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Sample records for breeder reactor plant

  1. Analysis of UF6 breeder reactor power plants

    Science.gov (United States)

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

    1976-01-01

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

  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. Large scale breeder reactor plant prototype mechanical pump conceptual design study

    Energy Technology Data Exchange (ETDEWEB)

    1976-07-01

    This final report is a complete conceptual design study of a mechanical pump for a large scale breeder reactor plant. The pumps are located in the cold leg side of the loops. This makes the net positive suction head available - NPSHA - low, and is, in fact, a major influencing factor in the design. Where possible, experience gained from the Clinch River Project and the FFTF is used in this study. Experience gained in the design, manufacturer, and testing of pumps in general and sodium pumps in particular is reflected in this report. The report includes estimated cost and time schedule for design, manufacture, and testing. It also includes a recommendation for development needs.

  4. An Evaluation of liquid metal leak detection methods for the Clinch River Breeder Reactor Plant

    Energy Technology Data Exchange (ETDEWEB)

    Morris, C.J.; Doctor, S.R.

    1977-12-01

    This report documents an independent review and evaluation of sodium leak detection methods described in the Clinch River Breeder Reactor Preliminary Safety Analysis Report. Only information in publicly available documents was used in making the assessments.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1981-03-11

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

  6. Investigation of molten salt fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kubota, Kenichi; Enuma, Yasuhiro; Tanaka, Yoshihiko; Konomura, Mamoru; Ichimiya, Masakazu [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center

    2000-06-01

    Phase I of Feasibility Studies on Commercialized Fast Reactor System is being performed for two years from Japanese Fiscal Year 1999. In this report, results of the study on fluid fuel reactors (especially a molten salt fast breeder reactor concept) are described from the viewpoint of technical and economical concerns of the plant system design. In JFY1999, we have started to investigate the fluid fuel reactors as alternative concepts of sodium cooled FBR systems with MOX fuel, and selected the unique concept of a molten chloride fast breeder reactor, whose U-Pu fuel cycle can be related to both light water reactors and fast breeder reactors on the basis of present technical data and design experiences. We selected a preliminary composition of molten fuel and conceptual plant design through evaluation of technical and economical issues essential for the molten salt reactors and then compared them with reference design concepts of sodium cooled FBR systems under limited information on the molten chloride fast breeder reactors. The following results were obtained. (1) The molten chloride fast breeder reactors have inherent safety features in the core and plant performances, ad the fluid fuel is quite promising for cost reduction of the fuel fabrication and reprocessing. (2) On the other hand, the inventory of the molten chloride fuel becomes high and thermal conductivity of the coolant is inferior compared to those of sodium cooled FBR systems, then, the size of main components such as IHX's becomes larger and the amount of construction materials is seems to be increased. (3) Furthermore economical vessel and piping materials which contact with the molten chloride salts are required to be developed. From the results, it is concluded that further steps to investigate the molten chloride fast breeder reactor concepts are too early to be conducted. (author)

  7. Estimated recurrence frequencies for initiating accident categories associated with the Clinch River Breeder Reactor Plant design

    Energy Technology Data Exchange (ETDEWEB)

    Copus, E R

    1982-04-01

    Estimated recurrence frequencies for each of twenty-five generic LMFBR initiating accident categories were quantified using the Clinch River Breeder Reactor Plant (CRBRP) design. These estimates were obtained using simplified systems fault trees and functional event tree models from the Accident Delineation Study Phase I Final Report coupled with order-of-magnitude estimates for the initiator-dependent failure probabilities of the individual CRBRP engineered safety systems. Twelve distinct protected accident categories where SCRAM is assumed to be successful are estimated to occur at a combined rate of 10/sup -3/ times per year while thirteen unprotected accident categories in which SCRAM fails are estimated to occur at a combined rate on the order of 10/sup -5/ times per year. These estimates are thought to be representative despite the fact that human performance factors, maintenance and repair, as well as input common cause uncertainties, were not treated explicitly. The overall results indicate that for the CRBRP design no single accident category appears to be dominant, nor can any be totally eliminated from further investigation in the areas of accident phenomenology for in-core events and post-accident phenomenology for containment.

  8. Exploding the myths about the fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Burns, S.

    1979-01-01

    This paper discusses the facts and figures about the effects of conservation policies, the benefits of the Clinch River Breeder Reactor demonstration plant, the feasibility of nuclear weapons manufacture from reactor-grade plutonium, diversion of plutonium from nuclear plants, radioactive waste disposal, and the toxicity of plutonium. The paper concludes that the U.S. is not proceeding with a high confidence strategy for breeder development because of a variety of false assumptions.

  9. Water cooled breeder program summary report (LWBR (Light Water Breeder Reactor) development program)

    Energy Technology Data Exchange (ETDEWEB)

    1987-10-01

    The purpose of the Department of Energy Water Cooled Breeder Program was to demonstrate pratical breeding in a uranium-233/thorium fueled core while producing electrical energy in a commercial water reactor generating station. A demonstration Light Water Breeder Reactor (LWBR) was successfully operated for more than 29,000 effective full power hours in the Shippingport Atomic Power Station. The reactor operated with an availability factor of 76% and had a gross electrical output of 2,128,943,470 kilowatt hours. Following operation, the expended core was examined and no evidence of any fuel element defects was found. Nondestructive assay of 524 fuel rods determined that 1.39 percent more fissile fuel was present at the end of core life than at the beginning, proving that breeding had occurred. This demonstrates the existence of a vast source of electrical energy using plentiful domestic thorium potentially capable of supplying the entire national need for many centuries. To build on the successful design and operation of the Shippingport Breeder Core and to provide the technology to implement this concept, several reactor designs of large breeders and prebreeders were developed for commercial-sized plants of 900--1000 Mw(e) net. This report summarizes the Water Cooled Breeder Program from its inception in 1965 to its completion in 1987. Four hundred thirty-six technical reports are referenced which document the work conducted as part of this program. This work demonstrated that the Light Water Breeder Reactor is a viable alternative as a PWR replacement in the next generation of nuclear reactors. This transition would only require a minimum of change in design and fabrication of the reactor and operation of the plant.

  10. Experimental Breeder Reactor I Preservation Plan

    Energy Technology Data Exchange (ETDEWEB)

    Julie Braun

    2006-10-01

    Experimental Breeder Reactor I (EBR I) is a National Historic Landmark located at the Idaho National Laboratory, a Department of Energy laboratory in southeastern Idaho. The facility is significant for its association and contributions to the development of nuclear reactor testing and development. This Plan includes a structural assessment of the interior and exterior of the EBR I Reactor Building from a preservation, rather than an engineering stand point and recommendations for maintenance to ensure its continued protection.

  11. Fuel Summary Report: Shippingport Light Water Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Illum, D.B.; Olson, G.L.; McCardell, R.K.

    1999-01-01

    The Shippingport Light Water Breeder Reactor (LWBR) was a small water cooled, U-233/Th-232 cycle breeder reactor developed by the Pittsburgh Naval Reactors to improve utilization of the nation's nuclear fuel resources in light water reactors. The LWBR was operated at Shippingport Atomic Power Station (APS), which was a Department of Energy (DOE) (formerly Atomic Energy Commission)-owned reactor plant. Shippingport APS was the first large-scale, central-station nuclear power plant in the United States and the first plant of such size in the world operated solely to produce electric power. The Shippingport LWBR was operated successfully from 1977 to 1982 at the APS. During the five years of operation, the LWBR generated more than 29,000 effective full power hours (EFPH) of energy. After final shutdown, the 39 core modules of the LWBR were shipped to the Expended Core Facility (ECF) at Naval Reactors Facility at the Idaho National Engineering and Environmental Laboratory (INEEL). At ECF, 12 of the 39 modules were dismantled and about 1000 of more than 17,000 rods were removed from the modules of proof-of-breeding and fuel performance testing. Some of the removed rods were kept at ECF, some were sent to Argonne National Laboratory-West (ANL-W) in Idaho and some to ANL-East in Chicago for a variety of physical, chemical and radiological examinations. All rods and rod sections remaining after the experiments were shipped back to ECF, where modules and loose rods were repackaged in liners for dry storage. In a series of shipments, the liners were transported from ECF to Idaho Nuclear Technology Engineering Center (INTEC), formerly the Idaho Chemical Processing Plant (ICPP). The 47 liners containing the fully-rodded and partially-derodded core modules, the loose rods, and the rod scraps, are now stored in underground dry wells at CPP-749.

  12. Conceptual design of Indian molten salt breeder reactor

    Indian Academy of Sciences (India)

    2015-08-28

    Aug 28, 2015 ... Home; Journals; Pramana – Journal of Physics; Volume 85; Issue 3. Conceptual design of Indian molten salt breeder ... India has recently started carrying out fundamental studies so as to arrive at a conceptual design of Indian molten salt breeder reactor (IMSBR). Presently, various design options and ...

  13. Designing a SCADA system simulator for fast breeder reactor

    Science.gov (United States)

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

    2016-04-01

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

  14. Interim Report on Fluid-Fuel Thermal Breeder Reactors

    Energy Technology Data Exchange (ETDEWEB)

    MacPherson, H. G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Alexander, L. G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Carter, W. L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Chapman, R. H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Kinyon, B. W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Miller, J. W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    1960-03-15

    The merits of aqueous-homogeneous {AHBR), graphite-moderated molten salt (MSBR) , and graphite-moderated liquid-bismuth (LBBR) breeder reactors operated at nearly comparable fuel-cycle costs (~1.5 mills/kwhr) were evaluated. The net electrical plant capability was assumed to be 1000 MwE, and the fuel and fertile streams were processed continuously on-site. The specific powers based on fuel were 1.2, 1.2, and 0.5 MwE/kg respectively, and 5.9, 3.7, and 5.3 MwE/tonne based on thorium. Net breeding ratios were 1.10, 1.07, and 1.07, giving doubling times of 5-1/2, 11, and 25 full power years . The fuel-cycle costs at the design points selected were 1.4, 1.3, and 1.6 mills/kwhr . The AHBR has an advantage in breeding ratio and doubling time because D2O is superior to graphite as a moderator in breeder reactors. MSBR has an advantage in fuel-cycle costs and in inventory of uranium in the fertile stream as a result of using a solution blanket.

  15. Dynamics and control of molten-salt breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sing, Vikram; Lish, Matthew R.; Chvala, Ondrej; Upadhyaya, Belle R. [Dept. of Nuclear Engineering, University of Tennessee, Knoxville (United States)

    2017-08-15

    Preliminary results of the dynamic analysis of a two-fluid molten-salt breeder reactor (MSBR) system are presented. Based on an earlier work on the preliminary dynamic model of the concept, the model presented here is nonlinear and has been revised to accurately reflect the design exemplified in ORNL-4528. A brief overview of the model followed by results from simulations performed to validate the model is presented. Simulations illustrate stable behavior of the reactor dynamics and temperature feedback effects to reactivity excursions. Stable and smooth changes at various nodal temperatures are also observed. Control strategies for molten-salt reactor operation are discussed, followed by an illustration of the open-loop load-following capability of the molten-salt breeder reactor system. It is observed that the molten-salt breeder reactor system exhibits “self-regulating” behavior, minimizing the need for external controller action for load-following maneuvers.

  16. ORNL breeder reactor safety quarterly technical progress report, July-September 1980

    Energy Technology Data Exchange (ETDEWEB)

    Fontana, M H; Wantland, J L

    1981-01-01

    Six tasks are reported upon: THORS (Thermal-Hydraulic Out-of-Reactor Safety) program, environmental assessment of alternate FBR fuels, model evaluation of breeder reactor radioactivity releases, nuclear safety information center activities, breeder reactor reliability data analysis center activities, and central data base for breeder reactor safety codes. (DLC)

  17. Development of fuels and structural materials for fast breeder reactors

    Indian Academy of Sciences (India)

    Fast breeder reactors (FBRs) are destined to play a crucial role inthe Indian nuclear power programme in the foreseeable future. FBR technology involves a multi-disciplinary approach to solve the various challenges in the areas of fuel and materials development. Fuels for FBRs have significantly higher concentration of ...

  18. Fission-suppressed hybrid reactor: the fusion breeder

    Energy Technology Data Exchange (ETDEWEB)

    Moir, R.W.; Lee, J.D.; Coops, M.S.

    1982-12-01

    Results of a conceptual design study of a /sup 233/U-producing fusion breeder are presented. The majority of the study was devoted to conceptual design and evaluation of a fission-suppressed blanket and to fuel cycle issues such as fuel reprocessing, fuel handling, and fuel management. Studies in the areas of fusion engineering, reactor safety, and economics were also performed.

  19. Economic Assessment of Russian Nuclear Strategies on the Basis of Fast Breeder Reactors

    Directory of Open Access Journals (Sweden)

    O. V. Marchenko

    2013-01-01

    Full Text Available The paper assesses the economic risk caused by the delay in commissioning innovative nuclear power plants with fast breeder reactors in Russia. The risk is quantitatively measured by the excessive costs for energy development and the possibility of implementing the considered variants that differ in power consumption, technical and economic indices of the reactors, and constraints on CO2 emissions. The probability distribution functions of economic losses for different strategies of nuclear energy development are constructed.

  20. Fuel Summary Report: Shippingport Light Water Breeder Reactor - Rev. 2

    Energy Technology Data Exchange (ETDEWEB)

    Olson, Gail Lynn; Mc Cardell, Richard Keith; Illum, Douglas Brent

    2002-09-01

    The Shippingport Light Water Breeder Reactor (LWBR) was developed by Bettis Atomic Power Laboratory to demonstrate the potential of a water-cooled, thorium oxide fuel cycle breeder reactor. The LWBR core operated from 1977-82 without major incident. The fuel and fuel components suffered minimal damage during operation, and the reactor testing was deemed successful. Extensive destructive and nondestructive postirradiation examinations confirmed that the fuel was in good condition with minimal amounts of cladding deformities and fuel pellet cracks. Fuel was placed in wet storage upon arrival at the Expended Core Facility, then dried and sent to the Idaho Nuclear Technology and Engineering Center for underground dry storage. It is likely that the fuel remains in good condition at its current underground dry storage location at the Idaho Nuclear Technology and Engineering Center. Reports show no indication of damage to the core associated with shipping, loading, or storage.

  1. Gas core reactors for actinide transmutation and breeder applications. Annual report

    Energy Technology Data Exchange (ETDEWEB)

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

    1978-04-01

    This work consists of design power plant studies for four types of reactor systems: uranium plasma core breeder, uranium plasma core actinide transmuter, UF6 breeder and UF6 actinide transmuter. The plasma core systems can be coupled to MHD generators to obtain high efficiency electrical power generation. A 1074 MWt UF6 breeder reactor was designed with a breeding ratio of 1.002 to guard against diversion of fuel. Using molten salt technology and a superheated steam cycle, an efficiency of 39.2% was obtained for the plant and the U233 inventory in the core and heat exchangers was limited to 105 Kg. It was found that the UF6 reactor can produce high fluxes (10 to the 14th power n/sq cm-sec) necessary for efficient burnup of actinide. However, the buildup of fissile isotopes posed severe heat transfer problems. Therefore, the flux in the actinide region must be decreased with time. Consequently, only beginning-of-life conditions were considered for the power plant design. A 577 MWt UF6 actinide transmutation reactor power plant was designed to operate with 39.3% efficiency and 102 Kg of U233 in the core and heat exchanger for beginning-of-life conditions.

  2. Decision analysis of the Liquid Metal Fast Breeder Reactor program

    Energy Technology Data Exchange (ETDEWEB)

    Seim, E.H.

    1983-01-01

    The decision-analysis methodology is employed to develop a model to examine the Liquid Metal Fast Breeder Reactor Program to provide guidance for US decision makers. Information relative to the nuclear fuel cycle, the decision analysis technique, and the supporting economic theory is provided for background purposes. The model consists of four courses of action, three decision times, and five critical factors with either two or three paths leading to 198 possible end results. The courses of action cover a range of the possible programs to develop a commercial LMFBR including scale-up, program timing, and plant schedules. Data developed from a number of recent studies along with probability assignments from three sources are run through the model and indicate that course of action one (Compressed Full Program) produces the greatest net benefits discounted to a present value at a real rate of 5%. An analysis is included to consider the foregone costs of coal usage for electrical generation when LMFBR capacity could be available. Ranking of the courses of action does not change compared to the analysis without foregone costs. The foregone costs are approximately five times greater than the LMFBR benefits alone. Recommendations for specific actions by decision makers conclude the study.

  3. Molten Salt Breeder Reactor Analysis Methods

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jinsu; Jeong, Yongjin; Lee, Deokjung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2015-05-15

    Utilizing the uranium-thorium fuel cycle shows considerable potential for the possibility of MSR. The concept of MSBR should be revised because of molten salt reactor's advantage such as outstanding neutron economy, possibility of continuous online reprocessing and refueling, a high level of inherent safety, and economic benefit by keeping off the fuel fabrication process. For the development of MSR research, this paper provides the MSBR single-cell, two-cell and whole core model for computer code input, and several calculation results including depletion calculation of each models. The calculations are carried out by using MCNP6, a Monte Carlo computer code, which has CINDER90 for depletion calculation using ENDF-VII nuclear data. From the calculation results of various reactor design parameters, the temperature coefficients are all negative at the initial state and MTC becomes positive at the equilibrium state. From the results of core rod worth, the graphite control rod alone cannot makes the core subcritical at initial state. But the equilibrium state, the core can be made subcritical state only by graphite control rods. Through the comparison of the results of each models, the two-cell method can represent the MSBR core model more accurately with a little more computational resources than the single-cell method. Many of the thermal spectrum MSR have adopted a multi-region single-fluid strategy.

  4. Training experience at Experimental Breeder Reactor II

    Energy Technology Data Exchange (ETDEWEB)

    Driscoll, J.W.; McCormick, R.P.; McCreery, H.I.

    1978-01-01

    The EBR-II Training Group develops, maintains,and oversees training programs and activities associated with the EBR-II Project. The group originally spent all its time on EBR-II plant-operations training, but has gradually spread its work into other areas. These other areas of training now include mechanical maintenance, fuel manufacturing facility, instrumentation and control, fissile fuel handling, and emergency activities. This report describes each of the programs and gives a statistical breakdown of the time spent by the Training Group for each program. The major training programs for the EBR-II Project are presented by multimedia methods at a pace controlled by the student. The Training Group has much experience in the use of audio-visual techniques and equipment, including video-tapes, 35 mm slides, Super 8 and 16 mm film, models, and filmstrips. The effectiveness of these techniques is evaluated in this report.

  5. Towards an intrinsically safe and economic thorium breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jagannathan, V. [Light Water Reactor Physics Section, Reactor Physics Design Division, Bhabha Atomic Research Centre, 5th Floor, Central Complex, Mumbai 400085 (India)]. E-mail: vjagan@magnum.barc.ernet.in; Pal, Usha [Light Water Reactor Physics Section, Reactor Physics Design Division, Bhabha Atomic Research Centre, 5th Floor, Central Complex, Mumbai 400085 (India)

    2006-10-15

    Thorium does not have intrinsic fissile content unlike uranium. {sup 232}Th has nearly three times thermal absorption cross section compared to {sup 238}U and hence requires much larger externally fed fissile content compared to uranium based fuel. These factors give a permanent economic competitive edge to uranium. Thus thorium is not inducted in any significant measure in present day power reactors, despite the fact that thorium is three times more abundant in the earth's crust than uranium. Uranium reserves vary from country to country and there is also difficulty in having equitable distribution of uranium. Thus when {sup 235}U would get exhausted, perhaps much sooner in countries having limited uranium reserve, there will be a need to switch over from the today's open fuel cycle programme based on {sup 235}U feed to closed fuel cycle based on Pu feed. At that stage thorium and (depleted) uranium would become equal candidates to form the fertile base. All economic considerations would have to be readdressed. The size and growth of the nuclear power programme based on closed fuel cycle would be dependent on maximizing the fissile conversion rate in those reactors. In this paper we reemphasize the principles and the details of the thermal reactor concept 'A Thorium Breeder Reactor' (ATBR), in which the use of PuO{sub 2} seeded thoria fuel is found to give excellent core characteristics like two years cycle length with nearly zero control maneuvers, fairly high seed output to input ratio and intrinsically safe reactivity coefficients [Jagannathan V, Ganesan S, Karthikeyan R. Sensitivity studies for a thorium breeder reactor design with the nuclear data libraries of WIMS library update project. In: Proceedings of the international conference on emerging nuclear energy systems ICENES-2000, September 25-28, 2000, Petten, The Netherlands].

  6. Multiple recycling of fuel in prototype fast breeder reactor in a closed ...

    Indian Academy of Sciences (India)

    Abstract. A fast breeder reactor (FBR) closed fuel cycle involves recycling of the discharged fuel, after reprocessing and refabrication, in order to utilize the unburnt fuel and the bred fissile material. Our previous study in this regard for the prototype fast breeder reactor (PFBR) indicated the possibility of multiple recycling with ...

  7. Multiple recycling of fuel in prototype fast breeder reactor in a closed ...

    Indian Academy of Sciences (India)

    A fast breeder reactor (FBR) closed fuel cycle involves recycling of the discharged fuel, after reprocessing and refabrication, in order to utilize the unburnt fuel and the bred fissile material. Our previous study in this regard for the prototype fast breeder reactor (PFBR) indicated the possibility of multiple recycling with ...

  8. Immediate relation of ING to fast breeder reactor programs

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, W.B

    1969-07-01

    The future large-scale use of nuclear energy is linked in the United States and other major countries to their fast breeder reactor development. Very serious basic problems have been discovered within the last two years, limiting the life in the high fast neutron flux at appropriate temperatures of materials, in particular of metals suitable for fuel cladding in sodium coolant. There is therefore a most urgent need for materials testing facilities under controlled conditions of temperature and neutron flux at sufficiently high ratings to match or surpass those required in commercially competitive fast breeder reactors. None of the test facilities yet planned for 1976 or sooner in the western world appears to match these conditions. The problem is mainly the difficulty of providing the high neutron flux effectively continuously. The spallation reaction in heavy elements was chosen as the basis of ING - the intense neutron generator, because it is the only known reaction that promises a fast neutron source density that is higher than can be controlled from the fission process. It is suggested that several countries will wish to consider urgently whether they should also explore the spallation reaction for the purpose of a fast neutron irradiation test facility. In view of the discontinuance of the ING project in Canada a favourable opportunity will exist over the next few months 10 obtain from Canada by direct personal contact details of the significant study that has been carried on for ING over the last five years. In the event that satisfactory materials are established within the lifetime of the spallation facilities they may continue to be used for the production of selected isotopes more profitably produced in high neutron fluxes. The facilities may be also used for the desirable preirradiation of thorium reactor fuel. The other research purposes planned for ING could also be served. (author)

  9. Accident analysis of heavy water cooled thorium breeder reactor

    Science.gov (United States)

    Yulianti, Yanti; Su'ud, Zaki; Takaki, Naoyuki

    2015-04-01

    Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of fuel and claddings during accident are still below limitations which are in secure condition.

  10. Comparative assessment of nuclear fuel cycles. Light-water reactor once-through, classical fast breeder reactor, and symbiotic fast breeder reactor cycles

    Energy Technology Data Exchange (ETDEWEB)

    Hardie, R.W.; Barrett, R.J.; Freiwald, J.G.

    1980-06-01

    The object of the Alternative Nuclear Fuel Cycle Study is to perform comparative assessments of nuclear power systems. There are two important features of this study. First, this evaluation attempts to encompass the complete, integrated fuel cycle from mining of uranium ore to disposal of waste rather than isolated components. Second, it compares several aspects of each cycle - energy use, economics, technological status, proliferation, public safety, and commercial potential - instead of concentrating on one or two assessment areas. This report presents assessment results for three fuel cycles. These are the light-water reactor once-through cycle, the fast breeder reactor on the classical plutonium cycle, and the fast breeder reactor on a symbiotic cycle using plutonium and /sup 233/U as fissile fuels. The report also contains a description of the methodology used in this assessment. Subsequent reports will present results for additional fuel cycles.

  11. Review of ORNL-TSF shielding experiments for the gas-cooled Fast Breeder Reactor Program

    Energy Technology Data Exchange (ETDEWEB)

    Abbott, L.S.; Ingersoll, D.T.; Muckenthaler, F.J.; Slater, C.O.

    1982-01-01

    During the period between 1975 and 1980 a series of experiments was performed at the ORNL Tower Shielding Facility in support of the shield design for a 300-MW(e) Gas Cooled Fast Breeder Demonstration Plant. This report reviews the experiments and calculations, which included studies of: (1) neutron streaming in the helium coolant passageways in the GCFR core; (2) the effectiveness of the shield designed to protect the reactor grid plate from radiation damage; (3) the adequacy of the radial shield in protecting the PCRV (prestressed concrete reactor vessel) from radiation damage; (4) neutron streaming between abutting sections of the radial shield; and (5) the effectiveness of the exit shield in reducing the neutron fluxes in the upper plenum region of the reactor.

  12. Steam condenser optimization using Real-parameter Genetic Algorithm for Prototype Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jayalal, M.L., E-mail: jayalal@igcar.gov.in [Indira Gandhi Centre for Atomic Research, Kalpakkam 603102, Tamil Nadu (India); Kumar, L. Satish, E-mail: satish@igcar.gov.in [Indira Gandhi Centre for Atomic Research, Kalpakkam 603102, Tamil Nadu (India); Jehadeesan, R., E-mail: jeha@igcar.gov.in [Indira Gandhi Centre for Atomic Research, Kalpakkam 603102, Tamil Nadu (India); Rajeswari, S., E-mail: raj@igcar.gov.in [Indira Gandhi Centre for Atomic Research, Kalpakkam 603102, Tamil Nadu (India); Satya Murty, S.A.V., E-mail: satya@igcar.gov.in [Indira Gandhi Centre for Atomic Research, Kalpakkam 603102, Tamil Nadu (India); Balasubramaniyan, V.; Chetal, S.C. [Indira Gandhi Centre for Atomic Research, Kalpakkam 603102, Tamil Nadu (India)

    2011-10-15

    Highlights: > We model design optimization of a vital reactor component using Genetic Algorithm. > Real-parameter Genetic Algorithm is used for steam condenser optimization study. > Comparison analysis done with various Genetic Algorithm related mechanisms. > The results obtained are validated with the reference study results. - Abstract: This work explores the use of Real-parameter Genetic Algorithm and analyses its performance in the steam condenser (or Circulating Water System) optimization study of a 500 MW fast breeder nuclear reactor. Choice of optimum design parameters for condenser for a power plant from among a large number of technically viable combination is a complex task. This is primarily due to the conflicting nature of the economic implications of the different system parameters for maximizing the capitalized profit. In order to find the optimum design parameters a Real-parameter Genetic Algorithm model is developed and applied. The results obtained are validated with the reference study results.

  13. Installation of the Light-Water Breeder Reactor at the Shippingport Atomic Power Station (LWBR Development Program)

    Energy Technology Data Exchange (ETDEWEB)

    Massimino, R.J.; Williams, D.A.

    1983-05-01

    This report summarizes the refueling operations performed to install a Light Water Breeder Reactor (LWBR) core into the existing pressurized water reactor vessel at the Shippingport Atomic Power Station. Detailed descriptions of the major installation operations (e.g., primary system preconditioning, fuel installation, pressure boundary seal welding) are included as appendices to this report; these operations are of technical interest to any reactor servicing operation, whether the reactor is a breeder or a conventional light water non-breeder core.

  14. Instrumentation and control system for the prototype fast breeder reactor 'MONJU' power station

    Energy Technology Data Exchange (ETDEWEB)

    Hara, Hiroshi (Power Reactor and Nuclear Fuel Development Corp., Tokyo (Japan)); Mae, Yoshinori; Ishida, Takayuki; Hashiura, Kazuhiko; Kasai, Shozo; Yamamoto, Hajime

    1989-10-01

    The fast breeder reactor 'Monju' power station is constructed as the nuclear power station of next generation in Tsuruga City, Fukui Prefecture. In order to realize high safety and operational reliability as the newest nuclear power station, the measurement and control system of Monju (electric power output 280 MW) has been designed and manufactured by reflecting the experiences of construction and operation of the experimental FBR 'Joyo' and the results of various research and development of sodium instrumentation and others, and by using the latest digital control technology and multiplexing system technology. In this paper, the results of development of the characteristic measurement and control technology as fast breeder reactors and the state of application to the measurement and control system which was designed and manufactured for Monju are described. Central monitoring panel, plant control system, sodium instrumentation, preheating control system and so on are reported. In the case of Monju, the heat capacity and thermal inertia of the primary and secondary cooling systems are large, and the system comprises three loops. (K.I.).

  15. Multiple recycling of fuel in prototype fast breeder reactor

    Indian Academy of Sciences (India)

    In the FBR closed fuel cycle, possibility of multi-recycle has been recognized. In the present study, Pu-239 equivalence approach is used to demonstrate the feasibility of achieving near constant input inventory of Pu and near stable Pu isotopic composition after a few recycles of the same fuel of the prototype fast breeder ...

  16. The International Breeder's Rights System and Crop Plant Innovation.

    Science.gov (United States)

    Barton, J H

    1982-06-04

    Legal arrangements governing a plant breeder's intellectual property rights to his inventions are likely to affect the future of crop research. Such systems, although controversial, are probably currently desirable for the developed world. The new genetic technologies may change this judgment, and certainly require redefinition of the lines between plant patents and regular patents. Several safeguards, present in the United States breeder's rights law, should be applied more broadly. A new safeguard-of ensuring that material be entered into germplasm banks-should be applied everywhere. For the developing world, the desirability of a plant patent system is much less clear; new agreements may be desirable to ensure the free flow and collection of germplasm.

  17. Summary of estimated doses and risks resulting from routine radionuclide releases from fast breeder reactor fuel cycle facilities

    Energy Technology Data Exchange (ETDEWEB)

    Miller, C.W.; Meyer, H.R.

    1985-01-01

    A project is underway at Oak Ridge National Laboratory to assess the human health and environment effects associated with operation of Liquid Metal Fast Breeder Reactor fuel cycle. In this first phase of the work, emphasis was focused on routine radionuclide releases from reactor and reprocessing facilities. For this study, sites for fifty 1-GW(e) capacity reactors and three reprocessing plants were selected to develop scenarios representative of US power requirements. For both the reactor and reprocessing facility siting schemes selected, relatively small impacts were calculated for locality-specific populations residing within 100 km. Also, the results of these analyses are being used in the identification of research priorities. 13 refs., 2 figs., 3 tabs.

  18. Acoustic monitoring of sodium boiling in a liquid metal fast breeder reactor from autoregressive models

    Energy Technology Data Exchange (ETDEWEB)

    Geraldo, Issa Cherif [Laboratoire d’Automatique, Génie Informatique et Signal (LAGIS UMR CNRS 8219), Université Lille 1, Sciences et technologies, Avenue Paul Langevin, BP 48, 59651 Villeneuve d’Ascq CEDEX (France); Bose, Tanmoy [Indian Institute of Technology Kharagpur, Kharagpur 721302, West Bengal (India); Pekpe, Komi Midzodzi, E-mail: midzodzi.pekpe@univ-lille1.fr [Laboratoire d’Automatique, Génie Informatique et Signal (LAGIS UMR CNRS 8219), Université Lille 1, Sciences et technologies, Avenue Paul Langevin, BP 48, 59651 Villeneuve d’Ascq CEDEX (France); Cassar, Jean-Philippe [Laboratoire d’Automatique, Génie Informatique et Signal (LAGIS UMR CNRS 8219), Université Lille 1, Sciences et technologies, Avenue Paul Langevin, BP 48, 59651 Villeneuve d’Ascq CEDEX (France); Mohanty, A.R. [Indian Institute of Technology Kharagpur, Kharagpur 721302, West Bengal (India); Paumel, Kévin [CEA, DEN, Nuclear Technology Department, F-13108 Saint-Paul-lez-Durance (France)

    2014-10-15

    Highlights: • The work deals with sodium boiling detection in a liquid metal fast breeder reactor. • The authors choose to use acoustic data instead of thermal data. • The method is designed to not to be disturbed by the environment noises. • A real time boiling detection methods are proposed in the paper. - Abstract: This paper deals with acoustic monitoring of sodium boiling in a liquid metal fast breeder reactor (LMFBR) based on auto regressive (AR) models which have low computational complexities. Some authors have used AR models for sodium boiling or sodium–water reaction detection. These works are based on the characterization of the difference between fault free condition and current functioning of the system. However, even in absence of faults, it is possible to observe a change in the AR models due to the change of operating mode of the LMFBR. This sets up the delicate problem of how to distinguish a change in operating mode in absence of faults and a change due to presence of faults. In this paper we propose a new approach for boiling detection based on the estimation of AR models on sliding windows. Afterwards, classification of the models into boiling or non-boiling models is made by comparing their coefficients by two statistical methods, multiple linear regression (LR) and support vectors machines (SVM). The proposed approach takes into account operating mode information in order to avoid false alarms. Experimental data include non-boiling background noise data collected from Phenix power plant (France) and provided by the CEA (Commissariat à l’Energie Atomique et aux énergies alternatives, France) and boiling condition data generated in laboratory. High boiling detection rates as well as low false alarms rates obtained on these experimental data show that the proposed method is efficient for boiling detection. Most importantly, it shows that the boiling phenomenon introduces a disturbance into the AR models that can be clearly detected.

  19. Advanced automation concepts applied to Experimental Breeder Reactor-II startup

    Energy Technology Data Exchange (ETDEWEB)

    Berkan, R.C.; Upadhyaya, B.R.; Bywater, R.L. (Tennessee Univ., Knoxville, TN (United States). Dept. of Nuclear Engineering); Kisner, R.A. (Oak Ridge National Lab., TN (United States))

    1991-08-01

    The major objective of this work is to demonstrate through simulations that advanced liquid-metal reactor plants can be operated from low power by computer control. Development of an automatic control system with this objective will help resolve specific issues and provide proof through demonstration that automatic control for plant startup is feasible. This paper presents an advanced control system design for startup of the Experimental Breeder Reactor-2 (EBR-2) located at Idaho Falls, Idaho. The design incorporates recent methods in nonlinear control with advanced diagnostics techniques such as neural networks to form an integrated architecture. The preliminary evaluations are obtained in a simulated environment by a low-order, valid nonlinear model. Within the framework of phase 1 research, the design includes an inverse dynamics controller, a fuzzy controller, and an artificial neural network controller. These three nonlinear control modules are designed to follow the EBR-2 startup trajectories in a multi-input/output regime. They are coordinated by a supervisory routine to yield a fault-tolerant, parallel operation. The control system operates in three modes: manual, semiautomatic, and fully automatic control. The simulation results of the EBR-2 startup transients proved the effectiveness of the advanced concepts. The work presented in this paper is a preliminary feasibility analysis and does not constitute a final design of an automated startup control system for EBR-2. 14 refs., 43 figs.

  20. Status of EC solid breeder blanket designs and R and D for demo fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Proust, E. [CEA Centre d`Etudes de Saclay, 91 - Gif-sur-Yvette (France). Dept. de Mecanique et de Technologie; Dalle Donne, M. [Kernforschungszentrum Karlsruhe GmbH (Germany); Anzidei, L. [ENEA, Frascati (Italy). Centro Ricerche Energia; Kwast, H. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Moons, F. [Centre d`Etude de l`Energie Nucleaire, Mol (Belgium)

    1994-12-31

    Within the European Community Fusion Technology Program two solid breeder blankets for a DEMO reactor are being developed. The two blankets have various features in common: helium as coolant and as tritium purge gas, the martensitic steel MANET as structural material and beryllium as neutron multiplier. The configurations of the two blankets are however different: in the B.I.T. (Breeder Inside Tube) concept the breeder materials are LiAlO{sub 2} or Li{sub 2}ZrO{sub 3} in the form of annular pellets contained in tubes surrounded by beryllium blocks, the coolant helium being outside the tubes, whereas in the B.O.T. (Breeder out of Tube) the breeder and multiplier material are Li{sub 4}SiO{sub 4} and beryllium pebbles forming a mixed bed placed outside the tubes containing the coolant helium. The main critical issues for both blankets are the behavior of the breeder ceramics and of beryllium under irradiation and the tritium control. Other issues are the low temperature irradiation induced embrittlement of MANET, the mechanical effects caused by major plasma disruptions, and safety and reliability. The R and D work concentrate on these issues. The development of martensitic steels including MANET is part of a separate program. Breeder ceramics and beryllium irradiations have been so far performed for conditions which do not cover the peak values injected in the DEMO blankets. Further irradiations in thermal reactors and in fast reactors, especially for beryllium, are required. An effective tritium control requires the development of permeation barriers and/or of methods of oxidation of the tritium in the main helium cooling systems. First promising results have been obtained also in field of mechanical effects from plasma disruptions and safety and reliability, however further work is required in the reliability field and to validate the codes for the calculations of the plasma disruption effects. (authors). 8 figs., 2 tabs., 53 refs.

  1. Atoms in Appalachia. Historical report on the Clinch River Breeder Reactor site

    Energy Technology Data Exchange (ETDEWEB)

    Schaffer, D

    1982-01-01

    The background information concerning the acquisition of the land for siting the Clinch River Breeder Reactor is presented. Historical information is also presented concerning the land acquisition for the Oak Ridge facilities known as the Manhattan Project during World War II.

  2. Engineering review of the core support structure of the Gas Cooled Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    None

    1978-09-01

    The review of the core support structure of the gas cooled fast breeder reactor (GCFR) covered such areas as the design criteria, the design and analysis of the concepts, the development plan, and the projected manufacturing costs. Recommendations are provided to establish a basis for future work on the GCFR core support structure.

  3. Nuclear Engineering Computer Modules, Thermal-Hydraulics, TH-2: Liquid Metal Fast Breeder Reactors.

    Science.gov (United States)

    Reihman, Thomas C.

    This learning module is concerned with the temperature field, the heat transfer rates, and the coolant pressure drop in typical liquid metal fast breeder reactor (LMFBR) fuel assemblies. As in all of the modules of this series, emphasis is placed on developing the theory and demonstrating the use with a simplified model. The heart of the module is…

  4. A FAST BREEDER REACTOR SPENT FUEL MEASUREMENTS PROGRAM FOR BN-350 REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    P. STAPLES; J. HALBIG; ET AL

    1999-04-01

    A project to verify the fissile content of fast breeder reactor spent nuclear fuel is underway in the Republic of Kasakhstan. There are a variety of assembly types with different irradiation histories and profiles in the reactor that require a variety of measurement and analysis procedures. These procedures will be discussed and compared as will the general process that has been designed to resolve any potential measurement discrepancies. The underwater counter is part of a system that is designed to assist the International Atomic Energy Agency (IAEA) in maintaining continuity of knowledge from the time of measurement until the measured item is placed in a welded container with a unique identification. In addition to satisfying IAEA requirements for the spent nuclear fuel, this measurement program is able to satisfy some of the measurement requirements for the Kasakhstan Atomic Energy Agency concerning the repackaging of the spent nuclear fuel into a standard canister. The project is currently operational in a mode requiring the IAEA's continuous presence.

  5. Helium Leak Detection of Vessels in Fuel Transfer Cell (FTC) of Prototype Fast Breeder Reactor (PFBR)

    Science.gov (United States)

    Dutta, N. G.

    2012-11-01

    Bharatiya Nabhikiya Vidyut Nigam (BHAVINI) is engaged in construction of 500MW Prototype Fast Breeder Reactor (PFBR) at Kalpak am, Chennai. In this very important and prestigious national programme Special Product Division (SPD) of M/s Kay Bouvet Engg.pvt. ltd. (M/s KBEPL) Satara is contributing in a major way by supplying many important sub-assemblies like- Under Water trolley (UWT), Airlocks (PAL, EAL) Container and Storage Rack (CSR) Vessels in Fuel Transfer Cell (FTC) etc for PFBR. SPD of KBEPL caters to the requirements of Government departments like - Department of Atomic Energy (DAE), BARC, Defense, and Government undertakings like NPCIL, BHAVINI, BHEL etc. and other precision Heavy Engg. Industries. SPD is equipped with large size Horizontal Boring Machines, Vertical Boring Machines, Planno milling, Vertical Turret Lathe (VTL) & Radial drilling Machine, different types of welding machines etc. PFBR is 500 MWE sodium cooled pool type reactor in which energy is produced by fissions of mixed oxides of Uranium and Plutonium pellets by fast neutrons and it also breeds uranium by conversion of thorium, put along with fuel rod in the reactor. In the long run, the breeder reactor produces more fuel then it consumes. India has taken the lead to go ahead with Fast Breeder Reactor Programme to produce electricity primarily because India has large reserve of Thorium. To use Thorium as further fuel in future, thorium has to be converted in Uranium by PFBR Technology.

  6. Multiple recycling of fuel in prototype fast breeder reactor

    Indian Academy of Sciences (India)

    Abstract. In a thermal neutron reactor, multiple recycle of U–Pu fuel is not possible due to degradation of fissile content of Pu in just one recycle. In the FBR closed fuel cycle, possibility of multi-recycle has been recognized. In the present study, Pu-239 equivalence approach is used to demonstrate the feasibility of achieving ...

  7. The manufacture of enriched uranium fuel slugs for the Experimental Breeder Reactor.

    Energy Technology Data Exchange (ETDEWEB)

    Shuck, Author B.

    1953-04-20

    This report describes the specifications, materials and the sequence of operations used to found and fabricate 4 the first charge of enriched uranium fuel in the Experimental Breeder Reactor. The work was governed by the following principles: a. That the fuel be of correct composition, dimension and metallurgical condition for use in the reactor. b. That a maximum yield of finished fuel slugs from the quantity of uranium available for the program be achieved. c. That the residues be in a form which can be recovered by chemical or other means. d. That a detailed record be kept in such form that a complete history of each fuel slug be available.

  8. Cellular convection in vertical annuli of fast breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hemanath, M.G. [Fast Reactor Technology Group, Indira Gandhi Center for Atomic Research, Kalpakkam (India)], E-mail: hemanath@igcar.gov.in; Meikandamurthy, C.; Ramakrishnan, V.; Rajan, K.K.; Rajan, M.; Vaidyanathan, G. [Fast Reactor Technology Group, Indira Gandhi Center for Atomic Research, Kalpakkam (India)

    2007-08-15

    In the pool type fast reactors the roof structure is penetrated by a number of pumps and heat exchangers that are cylindrical in shape. Sandwiched between the free surface of sodium and the roof structure, is stagnant argon gas, which can flow in the annular space between the components and roof structure, as a thermosyphon. These thermosyphons not only transport heat from sodium to roof structure, but also result in cellular convection in vertical annuli resulting in circumferential temperature asymmetry of the penetrating components. There is need to know the temperature asymmetry as it can cause tilting of the components. Experiments were carried out in an annulus model to predict the circumferential temperature difference with and without sodium in the test vessel. Three-dimensional analysis was also carried out using PHOENICS CFD code and compared with the experiment. This paper describes the experimental details, the theoretical analysis and their comparison.

  9. Investigation of zero-release cycle using fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-11-01

    The task force was organized for the main purpose of offering quantitative basic data to the study group on nuclear fuel cycle in February, 1997. The effect of so-called frontier technologies such as the isotope separation by laser method, the FP annihilation with electron beam accelerators and so on in the FBR cycle based on MOX fuel and PUREX reprocessing method was expected. It is aimed at to recycle the total amount of minor actinides. The object of recycling is the nuclides which contribute largely to toxicity, namely 11 elements, 12 nuclides. The preconditions and the target to be attained of the investigation are explained. As the results of investigation, the amount of reloading MA and FP into a reactor, squeezing the recycling scenario, the effect of reducing toxicity and the subject of the countermeasures to the nuclides with long half-life which cannot be reloaded are reported. As the technical evaluation required for realizing the concept, the concept of the core which excludes recriticality, the advance of reprocessing technology, isotope separation, the fabrication into the optimal form for recycling and so on are discussed. The economical efficiency of the recycling based on MOX and PUREX and the proposal of the development scenario are described. (K.I.)

  10. The long-term future for civilian nuclear power generation in France: The case for breeder reactors. Breeder reactors: The physical and physical chemistry parameters, associate material thermodynamics and mechanical engineering: Novelties and issues

    Science.gov (United States)

    Dautray, Robert

    2011-06-01

    , etc., radioprotection measures and treatment for the "transuranic" elements. For a long period of time, France was in the forefront of nuclear breeder power generation science, technological research and also in the knowledge base related to breeder reactors. It is in the country's interest to pursue these efforts and this could per se constitute one of the national priorities. Nous sommes naturellement bien conscients de l'énorme problème qui se pose au Japon actuellement comme suite au tremblement de terre et au tsunami de mars 2011 et leurs conséquences, notamment sur des installations électronucléaires. Le texte que nous présentons concerne des conditions totalement générales, indépendantes des problèmes spécifiques de sûreté qu'il faudra, de toute façon, traiter dans le cadre d'un développement éventuel de l'énergie nucléaire.We are aware, of course, of the huge problem that Japan has to deal with the aftermath of the quake and tsunami of March 2011 and their consequences on electronuclear power plants. The text that we present here concerns general physical topics independent of the specific safety problems, general physical topics which will have to be solved in the case of a contingent development of electronuclear power plants.

  11. Preliminary design of a Binary Breeder Reactor; Diseno preliminar de un reactor esferico de quema/cria

    Energy Technology Data Exchange (ETDEWEB)

    Garcia C, E. Y.; Francois, J. L.; Lopez S, R. C., E-mail: eliasgarcerv@hotmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac No. 8532, 62550 Jiutepec, Morelos (Mexico)

    2014-10-15

    A binary breeder reactor (BBR) is a reactor that by means of the transmutation and fission process can operates through the depleted uranium burning with a small quantity of fissile material. The advantages of a BBR with relation to other nuclear reactor types are numerous, taking into account their capacity to operate for a long time without requiring fuel reload or re-arrangement. In this work four different simulations are shown carried out with the MCNPX code with libraries Jeff-3.1 to 1200 K. The objective of this study is to compare two different models of BBR: a spherical reactor and a cylindrical one, using two fuel cycles for each one of them (U-Pu and Th-U) and different reflectors for the two different geometries. For all the models a super-criticality state was obtained at least 10.9 years without carrying out some fuel re-arrangement or reload. The plutonium-239 production was achieved in the models where natural uranium was used in the breeding area, while the production of uranium-233 was observed in the cases where thorium was used in the fertile area. Finally, a behavior of stationary wave reactor was observed inside the models of spherical reactor when contemplating the power uniform increment in the breeding area, while inside the cylindrical models was observed the behavior of a traveling wave reactor when registering the displacement of the burnt wave along the cylindrical model. (Author)

  12. Review of uncertainty estimates associated with models for assessing the impact of breeder reactor radioactivity releases

    Energy Technology Data Exchange (ETDEWEB)

    Miller, C.; Little, C.A.

    1982-08-01

    The purpose is to summarize estimates based on currently available data of the uncertainty associated with radiological assessment models. The models being examined herein are those recommended previously for use in breeder reactor assessments. Uncertainty estimates are presented for models of atmospheric and hydrologic transport, terrestrial and aquatic food-chain bioaccumulation, and internal and external dosimetry. Both long-term and short-term release conditions are discussed. The uncertainty estimates presented in this report indicate that, for many sites, generic models and representative parameter values may be used to calculate doses from annual average radionuclide releases when these calculated doses are on the order of one-tenth or less of a relevant dose limit. For short-term, accidental releases, especially those from breeder reactors located in sites dominated by complex terrain and/or coastal meteorology, the uncertainty in the dose calculations may be much larger than an order of magnitude. As a result, it may be necessary to incorporate site-specific information into the dose calculation under these circumstances to reduce this uncertainty. However, even using site-specific information, natural variability and the uncertainties in the dose conversion factor will likely result in an overall uncertainty of greater than an order of magnitude for predictions of dose or concentration in environmental media following shortterm releases.

  13. Optimization of material and production to develop fluoroelastomer inflatable seals for sodium cooled fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sinha, N.K., E-mail: nksinha@igcar.gov.i [Indira Gandhi Centre for Atomic Research (IGCAR), Department of Atomic Energy (DAE), Kalpakkam, Tamilnadu 603102 (India); Raj, Baldev, E-mail: dir@igcar.gov.i [Indira Gandhi Centre for Atomic Research (IGCAR), Department of Atomic Energy (DAE), Kalpakkam, Tamilnadu 603102 (India)

    2011-03-15

    Research highlights: Production of thin fluoroelastomer profiles by cold feed extrusion and continuous cure involving microwave and hot air heating. Use of peroxide curing in air during production. Use of fluoroelastomers based on advanced polymer architecture (APA) for the production of profiles. Use of the profiles in inflatable seals for critical application of Prototype Fast Breeder Reactor. Tailoring of material formulation by synchronized optimization of material and production technologies to ensure that the produced seal ensures significant gains in terms of performance and safety in reactor under synergistic influences of temperature, radiation, air and sodium aerosol. - Abstract: The feasibility of producing thin-walled fluoroelastomer profiles under continuous, atmospheric-pressure vulcanization conditions in air has been demonstrated by successful manufacture of {approx}2 m diameter test inflatable seals for the 500 MWe, Prototype Fast Breeder Reactor (PFBR) using a 50/50 blend formulation of Viton GBL-200S/600S based on advanced polymer architecture (APA). A commercial cold feed screw extruder with 90 mm diameter screw was used along with continuous cure by microwave (2.45 GHz) and hot air heating (190 {sup o}C) at a line speed of 1 m/min to produce the seals. The blend formulation promises significant improvement in the performance and safety of the seals. This article depicts the relevant characteristics of the original inflatable seal compound that was used as reference to achieve the objectives through synchronized optimization of material and production technologies. The production trials are outlined and the blend formulation used with minor factory modifications to produce the test seals is reported. Progressive refinements of the original, Viton A-401C based compound to the blend formulation is presented along with an assessment of potential performance gains. Possible uses of the reported formulation and production technique for other large

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

    Science.gov (United States)

    Harto, Andang Widi

    2012-06-01

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

  15. Compendium of computer codes for the safety analysis of fast breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    1977-10-01

    The objective of the compendium is to provide the reader with a guide which briefly describes many of the computer codes used for liquid metal fast breeder reactor safety analyses, since it is for this system that most of the codes have been developed. The compendium is designed to address the following frequently asked questions from individuals in licensing and research and development activities: (1) What does the code do. (2) To what safety problems has it been applied. (3) What are the code's limitations. (4) What is being done to remove these limitations. (5) How does the code compare with experimental observations and other code predictions. (6) What reference documents are available.

  16. TOKOPS: Tokamak Reactor Operations Study: The influence of reactor operations on the design and performance of tokamaks with solid-breeder blankets: Final report

    Energy Technology Data Exchange (ETDEWEB)

    Conn, R.W.; Ghoniem, N.M.; Firestone, M.A. (eds.)

    1986-09-01

    Reactor system operation and procedures have a profound impact on the conception and design of power plants. These issues are studied here using a model tokamak system employing a solid-breeder blanket. The model blanket is one which has evolved from the STARFIRE and BCSS studies. The reactor parameters are similar to those characterizing near-term fusion engineering reactors such as INTOR or NET (Next European Tokamak). Plasma startup, burn analysis, and methods for operation at various levels of output power are studied. A critical, and complicating, element is found to be the self-consistent electromagnetic response of the system, including the presence of the blanket and the resulting forces and loadings. Fractional power operation, and the strategy for burn control, is found to vary depending on the scaling law for energy confinement, and an extensive study is reported. Full-power reactor operation is at a neutron wall loading pf 5 MW/m/sup 2/ and a surface heat flux of 1 MW/m/sup 2/. The blanket is a pressurized steel module with bare beryllium rods and low-activation HT-9-(9-C-) clad LiAlO/sub 2/ rods. The helium coolant pressure is 5 MPa, entering the module at 297/sup 0/C and exiting at 550/sup 0/C. The system power output is rated at 1000 MW(e). In this report, we present our findings on various operational scenarios and their impact on system design. We first start with the salient aspects of operational physics. Time-dependent analyses of the blanket and balance of plant are then presented. Separate abstracts are included for each chapter.

  17. Be-Breeder - Learning: a new tool for teaching and learning plant breeding principles

    OpenAIRE

    Roberto Fritsche-Neto; Filipe Inácio Matias

    2016-01-01

    Abstract The Be-Breeder application is an on-line tool constructed through the R software for the purpose of assisting in some of the main genetic and statistical analyses related to the area of plant breeding. In addition, Be-Breeder provides a section called "Learning", which in a simple click-point manner allows explanation of theories related to the effect of inbreeding, population structure, qualitative and quantitative traits, heterosis, population size, effect of selection, and composi...

