WorldWideScience

Sample records for ans reactor studies

  1. Plasma nitriding monitoring reactor: A model reactor for studying plasma nitriding processes using an active screen

    Science.gov (United States)

    Hamann, S.; Börner, K.; Burlacov, I.; Spies, H.-J.; Strämke, M.; Strämke, S.; Röpcke, J.

    2015-12-01

    A laboratory scale plasma nitriding monitoring reactor (PLANIMOR) has been designed to study the basics of active screen plasma nitriding (ASPN) processes. PLANIMOR consists of a tube reactor vessel, made of borosilicate glass, enabling optical emission spectroscopy (OES) and infrared absorption spectroscopy. The linear setup of the electrode system of the reactor has the advantages to apply the diagnostic approaches on each part of the plasma process, separately. Furthermore, possible changes of the electrical field and of the heat generation, as they could appear in down-scaled cylindrical ASPN reactors, are avoided. PLANIMOR has been used for the nitriding of steel samples, achieving similar results as in an industrial scale ASPN reactor. A compact spectrometer using an external cavity quantum cascade laser combined with an optical multi-pass cell has been applied for the detection of molecular reaction products. This allowed the determination of the concentrations of four stable molecular species (CH4, C2H2, HCN, and NH3). With the help of OES, the rotational temperature of the screen plasma could be determined.

  2. Plasma nitriding monitoring reactor: A model reactor for studying plasma nitriding processes using an active screen

    Energy Technology Data Exchange (ETDEWEB)

    Hamann, S., E-mail: hamann@inp-greifswald.de; Röpcke, J. [INP-Greifswald, Felix-Hausdorff-Str. 2, 17489 Greifswald (Germany); Börner, K.; Burlacov, I.; Spies, H.-J. [TU Bergakademie Freiberg, Institute of Materials Engineering, Gustav-Zeuner-Str. 5, 09599 Freiberg (Germany); Strämke, M.; Strämke, S. [ELTRO GmbH, Arnold-Sommerfeld-Ring 3, 52499 Baesweiler (Germany)

    2015-12-15

    A laboratory scale plasma nitriding monitoring reactor (PLANIMOR) has been designed to study the basics of active screen plasma nitriding (ASPN) processes. PLANIMOR consists of a tube reactor vessel, made of borosilicate glass, enabling optical emission spectroscopy (OES) and infrared absorption spectroscopy. The linear setup of the electrode system of the reactor has the advantages to apply the diagnostic approaches on each part of the plasma process, separately. Furthermore, possible changes of the electrical field and of the heat generation, as they could appear in down-scaled cylindrical ASPN reactors, are avoided. PLANIMOR has been used for the nitriding of steel samples, achieving similar results as in an industrial scale ASPN reactor. A compact spectrometer using an external cavity quantum cascade laser combined with an optical multi-pass cell has been applied for the detection of molecular reaction products. This allowed the determination of the concentrations of four stable molecular species (CH{sub 4}, C{sub 2}H{sub 2}, HCN, and NH{sub 3}). With the help of OES, the rotational temperature of the screen plasma could be determined.

  3. Numerical study of the effects of lamp configuration and reactor wall roughness in an open channel water disinfection UV reactor.

    Science.gov (United States)

    Sultan, Tipu

    2016-07-01

    This article describes the assessment of a numerical procedure used to determine the UV lamp configuration and surface roughness effects on an open channel water disinfection UV reactor. The performance of the open channel water disinfection UV reactor was numerically analyzed on the basis of the performance indictor reduction equivalent dose (RED). The RED values were calculated as a function of the Reynolds number to monitor the performance. The flow through the open channel UV reactor was modelled using a k-ε model with scalable wall function, a discrete ordinate (DO) model for fluence rate calculation, a volume of fluid (VOF) model to locate the unknown free surface, a discrete phase model (DPM) to track the pathogen transport, and a modified law of the wall to incorporate the reactor wall roughness effects. The performance analysis was carried out using commercial CFD software (ANSYS Fluent 15.0). Four case studies were analyzed based on open channel UV reactor type (horizontal and vertical) and lamp configuration (parallel and staggered). The results show that lamp configuration can play an important role in the performance of an open channel water disinfection UV reactor. The effects of the reactor wall roughness were Reynolds number dependent. The proposed methodology is useful for performance optimization of an open channel water disinfection UV reactor.

  4. Numerical study of the effects of lamp configuration and reactor wall roughness in an open channel water disinfection UV reactor.

    Science.gov (United States)

    Sultan, Tipu

    2016-07-01

    This article describes the assessment of a numerical procedure used to determine the UV lamp configuration and surface roughness effects on an open channel water disinfection UV reactor. The performance of the open channel water disinfection UV reactor was numerically analyzed on the basis of the performance indictor reduction equivalent dose (RED). The RED values were calculated as a function of the Reynolds number to monitor the performance. The flow through the open channel UV reactor was modelled using a k-ε model with scalable wall function, a discrete ordinate (DO) model for fluence rate calculation, a volume of fluid (VOF) model to locate the unknown free surface, a discrete phase model (DPM) to track the pathogen transport, and a modified law of the wall to incorporate the reactor wall roughness effects. The performance analysis was carried out using commercial CFD software (ANSYS Fluent 15.0). Four case studies were analyzed based on open channel UV reactor type (horizontal and vertical) and lamp configuration (parallel and staggered). The results show that lamp configuration can play an important role in the performance of an open channel water disinfection UV reactor. The effects of the reactor wall roughness were Reynolds number dependent. The proposed methodology is useful for performance optimization of an open channel water disinfection UV reactor. PMID:27108375

  5. A planning study project of developing an integral reactor (SMART)

    Energy Technology Data Exchange (ETDEWEB)

    Kang, C. S.; Kim, B. S.; Park, K. C. [Korea Nuclear Society, Taejeon (Korea)

    2001-06-01

    Since July of 1997, Korea Atomic Energy Research Institute (Korea Atomic Energy Research Institute) has been developing the conceptual and basic designs of a small- and medium- sized reactor called SMART (System integrated Modular Advanced ReacTor). SMART is a 330MWt integral type pressurized water reactor which will be used not only for electricity generation but also for cogeneration, district heating and seawater desalination. The conceptual design was already completed in March of 1999, and the basic design is on-going, which will be finished by March of 2002. In order for the Government to make a decision whether the project of developing SMART should continue to be supported beyond the design phase, before the completion of its basic design in March 2002, it has decided to go through a systematic interim assessment of the current status of the project of SMART development. Hence, the specific objectives of this planning study are to review the technical soundness of the present SMART design, estimate the cost of completing its development, assess the economic incentives of its development in the future, and make necessary recommendations for the successful completion of the SMART development based on the results of technical review and economic evaluation. Based upon the results of technical review of the present design of SMART such as design appropriateness, licensibility, and equipment manufacturability, in general, it is concluded that there exist no fundamental problems in its design. However, for smooth licensing progress and eventual commercialization in the future, the following issues should be addressed: There are some unresolved safety issues found in the current design under existing licensing criteria such as violations of diverse reactivity control, SG in-service inspection, etc., which should be somehow resolved by either employing appropriate design changes or modifying existing licensing criteria through extensive discussion with regulatory

  6. Slurry reactor design studies

    Energy Technology Data Exchange (ETDEWEB)

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

    1990-06-01

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

  7. An experimental study of a VVER reactor's steam generator model operating in the condensing mode

    Science.gov (United States)

    Morozov, A. V.; Remizov, O. V.

    2012-05-01

    Results obtained from an experimental study of a VVER reactor's steam generator model operating in the condensing mode are presented. The obtained empirical dependence for calculating the power of heat exchangers operating in the steam condensation mode is presented.

  8. Study of the catalyst deactivation in an industrial gasoil HDS reactor using a mini-scale laboratory reactor

    Energy Technology Data Exchange (ETDEWEB)

    L.E. Kallinikos; G.D. Bellos; N.G. Papayannako [National Technical University of Athens, Athens (Greece). School of Chemical Engineering

    2008-09-15

    The activity of a hydrodesulphurization catalyst loaded in an industrial hydrotreater is studied at start up and end of run. Catalyst initial and final activity was determined by performing HDS experiments at industrial conditions in a laboratory mini-scale hydrotreater. The results show that the deactivation of the catalyst samples collected from three different places of the industrial reactor do not vary significantly, the maximum difference among the catalyst samples, being less than {+-}4%. The experimentally determined deactivation level of the catalyst samples is compared with the deactivation estimated for the same industrial reactor and the same load using a hybrid neural network model trained with operational data of the industrial and the results are in close agreement. Catalyst deactivation appears to be faster for hydrogen consumption reactions than for hydrodesulphurization reactions indicating a decreasing hydrogen consumption trend with time in operation for specific sulphur content in the product. 21 refs., 9 figs., 4 tabs.

  9. Chernobyl reactor transient simulation study

    International Nuclear Information System (INIS)

    This paper deals with the Chernobyl nuclear power station transient simulation study. The Chernobyl (RBMK) reactor is a graphite moderated pressure tube type reactor. It is cooled by circulating light water that boils in the upper parts of vertical pressure tubes to produce steam. At equilibrium fuel irradiation, the RBMK reactor has a positive void reactivity coefficient. However, the fuel temperature coefficient is negative and the net effect of a power change depends upon the power level. Under normal operating conditions the net effect (power coefficient) is negative at full power and becomes positive under certain transient conditions. A series of dynamic performance transient analysis for RBMK reactor, pressurized water reactor (PWR) and fast breeder reactor (FBR) have been performed using digital simulator codes, the purpose of this transient study is to show that an accident of Chernobyl's severity does not occur in PWR or FBR nuclear power reactors. This appears from the study of the inherent, stability of RBMK, PWR and FBR under certain transient conditions. This inherent stability is related to the effect of the feed back reactivity. The power distribution stability in the graphite RBMK reactor is difficult to maintain throughout its entire life, so the reactor has an inherent instability. PWR has larger negative temperature coefficient of reactivity, therefore, the PWR by itself has a large amount of natural stability, so PWR is inherently safe. FBR has positive sodium expansion coefficient, therefore it has insufficient stability it has been concluded that PWR has safe operation than FBR and RBMK reactors

  10. Kinetic Study of COS with Tertiary Alkanolamine Solutions. 1. Experiments in an Intensely Stirred Batch Reactor

    NARCIS (Netherlands)

    Littel, Rob J.; Versteeg, Geert F.; Swaaij, Wim P.M. van

    1992-01-01

    The reaction between COS and various tertiary alkanolamines in aqueous solutions has been studied in an intensely stirred batch reactor. Experiments for TEA, DMMEA, and DEMEA were carried out at 303 K; the reaction between COS and aqueous MDEA has been studied at temperatures ranging from 293 to 323

  11. A study on fast reactor core mechanics by an ex-reactor test and comparisons with calculations

    International Nuclear Information System (INIS)

    This paper presents and discusses the results of core bowing experiments performed with an ex-reactor rig holding a half hexagon array of 22 sub-assemblies (S/As) simulating the Japanese DFBR conditions and the comparisons of the measured results with calculations by individually developed codes--ARKAS, RAINBOW, SANBOW. The main conclusions of this study are (1) interwrapper loads and S/A displacements within the array were measured at selected positions for a series of five tests simulating the DFBR core bowing modes, (2) the overall comparison between the non-friction calculation and measurement showed good agreement for loads, displacements and their directions, and (3) validation of the friction algorithm has also been carried out and further improvement of the agreement was obtained

  12. Mirror reactor surface study

    International Nuclear Information System (INIS)

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

  13. Thermionic Reactor Design Studies

    Energy Technology Data Exchange (ETDEWEB)

    Schock, Alfred

    1994-06-01

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

  14. Fast Breeder Reactor studies

    International Nuclear Information System (INIS)

    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

  15. 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.

  16. Study and modelling of an innovative coprecipitation reactor for radioactive liquid wastes decontamination

    International Nuclear Information System (INIS)

    In order to decontaminate radioactive liquid wastes of low and intermediate levels, the coprecipitation is the process industrially used. The aim of this PhD work is to optimize the continuous process of coprecipitation. To do so, an innovative reactor is designed and modelled: the continuous reactor/classifier. Two model systems are studied: the coprecipitation of strontium by barium sulphate and the sorption of cesium by PPFeNi. The simulated effluent contains sodium nitrate in order to consider the high ionic strength of radioactive liquid wastes. First, each model system is studied on its own, and then a simultaneous treatment is performed. The kinetic laws of nucleation and crystal growth of barium sulphate are determined and incorporated into the coprecipitation model. Kinetic studies and sorption isotherms of cesium by PPFeNi are also performed in order to acquire the necessary data for process modelling. The modelling realised enables accurate prediction of the residual strontium and cesium concentrations according to the process used: it is a valuable tool for the optimization of existing units, but also the design of future units. The continuous reactor/classifier presents many advantages compared to the classical continuous process: the decontamination efficiency of strontium and cesium is highly improved while the volume of sludge generated by the process is reduced. A better liquid/solid separation is observed in the reactor/classifier and the global installation is significantly more compact. Thus, the radioactive liquid wastes treatment processes can be intensified by the continuous reactor/classifier, which represents a very promising technology for future industrial application. (author)

  17. CFD modelling of flow mal-distribution in an industrial ammonia oxidation reactor: A case study

    International Nuclear Information System (INIS)

    Ammonia oxidation reactor is widely used in nitric acid plant to cause the catalytic reaction between air and ammonia to produce nitrous gases. In this work, the flow distribution inside the ammonia oxidation reactor at Shiraz Petrochemical Complex (SPC) has been simulated using Computational Fluid Dynamics (CFD) code. The CFD results showed that the flow is non-uniformly distributed inside the reactor due to improper header design of the reactor. Measuring of the temperature distribution around the skin of the reactor has been carried out using thermograph. The thermograph experiment showed a considerable temperature difference between the left and right side of the reactor. It was found that the mal-distribution of the gas flow inside the reactor can directly affect the performance of the reactor. - Highlights: •A failure has been observed in an industrial ammonia oxidation reactor. •CFD code helps to simulate the flow inside the reactor. •The flow becomes non-uniformly distributed due to the reactor header mal-design. •The flow mal-distribution results in some drawbacks

  18. Techniques in Gas-Phase Thermolyses. Part 6. Pulse Pyrolysis: Gas Kinetic Studies in an Inductively Heated Flow Reactor

    DEFF Research Database (Denmark)

    Egsgaard, Helge; Bo, P.; Carlsen, Lars

    1985-01-01

    A prototype of an inductively heated flow reactor for gas kinetic studies is presented. The applicability of the system, which is based on a direct coupling between the reactor and the ion source of a mass spectrometer, is illustrated by investigations of a series of simple bond fission reactions...

  19. Studies on reactor physics

    International Nuclear Information System (INIS)

    Most of the peaceful applications of atomic energy are inherently dependent on advances in the science and technology of nuclear reactors, and aspects of this development are part of a major programme of the International Atomic Energy Agency. The most useful role that the Agency can play is as a co-ordinating body or central forum where the trends can be reviewed and the results assessed. Some of the basic studies are carried out by members of the Agency's own scientific staff. The Agency also convenes groups of experts from different countries to examine a particular problem in detail and make any necessary recommendations. Some of the important subjects are discussed at international scientific meetings held by the Agency. One of the subjects covered by such studies is the physics of nuclear reactors and a specific topic recently discussed was Codes for Reactor Computations, on which a seminar was held in Vienna in April this year. Another The members of the Panel described the development of heavy water reactors, the equipment and methods of research currently used, and plans for further development in their respective countries meeting of Panel of Experts on Heavy Water Lattices was held in Vienna in August 1959

  20. Nordic study on reactor waste

    International Nuclear Information System (INIS)

    In 1981, 14 nuclear power reactors are in operation and 2 under construction in the Nordic countries. So far, the reactor waste originating from day-to-day operation of these plants has been stored in solidified form at the reactor sites. Within a few years a satisfactory disposal procedure needs to be established. While the main R and D effects in the waste field have earlier been devoted to the question of irradiated fuel and waste from reprocessing, there is therefore now an increased interest in reactor waste with its much lower radioactivity but somewhat larger volumes. Since 1977, efforts have been made in a joint Nordic study to examine which facts need to be known in order to perform a comprehensive safety assessment of a reactor waste management system. In the present study a Reference system related to the waste generated over 30 years from six 500 MW-reactors is examined. The dominating radionuclides during storage and transportation accident scenarios are Cs-134, Cs-137 and Co-60. For most of the release scenarios from repositories Cs-137 and Sr-90 are dominating. Some scenarios are, however, dominated by the very longlived nuclides I-129 and C-14. A closer examination of the concentration in the waste of these nuclides and of their leaching properties indicates that their small - but significant - influence, as calculated, is probably grossly overestimated. The mechanical stability obtained in routine solidification processes of reactor waste products in conjunction with the outer container (steel drum, transport container, etc.) turns out to be sufficient. Difficulties were encountered in applying ICRP methodology and available dose calculation methods to calculation of population doses due to small activity releases, and effects extending into the far future. (EG)

  1. Operator reliability study for Probabilistic Safety Analysis of an operating research reactor

    International Nuclear Information System (INIS)

    Highlights: • Human Reliability Analysis (HRA) for Level 1 Probabilistic Safety Analysis (PSA) is performed on research nuclear reactor. • Implemented qualitative HRA framework is addressed. • Human Failure Events of significant impact to the reactor safety are derived. - Abstract: A Level 1 Probabilistic Safety Analysis (PSA) for the TRIGA Mark II research reactor of Malaysian Nuclear Agency has been developed to evaluate the potential risk in its operation. In conjunction to this PSA development, Human Reliability Analysis (HRA) is performed in order to determine human contribution to the risk. The aim of this study is to qualitatively analyze human actions (HAs) involved in the operation of this reactor according to the qualitative part of the HRA framework for PSA which is namely the identification, qualitative screening and modeling of HAs. By performing this framework, Human Failure Events (HFEs) of significant impact to the reactor safety are systematically analyzed and incorporated into the PSA structure. A part of the findings in this study will become the input for the subsequent quantitative part of the HRA framework, i.e. the Human Error Probability (HEP) quantification

  2. Helias reactor studies

    International Nuclear Information System (INIS)

