. Additional design considerations include the effect of varying the number of fuel elements, outlet Mach number in hot gas channel, etc. All of these variables and options result in a wide range of reactor weights and performance. Extremely light weight reactors (approximately 1 kg/MW) are possible with the appropriate choice of moderator/reflector and power density. Such systems are very attractive for propulsion systems where parasitic weight has to be minimized

  16. Jules Horowitz Reactor, basic design

    International Nuclear Information System (INIS)

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

    2003-01-01

    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)

  17. Jules Horowitz reactor, basic design

    International Nuclear Information System (INIS)

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

    2002-01-01

    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. (author)

  18. Combustion synthesis continuous flow reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-01-06

    The present invention is a reactor for combustion synthesis of inorganic powders. The reactor includes a reaction vessel having a length and a first end and a second end. The reaction vessel further has a solution inlet and a carrier gas inlet. The reactor further has a heater for heating both the solution and the carrier gas. In a preferred embodiment, the reaction vessel is heated and the solution is in contact with the heated reaction vessel. It is further preferred that the reaction vessel be cylindrical and that the carrier gas is introduced tangentially into the reaction vessel so that the solution flows helically along the interior wall of the reaction vessel. As the solution evaporates and combustion produces inorganic material powder, the carrier gas entrains the powder and carries it out of the reactor. 10 figs.

  19. Combustion synthesis continuous flow reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-01-01

    The present invention is a reactor for combustion synthesis of inorganic powders. The reactor includes a reaction vessel having a length and a first end and a second end. The reaction vessel further has a solution inlet and a carrier gas inlet. The reactor further has a heater for heating both the solution and the carrier gas. In a preferred embodiment, the reaction vessel is heated and the solution is in contact with the heated reaction vessel. It is further preferred that the reaction vessel be cylindrical and that the carrier gas is introduced tangentially into the reaction vessel so that the solution flows helically along the interior wall of the reaction vessel. As the solution evaporates and combustion produces inorganic material powder, the carrier gas entrains the powder and carries it out of the reactor.

  20. Overview of fusion reactor safety

    Science.gov (United States)

    Cohen, S.; Crocker, J. G.

    Use of deuterium-tritium fusion reactors requires examination of several major safety and environmental issues: (1) tritium inventory control; (2) neutron activation of structural materials, fluid streams and reactor hall environment; (3) release of radioactivity from energy sources including lithium spill reactions, superconducting magnet stored energy release, and plasma disruptions; (4) high magnetic and electromagnetic fields associated with fusion reactor superconducting magnets and radio frequency heating devices; and (5) handling and disposal of radioactive waste. Early recognition of potential safety problems with fusion reactors provides the opportunity for improvement in design and materials to eliminate or greatly reduce these problems. With an early start in this endeavor, fusion should be among the lower risk technologies for generation of commercial electrical power.

  1. Advanced Catalytic Hydrogenation Retrofit Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Reinaldo M. Machado

    2002-08-15

    Industrial hydrogenation is often performed using a slurry catalyst in large stirred-tank reactors. These systems are inherently problematic in a number of areas, including industrial hygiene, process safety, environmental contamination, waste production, process operability and productivity. This program proposed the development of a practical replacement for the slurry catalysts using a novel fixed-bed monolith catalyst reactor, which could be retrofitted onto an existing stirred-tank reactor and would mitigate many of the minitations and problems associated with slurry catalysts. The full retrofit monolith system, consisting of a recirculation pump, gas/liquid ejector and monolith catalyst, is described as a monolith loop reactor or MLR. The MLR technology can reduce waste and increase raw material efficiency, which reduces the overall energy required to produce specialty and fine chemicals.

  2. Overview of fusion reactor safety

    International Nuclear Information System (INIS)

    Cohen, S.; Crocker, J.G.

    1981-01-01

    Use of deuterium-tritium burning fusion reactors requires examination of several major safety and environmental issues: (1) tritium inventory control, (2) neutron activation of structural materials, fluid streams and reactor hall environment, (3) release of radioactivity from energy sources including lithium spill reactions, superconducting magnet stored energy release, and plasma disruptions, (4) high magnetic and electromagnetic fields associated with fusion reactor superconducting magnets and radio frequency heating devices, and (5) handling and disposal of radioactive waste. Early recognition of potential safety problems with fusion reactors provides the opportunity for improvement in design and materials to eliminate or greatly reduce these problems. With an early start in this endeavor, fusion should be among the lower risk technologies for generation of commercial electrical power

  3. Compact containment boiling water reactor (CCR)

    International Nuclear Information System (INIS)