  18. A ceramic breeder in a poloidal tube blanket for a tokamak reactor

    Energy Technology Data Exchange (ETDEWEB)

    Amici, A.; Anzidei, L.; Gallina, M.; Rado, V.; Simbolotti, G.; Violante, V.; Zampaglione, V.; Petrizzi, L. (Associazione Euratom-CNEN sulla Fusione, Centro di Frascati (Italy))

    1989-04-01

    A conceptual study of a helium-cooled solid breeder blanket for a tokamak reactor is presented. Tritium breeding capability together with system reliability are taken as the main design criteria. The blanket consists of tubular poloidal modules made of a central bundle of ceramic rods ({gamma}LiAlO/sub 2/) with a coaxial distribution of the inlet/outlet coolant flow (He) surrounded by a multiplier material (Be) in the form of bored bricks. The Be to {gamma}LiAlO/sub 2/ volume ratio is 4/1. The He inlet and outlet branches are cooling Be and {gamma}LiAlO/sub 2/, respectively. A purge He flow running through small central holes of the ceramic rods is derived from the main flow. Under the typical conditions of a tokamak reactor (neutron wall load=2 MW/m/sup 2/), a full coverage tritium breeding ratio of 1.47 is achieved for the following design and operating parameters: outlet He temperature=570/sup 0/C; inlet He temperature=250/sup 0/; total extracted power=2700 MW; He pumping power percentage=2%; minimum/maximum {gamma}LiAlO/sub 2/ temperature=400/900/sup 0/C; maximum structural temperature=475/sup 0/C; and maximum Be temperature=525/sup 0/C. (orig.).

  19. Evaluation of the Initial Isothermal Physics Measurements at the Fast Flux Test Facility, a Prototypic Liquid Metal Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess

    2010-03-01

    The Fast Flux Test Facility (FFTF) was a 400-MWt, sodium-cooled, low-pressure, high-temperature, fast-neutron flux, nuclear fission reactor plant designed for the irradiation testing of nuclear reactor fuels and materials for the development of liquid metal fast breeder reactors (LMFBRs). The FFTF was fueled with plutonium-uranium mixed oxide (MOX) and reflected by Inconel-600. Westinghouse Hanford Company operated the FFTF as part of the Hanford Engineering Development Laboratory (HEDL) for the U.S. Department of Energy on the Hanford Site near Richland, Washington. Although the FFTF was a testing facility not specifically designed to breed fuel or produce electricity, it did provide valuable information for LMFBR projects and base technology programs in the areas of plant system and component design, component fabrication, prototype testing, and site construction. The major objectives of the FFTF were to provide a strong, disciplined engineering base for the LMFBR program, provide fast flux testing for other U.S. programs, and contribute to the development of a viable self-sustaining competitive U.S. LMFBR industry. During its ten years of operation, the FFTF acted as a national research facility to test advanced nuclear fuels, materials, components, systems, nuclear power plant operating and maintenance procedures, and active and passive reactor safety technologies; it also produced a large number of isotopes for medical and industrial users, generated tritium for the U.S. fusion research program, and participated in cooperative, international research work. Prior to the implementation of the reactor characterization program, a series of isothermal physics measurements were performed; this acceptance testing program consisted of a series of control rod worths, critical rod positions, subcriticality measurements, maximum reactivity addition rates, shutdown margins, excess reactivity, and isothermal temperature coefficient reactivity. The results of these

  20. Be-Breeder - Learning: a new tool for teaching and learning plant breeding principles

    Directory of Open Access Journals (Sweden)

    Roberto Fritsche-Neto

    2016-09-01

    Full Text Available The Be-Breeder application is an on-line tool constructed through the R software for the purpose of assisting in some of the main genetic and statistical analyses related to the area of plant breeding. In addition, Be-Breeder provides a section called “Learning”, which in a simple click-point manner allows explanation of theories related to the effect of inbreeding, population structure, qualitative and quantitative traits, heterosis, population size, effect of selection, and composition of hybrids. Be-Breeder is available for network use on the website of the Allogamous Plant Breeding Laboratory (Laboratório de Melhoramento de Plantas Alógamas of ESALQ-USP through the link: http://www.genetica.esalq.usp.br/alogamas/R.html.

  1. Performance characterization of geopolymer composites for hot sodium exposed sacrificial layer in fast breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Haneefa, K. Mohammed, E-mail: mhkolakkadan@gmail.com [Department of Civil Engineering, IIT Madras, Chennai (India); Santhanam, Manu [Department of Civil Engineering, IIT Madras, Chennai (India); Parida, F.C. [Radiological Safety Division, Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

    2013-12-15

    Highlights: • Performance evaluation of geopolymers subjected to hot liquid sodium is performed. • Apart from mechanical properties, micro-analytical techniques are used for material characterization. • The geopolymer composite showed comparatively lesser damage than conventional cement composites. • Geopolymer technology can emerge as a new choice for sacrificial layer in SCFBRs. - Abstract: A sacrificial layer of concrete is used in sodium cooled fast breeder reactors (SCFBRs) to mitigate thermo-chemical effect of accidentally spilled sodium at and above 550 °C on structural concrete. Performance of this layer is governed by thermo-chemical stability of the ingredients of sacrificial layer concrete. Concrete with limestone aggregate is generally used as a sacrificial layer. Conventional cement based systems exhibit instability in hot liquid sodium environment. Geo-polymer composites are well known to perform excellently at elevated temperatures compared to conventional cement systems. This paper discusses performance of such composites subjected to exposure of hot liquid sodium in air. The investigation includes comprehensive evaluation of various geo-polymer composites before any exposure, after heating to 550 °C in air, and after immersing in hot liquid sodium initially heated to 550 °C in air. Results from the current study indicate that hot liquid sodium produces less damage to geopolymer composites than to the existing conventional cement based system. Hence, the geopolymer technology has potential application in mitigating the degrading effects of sodium fires and can emerge as a new choice for sodium exposed sacrificial layer in SCFBRs.

  2. Assessment of gel-sphere-pac fuel for fast breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lackey, W J; Selle, J E [comps.

    1978-10-01

    An assessment of the state of the art for the gel-sphere-pac process was undertaken to provide a sound basis for further development of the technology. Information is provided on sol preparation, sphere forming, drying, sintering, characterization, loading, fuel rod inspection, and irradiation performance. In addition, discussions are included on: evaluation of the potential for scale-up to production capacities, potential problems associated with remote operation, and future work required to further develop the technology. Three techniques are available for microsphere production: (1) internal gelation, (2) external gelation, and (3) gelation by water extraction. Each has its own advantages and disadvantages; for example, internal gelation appears better suited to the preparation of large spheres than the other processes. Numerous advantages and disadvantages are discussed in detail. Scale-up or remote operation of these techniques appears achievable, although some would require less development than others. Techniques have been developed for drying and sintering spheres. Extensive technology has been developed for sphere characterization, handling, and the loading and inspection of fuel pins. Data available to date indicates that sphere-pac oxide fuel will perform similarly to pellet oxide fuels under fast breeder reactor operating conditions. Gel-sphere-pac technology also appears attractive for carbide fuels.

  3. Final report for the Light Water Breeder Reactor proof-of-breeding analytical support project

    Energy Technology Data Exchange (ETDEWEB)

    Graczyk, D.G.; Hoh, J.C.; Martino, F.J.; Nelson, R.E.; Osudar, J.; Levitz, N.M.

    1987-05-01

    The technology of breeding /sup 233/U from /sup 232/Th in a light water reactor is being developed and evaluated by the Westinghouse Bettis Atomic Power Laboratory (BAPL) through operation and examination of the Shippingport Light Water Breeder Reactor (LWBR). Bettis is determining the end-of-life (EOL) inventory of fissile uranium in the LWBR core by nondestructive assay of a statistical sample comprising approximately 500 EOL fuel rods. This determination is being made with an irradiated-fuel assay gauge based on neutron interrogation and detection of delayed neutrons from each rod. The EOL fissile inventory will be compared with the beginning-of-life fissile loading of the LWBR to determine the extent of breeding. In support of the BAPL proof-of-breeding (POB) effort, Argonne National Laboratory (ANL) carried out destructive physical, chemical, and radiometric analyses on 17 EOL LWBR fuel rods that were previously assayed with the nondestructive gauge. The ANL work included measurements on the intact rods; shearing of the rods into pre-designated contiguous segments; separate dissolution of each of the more than 150 segments; and analysis of the dissolver solutions to determine each segment's uranium content, uranium isotopic composition, and loading of selected fission products. This report describes the facilities in which this work was carried out, details operations involved in processing each rod, and presents a comprehensive discussion of uncertainties associated with each result of the ANL measurements. Most operations were carried out remotely in shielded cells. Automated equipment and procedures, controlled by a computer system, provided error-free data acquisition and processing, as well as full replication of operations with each rod. Despite difficulties that arose during processing of a few rod segments, the ANL destructive-assay results satisfied the demanding needs of the parent LWBR-POB program.

  4. Software development methodology for computer based I&C systems of prototype fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Manimaran, M., E-mail: maran@igcar.gov.in; Shanmugam, A.; Parimalam, P.; Murali, N.; Satya Murty, S.A.V.

    2015-10-15

    Highlights: • Software development methodology adopted for computer based I&C systems of PFBR is detailed. • Constraints imposed as part of software requirements and coding phase are elaborated. • Compliance to safety and security requirements are described. • Usage of CASE (Computer Aided Software Engineering) tools during software design, analysis and testing phase are explained. - Abstract: Prototype Fast Breeder Reactor (PFBR) is sodium cooled reactor which is in the advanced stage of construction in Kalpakkam, India. Versa Module Europa bus based Real Time Computer (RTC) systems are deployed for Instrumentation & Control of PFBR. RTC systems have to perform safety functions within the stipulated time which calls for highly dependable software. Hence, well defined software development methodology is adopted for RTC systems starting from the requirement capture phase till the final validation of the software product. V-model is used for software development. IEC 60880 standard and AERB SG D-25 guideline are followed at each phase of software development. Requirements documents and design documents are prepared as per IEEE standards. Defensive programming strategies are followed for software development using C language. Verification and validation (V&V) of documents and software are carried out at each phase by independent V&V committee. Computer aided software engineering tools are used for software modelling, checking for MISRA C compliance and to carry out static and dynamic analysis. Various software metrics such as cyclomatic complexity, nesting depth and comment to code are checked. Test cases are generated using equivalence class partitioning, boundary value analysis and cause and effect graphing techniques. System integration testing is carried out wherein functional and performance requirements of the system are monitored.

  5. Choice of rotatable plug seals for prototype fast breeder reactor: Review of historical perspectives

    Energy Technology Data Exchange (ETDEWEB)

    Sinha, N.K., E-mail: nksinha@igcar.gov.in; Raj, Baldev, E-mail: baldev.dr@gmail.com

    2015-09-15

    Highlights: • Choice and arrangement of elastomeric inflatable and backup seals as primary and secondary barriers. • With survey (mid-1930s onwards) of reactor, sealing, R&D and rubber technology. • Load, reliability, safety, life and economy of seals and reactors are key factors. • PFBR blends concepts and experience of MOX fuelled FBRs with original solutions. • R&D indicates inflatable seal advanced fluoroelastomer pivotal in unifying nuclear sealing. - Abstract: Choice and arrangement of elastomeric primary inflatable and secondary backup seals for the rotatable plugs (RPs) of 500 MW (e), sodium cooled, pool type, 2-loop, mixed oxide (MOX) fuelled Prototype Fast Breeder Reactor (PFBR) is depicted with review of various historical perspectives. Static and dynamic operation, largest diameters (PFBR: ∼6.4 m, ∼4.2 m), widest gaps and variations (5 ± 2 mm) and demanding operating requirements make RP openings on top shield (TS) the most difficult to seal which necessitated extensive development from 1950s to early 1990s. Liquid metal freeze seals with life equivalent to reactor prevailed as primary barrier (France, Japan, U.S.S.R.) during pre-1980s in spite of bulk, cost and complexity due to the abilities to meet zero leakage and resist core disruptive accident (CDA). Redefinition of CDA as beyond design basis accident, tolerable leakage and enhanced economisation drive during post-1980s established elastomeric inflatable seal as primary barrier excepting in U.S.S.R. (MOX fuel, freeze seal) and U.S.A. (metallic fuel). Choice of inflatable seal for PFBR RPs considers these perspectives, inherent advantages of elastomers and those of inflatable seals which maximise seal life. Choice of elastomeric backup seal as secondary barrier was governed by reliability and minimisation as well as distribution of load (temperature, radiation, mist) to maximise seal life. The compact sealing combination brings the hanging RPs at about the same elevation to reduce

  6. Cost/performance comparison between pulse columns and centrifugal contactors designed to process Clinch River Breeder Reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ciucci, J.A. Jr.

    1983-12-01

    A comparison between pulse columns and centrifugal contactors was made to determine which type of equipment was more advantageous for use in the primary decontamination cycle of a remotely operated fuel reprocessing plant. Clinch River Breeder Reactor (CRBR) fuel was chosen as the fuel to be processed in the proposed 1 metric tonne/day reprocessing facility. The pulse columns and centrifugal contactors were compared on a performance and total cost basis. From this comparison, either the pulse columns or the centrifugal contactors will be recommended for use in a fuel reprocessing plant built to reprocess CRBR fuel. The reliability, solvent exposure to radiation, required time to reach steady state, and the total costs were the primary areas of concern for the comparison. The pulse column units were determined to be more reliable than the centrifugal contactors. When a centrifugal contactor motor fails, it can be remotely changed in less than one eight hour shift. Pulse columns expose the solvent to approximately five times as much radiation dose as the centrifugal contactor units; however, the proposed solvent recovery system adequately cleans the solvent for either case. The time required for pulse columns to reach steady state is many times longer than the time required for centrifugal contactors to reach steady state. The cost comparison between the two types of contacting equipment resulted in centrifugal contactors costing 85% of the total cost of pulse columns when the contactors were stacked on three levels in the module. If the centrifugal contactors were all positioned on the top level of a module with the unoccupied volume in the module occupied by other equipment, the centrifugal contactors cost is 66% of the total cost of pulse columns. Based on these results, centrifugal contactors are recommended for use in a remotely operated reprocessing plant built to reprocess CRBR fuel.

  7. Three core concepts for producing uranium-233 in commercial pressurized light water reactors for possible use in water-cooled breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Conley, G.H.; Cowell, G.K.; Detrick, C.A.; Kusenko, J.; Johnson, E.G.; Dunyak, J.; Flanery, B.K.; Shinko, M.S.; Giffen, R.H.; Rampolla, D.S.

    1979-12-01

    Selected prebreeder core concepts are described which could be backfit into a reference light water reactor similar to current commercial reactors, and produce uranium-233 for use in water-cooled breeder reactors. The prebreeder concepts were selected on the basis of minimizing fuel system development and reactor changes required to permit a backfit. The fuel assemblies for the prebreeder core concepts discussed would occupy the same space envelope as those in the reference core but contain a 19 by 19 array of fuel rods instead of the reference 17 by 17 array. An instrument well and 28 guide tubes for control rods have been allocated to each prebreeder fuel assembly in a pattern similar to that for the reference fuel assemblies. Backfit of these prebreeder concepts into the reference reactor would require changes only to the upper core support structure while providing flexibility for alternatives in the type of fuel used.

  8. Uncertainty evaluation of reliability of shutdown system of a medium size fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zeliang, Chireuding; Singh, Om Pal, E-mail: singhop@iitk.ac.in; Munshi, Prabhat

    2016-11-15

    Highlights: • Uncertainty analysis of reliability of Shutdown System is carried out. • Monte Carlo method of sampling is used. • The effect of various reliability improvement measures of SDS are accounted. - Abstract: In this paper, results are presented on the uncertainty evaluation of the reliability of Shutdown System (SDS) of a Medium Size Fast Breeder Reactor (MSFBR). The reliability analysis results are of Kumar et al. (2005). The failure rate of the components of SDS are taken from International literature and it is assumed that these follow log-normal distribution. Fault tree method is employed to propagate the uncertainty in failure rate from components level to shutdown system level. The beta factor model is used to account different extent of diversity. The Monte Carlo sampling technique is used for the analysis. The results of uncertainty analysis are presented in terms of the probability density function, cumulative distribution function, mean, variance, percentile values, confidence intervals, etc. It is observed that the spread in the probability distribution of SDS failure rate is less than SDS components failure rate and ninety percent values of the failure rate of SDS falls below the target value. As generic values of failure rates are used, sensitivity analysis is performed with respect to failure rate of control and safety rods and beta factor. It is discovered that a large increase in failure rate of SDS rods is not carried to SDS system failure proportionately. The failure rate of SDS is very sensitive to the beta factor of common cause failure between the two systems of SDS. The results of the study provide insight in the propagation of uncertainty in the failure rate of SDS components to failure rate of shutdown system.

  9. Gas-cooled fast breeder reactor. Quarterly progress report, February 1-April 30, 1980

    Energy Technology Data Exchange (ETDEWEB)

    1980-05-01

    Information is presented concerning the reactor vessel; reactivity control mechanisms and instrumentation; reactor internals; primary coolant circuits;core auxiliary cooling system; reactor core; systems engineering; and reactor safety and reliability;

  10. CerealsDB 2.0: an integrated resource for plant breeders and scientists

    Directory of Open Access Journals (Sweden)

    Wilkinson Paul A

    2012-09-01

    Full Text Available Abstract Background Food security is an issue that has come under renewed scrutiny amidst concerns that substantial yield increases in cereal crops are required to feed the world’s booming population. Wheat is of fundamental importance in this regard being one of the three most important crops for both human consumption and livestock feed; however, increase in crop yields have not kept pace with the demands of a growing world population. In order to address this issue, plant breeders require new molecular tools to help them identify genes for important agronomic traits that can be introduced into elite varieties. Studies of the genome using next-generation sequencing enable the identification of molecular markers such as single nucleotide polymorphisms that may be used by breeders to identify and follow genes when breeding new varieties. The development and application of next-generation sequencing technologies has made the characterisation of SNP markers in wheat relatively cheap and straightforward. There is a growing need for the widespread dissemination of this information to plant breeders. Description CerealsDB is an online resource containing a range of genomic datasets for wheat (Triticum aestivum that will assist plant breeders and scientists to select the most appropriate markers for marker assisted selection. CerealsDB includes a database which currently contains in excess of 100,000 putative varietal SNPs, of which several thousand have been experimentally validated. In addition, CerealsDB contains databases for DArT markers and EST sequences, and links to a draft genome sequence for the wheat variety Chinese Spring. Conclusion CerealsDB is an open access website that is rapidly becoming an invaluable resource within the wheat research and plant breeding communities.

  11. The evolving landscape of plant breeders rights regarding wheat varieties in South Africa

    Directory of Open Access Journals (Sweden)

    Charity R. Nhemachena

    2016-03-01

    Full Text Available Addressing the multiple challenges facing global agriculture requires integrated innovation in areas such as seeds, biotechnology, crop protection, grain storage and transport. Innovations related to plant improvement and the development of new or improved plant varieties will only happen at an optimal level if plant breeders rights (PBR are properly protected. The objective was to analyse the evolving landscape of wheat plant breeders rights to address the dearth of empirical evidence of the patterns and trends of wheat varietal improvements in South Africa. We compiled a detailed and novel count and attribute database of wheat varietal innovations in South Africa from 1979 to 2013 using various sources. This data set was then analysed to ascertain the main trends in, and ownership of PBRs for wheat varietal improvements in South Africa over this period. A total of 134 PBR wheat varietal innovations were lodged from 1979 to 2013, an average of 6 applications per year. The administrative delays in granting PBR applications were substantially reduced by 77 days during the post-deregulation period (after 1996, indicating increased efficiency. The main PBR applicants were Sensako (39%, the Agricultural Research Council Small Grains Institute (ARC-SGI (25% and Pannar (15%. The ARC-SGI contributed to some of the PBRs owned by private companies through shared genetic resources before Plant Variety Protection (PVP was implemented. Future innovations and dissemination of wheat innovations can be stimulated by plant variety protection, together with broader variety sector legislation that encourages both public and private sector investment.

  12. Computerized operating procedures for shearing and dissolution of segments from LWBR (Light Water Breeder Reactor) fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Osudar, J.; Deeken, P.G.; Graczyk, D.G.; Fagan, J.E.; Martino, F.J.; Parks, J.E.; Levitz, N.M.; Kessie, R.W.; Leddin, J.M.

    1987-05-01

    This report presents two detailed computerized operating procedures developed to assist and control the shearing and dissolution of irradiated fuel rods. The procedures were employed in the destructive analysis of end-of-life fuel rods from the Light Water Breeder Reactor (LWBR) that was designed by the Westinghouse Electric Corporation Bettis Atomic Power Laboratory. Seventeen entire fuel rods from the end-of-life core of the LWBR were sheared into 169 precisely characterized segments, and more than 150 of these segments were dissolved during execution of the LWBR Proof-of-Breeding (LWBR-POB) Analytical Support Project at Argonne National Laboratory. The procedures illustrate our approaches to process monitoring, data reduction, and quality assurance during the LWBR-POB work.

  13. DIissolution of low enriched uranium from the experimental breeder reactor-II fuel stored at the Idaho National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Daniel, G. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Rudisill, T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Almond, P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); O' Rourke, P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-06-28

    The Idaho National Laboratory (INL) is actively engaged in the development of electrochemical processing technology for the treatment of fast reactor fuels using irradiated fuel from the Experimental Breeder Reactor-II (EBR-II) as the primary test material. The research and development (R&D) activities generate a low enriched uranium (LEU) metal product from the electrorefining of the EBR-II fuel and the subsequent consolidation and removal of chloride salts by the cathode processor. The LEU metal ingots from past R&D activities are currently stored at INL awaiting disposition. One potential disposition pathway is the shipment of the ingots to the Savannah River Site (SRS) for dissolution in H-Canyon. Carbon steel cans containing the LEU metal would be loaded into reusable charging bundles in the H-Canyon Crane Maintenance Area and charged to the 6.4D or 6.1D dissolver. The LEU dissolution would be accomplished as the final charge in a dissolver batch (following the dissolution of multiple charges of spent nuclear fuel (SNF)). The solution would then be purified and the 235U enrichment downblended to allow use of the U in commercial reactor fuel. To support this potential disposition path, the Savannah River National Laboratory (SRNL) developed a dissolution flowsheet for the LEU using samples of the material received from INL.

  14. BRENDA: a dynamic simulator for a sodium-cooled fast reactor power plant

    Energy Technology Data Exchange (ETDEWEB)

    Hetrick, D.L.; Sowers, G.W.

    1978-06-01

    This report is a users' manual for one version of BRENDA (Breeder Reactor Nuclear Dynamic Analysis), which is a digital program for simulating the dynamic behavior of a sodium-cooled fast reactor power plant. This version, which contains 57 differential equations, represents a simplified model of the Clinch River Breeder Reactor Project (CRBRP). BRENDA is an input deck for DARE P (Differential Analyzer Replacement, Portable), which is a continuous-system simulation language developed at the University of Arizona. This report contains brief descriptions of DARE P and BRENDA, instructions for using BRENDA in conjunction with DARE P, and some sample output. A list of variable names and a listing for BRENDA are included as appendices.

  15. The passive nondestructive assay of the plutonium content of spent-fuel assemblies from the BN-350 fast-breeder reactor in the city of Aqtau, Kazakhstan

    CERN Document Server

    Lestone, J P; Rennie, J A; Sprinkle, J K; Staples, P; Grimm, K N; Hill, R N; Cherradi, I; Islam, N; Koulikov, J; Starovich, Z

    2002-01-01

    The International Atomic Energy Agency is presently interested in developing equipment and techniques to measure the plutonium content of breeder reactor spent-fuel assemblies located in storage ponds before they are relocated to more secure facilities. We present the first quantitative nondestructive assay of the plutonium content of fast-breeder reactor spent-fuel assemblies while still underwater in their facility storage pond. We have calibrated and installed an underwater neutron coincidence counter (Spent Fuel Coincidence Counter (SFCC)) in the BN-350 reactor spent-fuel pond in Aqtau, Kazakhstan. A procedure has been developed to convert singles and doubles (coincidence) neutron rates observed by the SFCC into the total plutonium content of a given BN-350 spent-fuel assembly. The plutonium content has been successfully determined for spent-fuel assemblies with a contact radiation level as high as approx 10 sup 5 Rads/h. Using limited facility information and multiple measurements along the length of spe...

  16. International treaty on plant genetic resources for food and agriculture as a basis for limiting intellectual property of plant breeders in Serbia

    Directory of Open Access Journals (Sweden)

    Veselinović Janko

    2014-01-01

    Full Text Available It is widely accepted that new plant varieties are protected as a form of intellectual property. Serbian legislation and international conventions protect new plant varieties, either through sui generis protection or by patent rights. However, International Treaty on Plant Genetic Resources for Food and Agriculture, also signed by Serbia, has endangered the rights of plant breeders in Serbia, because the greatest number of plant varieties in Serbia is not protected as intellectual property. On the other hand, the Treaty offers the possibility to natural and legal persons from other signatory countries to use new plant varieties of Serbian plant breeders. Seeing as the varieties are not protected, there is a possibility for the plant breeders' rights to be endangered. The goal of this paper is to try and consider the legal consequences, answering certain open questions regarding the protection of intellectual property in this field. A method of parallel comparison was used.

  17. Automated operator procedure prompting for startup of Experimental Breeder Reactor-2

    Energy Technology Data Exchange (ETDEWEB)

    Renshaw, A.W.; Ball, S.J.; Ford, C.E.

    1990-11-01

    This report describes the development of an operator procedure prompting aid for startup of a nuclear reactor. This operator aid is a preliminary design for a similar aid that eventually will be used with the Advanced Liquid Metal Reactor (ALMR) presently in the design stage. Two approaches were used to develop this operator procedure prompting aid. One method uses an expert system software shell, and the other method uses database software. The preliminary requirements strongly pointed toward features traditionally associated with both database and expert systems software. Database software usually provides data manipulation flexibility and user interface tools, and expert systems tools offer sophisticated data representation and reasoning capabilities. Both methods, including software and associated hardware, are described in this report. Proposals for future enhancements to improve the expert system approach to procedure prompting and for developing other operator aids are also offered. 25 refs., 14 figs.

  18. An express route to perfection: Ornamental plant breeders have the tools at their disposal to expedite breeding for specific traits

    Science.gov (United States)

    Plants inherit traits the same way Gregor Mendel observed in his famous experiments with peas in the 1800s. Breeders select parents with one or more desirable traits and then look for superior combinations of these traits in their offspring. Seedlings must be rigorously evaluated for the desirable...

  19. Techniques for processing remote field eddy current signals from bend regions of steam generator tubes of prototype fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Thirunavukkarasu, S. [Non Destructive Evaluation Division, Indira Gandhi Centre for Atomic Research, Kalpakkam, TN 603 102 (India); Rao, B.P.C., E-mail: bpcrao@igcar.gov.in [Non Destructive Evaluation Division, Indira Gandhi Centre for Atomic Research, Kalpakkam, TN 603 102 (India); Jayakumar, T.; Raj, Baldev [Non Destructive Evaluation Division, Indira Gandhi Centre for Atomic Research, Kalpakkam, TN 603 102 (India)

    2011-04-15

    Steam generator (SG) is one of the most critical components of sodium cooled fast breeder reactor. Remote field eddy current (RFEC) technique has been chosen for in-service inspection (ISI) of these ferromagnetic SG tubes made of modified 9Cr-1Mo steel (Grade 91). Expansion bends are provided in the SGs to accommodate differential thermal expansion. During ISI using RFEC technique, in expansion bend regions, exciter-receiver coil misalignment, bending stresses, probe wobble and magnetic permeability variations produce disturbing noise hindering detection of defects. Fourier filtering, cross-correlation and wavelet transform techniques have been studied for noise reduction as well as enhancement of RFEC signals of defects in bend regions, having machined grooves and localized defects. Performance of these three techniques has been compared using signal-to-noise ratio (SNR). Fourier filtering technique has shown better performance for noise reduction while cross-correlation technique has resulted in significant enhancement of signals. Wavelet transform technique has shown the combined capability of noise reduction and signal enhancement and resulted in unambiguous detection of 10% of wall loss grooves and localized defects in the bend regions with SNR better than 7 dB.

  20. Core loading pattern optimization of thorium fueled heavy water breeder reactor using genetic algorithm

    Energy Technology Data Exchange (ETDEWEB)

    Soewono, C. N.; Takaki, N. [Dept. of Applied Science Engineering, Faculty Tokai Univ., Kanagawa-ken, Hiratsuka-shi Kitakaname 4-1-1 (Japan)

    2012-07-01

    In this work genetic algorithm was proposed to solve fuel loading pattern optimization problem in thorium fueled heavy water reactor. The objective function of optimization was to maximize the conversion ratio and minimize power peaking factor. Those objectives were simultaneously optimized using non-dominated Pareto-based population ranking optimal method. Members of non-dominated population were assigned selection probabilities based on their rankings in a manner similar to Baker's single criterion ranking selection procedure. A selected non-dominated member was bred through simple mutation or one-point crossover process to produce a new member. The genetic algorithm program was developed in FORTRAN 90 while neutronic calculation and analysis was done by COREBN code, a module of core burn-up calculation for SRAC. (authors)

  1. Conceptual Design Studies of a Passively Safe Thorium Breeder Pebble Bed Reactor

    NARCIS (Netherlands)

    Wols, F.J.

    2015-01-01

    Nuclear power plants are expected to play an important role in the worldwide electricity production in the coming decades, since they provide an economically attractive, reliable and low-carbon source of electricity with plenty of resources available for at least the coming hundreds of years.

  2. Gas cooled fast breeder reactor design for a circulator test facility (modified HTGR circulator test facility)

    Energy Technology Data Exchange (ETDEWEB)

    1979-10-01

    A GCFR helium circulator test facility sized for full design conditions is proposed for meeting the above requirements. The circulator will be mounted in a large vessel containing high pressure helium which will permit testing at the same power, speed, pressure, temperature and flow conditions intended in the demonstration plant. The electric drive motor for the circulator will obtain its power from an electric supply and distribution system in which electric power will be taken from a local utility. The conceptual design decribed in this report is the result of close interaction between the General Atomic Company (GA), designer of the GCFR, and The Ralph M. Parson Company, architect/engineer for the test facility. A realistic estimate of total project cost is presented, together with a schedule for design, procurement, construction, and inspection.

  3. Nuclear reactor kinetics and plant control

    CERN Document Server

    Oka, Yoshiaki

    2013-01-01

    Understanding time-dependent behaviors of nuclear reactors and the methods of their control is essential to the operation and safety of nuclear power plants. This book provides graduate students, researchers, and engineers in nuclear engineering comprehensive information on both the fundamental theory of nuclear reactor kinetics and control and the state-of-the-art practice in actual plants, as well as the idea of how to bridge the two. The first part focuses on understanding fundamental nuclear kinetics. It introduces delayed neutrons, fission chain reactions, point kinetics theory, reactivit

  4. Superalloy applications in the fast breeder reactor. [Alloy-A-286; inconel 706; inconel 718; nimonic PE16; alloy-M-813

    Energy Technology Data Exchange (ETDEWEB)

    Powell, R.W.

    1976-01-01

    The economics of the LMFBR are dependent on the breeding of new fuel in the reactor core and this can be improved by the use of advanced alloys as core structural components. The environment of the core makes superalloys a natural choice for these components, but phenomena related directly to neutron irradiation necessitate extensive testing. Consequently, commercially-available superalloys, together with a number of developmental alloys are being tested in existing LMFBR's and by simulation techniques to determine the best alloy for use in the LMFBR core. It presently appears that such materials will indeed be capable of the performance required, and will greatly facilitate the commercial realization of the fast breeder reactor.

  5. The long-term future for civilian nuclear power generation in France: The case for breeder reactors. Breeder reactors: The physical and physical chemistry parameters, associate material thermodynamics and mechanical engineering: Novelties and issues; La production d'electricite d'origine nucleaire en France, dans le futur a long terme: Le cas des surgenerateurs: Les reacteurs nucleaires surgenerateurs: Les parametres physique et physico-chimiques, la thermodynamique associee des materiaux et de l'ingenierie mecanique: Nouveautes et options

    Energy Technology Data Exchange (ETDEWEB)

    Dautray, R. [Academie des sciences, 23, quai de Conti, 75270 Paris cedex 06 (France)

    2011-06-15

    The author gives a summary overview of the knowledge base acquired since the first breeder reactors became operational in the fifties. Neutron transport theory, thermal phenomena, reactor core cooling, various coolants used and envisioned for this function, fuel fabrication from separated materials, main equipment (pumps, valves, heat exchanges...) have now attained maturity, sufficient to implement sodium cooling circuits. However, the use of metallic sodium still raises certain severe questions in terms of safe handling and security considerations. The structural components, both inside the reactor core and outside (i.e. heat exchangers) are undergoing in-depth research so as to last longer. The fuel cycle, notably the re-fabrication of fuel elements and fertile elements, the case of transuranic elements, etc., call for studies into radiation induced phenomena, chemistry separation, separate or otherwise treatments for materials that have different radioactive, physical, thermodynamical, chemical and biological properties. The concerns that surround the definitive disposal of certain radioactive wastes could be qualitatively improved with respect to the pressurized water reactors (PWRs) in service today. Lastly, the author notes that breeder reactors eliminate the need for an isotope separation facility, and this constitutes a significant contribution to contain nuclear proliferation. France was in the forefront of nuclear breeder power generation science, technological research and also in the knowledge base related to breeder reactors. It is in the country's interest to pursue these efforts. (author)

  6. Capital cost: gas cooled fast reactor plant

    Energy Technology Data Exchange (ETDEWEB)

    1977-09-01

    The results of an investment cost study for a 900 MW(e) GCFR central station power plant are presented. The capital cost estimate arrived at is based on 1976 prices and a conceptual design only, not a mature reactor design.

  7. Generic small modular reactor plant design.

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-12-01

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

  8. Special Analysis for the Disposal of the INL Waste Associated with the Unirradiated Light Water Breeder Reactor (LWBR) Waste Stream at the Area 5 Radioactive Waste Management Site

    Energy Technology Data Exchange (ETDEWEB)

    Shott, Gregory [National Security Technologies, LLC, Las Vegas, NV (United States)

    2017-03-21

    This special analysis (SA) evaluates whether the Idaho National Laboratory (INL) Waste Associated with the Unirradiated Light Water Breeder Reactor (LWBR) waste stream (INEL167203QR1, Revision 0) is suitable for shallow land burial (SLB) at the Area 5 Radioactive Waste Management Site (RWMS) on the Nevada National Security Site (NNSS). Disposal of the INL Waste Associated with the Unirradiated LWBR waste meets all U.S. Department of Energy (DOE) Manual DOE M 435.1-1, “Radioactive Waste Management Manual,” Chapter IV, Section P performance objectives (DOE 1999). The INL Waste Associated with the Unirradiated LWBR waste stream is recommended for acceptance with the condition that the total uranium-233 (233U) inventory be limited to 2.7E13 Bq (7.2E2 Ci).

  9. 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...

  10. Feasibility of processing the experimental breeder reactor-II driver fuel from the Idaho National Laboratory through Savannah River Site's H-Canyon facility

    Energy Technology Data Exchange (ETDEWEB)

    Magoulas, V. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-07-28

    Savannah River National Laboratory (SRNL) was requested to evaluate the potential to receive and process the Idaho National Laboratory (INL) uranium (U) recovered from the Experimental Breeder Reactor II (EBR-II) driver fuel through the Savannah River Site’s (SRS) H-Canyon as a way to disposition the material. INL recovers the uranium from the sodium bonded metallic fuel irradiated in the EBR-II reactor using an electrorefining process. There were two compositions of EBR-II driver fuel. The early generation fuel was U-5Fs, which consisted of 95% U metal alloyed with 5% noble metal elements “fissium” (2.5% molybdenum, 2.0% ruthenium, 0.3% rhodium, 0.1% palladium, and 0.1% zirconium), while the later generation was U-10Zr which was 90% U metal alloyed with 10% zirconium. A potential concern during the H-Canyon nitric acid dissolution process of the U metal containing zirconium (Zr) is the explosive behavior that has been reported for alloys of these materials. For this reason, this evaluation was focused on the ability to process the lower Zr content materials, the U-5Fs material.

  11. Control rod calibration methods for fast breeder reactors applied to Phenix; Les methodes d'etalonnage des barres de commande des reacteurs a neutrons rapides application a Phenix

    Energy Technology Data Exchange (ETDEWEB)

    Lecourt, G

    1998-06-18

    The control and the emergency shutdown of a fast breeder reactor depends essentially on control rods. For this reason, it is imperative to know exactly how much anti reactivity is introduced with the rods in the reactor core. Different methods have been compared in order to see if they are compatible with Phenix reactor. Their limits have been studied. The shadow and anti shadow effects that can the rods make one to the other and then their effective weight of the rods screen have been clarified. (N.C.)

  12. Reactor safety determination Savannah River Plant

    Energy Technology Data Exchange (ETDEWEB)

    Menegus, R.I.; Ring, H.F.

    1956-02-01

    Large quantities of radioactive material are present in the Reactor and Separations Areas of the Savannah River Plant. This report deals with the precautions that were taken in the design of the plant to guard against the escape of radioactive material, and with the various devices and procedures that have been provided to assure safe operation under foreseeable operating conditions. Because no amount of care and foresight can guarantee against the risk of accident, one must assume that there is a remote possibility that a serious release of radioactive material can occur even in a very carefully designed plant. Once liberated to the surroundings, radioactive material may be spread by wind and water. No effective control of these agents is possible. The obvious preventive measure is isolation of the plant site from centers of population. So far as is practicable, this has been done at Savannah River, but the need for communications, power and a work force makes it impossible to secure the full benefits that might be derived from complete isolation of the plant. In order to calculate the magnitude of the hazard to people outside the plant site, it is necessary to assume the occurrence of highly improbable events - the very events that this design seeks to prevent by careful design and operation. Nevertheless, if one assumes a release of radioactivity, one must estimate the airborne and waterborne radioactive burdens carried off the plant site, if one wishes to evaluate the hazard to which the off-site populace is exposed. The climate and the geology of the plant site determine the direction of transport and the dilution of the radioactive material. This information, together with data on the distribution of the population around the site, may be used to estimate the damage to be anticipated in the improbable event of the release of large quantities of radioactive material.

  13. Capital Cost: Pressurized Water Reactor Plant Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    1977-06-01

    The investment cost study for the 1139-MW(e) pressurized water reactor (PWR) central station power plant consists of two volumes. This volume includes in addition to the foreword and summary, the plant description and the detailed cost estimate.

  14. Numerical analysis of thermal stratification phenomena in upper plenum of a fast breeder reactor (1). Evaluation of thermal stratification phenomena near the region of flow holes

    Energy Technology Data Exchange (ETDEWEB)

    Suda, Kazunori; Muramatu, Toshiharu; Yamaguchi, Akira [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center

    2002-12-01

    Thermal stratification phenomena are observed in an upper plenum of liquid metal fast breeder reactors (LMFBRs) under reactor scram conditions, which give rise to thermal stress on structural components. Therefore it is important to evaluate characteristics of the phenomena in the design on the internal structures in an LMFBR plenum. To evaluate long-term characteristics of thermal stratification phenomena in a typical LMFBR upper plenum, numerical analysis was carried out with a multi-dimensional thermohydraulics code AQUA for a scram event from full power operation condition. Thereafter the numerical results were compared with extrapolated results of measured transient data on the 40% operation condition. From the thermohydraulic analysis by the AQUA code, the following results have been obtained. (Long-term characteristics of thermal stratification phenomena) The cold fluid region near the inside inner barrel was expanded with accumulation of the cold fluid in the lower region of the plenum after 300 seconds from the reactor scram, so that the fluid from core flowed to the lower region of the upper plenum. The characteristics of axial temperature distributions in the upper plenum were similar to them at the 300 seconds. The thermal stratification interface was located initially around intermediate position between upper lower flow holes. And an another thermal stratification interface was formed around the inner barrel support plate after 300 seconds from the scram, so that the cold fluid accumulated in the lower region of the plenum. But the thermal stratification interface around the inner barrel support plate was disappeared by mixture and heat conduction of coolant of circumferential direction. The thermal stratification interface which was located below in the upper flow holes, rose to the upward position of the upper flow holes at the 720 seconds. In annular gap region between the inner barrel and the reactor vessel wall, thermal stratification interface

  15. Liquid metal cooled nuclear reactor plant system

    Science.gov (United States)

    Hunsbedt, Anstein; Boardman, Charles E.

    1993-01-01

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

  16. Large scale breeder reactor plant prototype mechanical pump conceptual design study

    Energy Technology Data Exchange (ETDEWEB)

    1976-07-01

    This report includes engineering memorandums, drawings, key feature descriptions, and other data. Some of the reports, such as manufacturability and some stress analysis, were done by consultants for Byron Jackson. Review of this report indicates that the design is feasible. The pump can be manufactured to system and specification requirements. The overall length and weight of some pieces will require special consideration, but is within the scope of equipment and technology available today. The fabricated parts are large and heavy, but can be manufactured and machined. Only the high temperature is unique to this size, since previous sodium pumps were smaller. Nondestructive tests as required by the Code are described and are feasible. The performance test of the prototype has been studied thoroughly. It is feasible for a cold water test. There are some problem areas. However, all of them can be solved. Development needs include building and testing a small scale model.

  17. Reactor core and plant design concepts of the Canadian supercritical water-cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yetisir, M.; Gaudet, M.; Bailey, J.; Rhodes, D.; Guzonas, D.; Hamilton, H.; Haque, Z.; Pencer, J.; Sartipi, A. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2014-07-01

    Canada is developing a 1200 MWe supercritical water-cooled reactor (SCWR), which has evolved from the well-established pressure-tube type CANDU{sup 1} reactor. This SCWR reactor concept, which is often referred to as the Canadian SCWR, uses supercritical water as a coolant, has a low-pressure heavy water moderator and a direct cycle for power production. The reactor concept incorporates advanced safety features, such as passive emergency core cooling, long-term decay heat rejection to the environment and fuel melt prevention via passive moderator cooling. These features significantly reduce the core damage frequency beyond existing nuclear reactors. This paper presents a description of the Canadian SCWR core design concept, the integration of in-core and out-of-core components and the mechanical plant design concept. Supporting systems for reactor safety, reactor control and moderator cooling are also described. (author)

  18. Thermal hydraulic parametric investigation of decay heat removal from degraded core of a sodium cooled fast Breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Verma, Lokesh [Department of Physics and Astrophysics, University of Delhi, Delhi 110007 (India); Kumar Sharma, Anil, E-mail: aksharma@igcar.gov.in [Reactor Design Group, Indira Gandhi Centre for Atomic Research, HBNI, Kalpakkam (India); Velusamy, K. [Reactor Design Group, Indira Gandhi Centre for Atomic Research, HBNI, Kalpakkam (India)

    2017-03-15

    Highlights: • Decay heat removal from degraded core of a typical SFR is highlighted. • Influence of number of DHXs in operation on PAHR is analyzed. • Investigations on structural integrity of the inner vessel and core catcher. • Feasibility study for retention of a part of debris in upper pool of SFR. - Abstract: Ensuring post accident decay heat removal with high degree of reliability following a Core Disruptive Accident (CDA) is very important in the design of sodium cooled fast reactors (SFR). In the recent past, a lot of research has been done towards the design of an in-vessel core catcher below the grid plate to prevent the core debris reaching the main vessel in a pool type SFR. However, during an energetic CDA, the entire core debris is unlikely to reach the core catcher. A significant part of the debris is likely to settle in core periphery between radial shielding subassemblies and the inner vessel. Failure of inner vessel due to the decay heat can lead to core debris reaching the main vessel and threatening its integrity. On the other hand, retention of a part of debris in core periphery can reduce the load on main core catcher. Towards achieving an optimum design of SFR and safety evaluation, it is essential to quantify the amount of heat generating core debris that can be retained safely within the primary vessel. This has been performed by a mathematical simulation comprising solution of 2-D transient form of the governing equations of turbulent sodium flow and heat transfer with Boussinesq approximations. The conjugate conduction-convection model adopted for this purpose is validated against in-house experimental data. Transient evolutions of natural convection in the pools and structural temperatures in critical components have been predicted. It is found that 50% of the core debris can be safely accommodated in the gap between radial shielding subassemblies and inner vessel without exceeding structural temperature limit. It is also

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

    Energy Technology Data Exchange (ETDEWEB)

    Lowry, L.L. (comp.)

    1977-09-01

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

  20. Thermal breeder fuel enrichment zoning

    Science.gov (United States)

    Capossela, Harry J.; Dwyer, Joseph R.; Luce, Robert G.; McCoy, Daniel F.; Merriman, Floyd C.

    1992-01-01

    A method and apparatus for improving the performance of a thermal breeder reactor having regions of higher than average moderator concentration are disclosed. The fuel modules of the reactor core contain at least two different types of fuel elements, a high enrichment fuel element and a low enrichment fuel element. The two types of fuel elements are arranged in the fuel module with the low enrichment fuel elements located between the high moderator regions and the high enrichment fuel elements. Preferably, shim rods made of a fertile material are provided in selective regions for controlling the reactivity of the reactor by movement of the shim rods into and out of the reactor core. The moderation of neutrons adjacent the high enrichment fuel elements is preferably minimized as by reducing the spacing of the high enrichment fuel elements and/or using a moderator having a reduced moderating effect.

  1. Helium-cooled molten-salt fusion breeder

    Energy Technology Data Exchange (ETDEWEB)

    Moir, R.W.; Lee, J.D.; Fulton, F.J.; Huegel, F.; Neef, W.S. Jr.; Sherwood, A.E.; Berwald, D.H.; Whitley, R.H.; Wong, C.P.C.; Devan, J.H.

    1984-12-01

    We present a new conceptual design for a fusion reactor blanket that is intended to produce fissile material for fission power plants. Fast fission is suppressed by using beryllium instead of uranium to multiply neutrons. Thermal fission is suppressed by minimizing the fissile inventory. The molten-salt breeding medium (LiF + BeF/sub 2/ + ThF/sub 4/) is circulated through the blanket and to the on-line processing system where /sup 233/U and tritium are continuously removed. Helium cools the blanket and the austenitic steel tubes that contain the molten salt. Austenitic steel was chosen because of its ease of fabrication, adequate radiation-damage lifetime, and low corrosion by molten salt. We estimate that a breeder having 3000 MW of fusion power will produce 6500 kg of /sup 233/U per year. This amount is enough to provide makeup for 20 GWe of light-water reactors per year or twice that many high-temperature gas-cooled reactors or Canadian heavy-water reactors. Safety is enhanced because the afterheat is low and blanket materials do not react with air or water. The fusion breeder based on a pre-MARS tandem mirror is estimated to cost $4.9B or 2.35 times a light-water reactor of the same power. The estimated cost of the /sup 233/U produced is $40/g for fusion plants costing 2.35 times that of a light-water reactor if utility owned or $16/g if government owned.

  2. Upgrading program of the experimental fast reactor Joyo

    Energy Technology Data Exchange (ETDEWEB)

    Yoshida, A.; Yogo, S. [Japan Nuclear Cycle Development Institute, Iibaraki-Ken (Japan)

    2001-07-01

    The experimental fast reactor Joyo finished its operation as an irradiation core in June, 2000. Throughout the operation of MK-I (breeder core) and MK-II (irradiation core), the net operation time has exceeded 60,000 hours. During these operations there were no fuel failures or serious plant problems. The MK-III modification program will improve irradiation capability to demonstrate advanced technologies for commercial Fast Breeder Reactor (FBR). When the MK-III core is started, it will support irradiation tests in feasibility studies for fast reactor and related fuel cycle research and development in Japan. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

    1979-11-01

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

  4. 77 FR 3009 - Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors

    Science.gov (United States)

    2012-01-20

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors..., ``Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors.'' DATES...

  5. 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

  6. Special Analysis for the Disposal of the Idaho National Laboratory Unirradiated Light Water Breeder Reactor Rods and Pellets Waste Stream at the Area 5 Radioactive Waste Management Site, Nevada National Security Site, Nye County, Nevada

    Energy Technology Data Exchange (ETDEWEB)

    None

    2014-08-31

    The purpose of this special analysis (SA) is to determine if the Idaho National Laboratory (INL) Unirradiated Light Water Breeder Reactor (LWBR) Rods and Pellets waste stream (INEL103597TR2, Revision 2) is suitable for disposal by shallow land burial (SLB) at the Area 5 Radioactive Waste Management Site (RWMS). The INL Unirradiated LWBR Rods and Pellets waste stream consists of 24 containers with unirradiated fabricated rods and pellets composed of uranium oxide (UO2) and thorium oxide (ThO2) fuel in zirconium cladding. The INL Unirradiated LWBR Rods and Pellets waste stream requires an SA because the 229Th, 230Th, 232U, 233U, and 234U activity concentrations exceed the Nevada National Security Site (NNSS) Waste Acceptance Criteria (WAC) Action Levels.

  7. The radiochemistry of nuclear power plants with light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Neeb, K.H.

    1997-12-31

    In this book, radioactivity and the chemical reactions of radionuclides within the different areas of a nucler power plant are discussed. The text concentrates on commercially operated light water reactors which currently represent the greatest fraction by far of the world`s nuclear power capacity. This book is not only intended for experts working in the various fields of radiochemistry in nuclear power plants. It also provides an overview of the topics dealt with for the operators of nuclear power plants, for people working in design and development and safety-related areas, as well as for those working in licensing and supervision. (orig.)

  8. Modeling Operating Modes for the Monju Nuclear Power Plant

    DEFF Research Database (Denmark)

    Lind, Morten; Yoshikawa, H.; Jørgensen, Sten Bay

    2012-01-01

    The specification of supervision and control tasks in complex processes requires definition of plant states on various levels of abstraction related to plant operation in start-up, normal operation and shut-down. Modes of plant operation are often specified in relation to a plant decomposition in...... for the Japanese fast breeder reactor plant MONJU....

  9. Study of reactor plant disturbed cooling condition modes caused by the VVER reactor secondary circuit

    Directory of Open Access Journals (Sweden)

    V.I. Belozerov

    2016-12-01

    Based on the RELAP-5, TRAC, and TRACE software codes, reactor plant cooling condition malfunction modes caused by the VVER-1000 secondary circuit were simulated and investigated. Experimental data on the mode with the turbine-generator stop valve closing are presented. The obtained dependences made it possible to determine the maximum values of pressure and temperature in the circulation circuit as well as estimate the Minimum Critical Heat Flux Ratio (MCHFR. It has been found that, if any of the initial events occurs, safety systems are activated according to the set points; transient processes are stabilized in time; and the Critical Heat Flux (CHF limit is provided. Therefore, in the event of emergency associated with the considered modes, the reactor plant safety will be ensured.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-12-11

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

  11. Fabrication, properties, and tritium recovery from solid breeder materials

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, C.E. (Argonne National Lab., IL (USA)); Kondo, T. (Japan Atomic Energy Research Inst., Tokyo (Japan)); Roux, N. (CEA Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France)); Tanaka, S. (Tokyo Univ. (Japan)); Vollath, D. (Kernforschungszentrum Karlsruhe GmbH (Germany, F.R.))

    1991-01-01

    The breeding blanket is a key component of the fusion reactor because it directly involves tritium breeding and energy extraction, both of which are critical to development of fusion power. The lithium ceramics continue to show promise as candidate breeder materials. This promise was recognized by the International Thermonuclear Experimental Reactor (ITER) design team in its selection of ceramics as the first option for the ITER breeder material. Blanket design studies have indicated properties in the candidate materials data base that need further investigation. Current studies are focusing on tritium release behavior at high burnup, changes in thermophysical properties with burnup, compatibility between the ceramic breeder and beryllium multiplier, and phase changes with burnup. Laboratory and in-reactor tests, some as part of an international collaboration for development of ceramic breeder materials, are underway. 133 refs., 1 fig.

  12. Design of a helium-cooled molten salt fusion breeder

    Energy Technology Data Exchange (ETDEWEB)

    Moir, R.W.; Lee, J.D.; Fulton, F.J.; Huegel, F.; Neef, W.S. Jr.; Sherwood, A.E.; Berwald, D.H.; Whitley, R.H.; Wong, C.P.C.; DeVan, J.H.

    1985-02-01

    A new conceptual blanket design for a fusion reactor produces fissile material for fission power plants. Fission is suppressed by using beryllium, rather than uranium, to multiply neutrons and also by minimizing the fissile inventory. The molten-salt breeding media (LiF + BeF/sub 2/ + TghF/sub 4/) is circulated through the blanket and on to the online processing system where /sup 233/U and tritium are continuously removed. Helium cools the blanket including the steel pipes containing the molten salt. Austenitic steel was chosen because of its ease of fabrication, adequate radiation-damage lifetime, and low corrosion rate by molten salt. We estimate the breeder, having 3000 MW of fusion power, produces 6400 kg of /sup 233/U per year, which is enough to provide make up for 20 GWe of LWR per year (or 14 LWR plants of 4440 MWt) or twice that many HTGRs or CANDUs. Safety is enhanced because the afterheat is low and the blanket materials do not react with air or water. The fusion breeder based on a pre-MARS tandem mirror is estimated to cost $4.9B or 2.35 times an LWR of the same power. The estimated present value cost of the /sup 2/anumber/sup 3/U produced is $40/g if utility financed or $16/g if government financed.