    The Helias reactor is an upgraded version of the Wendelstein 7-X experiment. The magnetic field has 5 field periods and the main optimization principle is the reduction of the Pfirsch-Schlueter currents and the Shafranov shift, which has been verified by computations with the NEMEC and MFBE-codes. The modular coil system comprises 50 coils, which are constructed using NbTi-superconducting cables. The basic dimensions are: major radius 22 m, average plasma radius 1.8 m, magnetic field on axis 5 T, maximum field on the coils 10 T. Forces and stresses in the coil system have been investigated with the aid of the ANSYS code, which found maximum stress values of about 650 MPa in the coil casing. Helias configurations with 4 and 3 field periods have been constructed by starting from the 5-period case and by eliminating one or two periods while the shape of the coils is kept nearly invariant. In a first survey blanket concepts, developed for the DEMO tokamak, have been adapted to the Helias geometry, in particular, the solid breeder concept developed by FZK (Karlsruhe) has been extrapolated to the Helias geometry identifying the drawbacks and advantages of this concept. Furthermore, the liquid breeder concept using Li7-Pb83 and water-cooling is an interesting alternative for the Helias reactor. Maintenance of blanket and plasma facing components is possible through the portholes between modular coils. Numerical simulations of the start-up phase of the Helias reactor using the TOTAL-P code have confirmed the zero-dimensional modeling of the fusion plasma with the aid of empirical scaling laws. (author)

  3. An Experimental Study of Natural Convection in The Hottest Channel of TRIGA 2000 k W Reactor

    International Nuclear Information System (INIS)

    With the increase of radioisotope demand, in 1995, National Nuclear Energy Agency of Indonesia made a decision to upgrade the power of the TRIGA Mark II reactor from 1 MW to 2 MW maximum power. The reactor reached its first criticality on May 13, 2000. To accomplish the safety evaluation of the reactor, a thermal hydraulic analysis was carried out by using thermal hydraulic computer code. This code calculates the natural convection flow through water coolant bounded by vertical cylindrical heat sources. In this paper, it will be reported the experimental study of natural convection in the hottest channel of TRIGA 2000 k W reactor. The purpose of the experimental study is to verify the theoretical analysis, especially the temperature distribution in the hottest coolant channel. In this experiment, a special probe for temperature detection has been designed and inserted to central thimble (CT). In the experiment, eight thermocouples were used to measure the bulk temperature of the water at different position in the cooling channel and simultaneous quantitative measurement of the temperature distribution were done by using a data acquisition cards system. The result obtained theoretically using the STAT code has been verified by this experimental study. (author)

  4. Parametric studies of tandem mirror reactors

    International Nuclear Information System (INIS)

    This report, along with its companion, An Improved Tandem Mirror Reactor, discusses the recent progress and present status of our tandem mirror reactor studies. This report presents the detailed results of parametric studies up to, but not including, the very new ideas involving thermal barriers

  5. Study on the use of slightly enriched uranium fuel cycle in an existing CANDU 6 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yeom, Choong Sub; Kim, Hyun Dae [Institute for Advanced Engineering, Seoul (Korea, Republic of)

    1997-12-31

    To test the viability of CANFLEX-SEU bundles in an existing CANDU 6 reactor, core follow-up simulation has been carried out using the reactor fueling simulation program of the CANDU 6, RFSP computer code, and a lattice physics code, WIMS-AECL. During the core follow-up, bundle and channel powers and zone levels have been checked against their operating limits at each simulation. It is observed from the simulation results that an equilibrium core loaded with 0.9 w/o CANFLEX-SEU bundles could be refueled and maintained for 550 FPD without any significant violations in the channel and bundle power limits and the permissible operating range of the liquid zone controllers. 8 refs., 2 figs., 1 tab. (Author)

  6. Co-combustion of coal and SRF in an entrained flow reactor: a preliminary study

    DEFF Research Database (Denmark)

    Wu, Hao; Glarborg, Peter; Frandsen, Flemming;

    2009-01-01

    Investigations on co-firing of SRF with two kinds of bituminous coal were carried out in an entrained flow reactor. The experimental results showed that co-combustion of coal and SRF increased the unburnt carbon in fly ashes. The emissions of NO and SO2 were reduced with an increasing share of SRF...... slightly with an increasing share of SRF. For SAKLEI coal and SRF cocombustion, the deposit formation rate showed an increasing trend up to 10 mass percent of SRF, and started to decrease at a higher SRF share. By analyzing the ash samples, it has been found that the concentrations of some trace elements...

  7. An analytic study on LBLOCA for CANDU type reactor using MARS-KS/CANDU

    International Nuclear Information System (INIS)

    This study provides the simulation results using MARS-KS/CANDU code for the Large Break LOCA of CANDU type reactor. The purpose of the study is to evaluate the capability of MARS-KS/CANDU for simulating the actual plants (Wolsong 2/3/4). The steady state and the transient analysis results were provided. After the sensitivity study depend on break size, the case that 35% of the inlet header known as the accident that has the most limiting effect on the temperature of the fuel sheath was calculated. In order to evaluate the results, the results were compared with those of CATHENA simulation. (author)

  8. Study of a loss of feedwater transient for an integral pressurized water reactor with a helical coil steam generator

    International Nuclear Information System (INIS)

    The Multi-Application Small Light Water Reactor (MASLWR) is a system-level test facility constructed by Oregon State University to examine the thermal hydraulic phenomena that are of importance to integral pressurized water reactors (IPWRs). These phenomena include natural circulation instability, helical coil steam generator (HCSG) heat transfer and coupled reactor-containment pressurization. In MASLWR, the steam generator, the pressurizer and the reactor coolant system are integrated in one reactor pressure vessel unit and a metal slab is used to conduct the heat transfer between the containment vessel and the cooling pool. MASLWR is a scaled model of the NuScale design, a small modular reactor (SMR), with 1:3 length scale, 1:254 volume scale, 1:1 time scale, and full operating pressure and temperature. An IAEA International Collaborative Standard Problem (ICSP) was conducted using MASLWR to generate experimental data for various natural circulation conditions. In the ICSP, two experiments were conducted – “Loss of feedwater transient with subsequent automatic depressurization system operation” and “Normal operating conditions at different power levels”. US Nuclear Regulatory Commission participated in this standard problem in order to validate the TRACE MASLWR model and to identify any potential challenges to the code in simulating this unique design. The aim of the study is to support the future licensing review of the NuScale-like integral reactor design. In this paper, the loss of feedwater transient simulation will be discussed. The modeling strategy for the IPWR's unique components, such as the helical coil steam generator and the coupled reactor and containment, will be described in details. TRACE results show qualitative agreement with the experimental data in the major parameters of interest. However, one important event time of the calculation - the pressure equalization between the reactor and the containment - deviates from the data

  9. Thermionic Reactor Design Studies

    Energy Technology Data Exchange (ETDEWEB)

    Schock, Alfred

    1994-08-01

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

  10. HIBALL-II - an improved conceptual heavy ion beam driven fusion reactor study

    International Nuclear Information System (INIS)

    An improved design of the HIBALL inertial-confinement fusion power station is presented. The new RF-linac based heavy ion driver has improved concepts for beam stacking, bunching and final focusing. The new target design takes into account radiation transport effects in a coarse approximation. The system of four reactors with a net total output of 3.8 GW electric is essentially the same as described earlier, however, progress in the analysis has enhanced its credibility and self-consistency. Considerations of environmental and safety aspects and cost estimates are given. (orig.)

  11. A Neutronic Feasibility Study on the Recycling of an Oxide Fuel in Sodium-Cooled Fast Reactors

    International Nuclear Information System (INIS)

    Neutronic feasibility was implemented for the recycling of a mixed oxide fuel in sodium-cooled fast reactors (SFR) through a thermal/mechanical dry process, which is recognized as one of the most proliferation- resistant recycling processes. In order to assess the applicability of a simple dry process which is not capable of completely removing all the fission products from a spent fuel, sensitivity calculations were performed for the reactor physics parameters with a dependency on the fission product removal rate of the recycled spent fuel. The equilibrium core calculations were performed by the REBUS-3 code for a BN-600 core without blanket fuels and a modified core with an increased fuel volume fraction. The reactor performance parameters such as the transuranic content, breeding ratio, peak linear power, burnup reactivity swing and reactivity coefficients were calculated for an equilibrium core under a fixed fuel management scheme. The results showed that a recycling of the oxide fuel in the SFR is feasible if the fission products are removed by more than 70% through the dry process as far as the material balance is concerned. However the physics analysis also showed that some of the physics design parameters are slightly deteriorated. The results of this study indicate that the recycling characteristics can be improved if the dry process can remove more fission products, and the reactor configuration is further optimized or the spent fuel composition is adjusted

  12. Stellarator fusion reactors - an overview

    International Nuclear Information System (INIS)

    The stellarator system offers a distinct alternative to the mainline approaches to magnetic fusion power and has several potentially major advantages. Since the first proposal of the stellarator concept many reactor studies have been published and these studies reflect the large variety of stellarator configurations. The main representatives are the continuous-coil configurations and the modular-coil configurations. As a continuation of the LHD experiment two reactor configurations, FFHR1 and FFHR2, have been investigated, which use continuous helical windings for providing the magnetic field. The modular coil concept has been realized in the MHH-reactor study (USA 1997) and in the Helias reactor. The Helias reactor combines the principle of plasma optimisation with a modular coil system. The paper also discusses the issues associated with the blanket and the maintenance process. Stellarator configurations with continuous coils such as LHD possess a natural helical divertor, which can be used favourably for impurity control. In advanced stellarators with modular coils the same goal can be achieved by the island divertor. Plasma parameters in the various stellarator reactors are computed on the basis of presently known scaling laws showing that confinement is sufficiently good to provide ignition and self-sustained burn. (author)

  13. An experimental study on in-vessel retention strategy by external reactor vessel cooling with liquid metal

    International Nuclear Information System (INIS)

    The present work ultimately aims to develop the IVR-ERVC (In-Vessel Retention through External Reactor Vessel Cooling) system with enough thermal margin adopting liquid metal coolant as the severe accident mitigation system even for high power reactor. For the purpose, the conceptual design of IVR-ERVC with liquid metal is evaluated by performing an experimental campaign for a scaled facility. The specific geometry was devised to contain the liquid metal coolant facing with water through the container wall. Through this system, the heat transfer area is enlarged up to 2 times compared to the original area of the reactor vessel. This effect is also named as 'liquid metal fin' in the current study. Improved heat transfer or reduced heat flux including large drop of focusing effect was confirmed by experimental results for a small-scaled facility to simulate the boiling phenomena under IVR-ERVC condition. It was found that significant reduction of focusing effect by liquid metal and extended surface area guarantee enough margin of successful IVR-ERVC without CHF issue even for large-sized power reactors. (author)

  14. An experimental and modeling study of propene oxidation. Part 1: Speciation measurements in jetstirred and flow reactors

    OpenAIRE

    Burke, Sinéad,; Metcalfe, Wayne; Herbinet, Olivier; Battin-Leclerc, Frédérique; Haas, Francis,; Santner, Jeffrey; Dryer, Frederick; J. Curran, Henry

    2014-01-01

    International audience Propene is a significant component of Liquefied Petroleum Gas (LPG) and an intermediate in the combustion of higher order hydrocarbons. To better understand the combustion characteristics of propene, this study and its companion paper present new experimental data from jet-stirred (JSR) and flow reactors (Part I) and ignition delay time and flame speed experiments (Part II). Species profiles from JSR experiments are presented and were obtained at near-atmospheric pre...

  15. Progress in Helias reactor studies

    International Nuclear Information System (INIS)

    The Helias reactor is an upgraded version of the Wendelstein 7-X experiment, which is under construction in the city of Greifswald. The modular coil system comprises 50 coils, which are constructed using NbTi-superconducting cables. The basic dimensions are: major radius 22 m, average plasma radius 1.8 m, magnetic field on axis 5 T, maximum field on coils 10 T. Over the past year progress toward better understanding of the fusion plasma has been made. In particular, the following issues have been addressed: Plasma equilibrium and MHD-stability; Neoclassical transport in the Helias configuration; Start-up scenarios and steady state burn; Alpha-particle orbits and alpha-particle losses; Drift waves in Helias configurations; Modelling the fusion plasma using empirical scaling laws Technical studies have been focussed on the optimization of the coil system with respect to magnetic field distribution, forces and stresses. In this context the ANSYS-code has been found useful for optimising the support system. In a first survey several blanket concepts, developed for the DEMO tokamak, have been adapted to the Helias geometry. Presently a water-cooled LiPb-blanket is favored in comparison with ceramic breeders, since safety properties and maintenance procedure seem to be more advantageous within this concept. Maintenance and replacement of blanket segments through portholes have also been studied with respect to their geometric compatibility. Finally parameter studies of low aspect ratio Helias reactors will be discussed. (orig.)

  16. A reverse flow catalytic membrane reactor for the production of syngas: an experimental study

    NARCIS (Netherlands)

    Smit, J.; Bekink, G.J.; Sint Annaland, van M.; Kuipers, J.A.M.

    2005-01-01

    In this paper experimental results are presented for a demonstration unit of a recently proposed novel integrated reactor concept (Smit et. al., 2005) for the partial oxidation of natural gas to syngas (POM), namely a Reverse Flow Catalytic Membrane Reactor (RFCMR). Natural gas has great potential a

  17. A computational fluid dynamics study of hydrogen bubbles in an electrochemical reactor

    Directory of Open Access Journals (Sweden)

    Renata da Silva Cavalcanti

    2005-06-01

    Full Text Available Most electrochemical reactors present reactions with the growth and departure of gas bubbles which influence on the reactor hydrodynamics and this study is usually complex, representing a vast field for research. The present paper had as objective to study a bi-phase (gas-liquid system aiming to foresee the influence of departure of hydrogen bubbles generated on effective electrode surface situated on cathodic semi-cell. Nevertheless, it was idealized that the gas was injected into the semi cell, through the effective electrode surface With this hypothesis, it was possible to study, and numerically analyze, the hydrodynamic behavior of the hydrogen bubbles in the interior of the study domain, applying concepts of computational fluid dynamics by using the computational applicative CFX-4 for the application of the MUSIG ("MUltiple-SIze-Group" model, taking into consideration the phenomena of coalescence and the distribution of the diameter of the bubbles.A maioria dos reatores eletroquímicos apresenta reações com crescimento e desprendimento de bolhas de gás influenciando na hidrodinâmica dos reatores e seu estudo é, geralmente, complexo representando um campo amplo para pesquisas. O presente artigo teve por objetivo estudar um sistema bifásico (gás-líquido visando prever a influência do desprendimento das bolhas de hidrogênio geradas na superfície específica do eletrodo localizada na semicélula catódica. No entanto, foi idealizado que o gás fora injetado no interior da semicélula através da superfície específica do eletrodo. Com esta hipótese, foi possível estudar e analisar numericamente o comportamento hidrodinâmico das bolhas de hidrogênio no interior do domínio de estudo, aplicando-se os conceitos de fluidodinâmica computacional usando o aplicativo computacional CFX-4 para aplicação do modelo MUSIG ("Multiple-size-group" levando em consideração os fenômenos da coalescência e da distribuição do diâmetro das

  18. Reactor based plutonium disposition - physics and fuel behaviour benchmark studies of an OECD/NEA experts group

    Energy Technology Data Exchange (ETDEWEB)

    D' Hondt, P. [SCK.CEN, Mol (Belgium); Gehin, J. [ORNL, Oak Ridge, TN (United States); Na, B.C.; Sartori, E. [Organisation for Economic Co-Operation and Development, Nuclear Energy Agency, 92 - Issy les Moulineaux (France); Wiesenack, W. [Organisation for Economic Co-Operation and Development/HRP, Halden (Norway)

    2001-07-01

    One of the options envisaged for disposing of weapons grade plutonium, declared surplus for national defence in the Russian Federation and Usa, is to burn it in nuclear power reactors. The scientific/technical know-how accumulated in the use of MOX as a fuel for electricity generation is of great relevance for the plutonium disposition programmes. An Expert Group of the OECD/Nea is carrying out a series of benchmarks with the aim of facilitating the use of this know-how for meeting this objective. This paper describes the background that led to establishing the Expert Group, and the present status of results from these benchmarks. The benchmark studies cover a theoretical reactor physics benchmark on a VVER-1000 core loaded with MOX, two experimental benchmarks on MOX lattices and a benchmark concerned with MOX fuel behaviour for both solid and hollow pellets. First conclusions are outlined as well as future work. (author)

  19. What drives innovation in nuclear reactors technologies? An empirical study based on patent counts

    International Nuclear Information System (INIS)

    This paper examines the evolution of innovation in nuclear power reactors between 1974 and 2008 in twelve OECD countries and assesses to what extent nuclear innovation has been driven by economic incentives, political decisions and safety regulation considerations. We use priority patent applications related to Nuclear Power Plants (NPPs) as a proxy for innovating activity. Our results highlight that nuclear innovation is partly driven by the conventional paradigm where both demand-pull, measured by NPPs constructions in the innovating country and in the rest of the world, and technology-push, measured by Research and Development (R and D) expenditures specific to NPPs, have a positive and significant impact on innovation. Our results also evidence that the impact of public R and D expenditures and national NPPs construction on innovation is stronger when the quality of innovation, measured by forward patent citations, is taken into account, and have a long run positive impact on innovation through the stock of knowledge available to innovators. In contrast, we show that political decisions following the Three Miles Island and Chernobyl nuclear accidents, measured by NPPs cancellations, have a negative impact on nuclear innovation. Finally, we find that the nuclear safety authority has an ambivalent effect on innovation. On one hand, regulatory inspections have a positive impact on innovation, one the other hand, regulatory decisions to temporarily close a NPP have an adverse impact on innovation. (author)

  20. An Experimental Study on the Breakup of Simulant Melt Jet Released from the Submerged Reactor Vessel Lower Head

    International Nuclear Information System (INIS)

    The potential risk of explosive molten fuel coolant interactions (FCI, steam explosion) has drawn substantial attention in the safety analysis of reactor severe accidents. The steam explosion intensity is largely dependent upon the degree of volumetric fractions of melt droplets and steam in the fuel-coolant mixture. The rate of melt jet breakup and droplet sizes are, therefore, the key physical parameters in the analysis of FCIs. An ex-vessel steam explosion may occur when the core melt is released from the failed reactor vessel lower head into the water-filled reactor cavity. The water level in the cavity can be either below or above the reactor vessel lower head depending on the severe accident management strategy. The former, a partially-filled cavity with free-fall space for the melt jet, has been the major condition for the steam explosion studies in the past. An In-Vessel Retention by External Vessel Cooling (IVR-ERVC), however, requires the water level in the cavity be above the reactor vessel lower head so that the vessel can be completely submerged in coolant water. In this case, the melt jet falls in liquid water without free fall. The jet breakup behavior in such a condition has been rarely studied, nor well identified. In this work, jet breakup of the melt released from the submerged vessel has been experimentally investigated using stimulant melt of Woods metal. The initial melt temperature was set below the boiling point of water so that only the hydrodynamic mechanism of jet breakup can be identified. High-speed videos were taken to visualize the jet breakup behavior and the post-test debris were collected and sieved to obtain debris size distributions. Non-boiling liquid jet breakup experiment was conducted using 50 mm-diameter Woods metal jet released from a vessel submerged in water pool. The post-test debris was sieved and the debris size distributions were obtained. The size of the largest mass of the debris shows a fair agreement with the