    2006-01-01

    investment risks and to facilitate public acceptance. To meet the output targets, the power range chosen for the LSBWR is from 100 to 300 MW(e). To overcome the disadvantages of economies of scale for small sized reactors, the following design approaches have been used for the LSBWR: - Simplification of systems by combining direct cycle and natural circulation and by relying on passive safety means. - Simplification of structure by integration of the reactor building and the turbine building. - Elimination of the fuel pool and refueling machine (for a 15-year operation cycle). - Using a modular design to ensure a short construction period. - Adoption of a seismic isolation and a ship hull structure (marine type protective enclosure structure). - Improvement in availability by implementing longer operating cycles (3 to 5 years). The reactor concept presented herein (the CCR) takes a follow-up on the LSBWR. It has a small power output, the capability of long operating cycles and a simplified and compact BWR type configuration with comprehensive safety features. To be economically competitive, the CCR design includes simplification of systems and compact structure, modular structures for short construction periods and improved availability. For comprehensive safety, the CCR targets to eliminate off-site emergency planning by using highly reliable equipment and systems, such as a reactor pressure vessel (RPV) with large inventory and the core being located at its bottom, the capability of in-vessel retention of a core melt (IVR) and a compact pressurized containment vessel (PCV) with high-integrity, incorporating several passive features. The CCR concept is to provide economic flexibility for a variety of site conditions and electricity demands, to mitigate investment risks, and to facilitate public acceptance. The principal stakeholders are: - Toshiba Corporation. - The Japan Atomic Power Company (JAPC)

  4. Decommissioning a nuclear reactor

    International Nuclear Information System (INIS)

    Montoya, G.M.

    1991-01-01

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

  5. Transient analysis of space reactors

    International Nuclear Information System (INIS)

    Giap, H.Q.; Best, F.

    1985-01-01

    Nuclear reactors are being considered as the primary energy source for an increasing number of space missions including the National Aeronautics and Space Administration's space station and the Strategic Defense Initiative's SP-100. Conceptual reactor designs include fast and thermal spectrum reactors with liquid-metal or gas cooling delivering thermal energy to Brayton, Stirling, thermoelectric, thermoionic, or Rankine cycle power conversion systems. Almost all designs utilize a thermal radiator for waste heat rejection, although the radiator may assume such exotic forms as fans of liquid-metal droplets or metal filaments floating in free space. The reactor, power conversion, and heat rejection systems must operate in zero gravity, be uniformly small and lightweight, and utilize exotic materials compared with terrestrial technology. Therefore, the problems of system design and analysis are unique to space applications, and new analysis tools must be developed. With this motivation, the Transient Analysis of Space Reactors (TASR) computer code has been developed to predict system performance under steady-state and transient conditions. The SP-100 space reactor system, being designed by General Electric (GE) Company, was used as the test system for the TASR code. At full power, the SP-100 produces 110 kW(e) from 2.21 MW(t) via a thermoelectric conversion system

  6. Development and design of control rod drive mechanisms for pressurized water reactors

    International Nuclear Information System (INIS)

    Leme, Francisco Louzano

    2003-01-01

    The Control Rod Drive Mechanisms (CRDM) for a Pressurized Water Reactor (PWR) are equipment, integrated to the reactor pressure vessel, incorporating mechanical and electrical components designed to move and position the control rods to guarantee the control of power and shutdown of the nuclear reactor, during normal operation, either in emergency or accidental situations. The type of CRDM used in PWR reactors, whose detailed individual description will be presented in this monograph are the Roller-Nut and Magnetic-Jack. The environment, where the CRDM performs its above presented operational functions, includes direct contact with the fluid used as coolant peculiar to the interior of the reactor, and its associated chemical characteristics, the radiation field next to the reactor core, and also the temperature and pressure in the reactor pressure vessel. So the importance of the CRDM design requirements related to its safety functions are emphasized. Finally, some aspects related to the mechanical and structural design of CRDM of a case study, considering the CRDM for a PWR from the experimental nuclear plant to be applied by CTMSP (Centro Tecnologico da Marinha em Sao Paulo), are pointed out. The design and development of these equipment (author)

  7. A 50-100 kWe gas-cooled reactor for use on Mars.

    Energy Technology Data Exchange (ETDEWEB)

    Peters, Curtis D. (.)

    2006-04-01

    In the space exploration field there is a general consensus that nuclear reactor powered systems will be extremely desirable for future missions to the outer solar system. Solar systems suffer from the decreasing intensity of solar radiation and relatively low power density. Radioisotope Thermoelectric Generators are limited to generating a few kilowatts electric (kWe). Chemical systems are short-lived due to prodigious fuel use. A well designed 50-100 kWe nuclear reactor power system would provide sufficient power for a variety of long term missions. This thesis will present basic work done on a 50-100 kWe reactor power system that has a reasonable lifespan and would function in an extraterrestrial environment. The system will use a Gas-Cooled Reactor that is directly coupled to a Closed Brayton Cycle (GCR-CBC) power system. Also included will be some variations on the primary design and their effects on the characteristics of the primary design. This thesis also presents a variety of neutronics related calculations, an examination of the reactor's thermal characteristics, feasibility for use in an extraterrestrial environment, and the reactor's safety characteristics in several accident scenarios. While there has been past work for space reactors, the challenges introduced by thin atmospheres like those on Mars have rarely been considered.