  13. Overview of pool hydraulic design of Indian prototype fast breeder ...

    Indian Academy of Sciences (India)

    Computational Fluid Dynamics (CFD) plays a critical role in the design of pool type reactors and becomes an increasingly popular tool, thanks to the advancements in computing technology. In this paper, thermal hydraulic characteristics of a fast breeder reactor, design limits and challenging thermal hydraulic investigations ...

  14. Initiating Events for Multi-Reactor Plant Sites

    Energy Technology Data Exchange (ETDEWEB)

    Muhlheim, Michael David [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Flanagan, George F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Poore, III, Willis P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2014-09-01

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

  15. A contribution to the analysis of the thermal behaviour of Fast Breeder fuel rods with UO{sub 2}-PuO{sub 2} fuel; Contribucion al analisis del comportamiento termico de las barras combustibles de UO{sub 2}-PuO{sub 2} de los reactores rapidos

    Energy Technology Data Exchange (ETDEWEB)

    Lopez Jimenez, J.; Elbel, H.

    1977-07-01

    The fuel of Fast Breeder Reactors which consists of Uranium and Plutonium dioxide is mainly characterized by the amount and distribution of void volume and Plutonium and the amount of oxygen. Irradiation experiments carried out with this fuel have shown that initial structure of the fuel pellet is subjected to large changes during operation. These are consequences of the radial and axial temperature gradients within the fuel rods. (Author) 54 refs.

  16. Tridimensional ultrasonic images analysis for the in service inspection of fast breeder reactors; Analyse d'images tridimensionnelles ultrasonores pour l'inspection en service des reacteurs a neutrons rapides

    Energy Technology Data Exchange (ETDEWEB)

    Dancre, M

    1999-11-01

    Tridimensional image analysis provides a set of methods for the intelligent extraction of information in order to visualize, recognize or inspect objects in volumetric images. In this field of research, we are interested in algorithmic and methodological aspects to extract surface visual information embedded in volume ultrasonic images. The aim is to help a non-acoustician operator, possibly the system itself, to inspect surfaces of vessel and internals in Fast Breeder Reactors (FBR). Those surfaces are immersed in liquid metal, what justifies the ultrasonic technology choice. We expose firstly a state of the art on the visualization of volume ultrasonic images, the methods of noise analysis, the geometrical modelling for surface analysis and finally curves and surfaces matching. These four points are then inserted in a global analysis strategy that relies on an acoustical analysis (echoes recognition), an object analysis (object recognition and reconstruction) and a surface analysis (surface defects detection). Few literature can be found on ultrasonic echoes recognition through image analysis. We suggest an original method that can be generalized to all images with structured and non-structured noise. From a technical point of view, this methodology applied to echoes recognition turns out to be a cooperative approach between morphological mathematics and snakes (active contours). An entropy maximization technique is required for volumetric data binarization. (author)

  17. Nuclear plant-aging research on reactor protection systems

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, L.C.

    1988-01-01

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

  18. Structural response of 1/20-scale models of the Clinch River Breeder Reactor to a simulated hypothetical core-disruptive accident

    Energy Technology Data Exchange (ETDEWEB)

    Romander, C M; Cagliostro, D J

    1978-10-01

    Five experiments were performed to help evaluate the structural integrity of the reactor vessel and head design and to verify code predictions. In the first experiment (SM 1), a detailed model of the head was loaded statically to determine its stiffness. In the remaining four experiments (SM 2 to SM 5), models of the vessel and head were loaded dynamically under a simulated 661 MW-s hypothetical core disruptive accident (HCDA). Models SM 2 to SM 4, each of increasing complexity, systematically showed the effects of upper internals structures, a thermal liner, core support platform, and torospherical bottom on vessel response. Model SM 5, identical to SM 4 but more heavily instrumented, demonstrated experimental reproducibility and provided more comprehensive data. The models consisted of a Ni 200 vessel and core barrel, a head with shielding and simulated component masses, and an upper internals structure (UIS).

  19. Power generation costs for alternate reactor fuel cycles

    Energy Technology Data Exchange (ETDEWEB)

    Smolen, G.R.; Delene, J.G.

    1980-09-01

    The total electric generating costs at the power plant busbar are estimated for various nuclear reactor fuel cycles which may be considered for power generation in the future. The reactor systems include pressurized water reactors (PWR), heavy-water reactors (HWR), high-temperature gas cooled reactors (HTGR), liquid-metal fast breeder reactors (LMFBR), light-water pre-breeder and breeder reactors (LWPR, LWBR), and a fast mixed spectrum reactor (FMSR). Fuel cycles include once-through, uranium-only recycle, and full recycle of the uranium and plutonium in the spent fuel assemblies. The U/sub 3/O/sub 8/ price for economic transition from once-through LWR fuel cycles to both PWR recycle and LMFBR systems is estimated. Electric power generation costs were determined both for a reference set of unit cost parameters and for a range of uncertainty in these parameters. In addition, cost sensitivity parameters are provided so that independent estimations can be made for alternate cost assumptions.

  20. Laser fusion driven breeder design study. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Berwald, D.H.; Massey, J.V.

    1980-12-01

    The results of the Laser Fusion Breeder Design Study are given. This information primarily relates to the conceptual design of an inertial confinement fusion (ICF) breeder reactor (or fusion-fission hybrid) based upon the HYLIFE liquid metal wall protection concept developed at Lawrence Livermore National Laboratory. The blanket design for this breeder is optimized to both reduce fissions and maximize the production of fissile fuel for subsequent use in conventional light water reactors (LWRs). When the suppressed fission blanket is compared with its fast fission counterparts, a minimal fission rate in the blanket results in a unique reactor safety advantage for this concept with respect to reduced radioactive inventory and reduced fission product decay afterheat in the event of a loss-of-coolant-accident.

  1. Dissipation of the reactor heat at the Savannah River Plant

    Energy Technology Data Exchange (ETDEWEB)

    Neill, J.S.; Babcock, D.F.

    1971-10-01

    The effluent cooling water from the heat exchangers of the Savannah River nuclear reactors is cooled by natural processes as it flows through the stream beds, canals, ponds, and swamps on the plant site. The Langhaar equation, which gives the rate of heat removal from the water surface as a function of the surface temperature, air temperature, relative humidity, and wind speed, is applied satisfactorily to calculate the cooling that occurs at all temperature levels and for all modes of water flow. The application of this equation requires an accounting of effects such as solar heating, shading, mixing, staging, stratification, underflow, rainfall, the imposed heat load, and the rate of change in heat content of the body of water.

  2. Structural response of 1/20-scale models of the Clinch River Breeder Reactor to a simulated hypothetical core disruptive accident. Technical report 4

    Energy Technology Data Exchange (ETDEWEB)

    Romander, C. M.; Cagliostro, D. J.

    1978-10-01

    Five experiments were performed to help evaluate the structural integrity of the reactor vessel and head design and to verify code predictions. In the first experiment (SM 1), a detailed model of the head was loaded statically to determine its stiffness. In the remaining four experiments (SM 2 to SM 5), models of the vessel and head were loaded dynamically under a simulated 661 MW-sec hypothetical core disruptive accident (HCDA). Models SM 2 to SM 4, each of increasing complexity, systematically showed the effects of upper internals structures, a thermal liner, core support platform, and torospherical bottom on vessel response. Model SM 5, identical to SM 4 but more heavily instrumented, demonstrated experimental reproducibility and provided more comprehensive data. The models consisted of a Ni 200 vessel and core barrel, a head with shielding and simulated component masses, an upper internals structure (UIS), and, in the more complex models SM 4 and SM 5, a Ni 200 thermal liner and core support structure. Water simulated the liquid sodium coolant and a low-density explosive simulated the HCDA loads.

  3. IAEA coordinated research program on `harmonization and validation of fast reactor thermomechanical and thermohydraulic codes using experimental data`. 1. Thermohydraulic benchmark analysis on high-cycle thermal fatigue events occurred at French fast breeder reactor Phenix

    Energy Technology Data Exchange (ETDEWEB)

    Muramatsu, Toshiharu [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1997-06-01

    A benchmark exercise on `Tee junction of Liquid Metal Fast Reactor (LMFR) secondary circuit` was proposed by France in the scope of the said Coordinated Research Program (CRP) via International Atomic Energy Agency (IAEA). The physical phenomenon chosen here deals with the mixture of two flows of different temperature. In a LMFR, several areas of the reactor are submitted to this problem. They are often difficult to design, because of the complexity of the phenomena involved. This is one of the major problems of the LMFRs. This problem has been encountered in the Phenix reactor on the secondary loop, where defects in a tee junction zone were detected during a campaign of inspections after an operation of 90,000 hours of the reactor. The present benchmark is based on an industrial problem and deal with thermal striping phenomena. Problems on pipes induced by thermal striping phenomena have been observed in some reactors and experimental facilities coolant circuits. This report presents numerical results on thermohydraulic characteristics of the benchmark problem, carried out using a direct numerical simulation code DINUS-3 and a boundary element code BEMSET. From the analysis with both the codes, it was confirmed that the hot sodium from the small pipe rise into the cold sodium of the main pipe with thermally instabilities. Furthermore, it was indicated that the coolant mixing region including the instabilities agrees approximately with the result by eye inspections. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-08-01

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

  5. Development of technology for next generation reactor - Research of evaluation technology for nuclear power plant -

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Kyun; Chang, Moon Heuy; Hwang, Yung Dong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)] [and others

    1993-09-01

    For development of next generation reactor, a project for evaluation technology for nuclear power plant is performed. Evaluation technology is essential to next generation reactor for reactor safety and system analysis. For design concept, detailed evaluation technologies are studied as follows: evaluation of safety margin, evaluation of safety facilities, evaluation of measurement and control technology; man-machine interface. Especially for thermal efficiency, thermal properties and chemical composition of inconel 690 tube, instead of inconel 600 tube, are measured for steam generator. (Author).

  6. Gas Reactor International Cooperative program. Pebble bed reactor plant: screening evaluation. Volume 3. Appendix A. Equipment list

    Energy Technology Data Exchange (ETDEWEB)

    1979-11-01

    This report consists of three volumes which describe the design concepts and screening evaluation for a 3000 MW(t) Pebble Bed Reactor Multiplex Plant (PBR-MX). The Multiplex plant produces both electricity and transportable chemical energy via the thermochemical pipeline (TCP). The evaluation was limited to a direct cycle plant which has the steam generators and steam reformers in the primary circuit. Volume 1 reports the overall plant and reactor system and was prepared by the General Electric Company. Core scoping studies were performed which evaluated the effects of annular and cylindrical core configurations, radial blanket zones, burnup, and ball heavy metal loadings. The reactor system, including the PCRV, was investigated for both the annular and cylindrical core configurations. Volume 3 is an Appendix containing the equipment list for the plant and was also prepared by United Engineers and Constructors, Inc. It tabulates the major components of the plant and describes each in terms of quantity, type, orientation, etc., to provide a basis for cost estimation.

  7. Measurement and Analysis of Structural Integrity of Reactor Core Support Structure in Pressurized Water Reactor (PWR) Plant

    Science.gov (United States)

    Ansari, Saleem A.; Haroon, Muhammad; Rashid, Atif; Kazmi, Zafar

    2017-02-01

    Extensive calculation and measurements of flow-induced vibrations (FIV) of reactor internals were made in a PWR plant to assess the structural integrity of reactor core support structure against coolant flow. The work was done to meet the requirements of the Fukushima Response Action Plan (FRAP) for enhancement of reactor safety, and the regulatory guide RG-1.20. For the core surveillance measurements the Reactor Internals Vibration Monitoring System (IVMS) has been developed based on detailed neutron noise analysis of the flux signals from the four ex-core neutron detectors. The natural frequencies, displacement and mode shapes of the reactor core barrel (CB) motion were determined with the help of IVMS. The random pressure fluctuations in reactor coolant flow due to turbulence force have been identified as the predominant cause of beam-mode deflection of CB. The dynamic FIV calculations were also made to supplement the core surveillance measurements. The calculational package employed the computational fluid dynamics, mode shape analysis, calculation of power spectral densities of flow & pressure fields and the structural response to random flow excitation forces. The dynamic loads and stiffness of the Hold-Down Spring that keeps the core structure in position against upward coolant thrust were also determined by noise measurements. Also, the boron concentration in primary coolant at any time of the core cycle has been determined with the IVMS.

  8. Capital cost: pressurized water reactor plant. Commercial electric power cost studies

    Energy Technology Data Exchange (ETDEWEB)

    1977-06-01

    The investment cost study for the 1139 MW(e) pressurized water reactor (PWR) central station power plant consists of two volumes. This volume contains the drawings, equipment list and site description.

  9. ITER solid breeder blanket materials database

    Energy Technology Data Exchange (ETDEWEB)

    Billone, M.C. [Argonne National Lab., IL (United States); Dienst, W. [Kernforschungszentrum Karlsruhe GmbH (Germany). Inst. fuer Material- und Festkoerperforschung; Flament, T. [CEA Centre d`Etudes de Fontenay-aux-Roses (France). Commissariat A L`Energie Atomique; Lorenzetto, P. [NET Team, Garching (Germany); Noda, K. [Japan Atomic Energy Research Inst., Takai, Ibaraki, (Japan); Roux, N. [CEA Centre d`Etudes et de Recherches Les Materiaux (France). Commissariat a L`Energie Atomique

    1993-11-01

    The databases for solid breeder ceramics (Li{sub 2},O, Li{sub 4}SiO{sub 4}, Li{sub 2}ZrO{sub 3} and LiAlO{sub 2}) and beryllium multiplier material are critically reviewed and evaluated. Emphasis is placed on physical, thermal, mechanical, chemical stability/compatibility, tritium, and radiation stability properties which are needed to assess the performance of these materials in a fusion reactor environment. Correlations are selected for design analysis and compared to the database. Areas for future research and development in blanket materials technology are highlighted and prioritized.

  10. Liquid Metal Fast Breeder Reactors: a bibliography

    Energy Technology Data Exchange (ETDEWEB)

    Raleigh, H.D. (ed.)

    1980-11-01

    This bibliography includes 5465 selected citations on LMFBR development. The citations were compiled from the DOE Energy Data Base covering the period January 1978 (EDB File No. 78R1087) through August 1980 (EDB File No. 80C79142). The references are to reports from the Department of Energy and its contractors, reports from other government or private organizations, and journal articles, books, conference papers, and monographs from US originators. Report citations are arranged alphanumerically by report number; nonreport literature citations are arranged chronologically. Corporate, Personal Author, Subject, and Report Number Indexes are provided in Volume 2.

  11. Liquid Metal Fast Breeder Reactors: a bibliography

    Energy Technology Data Exchange (ETDEWEB)

    Raleigh, H.D. (ed.)

    1980-11-01

    This bibliogralphy includes 5465 selected citations on LMFBR development. The citations were compiled from the DOE Energy Data Base covering the period January 1978 (EDB File No. 78R1087) through August 1980 (EDB File No. 80C79142). The references are to reports from the Department of Energy and its contractors, reports from other government or private organizations, and journal articles, books, conference papers, and monographs from US originators. Report citations are arranged alphanumerically by report number; nonreport literature citations are arranged chronologically. Corporate, Personal Author, Subject, and Report Number Indexes are provided in Volume 2.

  12. Reactor safety study methodology applications program: Calvert Cliffs No. 2 power plant

    Energy Technology Data Exchange (ETDEWEB)

    Hatch, S.W.; Kolb, G.J.; Cybulskis, P.; Wooton, R.O.

    1982-05-01

    This volume represents the results of the analysis of the Calvert Cliffs Unit 2 Nuclear Power Plant which was performed as part of the Reactor Safety Study Methodology Applications Program (RSSMAP). The RSSMAP was conducted to apply the methodology developed in the Reactor Safety Study (RSS) to an additional group of plants with the following objectives in mind: (1) identification of the risk dominating accident sequences for a broader group of reactor designs; (2) comparison of these accident sequences with those identified in the RSS; and (3) based on this comparison, identification of design differences which have a significant impact on risk.

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

    Energy Technology Data Exchange (ETDEWEB)

    1993-05-13

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

  14. New generation medium power nuclear station with VPBER-600 passive safety reactor plant

    Energy Technology Data Exchange (ETDEWEB)

    Mitenkov, F.M.; Antonovsky, G.M.; Panov, Yu.K.; Runov, B.I. [OKB Mech. Eng., Nizhny Novgorod (Russian Federation)

    1997-10-01

    Design principles and the main technical characteristics of a new generation medium size atomic station with passive safety VPBER-600 reactor plant (RP) are considered in the report. RP safety concept, design of the integral reactor and of the reactor unit, description of the principal diagram and plant safety system, layout decisions on reactor building and the concept of architectural-arrangement decisions on the site are given. The results of safety analysis show that the atomic station with VPBER-600 RP is an environmentally clean source of power supply with a minimal radiation impact on the environment. Information is given on the soundness of design solutions and data are presented concerning the respective cost and possible time schedule for the plant construction. (orig.) 4 refs.

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

    Energy Technology Data Exchange (ETDEWEB)

    Smith, R.R.

    1986-01-01

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

  16. Physical aspects of the Canadian generation IV supercritical water-cooled pressure tube reactor plant design

    Energy Technology Data Exchange (ETDEWEB)

    Gaudet, M.; Yetisir, M.; Haque, Z. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2014-07-01

    The form of the containment building is a function of the requirements imposed by various systems. In order to provide sufficient driving force for naturally-circulated emergency cooling systems, as well as providing a gravity-driven core flooding pool function, the Canadian SCWR reactor design relies on elevation differences between the reactor and the safety systems. These elevation differences, the required cooling pool volumes and the optimum layout of safety-related piping are major factors influencing the plant design. As a defence-in-depth, the containment building and safety systems also provide successive barriers to the unplanned release of radioactive materials, while providing a path for heat flow to the ultimate heat sink, the atmosphere. Access to the reactor for refuelling is from the top of the reactor, with water used as shielding during the refuelling operations. The accessibility to the reactor and protection of the environment are additional factors influencing the plant design. This paper describes the physical implementation of the major systems of the Canadian SCWR within the reactor building, and the position of major plant services relative to the reactor building. (author)

  17. Mutual Partnership Model between Plant Breeder and Gapoktan-Ponpes (A Traditional Farmer Group in Pre-releasing New Hybrid and Composite Varieties of Sweet Corn

    Directory of Open Access Journals (Sweden)

    Arifin Noor Sugiharto

    2014-12-01

    Full Text Available We have been genotyping the Filius (F6 sweet corn lines that currently ready to be evaluated by multi-location and heterosis test.Normally, in the step of breeding program, releasing varieties followed by seed production and distribution would be done after such tests that costly needed and time consuming. Traditional institution of farmer group called in Indonesia as “Gapoktan” and the Religious teaching stage called as “Ponpes” have been known as non-formal institution which have high potency and unique networking access to surrounding public citizen and other informal institutions. As so did, they can be as a producer and virtual guarantor or good marketing agent of many agricultural products as well. Therefore, a mutual partnership model established between breeder and them is very important and to be a prospective idea to short cut breeding process and hybrid seed production. Scientific accountability and efficiency of whole breeding – seed production would be kept into main consideration. The concept was “Gapoktan” and “Ponpes” firstly would be trained to adopt technology for crossing to set F1 Hybrid seed, then the results would be evaluated and compared with those done by breeder to determine whether they eligible or not. If they did so, partnership would be mutully continued and they would be encouraged to distribute F1 seeds to other “gapoktan” or public citizens. Three nominated Gapoktan and Ponpes from separated regencies, i.e., Pasuruan, Malang and Batu were determined as the model. The results showed that F1 seed production variable as indicator achievement of them were not significantly different from those of breeder in any place. Even though in between the places they did differ in several vegetative and productive factors each other. The differences were probably caused by ecological factors, not by their capability in agriculture practicing. Based on standard of common farming of corn or sweet corn in

  18. Flow rate analysis of wastewater inside reactor tanks on tofu wastewater treatment plant

    Science.gov (United States)

    Mamat; Sintawardani, N.; Astuti, J. T.; Nilawati, D.; Wulan, D. R.; Muchlis; Sriwuryandari, L.; Sembiring, T.; Jern, N. W.

    2017-03-01

    The research aimed to analyse the flow rate of the wastewater inside reactor tanks which were placed a number of bamboo cutting. The resistance of wastewater flow inside reactor tanks might not be occurred and produce biogas fuel optimally. Wastewater from eleven tofu factories was treated by multi-stages anaerobic process to reduce its organic pollutant and produce biogas. Biogas plant has six reactor tanks of which its capacity for waste water and gas dome was 18 m3 and 4.5 m3, respectively. Wastewater was pumped from collecting ponds to reactors by either serial or parallel way. Maximum pump capacity, head, and electrical motor power was 5m3/h, 50m, and 0.75HP, consecutively. Maximum pressure of biogas inside the reactor tanks was 55 mbar higher than atmosphere pressure. A number of 1,400 pieces of cutting bamboo at 50-60 mm diameter and 100 mm length were used as bacteria growth media inside each reactor tank, covering around 14,287 m2 bamboo area, and cross section area of inner reactor was 4,9 m2. In each reactor, a 6 inches PVC pipe was installed vertically as channel. When channels inside reactor were opened, flow rate of wastewater was 6x10-1 L.sec-1. Contrary, when channels were closed on the upper part, wastewater flow inside the first reactor affected and increased gas dome. Initially, wastewater flowed into each reactor by a gravity mode with head difference between the second and third reactor was 15x10-2m. However, head loss at the second reactor was equal to the third reactor by 8,422 x 10-4m. As result, wastewater flow at the second and third reactors were stagnant. To overcome the problem pump in each reactor should be installed in serial mode. In order to reach the output from the first reactor and the others would be equal, and biogas space was not filled by wastewater, therefore biogas production will be optimum.

  19. Preliminary neutronics design and analysis of helium cooled solid breeder blanket for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Lv, Zhongliang; Chen, Hongli, E-mail: hlchen1@ustc.edu.cn; Chen, Chong; Li, Min; Zhou, Guangming

    2015-06-15

    Highlights: • Neutronics design of a helium cooled solid breeder blanket for CFETR was presented. • The breeding zones parallel to FW and perpendicular to FW were optimized. • A series of neutronics analyses for the proposed blanket were shown. - Abstract: Chinese Fusion Engineering Test Reactor (CFETR) is a test tokamak reactor being designed in China to bridge the gap between ITER and future fusion power plant. Tritium self-sufficiency is one of the most important issues for CFETR and the tritium breeding ratio (TBR) is recommended not less than 1.2. As one of the candidates, a helium cooled solid breeder blanket for CFETR superconducting tokamak option was proposed. In the concept, radial arranged U-shaped breeding zones are adopted for higher TBR and simpler structure. In this work, three-dimensional neutronics design and analysis of the blanket were performed using the Monte Carlo N-Particle transport code MCNP with IAEA data library FENDL-2.1. Tritium breeding capability of the proposed blanket was assessed and the breeding zones parallel to first wall (FW) and perpendicular to FW were optimized. Meanwhile, the nuclear heating analysis and shielding performance were also presented for later thermal and structural analysis. The results showed that the blanket could well meet the tritium self-sufficiency target and the neutron shield could satisfy the design requirements.

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

    Energy Technology Data Exchange (ETDEWEB)

    M. J. Russell

    2006-06-01

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

  1. Development of a continuous rotating cone reactor pilot plant for the pyrolysis of polyethene and polypropene.

    NARCIS (Netherlands)

    Westerhout, R.W.J.; Westerhout, R.W.J.; Waanders, J.; Kuipers, J.A.M.; van Swaaij, Willibrordus Petrus Maria

    1998-01-01

    A pilot plant for the high-temperature pyrolysis of polymers to recycle plastic waste to valuable products was constructed based on the rotating cone reactor (RCR) technology. The RCR used in this pilot plant, termed the continuous RCR ([C]RCR) was an improved version of the bench-scale RCR ([B]RCR)

  2. Study on reactor vessel replacement (RVR) for 1100 MW class BWR plants in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Mizutani, J.; Kawamura, S. [Tokyo Electric Power Co., Inc. (Japan); Aoki, M. [Hitachi Ltd., Ibaraki (Japan); Mori, T. [Toshiba Corp., Yokihama (Japan)

    2001-07-01

    Plant Life Management (PLM) is being studied in Japan, and reactor vessel replacement (RVR) is being considered as one option. Since reactor internals, except for reusable parts, and the reactor pressure vessel (RPV) are replaced, the RVR provides an effective technology for extending the service life of nuclear power plants substantially. At ICONE 7, we reported on the technical viability of the RVR for BWR4-type 800 MWe class plants. This time, we rationalized the RVR method through a study for BWR5-type 1100 MWe class plants to reduce the RVR duration and evaluated the technical viability and the economic efficiency of the method. In addition, we discuss how to dispose of the RPV to complete a scenario of the process from the RVR to its final disposal. (author)

  3. Reactor safety study. An assessment of accident risks in U. S. commercial nuclear power plants. Appendices VII, VIII, IX, and X. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    1975-10-01

    Information is presented concerning the release of radioactivity in reactor accidents; physical processes in reactor meltdown accidents; safety design rationale for nuclear power plants; and design adequacy.

  4. PREMOR: a point reactor exposure model computer code for survey analysis of power plant performance

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.

    1979-10-01

    The PREMOR computer code was written to exploit a simple, two-group point nuclear reactor power plant model for survey analysis. Up to thirteen actinides, fourteen fission products, and one lumped absorber nuclide density are followed over a reactor history. Successive feed batches are accounted for with provision for from one to twenty batches resident. The effect of exposure of each of the batches to the same neutron flux is determined.

  5. Studies on environment safety and application of advanced reactor for inland nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Wei, L.; Jie, L. [Nuclear Power Institute of China, Southwestern Nuclear Power Plants Prepatory Office (China)

    2014-07-01

    To study environment safety assessment of inland nuclear power plants (NPPs), the impact of environment safety under the normal operation was researched and the environment risk of serious accidents was analyzed. Moreover, the requirements and relevant provisions of site selection between international nuclear power plant and China's are comparatively studied. The conclusion was that the environment safety assessment of inland and coastal nuclear power plant have no essential difference; the advanced reactor can meet with high criteria of environment safety of inland nuclear power plants. In this way, China is safe and feasible to develop inland nuclear power plant. China's inland nuclear power plants will be as big market for advanced reactor. (author)

  6. 76 FR 5220 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Future Plant...

    Science.gov (United States)

    2011-01-28

    ... small modular reactor applications. The Subcommittee will hear presentations by and hold discussions... COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Future Plant... rescheduling would result in a major inconvenience. Dated: January 24, 2011. Antonio Dias, Chief, Reactor...

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

    Science.gov (United States)

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

    2009-01-01

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

  8. Selection of power plant elements for future reactor space electric power systems

    Energy Technology Data Exchange (ETDEWEB)

    Buden, D.; Bennett, G.A.; Copper, K.

    1979-09-01

    Various types of reactor designs, electric power conversion equipment, and reject-heat systems to be used in nuclear reactor power plants for future space missions were studied. The designs included gas-cooled, liquid-cooled, and heat-pipe reactors. For the power converters, passive types such as thermoelectric and thermionic converters and dynamic types such as Brayton, potassium Rankine, and Stirling cycles were considered. For the radiators, heat pipes for transfer and radiating surface, pumped fluid for heat transfer with fins as the radiating surface, and pumped fluid for heat transfer with heat pipes as the radiating surface were considered. After careful consideration of weights, sizes, reliabilities, safety, and development cost and time, a heat-pipe reactor design, thermoelectric converters, and a heat-pipe radiator for an experimental program were selected.

  9. Boiling water reactor in a prestressed reinforced concrete vessel for an atomic central heating-and-power plant

    Energy Technology Data Exchange (ETDEWEB)

    Tokarev, Yu.I.; Sokolov, I.N.; Skvortsov, S.A.; Sidorov, A.M.; Krauze, L.V.

    1978-04-01

    The possibility of using a boiling water reactor in a prestressed reinforced concrete vessel for an atomic central heating-and-power plant (CHPP) was considered, with design features of the reactor intended for a two-purpose plant. A prestressed reinforced concrete vessel and integral arrangement of the primary circuit ensured reliability of the atomic CHPP using various CHPP flowsheets.

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

    Science.gov (United States)

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

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

  11. Pebble Bed Reactor: core physics and fuel cycle analysis

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.; Worley, B.A.

    1979-10-01

    The Pebble Bed Reactor is a gas-cooled, graphite-moderated high-temperature reactor that is continuously fueled with small spherical fuel elements. The projected performance was studied over a broad range of reactor applicability. Calculations were done for a burner on a throwaway cycle, a converter with recycle, a prebreeder and breeder. The thorium fuel cycle was considered using low, medium (denatured), and highly enriched uranium. The base calculations were carried out for electrical energy generation in a 1200 MW/sub e/ plant. A steady-state, continuous-fueling model was developed and one- and two-dimensional calculations were used to characterize performance. Treating a single point in time effects considerable savings in computer time as opposed to following a long reactor history, permitting evaluation of reactor performance over a broad range of design parameters and operating modes.

  12. Fast reactor programme in India

    Indian Academy of Sciences (India)

    2015-09-04

    Sep 4, 2015 ... criteria; passive shutdown and decay heat removal systems; fast breeder reactors in India. PACS No. 28.41.−i. 1. ... water reactors, mainly pressurized heavy water reactors (PHWRs) to extract ∼10 GWe capacity for ..... commissioning phase and most of the supporting systems have been commissioned and.

  13. Fast reactor programme in India

    Indian Academy of Sciences (India)

    2015-09-04

    Sep 4, 2015 ... Home; Journals; Pramana – Journal of Physics; Volume 85; Issue 3. Fast reactor programme in India. P Chellapandi P R ... Keywords. Sodium fast reactor; design challenges; construction challenges; emerging safety criteria; passive shutdown and decay heat removal systems; fast breeder reactors in India.

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

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover

    2009-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-03-21

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

  16. Conceptual design of solid breeder blanket system cooled by supercritical water

    Energy Technology Data Exchange (ETDEWEB)

    Enoeda, Mikio; Akiba, Masato [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Ohara, Yoshihiro [Japan Atomic Energy Research Inst., Takasaki, Gunma (Japan). Takasaki Radiation Chemistry Research Establishment] [and others

    2001-12-01

    This report is a summary of the design works, which was discussed in the design workshop held in 2000 for the demonstration (DEMO) blanket aimed to strengthen the commercial competitiveness and technical feasibility simultaneously. The discussion of the Fusion Council in 1999 updated the assessment of the mission of DEMO blanket. Updated mission of the DEMO blanket is to be the prototype of the commercially competitive power plant. The DEMO blanket must supply the feasibility and experience of the total design of the power plant and the materials. From such standing point, the conceptual design study was performed to determine the updated strategy and goal of the R and D of the DEMO blanket which applies the supercritical water cooling proposed in A-SSTR, taking into account the recent progress of the plasma research and reactor engineering technology. The DEMO blanket applies the solid breeder materials and supercritical water cooling. The product tritium is purged out by helium gas stream in the breeder region. In the breeder region, the pebble bed concept was applied to withstand instable cracking of the breeder and multiplier materials in high neutron irradiation and high temperature operation. Inlet temperature of the coolant is planned to be 280degC and final outlet temperature is 510degC to obtain high energy conversion efficiency up to 43%. Reduced activation ferritic steel, F82H and ODS ferritic steel were selected as the structural material. Lithium ceramics, Li{sub 2}TiO{sub 3} or Li{sub 2}O were selected as the breeder materials. Beryllium or its inter-metallic compound Be12Ti was selected as the neutron multiplier materials. Basic module structure was selected as the box type structure which enables the remote handling replacement of the module from in-vessel access. Dimension of the box is limited to 2 m x 2 m, or smaller, due to the dimension of the replacement port. In the supercritical water cooling, the high coolant temperature is the merit for

  17. Recent advances in AFB biomass gasification pilot plant with catalytic reactors in a downstream slip flow

    Energy Technology Data Exchange (ETDEWEB)

    Aznar, M.P.; Gil, J.; Martin, J.A.; Frances, E.; Olivares, A.; Caballero, M.A.; Perez, P. [Saragossa Univ. (Spain). Dept. of Chemistry and Environment; Corella, J. [Madrid Univ. (Spain)

    1996-12-31

    A new 3rd generation pilot plant is being used for hot catalytic raw gas cleaning. It is based on a 15 cm. i.d. fluidized bed with biomass throughputs of 400-650 kg/h.m{sup 2}. Gasification is performed using mixtures of steam and oxygen. The produced gas is passed in a slip flow by two reactors in series containing a calcined dolomite and a commercial reforming catalyst. Tars are periodically sampled and analysed after the three reactors. Tar conversions of 99.99 % and a 300 % increase of the hydrogen content in the gas are obtained. (author) (2 refs.)

  18. VPBER-600 passive safety reactor plant for the medium size NPP of new generation

    Energy Technology Data Exchange (ETDEWEB)

    Antonovsky, G.M.; Kuznetsov, L.E.; Novikov, V.V.; Panov, Yu.K.; Runov, B.I.

    1993-12-31

    VPBER-600 (RP) a vessel type PWR for the enhanced safety nuclear power plant of new generation as compared with the existing ones is discussed in the paper. The enhanced safety level is achieved through adoption of the following design decisions: primary circuit leak tight arrangement; integral reactor type; guard vessel for reactor and primary circuit systems accomodation; passive safety systems that do not require external power and control actions for actuation, or initiated from self-actuated devices. VPBER-600 RP safety level allows to use the plant both in NPP and in co-generation plant located near heat consumers. Scheme and design decisions adopted in the RP are described and main technical data are given in the paper. The focus is on the design decisions that provide realization of the formulated safety concept.

  19. Sodium Advanced Fast Reactor (SAFR). Volume II. Summary plant description

    Energy Technology Data Exchange (ETDEWEB)

    None

    1985-09-01

    The balance of plant (BOP) includes all buildings and structures, all steam and water facilities, all power generation and transmission systems, and all auxiliary systems which do not contain or handle sodium. In addition, the BOP includes all site facilities such as roads, fences, etc.

  20. An organic profile of a pressurised water reactor secondary plant

    Energy Technology Data Exchange (ETDEWEB)

    Eeden, Nestor van; Stwayi, Mandisibuntu; Gericke, Gerhard [Eskom Holdings SOC Ltd., Western Cape (South Africa). Koeberg Power Station

    2012-07-15

    Make-up water addition to the steam/water cycle at Koeberg Nuclear Power Station usually results in a corresponding increase of the chloride concentration in the steam generator blowdown system. During plant transients, when higher than normal make-up is required to the secondary plant, the concentration of chloride occasionally exceeds the limiting value for the station chemistry performance indicator. Irrespective of this, the demineralised water make-up supply tanks, which are routinely analysed for chloride, are within all recognised acceptable standards for secondary water make-up and therefore these tanks do not initially appear to be the source of chloride contamination. Water treatment at the plant relies essentially on ion exchange, which has been proven to be very effective in removing inorganic ionic species such as chloride. Organic compounds are less effectively removed by ion exchange and may pass through the treatment system, and these organics can reside undetected in the make-up water tanks. Historically, the elevated chloride concentration following high system make-up has been attributed to chlorinated organic compounds known as trihalomethanes being present in the make-up water tanks, but no rigorous study had been undertaken. As it has been assumed that the majority of chloride in the secondary system originates from the make-up water organic impurities, it was considered important to confirm this by compiling an organic profile of the secondary plant. The use of organic additives was also taken into account in the profile. This work has confirmed the contribution from trihalomethanes and has also found that other organochlorides contribute even more significantly to the overall chloride inventory of the secondary plant. (orig.)

  1. Environmental impacts of nonfusion power systems. [Data on environmental effects of all power sources that may be competitive with fusion reactor power plants

    Energy Technology Data Exchange (ETDEWEB)

    Brouns, R.J.

    1976-09-01

    Data were collected on the environmental effects of power sources that may be competitive with future fusion reactor power plants. Data are included on nuclear power plants using HTGR, LMBR, GCFR, LMFBR, and molten salt reactors; fossil-fuel electric power plants; geothermal power plants; solar energy power plants, including satellite-based solar systems; wind energy power plants; ocean thermal gradient power plants; tidal energy power plants; and power plants using hydrogen and other synthetic fuels as energy sources.

  2. Optimization of reload of nuclear power plants using ACO together with the GENES reactor physics code

    Energy Technology Data Exchange (ETDEWEB)

    Lima, Alan M.M. de; Freire, Fernando S.; Nicolau, Andressa S.; Schirru, Roberto, E-mail: alan@lmp.ufrj.br, E-mail: andressa@lmp.ufrj.br, E-mail: schirru@lmp.ufrj.br, E-mail: ffreire@eletronuclear.gov.br [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil); Eletrobras Termonuclear S.A. (ELETRONUCLEAR), Rio de Janeiro, RJ (Brazil)

    2017-11-01

    The Nuclear reload of a Pressurized Water Reactor (PWR) occurs whenever the burning of the fuel elements can no longer maintain the criticality of the reactor, that is, it cannot maintain the Nuclear power plant operates within its nominal power. Nuclear reactor reload optimization problem consists of finding a loading pattern of fuel assemblies in the reactor core in order to minimize the cost/benefit ratio, trying to obtain maximum power generation with a minimum of cost, since in all reloads an average of one third of the new fuel elements are purchased. This loading pattern must also satisfy constraints of symmetry and security. In practice, it consists of the placing 121 fuel elements in 121 core positions, in the case of the Angra 1 Brazilian Nuclear Power Plant (NPP), making this new arrangement provide the best cost/benefit ratio. It is an extremely complex problem, since it has around 1% of great places. A core of 121 fuel elements has approximately 10{sup 13} combinations and 10{sup 11} great locations. With this number of possible combinations it is impossible to test all, in order to choose the best. In this work a system called ACO-GENES is proposed in order to optimization the Nuclear Reactor Reload Problem. ACO is successfully used in combination problems, and it is expected that ACO-GENES will show a robust optimization system, since in addition to optimizing ACO, it allows important prior knowledge such as K infinite, burn, etc. After optimization by ACO-GENES, the best results will be validated by a licensed reactor physics code and will be compared with the actual results of the cycle. (author)

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

    Directory of Open Access Journals (Sweden)

    Fic Adam

    2015-03-01

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

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

    Science.gov (United States)

    Fic, Adam; Składzień, Jan; Gabriel, Michał

    2015-03-01

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

  5. Analysis of Possible Application of High-Temperature Nuclear Reactors to Contemporary Large-Output Steam Power Plants on Ships

    Directory of Open Access Journals (Sweden)

    Kowalczyk T.

    2016-04-01

    Full Text Available This paper is aimed at analysis of possible application of helium to cooling high-temperature nuclear reactor to be used for generating steam in contemporary ship steam-turbine power plants of a large output with taking into account in particular variable operational parameters. In the first part of the paper types of contemporary ship power plants are presented. Features of today applied PWR reactors and proposed HTR reactors are discussed. Next, issues of load variability of the ship nuclear power plants, features of the proposed thermal cycles and results of their thermodynamic calculations in variable operational conditions, are presented.

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

    Energy Technology Data Exchange (ETDEWEB)

    Wood, RT

    2004-09-27

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

  7. Selection of Technical Reactor Equipment for Modular, Continuous Small-Scale Plants

    OpenAIRE

    Krasberg, Nicolai; Hohmann, Lukas; Bieringer, Thomas; Bramsiepe, Christian; Kockmann, Norbert

    2014-01-01

    Fast process development, flexible production and the utilization of advanced process conditions are the main goals of modular and continuous small-scale plants (MCSPs). A configurable layout of the modules and the use of predefined equipment enable a quick and reliable conceptual process development and scale-up of continuous processes. Therefore, a computer-assisted selection methodology was developed and is presented, which allows the quick selection of plug flow reactor equipment for homo...

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

    Science.gov (United States)

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

    2016-09-01

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

  9. EBR-II Reactor Physics Benchmark Evaluation Report

    Energy Technology Data Exchange (ETDEWEB)

    Pope, Chad L. [Idaho State Univ., Pocatello, ID (United States); Lum, Edward S [Idaho State Univ., Pocatello, ID (United States); Stewart, Ryan [Idaho State Univ., Pocatello, ID (United States); Byambadorj, Bilguun [Idaho State Univ., Pocatello, ID (United States); Beaulieu, Quinton [Idaho State Univ., Pocatello, ID (United States)

    2017-12-28

    This report provides a reactor physics benchmark evaluation with associated uncertainty quantification for the critical configuration of the April 1986 Experimental Breeder Reactor II Run 138B core configuration.

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

    Energy Technology Data Exchange (ETDEWEB)

    J. K. Wright; R. N. Wright

    2008-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-06-01

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

  12. The Westinghouse AP1000 plant design: a generation III+ reactor with unique proven passive safety technology

    Energy Technology Data Exchange (ETDEWEB)

    Demetri, K. J.; Leipner, C. I.; Marshall, M. L., E-mail: demetrkj@westinghouse.com [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

    2015-09-15

    The AP1000 plant is an 1100-M We pressurized water reactor with passive safety features and extensive plant simplifications and standardization that simplify construction, operation, maintenance, safety, and cost. The AP1000 plant is based on proven pressurized water reactor (PWR) technology, with an emphasis on safety features that rely solely on natural forces. These passive safety features are combined with simple, active, defense-in-depth systems used during normal plant operations which also provide the first level of defense against more probable events. This paper focuses on specific safety and licensing topics: the AP1000 plant robustness to be prepared for extreme events that may lead to catastrophic loss of infrastructure, such as the Fukushima Dai-ichi event, and the AP1000 plant compliance with the safety objectives for new plants. The first deployment of the AP1000 plant formally began in July 2007 when Westinghouse Electric Company and its consortium partner, the Shaw Group, signed contracts for four AP1000 units on coastal sites of Sanmen and Haiyang, China. Both sites have the planned ability to accommodate at least six AP1000 units; construction is largely concurrent for all four units. Additionally, the United States (U.S.) Nuclear Regulatory Commission (NRC) issued combined licenses (COLs) to allow Southern Nuclear Operating Company (SNC) and South Carolina Electric and Gas Company (SCE and G) to construct and operate AP1000 plants. Within this paper, the various factors that contribute to an unparalleled level of design, construction, delivery, and licensing certainty for any new AP1000 plant projects are described. These include: 1) How the AP1000 plant design development and reviews undertaken in the United States, China and Europe increase licensing certainty. 2) How the AP1000 passive plant robustness against extreme events that result in large loss of infrastructure further contributes to the licensing certainty in a post

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

    Energy Technology Data Exchange (ETDEWEB)

    J. K. Wright; R. N. Wright

    2010-07-01

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

  14. Design Study of Modular Nuclear Power Plant with Small Long Life Gas Cooled Fast Reactors Utilizing MOX Fuel

    Science.gov (United States)

    Ilham, Muhammad; Su'ud, Zaki

    2017-01-01

    Growing energy needed due to increasing of the world’s population encourages development of technology and science of nuclear power plant in its safety and security. In this research, it will be explained about design study of modular fast reactor with helium gas cooling (GCFR) small long life reactor, which can be operated over 20 years. It had been conducted about neutronic design GCFR with Mixed Oxide (UO2-PuO2) fuel in range of 100-200 MWth NPPs of power and 50-60% of fuel fraction variation with cylindrical pin cell and cylindrical balance of reactor core geometry. Calculation method used SRAC-CITATION code. The obtained results are the effective multiplication factor and density value of core reactor power (with geometry optimalization) to obtain optimum design core reactor power, whereas the obtained of optimum core reactor power is 200 MWth with 55% of fuel fraction and 9-13% of percentages.

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1993-09-15

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

  16. Reactor safety study. An assessment of accident risks in U. S. commercial nuclear power plants. Executive summary: main report. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    1975-10-01

    Information is presented concerning the objectives and organization of the reactor safety study; the basic concepts of risk; the nature of nuclear power plant accidents; risk assessment methodology; reactor accident risk; and comparison of nuclear risks to other societal risks.

  17. Hydrodynamics of Pilot Plant Scale Airlift Reactor in Presence of Alcohols

    Directory of Open Access Journals (Sweden)

    Nandi Somnath

    2015-09-01

    Full Text Available The overall gas hold-up of a pilot plant scale internal loop airlift reactor was studied in presence of different alcohols in varied concentration. It has been observed that these simple alcohols can enhance overall gas hold-up of the airlift assembly possibly due to formation of protective thin layer over smaller gas bubbles thereby reducing chance of their coalescence. As the alcohols can be used as food source for the micro organisms present in the system, this green and environment friendly process have potential to replace usage of hazardous surfactants often used for enhancing overall hold-up in order to obtain desired mass transfer characteristics. An empirical relationship encompassing overall gas hold-up of the reactor as a function of superficial gas velocity and alcohol concentration is also developed.

  18. A full scale worm reactor for efficient sludge reduction by predation in a wastewater treatment plant.

    Science.gov (United States)

    Tamis, J; van Schouwenburg, G; Kleerebezem, R; van Loosdrecht, M C M

    2011-11-15

    Sludge predation can be an effective solution to reduce sludge production at a wastewater treatment plant. Oligochaete worms are the natural consumers of biomass in benthic layers in ecosystems. In this study the results of secondary sludge degradation by the aquatic Oligochaete worm Aulophorus furcatus in a 125 m(3) reactor and further sludge conversion in an anaerobic tank are presented. The system was operated over a period of 4 years at WWTP Wolvega, the Netherlands and was fed with secondary sludge from a low loaded activated sludge process. It was possible to maintain a stable and active population of the aquatic worm species A. furcatus during the full period. Under optimal conditions a sludge conversion of 150-200 kg TSS/d or 1.2-1.6 kg TSS/m(3)/d was established in the worm reactor. The worms grew as a biofilm on carrier material in the reactor. The surface specific conversion rate reached 140-180 g TSS/m(2)d and the worm biomass specific conversion rate was 0.5-1 g TSS sludge/g dry weight worms per day. The sludge reduction under optimal conditions in the worm reactor was 30-40%. The degradation by worms was an order of magnitude larger than the endogenous conversion rate of the secondary sludge. Effluent sludge from the worm reactor was stored in an anaerobic tank where methanogenic processes became apparent. It appeared that besides reducing the sludge amount, the worms' activity increased anaerobic digestibility, allowing for future optimisation of the total system by maximising sludge reduction and methane formation. In the whole system it was possible to reduce the amount of sludge by at least 65% on TSS basis. This is a much better total conversion than reported for anaerobic biodegradability of secondary sludge of 20-30% efficiency in terms of TSS reduction. Copyright © 2011 Elsevier Ltd. All rights reserved.

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

    Energy Technology Data Exchange (ETDEWEB)

    Schlereth, J.R.; Pennington, D.

    1996-12-01

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

  20. ROLE OF PASSIVE SAFETY FEATURES IN PREVENTION AND MITIGATION OF SEVERE PLANT CONDITIONS IN INDIAN ADVANCED HEAVY WATER REACTOR

    Directory of Open Access Journals (Sweden)

    VIKAS JAIN

    2013-10-01

    Full Text Available Pressing demands of economic competitiveness, the need for large-scale deployment, minimizing the need of human intervention, and experience from the past events and incidents at operating reactors have guided the evolution and innovations in reactor technologies. Indian innovative reactor ‘AHWR’ is a pressure-tube type natural circulation based boiling water reactor that is designed to meet such requirements, which essentially reflect the needs of next generation reactors. The reactor employs various passive features to prevent and mitigate accidental conditions, like a slightly negative void reactivity coefficient, passive poison injection to scram the reactor in event of failure of the wired shutdown systems, a large elevated pool of water as a heat sink inside the containment, passive decay heat removal based on natural circulation and passive valves, passive ECC injection, etc. It is designed to meet the fundamental safety requirements of safe shutdown, safe decay heat removal and confinement of activity with no impact in public domain, and hence, no need for emergency planning under all conceivable scenarios. This paper examines the role of the various passive safety systems in prevention and mitigation of severe plant conditions that may arise in event of multiple failures. For the purpose of demonstration of the effectiveness of its passive features, postulated scenarios on the lines of three major severe accidents in the history of nuclear power reactors are considered, namely; the Three Mile Island (TMI, Chernobyl and Fukushima accidents. Severe plant conditions along the lines of these scenarios are postulated to the extent conceivable in the reactor under consideration and analyzed using best estimate system thermal-hydraulics code RELAP5/Mod3.2. It is found that the various passive systems incorporated enable the reactor to tolerate the postulated accident conditions without causing severe plant conditions and core degradation.

  1. Synthesis of Model Based Robust Stabilizing Reactor Power Controller for Nuclear Power Plant

    Directory of Open Access Journals (Sweden)

    Arshad Habib Malik

    2011-04-01

    Full Text Available In this paper, a nominal SISO (Single Input Single Output model of PHWR (Pressurized Heavy Water Reactor type nuclear power plant is developed based on normal moderator pump-up rate capturing the moderator level dynamics using system identification technique. As the plant model is not exact, therefore additive and multiplicative uncertainty modeling is required. A robust perturbed plant model is derived based on worst case model capturing slowest moderator pump-up rate dynamics and moderator control valve opening delay. Both nominal and worst case models of PHWR-type nuclear power plant have ARX (An Autoregressive Exogenous structures and the parameters of both models are estimated using recursive LMS (Least Mean Square optimization algorithm. Nominal and worst case discrete plant models are transformed into frequency domain for robust controller design purpose. The closed loop system is configured into two port model form and H? robust controller is synthesized. The H?controller is designed based on singular value loop shaping and desired magnitude of control input. The selection of desired disturbance attenuation factor and size of the largest anticipated multiplicative plant perturbation for loop shaping of H? robust controller form a constrained multi-objective optimization problem. The performance and robustness of the proposed controller is tested under transient condition of a nuclear power plant in Pakistan and found satisfactory.

  2. Sequencing Batch Reactor pilot plant in waste water treatment plants; Impiego di un impianto pilota del tipo Seguencing Batch Reactor (SBR) come strumento di gestione di impianti di depurazione in aree sensibili : Indagine preliminare

    Energy Technology Data Exchange (ETDEWEB)

    Musacco, Alessandro; Beccari, Mario [Rome, Univ. La Sapienza (Italy). Dip. di Chimica; Cecchi, Franco [Aquila, Univ. (Italy). Dip. di Chimica, Ingegneria Chimica e Materiali]|[ASPIV, Venice (Italy)

    1997-07-01

    The flexibility of Sequencing Batch Reactors is ideally suited to provide useful guidelines for reliable operation of large waste water treatment plants such as the plant at Fusina (Venice). This paper shows the preliminary results of experimental investigations carried out in a pilot scale Sequencing Batch Reactor plant. The main purpose has been to verify the effect of the addition of readily biodegradable COD from the acidogenic fermentation of the organic fraction of municipal solid wastes. This has to be considered an external carbon source for denitrification. The results underline the advantages of an integration between solid waste management and wastewater treatment.

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

    Energy Technology Data Exchange (ETDEWEB)

    Hallbert, B.P.; Sebok, A.; Haugset, K. [OECD Halden Reactor Project (Norway)

    1996-03-01

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

  4. Pebble fabrication of super advanced tritium breeders using a solid solution of Li2+xTiO3+y with Li2ZrO3

    Directory of Open Access Journals (Sweden)

    Tsuyoshi Hoshino

    2016-12-01

    Full Text Available Lithium titanate with excess lithium (Li2+xTiO3+y is one of the most promising candidates among advanced tritium breeders for demonstration power plant reactors because of its good tritium release characteristics. However, the tritium breeding ratio (TBR of Li2+xTiO3+y is smaller than that of e.g., Li2O or Li8TiO6 because of its lower Li density. Therefore, new Li-containing ceramic composites with both high stability and high Li density have been developed. Thus, this study focused on the development of a solid solution with a new characteristic. The solid-solution pebbles of Li2+xTiO3+y with Li2ZrO3 (Li2+x(Ti,ZrO3+y, designated as LTZO, were fabricated by an emulsion method. The X-ray diffraction patterns of sintered LTZO pebbles are approximately the same as those of Li2+xTiO3+y pebbles, and no peaks attributable to Li2ZrO3 are observed. These results demonstrate that LTZO pebbles are not a two-phase material but rather a solid solution. Furthermore, LTZO pebbles were easily sintered under air. Thus, the LTZO solid solution is a candidate breeder material for super advanced (SA tritium breeders.