  1. On fast reactor kinetics studies

    Energy Technology Data Exchange (ETDEWEB)

    Seleznev, E. F.; Belov, A. A. [Nuclear Safety Inst. of the Russian Academy of Sciences IBRAE (Russian Federation); Matveenko, I. P.; Zhukov, A. M.; Raskach, K. F. [Inst. for Physics and Power Engineering IPPE (Russian Federation)

    2012-07-01

    The results and the program of fast reactor core time and space kinetics experiments performed and planned to be performed at the IPPE critical facility is presented. The TIMER code was taken as computation support of the experimental work, which allows transient equations to be solved in 3-D geometry with multi-group diffusion approximation. The number of delayed neutron groups varies from 6 to 8. The code implements the solution of both transient neutron transfer problems: a direct one, where neutron flux density and its derivatives, such as reactor power, etc, are determined at each time step, and an inverse one for the point kinetics equation form, where such a parameter as reactivity is determined with a well-known reactor power time variation function. (authors)

  2. Feasibility study for fast reactor and related fuel cycle. Preliminary studies in 1998

    International Nuclear Information System (INIS)

    Prior to the feasibility study for fast reactors (FRs) starting from the 1999 fiscal year, planned in the medium and long-term program of JNC, preliminarily studies were performed on 'FR systems except sodium cooled MOX fueled reactors'. Small scale or module type reactors, heavy metal (Pb or Pb-Bi) cooled reactors, gas cooled reactors, light water cooled reactors, and molten salt reactors were studied on the basis of literature. They were evaluated from the viewpoint of the technical possibility (the structure integrity, earthquake resistance, safety, productivity, operability, maintenance repair, difficulty of the development), the long-term targets (market competitiveness as an energy system, utilization of uranium resources, reduction of radioactive waste, security of the non-proliferation), and developmental risk. As the result, the following concepts should be studied for future commercialized FRs. Small scale and module type reactor: Middle-sized reactor with an excellent economical efficiency. Small power reactor with a multipurpose design concept. Gas cooled reactor: CO2 gas cooled reactor, He gas cooled reactor. Heavy metal cooled reactor: Russian type lead cooled reactor. Light water cooled reactor: Light water cooled high converter reactor and super critical pressure light water cooled reactor. Molten salt reactor: Trichloride molten salt reactor which matches the U-Pu cycle. (author)

  3. Reactor safety study. An assessment of accident risks in U.S. commercial nuclear power plants. Appendix XI. Analysis of comments on the draft WASH-1400 report

    International Nuclear Information System (INIS)

    Information is presented concerning comments on reactor safety by governmental agencies and civilian organizations; reactor safety study methodology; consequence model; probability of accident sequences; and various accident conditions

  4. Hybrid Molten Salt Reactor (HMSR) System Study

    Energy Technology Data Exchange (ETDEWEB)

    Woolley, Robert D [PPPL; Miller, Laurence F [PPPL

    2014-04-01

    Can the hybrid system combination of (1) a critical fission Molten Salt Reactor (MSR) having a thermal spectrum and a high Conversion Ratio (CR) with (2) an external source of high energy neutrons provide an attractive solution to the world's expanding demand for energy? The present study indicates the answer is an emphatic yes.

  5. An experimental low-pressure facility to study boron transients in the pressurizer of an integral modular nuclear reactor

    International Nuclear Information System (INIS)

    Small and medium size modular reactors offer many advantages when compared with typical nuclear plants in various circumstances, such as offering greater simplicity of design, economy of mass production, and reducing siting costs. The integral configuration is characterized by having most of its components inside the pressure vessel, eliminating the probability of accidents. However, for this configuration there is no spray system for boron homogenization, which may cause power transients. Thus, it is necessary to investigate boron mixing. The Federal University of Pernambuco (UFPE), in a partnership with the Regional Center of Nuclear Sciences of Northeast (CRCN-NE) and the Nuclear Engineering Institute (IEN/CNEN-RJ), is developing a project that aims to analyze transients in a compact modular integral reactor. This analysis will be made by using the data obtained from one experimental bench that is mounted at CRCN-NE. A study accomplished in 2012 using a simplified bench (built in reduced scale with a test section manufactured with transparent acrylic) showed that it was possible to obtain preliminary experimental results for the boron homogenizing process. (author)

  6. The development and use of a laboratory scale reactor to study aspects of gasification in an air blown fluidised bed

    Energy Technology Data Exchange (ETDEWEB)

    Cousins, A.; Zhuo, Y.; Reed, G.P.; Paterson, N.; Dugwell, D.R.; Kandiyoti, R. [Imperial College London, London (United Kingdom). Dept of Chemical Engineering

    2006-07-01

    A laboratory scale reactor has been used to study aspects of air blown, spouted bed gasifiers. The effects of operating conditions on the release of fuel-N has been studied using both coal and sewage sludge. The work has clarified the reactions involved and shown that steam has an important effect on the formation of NH{sub 3} from both volatile-N and char-N. The HCN concentration depends strongly on the residence time at temperature and on the presence (and depth) of a char bed. Trace element results indicate that bed temperatures above 900{sup o}C enhanced depletion of Ba, Pb and Zn from the bed residue and their enrichment in the fines. Mercury and selenium were released and their subsequent capture required low temperature filters operating below 120{sup o}C. The reactor was modified to enable char samples to be prepared and collected under controlled conditions. Results show the decreasing reactivity of the char with increasing temperature, time, pressure and particle size. There appears to be an initial decrease in reactivity during pyrolysis and a further longer- term decrease caused by graphitisation. 10 refs., 8 figs., 6 tabs.

  7. Fast breeder reactors an engineering introduction

    CERN Document Server

    Judd, A M

    1981-01-01

    Fast Breeder Reactors: An Engineering Introduction is an introductory text to fast breeder reactors and covers topics ranging from reactor physics and design to engineering and safety considerations. Reactor fuels, coolant circuits, steam plants, and control systems are also discussed. This book is comprised of five chapters and opens with a brief summary of the history of fast reactors, with emphasis on international and the prospect of making accessible enormous reserves of energy. The next chapter deals with the physics of fast reactors and considers calculation methods, flux distribution,

  8. The ARIES-I tokamak reactor study

    International Nuclear Information System (INIS)

    This report contains an overview of the Aries-I tokamak reactor study. The following topics are discussed on this tokamak: Systems studies; equilibrium, stability, and transport; summary and conclusions; current drive; impurity control system; tritium systems; magnet engineering; fusion-power-core engineering; power conversion; Aries-I safety design and analysis; design layout and maintenance; and start-up and operations

  9. Kinetic study of treatment of wastewater contains food preservative agent by anaerobic baffled reactor : An overview

    Energy Technology Data Exchange (ETDEWEB)

    Sumantri, Indro; Purwanto,; Budiyono [Chemical Engineering Department, Faculty of Engineering, Diponegoro University Jl. Prof. H. Soedarto, SH, Kampus Baru Tembalang, Semarang (Indonesia)

    2015-12-29

    The characteristic of wastewater of food industries with preservative substances is high content of organic substances, degradable and high total suspended solid. High organic content in this waste forced the treatment is biologically and pointed out to anaerobic treatment. Anaerobic showed the better performance of degradation than aerobic for high content organic and also for toxic materials. During that day the treatment of food wastewater is aerobically which is high consume of energy required and high volume of sludge produced. The advantage of anaerobic is save high energy, less product of sludge, less requirement of nutrients of microorganism and high efficiency reduction of organic load. The high efficiency of reduction will reduce the load of further treatment, so that, the threshold limit based on the regulation would be easy to achieve. Research of treatment of wastewater of food industries would be utilized by both big scale industries and small industries using addition of preservative substances. The type reactor of anaerobic process is anaerobic baffled reactor that will give better contact between wastewater and microorganism in the sludge. The variables conducted in this research are the baffled configuration, sludge height, preservative agent contents, hydralic retention time and influence of micro nutrients. The respons of this research are the COD effluent, remaining preservative agent, pH, formation of volatile fatty acid and total suspended solid. The result of this research is kinetic model of the anaerobic baffled reactor, reaction kinetic of preservative agent degradation and technology of treatment wastewater contains preservative agent. The benefit of this research is to solve the treatment of wastewater of food industries with preservative substance in order to achieve wastewater limit regulation and also to prevent the environmental deterioration.

  10. Kinetic study of treatment of wastewater contains food preservative agent by anaerobic baffled reactor : An overview

    Science.gov (United States)

    Sumantri, Indro; Purwanto, Budiyono

    2015-12-01

    The characteristic of wastewater of food industries with preservative substances is high content of organic substances, degradable and high total suspended solid. High organic content in this waste forced the treatment is biologically and pointed out to anaerobic treatment. Anaerobic showed the better performance of degradation than aerobic for high content organic and also for toxic materials. During that day the treatment of food wastewater is aerobically which is high consume of energy required and high volume of sludge produced. The advantage of anaerobic is save high energy, less product of sludge, less requirement of nutrients of microorganism and high efficiency reduction of organic load. The high efficiency of reduction will reduce the load of further treatment, so that, the threshold limit based on the regulation would be easy to achieve. Research of treatment of wastewater of food industries would be utilized by both big scale industries and small industries using addition of preservative substances. The type reactor of anaerobic process is anaerobic baffled reactor that will give better contact between wastewater and microorganism in the sludge. The variables conducted in this research are the baffled configuration, sludge height, preservative agent contents, hydralic retention time and influence of micro nutrients. The respons of this research are the COD effluent, remaining preservative agent, pH, formation of volatile fatty acid and total suspended solid. The result of this research is kinetic model of the anaerobic baffled reactor, reaction kinetic of preservative agent degradation and technology of treatment wastewater contains preservative agent. The benefit of this research is to solve the treatment of wastewater of food industries with preservative substance in order to achieve wastewater limit regulation and also to prevent the environmental deterioration.

  11. Kinetic study of treatment of wastewater contains food preservative agent by anaerobic baffled reactor : An overview

    International Nuclear Information System (INIS)

    The characteristic of wastewater of food industries with preservative substances is high content of organic substances, degradable and high total suspended solid. High organic content in this waste forced the treatment is biologically and pointed out to anaerobic treatment. Anaerobic showed the better performance of degradation than aerobic for high content organic and also for toxic materials. During that day the treatment of food wastewater is aerobically which is high consume of energy required and high volume of sludge produced. The advantage of anaerobic is save high energy, less product of sludge, less requirement of nutrients of microorganism and high efficiency reduction of organic load. The high efficiency of reduction will reduce the load of further treatment, so that, the threshold limit based on the regulation would be easy to achieve. Research of treatment of wastewater of food industries would be utilized by both big scale industries and small industries using addition of preservative substances. The type reactor of anaerobic process is anaerobic baffled reactor that will give better contact between wastewater and microorganism in the sludge. The variables conducted in this research are the baffled configuration, sludge height, preservative agent contents, hydralic retention time and influence of micro nutrients. The respons of this research are the COD effluent, remaining preservative agent, pH, formation of volatile fatty acid and total suspended solid. The result of this research is kinetic model of the anaerobic baffled reactor, reaction kinetic of preservative agent degradation and technology of treatment wastewater contains preservative agent. The benefit of this research is to solve the treatment of wastewater of food industries with preservative substance in order to achieve wastewater limit regulation and also to prevent the environmental deterioration

  12. 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.

  13. Mirror Advanced Reactor Study interim design report

    International Nuclear Information System (INIS)

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

  14. Mirror Advanced Reactor Study interim design report

    Energy Technology Data Exchange (ETDEWEB)

    1983-04-01

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

  15. Design study of an armor tile handling manipulator for the Fusion Experimental Reactor

    International Nuclear Information System (INIS)

    A conceptual design of the Fusion Experimental Reactor (FER), which is a D-T burning reactor following on JT-60 in Japan, has been developed by Japan Atomic Energy Research Institute (JAERI). In FER, a rail-mounted vehicle concept is planned to be adopted for in-vessel maintenance, such as maintenance of divertor plates and armor tiles. Advantages of this concept are the high stiffness of the rail as a base structure for maintenance and the high mobility of the vehicle along the rail. Twin armor tile handling manipulators installed on both sides of the vehicle have been designed. The respective manipulators for armor tile handling have 8 degrees of freedom in order to have access to any place of the first wall and to go through the horizontal port by operating manipulator joints. If the two types of manipulators for divertor plates and armor tiles are installed on the vehicle and the divertor handling manipulator carries a case filled with armor tiles, the replacement time of armor tiles will be reduced. In FER, moreover, maintenance of armor tiles, which is a scheduled maintenance, is planned to be carried out by the autonomous control using position sensors etc. In order to accumulate the data base for the development of the autonomous control of the manipulator in armor tile maintenance, the present paper describes basic mechanical characteristics (stress, deflection and natural frequency) of the armor tile handling manipulator calculated by static stress and dynamic eigenvalue analyses. (orig.)

  16. Study on the Adaptability of Etheriifcation Feedstock to Reactor Type

    Institute of Scientific and Technical Information of China (English)

    Mao Junyi; Yuan Qing; Wang Lei; Huang Tao

    2016-01-01

    A reactive C5 oleifns and methanol etheriifcation kinetic model based on E-R mechanism was established and three different types of reactors including the adiabatic ifxed-bed liquid reactor, the external loop reactor and the mixed-phase reactor were constructed by Aspen Plus. The adaptability of reactive C5 oleifns to these reactors was studied and simulated using various gasoline fractions with different oleifns content. After the theoretical model was validated by the experimental data of the etheriifcation of three C5 light cut fractions from different gasoline sources in different reactors, the simulated isoamylene conversion with reactive C5 olefin contents increasing from 10% to 60% was studied in the three different types of reactors for etheriifcation with methanol, respectively. Test results show that there is an obvious adaptability of the feedstock composition to the reactor type to achieve a high conversion.

  17. The ARIES tokamak reactor study

    Energy Technology Data Exchange (ETDEWEB)

    1989-10-01

    The ARIES study is a community effort to develop several visions of tokamaks as fusion power reactors. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Three ARIES visions are planned, each having a different degree of extrapolation from the present data base in physics and technology. The ARIES-I design assumes a minimum extrapolation from current tokamak physics (e.g., 1st stability) and incorporates technological advances that can be available in the next 20 to 30 years. ARIES-II is a DT-burning tokamak which would operate at a higher beta in the 2nd MHD stability regime. It employs both potential advances in the physics and expected advances in technology and engineering. ARIES-II will examine the potential of the tokamak and the D{sup 3}He fuel cycle. This report is a collection of 14 papers on the results of the ARIES study which were presented at the IEEE 13th Symposium on Fusion Engineering (October 2-6, 1989, Knoxville, TN). This collection describes the ARIES research effort, with emphasis on the ARIES-I design, summarizing the major results, the key technical issues, and the central conclusions.

  18. The ARIES tokamak reactor study

    International Nuclear Information System (INIS)

    The ARIES study is a community effort to develop several visions of tokamaks as fusion power reactors. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Three ARIES visions are planned, each having a different degree of extrapolation from the present data base in physics and technology. The ARIES-I design assumes a minimum extrapolation from current tokamak physics (e.g., 1st stability) and incorporates technological advances that can be available in the next 20 to 30 years. ARIES-II is a DT-burning tokamak which would operate at a higher beta in the 2nd MHD stability regime. It employs both potential advances in the physics and expected advances in technology and engineering. ARIES-II will examine the potential of the tokamak and the D3He fuel cycle. This report is a collection of 14 papers on the results of the ARIES study which were presented at the IEEE 13th Symposium on Fusion Engineering (October 2-6, 1989, Knoxville, TN). This collection describes the ARIES research effort, with emphasis on the ARIES-I design, summarizing the major results, the key technical issues, and the central conclusions

  19. Optimization design study of an innovative divertor concept for future experimental tokamak-type fusion reactors

    International Nuclear Information System (INIS)

    The design optimization study of an innovative divertor concept for future experimental tokamak-type fusion devices is both an answer to the actual problems encountered in the multilayer divertor proposals and an illustration of a rational modelling philosophy and optimization strategy for the development of a new divertor structure. Instead of using mechanical attachment or metallurgical bonding of the protective material to the heat sink as in most actual divertor concepts, the so-called brush divertor in this study uses an array of unidirectional fibers penetrating in both the protective armor and the underling composite heat sink. Although the approach is fully concentrated on the divertor performance, including both a description of its function from the theoretical point of view and an overview of the problems related to the materials choice and evaluation, both the approach followed in the numerical modelling and the judgment of the results are thought to be valid also for other applications. Therefore the spin-off of the study must be situated in both the technological progress towards a feasible divertor solution, which introduces no additional physical uncertainties, and in the general area of the thermo-mechanical finite-element modelling on both macro-and microscale. The brush divertor itself embodies the use, and thus the modelling, of advanced materials such as tailor-made metal matrix composites and dispersion strengthened metals, and is shown to offer large potential advantages, demanding however and experimental validation under working conditions. It is clearly indicated where the need originates for an integrated experimental program which must allow to verify the basic modelling assumptions in order to arrive at the use of numerical computation as a powerful and realistic tool of structural testing and life-time prediction

  20. Reliability studies in research reactors

    International Nuclear Information System (INIS)

    Fault trees and event trees are widely used in industry to model and to evaluate the reliability of safety systems. Detailed analyzes in nuclear installations require the combination of these two techniques. This study uses the methods of FT (Fault Tree) and ET (Event Tree) to accomplish the PSA (Probabilistic Safety Assessment) in research reactors. According to IAEA (lnternational Atomic Energy Agency), the PSA is divided into Level 1, Level 2 and Level 3. At the Level 1, conceptually, the security systems perform to prevent the occurrence of accidents, At the Level 2, once accidents happened, this Level seeks to minimize consequences, known as stage management of accident, and at Level 3 accident impacts are determined. This study focuses on analyzing the Level 1, and searching through the acquisition of knowledge, the consolidation of methodologies for future reliability studies. The Greek Research Reactor, GRR-1, is a case example. The LOCA (Loss of Coolant Accident) was chosen as the initiating event and from it, using ET, possible accidental sequences were developed, which could lead damage to the core. Moreover, for each of affected systems, probabilities of each event top of FT were developed and evaluated in possible accidental sequences. Also, the estimates of importance measures for basic events are presented in this work. The studies of this research were conducted using a commercial computational tool SAPHIRE. Additionally, achieved results thus were considered satisfactory for the performance or the failure of analyzed systems. (author)

  1. Study on an innovative fast reactor utilizing hydride neutron absorber development of coating technique on cladding inner surface

    International Nuclear Information System (INIS)

    The study to extend the control rod life of the Fast Reactor (FR) and to compress its excess reactivity are being performed by adopting the hafnium hydride (HfHx) for control rod material and by using the gadolinium hydride (GdHx) burnable poison (BP) for the reactivity recession, respectively. In the program named 'Study on an innovative Fast Reactor utilizing Hydride Neutron Absorber', the coating technique on inner surface of cladding has been developed to prevent hydrogen transfer through cladding at occasions of the temperature rise events. The Cr2O3 coating (chromizing) and the Al2O3 coating (calorizing) were selected for the coating techniques from the viewpoint of stability under in-core conditions. Following tests were performed for austenitic steel SUS316 which is widely used in FRs and for ferritic steel SUS430. The SUS430 was selected to simulate the ODS (Oxide Dispersion Strengthened ferritic steel) which is the attractive candidate material for the high burn-up FR. Examination of coating processing conditions by using short length claddings (100-200 mm). Approval of coating conditions to mock-up length cladding (1000 mm). Measurement of hydrogen transfer coefficient. Then appropriate conditions for coating were clarified and the formation of homogeneous films of both chromizing and calorizing was achieved on the inner surfaces of long length claddings (1000 mm). The hydrogen transfer experiments showed that the hydrogen transfer coefficient of coated SUS316 and SUS430 can be reduced to below 1/10 of SUS316 raw material. (author)

  2. STUDY OF MERCURY OXIDATION BY SCR CATALYST IN AN ENTRAINED-FLOW REACTOR UNDER SIMULATED PRB CONDITIONS

    Science.gov (United States)

    A bench-scale entrained-flow reactor system was constructed for studying elemental mercury oxidation under selective catalytic reduction (SCR) reaction conditions. Simulated flue gas was doped with fly ash collected from a subbituminous Powder River Basin (PRB) coal-fired boiler ...