  8. A 50-100 kWe gas-cooled reactor for use on Mars

    International Nuclear Information System (INIS)

    Peters, Curtis D.

    2006-01-01

    In the space exploration field there is a general consensus that nuclear reactor powered systems will be extremely desirable for future missions to the outer solar system. Solar systems suffer from the decreasing intensity of solar radiation and relatively low power density. Radioisotope Thermoelectric Generators are limited to generating a few kilowatts electric (kWe). Chemical systems are short-lived due to prodigious fuel use. A well designed 50-100 kWe nuclear reactor power system would provide sufficient power for a variety of long term missions. This thesis will present basic work done on a 50-100 kWe reactor power system that has a reasonable lifespan and would function in an extraterrestrial environment. The system will use a Gas-Cooled Reactor that is directly coupled to a Closed Brayton Cycle (GCR-CBC) power system. Also included will be some variations on the primary design and their effects on the characteristics of the primary design. This thesis also presents a variety of neutronics related calculations, an examination of the reactor's thermal characteristics, feasibility for use in an extraterrestrial environment, and the reactor's safety characteristics in several accident scenarios. While there has been past work for space reactors, the challenges introduced by thin atmospheres like those on Mars have rarely been considered

  9. Fischer-Tropsch Slurry Reactor modeling

    Energy Technology Data Exchange (ETDEWEB)

    Soong, Y.; Gamwo, I.K.; Harke, F.W. [Pittsburgh Energy Technology Center, PA (United States)] [and others

    1995-12-31

    This paper reports experimental and theoretical results on hydrodynamic studies. The experiments were conducted in a hot-pressurized Slurry-Bubble Column Reactor (SBCR). It includes experimental results of Drakeol-10 oil/nitrogen/glass beads hydrodynamic study and the development of an ultrasonic technique for measuring solids concentration. A model to describe the flow behavior in reactors was developed. The hydrodynamic properties in a 10.16 cm diameter bubble column with a perforated-plate gas distributor were studied at pressures ranging from 0.1 to 1.36 MPa, and at temperatures from 20 to 200{degrees}C, using a dual hot-wire probe with nitrogen, glass beads, and Drakeol-10 oil as the gas, solid, and liquid phase, respectively. It was found that the addition of 20 oil wt% glass beads in the system has a slight effect on the average gas holdup and bubble size. A well-posed three-dimensional model for bed dynamics was developed from an ill-posed model. The new model has computed solid holdup distributions consistent with experimental observations with no artificial {open_quotes}fountain{close_quotes} as predicted by the earlier model. The model can be applied to a variety of multiphase flows of practical interest. An ultrasonic technique is being developed to measure solids concentration in a three-phase slurry reactor. Preliminary measurements have been made on slurries consisting of molten paraffin wax, glass beads, and nitrogen bubbles at 180 {degrees}C and 0.1 MPa. The data show that both the sound speed and attenuation are well-defined functions of both the solid and gas concentrations in the slurries. The results suggest possibilities to directly measure solids concentration during the operation of an autoclave reactor containing molten wax.

  10. New perspectives on reactor safety

    International Nuclear Information System (INIS)

    Avery, R.

    1986-01-01

    Over the past few years a number of changes and new perspectives have come about in our approach to reactor safety. These changes have occurred over a period of time extending from as long ago as 1975, when WASH-1400 came out representing the first major application of probabilistic risk analysis (PRA) to US reactor plants. The period of change has extended from that time to the present, and includes new areas of focus such as safety goals, source term studies, and severe accident policy statement and approaches, including the IDCOR Program. It has also included a greatly increased interest in inherent safety. These areas are discussed in this paper

  11. Operational safety and reactor life improvements of Kyoto University Reactor

    International Nuclear Information System (INIS)

    Utsuro, M.; Fujita, Y.; Nishihara, H.

    1990-01-01

    Recent important experience in improving the operational safety and life of a reactor are described. The Kyoto University Reactor (KUR) is a 25-year-old 5 MW light water reactor provided with two thermal columns of graphite and heavy water as well as other kinds of experimental facilities. In the graphite thermal column, noticeable amounts of neutron irradiation effects had accumulated in the graphite blocks near the core. Before the possible release of the stored energy, all the graphite blocks in the column were successfully replaced with new blocks using the opportunity provided by the installation of a liquid deuterium cold neutron source in the column. At the same time, special seal mechanisms were provided for essential improvements to the problem of radioactive argon production in the column. In the heavy-water thermal column we have accomplished the successful repair of a slow leak of heavy water through a thin instrumentation tube failure. The repair work included the removal and reconstructions of the lead and graphite shielding layers and welding of the instrumentation tube under radiation fields. Several mechanical components in the reactor cooling system were also exchanged for new components with improved designs and materials. On-line data logging of almost all instrumentation signals is continuously performed with a high speed data analysis system to diagnose operational conditions of the reactor. Furthermore, through detailed investigations on critical components, operational safety during further extended reactor life will be supported by well scheduled maintenance programs

  12. Nuclear Reactor Safety; (USA)

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-01-01

    This publication announces on an monthly basis the current worldwide information available on all safety-related aspects of reactors, including: accident analysis, safety systems, radiation protection, decommissioning and dismantling, and security measures. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database (EDB) during the past month. Also included are other US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency's Technology Data Exchange, the International Atomic Energy Agency's International Nuclear Information System, or government-to-government agreements.