  5. Production behavior of irradiation defects in solid breeder materials

    Energy Technology Data Exchange (ETDEWEB)

    Moriyama, Hirotake; Moritani, Kimikazu [Kyoto Univ. (Japan)

    1998-03-01

    The irradiation effects in solid breeder materials are important for the performance assessment of fusion reactor blanket systems. For a clearer understanding of such effects, we have studied the production behavior of irradiation defects in some lithium ceramics by an in-situ luminescence measurement technique under ion beam irradiation. The luminescence spectra were measured at different temperatures, and the temperature-transient behaviors of luminescence intensity were also measured. The production mechanisms of irradiation defects were discussed on the basis of the observations. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Memmott, M. J.; Stansbury, C.; Taylor, C. [Westinghouse Electric Company LLC, 600 Cranberry Woods Drive, Cranberry Twp. PA 16066 (United States)

    2012-07-01

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

  7. FINAL–REPORT NO. 2: INDEPENDENT CONFIRMATORY SURVEY SUMMARY AND RESULTS FOR THE ENRICO FERMI ATOMIC POWER PLANT, UNIT 1, NEWPORT, MICHIGAN (DOCKET NO. 50 16; RFTA 10-004)

    Energy Technology Data Exchange (ETDEWEB)

    Erika Bailey

    2011-07-07

    The Enrico Fermi Atomic Power Plant, Unit 1 (Fermi 1) was a fast breeder reactor design that was cooled by sodium and operated at essentially atmospheric pressure. On May 10, 1963, the Atomic Energy Commission (AEC) granted an operating license, DPR-9, to the Power Reactor Development Company (PRDC), a consortium specifically formed to own and operate a nuclear reactor at the Fermi 1 site. The reactor was designed for a maximum capability of 430 megawatts (MW); however, the maximum reactor power with the first core loading (Core A) was 200 MW. The primary system was filled with sodium in December 1960 and criticality was achieved in August 1963.

  8. Updated reference design of a liquid metal cooled tandem mirror fusion breeder

    Energy Technology Data Exchange (ETDEWEB)

    Berwald, D.H.; Whitley, R.H.; Garner, J.K.; Gromada, R.J.; McCarville, T.J.; Moir, R.W.; Lee, J.D.; Bandini, B.R.; Fulton, F.J.; Wong, C.P.C.; Maya, I.; Hoot, C.G.; Schultz, K.R.; Miller, L.G.; Beeston, J.M.; Harris, B.L.; Westman, R.A.; Ghoniem, N.M.; Orient, G.; Wolfer, M.; DeVan, J.H.; Torterelli, P.

    1985-09-01

    Detailed studies of key techinical issues for liquid metal cooled fusion breeder (fusion-fission hybrid blankets) have been performed during the period 1983-4. Based upon the results of these studies, the 1982 reference liquid metal cooled tandem mirror fusion breeder blanket design was updated and is described. The updated reference blankets provides increased breeding and lower technological risk in comparison with the original reference blanket. In addition to the blanket design revisions, a plant concept, cost, and fuel cycle economics assessment is provided. The fusion breeder continues to promise an economical source of fissile fuel for the indefinite future.

  9. 78 FR 17945 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Future Plant...

    Science.gov (United States)

    2013-03-25

    ... based licensing framework for the Next Generation Nuclear Plant (NGNP). The Subcommittee will hear... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Future Plant...

  10. Interactive Virtual Reactor and Control Room for Education and Training at Universities and Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Satoh, Yoshinori; Li, Ye; Zhu, Xuefeng; Rizwan, Uddin [University of Illinois, Urbana (United States)

    2014-08-15

    Efficient and effective education and training of nuclear engineering students and nuclear workers are critical for the safe operation and maintenance of nuclear power plants. With an eye toward this need, we have focused on the development of 3D models of virtual labs for education, training as well as to conduct virtual experiments. These virtual labs, that are expected to supplement currently available resources, and have the potential to reduce the cost of education and training, are most easily developed on game-engine platforms. We report some recent extensions to the virtual model of the University of Illinois TRIGA reactor.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1980-02-01

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

  12. RESEARCHES ON THE DIGESTERS AND REACTORS WHICH CAN BE USED IN A FARM SCALE BIOGAS PLANT

    Directory of Open Access Journals (Sweden)

    Mariana DUMITRU

    2014-10-01

    Full Text Available In the general context of searching integrated system of renewable energy production, this paper present some researches on the reactors and the digesters, as a main part of a biogas plant at a farm scale. After we present the most used types of digesters, we also concentrated over the processes which take place into a digester, one of them being the removal of H2S from biogas (desulphurisation, which can be made by various methods, either biological or chemical, taking place inside or outside the digester. In the case of biological desulphurization outside the digester, we concentrate on the types of reactors which can be used in this case. Beside the well known types of reactors, we present the possibility of using an original self pressure membrane bioreactor. In this type of bioreactor, the metabolic activity of gas producing microorganisms, especially yeast, could obtain high pressure from gas produced in closed medium on the one hand, and separation of other products of metabolism through membrane on the other hand, using gas pressure as driving force. It is known that several strains of yeast resist on very high hydrostatic pressure heaving good activity. This fact give the possibility to use their energy for other purposes, such as producing mechanical work. Combination of both, gas pressure and alchool burning, increase the process efficiency.

  13. Beloyarsk Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-08-01

    The Beloyarsk Nuclear Power Plant (BNPP) is located in Zarechny, approximately 60 km east of Ekaterinberg along the Trans-Siberian Highway. Zarechny, a small city of approximately 30,000 residents, was built to support BNPP operations. It is a closed city to unescorted visitors. Residents must show identification for entry. BNPP is one of the first and oldest commercial nuclear power plants in Russia and began operations in 1964. As for most nuclear power plants in the Russian Federation, BNPP is operated by Rosenergoatom, which is subordinated to the Ministry of Atomic Energy of the Russian Federation (Minatom). BNPP is the site of three nuclear reactors, Units 1, 2, and 3. Units 1 and 2, which have been shut-down and defueled, were graphite moderated reactors. The units were shut-down in 1981 and 1989. Unit 3, a BN-600 reactor, is a 600 MW(electric) sodium-cooled fast breeder reactor. Unit 3 went on-line in April 1980 and produces electric power which is fed into a distribution grid and thermal power which provides heat to Zarechny. The paper also discusses the SF NIKIET, the Sverdiovsk Branch of NIKIET, Moscow, which is the research and development branch of the parent NIKEIT and is primarily a design institute responsible for reactor design. Central to its operations is a 15 megawatt IVV research reactor. The paper discusses general security and fissile material control and accountability at these two facilities.

  14. Proceedings of the NEACRP/IAEA Specialists meeting on the international comparison calculation of a large sodium-cooled fast breeder reactor at Argonne National Laboratory on February 7-9, 1978

    Energy Technology Data Exchange (ETDEWEB)

    LeSage, L.G.; McKnight, R.D.; Wade, D.C.; Freese, K.E.; Collins, P.J.

    1980-08-01

    The results of an international comparison calculation of a large (1250 MWe) LMFBR benchmark model are presented and discussed. Eight reactor configurations were calculated. Parameters included with the comparison were: eigenvalue, k/sub infinity/, neutron balance data, breeding reaction rate ratios, reactivity worths, central control rod worth, regional sodium void reactivity, core Doppler and effective delayed neutron fraction. Ten countries participated in the comparison, and sixteen solutions were contributed. The discussion focuses on the variation in parameter values, the degree of consistency among the various parameters and solutions, and the identification of unexpected results. The results are displayed and discussed both by individual participants and by groupings of participants (e.g., results from adjusted data sets versus non-adjusted data sets).

  15. Gel-sphere-pac reactor fuel fabrication and its application to a variety of fuels

    Energy Technology Data Exchange (ETDEWEB)

    Olsen, A.R.; Judkins, R.R. (comps.)

    1979-12-01

    The gel-sphere-pac fuel fabrication option was evaluated for its possible application to commercial scale fuel fabrication for 19 fuel element designs that use oxide fuel in metal clad rods. The dry gel spheres are prepared at the reprocessing plant and are then calcined, sintered, inspected, and loaded into fuel rods and packed by low-energy vibration. A fuel smear density of 83 to 88% theoretical can be obtained. All fuel fabrication process steps were defined and evaluated from fuel receiving to finished fuel element shipping. The evaluation also covers the feasibility of the process, the current status of technology, estimates of the required time and cost to develop the technology to commercial status, and the safety and licensability of commercial scale plants. The primary evaluation was for a Light-Water Reactor fuel element containing (U,Pu)O/sub 2/ fuel. The other 18 fuel element types - 3 for Light-Water Reactors, 1 for a Heavy-Water Reactor, 1 for a Gas-Cooled Fast Reactor, 7 for Liquid-Metal-Cooled Fast Breeder Reactors, and 3 pairs for Light-Water Prebreeder and Breeder Reactors - were compared with the Light-Water Reactor. The gel-sphere-pac option was found applicable to 17 of the 19 element types; the characteristics of a commercial scale plant were defined for these for making cost estimates for such plants. The evaluation clearly shows the gel-sphere-pac process to be a viable fuel fabrication option. Estimates indicate a significant potential fabrication cost advantage for the gel-sphere-pac process if a remotely operated and remotely maintained fuel fabrication plant is required.

  16. Operation and Control Simulation of a Modular High Temperature Gas Cooled Reactor Nuclear Power Plant

    Science.gov (United States)

    Li, Haipeng; Huang, Xiaojin; Zhang, Liangju

    2008-08-01

    Issues in the operation and control of the multi-modular nuclear power plant are complicated. The high temperature gas cooled reactor pebble-bed module (HTR-PM) plant with two-module will be built as a demonstration plant in China. To investigate the operation and control characteristics of the plant, a simplified dynamic model is developed and mathematically formulated based upon the fundamental conversation of mass, energy and momentum. The model is implemented in a personal computer to simulate the power increase process of the HTR-PM operation. The open loop operation with no controller is first simulated and the results show that the essential parameter steam temperature varies drastically with time, which is not allowable in the normal operation. According to the preliminary control strategy of the HTR-PM, a simple steam temperature controller is proposed. The controller is of Proportional-type with a time lag. The closed loop operation with a steam temperature controller is then implemented and the simulation results show that the steam temperature and also other parameters are all well controlled in the allowable range.

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-01

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

  19. On monitoring the tritium breeder in a lead-lithium cooled ceramic breeder (LLCB) module of the ITER

    Energy Technology Data Exchange (ETDEWEB)

    Kapyshev, V., E-mail: kapyshev@nikiet.ru [Federal State Unitary Enterprise ' Dollezhal Research and Development Institute of Power Engineering' , PO Box 788, Moscow 101000 (Russian Federation); Kartashev, I.; Kovalenko, V.; Leshukov, A.; Poliksha, V.; Rasmerov, A.; Strebkov, Yu.; Yukhnov, N.; Vladimirova, N. [Federal State Unitary Enterprise ' Dollezhal Research and Development Institute of Power Engineering' , PO Box 788, Moscow 101000 (Russian Federation)

    2012-08-15

    The operation of a tritium breeder is a most process among engineering problems of DEMO. In this study, a design for monitoring tritium-breeding in the reactor is discussed. Additionally, a system for the experimental estimation of the tritium-breeding ratio (TBR) and the tritium-breeding dynamics in a lead-lithium cooled ceramic breeder (LLCB) test module used in the ITER is proposed. The systems are based on tritium and neutron-flux measurements under the ITER plasma D-T experiments and the use of lithium ortho-silicate and lithium carbonate samples and neutron detectors. Different lithum-6 and lithium-7 isotope contents in the samples are used to measure neutron spectrum. The samples and detectors are delivered in containers to the test breeder module (TBM) on a monitor channel connecting the TBM to an operating zone of the ITER. The tritium content in the samples is measured in a laboratory by the liquid scintillation method. Pneumatic control is used to deliver the samples to the TBM and to extract the samples using the channel during plasma-operational pauses. Neutron calculation is performed to estimate the tritium content in the samples and the heat distribution in the materials of the channel under reactor irradiation. A measurement accuracy of the tritium content in the carbonate and orthosilicate samples can attain a level of 7% and 10%, respectively. The results of the channel-cooling calculation performed under the nominal operating conditions of the TBM (a plasma pulse) are presented in the paper.

  20. Reference neutron transport calculation note for Korea nuclear power plants with 3-loop PWR reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byung Cheol; Chang, Ki Oak

    1997-05-01

    Reactor pressure vessel (RPV) steels are subjected to neutron irradiation at a temperature of about 290 deg C. This radiation exposure alters the mechanical properties, leading to a shift of the brittle-to-ductile transition temperature toward higher temperatures and to a diminution of the rupture energy as determined by Charpy V-notch tests. This radiation embrittlement is one of the important aging factors of nuclear power plants. U.S. NRC recommended the basic requirements for the determination of the pressure vessel fluence by regulatory guide DG-1025 in order to reduce the uncertainty in the determination of neutron fluence calculation and measurements. The determination of the pressure vessel fluence is based on both calculations and measurements. The fluence prediction is made with a calculation and the measurements are used to qualify the calculational methodology. Because of the importance and the difficulty of these calculations, the method`s qualification by comparison to measurement must be made to ensure a reliable and accurate vessel fluence determination. This reference calculation note is to provide a series of forward and adjoint neutron transport calculations for use in the evaluation of neutron dosimetry from surveillance capsule irradiations at 3-loop PWR reactor as well as for use in the determination of the neutron exposure of the reactor vessel wall in accordance with U.S Regulatory Guide DG-1025 requirements. The calculations of the pressure vessel fluence consist of the following steps; (1) Determination of the geometrical and material input data, (2) Determination of the core neutron source, and (3) Propagation of the neutron fluence from the core to the vessel and into the cavity. (author). 12 tabs., 3 figs., 7 refs.

  1. Analysis of a bio-electrochemical reactor containing carbon fiber textiles for the anaerobic digestion of tomato plant residues.

    Science.gov (United States)

    Hirano, Shin-Ichi; Matsumoto, Norio

    2017-11-03

    A bio-electrochemical system packed with supporting material can promote anaerobic digestion for several types of organic waste. To expand the target organic matters of a BES, tomato plant residues (TPRs), generated year-round as agricultural and cellulosic waste, were treated using three methanogenic reactors: a continuous stirred tank reactor (CSTR), a carbon fiber textile (CFT) reactor, and a bio-electrochemical reactor (BER) including CFT with electrochemical regulation (BER + CFT). CFT had positive effects on methane fermentation and methanogen abundance. The microbial population stimulated by electrochemical regulation, including hydrogenotrophic methanogens, cellulose-degrading bacteria, and acetate-degrading bacteria, suppressed acetate accumulation, as evidenced by the low acetate concentration in the suspended fraction in the BER + CFT. These results indicated that the microbial community in the BER + CFT facilitated the efficient decomposition of TPR and its intermediates such as acetate to methane. Copyright © 2017 Elsevier Ltd. All rights reserved.

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

    Energy Technology Data Exchange (ETDEWEB)

    J. M. Beck; L. F. Pincock

    2011-04-01

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

  3. Selection of Technical Reactor Equipment for Modular, Continuous Small-Scale Plants

    Directory of Open Access Journals (Sweden)

    Nicolai Krasberg

    2014-03-01

    Full Text Available Fast process development, flexible production and the utilization of advanced process conditions are the main goals of modular and continuous small-scale plants (MCSPs. A configurable layout of the modules and the use of predefined equipment enable a quick and reliable conceptual process development and scale-up of continuous processes. Therefore, a computer-assisted selection methodology was developed and is presented, which allows the quick selection of plug flow reactor equipment for homogeneous liquid phase reactions. It identifies a favorable technical apparatus and the configuration in the early stages of process development. This can lead to the effective planning and guiding of scale-up experiments and closes the gap between lab and process development.

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

    Science.gov (United States)

    Higgy, R H; Pimpl, M

    1998-12-01

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

  5. Potential of Electric Power Production from Microbial Fuel Cell (MFC in Evapotranspiration Reactor for Leachate Treatment Using Alocasia macrorrhiza Plant and Eleusine indica Grass

    Directory of Open Access Journals (Sweden)

    Zaman Badrus

    2018-01-01

    Full Text Available Microbial fuel cell is one of attractive electric power generator from nature bacterial activity. While, Evapotranspiration is one of the waste water treatment system which developed to eliminate biological weakness that utilize the natural evaporation process and bacterial activity on plant roots and plant media. This study aims to determine the potential of electrical energy from leachate treatment using evapotranspiration reactor. The study was conducted using local plant, namely Alocasia macrorrhiza and local grass, namely Eleusine Indica. The system was using horizontal MFC by placing the cathodes and anodes at different chamber (i.e. in the leachate reactor and reactor with plant media. Carbon plates was used for chatode-anodes material with size of 40 cm x 10 cm x1 cm. Electrical power production was measure by a digital multimeter for 30 days reactor operation. The result shows electric power production was fluctuated during reactor operation from all reactors. The electric power generated from each reactor was fluctuated, but from the reactor using Alocasia macrorrhiza plant reach to 70 μwatt average. From the reactor using Eleusine Indica grass was reached 60 μwatt average. Electric power production fluctuation is related to the bacterial growth pattern in the soil media and on the plant roots which undergo the adaptation process until the middle of the operational period and then in stable growth condition until the end of the reactor operation. The results indicate that the evapotranspiration reactor using Alocasia macrorrhiza plant was 60-95% higher electric power potential than using Eleusine Indica grass in short-term (30-day operation. Although, MFC system in evapotranspiration reactor system was one of potential system for renewable electric power generation.

  6. 75 FR 5632 - Office of New Reactors; Interim Staff Guidance on the Review of Nuclear Power Plant Designs Using...

    Science.gov (United States)

    2010-02-03

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Office of New Reactors; Interim Staff Guidance on the Review of Nuclear Power Plant Designs Using... Commission (NRC). ACTION: Solicitation of public comment. SUMMARY: The NRC staff is soliciting public comment...

  7. Comparison of early socialization practices used for litters of small-scale registered dog breeders and nonregistered dog breeders.

    Science.gov (United States)

    Korbelik, Juraj; Rand, Jacquie S; Morton, John M

    2011-10-15

    OBJECTIVE-To compare early socialization practices between litters of breeders registered with the Canine Control Council (CCC) and litters of nonregistered breeders advertising puppies for sale in a local newspaper. DESIGN-Retrospective cohort study. Animals-80 litters of purebred and mixed-breed dogs from registered (n = 40) and non-registered (40) breeders. PROCEDURES-Registered breeders were randomly selected from the CCC website, and nonregistered breeders were randomly selected from a weekly advertising newspaper. The litter sold most recently by each breeder was then enrolled in the study. Information pertaining to socialization practices for each litter was obtained through a questionnaire administered over the telephone. RESULTS-Registered breeders generally had more breeding bitches and had more litters than did nonregistered breeders. Litters of registered breeders were more likely to have been socialized with adult dogs, people of different appearances, and various environmental stimuli, compared with litters of nonregistered breeders. Litters from registered breeders were also much less likely to have been the result of an unplanned pregnancy. CONCLUSIONS AND CLINICAL RELEVANCE-Among those breeders represented, litters of registered breeders received more socialization experience, compared with litters of nonregistered breeders. People purchasing puppies from nonregistered breeders should focus on socializing their puppies between the time of purchase and 14 weeks of age. Additional research is required to determine whether puppies from nonregistered breeders are at increased risk of behavioral problems and are therefore more likely to be relinquished to animal shelters or euthanized, relative to puppies from registered breeders.

  8. EFFICIENCY OF REMOVING ORGANIC POLLUTANTS FROM WASTEWATER IN A HOUSEHOLD WASTEWATER TREATMENT PLANT WITH A HYBRID REACTOR

    Directory of Open Access Journals (Sweden)

    Michał Marzec

    2016-11-01

    Full Text Available The purpose of the study is to evaluate the efficiency of removing organic pollutants from wastewater in a household wastewater treatment plant with a hybrid reactor. The efficiency of the plant was analysed in terms of decreasing BOD5, COD and the content of total suspended solids in differentiated technological conditions. The variable parameter was the length of the reactor aeration phase and the dwell phase. Six different aeration variants were analysed. The efficiency of a wastewater treatment plant with a hybrid reactor in terms of reducing BOD5 oscillated from 81 to 90%, COD – 64–84%, whereas total suspended solids – 66–89% which, as a rule, was not sufficient to achieve standard levels. The best results were achieved for the following variants: 2h aeration – 2h dwell and 1h aeration – 0.5h dwell. Lower efficiency of the plant could have been attributed to operational problems, mainly failures of the plant control system and uneven conditions such as air temperature, volume and composition of wastewater received in the plant.

  9. Practitioners'--breeders' approach to canine parturition.

    Science.gov (United States)

    Freak, M J

    1975-04-05

    Both veterinary surgeon and dog breeder should be involved in assessing the whelping capability of brood bitches as one essential point in the selection of sound stock. Normal parturition is described in a manner that might be used in the instruction of breeders and nursing auxiliaries in midwifery. In its classical three stages, the mechanics and hydraulics are explained in simple terms and the hormonal changes discussed in relation to each stage. The type of co-operation desired between veterinary surgeon and breeder is discussed and the instruction to be given on the recognition of dystocia is outlined. Some commonly met dystocias are described. Simple digitally-assisted delivery by the midwife/breeder is described as are the forceps techniques which may be applied by the veterinary surgeon in the course of diagnosis and delivery.

  10. Major welfare issues in broiler breeders

    NARCIS (Netherlands)

    Jong, de I.C.; Guemene, D.

    2011-01-01

    Under current practices, broiler parent stock (broiler breeders) encounter several welfare problems, such as feed restriction and injury during mating. Intensive selection for production traits, especially growth rate, is associated with increased nutritious requirement and thus feed consumption,

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

    Science.gov (United States)

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

    2016-01-01

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

  12. Study Neutronic of Small Pb-Bi Cooled Non-Refuelling Nuclear Power Plant Reactor (SPINNOR) with Hexagonal Geometry Calculation

    Science.gov (United States)

    Nur Krisna, Dwita; Su'ud, Zaki

    2017-01-01

    Nuclear reactor technology is growing rapidly, especially in developing Nuclear Power Plant (NPP). The utilization of nuclear energy in power generation systems has been progressing phase of the first generation to the fourth generation. This final project paper discusses the analysis neutronic one-cooled fast reactor type Pb-Bi, which is capable of operating up to 20 years without refueling. This reactor uses Thorium Uranium Nitride as fuel and operating on power range 100-500MWtNPPs. The method of calculation used a computer simulation program utilizing the SRAC. SPINNOR reactor is designed with the geometry of hexagonal shaped terrace that radially divided into three regions, namely the outermost regions with highest percentage of fuel, the middle regions with medium percentage of fuel, and most in the area with the lowest percentage. SPINNOR fast reactor operated for 20 years with variations in the percentage of Uranium-233 by 7%, 7.75%, and 8.5%. The neutronic calculation and analysis show that the design can be optimized in a fast reactor for thermal power output SPINNOR 300MWt with a fuel fraction 60% and variations of Uranium-233 enrichment of 7%-8.5%.

  13. Database structure and file layout of Nuclear Power Plant Database. Database for design information on Light Water Reactors in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Yamamoto, Nobuo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Izumi, Fumio

    1995-12-01

    The Nuclear Power Plant Database (PPD) has been developed at the Japan Atomic Energy Research Institute (JAERI) to provide plant design information on domestic Light Water Reactors (LWRs) to be used for nuclear safety research and so forth. This database can run on the main frame computer in the JAERI Tokai Establishment. The PPD contains the information on the plant design concepts, the numbers, capacities, materials, structures and types of equipment and components, etc, based on the safety analysis reports of the domestic LWRs. This report describes the details of the PPD focusing on the database structure and layout of data files so that the users can utilize it efficiently. (author).

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

    Energy Technology Data Exchange (ETDEWEB)

    Curtis Smith; Scott Beck; Bill Galyean

    2005-09-01

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

  15. Uranium resources and their implications for fission breeder and fusion hybrid development

    Energy Technology Data Exchange (ETDEWEB)

    Max, C.E.

    1984-05-15

    Present estimates of uranium resources and reserves in the US and the non-Communist world are reviewed. The resulting implications are considered for two proposed breeder technologies: the liquid metal fast breeder reactor (LMFBR) and the fusion hybrid reactor. Using both simple arguments and detailed scenarios from the published literature, conditions are explored under which the LMFBR and fusion hybrid could respectively have the most impact, considering both fuel-supply and economic factors. The conclusions emphasize strong potential advantages of the fusion hybrid, due to its inherently large breeding rate. A discussion is presented of proposed US development strategies for the fusion hybrid, which at present is far behind the LMFBR in its practical application and maturity.

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

    Energy Technology Data Exchange (ETDEWEB)

    1994-01-15

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

  17. Plutonium Worlds. Fast Breeders, Systems Analysis and Computer Simulation in the Age of Hypotheticality

    OpenAIRE

    Sebastian Vehlken

    2014-01-01

    This article examines the media history of one of the hallmark civil nuclear energy programs in Western Germany – the development of Liquid Metal Fast Breeder Reactor (LMFBR) technology. Promoted as a kind of perpetuum mobile of the Atomic Age, the "German Manhattan Project" not only imported big science thinking. In its context, nuclear technology was also put forth as an avantgarde of scientific inquiry, dealing with the most complex and critical technological endeavors. In the face of the ...

  18. Leak before break application in French PWR plants under operation

    Energy Technology Data Exchange (ETDEWEB)

    Faidy, C. [EDF SEPTEN, Villeurbanne (France)

    1997-04-01

    Practical applications of the leak-before break concept are presently limited in French Pressurized Water Reactors (PWR) compared to Fast Breeder Reactors. Neithertheless, different fracture mechanic demonstrations have been done on different primary, auxiliary and secondary PWR piping systems based on similar requirements that the American NUREG 1061 specifications. The consequences of the success in different demonstrations are still in discussion to be included in the global safety assessment of the plants, such as the consequences on in-service inspections, leak detection systems, support optimization,.... A large research and development program, realized in different co-operative agreements, completes the general approach.

  19. Development of self-cooled liquid metal breeder blankets

    Energy Technology Data Exchange (ETDEWEB)

    Malang, S.; Tillack, M.S. [comps.; Barleon, L.; Baumgaertner, S.; Borgstedt, H.U.; Buehler, L.; Buerkle, G.; Dammel, F.; Feuerstein, H.; Fischer, U.; Gabel, K.; Gerhardt, H.; Glasbrenner, H.; Heider, T.; Jordan, T.; Kleefeldt, K.; Kleykamp, H.; Lindau, R.; Moeslang, A.; Norajitra, F.; Reimann, G.; Reimann, J.; Riesch-Oppermann, H.; Ritzhaupt-Kleissl, H.J.; Schleisiek, K.; Schmitz, G.; Schnauder, H.; Stieglitz, R.; Tellini, B.; Tsige-Tamirat, H.

    1995-11-01

    The development of liquid metal breeder blankets for fusion reactors has been performed in the Forschungszentrum Karlsruhe as a part of the European fusion blanket development program with the aim to select the two most promising concepts in 1995 for further development. In this report are described the designs of self-cooled blankets together with the results of the accompanying R and D program of the years 1992-1995. The program includes design studies as well as theoretical and experimental work in the fields of neutronics, magneto-hydrodynamics, thermohydraulics, mechanical stresses, compatibility and purification of lead-lithium, tritium extraction and control, safety, reliability, electrical insulating coatings, and fabrication technologies for blanket segments. (orig.) 250 refs.

  20. Reacting flow simulations of supercritical water oxidation of PCB-contaminated transformer oil in a pilot plant reactor

    Directory of Open Access Journals (Sweden)

    V. Marulanda

    2011-06-01

    Full Text Available The scale-up of a supercritical water oxidation process, based on recent advancements in kinetic aspects, reactor configuration and optimal operational conditions, depends on the research and development of simulation tools, which allow the designer not only to understand the complex multiphysics phenomena that describe the system, but also to optimize the operational parameters to attain the best profit for the process and guarantee its safe operation. Accordingly, this paper reports a multiphysics simulation with the CFD software Comsol Multiphysics 3.3 of a pilot plant reactor for the supercritical water oxidation of a heavily PCB-contaminated mineral transformer oil. The proposed model was based on available information for the kinetic aspects of the complex mixture and the optimal operational conditions obtained in a lab-scale continuous supercritical water oxidation unit. The pilot plant simulation results indicate that it is not feasible to scale-up directly the optimal operational conditions obtained in the isothermal lab-scale experiments, due to the excess heat released by the exothermic oxidation reactions that result in outlet temperatures higher than 600°C, even at reactor inlet temperatures as low as 400°C. Consequently, different alternatives such as decreasing organic flowrates or a new reactor set-up with multiple oxidant injections should be considered to guarantee a safe operation.

  1. Formation of the research competence of future breeder-geneticists during the practical training

    Directory of Open Access Journals (Sweden)

    Наталія Павлівна Антіпова

    2016-03-01

    Full Text Available In the article were analyzed content, structure and special features of the practical training of the future breeder-geneticists. There was specified that for bachelor breeder-geneticists of the 6.090101 “Agronomy” training direction are provided the following practices: botany, tractors and cars, soil science, agricultural machines, fruit growing, vegetable growing, agriculture, agrochemistry, plant growing, entomology, phytopathology, technology of retreatment and preservation of plant growing production, selection and seed growing of agricultural crops; work practice on agronomy.In the article was grounded the structural-logic scheme of organization of practices of the future breeder-geneticists for forming their professional research competences. There were analyzed functions and methods of scientific-cognitive activity, forms, means and methods of the practical education. According to the types of practices there were elucidated and grounded the structure of practical training of the future breeder-geneticists – natural-technical and special-professional – educational and work practices.The natural-technical, special-professional and production-technological practices form in bachelor agronomists breeder-geneticists professional scientific-research competences: theoretical-analytical ones – geosphere, bio-geo-physical-chemical, meteorological-climatological, informative, fundamental and applied ecological, fundamental-scientific agrosphere; professional competences of professionally oriented and special practical training – agricultural, branch, techno-service, technological, selective, special-genetic-engineering ones; managerial ones

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

    Science.gov (United States)

    Juhasz, Albert J.

    2007-01-01

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

  3. Evaluation Of Communal Wastewater Treatment Plant Operating Anaerobic Baffled Reactor And Biofilter

    Directory of Open Access Journals (Sweden)

    Evy Hendriarianti

    2016-02-01

    Full Text Available Construction of communal Waste Water Treatment Plant, WWTP in city of Malang since 1998 but until recently had never done an evaluation the performance. Communal WWTP performance evaluation is needed to see how far the efficiency of processing result. Until now, Environmental Agency Malang City only measure effluent from WWTP Communal  to know the suitability  with domestic wastewater quality standards. Effluent quality data in 2014 showed value above the quality standard of domestic wastewater from East Java Governor Regulation No. 72 in 2013 for parameters BOD and COD. WWTP Communal USRI research objects are on a six (6 locations by involving the user community during the planning, construction, operation and maintenance. Technology choice of ABR followed by a biofilter reactor with the stone media proved capable of processing organic matter of BOD and COD with the removal levels respectively by 78% -99% and 71% -99%. As for the parameters of TSS, NO3 and PO4 have the ranges of removal respectively by 56% -100%, (43% - 72%, (2% - 13%. Ratio BOD and COD in influent are low and ranged from 0.22 to 0.41. From the evaluation shows that high organic matter concentrations in influent along with the HRT and operation time high will result in a higher removal level

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

    Energy Technology Data Exchange (ETDEWEB)

    Turbat, A.; Deschanels, H.; Sperandio, M. [and others

    1997-04-01

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

  5. Regularities pertinent to formation of hydraulic nonuniformities at the outlet from the reactor plant header system

    Science.gov (United States)

    Gabrianovich, B. N.; Del'nov, V. N.

    2014-05-01

    Maintaining the preset distribution of coolant mass flow rate at the outlet from the header system organized as a back turn with lateral admission of coolant to the header and central removal of coolant from it is one of the conditions essential for securing reliable and efficient operation of a reactor plant. In the header system, coolant passes through the annular lateral channel formed by the vessel and central barrel, after which it enters into the header, changes the motion direction in it, and goes out through the channels of the grid placed in the central barrel. The results obtained from experimental investigations of the hydrodynamics in the header system flow paths carried out on models with flat and cylindrical shapes are presented. A previously unknown regularity pertinent to formation of hydraulic irregularities at the header system outlet is revealed, and semi-empirical correlations for determining the coolant mass flow rate distribution at the header system outlet are obtained. The regularity connected with formation of hydrodynamic nonuniformities at the header system outlet lies in the fact that the position of the maximum coolant velocity at the grid outlet coincides with the position of the maximum velocity in the core of the incident jet attacking the grid, and that the maximal average coolant velocity in the grid holes is proportional to the average coolant velocity in the incident jet.

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

    Energy Technology Data Exchange (ETDEWEB)

    Zdarek, J.; Pecinka, L. [Nuclear Research Institute Rez (Czech Republic)

    1997-04-01

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

  7. Development of fuels and structural materials for fast breeder reactors

    Indian Academy of Sciences (India)

    R. Narasimhan (Krishtel eMaging) 1461 1996 Oct 15 13:05:22

    detrimental especially from welding considerations, and globular oxides are least harmful. For grades 304L(N) and 316L(N) SS, the grain size number is specified as larger than ASTM. No. 2 (i.e., a finer grain size), to achieve optimum high temperature mechanical properties and to permit meaningful ultrasonic examination.

  8. Conceptual design of Indian molten salt breeder reactor

    Indian Academy of Sciences (India)

    cooled options. Another option, which ... the quantity of waste. It is thus an optimum solution for meeting the energy needs of a large country like India in a sustainable manner, securing its energy freedom in the long term. Schematic of Indian ...

  9. Thermal control of solid breeder blankets

    Energy Technology Data Exchange (ETDEWEB)

    Raffray, A.R.; Ying, A.; Gorbis, Z.; Tillack, M.S.; Abdou, M.A.

    1991-12-31

    An assessment of the thermal control mechanisms applicable to solid breeder blanket designs under ITER-like operating conditions is presented in this paper. Four cases are considered: a helium gap; a sintered block Be region; a sintered block helium region with a metallic felt at the Be/clad interface; and a Be packed bed region. For these cases, typical operating are explored to determine the ranges of wall load which can be accommodated while maintaining the breeder within its allowable operating temperature window. The corresponding region thicknesses are calculated to help identify practicality and design tolerances.

  10. Thermal control of solid breeder blankets

    Energy Technology Data Exchange (ETDEWEB)

    Raffray, A.R.; Ying, A.; Gorbis, Z.; Tillack, M.S.; Abdou, M.A.

    1991-01-01

    An assessment of the thermal control mechanisms applicable to solid breeder blanket designs under ITER-like operating conditions is presented in this paper. Four cases are considered: a helium gap; a sintered block Be region; a sintered block helium region with a metallic felt at the Be/clad interface; and a Be packed bed region. For these cases, typical operating are explored to determine the ranges of wall load which can be accommodated while maintaining the breeder within its allowable operating temperature window. The corresponding region thicknesses are calculated to help identify practicality and design tolerances.

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

    Energy Technology Data Exchange (ETDEWEB)

    Lata

    1996-09-26

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

  12. Empirical Risk Analysis of Severe Reactor Accidents in Nuclear Power Plants after Fukushima

    Directory of Open Access Journals (Sweden)

    Jan Christian Kaiser

    2012-01-01

    Full Text Available Many countries are reexamining the risks connected with nuclear power generation after the Fukushima accidents. To provide updated information for the corresponding discussion a simple empirical approach is applied for risk quantification of severe reactor accidents with International Nuclear and Radiological Event Scale (INES level ≥5. The analysis is based on worldwide data of commercial nuclear facilities. An empirical hazard of 21 (95% confidence intervals (CI 4; 62 severe accidents among the world’s reactors in 100,000 years of operation has been estimated. This result is compatible with the frequency estimate of a probabilistic safety assessment for a typical pressurised power reactor in Germany. It is used in scenario calculations concerning the development in numbers of reactors in the next twenty years. For the base scenario with constant reactor numbers the time to the next accident among the world's 441 reactors, which were connected to the grid in 2010, is estimated to 11 (95% CI 3.7; 52 years. In two other scenarios a moderate increase or decrease in reactor numbers have negligible influence on the results. The time to the next accident can be extended well above the lifetime of reactors by retiring a sizeable number of less secure ones and by safety improvements for the rest.

  13. DESIGN ANALYSIS OF A PREPACKAGED NUCLEAR POWER PLANT (1000 eKW). VOLUME II. REACTOR DESIGN ANALYSIS

    Energy Technology Data Exchange (ETDEWEB)

    None

    1959-02-01

    Reactor and primary and secondary system design analyses of a prepackaged nuclear power plant are presented. The plant is of the pressurized water steam-electric generating type which is air transportable skid mounted and suitable for a remote site installation. The plant design incorporates the maximum degree of reliability and simplicity and the minimum degree of maintenance and cost. The plant is capable of producing a gross electrical output of 1000 kw(e) and 5 x 10/sup 6/ Btu/hr for space heating. The primary system generates a total of 6.5 Mw of heat in order to supply the required electrical and heating load. (W.D.M.)

  14. Design criteria -- Reactor plant modifications for increased production and 100-C Area Alterations (Sections A and B) CG-558. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Russ, M.H.

    1954-08-10

    This document defines the basic criteria to be used in the preparation of detailed design for Project CG-558, Reactor Plant Modification for Increased Production and for Project CG-600, 100-C Area Alterations. It has been determined that the most economical method of increasing plutonium production within the next five years is by the modernization and improvement of the 100-B, 100-C, 100-D, 100-DR, 100-F, and 100-H reactor plants. These reactors are currently incapable of operating at their maximum potential power levels because of a limited availability of process cooling water. As a result of this programs, it is estimated that 1650-2350 megawatts of total additional production will be achieved. The purpose of this document is to set forth the design for certain modifications and additions to Hanford reactors and their supporting facilities as required to obtain higher power levels and improve the safety of reactor operation.

  15. Analysis of the documents about the core envelopment of nuclear reactor at the Laguna Verde U-1 power plant; Analisis de documentos de los materiales de la envolvente del nucleo del reactor nuclear de la CLV U-1

    Energy Technology Data Exchange (ETDEWEB)

    Zamora R, L.; Medina F, A. [Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    1999-07-01

    The degradation of internal components at BWR type reactors is an important subject to consider in the performance availability of the power plant. The Wuergassen nuclear reactor license was confiscated due to the presence of cracking in the core envelopment. In consequence it is necessary carrying out a detailed study with the purpose to avoid these problems in the future. This report presents a review and analysis of documents and technical information referring to the core envelopment of a BWR/5/6 and the Laguna Verde Unit 1 nuclear reactor in Mexico. In this document are presented design data, documents about fabrication processes, and manufacturing of core envelopment. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Naus, D.J.; Oland, C.B. [Oak Ridge National Lab., TN (United States); Ellingwood, B.R. [Johns Hopkins Univ., Baltimore, MD (United States)] [and others

    1997-01-01

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

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

    Science.gov (United States)

    Nair, Abhilash T; Ahammed, M Mansoor

    2014-09-01

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

  18. Pilot Experimentation with Complete Mixing Anoxic Reactors to Improve Sewage Denitrification in Treatment Plants in Small Communities

    Directory of Open Access Journals (Sweden)

    Massimo Raboni

    2013-12-01

    Full Text Available This paper reports the results of two sewage treatment tests in a community of 15,000 inhabitants. The sewage treatment plant is subject to strong fluctuations in load (BOD5, COD, TKN, and in particular in the BOD5/TKN ratio. These fluctuations adversely affect the biological denitrification, as demonstrated by many pilot and real-scale plants. The plants we tested were subjected to two treatment types: anoxic-aerobic and simultaneous denitrification. Both processes are designed for complete mixing conditions in the reactors in order to level the fluctuations in the load and thus improve the denitrification efficiency. The results prove that an average denitrification efficiency of up to 80% can be achieved with the sludge loading close to 0.1 kg BOD5 (d∙kgMLVSS−1. The effect of the sludge loading and dissolved oxygen on the denitrification efficiency is highlighted.

  19. Efficient cycles for carbon capture CLC power plants based on thermally balanced redox reactors

    KAUST Repository

    Iloeje, Chukwunwike

    2015-10-01

    © 2015 Elsevier Ltd. The rotary reactor differs from most alternative chemical looping combustion (CLC) reactor designs because it maintains near-thermal equilibrium between the two stages of the redox process by thermally coupling channels undergoing oxidation and reduction. An earlier study showed that this thermal coupling between the oxidation and reduction reactors increases the efficiency by up to 2% points when implemented in a regenerative Brayton cycle. The present study extends this analysis to alternative CLC cycles with the objective of identifying optimal configurations and design tradeoffs. Results show that the increased efficiency from reactor thermal coupling applies only to cycles that are capable of exploiting the increased availability in the reduction reactor exhaust. Thus, in addition to the regenerative cycle, the combined CLC cycle and the combined-regenerative CLC cycle are suitable for integration with the rotary reactor. Parametric studies are used to compare the sensitivity of the different cycle efficiencies to parameters like pressure ratio, turbine inlet temperature, carrier-gas fraction and purge steam generation. One of the key conclusions from this analysis is that while the optimal efficiency for regenerative CLC cycle was the highest of the three (56% at 3. bars, 1200. °C), the combined-regenerative cycle offers a trade-off that combines a reasonably high efficiency (about 54% at 12. bars, 1200. °C) with much lower gas volumetric flow rate and consequently, smaller reactor size. Unlike the other two cycles, the optimal compressor pressure ratio for the regenerative cycle is weakly dependent on the design turbine inlet temperature. For the regenerative and combined regenerative cycles, steam production in the regenerator below 2× fuel flow rate improves exhaust recovery and consequently, the overall system efficiency. Also, given that the fuel side regenerator flow is unbalanced, it is more efficient to generate steam from the

  20. Design data and safety features of commerical nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Heddleson, F.A.

    1976-06-01

    Design data, safety features, and site characteristics are summarized for 34 nuclear power units in 17 power stations in the United States. Six pages of data are presented for each plant, consisting of thermal-hydraulic and nuclear factors, containment features, emergency-core-cooling systems, site features, circulating water system data, and miscellaneous factors. An aerial perspective is also presented for each plant. This volume covers Light Water Reactors (LWRs) with dockets 50-508 through 50-549, four HTGRs--50-171, 50-267, 50-450/451, 50-463/464, the Atlantic Floating Station 50-477/478, and the Clinch River Breeder 50-537.

  1. Plutonium Worlds. Fast Breeders, Systems Analysis and Computer Simulation in the Age of Hypotheticality

    Directory of Open Access Journals (Sweden)

    Sebastian Vehlken

    2014-09-01

    Full Text Available This article examines the media history of one of the hallmark civil nuclear energy programs in Western Germany – the development of Liquid Metal Fast Breeder Reactor (LMFBR technology. Promoted as a kind of perpetuum mobile of the Atomic Age, the "German Manhattan Project" not only imported big science thinking. In its context, nuclear technology was also put forth as an avantgarde of scientific inquiry, dealing with the most complex and critical technological endeavors. In the face of the risks of nuclear technology, German physicist Wolf Häfele thus announced a novel epistemology of "hypotheticality". In a context where traditional experimental engineering strategies became inappropiate, he called for the application of advanced media technologies: Computer Simulations (CS and Systems Analysis (SA generated computerized spaces for the production of knowledge. In the course of the German Fast Breeder program, such methods had a twofold impact. One the one hand, Häfele emphazised – as the "father of the German Fast Breeder" – the utilization of CS for the actual planning and construction of the novel reactor type. On the other, namely as the director of the department of Energy Systems at the International Institute for Applied Systems Analysis (IIASA, Häfele advised SA-based projections of energy consumption. These computerized scenarios provided the rationale for the conception of Fast Breeder programs as viable and necessary alternative energy sources in the first place. By focusing on the role of the involved CS techniques, the paper thus investigates the intertwined systems thinking of nuclear facilities’s planning and construction and the design of large-scale energy consumption and production scenarios in the 1970s and 1980s, as well as their conceptual afterlives in our contemporary era of computer simulation.

  2. High-temperature nuclear reactor power plant cycle for hydrogen and electricity production – numerical analysis

    Directory of Open Access Journals (Sweden)

    Dudek Michał

    2016-01-01

    Full Text Available High temperature gas-cooled nuclear reactor (called HTR or HTGR for both electricity generation and hydrogen production is analysed. The HTR reactor because of the relatively high temperature of coolant could be combined with a steam or gas turbine, as well as with the system for heat delivery for high-temperature hydrogen production. However, the current development of HTR’s allows us to consider achievable working temperature up to 750°C. Due to this fact, industrial-scale hydrogen production using copper-chlorine (Cu-Cl thermochemical cycle is considered and compared with high-temperature electrolysis. Presented calculations show and confirm the potential of HTR’s as a future solution for hydrogen production without CO2 emission. Furthermore, integration of a hightemperature nuclear reactor with a combined cycle for electricity and hydrogen production may reach very high efficiency and could possibly lead to a significant decrease of hydrogen production costs.

  3. Studying the capture cross sections of constructional elements from measurements of the neutron balance in breeder media

    Energy Technology Data Exchange (ETDEWEB)

    Golubev, V.I.; Dulin, V.A.; Kazanskii, Yu.A.; Darrouzet, M.; Martin-Deidier, L.; Rimpault, G.

    1987-04-01

    Until recently, the indeterminacy in the group capture cross sections of constructional elements at neutron energies above 1 keV were estimated at 15-20%, leading to an error of 0.2 and 1%, respectively, in calculating K/sub ef/ and the conversion factor of breeder reactors with oxide fuel and sodium coolant. In fact, calculations using the BNAB-78 group constants used in the USSR for the design development of fast reactors show that the mean neutron capture cross section of constructional elements (iron, nickel, and chromium) is approx. 1.4 times greater for a typical breeder reactor than in the case of calculation by the version of Carnaval IV used in France for the Superphoenix reactors. To refine the proportion of neutrons absorbed in stainless steel, the neutron balance in media consisting of uranium fuel and stainless steel with nickel in a proportion ensuring a near-unity breeder coefficient of infinite media of this composition (K/sub infinity/ = 1) has been measured at the Power-Physics Institute in Obninsk and at the Center for Nuclear Research in Caradache. The results obtained allow the accuracy of calculating the proportion of neutrons absorbed in constructional elements to be judged

  4. FOWL CHOLERA IN A BREEDER FLOCK

    Directory of Open Access Journals (Sweden)

    Z. Parveen, A. A. Nasir, K.Tasneem and A. Shah

    2003-12-01

    Full Text Available During January, 2003 Pasteurella multocida the causative agent of fowl cholera was isolated from a breeder flock in Lahore District. The age of the flock was 245 days. Increased mortality, swollen wattles and lameness were the clinical findings present in almost all the affected birds, while gross lesions were typical of fowl cholera. To prove the virulence of the organism, mice and six-week old cockerals were infected and P. multocida was reisolated.

  5. Ecological and toxicological aspects of the partial meltdown of the Chernobyl nuclear power plant reactor

    Science.gov (United States)

    Eisler, Ronald; Hoffman, David J.; Rattner, Barnett A.; Burton, G. Allen; Cairns, John

    1995-01-01

    the partial meltdown of the 1000-MW reactor at Chernobyl, Ukraine, on April 26, 1986, released large amounts of radiocesium and other radionuclides into the environment, causing widespread radioactive contamination of Europe and the former Soviet Union.1-7 At least 3,000,000 trillion becquerels (TBq) were released from the fuel during the accident (Table 24.1), dwarfing, by orders of magnitude, radiation released from other highly publicized reactor accidents at Windscale (U.K.) and three-Mile Island (U.S.)3,8 The Chernobyl accident happened while a test was being conducted during a normal scheduled shutdown and is attributed mainly to human error.3

  6. Management of waste from the International Thermonuclear Experimental Reactor and from future fusion power plants

    Energy Technology Data Exchange (ETDEWEB)

    Broden, K. [Association EURATOM, Nykoeping (Sweden); Lindberg, M. [Association EURATOM, Nykoeping (Sweden); Nisan, S. [The NET Team, Garching (Germany); Rocco, P. [European Commission, Institute for Advanced Materials, Joint Research Centre, Ispra (Vatican City State, Holy See) (Italy); Zucchetti, M. [Energetics Department, Polytechnic of Turin, Corso Duca degli Abruzzi 24, I-10129 Torino (Italy); Taylor, N. [Association EURATOM-UKAEA, UKAEA Fusion, Culham, Abingdon, Oxfordshire, OX14 3DB (United Kingdom); Forty, C. [Association EURATOM-UKAEA, UKAEA Fusion, Culham, Abingdon, Oxfordshire, OX14 3DB (United Kingdom)

    1997-04-01

    An important inherent advantage of fusion would be the total absence of high-level radioactive spent fuel as produced in fission reactors. Fusion will, however, produce activated material containing both activation products and tritium. Part of the material may also contain chemically toxic substances. This paper describes methods that could be used to manage these materials and also methods to reduce or entirely eliminate the waste quantities. The results are based on studies for the International Thermonuclear Experimental Reactor and also for future fusion power station designs currently under investigation within the European programme on the safety and environmental assessment of fusion power, long-term. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Donohue, M.L.; Price, M.E. (eds.)

    1984-07-01

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

  8. Initial design and test of the tritium breeder monitoring system for the lead-lithium cooled ceramic breeder (LLCB) module of the ITER

    Energy Technology Data Exchange (ETDEWEB)

    Kapyshev, V., E-mail: kapyshev@nikiet.ru [Federal State Unitary Enterprise “Dollezhal Research and Development Institute of Power Engineering”, PO Box 788, Moscow 101000 (Russian Federation); Danilov, I.; Kartashev, I.; Kovalenko, V.; Leshukov, A.; Poliksha, V.; Razmerov, A.; Strebkov, Yu.; Sviridenko, M.; Trusova, E.; Vladimirova, N. [Federal State Unitary Enterprise “Dollezhal Research and Development Institute of Power Engineering”, PO Box 788, Moscow 101000 (Russian Federation); Kalashnikov, A. [Rosatom (Russian Federation)

    2013-10-15

    Highlights: • We propose a system for measurement of tritium-breeding dynamics in module of the ITER. • Lithium carbonate sensors and neutron detectors are used for the measurements. • The sensors and detectors are irradiated by neutrons under ITER plasma operations. • A pneumatic concept is suggested for conveying of the samples in the module. • The results of the tritium measurements in the sensors after irradiation are discussed. -- Abstract: The demonstration of a tritium breeder is an important part of an ITER mission. A concept for an experimental estimation of the tritium-breeding ratio (TBR) and the tritium-breeding dynamics in a test breeder module (TBM) of ITER has been developed. A system for the experimental estimation of the values is proposed for the lead–lithium cooled ceramic breeder (LLCB) TBM of ITER. The system is based on tritium breeder and neutron flux measurements under ITER plasma D-T experiments and the use of lithium carbonate sensors and neutron detectors. Three capsules with lithium carbonate (Li{sub 2}CO{sub 3}) containing different relative abundances of lithum-6 and lithium-7 (Li-6/Li-7 ratios of that found in nature: 1/1 and 9/1) and three capsules with neutron detectors are placed in a container. The low-level activities of the structural materials of the container and the capsules are used to prevent operation in a hot cell after the reactor irradiation of the container. For the delivery/withdrawal of the containers into/from the TBM, a pneumatic concept is suggested with a monitor channel connecting the TBM and an operating zone for conveying the containers in the TBM before the pulse and extracting them after the pulse. The initial design of the container with the capsules for the samples and the channel part in the TBM are presented in this paper. A laboratory facility for the investigation of the pneumatic parameters and the container moving in the channel is proposed. Neutron calculation is performed to estimate

  9. Remote means of nondestructive testing of shell-type reactors of nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Grebennikov, V.V.; Lebedev, N.E.

    1984-11-01

    Equipment for the remote nondestructive testing of the welds in the shell of a water-cooled, water-moderated power reactor vessel is described. Television equipment is used for inspection. Ultrasonic monitoring is done in order to detect cracks, peeling, and separation of the austenite hard facing in the shells.

  10. Reactor operating procedures for start up of continuously operated chemical plants

    NARCIS (Netherlands)

    Verwijs, J.W.; Verwijs, J.W.; Kösters, P.H.; van den Berg, Henderikus; Westerterp, K.R.; Kosters, P.G.H.