  3. Reactor safety - an international task

    International Nuclear Information System (INIS)

    The dimensions and the significance of the task of ensuring reactor safety can be defined on the basis of experiences gained from Harrisburg and Chernobyl. The countries that use nuclear energy are tied together to a community by virtue of the risk they share. Therefore the GRS is working in close cooperation with the EC, OECD, IAEO and COMECON. This results in safety examinations of the Greifswald reactor, safety analyses of nuclear reactors in Germany, France and the USA and also considerations on the safety demands to be placed on new reactor concepts. (DG)

  4. Design study on the Advanced Recycling Reactor

    International Nuclear Information System (INIS)

    Full text: The design study on the Advanced Recycling Reactor (ARR) has been conducted. This paper presents the pre-conceptual design of the ARR that is a loop-typed sodium cooled reactor with MOX fuel. International Nuclear Recycling Alliance (INRA) takes advantage of international experience and uses the design based on Japan Sodium-cooled Fast Reactor (JSFR) as reference for FOA studies of US DOE, because Japan has conducted R and Ds for the JSFR incorporating thirteen technology enhancements expected to improve safety, enhance economics, and increase reactor reliability. The targets of the ARR are to generate electricity while consuming fuel containing transuranics and to attain cost competitiveness with the similar sized LWRs. INRA proposes 3 evolutions of the ARR; ARR1, a 500 MWe demonstration plant, online in 2025; ARR2, a 1,000 MWe commercial plant, online in 2035; ARR3, a 1,500 MWe full-scale commercial plant, online in 2050. INRA believes the scale-up factor of two is acceptable increase from manufacturing and licensing points of view. Major features of the ARR1 are the following: The reactor core is 70cm high and the volume fraction of fuel is approximately 32%. The conversion ratio of fissile is set up less than 0.6 and the amount of burned TRU is 45-51 kg/TWeh.Decay heat can be removed by natural circulation to improve safety. The primary cooling system consists of two-loop arrangement and the integrated IHX/Pump to improve economics. The steam generator with the straight double-walled tube is used to improve reliability. The ARR1 is co-located with a recycling facility. The overall plant facility arrangement is planned assuming to be constructed and installed in an inland area. The plant consists of a reactor building (including reactor auxiliary facilities and electrical/control systems), a turbine building, and a recycling building. The volume of the reactor building will be approximately 180,000 m3. The capital cost for the ARR1 and the ARR2 are

  5. Reactor safety study. An assessment of accident risks in U.S. commercial nuclear power plants. Appendix VI. Calculation of reactor accident consequences

    International Nuclear Information System (INIS)

    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

  6. 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.

  7. Reactor Neutrino Physics -- An Update

    OpenAIRE

    Boehm, Felix

    1999-01-01

    We review the status and the results of reactor neutrino experiments. Long baseline oscillation experiments at Palo Verde and Chooz have provided limits for the oscillation parameters while the recently proposed Kamland experiment at a baseline of more than 100km is now in the planning stage. We also describe the status of neutrino magnetic moment experiments at reactors.

  8. Reversed field pinch reactor study 3

    International Nuclear Information System (INIS)

    This report, the third of a series on the Reversed Field Pinch Reactor, describes a preliminary concept of the engineering design and layout of this pulsed toroidal reactor, which uses the stable plasma behaviour first observed in ZETA. The basic parameters of the 600 MW(e) reactor are taken from a companion study by Hancox and Spears. The plasma volume is 1.75m minor radius and 16m major radius surrounded by a 1.8m blanket-shield region - with the blanket divided into 14 removable segments for servicing. The magnetic confinement system consists of 28 toroidal field coils situated just outside the blanket and inside the poloidal and vertical field coils and all coils have normal copper conductors. The requirement to incorporate a conducting shell at the front of the blanket to provide a short-time plasma stability has a marked effect on the design. It sets the size of the blanket segment and the scale of the servicing operations, limits the breeding gain and complicates the blanket cooling and its integration with the heat engine. An extensive study will be required to confirm the overall reactor potential of the concept. (author)

  9. An approach for reactor vessel life assessment following an anneal

    International Nuclear Information System (INIS)

    The integrity of the reactor pressure vessel is critical to continued operation of nuclear power plants. Long term exposure to high energy neutrons can cause irradiation embrittlement of the reactor pressure vessel steel. Irradiation embrittlement may be a life limiting factor for some nuclear power plants. Annealing is the only option for reversing the effects of irradiation embrittlement. The feasibility and generic benefits of the annealing process have been demonstrated through numerous industry studies. The consideration of annealing as part of a reactor vessel aging management program requires the ability to predict the annealing and re-irradiation response of pressure vessel steel. Data for these predictions can be obtained through proper planning and implementation during the current years of reactor vessel operation. A comprehensive materials test plan enables a utility to gain significant information relative to reactor vessel annealing in a timely manner. This paper discusses a materials test plan for an example nuclear plant. The planning process, the type of data generated, and an approach for transforming the data into meaningful predictions for the vessel re-irradiation response are all illustrated. The intent is to provide guidelines for gathering and interpreting the data that is used to predict the life of a typical reactor vessel following an anneal

  10. D-D tokamak reactor studies

    International Nuclear Information System (INIS)

    A tokamak D-D reactor design, utilizing the advantages of a deuterium-fueled reactor but with parameters not unnecessarily extended from existing D-T designs, is presented. Studies leading to the choice of a design and initial studies of the design are described. The studies are in the areas of plasma engineering, first-wall/blanket/shield design, magnet design, and tritium/fuel/vacuum requirements. Conclusions concerning D-D tokamak reactors are stated

  11. Power reactor noise studies and applications

    International Nuclear Information System (INIS)

    The present thesis deals with the neutron noise arising in power reactor systems. Generally, it can be divided into two major parts: first, neutron noise diagnostics, or more specifically, novel methods and algorithms to monitor nuclear industrial reactors; and second, contributions to neutron noise theory as applied to power reactor systems. Neutron noise diagnostics is presented by two topics. The first one is a theoretical study on the possibility to use a newly proposed current-flux (C/F) detector in Pressurised Water Reactors (PWR) for the localisation of anomalies. The second topic concerns various methods to detect guide tube impacting in Boiling Water Reactors (BWR). The significance of these problems comes from the operational experience. The thesis describes a novel method to localise vibrating control rods in a PWR by using only one C/F detector. Another novel method, based on wavelet analysis, is put forward to detect impacting guide tubes in a BWR. Neutron noise theory is developed for both Accelerator Driven Systems (ADS) and traditional reactors. By design the accelerator-driven systems would operate in a subcritical mode with a strong external source. This calls for a revision of many concepts and methods that have been developed for traditional reactors and also it poses a number of new problems. As for the latter, the thesis investigates the space-dependent neutron noise caused by a fluctuating source. It is shown that the frequency-dependent spatial behaviour exhibits some new properties that are different from those known in traditional critical systems. On the other hand, various reactor physics approximations (point kinetic, adiabatic etc.) have not been defined yet for the subcritical systems. In this respect the thesis presents a systematic formulation of the above mentioned approximations as well as investigations of their properties. Another important problem in neutron noise theory is the treatment of moving boundaries. In this case one

  12. Power reactor noise studies and applications

    Energy Technology Data Exchange (ETDEWEB)

    Arzhanov, V

    2002-03-01

    The present thesis deals with the neutron noise arising in power reactor systems. Generally, it can be divided into two major parts: first, neutron noise diagnostics, or more specifically, novel methods and algorithms to monitor nuclear industrial reactors; and second, contributions to neutron noise theory as applied to power reactor systems. Neutron noise diagnostics is presented by two topics. The first one is a theoretical study on the possibility to use a newly proposed current-flux (C/F) detector in Pressurised Water Reactors (PWR) for the localisation of anomalies. The second topic concerns various methods to detect guide tube impacting in Boiling Water Reactors (BWR). The significance of these problems comes from the operational experience. The thesis describes a novel method to localise vibrating control rods in a PWR by using only one C/F detector. Another novel method, based on wavelet analysis, is put forward to detect impacting guide tubes in a BWR. Neutron noise theory is developed for both Accelerator Driven Systems (ADS) and traditional reactors. By design the accelerator-driven systems would operate in a subcritical mode with a strong external source. This calls for a revision of many concepts and methods that have been developed for traditional reactors and also it poses a number of new problems. As for the latter, the thesis investigates the space-dependent neutron noise caused by a fluctuating source. It is shown that the frequency-dependent spatial behaviour exhibits some new properties that are different from those known in traditional critical systems. On the other hand, various reactor physics approximations (point kinetic, adiabatic etc.) have not been defined yet for the subcritical systems. In this respect the thesis presents a systematic formulation of the above mentioned approximations as well as investigations of their properties. Another important problem in neutron noise theory is the treatment of moving boundaries. In this case one

  13. 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.

  14. Laboratory Experiments and Modeling for Interpreting Field Studies of Secondary Organic Aerosol Formation Using an Oxidation Flow Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez, Jose-Luis [Univ. of Colorado, Boulder, CO (United States)

    2016-02-01

    This grant was originally funded for deployment of a suite of aerosol instrumentation by our group in collaboration with other research groups and DOE/ARM to the Ganges Valley in India (GVAX) to study aerosols sources and processing. Much of the first year of this grant was focused on preparations for GVAX. That campaign was cancelled due to political reasons and with the consultation with our program manager, the research of this grant was refocused to study the applications of oxidation flow reactors (OFRs) for investigating secondary organic aerosol (SOA) formation and organic aerosol (OA) processing in the field and laboratory through a series of laboratory and modeling studies. We developed a gas-phase photochemical model of an OFR which was used to 1) explore the sensitivities of key output variables (e.g., OH exposure, O3, HO2/OH) to controlling factors (e.g., water vapor, external reactivity, UV irradiation), 2) develop simplified OH exposure estimation equations, 3) investigate under what conditions non-OH chemistry may be important, and 4) help guide design of future experiments to avoid conditions with undesired chemistry for a wide range of conditions applicable to the ambient, laboratory, and source studies. Uncertainties in the model were quantified and modeled OH exposure was compared to tracer decay measurements of OH exposure in the lab and field. Laboratory studies using OFRs were conducted to explore aerosol yields and composition from anthropogenic and biogenic VOC as well as crude oil evaporates. Various aspects of the modeling and laboratory results and tools were applied to interpretation of ambient and source measurements using OFR. Additionally, novel measurement methods were used to study gas/particle partitioning. The research conducted was highly successful and details of the key results are summarized in this report through narrative text, figures, and a complete list of publications acknowledging this grant.

  15. Extrap conceptual fusion reactor design study

    International Nuclear Information System (INIS)

    A study has recently been initiated to asses the fusion reactor potential of the Extrap concept. A reactor model is defined that fulfills certain economic and environmental criteria. This model is applied to Extrap and a reference reactor is outlined. The design is optimized by varying parameters subject to both physics and engineering constraints. Several design options are examined and key engineering issues are identified and addressed. Some preliminary results and conclusions of this work are summarized. (authors)

  16. A methodological study on organizing an intelligent CAD/CAE system for conceptual design of advanced nuclear reactor system

    International Nuclear Information System (INIS)

    In order to shorten the time span of design work and enhance both consistency and rationality of design products, the authors are now investigating an intelligent CAD/CAE system to support cooperative works by many specialists by adopting object-oriented approach. In this paper, the cognitive aspect of design activities of specialists in the conceptual design phase of nuclear reactors is discussed. The activities of the specialists in their design analysis process are highly knowledge-based and goal-oriented. The characteristics of the activities are 1) hierarchization of design goal into sub-goals, 2) prioritization of design sub-goals and step-by-step practise of design analysis, and 3) abstraction of real-world space structure into more simplified space structure to cope with theoretical treatment. Based on these consideration, a conceptual design model of specialists' activities composed of attribute modeling and design expertise knowledge base is proposed. The 'principle of functional independence' proposed by Sue is applied to bridge between the attribute modeling and design expertise knowledge base. The intelligent CAD/CAE system is now under development by focusing on the conceptual design of a space power reactor core utilizing thermo-ionic fuel elements as direct thermo-to-electric conversion. A program to calculate thermo-hydraulics of reactor core and thermo-ionic power generation has been developed. An interface has been also developed in order to communicate with the specialists at JAERI by E-mail concerning the interactive calculation between our calculation and the neutronics calculation of reactor core. (orig.)

  17. CARNSORE: Hypothetical reactor accident study

    International Nuclear Information System (INIS)

    Two types of design-basis accident and a series of hypothetical core-melt accidents to a 600 MWe reactor are described and their consequences assessed. The PLUCON 2 model was used to calculate the consequences which are presented in terms of individual and collective doses, as well as early and late health consequences. The site proposed for the nucelar power station is Carnsore Point, County Wexford, south-east Ireland. The release fractions for the accidents described are those given in WASH-1400. The analyses are based on the resident population as given in the 1979 census and on 20 years of data from the meteorological stations at Rosslare Harbour, 8.5 km north of the site. The consequences of one of the hypothetical core-melt accidents are described in detail in a meteorological parametric study. Likewise the consequences of the worst conceivable combination of situations are described. Finally, the release fraction in one accident is varied and the consequences of a proposed, more probable ''Class 9 accident'' are presented. (author)

  18. Wood ash amendment to biogas reactors as an alternative to landfilling? A preliminary study on changes in process chemistry and biology.

    Science.gov (United States)

    Podmirseg, Sabine M; Seewald, Martin S A; Knapp, Brigitte A; Bouzid, Ourdia; Biderre-Petit, Corinne; Peyret, Pierre; Insam, Heribert

    2013-08-01

    Wood ash addition to biogas plants represents an alternative to commonly used landfilling by improving the reactor performance, raising the pH and alleviating potential limits of trace elements. This study is the first on the effects of wood ash on reactor conditions and microbial communities in cattle slurry-based biogas reactors. General process parameters [temperature, pH, electrical conductivity, ammonia, volatile fatty acids, carbon/nitrogen (C/N), total solids (TS), volatile solids, and gas quantity and quality] were monitored along with molecular analyses of methanogens by polymerase chain reaction- denaturing gradient gel electrophoresis and modern microarrays (archaea and bacteria). A prompt pH rise was observed, as was an increase in C/N ratio and volatile fatty acids. Biogas production was inhibited, but recovered to even higher production rates and methane concentration after single amendment. High sulphur levels in the wood ash generated hydrogen sulphide and potentially hampered methanogenesis. Methanosarcina was the most dominant methanogen in all reactors; however, diversity was higher in ash-amended reactors. Bacterial groups like Firmicutes, Proteobacteria and Acidobacteria were favoured, which could improve the hydrolytic efficiency of the reactors. We recommend constant monitoring of the chemical composition of the used wood ash and suggest that ash amendment is adequate if added to the substrate at a rate low enough to allow adaptation of the microbiota (e.g. 0.25 g g(-1) TS). It could further help to enrich digestate with important nutrients, for example phosphorus, calcium and magnesium, but further experiments are required for the evaluation of wood ash concentrations that are tolerable for anaerobic digestion.

  19. Wood ash amendment to biogas reactors as an alternative to landfilling? A preliminary study on changes in process chemistry and biology.