  13. FBR type reactor

    International Nuclear Information System (INIS)

    Kimura, Kimitaka; Fukuie, Ken; Iijima, Tooru; Shimpo, Masakazu.

    1994-01-01

    In an FBR type reactor for exchanging fuels by pulling up reactor core upper mechanisms, a connection mechanism is disposed for connecting the top of the reactor core and the lower end of the reactor core upper mechanisms. In addition, a cylindrical body is disposed surrounding the reactor core upper mechanisms, and a support member is disposed to the cylindrical body for supporting an intermediate portion of the reactor core upper mechanisms. Then, the lower end of the reactor core upper mechanisms is connected to the top of the reactor core. Same displacements are caused to both of them upon occurrence of earthquakes and, as a result, it is possible to eliminate mutual horizontal displacement between a control rod guide hole of the reactor core upper mechanisms and a control rod insertion hole of the reactor core. In addition, since the intermediate portion of the reactor core upper mechanisms is supported by the support member disposed to the cylindrical body surrounding the reactor core upper mechanisms, deformation caused to the lower end of the reactor core upper mechanisms is reduced, so that the mutual horizontal displacement with respect to the control rod insertion hole of the reactor core can be reduced. As a result, performance of control rod insertion upon occurrence of the earthquakes is improved, so that reactor shutdown is conducted more reliably to improve reactor safety. (N.H.)

  14. Advances in laser solenoid fusion reactor design

    International Nuclear Information System (INIS)

    Steinhauer, L.C.; Quimby, D.C.

    1978-01-01

    The laser solenoid is an alternate fusion concept based on a laser-heated magnetically-confined plasma column. The reactor concept has evolved in several systems studies over the last five years. We describe recent advances in the plasma physics and technology of laser-plasma coupling. The technology advances include progress on first walls, inner magnet design, confinement module design, and reactor maintenance. We also describe a new generation of laser solenoid fusion and fusion-fission reactor designs

  15. Computerized reactor pressure vessel materials information system

    International Nuclear Information System (INIS)

    Strosnider, J.; Monserrate, C.; Kenworthy, L.D.; Tether, C.D.

    1980-10-01

    A computerized information system for storage and retrieval of reactor pressure vessel materials data was established, as part of Task Action Plan A-11, Reactor Vessel Materials Toughness. Data stored in the system are necessary for evaluating the resistance of reactor pressure vessels to flaw-induced fracture. This report includes (1) a description of the information system; (2) guidance on accessing the system; and (3) a user's manual for the system

  16. Advanced designs of VVER reactor plant

    International Nuclear Information System (INIS)

    Mokhov, V.A.

    2010-01-01

    The history of VVER reactors, current challenges and approaches to the challenges are highlighted. The VVER-1200 reactor of 3+ generation for AES-2006 units are under construction at the Leningrad 2 nuclear power plant (LNPP-2). The main parameters are listed and details are presented of the vessel, steam generator, and improved fuel. The issue of the NPP safety is discussed. Additional topics include the MIR-1200 reactor unit, VVER-600, and VVER-SCP (Generation 4). (P.A.)

  17. Research reactors; Les piles de recherche

    Energy Technology Data Exchange (ETDEWEB)

    Kowarski, L. [Commissariat a l' Energie Atomique, Paris (France). Centre d' Etudes Nucleaires]|[Organisation europeenne pour la Recherche Nucleaire, Geneve (Switzerland)

    1955-07-01

    It brings together the techniques data which are involved in the discussion about the utility for a research institute to acquire an atomic reactor for research purposes. This type of decision are often taken by non-specialist people who can need a brief presentation of a research reactor and its possibilities in term of research before asking advises to experts. In a first part, it draws up a list of the different research programs which can be studied by getting a research reactor. First of all is the reactor behaviour and kinetics studies (reproducibility factor, exploration of neutron density, effect of reactor structure, effect of material irradiation...). Physical studies includes study of the behaviour of the control system, studies of neutron resonance phenomena and study of the fission process for example. Chemical studies involves the study of manipulation and control of hot material, characterisation of nuclear species produced in the reactor and chemical effects of irradiation on chemical properties and reactions. Biology and medicine research involves studies of irradiation on man and animals, genetics research, food or medical tools sterilization and neutron beams effect on tumour for example. A large number of other subjects can be studied in a reactor research as reactor construction material research, fabrication of radioactive sources for radiographic techniques or applied research as in agriculture or electronic. The second part discussed the technological considerations when choosing the reactor type. The technological factors, which are considered for its choice, are the power of the reactor, the nature of the fuel which is used, the type of moderator (water, heavy water, graphite or BeO) and the reflector, the type of coolants, the protection shield and the control systems. In the third part, it described the characteristics (place of installation, type of combustible and comments) and performance (power, neutron flux ) of already existing