    1995-01-01

    Rules are presented for the startup of an adiabatic tubular reactor, based on a qualitative analysis of the dynamic behavior of continuously-operated vapor- and liquid-phase processes. The relationships between the process dynamics, operating criteria, and operating constraints are investigated,

  11. A catalytically membrane reactor for fast, highly exothermic, heterogeneous gas reactions : a pilot plant study

    NARCIS (Netherlands)

    Veldsink, J.W.; Veldsink, J.W.; Versteeg, Geert; van Swaaij, Willibrordus Petrus Maria

    1995-01-01

    Membrane reactors have been frequently studied because of their ability to combine chemical activity and separation properties into one device. Due to their thermal stability and mechanical strength, ceramic membranes are preferred over polymeric ones, but small transmembrane fluxes obstruct a

  12. A Catalytically Active Membrane Reactor for Fast, Highly Exothermic, Heterogeneous Gas Reactions. A Pilot Plant Study

    NARCIS (Netherlands)

    Veldsink, Jan W.; Versteeg, Geert F.; Swaaij, Wim P.M. van

    1995-01-01

    Membrane reactors have been frequently studied because of their ability to combine chemical activity and separation properties into one device. Due to their thermal stability and mechanical strength, ceramic membranes are preferred over polymeric ones, but small transmembrane fluxes obstruct a

  13. What determines hatchling weight: breeder age or incubated egg weight?

    Directory of Open Access Journals (Sweden)

    AB Traldi

    2011-12-01

    Full Text Available Two experiments were carried out to determine which factor influences weight at hatch of broiler chicks: breeder age or incubated egg weight. In Experiment 1, 2340 eggs produced by 29- and 55-week-old Ross® broiler breeders were incubated. The eggs selected for incubation weighed one standard deviation below and above average egg weight. In Experiment 2, 2160 eggs weighing 62 g produced by breeders of both ages were incubated. In both experiments, 50 additional eggs within the weight interval determined for each breeder age were weighed, broken, and their components were separated and weighed. At hatch, hatchlings were sexed and weighed, determining the average initial weight of the progeny of each breeder age. Data were analyzed using the Analyst program of SAS® software package. In Experiment 1, the weight difference between eggs produced by young and mature breeders was 10.92 g, and the component that mostly influenced this difference was the yolk (7.51 g heavier in mature breeders, compared with 4.23 g difference in albumen and 0.8 g in eggshell weights. Hatchling weight difference was 9.4 g higher in eggs from mature breeders. In Experiment 2, egg weight difference was only 0.74 g, but yolk weight was 4.59 g higher in the eggs of mature breeders. The results obtained in the present study indicate that hatchling weight is influenced by egg weight, and not by breeder age.

  14. Survey of Regulations Applicable to the Finned Containment in Korean Nuclear Power Plant for Light Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Noh, Hyung Gyun [Pohang University, Pohang (Korea, Republic of); Kang, Hie Chan [Kunsan University, Gunsan (Korea, Republic of)

    2016-05-15

    In severe accident, the molten corium would discharge into the reactor cavity and interact with water and concrete of cavity. Molten corium includes non-oxidation metals such as Zr, Fe and Cr. These metal species reacted with water emit hydrogen gas. In addition to this, a mount of steam can be emitted to the containment such as steam line break accident. As a result, steam and hydrogen gas can pressurize containment over the design pressure and threaten its integrity. For this reasons, a concept equipped with finned on the containment building was proposed for coping with prolonged accident. Finned containment can enhance heat transfer to the ambient, and the building itself is working as a heat sink. Multiple metal fins and metal rod are penetrated into containment wall, and the rods are working as an additional path of heat removal. To be accepted in the nuclear power plants, this configuration should satisfy the requirement of heat removal and follow all regulations related with containment also. For applying to Korean nuclear power plants, the finned containment should follow all regulations specialized in Korea such as Nuclear regulatory criteria for light water reactor and Guidelines of nuclear safety examination for light water reactor. A concept of containment as a passive cooling system has been proposed. Furthermore, the new containment concept can be applied on the real containment which satisfies the various regulations. Finned containment would be expected positive effects on heat removal from the containment. If the fins are properly welded to the liner, finned containment could satisfy the leak tightness and prevention of external influences. Finned containment could be favorable to protect external impact like aircraft crash because of the additional structural integrity by the fins.

  15. HEMERA: a 3D coupled core-plant system for accidental reactor transient simulation

    Energy Technology Data Exchange (ETDEWEB)

    Bruna, G.B.; Fouquet, F.; Dubois, F. [Institut de Radioprotection et de Surete Nucleaire, 92 - Fontenay aux Roses (France); Le Pallec, J.C.; Richebois, E.; Hourcade, E.; Poinot-Salanon, C.; Royer, E. [CEA Saclay, Dir. de l' Energie Nucleaire (DEN/DM2S), 91 - Gif sur Yvette (France)

    2007-07-01

    In the framework of their collaboration to develop a system to study reactor transients in safety-representative conditions, IRSN (Radioprotection and Nuclear Safety Institute) and Cea have launched the development of a fully coupled 3-dimensional computational chain, called HEMERA (Highly Evolutionary Methods for Extensive Reactor Analyses), based on the French SAPHYR code system, composed of APOLLO-2, CRONOS-2 and FLICA-4 codes, and the system code CATHARE. It includes cross sections generation, steady-state, depletion and transient computation capabilities in a consistent approach. Multi-level and multi-dimensional models are developed to account for neutronics, core thermal-hydraulics, fuel thermal analysis and system thermohydraulics. Currently Control Rod Ejection (RIA) and Main Steam Line Break (MSLB) accidents are investigated. The HEMERA system is presently applied to French PWR. The present paper outlines the main physical phenomena to be accounted for in such a coupled computational chain with significant time and space effects.

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

    Energy Technology Data Exchange (ETDEWEB)

    Donohue, M.L.; Price, M.E. (eds.)

    1984-07-01

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

  17. Empirical Risk Analysis of Severe Reactor Accidents in Nuclear Power Plants after Fukushima

    OpenAIRE

    Jan Christian Kaiser

    2012-01-01

    Many countries are reexamining the risks connected with nuclear power generation after the Fukushima accidents. To provide updated information for the corresponding discussion a simple empirical approach is applied for risk quantification of severe reactor accidents with International Nuclear and Radiological Event Scale (INES) level ≥5. The analysis is based on worldwide data of commercial nuclear facilities. An empirical hazard of 21 (95% confidence intervals (CI) 4; 62) severe accidents am...

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

    Energy Technology Data Exchange (ETDEWEB)

    1986-10-01

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

  19. Reactor safety study. An assessment of accident risks in U. S. commercial nuclear power plants. Appendix VI. Calculation of reactor accident consequences. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    1975-10-01

    Information is presented concerning the radioactive releases from the containment following accidents; radioactive inventory of the reactor core; atmospheric dispersion; reactor sites and meteorological data; radioactive decay and deposition from plumes; finite distance of plume travel; dosimetric models; health effects; demographic data; mitigation of radiation exposure; economic model; and calculated results with consequence model.

  20. Cyclic purging for low-temperature solid fusion reactor blanket operation

    Energy Technology Data Exchange (ETDEWEB)

    Was, G.S.; Lidsky, L.M.

    1979-05-01

    A purging process was developed that will permit operation of fusion reactor blankets employing solid LiAlO/sub 2/ as the breeder material at fuel temperatures of <600/sup 0/C. The low fuel temperature would greatly reduce the problems of fuel sintering, densification, and volume expansion that occur at fuel temperatures in excess of 900/sup 0/C without degrading the plant thermal efficiency. The process consists of heating the blanket to a specified temperature for a given time at regular intervals to release tritium held up in the breeding material. As an example, a detailed purging cycle was developed for the breeder rod shim rod blanket that uses LiAlO/sub 2/ in the form of micronsize particles compacted into millimeter-size pellets and is designed for low-temperature operation. Tritium inventory, doubling time, purging time, purging temperature, purging frequency, and particle size are the parameters used to evaluate the process. Calculations indicate that breeder particle sizes ranging from 20 to 50 ..mu..m and purging temperatures ranging from 600 to 700/sup 0/C can result in purge times of <1 h with three or more weeks between purges, and a doubling time of 7 yr for a blanket inventory limit of 5 kg and a breeding ratio of <1.02.

  1. [Reoccurrence of histomonosis in turkey breeder farm].

    Science.gov (United States)

    Aka, Johannes; Hauck, Rüdiger; Blankenstein, Petra; Balczulat, Stefanie; Hafez, Hafez Mohamed

    2011-01-01

    Histomonosis is a severe disease caused by the protozoan parasite Histomonas (H.) meleagridis, which can lead to high losses in turkeys. The present report describes the reoccurrence of histomonosis in a turkey breeder farm. The first outbreak occurred in 2005 in 17 weeks old hens, the second in 2009 in 8 weeks old hens. The disease remained restricted in one house and one compartment, respectively. Mortality rose to 26 and 65% respectively within few days in spite of therapy with various compounds. Both flocks had to be euthanized. In both cases H. meleagridis belonging to genotype A was detected. The source of infection remained unclear in both cases.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-05-01

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

  3. Guideline for Performing Systematic Approach to Evaluate and Qualify Legacy Documents that Support Advanced Reactor Technology Activity

    Energy Technology Data Exchange (ETDEWEB)

    Honma, George [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-10-01

    The establishment of a systematic process for the evaluation of historic technology information for use in advanced reactor licensing is described. Efforts are underway to recover and preserve Experimental Breeder Reactor II and Fast Flux Test Facility historical data. These efforts have generally emphasized preserving information from data-acquisition systems and hard-copy reports and entering it into modern electronic formats suitable for data retrieval and examination. The guidance contained in this document has been developed to facilitate consistent and systematic evaluation processes relating to quality attributes of historic technical information (with focus on sodium-cooled fast reactor (SFR) technology) that will be used to eventually support licensing of advanced reactor designs. The historical information may include, but is not limited to, design documents for SFRs, research-and-development (R&D) data and associated documents, test plans and associated protocols, operations and test data, international research data, technical reports, and information associated with past U.S. Nuclear Regulatory Commission (NRC) reviews of SFR designs. The evaluation process is prescribed in terms of SFR technology, but the process can be used to evaluate historical information for any type of advanced reactor technology. An appendix provides a discussion of typical issues that should be considered when evaluating and qualifying historical information for advanced reactor technology fuel and source terms, based on current light water reactor (LWR) requirements and recent experience gained from Next Generation Nuclear Plant (NGNP).

  4. Suggestion on the safety classification of spent fuel dry storage in China’s pressurized water reactor nuclear power plant

    Science.gov (United States)

    Liu, Ting; Qu, Yunhuan; Meng, De; Zhang, Qiaoer; Lu, Xinhua

    2018-01-01

    China’s spent fuel storage in the pressurized water reactors(PWR) is stored with wet storage way. With the rapid development of nuclear power industry, China’s NPPs(NPPs) will not be able to meet the problem of the production of spent fuel. Currently the world’s major nuclear power countries use dry storage as a way of spent fuel storage, so in recent years, China study on additional spent fuel dry storage system mainly. Part of the PWR NPP is ready to apply for additional spent fuel dry storage system. It also need to safety classificate to spent fuel dry storage facilities in PWR, but there is no standard for safety classification of spent fuel dry storage facilities in China. Because the storage facilities of the spent fuel dry storage are not part of the NPP, the classification standard of China’s NPPs is not applicable. This paper proposes the safety classification suggestion of the spent fuel dry storage for China’s PWR NPP, through to the study on China’s safety classification principles of PWR NPP in “Classification for the items of pressurized water reactor nuclear power plants (GB/T 17569-2013)”, and safety classification about spent fuel dry storage system in NUREG/CR - 6407 in the United States.

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

    Science.gov (United States)

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

    2014-11-01

    A laboratory plant consisting of two UASB reactors was used for the treatment of industrial wastewater from the wheat starch industry. Several load tests were carried out with starch wastewater and the synthetic substrates glucose, acetate, cellulose, butyrate and propionate to observe the impact of changing loads on gas yield and effluent quality. The measurement data sets were used for calibration and validation of the Anaerobic Digestion Model No. 1 (ADM1). For a precise simulation of the detected glucose degradation during load tests with starch wastewater and glucose, it was necessary to incorporate the complete lactic acid fermentation into the ADM1, which contains the formation and degradation of lactate and a non-competitive inhibition function. The modelling results of both reactors based on the modified ADM1 confirm an accurate calculation of the produced gas and the effluent concentrations. Especially, the modelled lactate effluent concentrations for the load cases are similar to the measurements and justified by literature. Copyright © 2014 Elsevier Ltd. All rights reserved.

  6. Characterization of an electrochemical pilot-plant filter-press reactor by hydrodynamic and mass transport studies

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez-Garcia, J.; Frias, A.; Exposito, E.; Montiel, V.; Aldaz, A.; Conesa, J.A.

    2000-05-01

    This work deals with the study of the influence of turbulence promoters in the hydrodynamic and mass transport behavior of a pilot-plant filter-press electrolyzer (a homemade UA200.08 with a 200 cm{sup 2} electrode area) in an undivided configuration. A simple experimental arrangement was used to generate data from electrolytic conductivity measurements in a series of impulse-response experiments. The presence and type of turbulence promoters influence the flow distribution inside the reactor. A new design of a model (presented in a previous work) has been used to analyze the residence time distributions. In this study a new parameter, the turbulence factor, given as N{sub {alpha}}{Phi}{sub {beta}}, was employed to classify the turbulence promoters. The optimization of the parameters indicates that the correct model is dispersed plug-flow behavior with a low axial dispersion that considers exchange between the dead and main zones of the reactor. It is also very interesting to highlight that the information obtained by means of the turbulence factors is similar to that obtained from the values of the mass transport coefficients measured using the limiting current technique.

  7. 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.

  8. Industrial solar breeder project using concentrator photovoltaics

    Energy Technology Data Exchange (ETDEWEB)

    Hamilton, R; Wohlgemuth, J; Burkholder, J; Levine, A; Storti, G; Wrigley, C; McKegg, A

    1979-08-01

    The purpose of this program is to demonstrate the use of a concentrating photovoltaic system to provide the energy for operating a silicon solar cell production facility, i.e., to demonstrate a solar breeder. Solarex has proposed to conduct the first real test of the solar breeder concept by building and operating a 200 kW(e) (peak) concentrating photovoltaic system based on the prototype and system design developed during Phase I. This system will provide all of the electrical and thermal energy required to operate a solar cell production line. This demonstration would be conducted at the Solarex Rockville facility, with the photovoltaic array located over the company parking lot and on an otherwise unusable flood plain. Phase I of this program included a comprehensive analysis of the application, prototype fabrication and evaluation, system design and specification, and a detailed plan for Phases II and III. A number of prototype tracking concentrator solar collectors were constructed and operated. Extensive system analysis was performed to design the Phase II system as a stand-alone power supply for a solar cell production line. Finally, a detailed system fabrication proposal for Phase II and an operation and evaluation plan for Phase III were completed. These proposals included technical, management, and cost plans for the fabrication and exercise of the proposed system.

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

    Science.gov (United States)

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

    2016-01-01

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

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

    Science.gov (United States)

    Juhasz, Albert J.

    2014-01-01

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

  11. FASTER test reactor preconceptual design report summary

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-02-29

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

  12. FASTER Test Reactor Preconceptual Design Report

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-03-31

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

  13. Numerical investigation on stress corrosion cracking behavior of dissimilar weld joints in pressurized water reactor plants

    Directory of Open Access Journals (Sweden)

    Lingyan Zhao

    2014-07-01

    Full Text Available There have been incidents recently where stress corrosion cracking (SCC observed in the dissimilar metal weld (DMW joints connecting the reactor pressure vessel (RPV nozzle with the hot leg pipe. Due to the complex microstructure and mechanical heterogeneity in the weld region, dissimilar metal weld joints are more susceptible to SCC than the bulk steels in the simulated high temperature water environment of pressurized water reactor (PWR. Tensile residual stress (RS, in addition to operating loads, has a great contribution to SCC crack growth. Limited experimental conditions, varied influence factors and diverging experimental data make it difficult to accurately predict the SCC behavior of DMW joints with complex geometry, material configuration, operating loads and crack shape. Based on the film slip/dissolution oxidation model and elastic-plastic finite element method (EPFEM, an approach is developed to quantitatively predict the SCC growth rate of a RPV outlet nozzle DMW joint. Moreover, this approach is expected to be a pre-analytical tool for SCC experiment of DMW joints in PWR primary water environment.

  14. Advanced Burner Reactor Preliminary NEPA Data Study.

    Energy Technology Data Exchange (ETDEWEB)

    Briggs, L. L.; Cahalan, J. E.; Deitrich, L. W.; Fanning, T. H.; Grandy, C.; Kellogg, R.; Kim, T. K.; Yang, W. S.; Nuclear Engineering Division

    2007-10-15

    The Global Nuclear Energy Partnership (GNEP) is a new nuclear fuel cycle paradigm with the goals of expanding the use of nuclear power both domestically and internationally, addressing nuclear waste management concerns, and promoting nonproliferation. A key aspect of this program is fast reactor transmutation, in which transuranics recovered from light water reactor spent fuel are to be recycled to create fast reactor transmutation fuels. The benefits of these fuels are to be demonstrated in an Advanced Burner Reactor (ABR), which will provide a representative environment for recycle fuel testing, safety testing, and modern fast reactor design and safeguard features. Because the GNEP programs will require facilities which may have an impact upon the environment within the meaning of the National Environmental Policy Act of 1969 (NEPA), preparation of a Programmatic Environmental Impact Statement (PEIS) for GNEP is being undertaken by Tetra Tech, Inc. The PEIS will include a section on the ABR. In support of the PEIS, the Nuclear Engineering Division of Argonne National Laboratory has been asked to provide a description of the ABR alternative, including graphics, plus estimates of construction and operations data for an ABR plant. The compilation of this information is presented in the remainder of this report. Currently, DOE has started the process of engaging industry on the design of an Advanced Burner Reactor. Therefore, there is no specific, current, vendor-produced ABR design that could be used for this PEIS datacall package. In addition, candidate sites for the ABR vary widely as to available water, geography, etc. Therefore, ANL has based its estimates for construction and operations data largely on generalization of available information from existing plants and from the environmental report assembled for the Clinch River Breeder Reactor Plant (CRBRP) design [CRBRP, 1977]. The CRBRP environmental report was chosen as a resource because it thoroughly

  15. Abrupt or gradual increases in photoperiod for broiler breeders ...

    African Journals Online (AJOL)

    Broiler breeders transferred from closed rearing to curtain-sided adult accommodation (similar to that in which a large proportion of broiler breeders are housed) respond similarly to abrupt and gradual increases in daylength as do birds maintained throughout in controlled environment accommodation, and modern ...

  16. The response of broiler breeder hens to dietary balanced protein

    African Journals Online (AJOL)

    Research

    2016-08-26

    Aug 26, 2016 ... overconsumption of protein and energy. Various estimates of the daily intake of protein required by broiler breeders to support egg production have been made by researchers in the past, these varying from 16.5 g/d, for individually housed breeders. (Pearson & Herron, 1982) to 22 g/d (Waldroup et al., ...

  17. Parameters for quantification of hunger in broiler breeders.

    NARCIS (Netherlands)

    Jong, de I.C.; Voorst, van A.S.; Blokhuis, H.J.

    2003-01-01

    The commercial restricted feeding programme of broiler breeders has a major negative effect on welfare, as the birds are continuously hungry. Objective parameters of hunger are needed to evaluate new management or feeding systems that may alleviate hunger and thus improve broiler breeder welfare.

  18. Staphylococcus agnetis, a potential pathogen in broiler breeders

    DEFF Research Database (Denmark)

    Poulsen, Louise Ladefoged; Thøfner, Ida; Bisgaard, Magne

    2017-01-01

    including bacteriological examination. In total 997 breeders were investigated and for the first time Staphylococcus agnetis was isolated in pure culture from cases of endocarditis and septicemia from 16 broiler breeders. In addition, the cloacal flora from newly hatched chickens originating from the same...

  19. Characteristics of Reproductive Tracts of Repeat Breeders in Cattle ...

    African Journals Online (AJOL)

    The study observed the characteristics and assessed abnormalities in reproductive tracts of repeat breeders in cattle. Fourty (40) herds were sampled during the study and fifty seven (57) repeat breeders were identified. The animals had normal reproductive tracts with good body condition scores (BCS) ranging from 2.5- 4.5 ...

  20. Contribution of reactor physics in past and future. Is reactor physics useful?

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Toshikazu [Osaka Univ. (Japan); Kosaka, Shinya [TEPCO Systems Co. (Japan); Tatsumi, Masahiro [Nuclear Fuel Industries Ltd., Tokyo (Japan)] (and others)

    2003-02-01

    Reactor Physics is a science to create rector and to play an important role in application to calculation science and safety evaluation. This feature articles contains topics, interested problems and development problems in the following field of reactor physics such as theory and experiment of reactor physics, core control, safety evaluation, criticality safety, accelerator driven subcritical reactor (ADS), new type reactor and evaluation of reactor physics. An original nuclear calculation method developed in Japan has been applied to design and analysis of fast breeder reactor. Interested problems are a proposal of fundamental principles of progressive reactor, development of calculation science, new knowledge by application of best estimate method to safety evaluation and investigation of complicated phenomena of criticality safety. (S.Y.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Oxstrand, Johanna [Idaho National Lab. (INL), Idaho Falls, ID (United States); Le Blanc, Katya [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bly, Aaron [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-09-01

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

  2. The TEX-I real-time expert system, applied to situation assessment for the SNR-300 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Schmal, N.; Leder, H.J.; Schade, H.J.

    1988-01-01

    Interatom, a subsidiary company of Siemens, is developing expert systems for the technical domain. These systems are operating in various industrial applications like flexible manufacturing or plant configuration, based on a domain-specific expert system shell, developed by Interatom. Additional projects are focusing on real-time diagnostics, e.g., for nuclear power plants. The authors report in this paper about a diagnosis expert system for the liquid-metal fast breeder reactor SNR-300, which uses new real-time tools, developed within the German TEX-I project (technical expert systems for data interpretation, diagnosis, and process control). The purpose of the system is to support the reactor operators in assessing plant status in real time, based on readings from many sensors. By on-line connection to the process control computer, it can monitor all incoming signal values, check the consistency of data, continuously diagnose the current plant status, detect unusual trends prior to accidents, localize faulty components, and recommended operator response in abnormal conditions. In the present knowledge acquisition and test phase, the expert system is connected to a real-time simulation of the reactor. The simulator is based on a thermohydraulic code for simulation of the transient behavior of temperatures and flow rates in the reactor core, plena, pipes, pumps, valves, intermediate heat exchangers, and cooling components. Additionally, the system's response to an asynchronous operator interaction can be simulated.

  3. Review of activities of Research Association of High Temperature Gas Cooled Reactor Plant (RAHP)

    Energy Technology Data Exchange (ETDEWEB)

    An, Shigehiro [Research Association of High Temperature Gas Cooled Reactor Plant RAHP, University of Tokyo, Tokyo (Japan); Tsuchie, Yasuo [Research Association of High Temperature Gas Cooled Reactor Plant RAHP, The Japan Atomic Power Co. JAPC, Tokyo (Japan)

    1998-09-01

    The Research Association of the HTGR Plant (RAHP) is the sole research association in the private or industrial sector of Japan with respect to HTGR Plants. It was established in 1985, composing of professors, representatives of electric power companies, and fabricators of nuclear plant and fuels Activities in these years were to analyze world trends of R and D, to identify techno-economical issues to be cleared, to set-up fundamental development strategies, and to put the results of the studies into actions towards commercialization of the HTGR. Conclusions obtained through the activities so far are: (1) From the view point of effective use of energy and reduction of environmental impacts on a global scale, development of nuclear power is essential, in particular of the HTGR, because of its very highly inherent safety and feasibility of high temperature heat uses. The role of the HTGR is inter-complementary with those of LWR and FBR; (2) Future subjects on the HTGR are technical demonstration of its unique characteristics, economic prospects, public acceptance (PA) and industrial acceptance, R and D through international cooperation and share in role, and successful realization of demonstration plant(s). RAHP is to start a survey on HTGR from nuclear fuel cycle point of view to have a better outlook on future needs of high temperature heat uses. 6 refs.

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

    Science.gov (United States)

    Bertch, Timothy Creston

    1998-12-01

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

  5. Plasma instabilities of a charge breeder ECRIS

    Science.gov (United States)

    Tarvainen, O.; Angot, J.; Izotov, I.; Skalyga, V.; Koivisto, H.; Thuillier, T.; Kalvas, T.; Lamy, T.

    2017-10-01

    Experimental observation of plasma instabilities in a charge breeder electron cyclotron resonance ion source (CB-ECRIS) is reported. It is demonstrated that the injection of 133Cs+ or 85Rb+ ion beam into the oxygen discharge of the CB-ECRIS can trigger electron cyclotron instabilities, which restricts the parameter space available for the optimization of the charge breeding efficiency. It is concluded that the transition from a stable to unstable plasma regime is caused by gradual accumulation and ionization of Cs/Rb and simultaneous change of the discharge parameters in 10-100 ms time scale, not by a prompt interaction between the incident ion beam and the ECRIS plasma. The instabilities lead to loss of ion confinement, which results in the sputtering of the surfaces in contact with the plasma, followed by up to an order of magnitude increase of impurity currents in the extracted n+ ion beam.

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

    Science.gov (United States)

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

    2015-02-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-03-01

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

  8. Comparison of the technical and economic parameters of different variants of the nuclear fuel cycle reactors of the nuclear power plants

    Science.gov (United States)

    Adamov, E. O.; Rachkov, V. I.; Tolstoukhov, D. A.; Panov, S. A.

    2016-12-01

    The article presents the comparative economic analysis of the variants of implementation of the fuel cycles for fast and thermal reactors. Calculations are carried out for the formed reference conditions by the cost of processing stages of the fuel cycles for the foreign and expert Russian information sources. The comparative data on the resource supply, absolute and specific costs of the fuel cycle variants for the whole life cycle of the thermal power plant projects with the fast and thermal reactors are considered. The conclusions of the efficiency of the fuel cycle variants for the assumed reference conditions of the calculation are made.

  9. STARFIRE: a commercial tokamak fusion power plant study

    Energy Technology Data Exchange (ETDEWEB)

    1980-09-01

    STARFIRE is a 1200 MWe central station fusion electric power plant that utilizes a deuterium-tritium fueled tokamak reactor as a heat source. Emphasis has been placed on developing design features which will provide for simpler assembly and maintenance, and improved safety and environmental characteristics. The major features of STARFIRE include a steady-state operating mode based on continuous rf lower-hybrid current drive and auxiliary heating, solid tritium breeder material, pressurized water cooling, limiter/vacuum system for impurity control and exhaust, high tritium burnup and low vulnerable tritium inventories, superconducting EF coils outside the superconducting TF coils, fully remote maintenance, and a low-activation shield. A comprehensive conceptual design has been developed including reactor features, support facilities and a complete balance of plant. A construction schedule and cost estimate are presented, as well as study conclusions and recommendations.

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

    Energy Technology Data Exchange (ETDEWEB)

    Reddy, D.P. (ed)

    1983-04-01

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

  11. Effect of operating conditions and reactor configuration on efficiency of full-scale biogas plants

    DEFF Research Database (Denmark)

    Angelidaki, Irini; Boe, Kanokwan; Ellegaard, L.

    2005-01-01

    was higher than approx. 4 g-N/l the degradation efficiency of the plant decreased and as a consequence, the residual methane potential was high. Decrease of the residual methane potential with increasing hydraulic retention time was found. Digestion temperature was very important for effective post...

  12. Modeling and monitoring cyclic and linear volatile methylsiloxanes in a wastewater treatment plant using constant water level sequencing batch reactors

    Energy Technology Data Exchange (ETDEWEB)

    Wang, De-Gao, E-mail: degaowang@dlmu.edu.cn; Du, Juan; Pei, Wei; Liu, Yongjun; Guo, Mingxing

    2015-04-15

    The fate of cyclic and linear volatile methylsiloxanes (VMSs) was evaluated in a wastewater treatment plant (WWTP) using constant water level sequencing batch reactors from Dalian, China. Influent, effluent, and sewage sludge samples were collected for seven consecutive days. The mean concentrations of cyclic VMSs (cVMSs) in influent and effluent samples are 1.05 μg L{sup −1} and 0.343 μg L{sup −1}; the total removal efficiency of VMSs is > 60%. Linear VMS (lVMS) concentration is under the quantification limitation in aquatic samples but is found in sludge samples with a value of 90 μg kg{sup −1}. High solid-water partition coefficients result in high VMS concentrations in sludge with the mean value of 5030 μg kg{sup −1}. No significant differences of the daily mass flows are found when comparing the concentration during the weekend and during working days. The estimated mass load of total cVMSs is 194 mg d{sup −1} 1000 inhabitants{sup −1} derived for the population. A mass balance model of the WWTP was developed and derived to simulate the fate of cVMSs. The removal by sorption on sludge increases, and the volatilization decreases with increasing hydrophobicity and decreasing volatility for cVMSs. Sensitivity analysis shows that the total suspended solid concentration in the effluent, mixed liquor suspended solid concentration, the sewage sludge flow rate, and the influent flow rate are the most influential parameters on the mass distribution of cVMSs in this WWTP. - Highlights: • A mass balance model for siloxanes was developed in sequencing batch reactor. • Total suspended solid in effluent has the most influence on removal efficiency. • Enhancement of suspended solid removal reduces the release to aquatic environment.

  13. Technical and economic proposal for the extension of the Laguna Verde Nuclear Power plant with an additional nuclear reactor; Propuesta tecnica y economica para la ampliacion de la Central Nucleoelectrica Laguna Verde con un reactor nuclear adicional

    Energy Technology Data Exchange (ETDEWEB)

    Leal C, C.D.; Francois L, J.L. [Facultad de Ingenieria, UNAM, Circuito Interior, C.U. Coyoacan, 04510 Mexico D.F. (Mexico)]. e-mail: carlosdanielleal@yahoo.com.mx

    2006-07-01

    The increment of the human activities in the industrial environments and of generation of electric power, through it burns it of fossil fuels, has brought as consequence an increase in the atmospheric concentrations of the calls greenhouse effect gases and, these in turn, serious repercussions about the environment and the quality of the alive beings life. The recent concern for the environment has provoked that industrialized countries and not industrialized carry out international agreements to mitigate the emission from these gases to the atmosphere. Our country, like part of the international community, not is exempt of this problem for what is necessary that programs begin guided toward the preservation of the environment. As for the electric power generation, it is indispensable to diversify the sources of primary energy; first, to knock down the dependence of the hydrocarbons and, second, to reduce the emission of polluting gases to the atmosphere. In this item, the nucleo electric energy not only has proven to be safe and competitive technical and economically, able to generate big quantities of electric power with a high plant factor and a considerable cost, but rather also, it is one of the energy sources that less pollutants it emits to the atmosphere. The main object of this work is to carry out a technical and economic proposal of the extension of the Laguna Verde Nuclear power plant (CNLV) with a new nuclear reactor of type A BWR (Advanced Boiling Water Reactor), evolutionary design of the BWR technology to which belong the two reactors installed at the moment in the plant, with the purpose of increasing the installed capacity of generation of the CNLV and of the Federal Commission of Electricity (CFE) with foundation in the sustainable development and guaranteeing the protection of the environment by means of the exploitation of a clean and sure technology that counts at the moment with around 12,000 year-reactor of operational experience in more of

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

    Directory of Open Access Journals (Sweden)

    Luciano Piani

    2014-01-01

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

  15. Upgrading of a wastewater treatment plant with a hybrid moving bed biofilm reactor (MBBR

    Directory of Open Access Journals (Sweden)

    Luigi Falletti

    2014-11-01

    Full Text Available The wastewater treatment plant of Porto Tolle (RO, Italy was originally projected for 2200 person equivalent (p.e. and it was made of a pumping station, an activated sludge oxidation tank (395 m3, a settler (315 m3, and two sludge drying beds. Other units were not yet in use in 2008: a fine screen, a sand and grit removal unit, a new settler (570 m3, a disinfection tank and a sludge thickener. Effective hydraulic load was 245% higher, organic load was 46% lower and nitrogen load was 39% higher than project values. Moreover, higher pollutant loads and more strict emission limits for nitrogen were expected. So the plant was upgraded: the old settler was divided into a sector of 180 m3 that was converted into a predenitrification tank, and a sector of 100 m3 that was converted into a hybrid MBBR tank filled with 50% AnoxKaldnesTM K3 carriers; the new settler was connected to the hybrid MBBR, and the other units were started. Biofilm growth was observed two months after plant restarting, its concentration reached 1.1 gTS/m2 (0.26 kgTS/m3, while activated sludge concentration was 2.0–2.8 kgTSS/m3 in all the period of study. The upgraded plant treats 1587 m3/d wastewater with 57 kgCOD/d, 23 kgBOD/d and 13.3 kgN/d, and has a significant residual capacity; the effluent respects all emission limits.

  16. Gas-cooled reactors: the importance of their development

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.

    1979-06-01

    The nearest term GCR is the steam-cycle HTGR, which can be used for both power and process steam production. Use of SC-HTGRs permits timely introduction of thorium fuel cycles and of high-thermal-efficiency reactors, decreasing the need for mined U/sub 3/O/sub 8/ before arrival of symbiotic fueling of fast-thermal reactor systems. The gas-turbine HTGR offers prospects of lower capital costs than other nuclear reactors, but it appears to require longer and more costly development than the SC-HTGR. Accelerated development of the GT-HTGR is needed to gain the advantages of timely introduction. The Gas-Cooled Fast Breeder Reactor (GCFR) offers the possibility of fast breeder reactors with lower capital costs and with higher breeding ratios from oxide fuels. The VHTR provides high-temperature heat for hydrogen production.

  17. An analysis system for in-reactor behavior, FANTASI

    Energy Technology Data Exchange (ETDEWEB)

    Uto, Nariaki; Tsukimori, Kazuyuki; Negishi, Hitoshi; Enuma, Yasuhiro; Sugaya, Toshio; Sakai, Kimiaki [Japan Nucler Cycle Developmnet Inst., Tokai, Ibaraki (Japan)

    2001-06-01

    The Japan Nuclear Fuel Cycle Development Institute developed FANTASI (A Computational System For Analyzing Coupled Neutronic, Thermal-Hydraulic And Structural Behaviors In A Fast Breeder Reactor Core) to simulate a conditions where nuclear reaction, thermal-hydraulic behavior of coolant and deformation of core construction progress under mutual relation in reactor of a fast breeder reactor by cooperation of engineers in the fields of physics, thermal-hydraulics, structure, and information system on reactor. Here was described on system construction of FANTASI after describing progress of this development. And then, after introducing a case study using this system, applicability to transient phenomena in nuclear reactor was described. At last, with summarizing results of this development, its future development was also mentioned. (G.K.)

  18. Iron requirements of broiler breeder hens.

    Science.gov (United States)

    Taschetto, Diogo; Vieira, Sergio Luiz; Angel, Clara Roselina; Stefanello, Catarina; Kindlein, Liris; Ebbing, Marco Antonio; Simões, Cristina Tonial

    2017-09-01

    A study was conducted to investigate Fe requirements of broiler breeders. One-hundred-fifty-six Cobb 500 broiler breeder hens were individually placed in electrostatically painted cages at 22 weeks. The study was composed of an adaptation phase, in which hens were fed corn-soy-wheat bran diets until 35 wks. An Fe deficient mash diet (24.6 ppm Fe) was provided from 35 to 46 wk in order to induce a partial body Fe depletion. A production phase followed from 47 to 70 wk when hens were fed 6 diets with increasing Fe sulfate supplementation, which, upon analyses had 24.6, 48.6, 74.3, 99.6, 125.6, and 148.2 ppm Fe. Thirty hatching eggs from each treatment were randomly collected in the last wk of each production period and incubated. Hemoglobin and hematocrit were analyzed from 6 hens as well as all hatched chicks per treatment. Analyses of production and hatching data were conducted using quadratic polynomial (QP), broken-line (BL), and exponential asymptotic (EA) models. Effects of dietary Fe were observed for total eggs and total hatching eggs, egg yolk Fe content, and hen and chick hematocrit and hemoglobin (P < 0.05). These responses to added Fe were optimized when dietary Fe were 96.8, 97.1, 130.6, 122.6, 120.0, and 125.0 ppm (QP) and 76.4, 89.3, 135.0, 128.4, 133.8, and 95.0 ppm (BL) for total hatching eggs, egg yolk Fe content, and hen and chick hematocrit and hemoglobin, respectively. Optimization with the EA model was obtained for total hatching eggs, egg yolk Fe, and hen and chick hemoglobin at 97.9, 111.0, 77.9, and 96.3 ppm Fe for total hatching eggs, egg yolk Fe, and hen and chick hemoglobin, respectively. Adequate Fe levels are needed to maintain egg production as well as hatching chicks' indexes. Fe concentration in the yolk and diet are positively influenced. The average of all Fe requirement estimates obtained in the present study was 106 ppm total Fe, whereas averaged values for BL, QP, and EA models were 107, 113, and 97 ppm Fe, respectively. © 2017

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

    Energy Technology Data Exchange (ETDEWEB)

    Ballinger, Ronald G. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Wang, Chun Yun [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Kadak, Andrew [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Todreas, Neil [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Mirick, Bradley [Concepts, Northern Engineering and Research, Woburn, MA (United States); Demetri, Eli [Concepts, Northern Engineering and Research, Woburn, MA (United States); Koronowski, Martin [Concepts, Northern Engineering and Research, Woburn, MA (United States)

    2004-08-30

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

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

    Science.gov (United States)

    Hölgye, Z; Filgas, R

    2006-04-01

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

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

    Science.gov (United States)

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

    2016-02-01

    An innovative plant configuration based in an UASB reactor coupled to a hybrid aerobic membrane bioreactor designed for sustainable treatment of municipal wastewater at ambient temperatures and low hydraulic retention time was studied in terms of organic micropollutants (OMPs) removal. OMPs removal mechanisms, as well as the potential influence of biomass activity and physical conformation were assessed. Throughout all periods of operation (150 days) high organic matter removals were maintained (>95%) and, regarding OMPs removal, this innovative system has shown to be more efficient than conventional technologies for those OMPs which are prone to be biotransformed under anaerobic conditions. For instance, sulfamethoxazole and trimethoprim have both shown to be biodegradable under anaerobic conditions with similar efficiencies (removal efficiencies above 84%). OMPs main removal mechanism was found to be biotransformation, except in the case of musk fragrances which showed medium sorption onto sludge. OMPs removal was strongly dependent on the efficiency of the primary metabolism (organic matter degradation and nitrification) and the type of biomass. Copyright © 2015 Elsevier Ltd. All rights reserved.

  2. The Performance of Advanced Sequencing Batch Reactor in Wastewater Treatment Plant to Remove Organic Materials and Linear Alkyl Benzene Sulfonates

    Directory of Open Access Journals (Sweden)

    Eslami

    2015-07-01

    Full Text Available Background Linear alkyl benzene sulfonates (LAS are the most important ionic detergents that produce negative ions in the environment. These compounds enter surface waters through domestic and industrial wastewaters and cause environmental hazards. Objectives The present study was aimed at assessing the performance of advanced sequencing batch reactor (SBR in wastewater treatment plant of Yazd, Iran, to remove organic materials and detergents. Materials and Methods The present research was a descriptive longitudinal study conducted on 96 input and output samples of SBR system over eight months from October 2012 to June 2013. The studied parameters were biochemical oxygen demand 5 (BOD5, chemical oxygen demand (COD, and detergents (LAS, which were assessed through standard methods. Finally, the study data were analyzed through analysis of variance (ANOVA using software package for statistical analysis (SPSS. Results The mean inputs of BOD5, COD, and LAS to the SBR system were 292.40 ± 45.28, 597.15 ± 97.30, and 3.29 ± 0.92 mg/L, and the mean outputs were 20.59 ± 3.54, 59.34 ± 9.47, and 0.606 ± 0.09 mg/L, respectively. The removal efficiency of BOD5, COD and LAS were respectively 92.95%, 90.06% and 81.6%. The results of ANOVA indicated that there was a significant relationship between the mean inputs and outputs of BOD5, COD, and the detergents (P ≤ 0.001. Conclusions With the proper operation of wastewater the treatment plant and increasing the retention time, the removal efficiency of the detergents increased. In addition, according to the environmental standards for BOD5, COD and the detergents, the results of the present study indicated that the outputs of these parameters from the SBR system were appropriate for agricultural irrigation.

  3. Upgrading of the Mirmohana Wastewater Treatment Plant in Kish Island, Iran, using a moving bed biofilm reactor

    Directory of Open Access Journals (Sweden)

    Mehdi Ahmadi

    2014-09-01

    Full Text Available Background and purpose: The performance of full-scale moving bed biofilm reactor (MBBR was evaluated as an alternative for upgrading of Kish Island Mirmohana wastewater treatment plant. In this study activated sludge process upgrade to MBBR process and different operating parameters were compared. Materials and Methods: Effect of upgrading on different parameters such as organic loading rate (OLR, mixed liquor suspended solid (MLSS, mixed liquor volatile suspended solid, sludge retention time (SRT, sludge volume index (SVI, hydraulic loading rate (HLR, also removal efficiency of chemical oxygen demand, biochemical oxygen demand and total suspended solid were investigated. Results: The study results show that with increasing the average flow influent (625.97 ± 38.6-1335.3 ± 102.06 m3/d and reducing of aeration tank volume (300-150 m3, OLR (0.29 ± 0.1-1.82 ± 0.15, MLSS (1291.14 ± 463.43 mg/L-7382.85 ± 272.42 mg/L, SRT (12.5 ± 4.2 d-28.79 ± 3.84 d, SVI (54.94 ± 15.82-51.2 ± 9.31, HLR (13.85 ± 0.85-29.45 ± 2.25 m/d, and hydraulic retention time (4.61 ± 0.27-2.17 ± 0.17 h were changed. Effluent concentrations under this operation condition were below the guidelines for irrigation water. Conclusion: Hence, MBBR process is a good alternative for upgrading wastewater plants especially when they have inadequacy space or need modification that will require a large investment.

  4. ECRIS as ion source and charge breeder

    CERN Document Server

    Sortais, P; Chauvin, N; Curdy, Jean Claude; Geller, R; Lamy, T; Solé, P; Vieux-Rochaz, J L

    2002-01-01

    We will recall the main characteristics of ECR ion sources and their very good property of ionization efficiency. We will start with a review of on line use of ECR ion sources for production of mono or multicharged radioactive ions, and then we will specially focus our attention on charge breeding process for multicharged ion production. Initially developed for the PIAFE project from ISN Grenoble, the ECR charge breeder shows that the beam injection of a primary beam inside an ECR ion source is a very general process for beam production. We will review the latest results obtained on the ISN Grenoble test bench for the production of CW or pulsed metallic ion beams with the so called '1 sup + /n sup + ' method. New results are given for CW operation where the efficiency is particularly optimized for application to multicharged radioactive ion production (for example, 3.5% for Zn sup 1 sup +->Zn sup 9 sup + , 4.2% for Pb sup 3 sup +->Pb sup 2 sup 4 sup + , 5% for Rb sup 1 sup +->Rb sup 1 sup 5 sup +). Different ...

  5. Vibroacoustic diagnostics of main circulating pump of primary coolant circuit of WWER reactor power plant

    Energy Technology Data Exchange (ETDEWEB)

    Samarin, A.A.; Adamenkov, K.A.

    1983-02-01

    The theoretical foundations are briefly described of the method and the noise sources given. Measurements on the main circulating pumps of the Novovoronezh nuclear power plant have shown that the range of frequencies of structural noises of the main circulating pumps lies within 10 and 20,000 Hz. These vibrations and noise have a broad band random character and therefore the mathematical apparatus of the theory of random processes is used for processing measured data. The ''signature'' method is used for diagnostics of the main circulating pumps - this is an analytical method consisting in that the measured spectrum is compared with a spectrum measured in a failure-free state. Experience with practical applications shows that a poor state of the main circulating pump is always accompanied by changes in noise spectrum.

  6. Life time of nuclear power plants and new types of reactors; La duree de vie des centrales nucleaires et les nouveaux types de reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-05-01

    This report, realized by the Evaluation Parliamentary Office of scientific and technological choices, aims to answer simple but fundamental questions for the french electric power production. What are the phenomena which may limit the exploitation time of nuclear power plants? How can we fight against the aging, at which cost and with which safety? The first chapter presents the management of the nuclear power plants life time, an essential element of the park optimization but not a sufficient element. The second chapter details the EPR and the other reactors for 2015 as a bond between the today and tomorrow parks. The last chapter deals with the necessity of efforts in the research and development to succeed in 2035 and presents other reactors in project. (A.L.B.)

  7. Environmental Enrichment for Broiler Breeders: An Undeveloped Field.

    Science.gov (United States)

    Riber, Anja B; de Jong, Ingrid C; van de Weerd, Heleen A; Steenfeldt, Sanna

    2017-01-01

    Welfare problems, such as hunger, frustration, aggression, and abnormal sexual behavior, are commonly found in broiler breeder production. To prevent or reduce these welfare problems, it has been suggested to provide stimulating enriched environments. We review the effect of the different types of environmental enrichment for broiler breeders, which have been described in the scientific literature, on behavior and welfare. Environmental enrichment is defined as an improvement of the environment of captive animals, which increases the behavioral opportunities of the animal and leads to improvements in biological function. This definition has been broadened to include practical and economic aspects as any enrichment strategy that adversely affects the health of animals (e.g., environmental hygiene), or that has too many economic or practical constraints will never be implemented on commercial farms and thus never benefit animals. Environmental enrichment for broiler breeders often has the purpose of satisfying the behavioral motivations for feeding and foraging, resting, and/or encouraging normal sexual behavior. Potentially successful enrichments for broiler breeders are elevated resting places, cover panels, and substrate (for broiler breeders housed in cage systems). However, most of the ideas for environmental enrichment for broiler breeders need to be further developed and studied with respect to the use, the effect on behavior and welfare, and the interaction with genotype and production system. In addition, information on practical use and the economics of the production system is often lacking although it is important for application in practice.

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

    Energy Technology Data Exchange (ETDEWEB)

    1994-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    1994-08-01

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

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

    Science.gov (United States)

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

    2016-06-01

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

  11. Hybrid moving bed biofilm reactors: an effective solution for upgrading a large wastewater treatment plant.

    Science.gov (United States)

    Mannina, Giorgio; Viviani, Gaspare

    2009-01-01

    Over the last few years there has been a growing attention regarding the receiving water body quality state. As a matter of a fact, the Directive 91/271 of the European Union (EU) replaced the "emission standard" concept that fixes discharge limits depending on polluting emission characteristics, with the "stream standard" concept that fixes discharge limits for each polluting substance depending on self-depurative characteristics of the RWB. In this context, several WWTPs need to be upgraded in order to meet stricter effluent limits. The need of WWTP upgrading was also emphasized by the growing urbanization that have led, in most cases, to get overloaded WWTP due to an overcoming of the maximum WWTP capacity. In order to upgrade existing WWTP basically two main possibilities can be chosen: building new tanks or modify the WWTP by introducing new technologies such as the HMBBR systems. In this paper, such latter possibility was explored and as a case study an existing Italian WWTP (Acqua dei Corsari) located in Palermo (IT) was analysed. The main goal was to test the effectiveness of HMBBR systems with respect to the WWTP upgrading. The survey was carried out by means of model simulation and an HMBBR pilot plant. This latter was employed for the evaluation of the model parameters as well as kinetic coefficients for the HMBBR. The model results are encouraging towards the WWTP upgrading by means of HMBBR. As a matter of fact, the model simulation results showed that the WWTP maximum capacity can be upgraded from 480,000 up to one million PE.

  12. Bubble column reactor fluid dynamic study at pilot plant scale for residue and extra heavy crude oil upgrading technology

    Energy Technology Data Exchange (ETDEWEB)

    Sardella, R.; Medina, H. [Infrastructure and Upgrading Department PDVSA-Intevep (Venezuela); Zacarias, L.; Paiva, M. [Refining Department. PDVSA-Intevep (Venezuela)

    2011-07-01

    Bubble column reactors are used in several applications because of their simplicity and low cost; a new technology was developed to convert extra heavy crude oil into upgraded crude using a bubble column reactor. To design this kind of reactor, a lot of parameters like flow regime, gas hold up and dispersion coefficient have to be taken into account. This study aimed at determining the fluid dynamic behaviour of a bubble column working under Aquaconversion operating conditions. Experiments were undertaken on air-tap water and air-light oil systems under atmospheric conditions with various gas superficial velocities and liquid flowrates. Results showed that gas hold up increases with superficial gas velocity but is independent of liquid flowrate and that both systems tested work at the same flow regimes. This paper showed that under the experimental conditions used, this reactor tends to be a complete mixing reactor.

  13. Analysis of the accident at Fukushima Daiichi nuclear power plant in an A BWR reactor; Analisis del accidente de la planta nucleoelectrica de Fukushima Daiichi en un reactor tipo ABWR

    Energy Technology Data Exchange (ETDEWEB)

    Escorcia O, D. [UNAM, Facultad de Ingenieria, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico); Salazar S, E., E-mail: daniel.escorcia.ortiz@gmail.com [UNAM, Facultad de Ingenieria, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, 62250 Jiutepec, Morelos (Mexico)

    2016-09-15

    The present work aims to recreate the accident occurred at the Fukushima Daiichi nuclear power plant in Japan on March 11, 2011, making use of an academic simulator of forced circulation of the A BWR reactor provided by the IAEA to know the scope of this simulator. The simulator was developed and distributed by the IAEA for academic purposes and contains the characteristics and general elements of this reactor to be able to simulate transients and failures of different types, allowing also to observe the general behavior of the reactor, as well as several phenomena and present systems in the same. Is an educational tool of great value, but it does not have a scope that allows the training of plant operators. To recreate the conditions of the Fukushima accident in the simulator, we first have to know what events led to this accident, as well as the actions taken by operators and managers to reduce the consequences of this accident; and the sequence of events that occurred during the course of the accident. Differences in the nuclear power plant behavior are observed and interpreted throughout the simulation, since the Fukushima plant technology and the simulator technology are not the same, although they have several elements in common. The Fukushima plant had an event that by far exceeded the design basis, which triggered in an accident that occurred in the first place by a total loss of power supply, followed by the loss of cooling systems, causing a level too high in temperature, melting the core and damaging the containment accordingly, allowing the escape of hydrogen and radioactive material. As a result of the simulation, was determined that the scope of the IAEA academic simulator reaches the entrance of the emergency equipment, so is able to simulate almost all the events occurred at the time of the earthquake and the arrival of the tsunami in the nuclear power plant of Fukushima Daiichi. However, due to its characteristics, is not able to simulate later

  14. Questions to the reactors power upgrade of the Nuclear Power Plant of Laguna Verde; Cuestionamientos al aumento de potencia de los reactores de la Central Nuclear de Laguna Verde

    Energy Technology Data Exchange (ETDEWEB)

    Salas M, B., E-mail: salasmarb@yahoo.com.mx [UNAM, Facultad de Ciencias, Departamento de Fisica, Circuito exterior s/n, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)

    2014-08-15

    The two reactors of the Nuclear Power Plant of Laguna Verde (NPP-L V) were subjected to power upgrade labors with the purpose of achieving 20% upgrade on the original power; these labors concluded in August 24, 2010 for the Reactor 1 and in January 16, 2011 for the Reactor 2, however in January of 2014, the NNP-L V has not received by part of the Comision Nacional de Seguridad Nuclear y Salvaguardias (CNSNS) the new Operation License to be able to work with the new power, because it does not fulfill all the necessary requirements of safety. In this work is presented and analyzed the information obtained in this respect, with data provided by the Instituto Federal de Acceso a la Informacion Publica y Proteccion de Datos (IFAI) and the Comision Federal de Electricidad (CFE) in Mexico, as well as the opinion of some workers of the NPP-L V. The Governing Board of the CFE announcement that will give special continuation to the behavior on the operation and reliability of the NPP-L V, because the frequency of not announced interruptions was increased 7 times more in the last three years. (Author)

  15. Some safety considerations in the selection of redans for pool-type LMR plants

    Energy Technology Data Exchange (ETDEWEB)

    Pan, Y.C.; Pedersen, D.R.; Wang, C.Y.

    1985-01-01

    Three basic safety issues in the selection of the redan design for a pool type liquid metal fast breeder reactor plant are examined. The first area examined is the effect of the redan selection on the integrity of the primary system pressure boundary in normal and offset conditions. The second area is on the consequence of the hypothetical core disruptive accident. The third area is on the consequence of the loss of heat sink accident. Some general discussion and numerical results are presented which may help in the selection of an optimum redan design. 3 refs., 7 figs.

  16. A comparative study of kinetics of nuclear reactors

    Directory of Open Access Journals (Sweden)

    Obaidurrahman Khalilurrahman

    2009-01-01

    Full Text Available The paper deals with the study of reactivity initiated transients to investigate major differences in the kinetics behavior of various reactor systems under different operating conditions. The article also states guidelines to determine the safety limits on reactivity insertion rates. Three systems, light water reactors (pressurized water reactors, heavy water reactors (pressurized heavy water reactors, and fast breeder reactors are considered for the sake of analysis. The upper safe limits for reactivity insertion rate in these reactor systems are determined. The analyses of transients are performed by a point kinetics computer code, PKOK. A simple but accurate method for accounting total reactivity feedback in kinetics calculations is suggested and used. Parameters governing the kinetics behavior of the core are studied under different core states. A few guidelines are discussed to project the possible kinetics trends in the next generation reactors.