    Science.gov (United States)

    Podmirseg, Sabine M; Seewald, Martin S A; Knapp, Brigitte A; Bouzid, Ourdia; Biderre-Petit, Corinne; Peyret, Pierre; Insam, Heribert

    2013-08-01

    Wood ash addition to biogas plants represents an alternative to commonly used landfilling by improving the reactor performance, raising the pH and alleviating potential limits of trace elements. This study is the first on the effects of wood ash on reactor conditions and microbial communities in cattle slurry-based biogas reactors. General process parameters [temperature, pH, electrical conductivity, ammonia, volatile fatty acids, carbon/nitrogen (C/N), total solids (TS), volatile solids, and gas quantity and quality] were monitored along with molecular analyses of methanogens by polymerase chain reaction- denaturing gradient gel electrophoresis and modern microarrays (archaea and bacteria). A prompt pH rise was observed, as was an increase in C/N ratio and volatile fatty acids. Biogas production was inhibited, but recovered to even higher production rates and methane concentration after single amendment. High sulphur levels in the wood ash generated hydrogen sulphide and potentially hampered methanogenesis. Methanosarcina was the most dominant methanogen in all reactors; however, diversity was higher in ash-amended reactors. Bacterial groups like Firmicutes, Proteobacteria and Acidobacteria were favoured, which could improve the hydrolytic efficiency of the reactors. We recommend constant monitoring of the chemical composition of the used wood ash and suggest that ash amendment is adequate if added to the substrate at a rate low enough to allow adaptation of the microbiota (e.g. 0.25 g g(-1) TS). It could further help to enrich digestate with important nutrients, for example phosphorus, calcium and magnesium, but further experiments are required for the evaluation of wood ash concentrations that are tolerable for anaerobic digestion. PMID:23831776

  20. Reactor application of an improved bundle divertor

    International Nuclear Information System (INIS)

    A Bundle Divertor was chosen as the impurity control and plasma exhaust system for the beam driven Demonstration Tokamak Hybrid Reactor - DTHR. In the context of a preconceptual design study of the reactor and associated facility a bundle divertor concept was developed and integrated into the reactor system. The overall system was found feasible and scalable for reactors with intermediate torodial field strengths on axis. The important design characteristics are: the overall average current density of the divertor coils is 0.73 kA for each tesla of toroidal field on axis; the divertor windings are made from super-conducting cables supported by steel structures and are designed to be maintainable; the particle collection assembly and auxiliary cryosorption vacuum pump are dual systems designed such that they can be reactivated alterntively to allow for continuous reactor operation; and the power requirement for energizing and operating the divertor is about 5 MW

  1. Conversion of an AVR reactor into a process heat reactor

    International Nuclear Information System (INIS)

    The proposal to convert the Juelich AVR into a process heat reactor aims at achieving a low-cost demonstration and system test of uncoupling process heat from a high-temperature reactor at a high useful temperature. The conversion is planned in such a way that the AVR can serve various conversion processes in succession as a nuclear source. For the first operational phase a methane-water vapor reformation by means of a tube fission furnace and a process steam generator as an element of direct heat uncoupling in a technical process is envisaged. For further operational phases, other coal processing methods are planned. The concept, safety features, licensability, cost, utilization and deadlines are outlined. (orig./DG)

  2. An experimental parametric study of the high pressure melt ejection from two different scale reactor cavity models

    International Nuclear Information System (INIS)

    A parametric study of the high pressure melt ejection(HPME) from two small-scale(1/25th and 1/41st) transparent reactor cavity models of the YoungGwang(unit 1 and 2) has been performed. Wood's metal and water have been used as melt simulants while high pressure nitrogen and carbon dioxide are used as driver gases to simulate the blowdown steam and gas from the breach of the reactor pressure vessel. Experimental data for the fraction of melt simulant retained in the cavity model(Yf) during a postulated scenario of the HPME from PWR pressure vessel have been obtained as a function of various test parameters. These data have been used to develop a correlation for Yf that fits all the data(a total of 313 data points) within the standard deviation of 0.054 by means of dimensional analysis and nonlinear least-squares optimization technique. The basic effects of important parameters used to describe the HPME accident sequence on the Yf are also determined based on the correlation obtained here and experimental evidence. (Author)

  3. An experimental study on the two-phase natural circulation flow through the gap between reactor vessel and insulation under ERVC

    International Nuclear Information System (INIS)

    As part of a study on a two-phase natural circulation flow between the outer reactor vessel and the insulation material in the reactor cavity under an external reactor vessel cooling of APR1400, T-HERMES-SMALL and HERMES-HALF experiments have been performed. For the T-HERMES-SMALL experiments, an 1/21.6 scaled experimental facility was prepared utilizing the results of a scaling analysis to simulate the APR1400 reactor and insulation system. The liquid mass flow rates driven by natural circulation loop were measured by varying the wall heat flux, upper outlet area and configuration, and water head condition. The experimental data were also compared with numerical ones given by simple loop analysis. And non-heating small-scaled experiments have also been performed to certify the hydraulic similarity of the heating experiments by injecting air equivalent to the steam generated in the heating experimental condition. The HERMES-HALF experiment is a half-scaled / non-heating experimental study on the two-phase natural circulation through the annular gap between the reactor vessel and the insulation. The behaviors of the two-phase natural circulation flow in the insulation gap were observed, and the liquid mass flow rates driven by natural circulation loop were measured by varying the air injection rate, the coolant inlet area and configuration, and the outlet area and also the water head condition of coolant reservoir. From the experimental flow observation, the recirculation flows in the near region of the shear key were identified. At a higher air injection rate condition, higher recirculation flows and choking phenomenon in the near region of the shear key were observed. As the water inlet areas increased, the natural circulation mass flow rates asymptotically increased, that is, they converged at a specific value. And the experimental correlations for the natural circulation mass flow rates along with a variation of the inlet / outlet area and wall heat flux were

  4. An experimental study on the two-phase natural circulation flow through the gap between reactor vessel and insulation under ERVC

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Kwang-Soon; Park, Rae-Joon; Cho, Young-Ro; Kim, Sang-Baik; Kim, Hwan-Yeol; Kim, Hee-dong

    2005-04-01

    As part of a study on a two-phase natural circulation flow between the outer reactor vessel and the insulation material in the reactor cavity under an external reactor vessel cooling of APR1400, T-HERMES-SMALL and HERMES-HALF experiments have been performed. For the T-HERMES-SMALL experiments, an 1/21.6 scaled experimental facility was prepared utilizing the results of a scaling analysis to simulate the APR1400 reactor and insulation system. The liquid mass flow rates driven by natural circulation loop were measured by varying the wall heat flux, upper outlet area and configuration, and water head condition. The experimental data were also compared with numerical ones given by simple loop analysis. And non-heating small-scaled experiments have also been performed to certify the hydraulic similarity of the heating experiments by injecting air equivalent to the steam generated in the heating experimental condition. The HERMES-HALF experiment is a half-scaled / non-heating experimental study on the two-phase natural circulation through the annular gap between the reactor vessel and the insulation. The behaviors of the two-phase natural circulation flow in the insulation gap were observed, and the liquid mass flow rates driven by natural circulation loop were measured by varying the air injection rate, the coolant inlet area and configuration, and the outlet area and also the water head condition of coolant reservoir. From the experimental flow observation, the recirculation flows in the near region of the shear key were identified. At a higher air injection rate condition, higher recirculation flows and choking phenomenon in the near region of the shear key were observed. As the water inlet areas increased, the natural circulation mass flow rates asymptotically increased, that is, they converged at a specific value. And the experimental correlations for the natural circulation mass flow rates along with a variation of the inlet / outlet area and wall heat flux were

  5. The ARIES-I tokamak reactor study

    International Nuclear Information System (INIS)

    This report discusses the following topics on the Aries-I Tokamak: Design description; systems studies and economics; reactor plasma physics; magnet engineering; fusion-power-ore engineering; and environmental and safety features

  6. Study on secondary shutdown systems in Tehran research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jalali, H.R.; Fadaei, A.H., E-mail: Fadaei_amir@aut.ac.ir; Gharib, M.

    2015-09-15

    Highlights: • A study was undertaken to summarize the techniques for secondary shutdown systems (SSS). • Neutronic calculation performed for proposed systems as SSS. • Dumping the heavy water stored in the reflector vessel is capable to shut down reactor. • Neutronic and transient calculation was done for validating the selected SSS. • All calculation shown that this system has advantages in safety and neutron economy. - Abstract: One important safety aspect of any research reactor is the ability to shut down the reactor. Usually, research reactors, currently in operation, have a single shutdown system based on the simultaneous insertion of the all control rods into the reactor core through gravity. Nevertheless, the International Atomic Energy Agency currently recommends use of two shutdown systems which are fully independent from each other to guarantee secure shutdown when one of them fails. This work presents an investigative study into secondary shutdown systems, which will be an important safety component in the research reactor and will provide another alternative way to shut down the reactor emergently. As part of this project, a study was undertaken to summarize the techniques that are currently used at world-wide research reactors for recognizing available techniques to consider in research reactors. Removal of the reflector, removal of the fuels, change in critical shape of reactor core and insertion of neutron absorber between the core and reflector are selected as possible techniques in mentioned function. In the next step, a comparison is performed for these methods from neutronic aspects. Then, chosen method is studied from the transient behavior point of view. Tehran research reactor which is a 5 MW open-pool reactor selected as a case study and all calculations are carried out for it. It has 5 control rods which serve the purpose of both reactivity control and shutdown of reactor under abnormal condition. Results indicated that heavy

  7. Development of an educational nuclear research reactor simulator

    International Nuclear Information System (INIS)

    This paper introduces the development of a research reactor educational simulator based on LabVIEW that allows the training of operators and studying different accident scenarios and the effects of operational parameters on the reactor behavior. Using this simulator, the trainee can test the interaction between the input parameters and the reactor activities. The LabVIEW acts as an engine implements the reactor mathematical models. In addition, it is used as a tool for implementing the animated graphical user interface. This simulator provides the training requirements for both of the reactor staff and the nuclear engineering students. Therefore, it uses dynamic animation to enhance learning and interest for a trainee on real system problems and provides better visual effects, improved communications, and higher interest levels. The benefits of conducting such projects are to develop the expertise in this field and save costs of both operators training and simulation courses.

  8. Development of an educational nuclear research reactor simulator

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-12-15

    This paper introduces the development of a research reactor educational simulator based on LabVIEW that allows the training of operators and studying different accident scenarios and the effects of operational parameters on the reactor behavior. Using this simulator, the trainee can test the interaction between the input parameters and the reactor activities. The LabVIEW acts as an engine implements the reactor mathematical models. In addition, it is used as a tool for implementing the animated graphical user interface. This simulator provides the training requirements for both of the reactor staff and the nuclear engineering students. Therefore, it uses dynamic animation to enhance learning and interest for a trainee on real system problems and provides better visual effects, improved communications, and higher interest levels. The benefits of conducting such projects are to develop the expertise in this field and save costs of both operators training and simulation courses.

  9. Development of an automatic reactor inspection system

    International Nuclear Information System (INIS)

    Using recent technologies on a mobile robot computer science, we developed an automatic inspection system for weld lines of the reactor vessel. The ultrasonic inspection of the reactor pressure vessel is currently performed by commercialized robot manipulators. Since, however, the conventional fixed type robot manipulator is very huge, heavy and expensive, it needs long inspection time and is hard to handle and maintain. In order to resolve these problems, we developed a new automatic inspection system using a small mobile robot crawling on the vertical wall of the reactor vessel. According to our conceptual design, we developed the reactor inspection system including an underwater inspection robot, a laser position control subsystem, an ultrasonic data acquisition/analysis subsystem and a main control subsystem. We successfully carried out underwater experiments on the reactor vessel mockup, and real reactor ready for Ulchine nuclear power plant unit 6 at Dusan Heavy Industry in Korea. After this project, we have a plan to commercialize our inspection system. Using this system, we can expect much reduction of the inspection time, performance enhancement, automatic management of inspection history, etc. In the economic point of view, we can also expect import substitution more than 4 million dollars. The established essential technologies for intelligent control and automation are expected to be synthetically applied to the automation of similar systems in nuclear power plants

  10. An option for the Brazilian nuclear project: necessity of fast breeder reactors and core design for an experimental fast reactor

    International Nuclear Information System (INIS)

    Aiming to assure the continued utilization of fission energy, the development of fast breeder reactors (FBRs) is a necessity. Binary fueled LMFBRs are proposed, as the best type for the Brazilian nuclear system in the future. The inherent safety characteristics are superior to current fast breeder reactors and an efficient utilization of thorium can be realized. The construction and operation of an experimental fast reactor is the first step and a basic tool for the development of FBRs technologies. A serie of core design for an 90 MW FBR is studied and the possible options and sizes of the main parameters are identified. (E.G.)

  11. Fusion reactor safety studies, FY 1977

    International Nuclear Information System (INIS)

    This report reviews the technical progress in the fusion reactor safety studies performed during FY 1977 in the Fusion Power Program at the Argonne National Laboratory. The subjects reported on include safety considerations of the vacuum vessel and first-wall design for the ANL/EPR, the thermal responses of a tokamak reactor first wall, the vacuum wall electrical resistive requirements in relationship to magnet safety, and a major effort is reported on considerations and experiments on air detritiation

  12. An overview of third generation reactors - Presentation

    International Nuclear Information System (INIS)

    Now, in 2013 about 69 nuclear reactors (67 GWe) are being built in the world, mostly in China and Russia. Although a few second generation reactors are being built (particularly in China), third generation reactors seem to be systematically chosen for any new construction. The French commercial offer is based on 3 models: the EPR, the ATMEA-1 and the KERENA, the first 2 being of PWR-type while the latter is a BWR. A lot of third generation reactor designs are available on the international market. Concerning the PWR technology we have: the EPR (AREVA), the AP1000 (Westinghouse), the AES 2006 and the VVER TOI (Rosatom), the APR1400 (KEPCO), the APWR (MHI), the ATMEA-1 (AREVA + MHI), the ACP 1000 (CNNC - China), ACPR1000 (CGN - China) and CAP 1400 (SNPTC - China). Concerning the BWR technology, the commercial offer is far less important we have: the ABWR (GE-Hitachi + Toshiba) and the ESBWR (GE-Hitachi). Some third generation reactors are operating now: the first ABWR and AES reactors but most of them (EPR, AP1000, APR1400 and AES 2006) are at an advanced stage of construction. (A.C.)

  13. Response of an embedded reactor containment to underground blast loading

    International Nuclear Information System (INIS)

    This paper presents a theoretical study of the dynamic response of an underground reactor containment structure to explosive loading. The work is motivated by the need for protecting a reactor against surface bombs and penetrating warheads. It is assumed that the outer containment is sufficiently strong to withstand the explosive loading without appreciable damage. The reactor systems, however, are subjected to high accelerations, due to the stress waves induced in the structure by the explosion. The level of these accelerations, under various loading conditions, is investigated in the present work. (orig.)

  14. Antineutrino reactor safeguards - a case study

    CERN Document Server

    Christensen, Eric; Jaffke, Patrick

    2013-01-01

    Antineutrinos have been proposed as a means of reactor safeguards for more than 30 years and there has been impressive experimental progress in neutrino detection. In this paper we conduct, for the first time, a case study of the application of antineutrino safeguards to a real-world scenario - the North Korean nuclear crisis in 1994. We derive detection limits to a partial or full core discharge in 1989 based on actual IAEA safeguards access and find that two independent methods would have yielded positive evidence for a second core with very high confidence. To generalize our results, we provide detailed estimates for the sensitivity to the plutonium content of various types of reactors, including most types of plutonium production reactors, based on detailed reactor simulations. A key finding of this study is that a wide class of reactors with a thermal power of less than 0.1-1 GWth can be safeguarded achieving IAEA goals for quantitative sensitivity and timeliness with detectors right outside the reactor ...

  15. Fusion reactor blanket/shield design study

    International Nuclear Information System (INIS)

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

  16. The SILENE reactor: an instrument suitable for studying the impact of intermediate and high radiation doses; Reacteur silene : outil adapte aux etudes appliquees a l'effet des moyennes et fortes doses

    Energy Technology Data Exchange (ETDEWEB)

    Verray, B.; Leo, Y.; Fouillaud, P. [Commissariat a l' Energie Atomique, CEA, Service de Recherches en Neutronique et Criticite, SNRC, Valduc, Is sur Tille (France)

    2002-07-01

    Designed in 1974 to study the phenomenology and consequences of a criticality accident, the SILENE experimental reactor, an intense source of mixed neutron and gamma radiation, is also suited to radiobiological studies. (author)

  17. Studies on Nitrogen Oxides Removal Using Plasma Assisted Catalytic Reactor

    Institute of Scientific and Technical Information of China (English)

    V. Ravi; Young Sun Mok; B. S. Rajanikanth; Ho-Chul Kang

    2003-01-01

    An electric discharge plasma reactor combined with a catalytic reactor was studied for removing nitrogen oxides. To understand the combined process thoroughly, discharge plasma and catalytic process were separately studied first, and then the two processes were combined for the study. The plasma reactor was able to oxidize NO to NO2 well although the oxidation rate decreased with temperature. The plasma reactor alone did not reduce the NOx (NO+NO2)level effectively, but the increase in the ratio of NO2 to NO as a result of plasma discharge led to the enhancement of NOx removal efficiency even at lower temperatures over the catalyst surface (V2O5-WOa/TiO2). At a gas temperature of 100℃, the NOx removal efficiency obtained using the combined plasma catalytic process was 88% for an energy input of 36 eV/molecule or 30 J/1.

  18. Ex vivo model of an immobilized-enzyme reactor.

    OpenAIRE

    Bernstein, H; Langer, R

    1988-01-01

    Immobilized-enzyme reactors are beginning to be studied for a variety of therapeutic applications. To facilitate the design of these devices for different clinical situations and a diverse patient population, mathematical models may be valuable. An immobilized-heparinase (EC 4.2.2.7) reactor was selected as a model system. The device removes heparin from blood that has been anticoagulated to prevent thrombus formation. Heparinase was immobilized to cross-linked agarose particles. A mathematic...