  18. Reactor core lower support

    International Nuclear Information System (INIS)

    1981-01-01

    This patent refers to the lower support of a nuclear reactor core, and is intended for supporting each fuel assembly of the core and for distributing the primary coolant through these assemblies. It is composed of: - A first thick plate supporting the fuel assemblies. Vertical channels are machined in this plate directly facing each assembly for the passage of the primary fluid: - A second thin plate drilled with orifices, fixed under the first plate, with no space between them, and so positioned that each orifice is directly facing one of the channels. The section of the orifices diminishes from the centre of the plate towards its periphery. The second plate can also be constituted of an assembly of juxtaposed smaller plates, each small plate being secured to the first plate independently of the neighbouring plates [fr

  19. Development and application of reactor noise diagnostics

    Energy Technology Data Exchange (ETDEWEB)

    Karlsson, Joakim K.H

    1999-04-01

    A number of problems in reactor noise diagnostics have been investigated within the framework of the present thesis. The six papers presented cover three relatively different areas, namely the use of analytical calculations of the neutron noise in simple reactor models, some aspects of boiling water reactor (BWR) stability and diagnostics of core barrel motion in pressurized water reactors (PWRs). The noise induced by small vibrations of a strong absorber has been the subject of several previous investigations. For a conventional {delta}-function source model, the equations can not be linearized in the traditional manner. Thus, a new source model, which is called the {epsilon}/d model, was developed. The correct solution has been derived in the {epsilon}/d model for both 1-D and 2-D reactor models. Recently, several reactor diagnostic problems have occurred which include a control rod partially inserted into the reactor core. In order to study such problems, we have developed an analytically solvable, axially non-homogenous, 2-D reactor model. This model has also been used to study the noise induced by a rod maneuvering experiment. Comparisons of the noise with the results of different reactor kinetic approximations have yielded information on the validity of the approximations in this relatively realistic model. In case of an instability event in a BWR, the noise may consist of one or several co-existing modes of oscillation and besides the fundamental mode, a regional first azimuthal mode has been observed in e.g. the Swedish BWR Ringhals-1. In order to determine the different stability characteristics of the different modes separately, it is important to be able to decompose the noise into its mode constituents. A separation method based on factorisation of the flux has been attempted previously, but without success. The reason for the failure of the factorisation method is the presence of the local component of the noise and its axial correlation properties. In

  20. Development and application of reactor noise diagnostics

    International Nuclear Information System (INIS)

    Karlsson, Joakim K.H.

    1999-04-01

    A number of problems in reactor noise diagnostics have been investigated within the framework of the present thesis. The six papers presented cover three relatively different areas, namely the use of analytical calculations of the neutron noise in simple reactor models, some aspects of boiling water reactor (BWR) stability and diagnostics of core barrel motion in pressurized water reactors (PWRs). The noise induced by small vibrations of a strong absorber has been the subject of several previous investigations. For a conventional δ-function source model, the equations can not be linearized in the traditional manner. Thus, a new source model, which is called the ε/d model, was developed. The correct solution has been derived in the ε/d model for both 1-D and 2-D reactor models. Recently, several reactor diagnostic problems have occurred which include a control rod partially inserted into the reactor core. In order to study such problems, we have developed an analytically solvable, axially non-homogenous, 2-D reactor model. This model has also been used to study the noise induced by a rod maneuvering experiment. Comparisons of the noise with the results of different reactor kinetic approximations have yielded information on the validity of the approximations in this relatively realistic model. In case of an instability event in a BWR, the noise may consist of one or several co-existing modes of oscillation and besides the fundamental mode, a regional first azimuthal mode has been observed in e.g. the Swedish BWR Ringhals-1. In order to determine the different stability characteristics of the different modes separately, it is important to be able to decompose the noise into its mode constituents. A separation method based on factorisation of the flux has been attempted previously, but without success. The reason for the failure of the factorisation method is the presence of the local component of the noise and its axial correlation properties. In the paper

  1. OECD Halden Reactor Project

    International Nuclear Information System (INIS)

    1988-01-01

    The OECD Halden Reactor project is an agreement between OECD member countries. It was first signed in 1958 and since then regularly renewed every third year. The activities at the Project is centred around the Halden heavy water rector, the HBWR. The reseach programme comprizes studies of fuel performance under various operating conditions, and the application of computers for process control. The HBWR is equipped for exposing fuel rods to temperatures and pressures, and at heat ratings met in modern BWR's and PWR's. A range of in-core instruments are available, permitting detailed measurements of the reactions of the fuel, including mechanical deformations, thermal behaviour, fission gas release, and corrosion. In the area of computer application, the studies of the communication between operator and process, and the surveillance and control of the reactor core, are of particular interst for reactor operation. 1988 represents the 30th year since the Project was started, and this publication is produced to mark this event. It gives and account of the activities and achievements of the Project through the years 1958-1988

  2. Tritium management in fusion reactors

    International Nuclear Information System (INIS)

    Galloway, T.R.