  17. Strategy to identify the causes and to solve a sludge granulation problem in methanogenic reactors: application to a full-scale plant treating cheese wastewater.

    Science.gov (United States)

    Macarie, Hervé; Esquivel, Maricela; Laguna, Acela; Baron, Olivier; El Mamouni, Rachid; Guiot, Serge R; Monroy, Oscar

    2017-08-26

    Granulation of biomass is at the basis of the operation of the most successful anaerobic systems (UASB, EGSB and IC reactors) applied worldwide for wastewater treatment. Despite of decades of studies of the biomass granulation process, it is still not fully understood and controlled. "Degranulation/lack of granulation" is a problem that occurs sometimes in anaerobic systems resulting often in heavy loss of biomass and poor treatment efficiencies or even complete reactor failure. Such a problem occurred in Mexico in two full-scale UASB reactors treating cheese wastewater. A close follow-up of the plant was performed to try to identify the factors responsible for the phenomenon. Basically, the list of possible causes to a granulation problem that were investigated can be classified amongst nutritional, i.e. related to wastewater composition (e.g. deficiency or excess of macronutrients or micronutrients, too high COD proportion due to proteins or volatile fatty acids, high ammonium, sulphate or fat concentrations), operational (excessive loading rate, sub- or over-optimal water upflow velocity) and structural (poor hydraulic design of the plant). Despite of an intensive search, the causes of the granulation problems could not be identified. The present case remains however an example of the strategy that must be followed to identify these causes and could be used as a guide for plant operators or consultants who are confronted with a similar situation independently of the type of wastewater. According to a large literature based on successful experiments at lab scale, an attempt to artificially granulate the industrial reactor biomass through the dosage of a cationic polymer was also tested but equally failed. Instead of promoting granulation, the dosage caused a heavy sludge flotation. This shows that the scaling of such a procedure from lab to real scale cannot be advised right away unless its operability at such a scale can be demonstrated.

  18. Reactivity control assembly for nuclear reactor. [LMFBR

    Science.gov (United States)

    Bollinger, L.R.

    1982-03-17

    This invention, which resulted from a contact with the United States Department of Energy, relates to a control mechanism for a nuclear reactor and, more particularly, to an assembly for selectively shifting different numbers of reactivity modifying rods into and out of the core of a nuclear reactor. It has been proposed heretofore to control the reactivity of a breeder reactor by varying the depth of insertion of control rods (e.g., rods containing a fertile material such as ThO/sub 2/) in the core of the reactor, thereby varying the amount of neutron-thermalizing coolant and the amount of neutron-capturing material in the core. This invention relates to a mechanism which can advantageously be used in this type of reactor control system.

  19. Modeling and monitoring cyclic and linear volatile methylsiloxanes in a wastewater treatment plant using constant water level sequencing batch reactors.

    Science.gov (United States)

    Wang, De-Gao; Du, Juan; Pei, Wei; Liu, Yongjun; Guo, Mingxing

    2015-04-15

    The fate of cyclic and linear volatile methylsiloxanes (VMSs) was evaluated in a wastewater treatment plant (WWTP) using constant water level sequencing batch reactors from Dalian, China. Influent, effluent, and sewage sludge samples were collected for seven consecutive days. The mean concentrations of cyclic VMSs (cVMSs) in influent and effluent samples are 1.05 μg L(-1) and 0.343 μg L(-1); the total removal efficiency of VMSs is >60%. Linear VMS (lVMS) concentration is under the quantification limitation in aquatic samples but is found in sludge samples with a value of 90 μg kg(-1). High solid-water partition coefficients result in high VMS concentrations in sludge with the mean value of 5030 μg kg(-1). No significant differences of the daily mass flows are found when comparing the concentration during the weekend and during working days. The estimated mass load of total cVMSs is 194 mg d(-1)1000 inhabitants(-1) derived for the population. A mass balance model of the WWTP was developed and derived to simulate the fate of cVMSs. The removal by sorption on sludge increases, and the volatilization decreases with increasing hydrophobicity and decreasing volatility for cVMSs. Sensitivity analysis shows that the total suspended solid concentration in the effluent, mixed liquor suspended solid concentration, the sewage sludge flow rate, and the influent flow rate are the most influential parameters on the mass distribution of cVMSs in this WWTP. Copyright © 2015 Elsevier B.V. All rights reserved.

  20. Pressurizing new reactors

    Energy Technology Data Exchange (ETDEWEB)

    Neill, J.S.

    1956-01-30

    The Technical Division was asked recently to consider designs for new reactors that would add 8000 MW capacity to the Savannah River Plant. One modification of the existing SRP design that would enable a higher power rating, and therefore require fewer new reactors, is an increase in the maximum pressure in the D{sub 2}O system. The existing reactors at SRP are designed for a maximum pressure in the gas plenum of only 5 psig. Higher pressures enable higher D{sub 2} temperatures and higher sheath temperatures without local boiling or burnout. The requirements in reactor cooling facilities at any given power level would therefore be reduced by pressurizing.

  1. Thermochemical comparison of the effectiveness of protium purging of fusion breeders

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, A.K.; Johnson, C.E.

    1985-07-01

    The tritium breeders for a fusion reactor, Li/sub 2/0, LiAl0/sub 2/, and Li/sub 4/Si0/sub 4/, are compared on a thermochemical basis in respect to their response to protium purging. Two oxygen activity levels, established by H/sub 2/O:H/sub 2/ ratios of 100: 1 and 1:100 are considered at the temperatures 900 and 1300K. In terms of tritium release (all gaseous forms), LiAl0/sub 2/ is better than Li/sub 2/0 and this in turn better than Li/sub 4/Si0/sub 4/. At 900K, Li/sub 2/0 and LiA10/sub 2/ release more tritium than at 1300K. Li/sub 4/Si0/sub 4/ releases more tritium at 1300K than at 900K.

  2. Thermochemical comparison of the effectiveness of protium purging of fusion breeders

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, A.K.; Johnson, C.E.

    1985-01-01

    The tritium breeders for a fusion reactor, Li/sub 2/O, LiAlO/sub 2/, and Li/sub 4/SiO/sub 4/, are compared on a thermochemical basis in respect to their response to protium purging. Two oxygen activity levels, established by H/sub 2/O:H/sub 2/ ratios of 100:1 and 1:100 are considered at the temperatures 900 and 1300K. In terms of tritium release (all gaseous forms). LiAlO/sub 2/ is better than Li/sub 2/O and this in turn better than Li/sub 4/SiO/sub 4/. At 900K, Li/sub 2/O and LiAlO/sub 2/ release more tritium than at 1300K. Li/sub 4/SiO/sub 4/ releases more tritium at 1300K than at 900K.

  3. DEM-CFD simulation of purge gas flow in a solid breeder pebble bed

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Hao [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230027 (China); Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Mianyang 621900 (China); Li, Zhenghong [Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Mianyang 621900 (China); University of Science and Technology of China, Hefei 230027 (China); Guo, Haibing [Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Mianyang 621900 (China); Ye, Minyou [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230027 (China); Huang, Hongwen, E-mail: inpclane@sina.com [Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Mianyang 621900 (China)

    2016-12-15

    Solid tritium breeding blanket applying pebble bed concept is promising for fusion reactors. Tritium bred in the pebble bed is purged out by inert gas. The flow characteristics of the purge gas are important for the tritium transport from the solid breeder materials. In this study, a randomly packed pebble bed was generated by Discrete Element Method (DEM) and verified by radial porosity distribution. The flow parameters of the purge gas in channels were solved by Computational Fluid Dynamics (CFD) method. The results show that the normalized velocity magnitudes have the same damped oscillating patterns with radial porosity distribution. Besides, the bypass flow near the wall cannot be ignored in this model, and it has a slight increase with inlet velocity. Furthermore, higher purging efficiency becomes with higher inlet velocity and especially higher in near wall region.

  4. Development of materials and manufacturing technologies for Indian fast reactor programme

    Energy Technology Data Exchange (ETDEWEB)

    Raj, Baldev; Jayakumar, T.; Bhaduri, A.K.; Mandal, Sumantra [Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

    2010-07-01

    Fast Breeder Reactors (FBRs) are vital towards meeting security and sustainability of energy for the growing economy of India. The development of FBRs necessitates extensive research and development in domains of materials and manufacturing technologies in association with a wide spectrum of disciplines and their inter-twining to meet the challenging technology. The paper highlight the work and the approaches adopted for the successful deployment of materials, manufacturing and inspection technologies for the in-core and structural components of current and future Indian Fast Breeder Reactor Programme. Indigenous development of in-core materials viz. Titanium modified austenitic stainless steel (Alloy D9) and its variants, ferritic/martensitic oxide-dispersion strengthened (ODS) steels as well as structural materials viz. 316L(N) stainless steel and modified 9Cr-1Mo have been achieved through synergistic interactions between Indira Gandhi Centre for Atomic Research (IGCAR), education and research institutes and industries. Robust manufacturing technology has been established for forming and joining of various components of 500 MWe Prototype Fast Breeder Reactor (PFBR) through 'science-based technology' approach. To achieve the strict quality standards of formed parts in terms of geometrical tolerances, residual stresses and microstructural defects, FEM-based modelling and experimental validation was carried out for estimation of spring-back during forming of multiple curvature thick plantes. Optimization of grain boundary character distribution in Alloy D9 was carried out by adopting the grain boundary engineering approach to reduce radiation induced segregation. Extensive welding is involved in the fabrication of reactor vessels, piping, steam generators, fuel sub-assemblies etc. Activated Tungsten Inert Gas Welding process along with activated flux developed at IGCAR has been successfully used in fabrication of dummy fuel subassemblies (DFSA) required

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

    Indian Academy of Sciences (India)

    The challenges and achievements in science and technology of FBRs focusing on safety are described with the particular reference to 500 MWe capacity Prototype Fast Breeder Reactor (PFBR), being commissioned at Kalpakkam. Roadmap with comprehensive R&D for the large scale deployment of Sodium Cooled Fast ...

  6. Engineering aspects of heterogeneous and homogeneous reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dickson, Jr, P W

    1977-01-01

    The core arrangement in an LMFBR can potentially affect the lower internals, upper internals, radial shielding, vessel, hot leg transients, head access area and control systems as well as breeding ratio, doubling time and core inventory. This paper describes the results of a study of the impact on these components and parameters that would result if the Clinch River Breeder Reactor were to incorporate a heterogeneous core.

  7. Demonstration of solar hydrogen production from water splitting via monolithic honeycomb reactors in a 100-kW-scale pilot plant

    Energy Technology Data Exchange (ETDEWEB)

    Agrafiotis, C.; Konstandopoulos, A.G.; Lorentzou, S. [Aerosol and Particle Technology Laboratory (APTL) (GR)] (and others)

    2010-07-01

    A 100 kW pilot plant for two-step solar thermo-chemical water splitting via monolithic honeycomb solar reactors based on the HYDROSOL two-chamber solar receiver-reactor technology has been developed, installed and test operated at the SSPS solar tower plant at PSA in Spain. At first, thermal qualification of the pilot plant has been carried out using uncoated ceramic honeycombs as absorbers, in order to develop and validate operational and measurement strategy as well as to create essential knowledge on the dynamics of the system during thermal cycling. Potential control parameters capable of ensuring sufficient constant temperature levels have been analyzed. It was concluded that the preferred way of controlling the process temperature is by the heliostats themselves which can be applied to ensure sufficiently steady states and compensation against solar flux fluctuations. A number of hydrogen production cycles applying honeycombs coated with redox material in several experimental series have been successfully carried out demonstrating the feasibility of solar water splitting via the HYDROSOL technology in the 100 kW-scale pilot scale under real conditions. Significant concentrations of hydrogen were produced with a conversion of steam of up to 30 %. Pilot plant tests on optimizing the process conditions are being continued by using absorber monoliths with different coatings from on-going material development studies, to demonstrate the high potential of the particular technology for further scale-up. (orig.)

  8. Nuclear reactor design

    CERN Document Server

    2014-01-01

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

  9. Compatibility of sodium with ceramic oxides employed in nuclear reactors; Compatibilidad del sodio con oxidos ceramicos utilizados en reactores nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Acena Moreno, V.

    1981-07-01

    This work is a review of experiments carried out up to the present time on the corrosion and compatibility of ceramic oxides with liquid sodium at temperatures corresponding to those in fast breeder reactors. The review also includes the results of a thermo-dynamic/liquid sodium reactions. The exercise has been conducted with a view to effecting experimental studies in the future. (Author)

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

    Science.gov (United States)

    Lim, Yesul; Kim, Dong-Jin

    2014-08-01

    This study proposes a simplified method for the quantification of N2O emission from a biological nutrient removal wastewater treatment plant (WWTP). The method incorporates a laboratory-scale batch reactor which had almost the same operational (wastewater and sludge flow rates) condition of a unit operation/process of the WWTP. Cumulative N2O emissions from the batch reactor at the corresponding hydraulic retention times of the full-scale units (primary and secondary clarifiers, pre-anoxic, anaerobic, anoxic and aerobic basins) were used for the quantification of N2O emission. The analysis showed that the aerobic basin emitted 95% of the total emission and the emission factor (yield) reached 0.8% based on the influent nitrogen load. The method successfully estimated N2O emission from the WWTP and it has shown advantages in measurement time and cost over the direct field measurement (floating chamber) method. Copyright © 2014 Elsevier Ltd. All rights reserved.

  11. Advanced Safeguards Approaches for New Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-12-15

    This third report in the series reviews possible safeguards approaches for new fast reactors in general, and the ABR in particular. Fast-neutron spectrum reactors have been used since the early 1960s on an experimental and developmental level, generally with fertile blanket fuels to “breed” nuclear fuel such as plutonium. Whether the reactor is designed to breed plutonium, or transmute and “burn” actinides depends mainly on the design of the reactor neutron reflector and the whether the blanket fuel is “fertile” or suitable for transmutation. However, the safeguards issues are very similar, since they pertain mainly to the receipt, shipment and storage of fresh and spent plutonium and actinide-bearing “TRU”-fuel. For these reasons, the design of existing fast reactors and details concerning how they have been safeguarded were studied in developing advanced safeguards approaches for the new fast reactors. In this regard, the design of the Experimental Breeder Reactor-II “EBR-II” at the Idaho National Laboratory (INL) was of interest, because it was designed as a collocated fast reactor with a pyrometallurgical reprocessing and fuel fabrication line – a design option being considered for the ABR. Similarly, the design of the Fast Flux Facility (FFTF) on the Hanford Site was studied, because it was a successful prototype fast reactor that ran for two decades to evaluate fuels and the design for commercial-scale fast reactors.

  12. Environmental Enrichment for Broiler Breeders: An Undeveloped Field

    DEFF Research Database (Denmark)

    Riber, Anja Brinch; Jong, Ingrid de; van de Werd, Heleen A.

    2017-01-01

    of environmental enrichment for broiler breeders, which have been described in the scientific literature, on behavior and welfare. Environmental enrichment is defined as an improvement of the environment of captive animals, which increases the behavioral opportunities of the animal and leads to improvements...... in biological function. This definition has been broadened to include practical and economic aspects as any enrichment strategy that adversely affects the health of animals (e.g., environmental hygiene), or that has too many economic or practical constraints will never be implemented on commercial farms...... and thus never benefit animals. Environmental enrichment for broiler breeders often has the purpose of satisfying the behavioral motivations for feeding and foraging, resting, and/or encouraging normal sexual behavior. Potentially successful enrichments for broiler breeders are elevated resting places...

  13. Analysis of gamma ray intensity on the S/C vent pipes area in the unit 2 reactor building of the Fukushima Daiichi Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jai Wan; Choi, Young Soo; Jeong, Kyung Min [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The robot is equipped with cameras, a dosimeter, and 2 DOF (degree of freedom) manipulation arms. It loads a small vehicle equipped with a camera that can access and inspect narrow areas. TEPCO is using the four-legged walking robot to inspect the suppression chamber (S/C) area of the unit 2 reactor building basement in the Fukushima Daiichi Nuclear Power Plant. The robot carried out 6 missions for about four months, from 11 December, 2012 to 15 March, 2013, where it examined an evidence of a leakage of radioactivity contaminated water in the S/C area of unit 2 reactor building. When a camera's signal processing unit, which is consist of ASIC and FPGA devices manufactured by a CMOS fabrication process, is exposed to a higher dose rate gamma ray, the speckle distribution in the camera image increase more. From the inspection videos, released by TEPCO, of the underground 8 vent pipes in the unit 2 reactor building, we analyzed the speckle distribution from the high dose-rate gamma rays. Based on the distribution of the speckle, we attempted to characterize the vent pipe with much radioactivity contaminated materials among the eight vent pipes connected to the PCV. The numbers of speckles viewed in the image of a CCD (or CMOS) camera are related to an intensity of the gamma ray energy emitted by a nuclear fission reaction from radioactivity materials. The numbers of speckles generated by gamma ray irradiation in the camera image are calculated by an image processing technique. Therefore, calculating the speckles counts, we can determine the vent pipe with relatively most radioactivity-contaminated materials among the other vent pipes. From the comparison of speckles counts calculated in the inspection image of the vent pipe with the speckles counts extracted by gamma ray irradiation experiment of the same small vehicle camera model loaded with the four-legged walking robot, we can qualitatively estimate the gamma ray dose-rate in the S/C vent pipe area of the

  14. 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...

  15. Prevalence of Campylobacter jejuni in poultry breeder flocks

    Directory of Open Access Journals (Sweden)

    Ludovico Dipineto

    2010-01-01

    Full Text Available The aim of this work is to present the preliminary results of a study about the prevalence of Campylobacter jejuni in poultry breeder flocks. It was examined three different breeder flocks of Bojano in Molise region. A total of 360 cloacal swabs and 80 enviromental swabs was collected. Of the 3 flocks studied, 6.9% tested were positive for Campylobacter spp. The most-prevalent isolated species is C. jejuni (8.2%. Only 3 of the 360 cloacal swabs samples examined were associated with C. coli. The environmental swabs resulted negative. This results confirms again that poultry is a reservoir of this germ.

  16. Application OF LIBS To Estimate The Age Of Broiler Breeders

    Science.gov (United States)

    Salam, Z. Abdel; Harith, M. A.

    2011-09-01

    Laser Induced Breakdown Spectroscopy (LIBS) is a well-known spectrochemical elemental analysis technique. In our investigations of the LIBS spectra it has been found that there is a remarkable correlation between the ionic to atomic spectral lines emission ratio and the surface hardness of eggshell for two Different Broiler Breeder at different age. The proposed technique has been applied successfully in poultry science to estimate the age of broiler breeders by measuring the surface hardness of their eggshell. The experiments have been performed on two different strains, Arbor Acres plus (AAP) and Hubard Classic (HC), and the results were satisfactory.

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

    Science.gov (United States)

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

    1998-10-01

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

  18. Simulation and design of an electron beam ion source charge breeder for the californium rare isotope breeder upgrade

    Directory of Open Access Journals (Sweden)

    Clayton Dickerson

    2013-02-01

    Full Text Available An electron beam ion source (EBIS will be constructed and used to charge breed ions from the californium rare isotope breeder upgrade (CARIBU for postacceleration into the Argonne tandem linear accelerator system (ATLAS. Simulations of the EBIS charge breeder performance and the related ion transport systems are reported. Propagation of the electron beam through the EBIS was verified, and the anticipated incident power density within the electron collector was identified. The full normalized acceptance of the charge breeder with a 2 A electron beam, 0.024π  mm mrad for nominal operating parameters, was determined by simulating ion injection into the EBIS. The optics of the ion transport lines were carefully optimized to achieve well-matched ion injection, to minimize emittance growth of the injected and extracted ion beams, and to enable adequate testing of the charge bred ions prior to installation in ATLAS.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-02-01

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

  20. 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.

  1. Calibration of the hydraulic model of a full-scale activated sludge plant; Calibracion hidraulica a escala real de un reactor de lodos activados

    Energy Technology Data Exchange (ETDEWEB)

    Fall, Cheikh [Universidad Autonoma del Estado de Mexico (Mexico); Loaiza-Navia, Jimmy [Servicios de Agua y Drenaje de Monterrey (Mexico)

    2008-04-15

    When planning to simulate a wastewater treatment plant (WWTP) with the activated sludge model number 1 (ASM1), one of the first requirements is to determine the hydraulic model of the reactor. The aim of this study was to evaluate the hydrodynamic regime of the aeration tank of a municipal WWTP by using a rhodamine tracer test and the Aquasim simulation software. A pre-simulation was performed in order to quantify the appropriate colorant mass, set up a sampling plan and evaluate the anticipated visual impact of the tracer test in the river receiving the treated effluents. A tracer test and dynamic flow measurements were carried out, the results of which served to establish and calibrate the hydraulic model. The evaluated tank was physically built as a plug-flow reactor subdivided in 7 compartments, but the study revealed that it is best represented by a model with 5 virtual mixed reactors in series. Through the study, the approach of using a WWTP simulator for hydraulics calibration was shown to be a powerful and flexible tool for designing a tracer test and for identifying adequate tank-in-series models of full-scale activated sludge aeration tanks. [Spanish] Cuando se planea simular una planta de tratamiento con base en el modelo numero 1 de lodos activados (ASM1), uno de los primeros requisitos es determinar el modelo hidraulico del reactor. En este trabajo se estudio el regimen hidrodinamico del tanque de accion de una planta de tratamiento de aguas residuales municipales (PTAR), utilizando una prueba de trazador con rodamina y un programa de simulacion (Aquasim). Se realizo una prueba de trazador con el experimento, lo que permitio determinar la cantidad requerida de trazador, fijar los intervalos de muestreo y limitar el impacto visual anticipado del colorante sobre el rio que recibe el efluente tratado. Se llevaron a cabo la prueba de trazador y la medicion de los perfiles dinamicos de caudales, cuyos resultados sirvieron para establecer y calibrar el

  2. APPLICATIONS OF LASERS AND OTHER TOPICS IN LASER PHYSICS AND TECHNOLOGY: Hybrid reactor based on laser thermonuclear fusion

    Science.gov (United States)

    Basov, N. G.; Belousov, N. I.; Grishunin, P. A.; Kalmykov, Yu K.; Lebo, I. G.; Rozanov, Vladislav B.; Sklizkov, G. V.; Subbotin, V. I.; Finkel'shteĭn, K. I.; Kharitonov, V. V.; Sherstnev, K. B.

    1987-10-01

    A physicotechnical and parametric analysis is used as the basis for a conceptual design of a thermonuclear inertial-confinement hybrid reactor as a breeder of fuel for fission nuclear power stations. It is proposed to use a laser as a driver in this reactor.

  3. Quantitative feed restriction of Pekin breeder ducks during the ...

    African Journals Online (AJOL)

    Quantitative feed restriction of Pekin breeder ducks during the rearing period and its effect on subsequent productivity. M.D. Olver. Animal and Dairy Science Research Institute, Irene. Six male and 24 female Pekin ducks per pen were randomly allocated to 12pens to test the effects of quantitative feed restriction during the ...

  4. Semen bacterial flora of Rhode Island Breeder cocks in Zaria ...

    African Journals Online (AJOL)

    The semen used in this study was collected from 77 Rhode Island Breeder cocks reared in battery cages under intensive management from a private farm in Zaria, Kaduna State, Nigeria using the back massage procedure, 27 of the 77 semen samples (35.1%) contained bacterial isolates. None of the samples grew fungi.

  5. The response of broiler breeder hens to dietary balanced protein ...

    African Journals Online (AJOL)

    Two basal feeds (118 and 175 g protein/kg) with similar balanced amino acid mixtures were appropriately blended to produce six experimental diets differing in protein. These were fed for six weeks to 180 broiler breeder hens (Ross 308) housed in individual cages from 26 w of age. A 13 h photoperiod was applied. Half the ...

  6. Environmental Enrichment for Broiler Breeders: An Undeveloped Field

    NARCIS (Netherlands)

    Riber, Anja B.; Jong, de Ingrid; Weerd, van de Heleen A.; Steenfeldt, Sanna

    2017-01-01

    Welfare problems, such as hunger, frustration, aggression, and abnormal sexual behavior, are commonly found in broiler breeder production. To prevent or reduce these welfare problems, it has been suggested to provide stimulating enriched environments. We review the effect of the different types of

  7. Impact of nutrition on welfare aspects of broiler breeder flocks

    NARCIS (Netherlands)

    Krimpen, van M.M.; Jong, de I.C.

    2014-01-01

    To ensure health and reproductive performance, broiler breeders are feed restricted during the rearing period and, to a lesser extent, during the production period. Although restricted feeding improves health and bird welfare, on the other hand the birds are chronically hungry and suffer from

  8. Feeding broiler breeder flocks in relation to bird welfare aspects

    NARCIS (Netherlands)

    Jong, de I.C.; Krimpen, van M.M.

    2011-01-01

    To ensure health and reproductive capacity of the birds, broiler breeders are fed restricted during the rearing period, and to a lesser extent also during the production period. Although restricted feeding improves health and thereby bird welfare, on the other hand the birds are chronically hungry

  9. Response of Japanese Breeder Quails to Varying Dietary Protein ...

    African Journals Online (AJOL)

    A three months feeding trial was conducted to assess the effects of feeding varying levels of dietary protein on productive performance, egg quality characteristics and hatchability using 930 six weeks breeders quails. The birds were randomly divided into five dietary treatment groups of 186 birds, and were further replicated ...

  10. Feeding broiler breeders to improve their welfare whilst maintaining productivity

    DEFF Research Database (Denmark)

    Steenfeldt, Sanna; Nielsen, Birte Lindstrøm

    of their litter may have affected their thermoregulation. This experiment indicates that high fibre diets can alleviate the feeling of hunger currently experienced by broiler breeders, and a high ratio of insoluble fibre can reduce stereotypies and may improve the well-being of the birds....

  11. Utilisation of synthetic amino acids by broiler breeder hens | Nonis ...

    African Journals Online (AJOL)

    This study was conducted to examine the response of broiler breeder hens to feeds supplemented with synthetic lysine and methionine when fed once or twice daily during peak production. Replacing intact protein with increasing amounts of free lysine and methionine, up to 2.3 g/kg feed, had no effect on feed intake, ...

  12. Vaccination Strategies in Breeder and Commercial Farms and ...

    African Journals Online (AJOL)

    In Nigeria infectious bursal disease (IBD) outbreaks have persisted despite routine vaccination. In a quest to determine some of the causes of the vaccination failures, the type of vaccines, vaccination schedules and seromonitoring for antibodies in breeder and commercial farms were investigated using structured ...

  13. Feeding Programs for Broiler Breeders in the Start Phase

    Directory of Open Access Journals (Sweden)

    J Tremarin

    Full Text Available ABSTRACT The fast-growing Brazilian aviculture requires studies to improve zootechnical performance indexes for broiler breeders. The purpose of this study was to assess different feeding programs for broiler breeders on performance and development of digestive organs. A total of 48,000, 1d-old, Cobb 500 broiler breeders were divided into two sheds with 24,000 birds each. The experiment was randomized in block design, considering each shed a block, with 4 treatments and 6 replications per treatment with 2,000 birds in each. Treatments consisted of: Shed 1 T1 = starter feed; T2 = T1 + probiotics; Shed 2 T3 = pre-starter feed; T4 = T3 + probiotics. The productive performance characteristics (bird weight, weight gain, feed intake and feed conversion, the development of digestive organs (gizzard, proventriculus, spleen, bursa of Fabricius and small bowel as well as the small bowel length were assessed weekly for all experimental groups, in samples of 10 birds per treatment. The best feed conversion and weight gain were observed with pre-starter feed in the first 7 days of age, with or without probiotic. Small bowel, gizzard and proventriculus development at 28 days was better for birds on pre-starter feed compared to those on starter feed alone. It is possible to conclude that broiler breeders on pre-starter feed during the first 7 days of age are likely to show better physical and productive performances in the adult phase.

  14. Plant breeding and rural development in the United States.

    Science.gov (United States)

    KE Woeste; SB Blanche; KA Moldenhauer; CD Nelson

    2010-01-01

    Plant breeders contributed enormously to the agricultural and economic development of the United States. By improving the profitability of farming, plant breeders improved the economic condition of farmers and contributed to the growth and structure of rural communities. In the years since World War II, agriculture and the quality of rural life have been driven by...

  15. Intellectual property rights, international trade and plant breeding

    NARCIS (Netherlands)

    Eaton, D.J.F.

    2013-01-01

    Seed is the physical embodiment of the invention of the plant breeder. Plant varieties thus constitute a special form of innovation, and an assessment of intellectual property right (IPR) systems needs to take this into account. This thesis concentrates on IPRs but breeders do have a number of means

  16. Fast neutron breeder reactor Rapsodie - situation of physics, hydraulic, thermal and dynamics studies and studies of stability early in 1963; Pile rapide rapsodie - point des etudes neutroniques, hydrauliques, thermiques et dynamiques et des etudes de stabilite au debut de l'annee 1963

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1964-07-01

    Early in 1963, it was necessary to make a choice among the two fuels examined for Rapsodie: the UPuMo alloy with double cladding, Nb and stainless steel, and the UO{sub 2}-PuO{sub 2} mix oxide. This report presents the results of the studies effected with the two types of fuel. We reconsider at first the different models which have been studied and we give a detailed description of the alloy and oxide cores as they are envisaged early in 1963. We give then the most important physics performances of the two cores: neutron flux and spectrum, reactivity of the compensation find safety rods, neutrons balance, specific power, effective fraction of delayed neutrons, lifetime of the prompt neutrons, reactivity coefficient. We describe the hydraulic studies and experiments which have been done concerning the two cores. We discuss the criteria adopted as basis for the flow calculations. We give the results of pressure drop and sub-assembly lifting, force measurements, and vibration and pin flow distribution experiments. We discuss the constants utilized for the thermal calculations and we give the temperatures of sodium and alloy or oxide fuel, the temperature increases due to the hot points, and the limitation of the oxide fuel burn-up, originated by the pressure of the fission gases. We treat the hypotheses having been utilized for the dynamics calculations and we describe the different accidents which have been studied. We give the results of the calculations for every accident and each fuel, and we show fuel melting or sodium boiling can be avoided, even in case of the most pessimistic hypotheses, by modifying reactor characteristics (shim-rod reactivity or power of the reactor with only one cooling circuit). The reactor stability has been evaluated with the hypotheses utilized for the dynamics calculations, except of the Doppler coefficient which was intentionally increased. We show that the alloy and oxide cores are stable for every envisaged reactor power. (authors

  17. Methods of assessing the effects of interface oxide growth in Magnox and advanced gas-cooled reactor plant

    Energy Technology Data Exchange (ETDEWEB)

    McLauchlin, I.R.; Wooton, M.R.; Morgan, J.D.; Watson, L.H.

    1986-10-01

    Growth of oxide at interfaces between structural steel components in CO/sub 2/-cooled reactors can deform fastenings such as bolts and welds. The mechanical response of joint members to oxide growth is discussed, and methods of assessment are outlined which contribute to procedures for ensuring continued structural integrity.

  18. Studies on serum macro and micro minerals status in repeat breeder ...

    African Journals Online (AJOL)

    ... 87%; whereas in repeat breeder control buffaloes, the overall pregnancy rate was 21%. In conclusion, the concentrations of macro and micro minerals were significantly lower in repeat breeder buffaloes and mineral mixtures should be added in the food stuff to improve reproductive efficiency of repeat breeder buffaloes.

  19. Economic and welfare benefits of environmental enrichment for broiler breeders.

    Science.gov (United States)

    Leone, E H; Estévez, I

    2008-01-01

    Designs to enrich the environment are crucial in the effort to fully address the biological needs of domestic animals. Although enrichment programs have been shown to improve health and welfare, little is known of their potential for application to commercial broiler breeder environments. We investigated the potential benefits of cover panels for broiler breeder reproductive performance in a commercial setting. This demonstration trial occurred on 5 commercial broiler breeder farms, each with a control and panel treatment room containing approximately 7,000 females and 800 males. Reproductive performance was measured from 25 to 60 wk by the number of eggs laid per female per week as well as weekly fertility and hatchability rates. The location of marked males was recorded weekly to quantify male movement. Access to cover panels improved egg production by 2.1% and maintained better hatchability and fertility throughout the breeding cycle (significant interactions of age and panel treatment) leading to an additional 4.5 chicks/female. Male home ranges, based on minimum convex polygons, were larger in the enriched (259 +/- 24.4 m(2)) vs. control flocks (184 +/- 23.1 m(2)). Providing enrichment in the form of cover panels improved reproductive performance, most likely by increasing males' mating opportunities and reducing female stress. We found a clear economic benefit to providing enrichment, an estimated $3 million if all breeder houses within the participating company were outfitted with the panels. These results demonstrate that environmental enrichment is not only beneficial for broiler breeder welfare, but can also be economically advantageous, resulting in a win-win situation for poultry welfare and production.

  20. CYCLOXYDIM-TOLERANT MAIZE – BREEDERS STANDPOINT

    Directory of Open Access Journals (Sweden)

    G. Bekavac

    2008-09-01

    Full Text Available Cycloxydim-Tolerant Maize (CTM was developed by researches at the University of Minnesota. CTM plants were regenerated from tissue culture selected for callus growth in the presence of cycloxydim, and the resulting plants were shown to contain a nuclear mutation, expressed as a single, partially dominant gene (known as Acc1 that conferred tolerance to the herbicide. Cycloxydim is a systemic herbicide for post emergence application in dicot crops to selectively control grass weeds. Corn, like most grasses is susceptible to cycloxydim due to inhibited acetyl-coenzyimeA carboxylaze enzyme activity. There are two key benefits of this technology: first, cycloxydim applications in CTM hybrids can be delayed until the weed spectrum and population density exceed agro-economic threshold; second, cycloxydim can be applied at either stage of plant development with no effect on basic agronomic traits, compared to non treated plants. Nevertheless, this type of tolerance requires 2 genes to be fully effective, i.e. gene must be present in both inbred parents to provide complete tolerance in the resulting hybrid. Such type of tolerance doubles the chances for yield drag and doubles the number of inbred conversions needed. This also limits germplasm integration and increases time lag in developing hybrids. Despite these difficulties, many seed companies introduce tolerance to cycloxydim into their commercial inbreds, and many of them have already commercialized CTM hybrids. Finally, it came as a logical question what is more important – hybrid performance or new trait? Critical to the success of this technology has been yield performance of CTM hybrids. At the same time, performance and herbicide tolerance do not exclude each other and can surely co-exist. To be accepted, this coexistence must secure high profitability to corn producers. However, CTM hybrids will not replace conventional ones on a large scale, but could be used as a specific tool, or could

  1. Development of heavy water reactor basic design technology. Development of plant-control simulation analyzer supporting design and operation of CANDU NPP

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sang Yong; Chun, Chong Sun; Cho, Chun Hye; Koo, Jun Moo; Chung, Dae Won; Sin, Ji Tae; Lee, Kyung Hoo [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of); Oh, Se Kee; Chung, Kun Moo; Kim, Seong Bae; Lim, Jae Chun; Kim, Hyun Dae; Park, Ji Won; Park, Jong Woun; Jung, Woo Sick; Cho, Seung Yun; Yeom, Choung Sup; Suh, Jae Seung; Park, Kyung Seok; Lee, Suk Hyung; Jung Ji Bum [Insititute for Advanced Engineering, (Korea, Republic of)

    1995-08-01

    Due to the rapid advances in computer technology, it becomes possible to use design codes rather than simplified models for training simulators, thereby allowing the convergence of training and engineering simulators. Such trend naturally leads to the idea of an integrated engineering design and analysis system based upon such simulators. For above purposes, A medium-fidelity simulator was developed for CANDU-6 including major process and control systems for normal operation. The display system of the simulator was developed with a customized SCADA GUI over a computer network. In particular, a multi-point 3D reactor model was developed which can be used as either a single point model for simple low-fidelity simulator an accurate high-fidelity model for reactor design. Though, the simulation program does not equipped with a good heat transfer model of two-phase flow, it was observed that the program can simulate plant states with good accuracy under normal and mild transient operations. Furthermore, synchronized multiple display system utilizing dynamic data link for plant monitoring and control was shown to be effective for simulating control room display environment. summarized. (author). 45 refs., 88 figs., 24 tabs.

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

    Science.gov (United States)

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

    2013-01-01

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

  3. Simplified method for measuring the response time of scram release electromagnet in a nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Patri, Sudheer, E-mail: patri@igcar.gov.in; Mohana, M.; Kameswari, K.; Kumar, S. Suresh; Narmadha, S.; Vijayshree, R.; Meikandamurthy, C.; Venkatesan, A.; Palanisami, K.; Murthy, D. Thirugnana; Babu, B.; Prakash, V.; Rajan, K.K.

    2015-04-15

    Highlights: • An alternative method for estimating the electromagnet clutch release time. • A systematic approach to develop a computer based measuring system. • Prototype tests on the measurement system. • Accuracy of the method is ±6% and repeatability error is within 2%. - Abstract: The delay time in electromagnet clutch release during a reactor trip (scram action) is an important safety parameter, having a bearing on the plant safety during various design basis events. Generally, it is measured using current decay characteristics of electromagnet coil and its energising circuit. A simplified method of measuring the same in a Sodium cooled fast reactors (SFR) is proposed in this paper. The method utilises the position data of control rod to estimate the delay time in electromagnet clutch release. A computer based real time measurement system for measuring the electromagnet clutch delay time is developed and qualified for retrofitting in prototype fast breeder reactor. Various stages involved in the development of the system are principle demonstration, experimental verification of hardware capabilities and prototype system testing. Tests on prototype system have demonstrated the satisfactory performance of the system with intended accuracy and repeatability.

  4. Reactor transient

    Energy Technology Data Exchange (ETDEWEB)

    Menegus, R.L.

    1956-05-31

    The authors are planning a calculation to be done on the Univac at the Louviers Building to estimate the effect of xenon transients, a high reactor power. This memorandum outlines the reasons why they prefer to do the work at Louviers rather than at another location, such as N.Y.U. They are to calculate the response of the reactor to a sudden change in position of the half rods. Qualitatively, the response will be a change in the rooftop ratio of the neutron flux. The rooftop ratio may oscillate with high damping, or, instead, it may oscillate for many cycles. It has not been possible for them to determine this response by hand calculation because of the complexity of the problem, and yet it is important for them to be certain that high power operation will not lead us to inherently unstable operation. Therefore they have resorted to machine computation. The system of differential equations that describes the response has seven dependent variables; therefore there are seven equations, each coupled with one or more of the others. The authors have discussed the problem with R.R. Haefner at the plant, and it is his opinion that the IBM 650 cannot adequately handle the system of seven equations because the characteristic time constants vary over a range of about 10{sup 8}. The Univac located at the Louviers Building is said to be satisfactory for this computation.

  5. Equipment specifications for an electrochemical fuel reprocessing plant

    Energy Technology Data Exchange (ETDEWEB)

    Hemphill, Kevin P [Los Alamos National Laboratory

    2010-01-01

    Electrochemical reprocessing is a technique used to chemically separate and dissolve the components of spent nuclear fuel, in order to produce new metal fuel. There are several different variations to electrochemical reprocessing. These variations are accounted for by both the production of different types of spent nuclear fuel, as well as different states and organizations doing research in the field. For this electrochemical reprocessing plant, the spent fuel will be in the metallurgical form, a product of fast breeder reactors, which are used in many nuclear power plants. The equipment line for this process is divided into two main categories, the fuel refining equipment and the fuel fabrication equipment. The fuel refining equipment is responsible for separating out the plutonium and uranium together, while getting rid of the minor transuranic elements and fission products. The fuel fabrication equipment will then convert this plutonium and uranium mixture into readily usable metal fuel.

  6. The Westinghouse Small Modular Reactor Design

    Energy Technology Data Exchange (ETDEWEB)

    Gutierrez Garcia, F. J.; Memmott, M. J.

    2012-07-01

    Westinghouse has developed a small modular reactor (SMR) which incorporates an integral pressurized water reactor (PWR) design. The Westinghouse SMR design also utilizes many of the key features and innovative concepts from the AP1000 plant, including passive safety systems.

  7. Updated neutronics analyses of a water cooled ceramic breeder blanket for the CFETR

    Science.gov (United States)

    Xiaokang, ZHANG; Songlin, LIU; Xia, LI; Qingjun, ZHU; Jia, LI

    2017-11-01

    The water cooled ceramic breeder (WCCB) blanket employing pressurized water as a coolant is one of the breeding blanket candidates for the China Fusion Engineering Test Reactor (CFETR). Some updating of neutronics analyses was needed, because there were changes in the neutronics performance of the blanket as several significant modifications and improvements have been adopted for the WCCB blanket, including the optimization of radial build-up and customized structure for each blanket module. A 22.5 degree toroidal symmetrical torus sector 3D neutronics model containing the updated design of the WCCB blanket modules was developed for the neutronics analyses. The tritium breeding capability, nuclear heating power, radiation damage, and decay heat were calculated by the MCNP and FISPACT code. The results show that the packing factor and 6Li enrichment of the breeder should both be no less than 0.8 to ensure tritium self-sufficiency. The nuclear heating power of the blanket under 200 MW fusion power reaches 201.23 MW. The displacement per atom per full power year (FPY) of the plasma-facing component and first wall reach 0.90 and 2.60, respectively. The peak H production rate reaches 150.79 appm/FPY and the peak He production reaches 29.09 appm/FPY in blanket module #3. The total decay heat of the blanket modules is 2.64 MW at 1 s after shutdown and the average decay heat density can reach 11.09 kW m-3 at that time. The decay heat density of the blanket modules slowly decreases to lower than 10 W m-3 in more than ten years.

  8. Optimization of mass-production conditions for tritium breeder pebbles based on slurry droplet wetting method

    Energy Technology Data Exchange (ETDEWEB)

    Park, Yi-Hyun, E-mail: yhpark@nfri.re.kr [National Fusion Research Institute, Daejeon (Korea, Republic of); Min, Kyung-Mi; Ahn, Mu-Young; Cho, Seungyon; Lee, Young-Min [National Fusion Research Institute, Daejeon (Korea, Republic of); Park, Sang-Jin; Danish, Rehan; Lim, Chul-Hwan; Jo, Yong-Dae [IVT Co., Ltd., Daegu (Korea, Republic of)

    2016-11-01

    Highlights: • An automatic dispensing system was developed to improve uniformity and production rate of breeder pebbles. • The production rate of this system for Li{sub 2}TiO{sub 3} pebble was estimated at 50 kg/year. • The optimization of dispensing and sintering conditions for the mass-production of Li{sub 2}TiO{sub 3} pebble was conducted. • Integrity of Li{sub 2}TiO{sub 3} pebble was able to be ensured during mass-production process, especially during batch process. - Abstract: Lithium metatitanate (Li{sub 2}TiO{sub 3}) is being considered as tritium breeding material for solid-type breeding blanket, which are used in pebble-bed form. The total amount of Li{sub 2}TiO{sub 3} pebbles in Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) is approximately 80 kg. Furthermore, DEMO reactor requires a great deal of breeder pebbles. Therefore, the development of mass-production system for breeder pebbles is necessary. The slurry droplet wetting method was adopted in the mass-production process for Li{sub 2}TiO{sub 3} pebbles, which had been developed in Korea. In this method, an automatic slurry dispensing system is one of the key apparatuses because the uniformity of pebbles and production rate are able to be improved. The system was successfully manufactured, which was consisted of a dispensing unit for instillation of Li{sub 2}TiO{sub 3} slurry, a glycerin bath for hardening of droplets, and an automatic maintaining unit for constant distance between syringe needle and glycerin surface. The production rate of this system for Li{sub 2}TiO{sub 3} pebble was estimated at 50 kg/year. In this study, it was investigated that the effect of dispensing and sintering conditions on the mass-production of Li{sub 2}TiO{sub 3} pebbles.

  9. An numerical analysis of high-temperature helium reactor power plant for co-production of hydrogen and electricity

    Science.gov (United States)

    Dudek, M.; Podsadna, J.; Jaszczur, M.

    2016-09-01

    In the present work, the feasibility of using a high temperature gas cooled nuclear reactor (HTR) for electricity generation and hydrogen production are analysed. The HTR is combined with a steam and a gas turbine, as well as with the system for heat delivery for medium temperature hydrogen production. Industrial-scale hydrogen production using copper-chlorine (Cu-Cl) thermochemical cycle is considered and compared with high temperature electrolysis. Presented cycle shows a very promising route for continuous, efficient, large-scale and environmentally benign hydrogen production without CO2 emissions. The results show that the integration of a high temperature helium reactor, with a combined cycle for electric power generation and hydrogen production, may reach very high efficiency and could possibly lead to a significant decrease of hydrogen production costs.

  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. Emergency Cooling of Nuclear Power Plant Reactors With Heat Removal By a Forced-Draft Cooling Tower

    Energy Technology Data Exchange (ETDEWEB)

    Murav’ev, V. P., E-mail: murval1@mail.ru

    2016-07-15

    The feasibility of heat removal during emergency cooling of a reactor by a forced-draft cooling tower with accumulation of the peak heat release in a volume of precooled water is evaluated. The advantages of a cooling tower over a spray cooling pond are demonstrated: it requires less space, consumes less material, employs shorter lines in the heat removal system, and provides considerably better protection of the environment from wetting by entrained moisture.

  12. Development of work, time and output utilisation of nuclear power plants with light water reactors in the world for the period 1968 to 1987 i. e. from the 1. to the 17. year of operation

    Energy Technology Data Exchange (ETDEWEB)

    Held, C.

    The cumulated work utilisation of light water reactors in the world rose as function of the calendar years from 56% in 1968 to 65% in 1986. Even more remarkable is the almost consistent value of 65% throughout 17 years of operation. In some countries values of over 80% are attained. Nuclear power plants with light water reactors could, as planned, be operated as basic output power plants with an output utilisation rising to over 90%. The cumulated work, time and output utilisation followed as function of the calendar year as well as operating year the law of the equation y = a + b . ln x.

  13. Plants Growth Rate in Evapotranspiration continuous system reactors as the 2nd Treatment at Anaerobic-evapotranspiration system with High Strength Ammonium in Leachate Influent

    Directory of Open Access Journals (Sweden)

    Badrus Zaman

    2014-05-01

    Full Text Available Ammonium is one of parameter which responsible to leachate toxicity. Preliminary research was shown that the Fimbristylis globulosa (water plant, Alocasia macrorrhiza (terrestrial plant and Eleusine indica (terrestrial grass were potential plants for used as object in evaporation reactor system with high strength ammonium  concentration in leachate treatment. This research was integrated of anaerobic system with evapotranspiration system with continuous influent using ammonium concentration in leachate was 2000 mg/l NH4-N. Plants growth rate was analyzed for 25 days operated. The result shown that average of thallus growth rate of Fimbristylis globulosa was 17,5 cm d-1. The average of leaf and thallus growth rate of Alocasia macrorrhiza was 18,1 cm d-1 and 3,2 cm d-1 respectively. The average of blade and thallus of Eleusine indica were same that was 4,7 cm d-1.This research conclude that integration system of anaerobic and evpotranspiration was be potential used for high strength ammonium in leachate treatment.

  14. Fast reactor safety and related physics. Volume I. Invited papers; panels; summary

    Energy Technology Data Exchange (ETDEWEB)

    1976-01-01

    Separate abstracts were prepared for each of the twenty invited papers included. The papers covered sessions on licensing aspects of safety design bases, safety of demonstration plants, safety aspects of large commercial fast breeders, and safety test facilities.

  15. 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

  16. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Moisseytsev, A.; Sienicki, J. J. (Nuclear Engineering Division)

    2012-05-10

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

  18. The Effect of Degree of Photostimulation on Male Broiler Breeder ...

    African Journals Online (AJOL)

    Nicky

    Abstract. The effect of photostimulation of male broiler breeders (n = 144) to different photoperiods (8-h control and 9, 9.5, 10, 10.5, 11, 11.5, 12, 12.5, 13, 14 and 18 h) applied at 20 weeks of age, on age at first semen production, testis weights, as predicted by comb area, and semen characteristics at the reported age at first.

  19. Function analysis of nuclear power plants for developing of man-machine interface system for Korean next generation reactor

    Energy Technology Data Exchange (ETDEWEB)

    Goo, In Soo; Kim, Jang Yyul; Kim, Jung Soo; Kim, Chang Hoi; Na, Nan Joo; Park, Keun Ok; Park, Won Man; Park, Jae Chang; Suh, Sang Moon; Oh, In Suk; Lee, Dong Young; Lee, Yong Hee; Cha, Kyung Ho; Chun, Se Woo; Hur, Sup [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Jung, Kyung Hoon [Korea Electric Power Co., Seoul (Korea, Republic of); Park, Yeon Sik; Lee, Bum Joo [Korea Power Engineering Company, Inc., Seoul (Korea, Republic of)

    1995-02-01

    In this study, we developed the methodology and implementation plant of function analysis and performed function analysis, which is one of the major activities for the development of Man-Machine Interface System of the KNGR. Identifying the functions of existing plants followed by structuring the functions, we established functions structured at the conceptual and middle levels. This structure was further checked if it would satisfy regulatory requirements and developed to include the aspects of plant performance and other plant features to emphasize its practicality for the application to the design. (Author) 13 refs., 8 figs., 4 tabs.

  20. Methanogenesis in Thermophilic Biogas Reactors

    DEFF Research Database (Denmark)

    Ahring, Birgitte Kiær

    1995-01-01

    Methanogenesis in thermophilic biogas reactors fed with different wastes is examined. The specific methanogenic activity with acetate or hydrogen as substrate reflected the organic loading of the specific reactor examined. Increasing the loading of thermophilic reactors stabilized the process...... as indicated by a lower concentration of volatile fatty acids in the effluent from the reactors. The specific methanogenic activity in a thermophilic pilot-plant biogas reactor fed with a mixture of cow and pig manure reflected the stability of the reactor. The numbers of methanogens counted by the most...... against Methanothrix soehngenii or Methanothrix CALS-I in any of the thermophilic biogas reactors examined. Studies using 2-14C-labeled acetate showed that at high concentrations (more than approx. 1 mM) acetate was metabolized via the aceticlastic pathway, transforming the methyl-group of acetate...

  1. 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.

  2. Study of the pyrochemical treatment-recycling process of the Molten Salt Reactor fuel; Estudio de sistema de un proceso de tratamiento-reciclaje piroquimico del combustible de un reactor de sales fundidas

    Energy Technology Data Exchange (ETDEWEB)

    Boussier, H.; Heuer, D.

    2010-07-01

    The Separation Processes Studies Laboratory (Commissariat a l'energie Atomique) has made a preliminary assessment of the reprocessing system associated with Molten Salt Fast Reactor (MSFR). The scheme studied in this paper is based on the principle of reductive extraction and metal transfer that constituted the core process designed for the Molten Salt Breeder Reactor (MSBR), although the flow diagram has been adapted to the current needs of the Molten Salt Fast Reactor (MSFR).

  3. Fossil fuel furnace reactor

    Science.gov (United States)

    Parkinson, William J.

    1987-01-01

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

  4. Nuclear proliferation and civilian nuclear power: report of the Nonproliferation Alternative Systems Assessment Program. Volume IX. Reactor and fuel cycle descriptions

    Energy Technology Data Exchange (ETDEWEB)

    1979-12-01

    The Nonproliferation Alternative Systems Assessment Program (NASAP) has characterized and assessed various reactor/fuel-cycle systems. Volume IX provides, in summary form, the technical descriptions of the reactor/fuel-cycle systems studied. This includes the status of the system technology, as well as a discussion of the safety, environmental, and licensing needs from a technical perspective. This information was then used in developing the research, development, and demonstration (RD and D) program, including its cost and time frame, to advance the existing technology to the level needed for commercial use. Wherever possible, the cost data are given as ranges to reflect the uncertainties in the estimates. Volume IX is divided into three sections: Chapter 1, Reactor Systems; Chapter 2, Fuel-Cycle Systems; and the Appendixes. Chapter 1 contains the characterizations of the following 12 reactor types: light-water reactor; heavy-water reactor; water-cooled breeder reactor; high-temperature gas-cooled reactor; gas-cooled fast reactor; liquid-metal fast breeder reactor; spectral-shift-controlled reactor; accelerator-driven reactor; molten-salt reactor; gaseous-core reactor; tokamak fusion-fisson hybrid reactor; and fast mixed-spectrum reactor. Chapter 2 contains similar information developed for fuel-cycle facilities in the following categories: mining and milling; conversion and enrichment; fuel fabrication; spent fuel reprocessing; waste handling and disposal; and transportation of nuclear materials.