  19. Design study on sodium-cooled large-scale reactor

    International Nuclear Information System (INIS)

    In Phase 1 of the 'Feasibility Study on Commercialized Fast Reactor Cycle Systems (F/S)', an advanced loop type reactor has been selected as a promising concept of sodium-cooled large-scale reactor, which has a possibility to fulfill the design requirements of the F/S. In Phase 2 of the F/S, it is planed to precede a preliminary conceptual design of a sodium-cooled large-scale reactor based on the design of the advanced loop type reactor. Through the design study, it is intended to construct such a plant concept that can show its attraction and competitiveness as a commercialized reactor. This report summarizes the results of the design study on the sodium-cooled large-scale reactor performed in JFY2001, which is the first year of Phase 2. In the JFY2001 design study, a plant concept has been constructed based on the design of the advanced loop type reactor, and fundamental specifications of main systems and components have been set. Furthermore, critical subjects related to safety, structural integrity, thermal hydraulics, operability, maintainability and economy have been examined and evaluated. As a result of this study, the plant concept of the sodium-cooled large-scale reactor has been constructed, which has a prospect to satisfy the economic goal (construction cost: less than 200,000yens/kWe, etc.) and has a prospect to solve the critical subjects. From now on, reflecting the results of elemental experiments, the preliminary conceptual design of this plant will be preceded toward the selection for narrowing down candidate concepts at the end of Phase 2. (author)

  20. L-Reactor Habitat Mitigation Study

    International Nuclear Information System (INIS)

    The L-Reactor Fish and Wildlife Resource Mitigation Study was conducted to quantify the effects on habitat of the L-Reactor restart and to identify the appropriate mitigation for these impacts. The completed project evaluated in this study includes construction of a 1000 acre reactor cooling reservoir formed by damming Steel Creek. Habitat impacts identified include a loss of approximately 3,700 average annual habitat units. This report presents a mitigation plan, Plan A, to offset these habitat losses. Plan A will offset losses for all species studied, except whitetailed deer. The South Carolina Wildlife and Marine Resources Department strongly recommends creation of a game management area to provide realistic mitigation for loss of deer habitats. 10 refs., 5 figs., 3 tabs

  1. Study of an anaerobic - aerobic reactor of alternate phases for the removal of organic matter and nutritious, nitrogen and phosphorus

    International Nuclear Information System (INIS)

    In this article the results of the behavior from a reactor to laboratory scale with half fixed and anaerobic-aerobic alternate phases, for the organic and nutritious simultaneous removal of matter is presented. The laboratory assembly it makes with two reactors, of 2,2 l (litre) of capacity each one, connected in series, low continuous feeding, and the control of a system of solenoids valves that it allowed to automatically program of anaerobic-aerobic alternate phases in each reactor. Under this outline efficiency was obtained in removal of COD (Chemical Oxygen Demand) between 88 and 92%, of phosphorus of 87% and nitrogen of 77 to 60%, and lower volumetric loads between 2,5 and 0,38 g COD/l-d. The process generates a low sludge production, presenting economic advantages in front of aerobic traditional systems and to the removal technologies of nutritious

  2. Sensitivity and Uncertainty Study for Thermal Molten Salt Reactors

    Science.gov (United States)

    Bidaud, Adrien; Ivanona, Tatiana; Mastrangelo, Victor; Kodeli, Ivo

    2006-04-01

    The Thermal Molten Salt Reactor (TMSR) using the thorium cycle can achieve the GEN IV objectives of economy, safety, non-proliferation and durability. Its low production of higher actinides, coupled with its breeding capabilities - even with a thermal spectrum - are very valuable characteristics for an innovative reactor. Furthermore, the thorium cycle is more flexible than the uranium cycle since only a small fissile inventory (reactor. The potential of these reactors is currently being extensively studied at the CNRS and EdF /1,2/. A simplified chemical reprocessing is envisaged compared to that used for the former Molten Salt Breeder Reactor (MSBR). The MSBR concept was developed at Oak Ridge National Laboratory (ORNL) in the 1970's based on the Molten Salt Reactor Experiment (MSRE). The main goals of our current studies are to achieve a reactor concept that enables breeding, improved safety and having chemical reprocessing needs reduced and simplified as much as reasonably possible. The neutronic properties of the new TMSR concept are presented in this paper. As the temperature coefficient is close to zero, we will see that the moderation ratio cannot be chosen to simultaneously achieve a high breeding ratio, long graphite lifetime and low uranium inventory. It is clear that any safety margin taken due to uncertainty in the nuclear data will significantly reduce the capability of this concept, thus a sensitivity analysis is vital to propose measurements which would allow to reduce at present high uncertainties in the design parameters of this reactor. Two methodologies, one based on OECD/NEA deterministic codes and one on IPPE (Obninsk) stochastic code, are compared for keff sensitivity analysis. The uncertainty analysis of keff using covariance matrices available in evaluated files has been performed. Furthermore, a comparison of temperature coefficient sensitivity profiles is presented for the most important reactions. These results are used to review the

  3. Recycle of LWR [Light Water Reactor] actinides to an IFR [Integral Fast Reactor

    International Nuclear Information System (INIS)

    A large quantity of actinide elements is present in irradiated Light Water Reactor (LWR) fuel that is stored throughout the world. Because of the high fission-to-capture ratio for the transuranium (TRU) elements with the high-energy neutrons in the metal-fueled Integral Fast Reactor (IFR), that reactor can consume these elements effectively. The stored fuel represents a valuable resource for an expanding application of fast power reactors. In addition, removal of the TRU elements from the spent LWR fuel has the potential for increasing the capacity of a high-level waste facility by reducing the heat loads and increasing the margin of safety in meeting licensing requirements. Argonne National Laboratory (ANL) is developing a pyrochemical process, which is compatible with the IFR fuel cycle, for the recovery of TRU elements from LWR fuel. The proposed product is a metallic actinide ingot, which can be introduced into the electrorefining step of the IFR process. The major objective of the LWR fuel recovery process is high TRU element recovery, with decontamination a secondary issue, because fission product removal is accomplished in the IFR process. The extensive pyrochemical processing studies of the 1960s and 1970s provide a basis for the design of possible processes. Two processes were selected for laboratory-scale investigation. One is based on the Salt Transport Process studied at ANL for mixed-oxide fast reactor fuel, and the other is based on the blanket processing studies done for ANL's second Experimental Breeder Reactor (EBR-2). This paper discusses the two processes and is a status report on the experimental studies. 5 refs., 2 figs., 2 tabs

  4. CONTROL OF VOLATILE ORGANIC COMPOUNDS BY AN AC ENERGIZED FERROELECTRIC PELLET REACTOR AND A PULSED CORONA REACTOR

    Science.gov (United States)

    The paper gives results of a study to develop baseline engineering data to demonstrate the feasibility of application of plasma reactors to the destruction of various volatile organic compounds at ppm levels. Two laboratory-scale reactors, an alternating current energized ferroel...

  5. High organic loading rate on thermophilic hydrogen production and metagenomic study at an anaerobic packed-bed reactor treating a residual liquid stream of a Brazilian biorefinery.

    Science.gov (United States)

    Ferraz Júnior, Antônio Djalma Nunes; Etchebehere, Claudia; Zaiat, Marcelo

    2015-06-01

    This study evaluated the influence of a high organic loading rate (OLR) on thermophilic hydrogen production at an up-flow anaerobic packed-bed reactor (APBR) treating a residual liquid stream of a Brazilian biorefinery. The APBR, filled with low-density polyethylene, was operated at an OLR of 84.2 kg-COD m(-3) d(-1). This value was determined in a previous study. The maximum values of hydrogen production and yield were 5,252.6 mL-H2 d(-1) and 3.7 mol-H2 mol(-1)(total carbohydrates), respectively. However, whereas the OLR remained constant, the specific organic load rate (sOLR) decreased throughout operation from 1.38 to 0.72 g-Total carbohydratesg-VS(-1) h(-1), this decrease negatively affected hydrogen production. A sOLR of 0.98 g-Total carbohydratesg-VS(-1) h(-1) was optimal for hydrogen production. The microbial community was studied using 454-pyrosequencing analysis. Organisms belonging to the genera Caloramator, Clostridium, Megasphaera, Oxobacter, Thermoanaerobacterium, and Thermohydrogenium were detected in samples taken from the reactor at operation days 30 and 60, suggesting that these organisms contribute to hydrogen production.

  6. Studies of Neutrino Oscillations at Reactors

    OpenAIRE

    Boehm, Felix

    2000-01-01

    Experiments with reactor neutrinos continue to shed light on our understanding of neutrino oscillations. We review some of the early decisive experiments. We then turn to the recent long baseline oscillation experiments at Palo Verde and Chooz which are leading to the conclusion that the atmospheric neutrino anomaly if attributed to oscillations does not involve an appreciable mixing with the $\\bar\

  7. Reactor safety study. An assessment of accident risks in U. S. commercial nuclear power plants. Appendix XI. Analysis of comments on the draft WASH-1400 report. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    1975-10-01

    Information is presented concerning comments on reactor safety by governmental agencies and civilian organizations; reactor safety study methodology; consequence model; probability of accident sequences; and various accident conditions.

  8. Aqueous two-phase micellar systems in an oscillatory flow micro-reactor: Study of perspectives and experimental performance

    OpenAIRE

    A. M. LOPES; Silva, Daniel Pereira da; A.A. Vicente; Pessoa Júnior, Adalberto; Teixeira, J. A.

    2011-01-01

    Aqueous two-phase micellar systems (ATPMS) are micellar surfactant solutions with physical properties that make them very efficient for the extraction/concentration of biological products. In this work the main proposal that has been discussed is the possible applicability and importance of a novel oscillatory flow micro-reactor (micro-OFR) envisaged for parallel screening and/or development of industrial bioprocesses in ATPMS. Based on the technology of oscillatory flow mixing (OFM), this ba...

  9. ELMO Bumpy Torus Reactor and power plant: conceptual design study

    International Nuclear Information System (INIS)

    A complete power plant design of a 1200-MWe ELMO Bumpy Torus Reactor (EBTR) is presented. An emphasis is placed on those features that are unique to the EBT confinement concept, with subsystems and balance-of-plant items that are more generic to magnetic fusion being adapted from past, more extensive tokamak reactor designs. Similar to the latter tokamak studies, this conceptual EBTR design also emphasizes the use of conventional or near state-of-the-art engineering technology and materials. An emphasis is also placed on system accessibility, reliability, and maintainability, as these crucial and desirable characteristics relate to the unique high-aspect-ratio configuration of EBTs. Equal and strong emphasis is given to physics, engineering/technology, and costing/economics components of this design effort. Parametric optimizations and sensitivity studies, using cost-of-electricity as an object function, are reported. Based on these results, the direction for future improvement on an already attractive reactor design is identified

  10. Application of SAFE to an operating reactor

    International Nuclear Information System (INIS)

    A method for the evaluation of physical protection systems at nuclear facilities has been developed. The evaluation process consists of five major phases: (1) Facility Characterization, (2) Facility Representation, (3) Component Performance, (4) Adversary Path Analysis, and (5) Effectiveness Evaluation. Each of these phases will be described in some detail and illustrated by examples. The process for evaluation of physical protection system effectiveness against an outside threat will be presented for a reactor facility

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

    International Nuclear Information System (INIS)

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

  12. National assessment study in Armenia using innovative nuclear reactors and fuel cycles methodology for an innovative nuclear systems in a country with small grid

    International Nuclear Information System (INIS)

    The International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) was launched in November 2000 under the aegis of the IAEA. Phases 1A and IB (first Part) of the Project were dedicated to elaboration, testing and validation of the INPRO Methodology. At the Technical Meeting in Vienna (13-15 October 2004) Armenia has proposed an assessment using the INPRO Methodology for an Innovative Nuclear Energy System in a country with a small electrical grid. Such kind of study helps Armenia in analysis of Innovative Nuclear Energy System (INS), including fuel cycle options, as well as shows applicability of INPRO methodology for small countries, like Armenia. This study was based on the results given in [3] and [4], and also on the main objectives, declared by the Government of Armenia in the paper 'Energy Sector Development Strategies in the Context of Economic Development in Armenia'

  13. Comparison of three combined sequencing batch reactor followed by enhanced Fenton process for an azo dye degradation: Bio-decolorization kinetics study.

    Science.gov (United States)

    Azizi, A; Alavi Moghaddam, M R; Maknoon, R; Kowsari, E

    2015-12-15

    The purpose of this research was to compare three combined sequencing batch reactor (SBR) - Fenton processes as post-treatment for the treatment of azo dye Acid Red 18 (AR18). Three combined treatment systems (CTS1, CTS2 and CTS3) were operated to investigate the biomass concentration, COD removal, AR18 dye decolorization and kinetics study. The MLSS concentration of CTS2 reached 7200 mg/L due to the use of external feeding in the SBR reactor of CTS2. The COD concentration remained 273 mg/L and 95 mg/L (initial COD=3270 mg/L) at the end of alternating anaerobic-aerobic SBR with external feeding (An-A MSBR) and CTS2, respectively, resulting in almost 65% of Fenton process efficiency. The dye concentration of 500 mg/L was finally reduced to less than 10mg/L in all systems indicating almost complete AR18 decolorization, which was also confirmed by UV-vis analysis. The dye was removed following two successive parts as parts 1 and 2 with pseudo zero-order and pseudo first-order kinetics, respectively, in all CTSs. Higher intermediate metabolites degradation was obtained using HPLC analysis in CTS2. Accordingly, a combined treatment system can be proposed as an appropriate and environmentally-friendly system for the treatment of the azo dye AR18 in wastewater. PMID:26143197

  14. Comparison of three combined sequencing batch reactor followed by enhanced Fenton process for an azo dye degradation: Bio-decolorization kinetics study.

    Science.gov (United States)

    Azizi, A; Alavi Moghaddam, M R; Maknoon, R; Kowsari, E

    2015-12-15

    The purpose of this research was to compare three combined sequencing batch reactor (SBR) - Fenton processes as post-treatment for the treatment of azo dye Acid Red 18 (AR18). Three combined treatment systems (CTS1, CTS2 and CTS3) were operated to investigate the biomass concentration, COD removal, AR18 dye decolorization and kinetics study. The MLSS concentration of CTS2 reached 7200 mg/L due to the use of external feeding in the SBR reactor of CTS2. The COD concentration remained 273 mg/L and 95 mg/L (initial COD=3270 mg/L) at the end of alternating anaerobic-aerobic SBR with external feeding (An-A MSBR) and CTS2, respectively, resulting in almost 65% of Fenton process efficiency. The dye concentration of 500 mg/L was finally reduced to less than 10mg/L in all systems indicating almost complete AR18 decolorization, which was also confirmed by UV-vis analysis. The dye was removed following two successive parts as parts 1 and 2 with pseudo zero-order and pseudo first-order kinetics, respectively, in all CTSs. Higher intermediate metabolites degradation was obtained using HPLC analysis in CTS2. Accordingly, a combined treatment system can be proposed as an appropriate and environmentally-friendly system for the treatment of the azo dye AR18 in wastewater.

  15. Reactor Physics Scoping and Characterization Study on Implementation of TRIGA Fuel in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jennifer Lyons; Wade R. Marcum; Mark D. DeHart; Sean R. Morrell

    2014-01-01

    The Advanced Test Reactor (ATR), under the Reduced Enrichment for Research and Test Reactors (RERTR) Program and the Global Threat Reduction Initiative (GTRI), is conducting feasibility studies for the conversion of its fuel from a highly enriched uranium (HEU) composition to a low enriched uranium (LEU) composition. These studies have considered a wide variety of LEU plate-type fuels to replace the current HEU fuel. Continuing to investigate potential alternatives to the present HEU fuel form, this study presents a preliminary analysis of TRIGA® fuel within the current ATR fuel envelopes and compares it to the functional requirements delineated by the Naval Reactors Program, which includes: greater than 4.8E+14 fissions/s/g of 235U, a fast to thermal neutron flux ratio that is less than 5% deviation of its current value, a constant cycle power within the corner lobes, and an operational cycle length of 56 days at 120 MW. Other parameters outside those put forth by the Naval Reactors Program which are investigated herein include axial and radial power profiles, effective delayed neutron fraction, and mean neutron generation time.

  16. Reactor Physics Scoping and Characterization Study on Implementation of TRIGA Fuel in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR), under the Reduced Enrichment for Research and Test Reactors (RERTR) Program and the Global Threat Reduction Initiative (GTRI), is conducting feasibility studies for the conversion of its fuel from a highly enriched uranium (HEU) composition to a low enriched uranium (LEU) composition. These studies have considered a wide variety of LEU plate-type fuels to replace the current HEU fuel. Continuing to investigate potential alternatives to the present HEU fuel form, this study presents a preliminary analysis of TRIGA® fuel within the current ATR fuel envelopes and compares it to the functional requirements delineated by the Naval Reactors Program, which includes: greater than 4.8E+14 fissions/s/g of 235U, a fast to thermal neutron flux ratio that is less than 5% deviation of its current value, a constant cycle power within the corner lobes, and an operational cycle length of 56 days at 120 MW. Other parameters outside those put forth by the Naval Reactors Program which are investigated herein include axial and radial power profiles, effective delayed neutron fraction, and mean neutron generation time.

  17. An experimental study of the partial oxidation of ethane to ethylene in a shallow fluidized bed reactor

    Directory of Open Access Journals (Sweden)

    DANICA BRZIC

    2007-02-01

    Full Text Available The partial catalytic oxidation of ethane to ethylene was investigated experimentally in a shallow fluidized bed. The performaces of two catalyst types, pure g‑Al2O3 and V2O5/ g-Al2O3 particles 1.8 mm in diameter, were analyzed. A pilot fluidized bed reactor with rectangular cross-section of 100mm´100mm was used. The experiments were carried out under atmospheric pressure in a dilute system under oxygen excess conditions. V2O5/g-Al2O3 showed good catalytic performances regarding ethylene selectivity. The influence of the temperature (in the range of 400–600 °C and the contact time (in the range of 35 – 85 kg sm-3 on the conversion of ethane and the selectivity to ethylene was analyzed. The highest yield of ethylene was 18 %.

  18. Conceptual design study of JSFR reactor building

    Energy Technology Data Exchange (ETDEWEB)

    Yamamoto, T.; Katoh, A.; Chikazawa, Y. [Japan Atomic Energy Agency (JAEA), 4002 Narita, Oarai, Ibaraki 311-1393 (Japan); Ohya, T.; Iwasaki, M.; Hara, H.; Akiyama, Y. [Mitsubishi FBR Systems, Inc. MFBR, 34-17, Jingumae 2-chome, Shibuya, Tokyo 150-0001 (Japan)

    2012-07-01

    Japan Sodium-cooled Fast Reactor (JSFR) is planning to adopt the new concepts of reactor building. One is that the steel plate reinforced concrete is adopted for containment vessel and reactor building. The other is the advanced seismic isolation system. This paper describes the detail of new concepts for JSFR reactor building and engineering evaluation of the new concepts. (authors)

  19. An experimental study on heat transfer from a mixture of solid-fuel and liquid-steel during core disruptive accidents in sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    The relocation of degraded core material through the Control Rod Guide Tubes (CRGTs) is one of essential subjects to achieve the in-vessel retention (IVR) in the case of postulated core disruptive accidents (CDAs) of sodium-cooled fast reactors (SFRs). The CRGT is available as the discharge path by its failure in the core region and heat-transfer from the core-material to the CRGT is one of dominant factors in its failure. In case of a core design into which a fuel subassembly with an inner duct structure (FAIDUS) is introduced, a mixture of solid-fuel and liquid-steel is supposed to remain in the core region since the FAIDUS could effectively eliminate fuel in liquid-state from the core region. Therefore, the objective of the present study is to obtain experimental knowledge for the evaluation of heat-transfer from the mixture of solid-fuel and liquid-steel to the CRGT. In the present study, an experiment was conducted using Impulse Graphite Reactor which is an experimental facility in National Nuclear Center of the Republic of Kazakhstan. In the experiment, the mixture of solid-fuel and liquid-steel was generated by a low-power nuclear heating of fuel and transferring its heat to steel, and then, data to consider the heat-transfer characteristics from the mixture of solid-fuel and liquid-steel to the CRGT were obtained. The heat-transfer characteristic was revealed by evaluating thermocouple responses observed in the experiment. Through the present study, knowledge was obtained to evaluate heat-transfer from the remaining core-materials to the CRGT. (author)

  20. A Downflow Hanging Sponge (DHS) reactor for faecal coliform removal from an Upflow Anaerobic Sludge Bed (UASB) effluent

    NARCIS (Netherlands)

    Yaya Beas, R.E.; Kujawa-Roeleveld, K.; Lier, van J.B.; Zeeman, G.