    1978-05-01

    This is a review paper covering the key environmental and safety issues and how they have been handled in the various magnetic and inertial confinement concepts and reference designs. The issues treated include: tritium accident analyses, tritium process control, occupational safety, HTO formation rate from the gas-phase, disposal of tritium contaminated wastes, and environmental impact--each covering the Joint European Tokamak (J.E.T. experiment), Tokamak Fusion Test Reactor (TFTR), Russian T-20, The Next Step (TNS) designs by Westinghouse/ORNL and General Atomic/ANL, the ANL and ORNL EPR's, the G.A. Doublet Demonstration Reactor, the Italian Fintor-D and the ORNL Demo Studies. There are also the following full scale plant reference designs: UWMAK-III, LASL's Theta Pinch Reactor Design (RTPR), Mirror Fusion Reactor (MFR), Tandem Mirror Reactor (TMR), and the Mirror Hybrid Reactor (MHR). There are four laser device breakeven experiments, SHIVA-NOVA, LLL reference designs, ORNL Laser Fusion power plant, the German ''Saturn,'' and LLL's Laser Fusion EPR I and II

  3. Evaluation of Torsatrons as reactors

    International Nuclear Information System (INIS)

    Lyon, J.F.; Gulec, K.; Miller, R.L.; El-Guebaly, L.

    1994-03-01

    Stellarators have significant operational advantages over tokamaks as ignited steady-state reactors. This scoping study, which uses an integrated cost-minimization code that incorporates costing and reactor component models self-consistently with a 1-D energy transport calculation, shows that a torsatron reactor could also be economically competitive with a tokamak reactor. The projected cost of electricity (COE) estimated using the Advanced Reactor Innovation and Evaluation Studies (ARIES) costing algorithms is 65.6 mill/kW(e)h in constant 1992 dollars for a reference 1-GW(e) Compact Torsatron reactor case. The COE is relatively insensitive (<10% variation) over a wide range of assumptions, including variations in the maximum field allowed on the coils, the coil elongation, the shape of the density profile, the beta limit, the confinement multiplier, and the presence of a large loss region for alpha particles. The largest variations in the COE occur for variations in the electrical power output demanded and the plasma-coil separation ratio

  4. Environmental aspects of fusion reactors

    International Nuclear Information System (INIS)

    Coffman, F.E.; Williams, J.M.

    1975-01-01

    With the continued depletion of fossil and uranium resources in the coming decades, the U. S. will be forced to look more toward renewable energy resources (e.g., wind, tidal, geothermal, and solar power) and toward such longer-term and nondepletable energy resources as fissile fast breeder reactors and fusion power. Several reference reactor designs have been completed for full-scale fusion power reactors that indicate that the environmental impacts from construction, operation, and eventual decommissioning of fusion reactors will be quite small. The principal environmental impact from fusion reactor operation will be from thermal discharges. Some of the safety and environmental characteristics that make fusion reactors appear attractive include an effectively infinite fuel supply at low cost, inherent incapability for a ''nuclear explosion'' or a ''nuclear runaway,'' the absence of fission products, the flexibility of selecting low neutron-cross-section structural materials so that emergency core cooling for a loss-of-coolant or other accident will not be necesary, and the absence of special nuclear materials such as 235 U or 239 Pu, so that diversion of nuclear weapons materials will not be possible and nuclear blackmail will not be a serious concern

  5. Nuclear Reactor Engineering Analysis Laboratory

    International Nuclear Information System (INIS)

    Carlos Chavez-Mercado; Jaime B. Morales-Sandoval; Benjamin E. Zayas-Perez

    1998-01-01

    The Nuclear Reactor Engineering Analysis Laboratory (NREAL) is a sophisticated computer system with state-of-the-art analytical tools and technology for analysis of light water reactors. Multiple application software tools can be activated to carry out different analyses and studies such as nuclear fuel reload evaluation, safety operation margin measurement, transient and severe accident analysis, nuclear reactor instability, operator training, normal and emergency procedures optimization, and human factors engineering studies. An advanced graphic interface, driven through touch-sensitive screens, provides the means to interact with specialized software and nuclear codes. The interface allows the visualization and control of all observable variables in a nuclear power plant (NPP), as well as a selected set of nonobservable or not directly controllable variables from conventional control panels

  6. Damping of reactor internals

    International Nuclear Information System (INIS)

    Singleton, N.R.; Bohm, G.J.