  5. CONVECTION REACTOR

    Science.gov (United States)

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

    1960-03-22

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

  6. Progress reports for Gen IV sodium fast reactor activities FY 2007.

    Energy Technology Data Exchange (ETDEWEB)

    Cahalan, J. E.; Tentner, A. M.; Nuclear Engineering Division

    2007-10-04

    identified as being for prevention of progression into severe accident conditions (prevention of core melting) or for mitigation of severe accident consequences (mitigation of the impact of core melting to protect public health and safety). Because design measures for severe accident prevention and mitigation are beyond the normal design basis, established regulatory guidelines and codes do not provide explicit identification of the design performance requirements for severe accident accommodation. The treatment of severe accidents is one of the key issues of R&D plans for the Gen IV systems in general, and for the Sodium Fast Reactor (SFR) in particular. Despite the lack of an unambiguous definition of safety approach applicable for severe accidents, there is an emerging consensus on the need for their consideration for the design. The US SFR program and Argonne National Laboratory (ANL) in particular have actively studied the potential scenarios and consequences of Hypothetical Core Disruptive Accidents (HCDA) for SFRs with oxide fuel during the Fast Flux Test Facility (FFTF) and Clinch River Breeder Reactor Plant (CRBRP) programs in the 70s and 80s. Later, the focus of the US SFR safety R&D activities shifted to the prevention of all HCDAs through passive safety features of the SFRs with metal fuel in the Integral Fast Reactor (IFR) program, and the study of severe accident consequences was de-emphasized. The goal of this paper is to provide an overview of the current SFR safety approach and the role of severe accidents in Japan and France, in preparation for an expected and more active collaboration in this area between the US, Japan, and France.

  7. Application of the SPA in the design of a hydrogen producer plant coupled to a nuclear reactor; Aplicacion del APS en el diseno de una planta productora de hidrogeno acoplada a un reactor nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Ruiz S, T.; Nelson, P. F.; Francois, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac No. 8532, Col. Progreso, 62550 Jiutepec, Morelos (Mexico); Cruz G, M. J., E-mail: truizsmx@yahoo.com.mx [UNAM, Facultad de Quimica, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)

    2013-10-15

    At the present time, one of the processes that is broadly investigated and that, theoretically demonstrates to be one of the most efficient for the hydrogen production, is the thermal-chemistry cycle Sulfur-Iodine (S-I) coupled to a nuclear reactor of very high temperature (VHTR). Because this chemical process of hydrogen production requires of a great inventory of toxic materials (sulphide compounds, hydriodic acid and iodine), is necessary the design of emergency systems with the purpose of protecting the facilities and the equipment s, the environment, as well as the near population. Given the impact of an accidental liberation of the process materials, as well as the proximity with the nuclear plant, is necessary that these emergency systems are the most reliable possible. This way, the results of the consequences analysis are utilized for the optimal localization of the gas sensors that activate the emergency systems, and the flows of the substances that are used for the leakage control. For all this, the use of the Safety Probabilistic Analysis methodology, as well as some standards of the nuclear industry, can be applied to the chemical installation to determine the fault sequences that can take to final states of not controlled leakage. This way, the use of methodologies of Event Tree Analysis and Fault Trees show in their results the components that but contribute in fault of such systems. In this work, is presented the evaluation of the joined models of event and fault trees and like with the obtained results, some proposals to increase the safety of the facilities are exposed. Also, the results of the evaluations of these proposals, and their impact of the probability of the not controlled fault sequences in a plant that is still in design stage are showed. (Author)

  8. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Smith, C

    2010-02-22

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

  10. Coupling of high temperature nuclear reactor with chemical plant by means of steam loop with heat pump

    Directory of Open Access Journals (Sweden)

    Kopeć Mariusz

    2017-01-01

    Full Text Available High temperature nuclear reactors (HTR can be used as an excellent, emission-free source of technological heat for various industrial applications. Their outlet helium temperature (700°-900°C allows not only for heat supply to all processes below 600°C (referred to as “steam class”, but also enables development of clean nuclear-assisted hydrogen production or coal liquefaction technologies with required temperatures up to 900°C (referred to as “chemical class”. This paper presents the results of analyses done for various configurations of the steam transport loop coupled with the high-temperature heat pump designed for “chemical class” applications. The advantages and disadvantages as well as the key issues are discussed in comparison with alternative solutions, trying to answer the question whether the system with the steam loop and the hightemperature heat pump is viable and economically justified.

  11. 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)

  12. Advanced reactor safety research quarterly report, October-December 1982. Volume 24

    Energy Technology Data Exchange (ETDEWEB)

    None

    1984-04-01

    This report describes progress in a number of activities dealing with current safety issues relevant to both light water reactors (LWRs) and breeder reactors. The work includes a broad range of experiments to simulate accidental conditions to provide the required data base to understand important accident sequences and to serve as a basis for development and verification of the complex computer simulation models and codes used in accident analysis and licensing reviews. Such a program must include the development of analytical models, verified by experiment, which can be used to predict reactor and safety system performance under a broad variety of abnormal conditions. Current major emphasis is focused on providing information to NRC relevant to (1) its deliberations and decisions dealing with severe LWR accidents and (2) its safety evaluation of the proposed Clinch River Breeder Reactor.

  13. Outline of the safety research results, in the power reactor field, fiscal year 1996

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-11-01

    The Power Reactor and Nuclear Fuel Development Corporation (PNC) has promoted the safety research in fiscal year of 1996 according to the Fundamental Research on Safety Research (fiscal year 1996 to 2000) prepared on March, 1996. Here is described on the research results in fiscal year 1996, the first year of the 5 years programme, and whole outline of the fundamental research on safety research, on the power reactor field (whole problems on the new nuclear converter and the fast breeder reactor field and problems relating to the power reactor in the safety for earthquake and probability theoretical safety evaluation field). (G.K.)

  14. Selection of Breeding Stock among Australian Purebred Dog Breeders, with Particular Emphasis on the Dam

    Directory of Open Access Journals (Sweden)

    Veronika Czerwinski

    2016-11-01

    Full Text Available Every year, thousands of purebred domestic dogs are bred by registered dog breeders. Yet, little is known about the rearing environment of these dogs, or the attitudes and priorities surrounding breeding practices of these dog breeders. The objective of this study was to explore some of the factors that dog breeders consider important for stock selection, with a particular emphasis on issues relating to the dam. Two-hundred and seventy-four Australian purebred dog breeders, covering 91 breeds across all Australian National Kennel Club breed groups, completed an online survey relating to breeding practices. Most breeders surveyed (76% reported specialising in one breed of dog, the median number of dogs and bitches per breeder was two and three respectively, and most breeders bred two litters or less a year. We identified four components, relating to the dam, that were considered important to breeders. These were defined as Maternal Care, Offspring Potential, Dam Temperament, and Dam Genetics and Health. Overall, differences were observed in attitudes and beliefs across these components, showing that there is variation according to breed/breed groups. In particular, the importance of Maternal Care varied according to dog breed group. Breeders of brachycephalic breeds tended to differ the most in relation to Offspring Potential and Dam Genetics and Health. The number of breeding dogs/bitches influenced breeding priority, especially in relation to Dam Temperament, however no effect was found relating to the number of puppies bred each year. Only 24% of breeders used their own sire for breeding. The finding that some breeders did not test for diseases relevant to their breed, such as hip dysplasia in Labrador Retrievers and German Shepherds, provides important information on the need to educate some breeders, and also buyers of purebred puppies, that screening for significant diseases should occur. Further research into the selection of breeding dams

  15. Selection of Breeding Stock among Australian Purebred Dog Breeders, with Particular Emphasis on the Dam.

    Science.gov (United States)

    Czerwinski, Veronika; McArthur, Michelle; Smith, Bradley; Hynd, Philip; Hazel, Susan

    2016-11-16

    Every year, thousands of purebred domestic dogs are bred by registered dog breeders. Yet, little is known about the rearing environment of these dogs, or the attitudes and priorities surrounding breeding practices of these dog breeders. The objective of this study was to explore some of the factors that dog breeders consider important for stock selection, with a particular emphasis on issues relating to the dam. Two-hundred and seventy-four Australian purebred dog breeders, covering 91 breeds across all Australian National Kennel Club breed groups, completed an online survey relating to breeding practices. Most breeders surveyed (76%) reported specialising in one breed of dog, the median number of dogs and bitches per breeder was two and three respectively, and most breeders bred two litters or less a year. We identified four components, relating to the dam, that were considered important to breeders. These were defined as Maternal Care, Offspring Potential, Dam Temperament, and Dam Genetics and Health. Overall, differences were observed in attitudes and beliefs across these components, showing that there is variation according to breed/breed groups. In particular, the importance of Maternal Care varied according to dog breed group. Breeders of brachycephalic breeds tended to differ the most in relation to Offspring Potential and Dam Genetics and Health. The number of breeding dogs/bitches influenced breeding priority, especially in relation to Dam Temperament, however no effect was found relating to the number of puppies bred each year. Only 24% of breeders used their own sire for breeding. The finding that some breeders did not test for diseases relevant to their breed, such as hip dysplasia in Labrador Retrievers and German Shepherds, provides important information on the need to educate some breeders, and also buyers of purebred puppies, that screening for significant diseases should occur. Further research into the selection of breeding dams and sires will

  16. Gas-cooled fast reactor program. Progress report, January 1, 1980-June 30, 1981

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.

    1981-09-01

    Since the national Gas-Cooled Fast Breeder Reactor Program has been terminated, this document is the last progress report until reinstatement. It is divided into three sections: Core Flow Test Loop, GCFR shielding and physics, and GCFR pressure vessel and closure studies. (DLC)

  17. Fabrication Technological Development of the Oxide Dispersion Strengthened Alloy MA957 for Fast Reactor Applications

    Energy Technology Data Exchange (ETDEWEB)

    Hamilton, Margaret L.; Gelles, David S.; Lobsinger, Ralph J.; Johnson, Gerald D.; Brown, W. F.; Paxton, Michael M.; Puigh, Raymond J.; Eiholzer, Cheryl R.; Martinez, C.; Blotter, M. A.

    2000-02-28

    A significant amount of effort has been devoted to determining the properties and understanding the behavior of the alloy MA957 to define its potential usefulness as a cladding material in the fast breeder reactor program. The numerous characterization and fabrication studies that were conducted are documented in this report.

  18. Engineering studies to support a liquid sodium cooled pebble bed target/blanket for an accelerator breeder

    Energy Technology Data Exchange (ETDEWEB)

    Cooper, K.C.; Malenfant, R.E.

    1977-01-01

    Preliminary calculations on using a liquid sodium cooled pebble bed of thorium metal for the target and blanket of a particle accelerator to convert fertile material to fissile have indicated the concept has considerable merit. Fuel management is simple and continuous on-line feed is effected. Additional features include excellent heat transfer, low pressure drop, minimization of structures and enhanced safety. Promising features strongly suggest that the work be continued to further refine the studies. Although specific to the target and blanket of the accelerator breeder concept, the work is readily extended to a high efficiency liquid metal cooled nuclear reactor employing a pebble bed as the fuel. 10 refs., 4 figs., 1 tab.

  19. Lay-out and materials for in pile tritium transport testing of breeder-inside-tube pin assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Alvani, C. [ENEA, CR Casaccia, Rome (Italy); Avon, J. [ENEA, CR Casaccia, Rome (Italy)]|[CEA, CEN, Grenoble (France); Casadio, S. [ENEA, CR Casaccia, Rome (Italy); Mancini, M.R. [ENEA, CR Casaccia, Rome (Italy); Nannetti, C.A. [ENEA, CR Casaccia, Rome (Italy); Pruzzo, G. [ENEA, CR Casaccia, Rome (Italy)]|[FN, Bosco Marengo-Alessandria (Italy); Ravel, S. [ENEA, CR Casaccia, Rome (Italy)]|[CEA, CEN, Grenoble (France); Roux, N. [ENEA, CR Casaccia, Rome (Italy)]|[CEA, CEN Saclay, Gif-sur-Yvette (France); Terlain, A. [ENEA, CR Casaccia, Rome (Italy)]|[CEA, CEN, Fontenay-aux-Roses (France); Terrosi [ENEA, CR Casaccia, Rome (Italy)]|[FN, Bosco Marengo-Alessandria (Italy); Zaghini, P. [ENEA, CR Casaccia, Rome (Italy)]|[FIN-Ceramica, Faenza, Ravenna (Italy); Zanardi, P. [ENEA, CR Casaccia, Rome (Italy)]|[ENEA, CR Clementel-Bologna (Italy); Zanotti, M. [ENEA, CR Casaccia, Rome (Italy)]|[ENEA, CR Clementel-Bologna (Italy)

    1995-12-31

    An irradiation experiment (90 FPD in SILOE reactor) has been designed in order to evaluate the in-situ effect of red-ox power of sweeping gas (helium with 100 vpm of H{sub 2}/H{sub 2}O with relative concentrations varying from pure H{sub 2} to pure H{sub 2}O) on (a) Tritium removal from LiAlO{sub 2} and Li{sub 2}ZrO{sub 3}; (b) Tritium permeation through AISI-316L SS tubes with bare and coated surfaces. The conditions and materials explored were selected in order to test possible improvements with respect to critical issues for the `Breeder Inside Tube` (BIT) blanket concept development. (orig.).

  20. Preliminary risk analysis of an Hydrogen production plant using the reformed process of methane with vapor coupled to a high temperature nuclear reactor; Analisis preliminar de riesgo de una planta de produccion de hidrogeno utilizando el proceso de reformado de metano con vapor acoplada a un reactor nuclear de alta temperatura

    Energy Technology Data Exchange (ETDEWEB)

    Flores y Flores, A. [Facultad de Ingenieria, UNAM, 04510 Mexico D.F. (Mexico)]. E-mail: alain_fyf@yahoo.com; Nelson E, P.F.; Francois L, J.L. [Facultad de Ingenieria, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, UNAM, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)

    2004-07-01

    It is necessary to identify the different types of dangers, as well as their causes, probabilities and consequences of the same ones, inside plants, industries and any process to classify the risks. This work is focused in particular to a study using the technical HAZOP (Hazard and Operability) for a plant of reformed of methane with vapor coupled to a nuclear reactor of the type HTTR (High Temperature Test Reactor), which is designed to be built in Japan. In particular in this study the interaction is analyzed between the nuclear reactor and the plant of reformed of methane with vapor. After knowing the possible causes of risk one it is built chart of results of HAZOP to have a better vision of the consequences of this faults toward the buildings and constructions, to people and the influence of the fault on each plant; for what there are proposed solutions to mitigate these consequences or to avoid them. The work is divided in three sections: a brief introduction about the technique of HAZOP; some important aspects of the plant of reformed of methane with vapor; and the construction of the chart of results of HAZOP. (Author)

  1. Ochratoxicosis in White Leghorn breeder hens: Production and breeding performance

    Directory of Open Access Journals (Sweden)

    Zahoor Ul Hassan*, Muhammad Zargham Khan, Ahrar Khan, Ijaz Javed1, Umer Sadique2 and Aisha Khatoon

    2012-10-01

    Full Text Available This study was designed to evaluate the effect of Ochratoxin A (OTA upon production and breeding parameters in White Leghorn (WL breeder hens. For this purpose, 84 WL breeder hens were divided into seven groups (A-G. The hens in these groups were maintained on feed contaminated with OTA @ 0.0 (control, 0.1, 0.5, 1.0, 3.0, 5.0 and 10.0 mg/Kg, respectively for 21 days. These hens were artificially inseminated with semen obtained from healthy roosters kept on OTA free feed. Egg production and their quality parameters were recorded. Fertile eggs obtained from each group were set for incubation on weekly basis. At the end of the experiment, hens in each group were killed to determined gross and microscopic lesions in different organs. OTA residue concentrations were determined in extracts of liver, kidneys and breast muscles by immunoaffinity column elution and HPLC-Fluorescent detection techniques. Feeing OTA contaminated diet resulted in a significant decrease in egg mass and egg quality parameters. Liver and kidneys showed characteristic lesions of ochratoxicosis. Residue concentration (ng/g of OTA in the hens fed 10 mg/kg OTA, was the highest in liver (26.336±1.16 followed by kidney (8.223±0.85 and were least in breast muscles (1.235±0.21. Embryonic mortalites were higher, while hatachabilites of the chicks were lower in the groups fed higher doses of OTA. Feeding OTA contaminated diets to breeder hen resulted in residues accumulation in their tissues along with significantly reduced production and breeding performance.

  2. Campylobacter epidemiology from breeders to their progeny in Eastern Spain.

    Science.gov (United States)

    Ingresa-Capaccioni, S; Jiménez-Trigos, E; Marco-Jiménez, F; Catalá, P; Vega, S; Marin, C

    2016-03-01

    While horizontal transmission is a route clearly linked to the spread of Campylobacter at the farm level, few studies support the transmission of Campylobacter spp. from breeder flocks to their offspring. Thus, the present study was carried out to investigate the possibility of vertical transmission. Breeders were monitored from the time of housing day-old chicks, then throughout the laying period (0 to 60 wk) and throughout their progeny (broiler fattening, 1 to 42 d) until slaughter. All samples were analyzed according with official method ISO 10272:2006. Results revealed that on breeder farms, Campylobacter isolation started from wk 16 and reached its peak at wk 26, with 57.0% and 93.2% of positive birds, respectively. After this point, the rate of positive birds decreased slightly to 86.0% at 60 wk. However, in broiler production all day-old chicks were found negative for Campylobacter spp, and the bacteria was first isolated at d 14 of age (5.0%), with a significant increase in detection during the fattening period with 62% of Campylobacter positive animals at the end of the production cycle. Moreover, non-positive sample was determined from environmental sources. These results could be explained because Campylobacter may be in a low concentration or in a non-culturable form, as there were several studies that successfully detected Campylobacter DNA, but failed to culture. This form can survive in the environment and infect successive flocks; consequently, further studies are needed to develop more modern, practical, cost-effective and suitable techniques for routine diagnosis. © 2016 Poultry Science Association Inc.

  3. Selection for early response to photostimulation in broiler breeders.

    Science.gov (United States)

    Tyler, N C; Gous, R M

    2011-08-01

    1. To determine if selection for early response to photostimulation could be successful, 150 male broiler breeders were photostimulated at 8 weeks of age. The first 20 to produce a semen sample and have a reddened comb with an area > 10 cm2 were selected as responders (R) and 20 birds that did not show these signs of sexual development were chosen as non-responders (NR). Once sexually mature, 8 birds from each group that consistently produced a semen sample were mated with both egg-type hybrids and broiler breeder females to observe the response to 8-week photostimulation in the as-hatched offspring. 2. The AFE of the F1 females with NR or R paternity and egg-type hybrid layer maternity (F1L) were similar, but AFE was advanced in birds from R relative to NR paternity when they had broiler breeder maternity (F1B). 3. Date following a normal distribution of AFE were extracted from the overall data set. This group included offspring from both NR and R paternity, but AFE in F1L and F1B females with R paternity was advanced compared to those with NR paternity. 4. Mean testis weights, or age at most rapid testis growth predicted using parameters from Tyler and Gous (2009), of F1 males were not significantly different in birds with NR or R paternity. A strong correlation was found between predicted age at most rapid growth and AFE of full sibs and so it is likely that an advance in AFE in female offspring would also result in an advance in age of testis development of males. 5. There was no significant difference in 21-d body weight of F1B females of NR or R paternity, but the 21-d body weights of F1L females were higher from R than from NR sires, suggesting that although fertility and meat-type traits are often negatively correlated, there was no adverse effect of selection for responsiveness to early photostimulation and broiler growth rates to 21 d. 6. These findings showed that the response to early stimulation is heritable, and should be useful to the broiler breeder

  4. Uterine adenocarcinoma with transcoelomic metastases in breeder hens (Gallus domesticus

    Directory of Open Access Journals (Sweden)

    D. G. Bwala

    2011-04-01

    Full Text Available Hens involved in a Newcastle disease study were euthanased at regular intervals according to a designed protocol. Of these, 7.14%(n = 42 of the 82-week-old specific pathogen-free breeder hens were found to have well-delineated firm white to yellowish nodules of varying sizes in the abdominal cavity. Histologically, the nodules were identified as an adenocarcinoma originating in the uterus. Transcoelomic spread was evidenced by the presence of similar neoplastic cells embedded in the serosa and outer longitudinal muscle layer of the intestines as well as the liver.

  5. REKONFIGURASI JARING DISTRIBUSI TENAGA LISTRIK MENGGUNAKAN BREEDER GENETIC ALGORITHM (BGA

    Directory of Open Access Journals (Sweden)

    Cok. Gede Indra Partha

    2009-05-01

    Full Text Available Kebutuhan akan daya listrik saat ini semakin meningkat, seiring dengan perkembangan teknologi, cara hidup, kebutuhan dan budaya di daerah tersebut. Untuk itu keandalan dan kontinuitas pelayanan, sistem transmisidan distribusi perlu ditingkatkan untuk memperoleh pelayanan yang optimal dengan losses terendah. Padapenelitian ini digunakan metode Breeder Algoritma Genetika (BGA yang telah dikembangkan dalam optimasibeban seimbang untuk rekonfigurasi jaring distribusi tegangan menengah (JTM. Proses optimasi beban dilakukandengan cara merubah switch-switch pada penyulang (sebagai gen-gen dalam kromosom jaring distribusi sehinggadiperoleh jaring distribusi yang paling optimal. Hasil analisis menggunakan BGA menunjukkan konfigurasi baruyang optimal dengan losses terendah serta lebih cepat konvergen jika dibandingkan dengan Genetic Algorithm(GA biasa.

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

    Science.gov (United States)

    Hartley, Ken; Lant, Paul

    2006-10-20

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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  8. Information system of corrosion and mechanical properties for steels used in nuclear power plants with PWR reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lahodova, M.; Novotny, R.; Sajdl, P. [Inst. of Chemical Technology, Prague (Czech Republic). Dept. of Power Engineering

    1998-11-01

    This paper gives information about a new developed database system which contains information about chemical constitution of steels used in nuclear power plants. It enables to hold data from corrosion tests and allows to insert graphs and pictures into the form. This system is an application of MS Access. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Changyong [State Key Laboratory of Environmental Criteria and Risk Assessment, Chinese Research Academy of Environmental Sciences, Beijing 100012 (China); Research Center of Water Pollution Control Technology, Chinese Research Academy of Environment Sciences, Beijing 100012 (China); Zhou, Yuexi, E-mail: zhouyuexi@263.net [State Key Laboratory of Environmental Criteria and Risk Assessment, Chinese Research Academy of Environmental Sciences, Beijing 100012 (China); Research Center of Water Pollution Control Technology, Chinese Research Academy of Environment Sciences, Beijing 100012 (China); Sun, Qingliang [School of Chemical Engineering, Beijing University of Chemical Technology, Beijing 100029 (China); Fu, Liya [Research Center of Water Pollution Control Technology, Chinese Research Academy of Environment Sciences, Beijing 100012 (China); School of Environment, Tsinghua University, Beijing 100084 (China); Xi, Hongbo; Yu, Yin [State Key Laboratory of Environmental Criteria and Risk Assessment, Chinese Research Academy of Environmental Sciences, Beijing 100012 (China); Research Center of Water Pollution Control Technology, Chinese Research Academy of Environment Sciences, Beijing 100012 (China); Yu, Ruozhen [State Key Laboratory of Environmental Criteria and Risk Assessment, Chinese Research Academy of Environmental Sciences, Beijing 100012 (China)

    2016-05-15

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

  10. Replacement of the feedwater pipe system in reactor building outside containment at the nuclear power plant Philippsburg; Austausch der Speisewasserleitung im Reaktorgebaeude ausserhalb SHB im Kernkraftwerk Philippsburg I

    Energy Technology Data Exchange (ETDEWEB)

    Kessler, A. [Energie-Versorgung Schwaben AG, Stuttgart (Germany); Labes, M. [Siemens AG Unternehmensbereich KWU, Offenbach am Main (Germany); Schwenk, B. [Kernkraftwerk Philippsburg GmbH (Germany)

    1998-11-01

    After full replacement of the feedwater pipe system during the inspection period in 1997, combined with a modern materials, manufacturing and analysis concept, the entire pipe system of the water/steam cycle in the reactor building of KKP 1 now consists of high-toughness materials. The safety level of the entire plant has been increased by leaving aside postulation of F2 breaks in the reactor building and providing for protection against 0.1 leaks. Based on fluid-dynamic calculations for the cases of pump failure and pipe break, as well as pipe system calculations in 5 extensive calculation cycles, about 130 documents were filed for inspection and approval (excluding preliminary test documents on restraints). Points of main interest for safety analysis in this context were the optimised closing performance of the 3rd check valves and the integrity of the nozzle region at the RPV. (oirg./CB) [Deutsch] Durch den Restaustausch der Speisewasserleitungen in der Revision 1997, verbunden mit einem modernen Werkstoff-, Fertigungs- und Nachweiskonzept, sind im Reaktorgebaeude von KKP 1 in den Hauptleitungen des Wasser-Dampf-Kreislaufes nur noch hochzaehe Werkstoffe eingesetzt. Durch den Verzicht auf das Postulat von 2F-Bruechen im Reaktorgebaeude und durch die Auslegung gegen 0,1F-Lecks wird das Sicherheitsniveau der Anlage insgesamt gesteigert. Ausgehend von fluiddynamischen Berechnungen fuer Pumpenausfall und Rohrbruch sowie Rohrsystem-Berechnungen in 5 umfangreichen Berechnungskreisen wurden fuer die Genehmigung und Begutachtung ca. 130 Unterlagen (ohne Halterungs-Vorpruefunterlagen) eingereicht und vom Gutachter geprueft. Schwerpunkte der Nachweisfuehrung waren die Optimierung des Schliessverhaltens der 3. Rueckschlagarmaturen sowie der Integritaetsnachweis des RDB-Anschlusses. (orig./MM)

  11. Engineering development studies for molten-salt breeder reactor processing No. 22

    Energy Technology Data Exchange (ETDEWEB)

    Hightower, J.R. Jr. (comp.)

    1976-06-01

    Processing methods are being developed for use in a close-coupled facility for removing fission products, corrosion products, and fissile materials from the MSBR fuel. This report discusses the autoresistance heating for the continuous fluorinator, the metal transfer experiment, experiments for the salt-metal contactor, and fuel reconstitution. 10 fig. (DLC)

  12. AB INITIO STUDY OF ADVANCED METALLIC NUCLEAR FUELS FOR FAST BREEDER REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Landa, A; Soderlind, P; Grabowski, B; Turchi, P A; Ruban, A V; Vitos, L

    2012-04-23

    Density-functional formalism is applied to study the ground state properties of {gamma}-U-Zr and {gamma}-U-Mo solid solutions. Calculated heats of formation are compared with CALPHAD assessments. We discuss how the heat of formation in both alloys correlates with the charge transfer between the alloy components. The decomposition curves for {gamma}-based U-Zr and U-Mo solid solutions are derived from Ising-type Monte Carlo simulations. We explore the idea of stabilization of the {delta}-UZr{sub 2} compound against the {alpha}-Zr (hcp) structure due to increase of Zr d-band occupancy by the addition of U to Zr. We discuss how the specific behavior of the electronic density of states in the vicinity of the Fermi level promotes the stabilization of the U{sub 2}Mo compound. The mechanism of possible Am redistribution in the U-Zr and U-Mo fuels is also discussed.

  13. Seismic design technology for breeder reactor structures. Volume 4. Special topics in piping and equipment

    Energy Technology Data Exchange (ETDEWEB)

    Reddy, D.P.

    1983-04-01

    This volume is divided into five chapters: experimental verification of piping systems, analytical verification of piping restraint systems, seismic analysis techniques for piping systems with multisupport input, development of floor spectra from input response spectra, and seismic analysis procedures for in-core components. (DLC)

  14. Calculation of a materials relocation experiment simulating a core disruptive accident condition in fast breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sawada, T. [Tokyo Inst. of Tech. (Japan). Research Lab. for Nuclear Reactors; Ninokata, H. [Tokyo Inst. of Tech. (Japan). Research Lab. for Nuclear Reactors; Shimizu, A. [Tokyo Inst. of Tech. (Japan). Research Lab. for Nuclear Reactors

    1995-07-01

    This paper describes an interpretation of the SIMBATH (Simulationsexperimente in Brennelementattrappen mit Thermit) experiments that use the SIMMER-II code. A series of SIMBATH experiments has aimed at simulating fuel pin disintegration and following materials relocation in the test sections of a single pin to 37-pin bundles. In the calculation, three modifications were incorporated into the SIMMER-II code. With these modifications, the calculation showed good agreement with the experimental measurements with respect to the void region propagation in sodium flow and the molten materials relocation leading to flow blockage. A set of parametric calculations has clarified the range of applicability of parameters for materials relocation and flow blockage formation. The particle radius r{sub p} in blockage regions and the mutiplier for particle viscosity (PARVIS) are recommended to be r{sub p}>or{approx}1/2D{sub h} and 0.001Pas

  15. Seismic design technology for breeder reactor structures. Volume 1. Special topics in earthquake ground motion

    Energy Technology Data Exchange (ETDEWEB)

    Reddy, D.P.

    1983-04-01

    This report is divided into twelve chapters: seismic hazard analysis procedures, statistical and probabilistic considerations, vertical ground motion characteristics, vertical ground response spectrum shapes, effects of inclined rock strata on site response, correlation of ground response spectra with intensity, intensity attenuation relationships, peak ground acceleration in the very mean field, statistical analysis of response spectral amplitudes, contributions of body and surface waves, evaluation of ground motion characteristics, and design earthquake motions. (DLC)

  16. Experimental Breeder Reactor II (EBR-II) Fuel-Performance Test Facility (FPTF)

    Energy Technology Data Exchange (ETDEWEB)

    Pardini, J.A.; Brubaker, R.C.; Veith, D.J.; Giorgis, G.C.; Walker, D.E.; Seim, O.S.

    1982-01-01

    The Fuel-Performance Test Facility (FPTF) is the latest in a series of special EBR-II instrumented in-core test facilities. A flow control valve in the facility is programmed to vary the coolant flow, and thus the temperature, in an experimental-irradiation subassembly beneath it and coupled to it. In this way, thermal transients can be simulated in that subassembly without changing the temperatures in surrounding subassemblies. The FPTF also monitors sodium flow and temperature, and detects delayed neutrons in the sodium effluent from the experimental-irradiation subassembly beneath it. This facility also has an acoustical detector (high-temperature microphone) for detecting sodium boiling.

  17. Feasibility studies for production of {sup 89}Sr in the Fast Breeder Test Reactor (FBTR)

    Energy Technology Data Exchange (ETDEWEB)

    Saha, Debasish; Vithya, J.; Ashok Kumar, G.V.S.; Swaminathan, K.; Kumar, R.; Venkata Subramani, C.R.; Vasudeva Rao, P.R. [Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam (India). Fuel Chemistry Div.

    2013-07-01

    {sup 89}Sr, a pure beta emitter with half life of 50.53 d is used as its chloride solution for palliative care of bone metastases. This paper describes the feasibility studies that have been conducted at FBTR, IGCAR for production of this radionuclide using the {sup 89}Y(n, p){sup 89}Sr reaction. Yttria pellets were irradiated in a special subassembly at the core centre for a total of 73 d in two steps of 35 d and 38 d with a time gap of 38 d. The irradiated yttria target was dissolved in nitric acid and the bulk Y was separated by solvent extraction using the TBP-HNO{sub 3} complex. The {sup 89}Sr fraction was purified using the cation exchange resin DOWEX 50W x 8 (100-200 mesh size) from the other radioactive impurities seen. The eluted {sup 89}Sr fraction was assayed using a GM counting system. The {sup 89}Sr activity produced in 1 g of yttria pellet was found to be 19 mCi. (orig.)

  18. Thermal-performance study of liquid metal fast breeder reactor insulation

    Energy Technology Data Exchange (ETDEWEB)

    Shiu, Kelvin K.

    1980-09-01

    Three types of metallic thermal insulation were investigated analytically and experimentally: multilayer reflective plates, multilayer honeycomb composite, and multilayer screens. Each type was subjected to evacuated and nonevacuated conditions, where thermal measurements were made to determine thermal-physical characteristics. A variation of the separation distance between adjacent reflective plates of multilayer reflective plates and multilayer screen insulation was also experimentally studied to reveal its significance. One configuration of the multilayer screen insulation was further selected to be examined in sodium and sodium oxide environments. The emissivity of Type 304 stainless steel used in comprising the insulation was measured by employing infrared technology. A comprehensive model was developed to describe the different proposed types of thermal insulation. Various modes of heat transfer inherent in each type of insulation were addressed and their relative importance compared. Provision was also made in the model to allow accurate simulation of possible sodium and sodium oxide contamination of the insulation. The thermal-radiation contribution to heat transfer in the temperature range of interest for LMFBR's was found to be moderate, and the suppression of natural convection within the insulation was vital in preserving its insulating properties. Experimental data were compared with the model and other published results. Moreover, the three proposed test samples were assessed and compared under various conditions as viable LMFBR thermal insulations.

  19. Present status and trend of development of operation and maintenance techniques of fast breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ozawa, Kenji; Yamashita, Yoshioki; Sunaoshi, Hiroshi

    1987-10-01

    Recently, accompanying the conspicuous improvement of the reliability and economical efficiency of LWRs, the level of attaining the high order technology required for FBRs has become high. The largest subject of FBRs is to heighten the economical efficiency while ensuring the safety, and many technical developments such as the heightening of the performance of FBR fuel, the shortening of pipings, the omission of secondary system and the rationalization of containment vessels have been carried out. Also the improvement of the capacity factor including the shortening of regular inspection period is important in addition to the safe and stable operation of FBRs, and it is necessary to upgrade the operation and maintenance techniques. The features of the operation and maintenance techniques of FBRs, the concept of the protection with deep strata in the operation techniques, the upgrading of the operation procedure at the time of abnormality, the development of operation-assisting systems, the upgrading of the education and training on operation, the shortening of regular inspection period, the techniques for reducing radiation exposure, the elucidation of the behavior of corrosion products, the international cooperation and others are described. (Kako, I.).

  20. An axially multilayered low void worth liquid-metal fast breeder reactor core concept

    Energy Technology Data Exchange (ETDEWEB)

    Kamei, T.; Yamaoka, M. (Toshiba Corp., Nuclear Engineering Lab., 4-1 Ukishima-cho, Kawasaki-ku, Kawasaki-shi 210 (JP))

    1992-03-01

    A new core concept with a negative sodium void reactivity coefficient has evolved. The core is composed of two core layers in the axial direction. The core layers are separated by an internal blanket, the central region of which comprises a neutron-absorbing material such as boron carbide or tantalum. Consequently, the two core layers are completely decoupled as regards neutronics, leading to an effective increase in neutron leakage from the core region when sodium is voided. This design is expected to be free from the disadvantages of a large core radius, as seen in a conventional spoiled core such as a pancake core. In this paper the design is described in detail, and its application to a 300-MW (electronic) metal fuel core and to a 450-MW (electric) minor actinide burned core is given as an example.

  1. REACTOR COOLING

    Science.gov (United States)

    Quackenbush, C.F.

    1959-09-29

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

  2. Design and startup of a membrane-biological-reactor system at a Ford-engine plant for treating oily wastewater.

    Science.gov (United States)

    Kim, B R; Anderson, J E; Mueller, S A; Gaines, W A; Szafranski, M J; Bremmer, A L; Yarema, G J; Guciardo, C D; Linden, S; Doherty, T E

    2006-04-01

    A wastewater-treatment facility at Ford (Dearborn, Michigan) was recently upgraded from chemical de-emulsification to ultrafiltration (UF) followed by a membrane-biological reactor (MBR). This paper describes the design, startup, and initial operational performance of the facility. Primary findings are as follows: (1) the MBR proved resilient; (2) the MBR removed approximately 90% of chemical-oxygen demand (COD) after primary UF; (3) the removal of total Kjeldahl nitrogen by MBR appeared to be more sensitive to operating conditions than COD removal; (4) nitrification and denitrification were established in one month; (5) the MBR removed oil and grease and phenolics to below detection levels consistently, in contrast to widely fluctuating concentrations in the past; (6) permeate fluxes of the primary and MBR UF were adversely affected by inadvertent use of a silicone-based defoamer; and (7) zinc concentrations in the effluent increased, which might have been a result of ethylenediaminetetraacetic acid used in membrane washing solutions and/or might have been within typical concentration ranges.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-12-01

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

  4. ALGERIAN SHEEP ARE NONSEASONAL BREEDERS: "CLINICAL, CYTOLOGICAL AND HISTOLOGICAL STUDIES"

    Directory of Open Access Journals (Sweden)

    A NIAR

    2001-12-01

    Full Text Available 4377 pregnant ewes have been used in this study to realise a curve of lambing. This curve prove that the three most important breeds of Algerian sheep can breed all year round.                 To prove also that Algerian sheep are no  seasonal breeder, a cytological and histological study of ewes vaginal mucosa have been done. The exfoliative cytology over all the phases of the oestrous cycle and pregnancy consistent in appearance and appear to be influenced directly by the changes in endogenous ovarian steroid hormones. The histological features of the anterior vaginal epithelium of the ewe sampled all year round, before, during, after oestrus and at pregnancy are described. In the present study, we have never found an arrest of the cyclical genital activity of ewes at any moment of the year. This finding agree with results of the curve of lambing. We can conclude that Algerian ewes are nonseasonal breeders.

  5. Humans are not cooperative breeders but practice biocultural reproduction.

    Science.gov (United States)

    Bogin, Barry; Bragg, Jared; Kuzawa, Christopher

    2014-01-01

    Alloparental care and feeding of young is often called "cooperative breeding" and humans are increasingly described as being a cooperative breeding species. To critically evaluate whether the human offspring care system is best grouped with that of other cooperative breeders. (1) Review of the human system of offspring care in the light of definitions of cooperative, communal and social breeding; (2) re-analysis of human lifetime reproductive effort. Human reproduction and offspring care are distinct from other species because alloparental behaviour is defined culturally rather than by genetic kinship alone. This system allows local flexibility in provisioning strategies and ensures that care and resources often flow between unrelated individuals. This review proposes the term "biocultural reproduction" to describe this unique human reproductive system. In a re-analysis of human life history data, it is estimated that the intense alloparenting typical of human societies lowers the lifetime reproductive effort of individual women by 14-29% compared to expectations based upon other mammals. Humans are not cooperative breeders as classically defined; one effect of the unique strategy of human biocultural reproduction is a lowering of human lifetime reproductive effort, which could help explain lifespan extension.

  6. Reactor Materials Research

    Energy Technology Data Exchange (ETDEWEB)

    Van Walle, E

    2002-04-01

    The activities of SCK-CEN's Reactor Materials Research Department for 2001 are summarised. The objectives of the department are: (1) to evaluate the integrity and behaviour of structural materials used in nuclear power industry; (2) to conduct research to unravel and understand the parameters that determine the material behaviour under or after irradiation; (3) to contribute to the interpretation, the modelling of the material behaviour and to develop and assess strategies for optimum life management of nuclear power plant components. The programmes within the department are focussed on studies concerning (1) Irradiation Assisted Stress Corrosion Cracking (IASCC); (2) nuclear fuel; and (3) Reactor Pressure Vessel Steel.

  7. Analysis of the micro-structural damages by neutronic irradiation of the steel of reactor vessels of the nuclear power plant of Laguna Verde. Characterization of the design steel; Analisis de los danos micro-estructurales por irradiacion neutronica del acero de la vasija de los reactores de la Central Nuclear de Laguna Verde. Caracterizacion del acero de diseno

    Energy Technology Data Exchange (ETDEWEB)

    Moranchel y Rodriguez, M.; Garcia B, A. [IPN, Escuela Superior de Fisica y Matematicas, Departamento de Fisica, Av. Luis Enrique Erro s/n, Unidad Profesional Adolfo Lopez Mateos, Col. Lindavista, 07738 Mexico D. F. (Mexico); Longoria G, L. C., E-mail: mmoranchel@ipn.m [ININ, Direccion de Investigacion Cientifica, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2010-09-15

    The vessel of a nuclear reactor is one of the safety barriers more important in the design, construction and operation of the reactor. If the vessel results affected to the grade of to have fracture and/or cracks it is very probable the conclusion of their useful life in order to guarantee the nuclear safety and the radiological protection of the exposure occupational personnel, of the public and the environment avoiding the exposition to radioactive sources. The materials of the vessel of a nuclear reactor are exposed continually to the neutronic irradiation that generates the same nuclear reactor. The neutrons that impact to the vessel have the sufficient energy to penetrate certain depth in function of the energy of the incident neutron until reaching the repose or to be absorbed by some nucleus. In the course of their penetration, the neutrons interact with the nuclei, atoms, molecules and with the same crystalline nets of the vessel material producing vacuums, interstitial, precipitate and segregations among other defects that can modify the mechanical properties of the steel. The steel A533-B is the material with which is manufactured the vessel of the nuclear reactors of nuclear power plant of Laguna Verde, is an alloy that, among other components, it contains atoms of Ni that if they are segregated by the neutrons impact this would favor to the cracking of the same vessel. This work is part of an investigation to analyze the micro-structural damages of the reactor vessels of the nuclear power plant of Laguna Verde due to the neutronic irradiation which is exposed in a continuous way. We will show the characterization of the design steel of the vessel, what offers a comprehension about their chemical composition, the superficial topography and the crystalline nets of the steel A533-B. It will also allow analyze the existence of precipitates, segregates, the type of crystalline net and the distances inter-plains of the design steel of the vessel. (Author)

  8. Nuclear reactor for breeding U.sup.233

    Science.gov (United States)

    Bohanan, Charles S.; Jones, David H.; Raab, Jr., Harry F.; Radkowsky, Alvin

    1976-01-01

    A light-water-cooled nuclear reactor capable of breeding U.sup.233 for use in a light-water breeder reactor includes physically separated regions containing U.sup.235 fissile material and U.sup.238 fertile material and Th.sup.232 fertile material and Pu.sup.239 fissile material, if available. Preferably the U.sup.235 fissile material and U.sup.238 fertile material are contained in longitudinally movable seed regions and the Pu.sup.239 fissile material and Th.sup.232 fertile material are contained in blanket regions surrounding the seed regions.

  9. Ultrasonic meters in the feedwater flow to recover thermal power in the reactor of nuclear power plant of Laguna Verde U1 and U2; Medidores ultrasonicos en el flujo de agua de alimentacion para recuperar potencia termica en el reactor de la Central Nuclear Laguna Verde U1 and U2

    Energy Technology Data Exchange (ETDEWEB)

    Tijerina S, F. [CFE, Central Laguna Verde, Km. 42.5 Carretera Cardel-Nautla, Veracruz (Mexico)]. e-mail: francisco.tijerina@cfe.gob.mx

    2008-07-01

    The engineers in nuclear power plants BWRs and PWRs based on the development of the ultrasonic technology for the measurement of the mass, volumetric flow, density and temperature in fluids, have applied this technology in two primary targets approved by the NRC: the use for the recovery of thermal power in the reactor and/or to be able to realize an increase of thermal power licensed in a 2% (MUR) by 1OCFR50 Appendix K. The present article mentions the current problem in the measurement of the feedwater flow with Venturi meters, which affects that the thermal balance of reactor BWRs or PWRs this underestimated. One in broad strokes describes the application of the ultrasonic technology for the ultrasonic measurement in the flow of the feedwater system of the reactor and power to recover thermal power of the reactor. One is to the methodology developed in CFE for a calibration of the temperature transmitters of RTD's and the methodology for a calibration of the venturi flow transmitters using ultrasonic measurement. Are show the measurements in the feedwater of reactor of the temperature with RTD's and ultrasonic measurement, as well as the flow with the venturi and the ultrasonic measurement operating the reactor to the 100% of nominal thermal power, before and after the calibration of the temperature transmitters and flow. Finally, is a plan to be able to realize a recovery of thermal power of the reactor, showing as carrying out their estimations. As a result of the application of ultrasonic technology in the feedwater of reactor BWR-5 in Laguna Verde, in the Unit 1 cycle 13 it was recover an equivalent energy to a thermal power of 25 MWt in the reactor and an exit electrical power of 6 M We in the turbogenerator. Also in the Unit 2 cycle 10 it was recover an equivalent energy to a thermal power of 40 MWt in the reactor and an exit electrical power of 16 M We in the turbogenerator. (Author)

  10. Multifunctional reactors

    NARCIS (Netherlands)

    Westerterp, K.R.

    1992-01-01

    Multifunctional reactors are single pieces of equipment in which, besides the reaction, other functions are carried out simultaneously. The other functions can be a heat, mass or momentum transfer operation and even another reaction. Multifunctional reactors are not new, but they have received much

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

    Energy Technology Data Exchange (ETDEWEB)

    De Bruyn, D.; Engelen, J. [Belgian Nuclear Research Centre SCK CEN, Boeretang 200, 2400 Mol (Belgium); Ortega, A.; Aguado, M. P. [Empresarios Agrupados A.I.E., Magallanes 3, 28015 Madrid (Spain)

    2012-07-01

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

  12. Hydrodynamic study to the bioreactor at the Granollers wastewater plant; Estudio hidrodinamico del reactor biologia de la EDAR Granollers

    Energy Technology Data Exchange (ETDEWEB)

    Olivet Santana, D.; Valls Puig, J.; Gordillo Bolasell, M. A.; Sanchez Ferrer, A.; Freixo Rey, A.

    2002-07-01

    One of the most important parameters affecting the operation of bioreactors used in the wastewater treatment is their hydrodynamic behaviour. Usually, bioreactors are designed according to ideal flow models, concretely stirred tank and plug flow models. However, actual bioreactors show a different behaviour from that predicted from ideal modelling. This fact is due to the presence of multiphase systems (e. g. biomass, liquid phase, air, etc.) and the yield obtained can significantly differ from theoretical design values. In the present work, residence-time distribution (RTD) is used to study a plug-flow bioreactor, which is currently working at the plant placed in Granollers (Barcelona). (Author) 5 refs.

  13. Effects of breeder age, strain, and eggshell temperature on nutrient metabolism of broiler embryos

    NARCIS (Netherlands)

    Nangsuay, A.; Meijerhof, R.; Anker-Hensen, van den I.; Heetkamp, M.J.W.; Kemp, B.; Brand, van den H.

    2017-01-01

    Breeder age and broiler strain influence the availability of nutrients and oxygen through yolk size and eggshell conductance, and the effects of these egg characteristics on nutrient metabolism might be influenced by eggshell temperature (EST). This study aims to determine effects of breeder age,

  14. Egg fertility and hatchability in Avians broiler-breeder hens under ...

    African Journals Online (AJOL)

    Impact of temperature and humidity in different month of lay was evaluated on fertility and hatchability in broiler breeder hens reared in Sapele, Nigeria. Six million, six hundred and nineteen thousand, seven hundred and forty six eggs from flocks of Avians broiler-breeder hens reared between 2005 and 2006 in a farm ...

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

    Science.gov (United States)

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

    2012-06-01

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

  16. Development of Tritium Plant System for Fusion Reactors - Achievements in the 14-year US-Japan Collaboration -

    Science.gov (United States)

    Nishi, Masataka; Yamanishi, Toshihiko; Shu, Wataru

    Fuel processing technology and tritium safe-handling technology have been developed through US/DOE-JAERI collaboration from 1987 till 2001, and the technologies to construct the tritium plant system of ITER have been made currently available. This paper overviews the major achievements of this collaborative researches over fourteen years, which were performed mainly at the Tritium Systems Test Assembly (TSTA) of the Los Alamos National Laboratory (LANL). The tritium plant system consists mainly of a fuel processing system, which includes a fuel cycle system and a blanket tritium recovery system, and a tritium confinement/removal system. The fuel cycle system recovers fuel from plasma exhaust gas and recycles it. In the collaboration, major key components and subsystems were developed, and the performance of the integrated system was successfully demonstrated over its one-month operation in which plasma exhaust model gas was processed at a processing rate of up to 1/6 level of the ITER. The technological basis of the fuel cycle system was thus established. Blanket tritium recovery technology was also successfully demonstrated using the TSTA system. Through the successful safeoperation of the TSTA, reliability of tritium confinement/removal system was verified basically. In addition, much data to confirm or enhance safety were accumulated by experiments such as intentional tritium release in a large room. Furthermore,distribution of tritium contamination in the vacuum vessel of the TFTR, a large tokamak of the Princeton Plasma Physics Laboratory (PPPL), was investigated in this work.

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

    Science.gov (United States)

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

    2013-05-21

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

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

    Science.gov (United States)

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

    2009-04-01

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

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

    Science.gov (United States)

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

    2014-01-01

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

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

    Science.gov (United States)

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

    2011-10-01

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

  1. Impacts of breeder loss on social structure, reproduction and population growth in a social canid.

    Science.gov (United States)

    Borg, Bridget L; Brainerd, Scott M; Meier, Thomas J; Prugh, Laura R

    2015-01-01

    The importance of individuals to the dynamics of populations may depend on reproductive status, especially for species with complex social structure. Loss of reproductive individuals in socially complex species could disproportionately affect population dynamics by destabilizing social structure and reducing population growth. Alternatively, compensatory mechanisms such as rapid replacement of breeders may result in little disruption. The impact of breeder loss on the population dynamics of social species remains poorly understood. We evaluated the effect of breeder loss on social stability, recruitment and population growth of grey wolves (Canis lupus) in Denali National Park and Preserve, Alaska using a 26-year dataset of 387 radiocollared wolves. Harvest of breeding wolves is a highly contentious conservation and management issue worldwide, with unknown population-level consequences. Breeder loss preceded 77% of cases (n = 53) of pack dissolution from 1986 to 2012. Packs were more likely to dissolve if a female or both breeders were lost and pack size was small. Harvest of breeders increased the probability of pack dissolution, likely because the timing of harvest coincided with the breeding season of wolves. Rates of denning and successful recruitment were uniformly high for packs that did not experience breeder loss; however, packs that lost breeders exhibited lower denning and recruitment rates. Breeder mortality and pack dissolution had no significant effects on immediate or longer term population dynamics. Our results indicate the importance of breeding individuals is context dependent. The impact of breeder loss on social group persistence, reproduction and population growth may be greatest when average group sizes are small and mortality occurs during the breeding season. This study highlights the importance of reproductive individuals in maintaining group cohesion in social species, but at the population level socially complex species may be resilient

  2. Effects of breeder turnover and harvest on group composition and recruitment in a social carnivore

    Science.gov (United States)

    Ausband, David E.; Mitchell, Michael S.; Waits, Lisette P.

    2017-01-01

    Breeder turnover can influence population growth in social carnivores through changes to group size, composition and recruitment.Studies that possess detailed group composition data that can provide insights about the effects of breeder turnover on groups have generally been conducted on species that are not subject to recurrent annual human harvest. We wanted to know how breeder turnover affects group composition and how harvest, in turn, affects breeder turnover in cooperatively breeding grey wolves (Canis lupus Linnaeus 1758).We used noninvasive genetic sampling at wolf rendezvous sites to construct pedigrees and estimate recruitment in groups of wolves before and after harvest in Idaho, USA.Turnover of breeding females increased polygamy and potential recruits per group by providing breeding opportunities for subordinates although resultant group size was unaffected 1 year after the turnover. Breeder turnover had no effect on the number of nonbreeding helpers per group. After breeding male turnover, fewer female pups were recruited in the new males’ litters. Harvest had no effect on the frequency of breeder turnover.We found that breeder turnover led to shifts in the reproductive hierarchies within groups and the resulting changes to group composition were quite variable and depended on the sex of the breeder lost. We hypothesize that nonbreeding females direct help away from non-kin female pups to preserve future breeding opportunities for themselves. Breeder turnover had marked effects on the breeding opportunities of subordinates and the number and sex ratios of subsequent litters of pups. Seemingly subtle changes to groups, such as the loss of one individual, can greatly affect group composition, genetic content, and short-term population growth when the individual lost is a breeder.

  3. Effects of breeder turnover and harvest on group composition and recruitment in a social carnivore.

    Science.gov (United States)

    Ausband, David E; Mitchell, Michael S; Waits, Lisette P

    2017-09-01

    Breeder turnover can influence population growth in social carnivores through changes to group size, composition and recruitment. Studies that possess detailed group composition data that can provide insights about the effects of breeder turnover on groups have generally been conducted on species that are not subject to recurrent annual human harvest. We wanted to know how breeder turnover affects group composition and how harvest, in turn, affects breeder turnover in cooperatively breeding grey wolves (Canis lupus Linnaeus 1758). We used noninvasive genetic sampling at wolf rendezvous sites to construct pedigrees and estimate recruitment in groups of wolves before and after harvest in Idaho, USA. Turnover of breeding females increased polygamy and potential recruits per group by providing breeding opportunities for subordinates although resultant group size was unaffected 1 year after the turnover. Breeder turnover had no effect on the number of nonbreeding helpers per group. After breeding male turnover, fewer female pups were recruited in the new males' litters. Harvest had no effect on the frequency of breeder turnover. We found that breeder turnover led to shifts in the reproductive hierarchies within groups and the resulting changes to group composition were quite variable and depended on the sex of the breeder lost. We hypothesize that nonbreeding females direct help away from non-kin female pups to preserve future breeding opportunities for themselves. Breeder turnover had marked effects on the breeding opportunities of subordinates and the number and sex ratios of subsequent litters of pups. Seemingly subtle changes to groups, such as the loss of one individual, can greatly affect group composition, genetic content, and short-term population growth when the individual lost is a breeder. © 2017 The Authors. Journal of Animal Ecology © 2017 British Ecological Society.