    2015-01-01

    This research was conducted to study the faecal coliforms removal capacity of Downflow Hanging Sponge (DHS) reactors as a post-treatment for an Upflow Anaerobic Sludge Blanket (UASB) reactor. Three long-term continuous lab-scale DHS reactors i.e. a reactor with cube type sponges without recirculatio

  1. 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

  2. The TITAN reversed-field-pinch fusion reactor study

    International Nuclear Information System (INIS)

    This paper on titan plasma engineering contains papers on the following topics: reversed-field pinch as a fusion reactor; parametric systems studies; magnetics; burning-plasma simulations; plasma transient operations; current drive; and physics issues for compact RFP reactors

  3. The TITAN reversed-field-pinch fusion reactor study

    Energy Technology Data Exchange (ETDEWEB)

    1990-01-01

    This paper on titan plasma engineering contains papers on the following topics: reversed-field pinch as a fusion reactor; parametric systems studies; magnetics; burning-plasma simulations; plasma transient operations; current drive; and physics issues for compact RFP reactors.

  4. Conceptual design study of fusion experimental reactor (FY86 FER)

    International Nuclear Information System (INIS)

    This report describes the results of conceptual design study on plant systems for the Fusion Experimental Reactor (FY86 FER). Design studies for FER plant systems have been continued from FY85, especially for design modifications made in accordance with revisions of plasma scaling parameters and system improvements. This report describes 1) system construction, 2) site and reactor building plan, 3) repaire and maintenance system, 4) tritium circulation system, 5) heating, ventilation and air conditioning system, 6) tritium clean-up system, 7) cooling and baking system, 8) waste treatment and storage system, 9) control system, 10) electric power system, 11) site factory plan, all of which are a part of FY86 design work. The plant systems described in this report generally have been based on the FY86 FER (ACS Reactor) which is an one of the six candidates for FER. (author)

  5. Thermal study of a non adiabatic differential calorimeter used for nuclear heating measurements inside an experimental channel of the Jules Horowitz Reactor

    Science.gov (United States)

    Reynard-Carette, C.; Lyoussi, A.; Brun, J.; Muraglia, M.; Carette, M.; Janulyte, A.; Zerega, Y.; André, J.; Bignan, G.; Chauvin, J.-P.; Fourmentel, D.; Gonnier, C.; Guimbal, P.; Malo, J.-Y.; Villard, J.-F.

    2012-11-01

    New online in-pile measurement methods are crucial during irradiations in Material Testing Reactors (MTR) for a better understanding of accelerated material ageing and nuclear fuel behaviour. In particular, instrumentation for measurements of one relevant parameter: nuclear heat deposition rate, called nuclear heating, has to be improved. The knowledge of this quantity is a great interest for various safety, scientist and end-user requirements (design of specific irradiation devices and associated cooling systems with imposed conditions). This paper focuses on thermal experimental and numerical studies carried out under non irradiation conditions on an in-pile calorimeter dedicated to nuclear heating quantification inside a new experimental device which will be dedicated to the experimental condition mapping (neutron and photon fluxes and nuclear heating) inside the JHR experimental channels. Experimental results concerning the calorimeter response during its electrical calibration (calorimeter design is compared to a single calorimeter which gives promising results (miniaturization, higher sensitivity).

  6. A comparative study of fuel management in PWR reactors

    International Nuclear Information System (INIS)

    A study about fuel management in PWR reactors, where not only the conventional uranium cycle is considered, but also the thorium cycle as an alternative is presented. The final results are presented in terms of U3O8 demand and SWU and the approximate costs of the principal stages of the fuel cycle, comparing with the stardand cycle without recycling. (E.G.)

  7. Mirror Advanced Reactor Study (MARS) final report summary

    International Nuclear Information System (INIS)

    The Mirror Advanced Reactor Study (MARS) has resulted in an overview of a first-generation tandem mirror reactor. The central cell fusion plasma is self-sustained by alpha heating (ignition), while electron-cyclotron resonance heating and negative ion beams maintain the electrostatic confining potentials in the end plugs. Plug injection power is reduced by the use of high-field choke coils and thermal barriers, concepts to be tested in the Tandem Mirror Experiment-Upgrade (TMX-U) and Mirror Fusion Test Facility (MFTF-B) at Lawrence Livermore National Laboratory

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

    Energy Technology Data Exchange (ETDEWEB)

    Rosenthal, Murray Wilford [ORNL

    2009-08-01

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

  9. Experimental study on the operating characteristics of an inner preheating transpiring wall reactor for supercritical water oxidation: Temperature profiles and product properties

    International Nuclear Information System (INIS)

    A new process to generate multiple thermal fluids by supercritical water oxidation (SCWO) was proposed to enhance oil recovery. An inner preheating transpiring wall reactor for SCWO was designed and tested to avoid plugging in the preheating section. Hot water (400–600 °C) was used as auxiliary heat source to preheat the feed to the reaction temperature. The effect of different operating parameters on the performance of the inner preheating transpiring wall reactor was investigated, and the optimized operating parameters were determined based on temperature profiles and product properties. The reaction temperature is close to 900 °C at an auxiliary heat source flow of 2.79 kg/h, and the auxiliary heat source flow is determined at 6–14 kg/h to avoid the overheating of the reactor. The useful reaction time is used to quantitatively describe the feed degradation efficiency. The outlet concentration of total organic carbon (TOCout) and CO in the effluent gradually decreases with increasing useful reaction time. The useful reaction time needed for complete oxidation of the feed is 10.5 s for the reactor. - Highlights: • A new process to generate multiple thermal fluids by SCWO was proposed. • An inner preheating transpiring wall reactor for SCWO was designed and tested. • Hot water was used as auxiliary heat source to preheat the feed at room temperature. • Effect of operating parameters on the performance of the reactor was investigated. • The useful reaction time required for complete oxidation of the feed is 10.5 s

  10. 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.

  11. Studying the effects of dynamical parameters on reactor core temperature

    Directory of Open Access Journals (Sweden)

    R Khodabakhsh

    2015-01-01

    Full Text Available In order to increase productivity, reduce depreciation, and avoid possible accidents in a system such as fuel rods' melting and overpressure, control of temperature changes in the reactor core is an important factor. There are several methods for solving and analysing the stability of point kinetics equations. In most previous analyses, the effects of various factors on the temperature of the reactor core have been ignored. In this work, the effects of various dynamical parameters on the temperature of the reactor core and stability of the system in the presence of temperature feedback reactivity with external reactivity step, ramp and sinusoidal for six groups of delayed neutrons were studied using the method of Lyapunov exponent. The results proved to be in good agreement with other works

  12. Studies of a modular advanced stellarator reactor ASRA6C

    International Nuclear Information System (INIS)

    This study is directed towards the clarification of critical issues of advanced modular stellerator reactors exploiting the inherent potential of steady state operation, and is not a point design study of a reactor. Critical technology issues arise from the three-dimensional magnetic field structure. The first wall, blanket and shield are more complex than those of axi-symmetric systems, but this is eased at moderate to large aspect ratio typical of stellarators. Several blanket options have been studied and a thin blanket (21 cm) was the first choice for the design. Superconducting modular coils were investigated with respect to the conductor and mechanical supports. From the analysis of forces and stresses caused by the electromagnetic loads the coils are considered to be feasible, although shear stresses might pose a critical issue. Demountable intermagnetic support elements were designed for use at separation areas between the cryostat modules. A scheme for remote reactor maintenance was also developed. The plasma physics issues of different configurations were studied using extrapolations of transport behaviour and equilibrium from theory and present experiments. These studies indicate that the confinement and equilibrium behaviour is adequate for ignited operation at an average value of 5% beta. Impurities may pose a critical issue. Several impurity control operations were investigated; a pumped limiter configuration utilizing the 'ergodic layer' at the plasma edge was chosen for edge plasma and impurity control. A general conclusion of the study is that the modular stellerator configuration offers interesting prospects regarding the development towards steady-state reactors. (orig.)

  13. Studies of a modular advanced stellarator reactor ASRA6C

    International Nuclear Information System (INIS)

    This study is directed towards the clarification of critical issues of advanced modular stellerator reactors exploiting the inherent potential of steady state operation, and is not a point design study of a reactor. Critical technology issues arise from the three-dimensional magnetic field structure. The first wall, blanket and shield are more complex than those of axi-symmetric systems, but this is eased at moderate to large aspect ratio typical of stellerators. Several blanket options have been studied and a thin blanket (21 cm) was the first choice for the design. Superconducting modular coils were investigated with respect to the conductor and mechanical supports. From the analysis of forces and stresses caused by the electromagnetic loads the coils are considered to be feasible, although shear stresses might pose a critical issue. Demountable intermagnetic support elements were designed for use at separation areas between the cryostat modules. A scheme for remote reactor maintenance was also developed. The plasma physics issues of different configurations were studied using extrapolations of transort behaviour and equilibrium from theory and present experiments. These studies indicate that the confinement and equilibrium behaviour is adequate for ignited operation at an average value of 5% beta. Impurities may pose a critical issue. Several impurity control operations were investigated; a pumped limiter configuration utilizing the 'ergodic layer' at the plasma edge was chosen for edge plasma and impurity control. A general conclusion of the study is that the modular stellerator configuration offers interesting prospects regarding the development towards steady-state reactors. (orig.)

  14. Reactor safety study. An assessment of accident risks in U.S. commercial nuclear power plants. Appendices III and IV

    International Nuclear Information System (INIS)

    The items listed below summarize the detail sections which follow: a listing of definitions and a discussion of the general treatment of data within the random variable approach as utilized by the study; a tabulation of the assessed data base containing failure classifications, final assessed ranges utilized in quantification and reference source values considered in determining the ranges; a discussion of nuclear power plant experience that was used to validate the data assessment by testing its applicability as well as to check on the adequacy of the model to incorporate typical real incidents; an expanded presentation of the data assessment giving information on applicability considerations; a discussion of test and maintenance data including comparisons of models with experience data; and special topics, including assessments required for the initiating event probabilities and human error data and modeling

  15. An experimental study of the liquid entrainment from swelled two-phase mixture surface in a reactor vessel

    International Nuclear Information System (INIS)

    An experimental study of liquid entrainment by rapid surface swellig of a two-phase mixture in a vessel has been performed. To investigate the effects of air flow and initial water level on the liquid entrainment, a series of experiments have been performed using air and water as working fluids. A total of 64 experimental liquid entrainment rate data have been obtained for various combinations of test parameters (i.e., six different initial water levels and varying air flow rates from 300 to 1,200 lpm) using two test vessels of the same height and different inner diameters (D=0.15 and 0.30m, respectively) for vertical bubbly and churn-turbulent flow conditions. An empirical correlation for the liquid entrainment rate, E, has been developed in terms of the superficial velocity of air, the initial water level, the density of gas, the surface tension, and the gravity. This correlation shows a good agreement with the present experimental data within 30% over a wide range of flow parameters

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

    International Nuclear Information System (INIS)

    This report examines the feasibility of a laser heated solenoid as a fusion or fusion-fission reactor system. The objective of this study, was an assessment of the laser heated solenoid reactor concept in terms of its plasma physics, engineering design, and commercial feasibility. Within the study many pertinent reactor aspects were treated including: physics of the laser-plasma interaction; thermonuclear behavior of a slender plasma column; end-losses under reactor conditions; design of a modular first wall, a hybrid (both superconducting and normal) magnet, a large CO2 laser system; reactor blanket; electrical storage elements; neutronics; radiation damage, and tritium processing. Self-consistent reactor configurations were developed for both pure fusion and fusion-fission designs, with the latter designed both to produce power and/or fissile fuels for conventional fission reactors. Appendix A is a bibliography with commentary of theoretical and experimental studies that have been directed at the laser heated solenoid

  17. Study of space reactors for exploration missions

    Energy Technology Data Exchange (ETDEWEB)

    Cliquet, Elisa; Ruault, Jean-Marc; Masson, Frederic, E-mail: elisa.cliquet@cnes.fr, E-mail: frederic.masson@cnes.fr [Centre National d' Etudes Spatiales (CNES), Paris (France); Roux, Jean-Pierre; Paris, Nicolas; Cazale, Brice; Manifacier, Laurent, E-mail: jean-pierre.roux@areva.com [AREVA TA, Aix en Provence, (France); Poinot-Salanon, Christine, E-mail: christine.poinot@cea.fr [Comissariado a l' Energie Atomique et Aux Energies alternatives (CEA), Paris (France)

    2013-07-01

    Nuclear propulsion has been studied for many decades. The power density of nuclear fission is much higher than chemical process, and for missions to outer solar system requiring several hundred of kilowatts, or for flexible manned missions to Mars requiring several megawatts, nuclear electric propulsion might be the only option offering a reasonable mass in low earth orbit. Despite the existence of low power experiences - SNAP10 in the 60's or Buk/Topaz in the 60-80's - no high power reactor has been developed: investment cost, long term time frame, high technological challenges and radioactive hazards are the main challenges we must overtake. However, it seems reasonable to look at the technical challenges that have to be overcome for a next generation of nuclear electric systems for space exploration. This paper will present some recent studies going on in France, on space reactors for exploration. Three classes of power have been considered: 10kWe, 100kWe, and several megawatts. Available data from previous studies and developments performed in Russia, USA], and Europe, have been collected and gave us a large overview of potential technical solutions. This was the starting point of a trade-off analysis aiming at the selection of the best options, with regards to the technological readiness level in France and Europe. The resulting preliminary designs will be presented and critical technologies needing maturation activities will be highlighted. (author)

  18. The TITAN Reversed-Field Pinch fusion reactor study

    International Nuclear Information System (INIS)

    The TITAN Reversed-Field Pinch (RFP) fusion reactor study is a multi-institutional research effort to determine the technical feasibility and key developmental issues of an RFP fusion reactor, especially at high power density, and to determine the potential economics, operations, safety, and environmental features of high-mass-power-density fusion systems. The TITAN conceptual designs are DT burning, 1000 MWe power reactors based on the RFP confinement concept. The designs are compact, have a high neutron wall loading of 18 MW/m2 and a mass power density of 700 kWe/tonne. The inherent characteristics of the RFP confinement concept make fusion reactors with such a high mass power density possible. Two different detailed designs have emerged: the TITAN-I lithium-vanadium design, incorporating the integrated-blanket-coil concept; and the TITAN-II aqueous loop-in-pool design with ferritic steel structure. This report contains a collection of 16 papers on the results of the TITAN study which were presented at the International Symposium on Fusion Nuclear Technology. This collection describes the TITAN research effort, and specifically the TITAN-I and TITAN-II designs, summarizing the major results, the key technical issues, and the central conclusions and recommendations. Overall, the basic conclusions are that high-mass power-density fusion reactors appear to be technically feasible even with neutron wall loadings up to 20 MW/m2; that single-piece maintenance of the FPC is possible and advantageous; that the economics of the reactor is enhanced by its compactness; and the safety and environmental features need not to be sacrificed in high-power-density designs. The fact that two design approaches have emerged, and others may also be possible, in some sense indicates the robustness of the general findings

  19. The TITAN Reversed-Field Pinch fusion reactor study

    Energy Technology Data Exchange (ETDEWEB)

    1988-03-01

    The TITAN Reversed-Field Pinch (RFP) fusion reactor study is a multi-institutional research effort to determine the technical feasibility and key developmental issues of an RFP fusion reactor, especially at high power density, and to determine the potential economics, operations, safety, and environmental features of high-mass-power-density fusion systems. The TITAN conceptual designs are DT burning, 1000 MWe power reactors based on the RFP confinement concept. The designs are compact, have a high neutron wall loading of 18 MW/m{sup 2} and a mass power density of 700 kWe/tonne. The inherent characteristics of the RFP confinement concept make fusion reactors with such a high mass power density possible. Two different detailed designs have emerged: the TITAN-I lithium-vanadium design, incorporating the integrated-blanket-coil concept; and the TITAN-II aqueous loop-in-pool design with ferritic steel structure. This report contains a collection of 16 papers on the results of the TITAN study which were presented at the International Symposium on Fusion Nuclear Technology. This collection describes the TITAN research effort, and specifically the TITAN-I and TITAN-II designs, summarizing the major results, the key technical issues, and the central conclusions and recommendations. Overall, the basic conclusions are that high-mass power-density fusion reactors appear to be technically feasible even with neutron wall loadings up to 20 MW/m{sup 2}; that single-piece maintenance of the FPC is possible and advantageous; that the economics of the reactor is enhanced by its compactness; and the safety and environmental features need not to be sacrificed in high-power-density designs. The fact that two design approaches have emerged, and others may also be possible, in some sense indicates the robustness of the general findings.

  20. The AFR. An approved network of research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hampel, Gabriele [Mainz Univ. (Germany). Arbeitsgemeinschaft fuer Betriebs- und Sicherheitsfragen an Forschungsreaktoren (AFR)

    2012-10-15

    AFR (Arbeitsgemeinschaft fuer Betriebs- und Sicherheitsfragen an Forschungsreaktoren) is the German acronym for 'Association for Research Reactor Operation and Safety Issues' which was founded in 1959. Reactor managers of European research reactors mainly from the German linguistic area meet regularly for their mutual benefit to exchange experience and knowledge in all areas of operating, managing and utilization of research reactors. In the last 2 years joint meetings were held together with the French association of research reactors CER (Club d'Exploitants des Reacteurs). In this contribution the AFR, its members, work and aims as well as the French partner CER are presented. (orig.)