    1977-01-01

    This paper presents and discusses the results of a study of internals damping using data obtained from wave analysis of PWR plant flow tests, and shaker tests. The damping values were obtained from vibration data taken during the pre-operational testing of several reactor plants and some in air shaker tests. Parameters available in the data include the presence of the core, the presence and position of the control rod drive line, reactor coolant temperature, and combination of reactor coolant pumps in operation. Modal damping values for the structures reported were obtained from the decay of autocorrellograms or from the modal response half-power bandwidths of frequency spectra. For the lower frequency core barrel-reactor vessel beam modes 2% to 5% damping values were found for minimum damping values. Significantly higher values are found in the data when, for example, intermittent contact occurs at the core barrel level supports. Core barrel and thermal shield shell modes having natural frequencies in the frequency range of interest for seismic response calculations exhibit damping values generally on the order of 1% to 2%. Higher frequency, very low amplitude, shell modes of these structures can have damping values of less than 1%. Damping values for guide tubes were found to have minimum values of 2% to 5% depending on their core location. The cross flow velocity and thus the floor turbulence excited amplitude is higher for guide tubes in core locations near the outlet nozzles. Information on the damping of upper support columns which are similarly excited is also given. Damping values reported are suitable for normal operation design conditions, i.e., for oscillatory behavior and relatively small amplitudes. The extrapolation of the data to obtain realistic values for large seismic events and for loss of coolant accidents is also discussed

  7. Partially ionized plasmas including the third symposium on uranium plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Krishnan, M. [ed.

    1976-09-01

    Separate abstracts are included for 28 papers on electrically generated plasmas, fission generated plasmas, nuclear pumped lasers, gaseous fuel reactor research, and applications. Five papers have been previously abstracted and included in ERA.

  8. Fast breeder reactors

    International Nuclear Information System (INIS)

    Heinzel, V.

    1975-01-01

    The author gives a survey of 'fast breeder reactors'. In detail the process of breeding, the reasons for the development of fast breeders, the possible breeder reactors, the design criteria, fuels, cladding, coolant, and safety aspects are reported on. Design data of some experimental reactors already in operation are summarized in stabular form. 300 MWe Prototype-Reactors SNR-300 and PFR are explained in detail and data of KWU helium-cooled fast breeder reactors are given. (HR) [de

  9. Determining Reactor Neutrino Flux

    OpenAIRE

    Cao, Jun

    2011-01-01

    Flux is an important source of uncertainties for a reactor neutrino experiment. It is determined from thermal power measurements, reactor core simulation, and knowledge of neutrino spectra of fuel isotopes. Past reactor neutrino experiments have determined the flux to (2-3)% precision. Precision measurements of mixing angle $\\theta_{13}$ by reactor neutrino experiments in the coming years will use near-far detector configurations. Most uncertainties from reactor will be canceled out. Understa...

  10. Nuclear reactor coolant channels

    International Nuclear Information System (INIS)

    Macbeth, R.V.

    1978-01-01

    A nuclear reactor coolant channel is described that is suitable for sub-cooled reactors as in pressurised water reactors as well as for bulk boiling, as in boiling water reactors and steam generating nuclear reactors. The arrangement aims to improve heat transfer between the fuel elements and the coolant. Full constructional details are given. See also other similar patents by the author. (U.K.)

  11. 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...

  12. Updating reactor control: mini-computers

    International Nuclear Information System (INIS)

    Crawford, K.C.; Sandquist, G.M.

    1984-01-01

    An aging reactor control console and a limited operating budget have impeded many research projects in the TRIGA reactor facility at the University of Utah. The, University's present console is Circa 1959 vintage and repairs to the console are frequently required which present many electronic problems to a staff with little electronic training. As an alternative to a single function control console we are developing a TRIGA control system based upon a mini-computer. The system hardware has been specified and the hardware is currently being acquired. The software will be programmed by the staff to customize the system to the reactor's physical systems and technical specifications. The software will be designed to monitor and control all reactor functions, control a pneumatic sample transfer system, acquire and analyze neutron activation data, provide reactor facility security surveillance, provide reactor documentation including online logging of physical parameters, and record regularly scheduled reactor calibrations and laboratory accounting procedures. The problem of hardware rewiring and changing technical specifications and changing safety system characteristics can be easily handled in the software. Our TRIGA reactor also functions as a major educational resource using available reactor based software. The computer control system can be employed to provide on-line training in reactor physics and kinetics. (author)

  13. Scanning tunneling microscope assembly, reactor, and system

    Science.gov (United States)

    Tao, Feng; Salmeron, Miquel; Somorjai, Gabor A

    2014-11-18

    An embodiment of a scanning tunneling microscope (STM) reactor includes a pressure vessel, an STM assembly, and three spring coupling objects. The pressure vessel includes a sealable port, an interior, and an exterior. An embodiment of an STM system includes a vacuum chamber, an STM reactor, and three springs. The three springs couple the STM reactor to the vacuum chamber and are operable to suspend the scanning tunneling microscope reactor within the interior of the vacuum chamber during operation of the STM reactor. An embodiment of an STM assembly includes a coarse displacement arrangement, a piezoelectric fine displacement scanning tube coupled to the coarse displacement arrangement, and a receiver. The piezoelectric fine displacement scanning tube is coupled to the coarse displacement arrangement. The receiver is coupled to the piezoelectric scanning tube and is operable to receive a tip holder, and the tip holder is operable to receive a tip.