  4. Plant mutation breeding and biotechnology

    National Research Council Canada - National Science Library

    Shu, Q. Y; Forster, Brian P; Nakagawa, H

    2012-01-01

    ... (FAO / IAEA) Division of Nuclear Techniques in Food and Agriculture, with its global coordinating and synergistic roles, that plant mutation breeding became a common tool available to plant breeders worldwide. Since these early days the Joint Division continues to play a considerable role in fostering the use of mutation techni...

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

    Science.gov (United States)

    Vickers, Lisa Rene

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

  6. Design study of an upgraded charge breeder for ISOLDE

    CERN Document Server

    Shornikov, A; Wenander, F; Pikin, A

    2013-01-01

    In this work we present our progress in the design study of a new Electron Beam Ion Source (EBIS) to be installed as a charge breeder for reacceleration of rare ions at ISOLDE. The work is triggered by the HIE-ISOLDE upgrade {[}1] and the planned TSR@ISOLDE project {[}2]. To fulfill the requests of the user community the new EBIS should reach an electron beam density of 10(4) A/cm(2) at electron energies up to 150 key and, provide UHV environment and ion cooling in the breeding region to ensure confinement of the ions long enough to reach the requested charge states. We report on the established design parameters and first prototyping steps towards production and testing of suitable equipment. (C) 2013 Elsevier B.V. All rights reserved.

  7. Radiocesium in reindeer breeders in Northern Norway since 1965

    Energy Technology Data Exchange (ETDEWEB)

    Selnaes, T.D. [Institutt for Energiteknikk, Kjeller (Norway); Strand, P. [Statens Straalevern, Oesteraas (Norway)

    1995-12-31

    Reindeer breeders in Kautokeino, Northern Norway, have been monitored for whole body content of {sup 137}Cs every years since 1965. The group was chosen because of it`s large intake of reindeer meat (65 kg/year). The same group has been monitored throughout the years. Some supplements have been made, to maintain the average age within the group. Whole body countings have been performed with a scintillation detector and i single- or multichannel analyzer. The highest values were monitored in 1966, when the average body content of {sup 137}Cs was 39,500 Bq for men, and 18,600 Bq for women. Whole body contents are today around 2,000 Bq, giving a calculated whole body dose of 0.1 mSv/year. 5 refs., 4 figs.

  8. Relationship between ecological concepts and biosafety in broiler breeder farms

    Directory of Open Access Journals (Sweden)

    CA Santos

    2007-09-01

    Full Text Available The entrance of poultry products into the trade world requires changes in the configuration of these products, such as programs that ensure their quality and biosafety for the consumers. This article aims at presenting new perception on poultry biosafety programs in broiler breeder farms from an ecological perspective, making these programs more efficient and cost-effective, i.e., more competitive. Using literature review, some convergences were found between ecology concepts and biosafety programs. One of these convergences is understanding the farm as an open ecosystem, integrating through adaptation the natural environment with the exotic environment. This also allows understanding how the production area interacts with the environment as to energy substrate input and output or as to the dissemination of poultry pathogens by vectors outside the farm or from the production area to the environment. This allows building a theoretical reference for further studies on ecological models for the improvement of poultry biosafety programs.

  9. Induction of eggshell apex abnormalities in broiler breeder hens.

    Science.gov (United States)

    Feberwee, A; Landman, W J M

    2010-04-01

    Recently, the causal relationship between eggshell apex abnormalities (EAA) and Mycoplasma synoviae was described. This eggshell pathology has only been documented in table egg layers both spontaneously and experimentally infected with M. synoviae, suggesting that meat-type layers are less prone to this condition. In this study the susceptibility of specified pathogen free (SPF) broiler breeder hens to produce eggs with EAA after M. synoviae infection was assessed. Five groups of 12 hens each were made: a negative control group, a group inoculated intratracheally (i.t.) with a M. synoviae EAA strain at 19 weeks of age, a group inoculated i.t. with this strain at 19 and 26 weeks of age, a group inoculated with M. synoviae i.t. at 19 weeks of age and infected 5 days earlier with infectious bronchitis virus D1466 (IBV), and a fifth group similar to the former but inoculated i.t. twice with an M. synoviae EAA strain at 19 and 26 weeks of age. Eggs with EAA were only produced after a single i.t. inoculation with the M. synoviae EAA strain if preceded by an infection with IBV. The production of eggs with EAA started 6 weeks after M. synoviae EAA inoculation and the proportion of eggs with EAA during the experiment was 9/449 (2%), which was much lower than that in SPF layer hens (14-22%). The present results suggest that broiler breeder hens are less susceptible to producing eggs with EAA after an infection with a M. synoviae EAA strain preceded by an IBV infection, compared with table egg layers. Similar to SPF egg layers, the mean daily egg production per hen was significantly reduced by the M. synoviae EAA strain and there was a general negative effect on eggshell strength by this strain, suggesting it could also have a detrimental effect on hatching egg quality.

  10. Estimation of in-plant source term release behaviors from Fukushima daiichi reactor cores by forward method and comparison with reverse method

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Tae Won; Rhee, Bo Wook; Song, Jin Ho; Kim, Sung Il; Ha, Kwang Soon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2017-06-15

    The purpose of this paper is to confirm the event timings and the magnitude of fission product aerosol release from the Fukushima accident. Over a few hundreds of technical papers have been published on the environmental impact of Fukushima Daiichi accident since the accident occurred on March 11, 2011. However, most of the research used reverse or inverse method based on the monitoring of activities in the remote places and only few papers attempted to estimate the release of fission products from individual reactor core or from individual spent fuel pool. Severe accident analysis code can be used to estimate the radioactive release from which reactor core and from which radionuclide the peaks in monitoring points can be generated. The basic material used for this study are the initial core inventory obtained from the report JAEA-Data/Code 2012–018 and the given accident scenarios provided by Japanese Government or Tokyo Electric Power Company (TEPCO) in official reports. In this research a forward method using severe accident progression code is used as it might be useful for justifying the results of reverse or inverse method or vice versa. The release timing and amounts to the environment are estimated for volatile radioactive fission products such as noble gases, cesium, iodine, and tellurium up to 184 hours (about 7.7 days) after earthquake occurs. The in-plant fission product behaviors and release characteristics to environment are estimated using the severe accident progression analysis code, MELCOR, for Fukushima Daiichi accident. These results are compared with other research results which are summarized in UNSCEAR 2013 Report and other technical papers. Also it may provide the physically based arguments for justifying or suspecting the rationale for the scenarios provided in open literature. The estimated results by MELCOR code simulation of this study indicate that the release amount of volatile fission products to environment from Units 1, 2, and 3

  11. The Effect of Low-Density Broiler Breeder Diets on Performance and Immune Status of their Offspring

    NARCIS (Netherlands)

    Enting, H.; Boersma, W.J.A.; Cornelissen, J.B.W.J.; Winden, van S.C.L.; Verstegen, M.W.A.; Aar, van de P.J.

    2007-01-01

    Effects of low-density broiler breeder diets on offspring performance and mortality were studied using 2,100 female and 210 male Cobb 500 breeders. Breeder treatments involved 4 experimental groups and a control group with normal density diets (ND, 2,600 kcal of AME/kg during rearing and 2,800 kcal

  12. 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.

  13. The effect of γ-radiation in Li{sub 4}SiO{sub 4} ceramic breeder blankets

    Energy Technology Data Exchange (ETDEWEB)

    Carella, E., E-mail: elisabetta.carella@externos.ciemat.es [UNED Foundation, C/Francisco de Rojas, 2, 28010 Madrid (Spain); National Fusion Laboratory, CIEMAT, Av. Complutense 40, 28040 Madrid (Spain); Hernández, T. [National Fusion Laboratory, CIEMAT, Av. Complutense 40, 28040 Madrid (Spain)

    2015-01-15

    Highlights: • Li{sub 4}SiO{sub 4} is γ-irradiated with different doses (5, 10 and 13 MGy) and its electrical bulk conductivity measured by EIS. • The electrical measurements are compared with SiO{sub 2} confirming the charge carrier role of Li{sup +} under thermal activation. • The recombination effect of temperature during irradiation has been observed. - Abstract: Lithium orthosilicate (Li{sub 4}SiO{sub 4}) is considered one of the best candidates for the solid breeder blanket system (BBs) of future thermonuclear reactors. During reactor operation it will be bombarded by neutrons and gamma radiation which may alter its composition and properties, affecting its shielding role. The electrochemical impedance spectroscopy (EIS) is here used as a non-destructive tool for monitoring the electrical bulk conductivity of Li{sub 4}SiO{sub 4} ceramic after different ionizing damaging treatments. The compound fabricated in our laboratories, was irradiated by a {sup 60}Co in the Nayade-facility (CIEMAT-Spain). Several studies with slight experimental variations were realized and here presented to understand the dynamic of the charge carriers’ movement and the role of intrinsic and extrinsic defects on the electrical properties of this candidate ceramic.

  14. NUCLEAR REACTOR

    Science.gov (United States)

    Grebe, J.J.

    1959-07-14

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

  15. Studies of reactor irradiation effect on hydrogen isotope release from vanadium alloy V4Cr4Ti

    Energy Technology Data Exchange (ETDEWEB)

    Kulsartov, T. [Kazakhstan State University, Tole-bi-str. 96a., Almaty (Kazakhstan); Shestakov, V. [Kazakhstan State University, Tole-bi-str. 96a., Almaty (Kazakhstan); Chikhray, Y. [Kazakhstan State University, Tole-bi-str. 96a., Almaty (Kazakhstan); Kenzhin, Y. [Institute of Atomic Energy NNC RK, Krasnoarmeyskaya-str. 10, Kurchatov (Kazakhstan); Kolbayenkov, A. [Institute of Atomic Energy NNC RK, Krasnoarmeyskaya-str. 10, Kurchatov (Kazakhstan); Tazhibayeva, I. [National Nuclear Center, Lenin-str. 6, Kurchatov (Kazakhstan)

    2007-08-01

    Vanadium alloys are most promising materials being considered for lithium blanket-breeder in future fusion reactors. The primary reason for these stems from good combination of physical-mechanical and radiation properties of vanadium alloys. In operational conditions of fusion reactors the very important issue is behavior of vanadium alloy with respect to hydrogen isotopes under neutron and gamma irradiation. This paper shows results of the experimental studies of reactor irradiation influence on parameters of hydrogen release from vanadium alloys. Experiments were carried out for various levels of reactor irradiation and showed the effect of irradiation on parameters of hydrogen release from vanadium alloy V4Cr4Ti.

  16. Safety of VVER-440 reactors

    CERN Document Server

    Slugen, Vladimir

    2011-01-01

    Safety of VVER-440 Reactors endeavours to promote an increase in the safety of VVER-440 nuclear reactors via the improvement of fission products limitation systems and the implementation of special non-destructive spectroscopic methods for materials testing. All theoretical and experimental studies performed the by author over the last 25 years have been undertaken with the aim of improving VVER-440 defence in depth, which is one of the most important principle for ensuring safety in nuclear power plants. Safety of VVER-440 Reactors is focused on the barrier system through which the safety pri

  17. Fuel Fabrication and Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Karpius, Peter Joseph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-02-02

    The uranium from the enrichment plant is still in the form of UF6. UF6 is not suitable for use in a reactor due to its highly corrosive chemistry as well as its phase diagram. UF6 is converted into UO2 fuel pellets, which are in turn placed in fuel rods and assemblies. Reactor designs are variable in moderators, coolants, fuel, performance etc.The dream of energy ‘too-cheap to meter’ is no more, and now the nuclear power industry is pushing ahead with advanced reactor designs.

  18. Light Water Reactor Sustainability Program: Evaluation of Localized Cable Test Methods for Nuclear Power Plant Cable Aging Management Programs

    Energy Technology Data Exchange (ETDEWEB)

    Glass, Samuel W. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Fifield, Leonard S. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Hartman, Trenton S. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-05-30

    This Pacific Northwest National Laboratory (PNNL) milestone report describes progress to date on the investigation of nondestructive test (NDE) methods focusing particularly on local measurements that provide key indicators of cable aging and damage. The work includes a review of relevant literature as well as hands-on experimental verification of inspection capabilities. As NPPs consider applying for second, or subsequent, license renewal (SLR) to extend their operating period from 60 years to 80 years, it important to understand how the materials installed in plant systems and components will age during that time and develop aging management programs (AMPs) to assure continued safe operation under normal and design basis events (DBE). Normal component and system tests typically confirm the cables can perform their normal operational function. The focus of the cable test program is directed toward the more demanding challenge of assuring the cable function under accident or DBE. Most utilities already have a program associated with their first life extension from 40 to 60 years. Regrettably, there is neither a clear guideline nor a single NDE that can assure cable function and integrity for all cables. Thankfully, however, practical implementation of a broad range of tests allows utilities to develop a practical program that assures cable function to a high degree. The industry has adopted 50% elongation at break (EAB) relative to the un-aged cable condition as the acceptability standard. All tests are benchmarked against the cable EAB test. EAB is a destructive test so the test programs must apply an array of other NDE tests to assure or infer the overall set of cable’s system integrity. These cable NDE programs vary in rigor and methodology. As the industry gains experience with the efficacy of these programs, it is expected that implementation practice will converge to a more common approach. This report addresses the range of local NDE cable tests that are

  19. Improved Performance of Broilers and Broiler Breeders Associated with an Amended Vaccination Program Against Reovirosis.

    Science.gov (United States)

    De Herdt, Peter; Broeckx, Marlies; Van Driessche, Filip; Vermeiren, Bart; Van Den Abeele, Geert; Van Gorp, Stefaan

    2016-12-01

    A vertically integrated monitoring program was set up for breeders hatched in 2013 and their offspring to detect differences in performance related to the reovirus vaccination schedule. Within the same organization in Belgium, 17 breeder flocks were vaccinated with one dose of live and one dose of inactivated reovirus vaccine, while 14 flocks received two doses of inactivated vaccine without live priming. The hatchability of the eggs produced by these birds was examined. Further, the daily growth, feed conversion, mortality, slaughterhouse condemnation, production index, and antibiotic use were monitored in 110 broiler flocks derived from the breeders. All gathered data were examined statistically. In eggs obtained from breeders vaccinated twice with inactivated reovirus vaccine, a significant 2.88% higher hatchability rate was observed. The progeny broiler flocks of these breeders showed a significant 18.2% lower mortality during the fattening period. Although not statistically significant, the slaughterhouse condemnation rate was 10.1% lower as well. The results may indicate that-under the epidemiologic conditions of this study-double administration of inactivated reovirus vaccine in broiler breeders can at least contribute to higher hatchability of breeder eggs and lower broiler mortality.

  20. Fungal plant pathogens and the plant immune system

    OpenAIRE

    Wit, de, J.C.M.

    2014-01-01

    Fungi are notorious plant pathogens and continuously threat global food production. In the last decades we have obtained a better understanding of infection strategies of fungi and the plant immune system. This has facilitated more efficient introduction of disease resistance genes in crop plants by plant breeders. A brief overview of progress in research and applications will be provided as well as a glimpse into the future.

  1. Isotope exchange reactions on ceramic breeder materials and their effect on tritium inventory

    Energy Technology Data Exchange (ETDEWEB)

    Nishikawa, M.; Baba, A. [Kyushu Univ., Fukuoka (Japan). Faculty of Engineering; Kawamura, Y.; Nishi, M.

    1998-03-01

    Though lithium ceramic materials such as Li{sub 2}O, LiAlO{sub 2}, Li{sub 2}ZrO{sub 3}, Li{sub 2}TiO{sub 3} and Li{sub 4}SiO{sub 4} are considered as breeding materials in the blanket of a D-T fusion reactor, the release behavior of the bred tritium in these solid breeder materials has not been fully understood. The isotope exchange reaction rate between hydrogen isotopes in the purge gas and tritium on the surface of breeding materials have not been quantified yet, although helium gas with hydrogen or deuterium is planned to be used as the blanket purge gas in the recent blanket designs. The mass transfer coefficient representing the isotope exchange reaction between H{sub 2} and D{sub 2}O or that between D{sub 2} and H{sub 2}O in the ceramic breeding materials bed is experimentally obtained in this study. Effects of isotope exchange reactions on the tritium inventory in the bleeding blanket is discussed based on data obtained in this study where effects of diffusion of tritium in the grain, absorption of water in the bulk of grain, and adsorption of water on the surface of grain, together with two types of isotope exchange reactions are considered. The way to estimate the tritium inventory in a Li{sub 2}ZrO{sub 3} blanket used in this study shows a good agreement with data obtained in such in-situ experiments as MOZART, EXOTIC-5, 6 and TRINE experiments. (author)

  2. Thermal-hydraulic analysis on the whole module of water cooled ceramic breeder blanket for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Jiang, Kecheng; Ma, Xuebin [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); University of Science and Technology of China, Hefei, Anhui, 230027 (China); Cheng, Xiaoman [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); Lin, Shuang [University of Science and Technology of China, Hefei, Anhui, 230027 (China); Huang, Kai [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); Liu, Songlin, E-mail: slliu@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); University of Science and Technology of China, Hefei, Anhui, 230027 (China)

    2016-11-15

    Highlights: • The 3D thermal hydraulic analysis on the whole module of WCCB is performed by CFD method. • Temperature field and mass flow distribution have been obtained. • The design of WCCB is reasonable from the perspective of thermal-hydraulics. • The scheme for further optimization has been proposed. - Abstract: The Water Cooled Ceramic Breeder blanket (WCCB) is being researched for Chinese Fusion Engineering Test Reactor (CFETR). The thermal-hydraulic analysis is essential because the blanket should remove the high heat flux from the plasma and the volumetric heat generated by neutrons. In this paper, the detailed three dimensional (3D) thermal hydraulic analysis on the whole module of WCCB blanket has been performed by Computational Fluid Dynamics (CFD) method, which is capable of solving conjugate heat transfer between solid structure and fluid. The main results, including temperature field, distribution of mass flow rate and coolant pressure drop, have been calculated simultaneously. These provides beneficial guidance data for the further structural optimization and for the design arrangement of primary and secondary circuit. Under the total heat source of 1.23 MW, the coolant mass flow rate of 5.457 kg/s is required to make coolant water corresponding to the Pressurized Water Reactor (PWR) condition (15.5 MPa, 285 °C–325 °C), generating the total coolant pressure drop (△P) of 0.467 MPa. The results show that the present structural design can make all the materials effectively cooled to the allowable temperature range, except for a few small modifications on the both sides of FW. The main components, including the first wall (FW), cooling plates (CPs), side wall (SWs)&stiffening plates (SPs) and the manifold(1–4), dominate 4.7%/41.7%/13%/40.6% of the total pressure drop, respectively. Additionally, the mass flow rate of each channel has been obtained, showing the peak relative deviation of 3.4% and 2% from the average for the paratactic

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

    Energy Technology Data Exchange (ETDEWEB)

    Moisseytsev, A.; Sienicki, J. J.

    2007-03-08

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

  4. Parentage analysis with few contributing breeders: validation and improvement.

    Science.gov (United States)

    Duchesne, Pierre; Meldgaard, Torben; Berrebi, Patrick

    2008-01-01

    Validation of parental allocation using PAPA software (Duchesne P, Godbout MH, Bernatchez L. 2002. PAPA (package for the analysis of parental allocation): a computer program for simulated and real parental allocation. Mol Ecol Notes. 2:191-193.) was investigated under the assumption that only a small proportion of potential breeders contributed to the offspring sample. Inbreeding levels proved to have a large impact on allocation error rate. Consequently, simulations from artificial, unrelated parents may strongly underestimate allocation error, and so, whenever possible, simulations based on the actual parental genotypes should be run. An unexpected and interesting finding was that ambiguity (the highest likelihood is shared by several parental pairs) rates below 10% stood very close to exact allocation error rates (true proportions of wrong allocations). Hence, the ambiguity rate statistic may be viewed as a ready-made indicator of the resolution power of a specific parental allocation run and, if not exceeding 10%, used as an estimate of allocation error rate. It was found that the PAPA simulator, even with few contributing breeders, can be trusted to output reasonably accurate estimates of allocation error as long as those estimates do not exceed 15%. Indeed, most discrepancies between exact and estimated error then stood below 3%. Reproductive success variance had little impact on error estimate discrepancies within the same range. Finally, a (focal set) method was described to correct the estimated family sizes computed directly from parental allocations. Essentially, this method makes use of the detailed structure of the allocation probabilities associated with each parental pair with at least 1 allocated offspring. The allocation probabilities are expressed in matrix form, and the subsequent calculations are run based on standard matrix algebra. On average, this method provided better estimates of family sizes for each investigated combination of parameter

  5. Reactor Neutrinos

    CERN Document Server

    Lasserre, T; Lasserre, Thierry; Sobel, Henry W.

    2005-01-01

    We review the status and the results of reactor neutrino experiments, that toe the cutting edge of neutrino research. Short baseline experiments have provided the measurement of the reactor neutrino spectrum, and are still searching for important phenomena such as the neutrino magnetic moment. They could open the door to the measurement of coherent neutrino scattering in a near future. Middle and long baseline oscillation experiments at Chooz and KamLAND have played a relevant role in neutrino oscillation physics in the last years. It is now widely accepted that a new middle baseline disappearance reactor neutrino experiment with multiple detectors could provide a clean measurement of the last undetermined neutrino mixing angle theta13. We conclude by opening on possible use of neutrinos for Society: NonProliferation of Nuclear materials and Geophysics.

  6. NEUTRONIC REACTORS

    Science.gov (United States)

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

    1958-10-14

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

  7. The influence of operational and water chemistry parameters on the deposits of corrosion products on fuel assemblies at nuclear power plants with VVER reactors

    Science.gov (United States)

    Kritskii, V. G.; Berezina, I. G.; Rodionov, Yu. A.; Gavrilov, A. V.

    2011-07-01

    The phenomenon involving a growth of pressure drop in the reactor core and redistribution of deposits in the reactor core and primary coolant circuit of a nuclear power station equipped with VVER-440 reactors is considered. A model is developed, the physicochemical foundation of which is based on the dependence of corrosion product transfer on the temperature and pH t value of coolant and on the correlation between the formation rate of corrosion products (Fe) (after subjecting the steam generators to decontamination) and rate with which they are removed from the circuit. The purpose of the simulation carried on the model is to predict the growth of pressure drop on the basis of field data obtained from nuclear power installations and correct the water chemistry (by adjusting the concentrations of KOH, H2, and NH3) so as to keep the pressure drop in the reactor at a stable level.

  8. Overview of pool hydraulic design of Indian prototype fast breeder ...

    Indian Academy of Sciences (India)

    out for validation of the computer codes have also been described. Keywords. Fast reactor ..... water experiments and limited sodium experiments to validate computer codes and then using these validated computer ..... Based on detailed thermal hydraulic assessment of circumferential temperature non-uniformity and jet.

  9. Analysis of changes in egg quality of broiler breeders during the first reproduction period

    National Research Council Canada - National Science Library

    Helena Kontecka; Sebastian Nowaczewski; Marta M. Sierszuła

    2012-01-01

    .... The aim of the study was to analyse the physical characteristics, morphological composition and quality of individual components of the eggs of Cobb 500 broiler breeders during the first year of reproduction...

  10. Innovatieve huisvesting voor vleeskuikenouderdieren: "Quality Time" stal = Innovative broiler breeder housing system: 'Quality Time' house

    OpenAIRE

    Emous, van, R.A.; Gunnink, H.

    2011-01-01

    In this study results are presented of the research of implementing the innovative housing system "Quality Time" concept in a practical poultry house for broiler breeders. The "Quality Time" concept can improve the sexual behaviour and fertility of the eggs.

  11. Serological profiles of commercial broiler breeders and their progeny. 2. Newcastle disease virus.

    Science.gov (United States)

    King, D J

    1986-01-01

    Newcastle disease virus (NDV) hemagglutination-inhibition (HI) titers were determined for serum samples from eight commercial broiler breeder flocks and their progeny. The chickens sampled had been vaccinated and reared by different producers in different regions of the United States. Breeder flocks had the highest number of NDV-positive HI titers (greater than or equal to 1:10). Eighty percent or more of the samples from six of eight breeder flocks were positive; the geometric mean titers (GMTs) for those six breeder flocks ranged from 19 to 92. Only 3 of 8 broiler flocks had an increased frequency of positive titers and higher GMTs after vaccination. The frequency of positive titers was greater than 80% in only 2 of 8 of the oldest broiler flocks. The number of NDV-negative titers (less than 1:10) increased with age in most broiler flocks, even though all had been vaccinated once or more with live NDV vaccines.

  12. Helper effects on breeder allocations to direct care.

    Science.gov (United States)

    Kushnick, Geoff

    2012-01-01

    Mothers receive childcare and productive assistance from allomaternal helpers in many societies. Although much effort has been aimed toward showing helper effects on maternal reproductive success, less has been directed toward highlighting the full range of potential effects on breeder behavior. I present a model of optimal maternal care with helpers, and tests of derived hypotheses with data collected among the Karo Batak-a group of Indonesian agriculturalists. To test the model's predictions I compared the effect of women receiving help from patrilateral versus matrilateral kin because those kin may provide help with different maternal responsibilities. The model predicts a decrease in maternal allocation to care that is substitutable with the helper contribution and the helper assists with that type of care; it predicts an increase in care that is nonsubstitutable with the helper contribution or substitutable care when the helper assists with other responsibilities. With the exception of one other, most models have failed to account for an increase. Analyses of time spent carrying children supported the model. With matrilateral helpers, women increased carrying; with patrilateral helpers, they decreased it. Time spent farmworking showed the opposite pattern, suggesting that matrilateral helpers effectively decrease costs, nudging optimal maternal care upward. Patterns of breastfeeding provided little support for the model. The results do, however, suggest potential proximate mechanisms by which helpers influence maternal reproductive success in cooperative breeding societies. Copyright © 2012 Wiley Periodicals, Inc.

  13. Argon generation in fusion reactor materials

    Energy Technology Data Exchange (ETDEWEB)

    Khripunov, Vladimir, E-mail: Khripunov_VI@NRCKI.ru

    2015-10-15

    Highlights: • A relatively long-lived Ar-39 (T{sub 1/2} = 269 yr) may appear in fusion reactor materials. • Ar-39 activities may become apparent after tritium removal. • Initial impurity control of K is definitely recommended. • A substantiation of the effective dose rates for exposure to inert argon is urgent. - Abstract: Different candidate plasma facing materials (as tungsten, beryllium), the low activation structure materials (as vanadium alloys, silicon carbides), liquid breeders (lithium and lithium-lead) and some others have been suggested for future fusion power reactor cores as corresponding to maintenance, recycling and for waste disposal acceptance after 50 and 100 years of cooling. It is shown by the neutron activation analysis that a relatively short-lived Ar-41 (T{sub 1/2} = 1.85 h), Ar-37 (T{sub 1/2} = 35 days) and rather long lived Ar-39 (T{sub 1/2} = 269 yr) may appear in these materials under the fusion neutron irradiation conditions. While argon production is essentially less than helium production in irradiated materials, at other times its impact, e.g., in the inhalation dose, becomes significant. In some cases the Ar-39 activity is comparable or even exceeds the C-14 activity and may become apparent after tritium removal from plasma exhaust and dust, from the liquid breeders, during plasma-facing and structural component recycling and waste management. The main source terms of argon-39 activity for these materials are identified and the specific production rates are evaluated relative to radiation conditions of a power or DEMO fusion reactor and to electric power production.

  14. Evaluation of permeable and non-permeable tritium in normal condition in a fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Marta, V; Manuel, P J [Instituto de Fusion Nuclear (DENIM)/ETSII, Universidad Politecnica Madrid (UPM) (Spain); Sedano Luis, A [Ministerio de Educacion y Ciencia, Ciemat (Spain)], E-mail: marta@denim.upm.es

    2008-05-15

    The tritium cycle, technologies of process and control of the tritium in the plant will constitute a fraction of the environmental impact of the first generation of DT fusion reactors. The efforts of conceptual development of the tritium cycle are centered in the Internal Regenerator Cycle. The tritium could be recovered from a flow of He gas, or directly from solid breeder. The limits of transfers to the atmosphere are assumed {approx} 1 gr-T/a ({approx}20 Ci/a) (without species distinction). In the case of ITER, for example, we have global demands of control of 5 orders of magnitude have been demonstrated at experimental level. The transfer limits determine the key parameters in tritium Cycle (HT, HTO, as dominant, and T2, T2O as marginal). Presently, the transfer from the cycle to the environment is assumed through the exchange system of the power plant (primary to secondary). That transport is due to the permeation through HT, T2, or leakage to the coolant in the primary system. It is key the chemical optimization in the primary system, that needs to be reanalyzed in terms of radiological impact both for permeable, HT, T2, and non-permeable HTO, T2O. It is necessary considered the pathway of tritium from the reactor to the atmosphere, these processes are modelled adequately. Results of the assessments were early and chronic doses which have been evaluated for the Most Exposed Individual at particular distance bands from the release point. The impact evaluations will be performed with the computational tools (NORMTRI), besides national regulatory models, internationally accepted computer these code for dosimetric evaluations of tritiated effluents in operational conditions.

  15. Neutronic reactor

    Science.gov (United States)

    Wende, Charles W. J.; Babcock, Dale F.; Menegus, Robert L.

    1983-01-01

    A nuclear reactor includes an active portion with fissionable fuel and neutron moderating material surrounded by neutron reflecting material. A control element in the active portion includes a group of movable rods constructed of neutron-absorbing material. Each rod is movable with respect to the other rods to vary the absorption of neutrons and effect control over neutron flux.

  16. Neutronic reactor

    Energy Technology Data Exchange (ETDEWEB)

    Babcock, D.F.; Menegus, R.L.; Wende, C.W.

    1983-01-04

    A nuclear reactor includes an active portion with fissionable fuel and neutron moderating material surrounded by neutron reflecting material. A control element in the active portion includes a group of movable rods constructed of neutron-absorbing material. Each rod is movable with respect to the other rods to vary the absorption of neutrons and effect control over neutron flux.

  17. Digestible Threonine Levels in the Starter Diet of Broilers Derived from Breeders of Different Ages

    Directory of Open Access Journals (Sweden)

    CBGS Tanure

    2015-12-01

    Full Text Available ABSTRACT The aim of this study was to evaluate the effect of digestible threonine supplementation in the starter diet on the performance, intestinal parameters, and nutrient metabolism of broilers derived from breeders of different ages. In total, 480 one-day-old Cobb chicks, derived from 38-or 49-week-oldbreeders, were housed in experimental battery cages until 21 days of age and fed four different threonine levels (800, 900, 1,000, or 1,100 mg/kg in the starter feed. A completely randomized experimental design in a 2x4 factorial arrangement (breeder age x threonine levels was applied, totaling eight treatments with five replicates of 12 birds each. Broilers from older breeders fed 800 mg digestible threonine/kg of diet presented higher weight gain, with a positive linear effect. There was also an interaction between breeder age and threonine levels for the weight gain of 21-d-old broilers supplemented at maximum level of 1,003 mg Thr/kg diet during the starter phase. There was no effect of breeder age or threonine levels on nutrient metabolism during the period of 17-21 days. There was no influence of breeder age or threonine levels in the starter diet on intestinal morphometric measurements, absorption area, or percentage of goblet cells.

  18. Space reactors. Progress report, July-September 1981

    Energy Technology Data Exchange (ETDEWEB)

    Ranken, W.A. (comp.)

    1982-06-01

    Progress in technology development for the Space Power Advanced Reactor (SPAR) project is reported for the period July 1-September 30, 1981. The weights of neutron and gamma shields required to protect the SPAR system payloads for a range of permissible exposures have been determined, and initial results are reported. SPAR reactor safety in the case of water immersion has been modeled. Approval-in-Principle has been received for the SPAR fuel test in the Experimental Breeder Reactor-II (EBR-II); the heat pipe developed for this test is performing well. SPAR system design variations are being examined under the possibility of using long core heat pipes. Testing of the initial molybdenum/sodium artery heat pipe continued, with ambiguous results. Fabrication of the first all-bonded thermoelectric units has been completed and testing has been initiated.

  19. Selective breeding of Arabian and Thoroughbred racehorses in Algeria: perceptions, objectives and practices of owners-breeders

    Directory of Open Access Journals (Sweden)

    Safia Tennah

    2014-04-01

    Full Text Available This survey, conducted with 461 racehorse owners-breeders in Algeria between 2009 and 2011, investigates their perceptions, objectives and practices regarding selective breeding. Racehorse breeding is a full-time professional activity for a third of interviewees. The holdings are small-sized with 77% owning one or two mares. The regular practice of mating is here used to categorize breeders according to their degree of professionalization (38.4% professional vs. 61.6% occasional breeders. Experience in the sector was also used to classify breeders, considering as "junior" the breeders under 10 years experience (38.8% and as "senior" those above 10 years (61.2%. More than professionalization, experience shows a significant impact on practices and objectives. Thus, experience influences breed choice (junior breeders tend to specialize while senior own both Arabian and Thoroughbreds, age at first foaling (sooner among senior breeders, information sources considered for selecting stallions (senior use more diversified sources, the importance granted to the price of mating (greater for junior breeders, the importance granted to the ranking compared to earnings (the ranking being more important to junior breeders, and the priority given to breeding (junior breeders give higher priority to a buy-race-resell activity. Finally, racehorse breeding is poorly professionalized, the only financial goal being cost coverage. Despite inappropriate practices, an interest for selection is noticed.

  20. Li vaporization property of two-phase material of Li{sub 2}TiO{sub 3} and Li{sub 2}SiO{sub 3} for tritium breeder

    Energy Technology Data Exchange (ETDEWEB)

    Ogawa, Seiya [Course of Mechanical Engineering, Graduate School of Engineering, Tokai University, 4-1-1 Kitakaname, Hiratsuka, Kanagawa 259-1292 (Japan); Masuko, Yuki; Kato, Hirokazu; Yuyama, Hayato; Sakai, Yutaro [Department of Prime Mover Engineering, School of Engineering, Tokai University, 4-1-1 Kitakaname, Hiratsuka, Kanagawa 259-1292 (Japan); Niwa, Eiki; Hashimoto, Takuya [Department of Physics, College of Humanities and Sciences, Nihon University, 3-8-1 Sakurajousui, Setagaya-ku, Tokyo 156-8550 (Japan); Mukai, Keisuke [Department of Nuclear Engineering and Management, School of Engineering, The University of Tokyo, 7-3-1 Bunkyo-ku, Tokyo 113-8656 (Japan); Hosino, Tsuyoshi [Breeding Functional Materials Development Group, Department of Blanket Systems Research, Rokkasho Fusion Institute, Sector of Fusion Research and Development, Japan Atomic Energy Agency, 2-166 Obuch, Omotedate, Rokkasho-mura, Kamikita-gun, Aomori 039-3212 (Japan); Sasaki, Kazuya, E-mail: k_sasaki@tokai-u.jp [Course of Mechanical Engineering, Graduate School of Engineering, Tokai University, 4-1-1 Kitakaname, Hiratsuka, Kanagawa 259-1292 (Japan); Department of Prime Mover Engineering, School of Engineering, Tokai University, 4-1-1 Kitakaname, Hiratsuka, Kanagawa 259-1292 (Japan); Course of Mechanical Engineering and Aeronautics and Astronautics, Graduate School of Science and Technology, Tokai University, 4-1-1 Kitakaname, Hiratsuka, Kanagawa 259-1292 (Japan)

    2015-10-15

    Highlights: • We synthesized two phase materials based on Li{sub 2}SiO{sub 3} and Li{sub 2}TiO{sub 3}. • We investigated the Li vaporization property of the two-phase materials. • Li vaporization occurs significantly from only Li{sub 2}SiO{sub 3} grains in the vicinity of the surface of the pellets. • The Li vaporization is remarkable only for an early short time for the vaporization from Li{sub 2}SiO{sub 3} grains at the vicinity of the surface. • The second stable phase added functions effectively for inhibition of the Li vaporization. - Abstract: Li vaporization property of two-phase materials of Li{sub 2}TiO{sub 3} and Li{sub 2}SiO{sub 3} in a working condition for the solid tritium breeder used in the demonstration power plant of fusion reactor was investigated, and the suppression mechanism of the vaporization was considered. The Li vaporization rate from the specimen pellet was measured by gravimetric method, and the change of Li concentration distribution in the pellet was analyzed by time-of-flight secondary ion mass spectrometer. Li was vaporized only from the Li{sub 2}SiO{sub 3} at the vicinity of the surface of the pellet. The remarkable vaporization of Li arose only in an early short time. The inhibition of the vaporization from the Li{sub 2}SiO{sub 3} was successful by adding the small amount of the stable secondary phase of Li{sub 2}TiO{sub 3}.

  1. Helium-cooled high temperature reactors

    Energy Technology Data Exchange (ETDEWEB)

    Trauger, D.B.

    1985-01-01

    Experience with several helium cooled reactors has been favorable, and two commercial plants are now operating. Both of these units are of the High Temperature Graphite Gas Cooled concept, one in the United States and the other in the Federal Republic of Germany. The initial helium charge for a reactor of the 1000 MW(e) size is modest, approx.15,000 kg.

  2. Selective purge for hydrogenation reactor recycle loop

    Science.gov (United States)

    Baker, Richard W.; Lokhandwala, Kaaeid A.

    2001-01-01

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

  3. Final Report for the 1st Surveillance Test of the Reactor Pressure Vessel Material (Capsule 2) of Ulchin Nuclear Power Plant Unit 4

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai (and others)

    2007-04-15

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 1st surveillance testing was performed completely by Korea Atomic Energy Research Institute at Daejon after the capsule was transported from Ulchin site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Ulchin Unit 4 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsule 2 is 4.306E+18n/cm{sup 2}. The bias factor, the ratio of calculation/measurement, was 0.918 for the 1st testing and the calculational uncertainty,7.0% satisfied the requirement of USNRC Reg.Guide 1.190, 20%. The best estimated neutron fluence for reactor vessel inside surface was 3.615E+18n/cm{sup 2} based on the end of 6th fuel cycle and it was predicted that the fluences of vessel inside surface at 16 and 32EFPY would reach 8.478E+18 and 1.673E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Ulchin Unit 4 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life.

  4. Final Report for the 1st Surveillance Test of the Reactor Pressure Vessel Material (CAPSULE 2) of Ulchin Nuclear Power Plant Unit 3

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai (and others)

    2006-12-15

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 1st surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejon after the capsule was transported from Ulchin site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Ulchin unit 3 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsule 2 is 4.674E 18n/cm{sup 2}. The bias factor, the ratio of calculation/measurement, was 0.920 for the 1st testing and the calculational uncertainty,7.0% satisfied the requirement of USNRC Reg.Guide 1.190, 20%. The best estimated neutron fluence for reactor vessel inside surface was 3.913E 18n/cm{sup 2} based on the end of 6th fuel cycle and it was predicted that the fluences of vessel inside surface at 16 and 32EFPY would reach 9.249E 18 and 1.834E 19n/cm{sup 2} based on the current calculation. The result through this analysis for Ulchin unit 3 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life.

  5. Microwave cavity studies for thermal testing of ceramic breeder materials

    Energy Technology Data Exchange (ETDEWEB)

    Kuston, R.L.

    1987-01-01

    Dielectric heating of proposed ceramic tritium breeder material to study the thermomechanical and thermal-hydraulic properties of the material has been previously suggested. Recent computer studies using codes capable of modeling three-dimensional EM cavities with enclosed dielectric material have been used to determine the size limitations of cavity designs at 200 MHz. The sample can be as large as 0.44 /times/ 0.72m in the plane that is transverse to the direction of neutron flux. The uniformity of volumetric heating over the transverse plane is constant to within a few percent. The sample can be as long as 10cm in the direction of the heat flux and match the expected exponential decay of heat generation, exp( /minus/z/lambda), to within +/minus/8%. The design of the chamber is decribed, including the sample region, additional dielectric loading blocks on two sides of the sample region that are required to generate the field uniformity in the transverse plane, and a description of the matching-section portion of the cavity which provides the correct geometry to cause the cavity to resonate at 200 MHz with the right z dependence to stimulate the exponentially-decaying heat profile in the sample region. The matching section consists of two dielectric slabs, one on each wall of the chamber, and an air or free space region in the center of the matching section. The coupling loop is located near the wall end of the matching section in the free space region. 7 refs., 2 figs.

  6. Numerical research on the neutronic/thermal-hydraulic/mechanical coupling characteristics of the optimized helium cooled solid breeder blanket for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Cui, Shijie; Zhang, Dalin, E-mail: dlzhang@mail.xjtu.edu.cn; Cheng, Jie; Tian, Wenxi; Su, G.H.

    2017-01-15

    As one of the candidate tritium breeding blankets for Chinese Fusion Engineering Test Reactor (CFETR), a conceptual structure of the helium cooled solid breeder blanket has recently been proposed. The neutronic, thermal-hydraulic and mechanical characteristics of the blanket directly affect its tritium breeding and safety performance. Therefore, neutronic/thermal-hydraulic/mechanical coupling analyses are of vital importance for a reliable blanket design. In this work, first, three-dimensional neutronics analysis and optimization of the typical outboard equatorial blanket module (No. 12) were performed for the comprehensive optimal scheme. Then, thermal and fluid dynamic analyses of the scheme under both normal and critical conditions were performed and coupled with the previous neutronic calculation results. With thermal-hydraulic boundaries, thermo-mechanical analyses of the structure materials under normal, critical and blanket over-pressurization conditions were carried out. In addition, several parametric sensitivity studies were also conducted to investigate the influences of the main parameters on the blanket temperature distributions. In this paper, the coupled analyses verify the reasonability of the optimized conceptual design preliminarily and can provide an important reference for the further analysis and optimization design of the CFETR helium cooled solid breeder blanket.

  7. Comparison of serum leptin, glucose, total cholesterol and total protein levels in fertile and repeat breeder cows

    Directory of Open Access Journals (Sweden)

    Saime Guzel

    2014-12-01

    Full Text Available In the present study we measured serum glucose, leptin, total cholesterol and total protein concentrations in repeat breeder cows and compared them with fertile cows. For this aim, 20 repeat breeder cows and 20 fertile cows were used as material. Repeat breeder cows were found to have lower levels of leptin and glucose as compared with fertile ones. No significant differences in total cholesterol and total protein levels were observed between the two groups. No significant correlation of leptin with glucose, total cholesterol and total protein was observed in fertile and repeat breeder cows. Low concentrations of glucose and leptin can have some effects on reproductive problems as repeat breeder and help to understand potential mechanisms impairing fertility in repeat breeder cows.

  8. OECD - HRP Summer School on Light Water Reactor Structural Materials. August 26th - 30th, 2002

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-07-01

    In cooperation with the OECD Nuclear Energy Agency (NEA), the Halden Reactor Project organised a Summer School on Light Water Reactor Structural Materials in the period August 26 - 30, 2002. The summer school was primarily intended for people who wanted to become acquainted with materials-related subjects and issues without being experts. It is especially hoped that the summer school served to transfer knowledge to the ''young generation'' in the field of nuclear. Experts from Halden Project member organisations were solicited for the following programme: (1) Overview of The Nuclear Community and Current Issues, (2) Regulatory Framework for Ensuring Structural Integrity, (3) Non-Destructive Testing for Detection of Cracks, (4) Part I - Basics of Radiation and Radiation Damage, (5) Part II - Radiation Effects on Reactor Internal Materials, (6) Water Chemistry and Radiolysis Effects in LWRs, (7) PWR and Fast Breeder Reactor Internals, (8) PWR and Fast Breeder Reactor Internals, (9) Secondary Side Corrosion Cracking of PWR Steam Generator Tubes, (10) BWR Materials and Their Interaction with the Environment, (11) Radiation Damage in Reactor Pressure Vessels.

  9. Radionuclide inventories in the discharged fuels of PHWR-220, BWR-160, VVER-1000 and the conceptual ATBR-600 reactors - A case study

    Energy Technology Data Exchange (ETDEWEB)

    Pal, Usha [Light Water Reactors Physics Section, Reactor Physics Design Division, Bhabha Atomic Research Centre, Mumbai 400 085 (India)], E-mail: ushapal@barc.gov.in; Jagannathan, V. [Light Water Reactors Physics Section, Reactor Physics Design Division, Bhabha Atomic Research Centre, Mumbai 400 085 (India)], E-mail: vjagan@barc.gov.in

    2008-10-15

    Radionuclides content in the discharged fuel of the conceptual thorium breeder reactor ATBR-600 has been assessed and compared against other thermal power reactors considered in Indian nuclear power programme. The contribution of actinides and the fission products inventories in the discharged fuels are separately estimated and assessed. The ATBR-600 reactor is suggested for closed fuel cycle option. The relatively large presence of the unspent plutonium would in fact be recycled. Nonetheless, the data has been presented in the event of operating ATBR-600 like other present day power reactors in a once through fuel cycle mode.

  10. Nuclear Power Plant Thermocouple Sensor-Fault Detection and Classification Using Deep Learning and Generalized Likelihood Ratio Test

    Science.gov (United States)

    Mandal, Shyamapada; Santhi, B.; Sridhar, S.; Vinolia, K.; Swaminathan, P.

    2017-06-01

    In this paper, an online fault detection and classification method is proposed for thermocouples used in nuclear power plants. In the proposed method, the fault data are detected by the classification method, which classifies the fault data from the normal data. Deep belief network (DBN), a technique for deep learning, is applied to classify the fault data. The DBN has a multilayer feature extraction scheme, which is highly sensitive to a small variation of data. Since the classification method is unable to detect the faulty sensor; therefore, a technique is proposed to identify the faulty sensor from the fault data. Finally, the composite statistical hypothesis test, namely generalized likelihood ratio test, is applied to compute the fault pattern of the faulty sensor signal based on the magnitude of the fault. The performance of the proposed method is validated by field data obtained from thermocouple sensors of the fast breeder test reactor.

  11. Seeds that give: Participatory plant breeding

    International Development Research Centre (IDRC) Digital Library (Canada)

    IDRC

    One method the researchers used to introduce farmers to new or unknown varieties or lines was the seed fair. Fairs are organized by plant breeders and take place at the .... unique philosophy of plant breeding. “For me, when I think about varieties, they are like people. We are not all the same: some of us like to work hard, ...

  12. Nuclear reactor PBMR and cogeneration; Reactor nuclear PBMR y cogeneracion

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez S, J. R.; Alonso V, G., E-mail: ramon.ramirez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2013-10-15

    In recent years the nuclear reactor designs for the electricity generation have increased their costs, so that at the moment costs are managed of around the 5000 US D for installed kw, reason for which a big nuclear plant requires of investments of the order of billions of dollars, the designed reactors as modular of low power seek to lighten the initial investment of a big reactor dividing the power in parts and dividing in modules the components to lower the production costs, this way it can begin to build a module and finished this to build other, differing the long term investment, getting less risk therefore in the investment. On the other hand the reactors of low power can be very useful in regions where is difficult to have access to the electric net being able to take advantage of the thermal energy of the reactor to feed other processes like the water desalination or the vapor generation for the processes industry like the petrochemical, or even more the possible hydrogen production to be used as fuel. In this work the possibility to generate vapor of high quality for the petrochemical industry is described using a spheres bed reactor of high temperature. (Author)

  13. Fusion reactor blanket with Li17-Pb83 eutectic

    Energy Technology Data Exchange (ETDEWEB)

    Antipenkov, A.; Danilov, I.; Epinatiev, A.; Eremin, S.; Kalinin, G.; Kolganov, V.; Poliksha, V.; Shchipakin, O.; Shiverski, E.; Sidorov, A.; Skladnov, K.; Strebkov, Yu. (Research and Development Inst. of Power Engineering, Moscow (USSR)); Butko, A.; Kuzmin, A. (Moscow Inst. for Hydraulic Engineering and Land-Reclamation (USSR)); Chepovski, A.; Khripunov, V.; Shatalov, G. (Kurchatov Inst. of Atomic Energy, Moscow (USSR))

    1991-04-01

    The article contains some features of using Li17-Pb83 eutectic as a breeder for ITER/OTR fusion reactor. Described blanket design options aim to reduce electromagnetic loads or relieve eutectic/channel interaction. Eutectic channel stress analysis confirms design feasibility. Channel temperature behaviour is analyzed for loss of blanket cooling system power and rupture of a distribution header accidents. First wall and blanket failure rates were evaluated. The results of neutron power density distribution and tritium breeding ratio estimation are presented. Polonium concentration dynamics is estimated for accidental conditions. Some guide-lines for future work and design progress are advised. (orig.).

  14. Lead-cooled fast reactor (LFR) overview and perspectives

    OpenAIRE

    CINOTTI Luciano; Smith, Craig F.; SEKIMOTO, HIROSHI

    2009-01-01

    The GIF Technology Roadmap identified the Lead-cooled Fast Reactor (LFR) as a technology with great potential to meet the small-unit electricity needs of remote sites while also offering advantages as a large system for grid-connected power stations. The LFR features a fast- neutron spectrum and a closed fuel cycle for efficient conversion of fertile uranium. It can also be used as a burner of minor actinides from spent fuel and as a burner/breeder. An important feature of the LFR is the ...

  15. The 4th surveillance test and evaluation of the reactor pressure vessel material (capsule W) of Yonggwang nuclear power plant unit 2

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwon Jae; Gong, Un Sik; Chang, Jong Hwa; Joo, Yong Sun; Ahn, Sang Bok; Hong, Joon Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2001-02-01

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 4th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Kori site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Yonggwang unit 2 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X and W are 5.762E+18, 1.5391E+19, 3.5119E+19, and 4.2610E+19 n/cm{sup 2}, respectively. The bias factor, the ratio of measurement versus calculation, was 0.899 for the 1st through 4th testing and the calculational uncertainty, 12.3% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.357E+19 n/cm{sup 2} based on the end of 11th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 3.525E+19, 4.337E+19, 5.148E+19 and 5.960E+19 n/cm{sup 2} based on the current calculation. The result through this analysis for Yonggwang unit 2 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 48 refs., 35 figs., 42 tabs. (Author)

  16. The 4th surveillance test and evaluation of the reactor pressure vessel material (capsule W) of Younggwang nuclear power plant unit1

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwon Jae [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2001-08-01

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 4th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Yonggwang site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Yonggwang unit 1 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X and W are 5.555E+18, 1.662E+19, 3.358E+19, and 4.521E+19 n/cm{sup 2}, respectively. The bias factor, the ratio of measurement versus calculation, was 0.859 for the 1st through 4th testing and the calculational uncertainty, 11.80% satisfied the requirement of USNRC Reg.Guide 1.190, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.551E+19 n/cm{sup 2} based on the end of 12th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 3.929E+19, 4.880E+19, 5.831E+19 and 6.782E+19 n/cm{sup 2} based on the current calculation. The result through this analysis for Yonggwang unit 1 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 4 refs., 41 figs., 35 tabs. (Author)

  17. Neutronic performance of two European breeder-inside-tube (BIT) blankets for DEMO: the helium-cooled ceramic LiAlO{sub 2} with Be multiplier and the water-cooled liquid Li{sub 17}Pb

    Energy Technology Data Exchange (ETDEWEB)

    Petrizzi, L. [Associazione EURATOM-ENEA sulla Fusione, Frascati (Italy); Rado, V. [Associazione EURATOM-ENEA sulla Fusione, Frascati (Italy)

    1995-03-01

    In support of ENEA activity in the European Community Test Programme, neutron analysis has been performed on the two latest blanket designs: helium-cooled ceramic breeder-inside-tube (BIT) (with LiAlO{sub 2} and Be multiplier) and water-cooled liquid Li{sub 17}Pb in cylindrical modules (CM). The powerful MCNP Monte Carlo code was used (version 4.2). A detailed and accurate description of the geometrical model has been performed by inserting the main reactor details and avoiding breeder material dilution inside the modules. The tritium breeding ratio (TBR) performance is low for the solid breeder BIT blanket (with 10 ports 1.011) due mainly to low blanket coverage near the exhaust duct, and this solution should be revised. The CM Li{sub 17}Pb blanket reaches a sufficient TBR (1.059, with ports) to rely on tritium self-sufficiency. Shielding properties, with respect to the toroidal field coils, have been estimated in a simplified model by means of the ANISN code, supplied with a nuclear data library consistent with that used by MCNP. The analysis suggests that a careful shield thickness/composition design should be used to ensure the shielding capability of the whole blanket plus shield system. (orig.).

  18. Rapid starting methanol reactor system

    Science.gov (United States)

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

    1984-01-01

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

  19. Nuclear Reactors. Revised.

    Science.gov (United States)

    Hogerton, John F.

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-02-01

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