  14. Solid State Reactor Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Mays, G.T.

    2004-03-10

    The Solid State Reactor (SSR) is an advanced reactor concept designed to take advantage of Oak Ridge National Laboratory's (ORNL's) recently developed graphite foam that has enhanced heat transfer characteristics and excellent high-temperature mechanical properties, to provide an inherently safe, self-regulated, source of heat for power and other potential applications. This work was funded by the U.S. Department of Energy's Nuclear Energy Research Initiative (NERI) program (Project No. 99-064) from August 1999 through September 30, 2002. The initial concept of utilizing the graphite foam as a basis for developing an advanced reactor concept envisioned that a suite of reactor configurations and power levels could be developed for several different applications. The initial focus was looking at the reactor as a heat source that was scalable, independent of any heat removal/power conversion process. These applications might include conventional power generation, isotope production and destruction (actinides), and hydrogen production. Having conducted the initial research on the graphite foam and having performed the scoping parametric analyses from neutronics and thermal-hydraulic perspectives, it was necessary to focus on a particular application that would (1) demonstrate the viability of the overall concept and (2) require a reasonably structured design analysis process that would synthesize those important parameters that influence the concept the most as part of a feasible, working reactor system. Thus, the application targeted for this concept was supplying power for remote/harsh environments and a design that was easily deployable, simplistic from an operational standpoint, and utilized the new graphite foam. Specifically, a 500-kW(t) reactor concept was pursued that is naturally load following, inherently safe, optimized via neutronic studies to achieve near-zero reactivity change with burnup, and proliferation resistant. These four major areas

  15. Reactor core performance estimating device

    International Nuclear Information System (INIS)

    Tanabe, Akira; Yamamoto, Toru; Shinpuku, Kimihiro; Chuzen, Takuji; Nishide, Fusayo.

    1995-01-01

    The present invention can autonomously simplify a neural net model thereby enabling to conveniently estimate various amounts which represents reactor core performances by a simple calculation in a short period of time. Namely, a reactor core performance estimation device comprises a nerve circuit net which divides the reactor core into a large number of spacial regions, and receives various physical amounts for each region as input signals for input nerve cells and outputs estimation values of each amount representing the reactor core performances as output signals of output nerve cells. In this case, the nerve circuit net (1) has a structure of extended multi-layered model having direct coupling from an upper stream layer to each of downstream layers, (2) has a forgetting constant q in a corrected equation for a joined load value ω using an inverse error propagation method, (3) learns various amounts representing reactor core performances determined using the physical models as teacher signals, (4) determines the joined load value ω decreased as '0' when it is to less than a predetermined value upon learning described above, and (5) eliminates elements of the nerve circuit net having all of the joined load value decreased to 0. As a result, the neural net model comprises an autonomously simplifying means. (I.S.)

  16. SOLASE: a conceptual laser fusion reactor design

    International Nuclear Information System (INIS)

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

    1977-12-01

    The SOLASE conceptual laser fusion reactor has been designed to elucidate the technological problems posed by inertial confinement fusion reactors. This report contains a detailed description of all aspects of the study including the physics of pellet implosion and burn, optics and target illumination, last mirror design, laser system analysis, cavity design, pellet fabrication and delivery, vacuum system requirements, blanket design, thermal hydraulics, tritium analysis, neutronics calculations, radiation effects, stress analysis, shield design, reactor and plant building layout, maintenance procedures, and power cycle design. The reactor is designed as a 1000 MW/sub e/ unit for central station electric power generation

  17. Supercritical-pressure light water cooled reactors

    CERN Document Server

    Oka, Yoshiaki

    2014-01-01

    This book focuses on the latest reactor concepts, single pass core and experimental findings in thermal hydraulics, materials, corrosion, and water chemistry. It highlights research on supercritical-pressure light water cooled reactors (SCWRs), one of the Generation IV reactors that are studied around the world. This book includes cladding material development and experimental findings on heat transfer, corrosion and water chemistry. The work presented here will help readers to understand the fundamental elements of reactor design and analysis methods, thermal hydraulics, materials and water

  18. Heat dissipating nuclear reactor with metal liner

    Science.gov (United States)

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

    1985-11-21

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

  19. Non-equilibrium radiation nuclear reactor

    Science.gov (United States)

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

    1978-01-01

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

  20. Conceptual design of fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    1984-02-01

    This report describes the engineering conceptual design of Fusion Experimental Reactor (FER) which is to be built as a next generation tokamak machine. This design covers overall reactor systems including MHD equilibrium analysis, mechanical configuration of reactor, divertor, pumped limiter, first wall/breeding blanket/shield, toroidal field magnet, poloidal field magnet, cryostat, electromagnetic analysis, vacuum system, power handling and conversion, NBI, RF heating device, tritium system, neutronics, maintenance, cooling system and layout of facilities. The engineering comparison of a divertor with pumped limiters and safety analysis of reactor systems are also conducted. (author)