  1. Preliminary design concept of an advanced integral reactor

    International Nuclear Information System (INIS)

    An integral reactor on the basis of PWR technology is being conceptually developed at KAERI. Advanced technologies such as intrinsic and passive safety features are implemented in establishing the design concepts of the reactor to enhance the safety and performance. Research and development including laboratory-scale tests are concurrently underway for confirming the technical adoption of those concepts to the reactor design. The power output of the reactor will be in the range of 100MWe to 600MWe which is relatively small compared to the existing loop type reactors. The detailed analysis to assure the design concepts is in progress. (author)

  2. Study of an electromagnetic pump applied to a primary main pump of a large scale sodium cooled reactor

    International Nuclear Information System (INIS)

    This paper describes a future innovative design options with a parallel electromagnetic pump (EMP) system as the main circulating pump of the JSFR design. A conceptual design of EMPs integrated with an intermediate heat exchanger (IHX) is carried out. The major design parameters are consistent with the current JSFR design, where the main flow rate is 630 m3/min and the flow halving time is the same of the mechanical pump with the similar reliability. As a result of several design studies, a five parallel EMPs with IHX system has been selected from the geometry suitability for JSFR design. The EMP advantages comparing with mechanical pumps are investigated from the views of in-service inspection, maintenance and reliability. Numerical analysis with two dimensional MHD codes is conducted on a former experiment of a 160 m3/min flow rate EMP. The overall trend of the experimental data and the numerical results agrees with that in small-scale EMPs. However, the difference between the experimental data and the numerical results seems larger compared with the small-scale EMPs, which comes from large magnetic Reynolds number and interaction parameter of 160 m3/min EMP. (author)

  3. Towards an efficient multiphysics model for nuclear reactor dynamics

    Directory of Open Access Journals (Sweden)

    Obaidurrahman K.

    2015-01-01

    Full Text Available Availability of fast computer resources nowadays has facilitated more in-depth modeling of complex engineering systems which involve strong multiphysics interactions. This multiphysics modeling is an important necessity in nuclear reactor safety studies where efforts are being made worldwide to combine the knowledge from all associated disciplines at one place to accomplish the most realistic simulation of involved phenomenon. On these lines coupled modeling of nuclear reactor neutron kinetics, fuel heat transfer and coolant transport is a regular practice nowadays for transient analysis of reactor core. However optimization between modeling accuracy and computational economy has always been a challenging task to ensure the adequate degree of reliability in such extensive numerical exercises. Complex reactor core modeling involves estimation of evolving 3-D core thermal state, which in turn demands an expensive multichannel based detailed core thermal hydraulics model. A novel approach of power weighted coupling between core neutronics and thermal hydraulics presented in this work aims to reduce the bulk of core thermal calculations in core dynamics modeling to a significant extent without compromising accuracy of computation. Coupled core model has been validated against a series of international benchmarks. Accuracy and computational efficiency of the proposed multiphysics model has been demonstrated by analyzing a reactivity initiated transient.

  4. Feasibility study of an optical resonator for applications in neutral-beam injection systems for the next generation of nuclear fusion reactors

    International Nuclear Information System (INIS)

    This work is part of a larger project called SIPHORE (Single gap Photo-neutralizer energy Recovery injector), which aims to enhance the overall efficiency of one of the mechanisms through which the plasma is heated, in a nuclear fusion reactor, i.e. the Neutral Beam Injection (NBI) system. An important component of a NBI system is the neutralizer of high energetic ion beams. SIPHORE proposes to substitute the gas cell neutralizer, used in the current NBI systems, with a photo-neutralizer exploiting the photo-detachment process within Fabry Perot cavities. This mechanism should allow a relevant NBI global efficiency of η≥ 60%, significantly higher than the one currently possible (η≤25% for ITER). The present work concerns the feasibility study of an optical cavity with suitable properties for applications in NBI systems. Within this context, the issue of the determination of an appropriated optical cavity design has been firstly considered and the theoretical and experimental analysis of a particular optical resonator has been carried on. The problems associated with the high levels of intracavity optical power (∼3 MW) required for an adequate photo-neutralization rate have then been faced. In this respect, we addressed both the problem of the thermal effects on the cavity mirrors due to their absorption of intra-cavity optical power (∼1 W) and the one associated to the necessity of a high powerful input laser beam (∼1 kW) to feed the optical resonator. (author)

  5. 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.

  6. Entrained Flow Reactor Study of KCl Capture by Solid Additives

    OpenAIRE

    Wang, Guoliang; Jensen, Peter Arendt; Wu, Hao; Jappe Frandsen, Flemming; Bøjer, M.; Glarborg, Peter

    2016-01-01

    An option for abating deposition and corrosion caused byalkali species during biomass combustion, is the introduction of additivesinto boilers for transforming harmful gaseous alkali compounds (e.g. KCl,KOH) into less corrosive ash species with a higher melting point. Kaolin andcoal fly ash have been proved to be very promising additives and havereceived extensive studies during the past decades. However, mostprevious studies were carried out in fixed-bed reactors where the reaction condition...

  7. Study on operation conditions and an operation system of a nuclear powered submersible research vessel, 'report of working group on application of a very small nuclear reactor to an ocean research'

    Energy Technology Data Exchange (ETDEWEB)

    Ura, Tamaki [Tokyo Univ., Tokyo (Japan); Takamasa, Tomoji [Tokyo Univ. of Mercantile Marine, Tokyo (Japan); Nishimura, Hajime [Japan Marine Science and Technology Center, Yokosuka, Kanagawa (JP)] [and others

    2001-07-01

    JAERI has studied on design of a nuclear powered submersible research vessel, which will navigate under sea mainly in the Arctic Ocean, as a part of the design activity of advanced marine reactors. This report describes operation conditions and an operating system of the vessel, which were discussed by the specialists of hull design, sound positioning, ship motions and oceanography, etc. The design conditions on ship motions for submersible vessels were surveyed considering regulations in our country, and ship motions were evaluated in the cases of underwater and surface navigations taking account of observation activities in the Arctic Ocean. The effect of ship motions on the compact nuclear reactor SCR was assessed. A submarine transponder system and an on-ice communication buoy system were examined as a positioning and communication system, supposing the activity under ice. The interval between transponders or communication buoys was recommended as 130 km. Procedures to secure safety of nuclear powered submersible research vessel were discussed according to accidents on the hull or the nuclear reactor. These results were reflected to the concept of the nuclear powered submersible research vessel, and subjects to be settled in the next step were clarified. (author)

  8. An initial study on modeling the global thermal and fast reactor fuel cycle mass flow using Vensim

    Energy Technology Data Exchange (ETDEWEB)

    Brinton, Samuel [Kansas State University, Mechanical Engineering, Manhattan, KS 66506 (United States)

    2008-07-01

    This study concentrated on modeling the construction and decommissioning rates of five major facilities comprising the nuclear fuel cycle: (1) current LWRs with a 60-year service life, (2) new LWRs burning MOX fuel, (3) new LWRs to replace units in the current fleet, (4) new FRs to be added to the fleet, and (5) new spent fuel reprocessing facilities. This is a mass flow mode starting from uranium ore and following it to spent forms. The visual dynamic modeling program Vensim was used to create a system of equations and variables to track the mass flows from enrichment, fabrication, burn-up, and the back-end of the fuel cycle. The scenarios considered provide estimates of the uranium ore requirements, quantities of LLW and HLW production, and the number of reprocessing facilities necessary to reduce recently reported levels of spent fuel inventory. Preliminary results indicate that the entire national spent fuel inventory produced in the next 100 years can be reprocessed with a reprocessing plant built every 11 years (small capacity) or even as low as every 23 years (large capacity). (authors)

  9. Kinetics study of heterogeneous reactions of ozone with erucic acid using an ATR-IR flow reactor.

    Science.gov (United States)

    Leng, Chunbo; Hiltner, Joseph; Pham, Hai; Kelley, Judas; Mach, Mindy; Zhang, Yunhong; Liu, Yong

    2014-03-01

    The ozone initiated heterogeneous oxidation of erucic acid (EA) thin film was investigated using a flow system combined with attenuated total reflection infrared spectroscopy (ATR-IR) over wide ranges of ozone concentrations (0.25-60 ppm), thin film thickness (0.1-1.0 μm), temperatures (263-298 K), and relative humidities (0-80% RH) for the first time. Pseudo-first-order rate constants, kapp, and overall reactive uptake coefficients, γ, were obtained through changes in the absorbance of C[double bond, length as m-dash]O stretching bands at 1695 cm(-1), which is assigned to the carbonyl group in carboxylic acid. Results showed that the reaction followed the Langmuir-Hinshelwood mechanism and kapp was largely dominated by surface reaction over bulk phase reaction. In addition, both the kapp and the γ values showed very strong temperature dependences (∼two orders of magnitude) over the temperature range; in contrast, they only slightly increased with increasing RH values from 0-80%. According to the kapp values as a function of temperature, the activation energy for the heterogeneous reaction was estimated to be 80.6 kJ mol(-1). Our results have suggested that heterogeneous reactions between ozone and unsaturated solid surfaces likely have a substantially greater temperature dependence than liquid ones. Moreover, the hygroscopic properties of EA thin films before and after exposure to ozone were also studied by measurement of water uptake. Based on the hygroscopicity data, the insignificant RH effect on reaction kinetics was probably due to the relatively weak water uptake by the unreacted and reacted EA thin films.

  10. The impact of fuel cladding failure events on occupational radiation exposures at nuclear power plants: Case study, PWR [pressurized-water reactor] during an outage

    International Nuclear Information System (INIS)

    This report is the second in a series of case studies designed to evaluate the magnitude of increase in occupational radiation exposures at commercial US nuclear power plants resulting from small incidents or abnormal events. The event evaluated is fuel cladding failure, which can result in elevated primary coolant activity and increased radiation exposure rates within a plant. For this case study, radiation measurements were made at a pressurized-water reactor (PWR) during a maintenance and refueling outage. The PWR had been operating for 22 months with fuel cladding failure characterized as 105 pin-hole leakers, the equivalent of 0.21% failed fuel. Gamma spectroscopy measurements, radiation exposure rate determinations, thermoluminescent dosimeter (TLD) assessments, and air sample analyses were made in the plant's radwaste, pipe penetration, and containment buildings. Based on the data collected, evaluations indicate that the relative contributions of activation products and fission products to the total exposure rates were constant over the duration of the outage. This constancy is due to the significant contribution from the longer-lived isotopes of cesium (a fission product) and cobalt (an activation product). For this reason, fuel cladding failure events remain as significant to occupational radiation exposure during an outage as during routine operations. As documented in the previous case study (NUREG/CR-4485 Vol. 1), fuel cladding failure events increased radiation exposure rates an estimated 540% at some locations of the plant during routine operations. Consequently, such events can result in significantly greater radiation exposure rates in many areas of the plant during the maintenance and refueling outages than would have been present under normal fuel conditions

  11. Yale High Energy Physics Research: Precision Studies of Reactor Antineutrinos

    Energy Technology Data Exchange (ETDEWEB)

    Heeger, Karsten M. [Yale Univ., New Haven, CT (United States)

    2014-09-13

    This report presents experimental research at the intensity frontier of particle physics with particular focus on the study of reactor antineutrinos and the precision measurement of neutrino oscillations. The experimental neutrino physics group of Professor Heeger and Senior Scientist Band at Yale University has had leading responsibilities in the construction and operation of the Daya Bay Reactor Antineutrino Experiment and made critical contributions to the discovery of non-zero$\\theta_{13}$. Heeger and Band led the Daya Bay detector management team and are now overseeing the operations of the antineutrino detectors. Postdoctoral researchers and students in this group have made leading contributions to the Daya Bay analysis including the prediction of the reactor antineutrino flux and spectrum, the analysis of the oscillation signal, and the precision determination of the target mass yielding unprecedented precision in the relative detector uncertainty. Heeger's group is now leading an R\\&D effort towards a short-baseline oscillation experiment, called PROSPECT, at a US research reactor and the development of antineutrino detectors with advanced background discrimination.

  12. Feasibility study of a pilot scale molten salt reactor demonstration

    International Nuclear Information System (INIS)

    Energy Process Developments Ltd. was awarded a grant by Innovate UK in July 2014 to undertake a year-long project to determine the feasibility of developing a pilot scale molten salt reactor in the UK. The study looked at six current available proposed MSR configurations and proposed the immediate next steps for design and build of a chosen demonstrator reactor project. Tremendous knowledge growth in the 60 years of the first nuclear era has not seen substantial advances in nuclear fission technology much beyond the Pressurised Water Reactor, initially a hastily adopted device for military and civil applications, and essentially comprising water cooling of solid fuel elements. The imminent second nuclear era requires introduction of inherently more efficient, safer, cheaper, nuclear power obtainable with liquid-fuelled - namely Molten Salt Reactor (MSR) technology - the best out of the six Gen IV options. This Gen IV option, when considered in 2002, was believed to be decades away from readiness. This study reviews more recent work. The evidence is that the MSR is ready now. In the immediate urgency of the present, this liquid-fuelled reactor technology can be seen as highly innovative, necessary and rewarding. It is ready to form a key part of any affordable policy proposals for the UK energy supply. This feasibility study is seen as the first step towards full scale implementation of the technology. MSRs are passively safe, operate at atmospheric pressure, at higher efficiencies than PWRs and can be load following. Thorium is the ultimate fuel of choice which can provide the world with a near limitless supply of energy. A demonstration reactor will show the media, public and investors that this technology exists as a clean source of cheap sustainable power. The project reviewed the status of all MSR activity internationally, the regulatory regime in the UK and potential sites. Nuclear insurers were consulted on their insurability and the outlook of an energy

  13. Neutronic study regarding transmutation fuel research at Jules Horowitz Reactor

    International Nuclear Information System (INIS)

    In order to estimate the possibilities for transmutation experiments at the Jules Horowitz Reactor several ideas for neutronic and fuel behaviour studies are investigated at CEA Cadarache. Naturally an exact replication of the burning of minor actinides in fast reactors, as expected in most transmutation scenarios, is impossible, but some key transmutation parameters can be investigated in a MTR neutron spectrum. In this paper a parametric study regarding fuel damage by He and fission products in AmUO2 is presented. By varying flux level, uranium enrichment and americium content of the sample in the JHR reflector a He production to fission ratio comparable to reference samples in the core of a SFR can be achieved. The calculations were done with the depletion code DARWIN2.2 using JEF2.2 data and spectra from a TRIPOLI model of JHR and an ERANOS model for the SFR respectively. (author)

  14. Oregon State TRIGA reactor power calibration study

    International Nuclear Information System (INIS)

    As a result of a recent review of the Oregon State TRIGA Reactor (OSTR) power calibration procedure, an investigation was performed on the origin and correctness of the OSTR tank factor and the calibration method. It was determined that there was no clear basis for the tank factor which was being used (0.0525 deg. C/kwh) and therefore a new value was calculated (0.0493 deg. C/kwh). The calculational method and likely errors are presented in the paper. In addition, a series of experimental tests were conducted to decide if the power calibration was best performed with or without a mixer, at 100 KW or at 1 MW. The results of these tests along with the final recommendation are presented. (author)

  15. Experimental studies on catalytic hydrogen recombiners for light water reactors

    International Nuclear Information System (INIS)

    In the course of core melt accidents in nuclear power plants a large amount of hydrogen can be produced and form an explosive or even detonative gas mixture with aerial oxygen in the reactor building. In the containment atmosphere of pressurized water reactors hydrogen combines a phlogistically with the oxygen present to form water vapor even at room temperature. In the past, experimental work conducted at various facilities has contributed little or nothing to an understanding of the operating principles of catalytic recombiners. Hence, the purpose of the present study was to conduct detailed investigations on a section of a recombiner essentially in order to deepen the understanding of reaction kinetics and heat transport processes. The results of the experiments presented in this dissertation form a large data base of measurements which provides an insight into the processes taking place in recombiners. The reaction-kinetic interpretation of the measured data confirms and deepens the diffusion theory - proposed in an earlier study. Thus it is now possible to validate detailed numeric models representing the processes in recombiners. Consequently the present study serves to broaden and corroborate competence in this significant area of reactor technology. In addition, the empirical knowledge thus gained may be used for a critical reassessment of previous numeric model calculations. (orig.)

  16. An apparatus for the study of high temperature water radiolysis in a nuclear reactor: calibration of dose in a mixed neutron/gamma radiation field.

    Science.gov (United States)

    Edwards, Eric J; Wilson, Paul P H; Anderson, Mark H; Mezyk, Stephen P; Pimblott, Simon M; Bartels, David M

    2007-12-01

    The cooling water of nuclear reactors undergoes radiolytic decomposition induced by gamma, fast electron, and neutron radiation in the core. To model the process, recombination reaction rates and radiolytic yields for the water radical fragments need to be measured at high temperature and pressure. Yields for the action of neutron radiation are particularly hard to determine independently because of the beta/gamma field also present in any reactor. In this paper we report the design of an apparatus intended to measure neutron radiolysis yields as a function of temperature and pressure. A new methodology for separation of neutron and beta/gamma radiolysis yields in a mixed radiation field is proposed and demonstrated.

  17. Economics of an advanced pressurized water reactor (APWR)

    International Nuclear Information System (INIS)

    Limited natural uranium resources together with their low utilization in current lightwater reactors (LWR) on the one hand and the high capital investments for a LWR and fast breeder reactor system resulting in a high fuel utilization are the most important reasons for research and development (R+D) work related to a high converting APWR system. It is the main task of this analysis to study the economic behaviour of an APWR combining its technical and physical parameters with an economic data base. After introductional remarks chapter II presents the potential improvements of the uranium utilization and their importance for a whole national economy. Chapter III shows the micro-economic aspects for an electricity producing utility using such an APWR plant. The chapter is restricted to the fuel cycle costs and their dependence on various parameters. The corresponding costs of other nuclear power plants are described in chapter IV and compared to those of the APWR in chapter V. Finally a cost comparison on the basis of the electricity generating costs will complete the economic picture of an advanced pressurized water reactor. (orig./UA)

  18. An advanced method of heterogeneous reactor theory

    International Nuclear Information System (INIS)

    Recent approaches to heterogeneous reactor theory for numerical applications were presented in the course of 8 lectures given in JAERI. The limitations of initial theory known after the First Conference on Peacefull Uses of Atomic Energy held in Geneva in 1955 as Galanine-Feinberg heterogeneous theory:-matrix from of equations, -lack of consistent theory for heterogeneous parameters for reactor cell, -were overcome by a transformation of heterogeneous reactor equations to a difference form and by a development of a consistent theory for the characteristics of a reactor cell based on detailed space-energy calculations. General few group (G-number of groups) heterogeneous reactor equations in dipole approximation are formulated with the extension of two-dimensional problem to three-dimensions by finite Furie expansion of axial dependence of neutron fluxes. A transformation of initial matrix reactor equations to a difference form is presented. The methods for calculation of heterogeneous reactor cell characteristics giving the relation between vector-flux and vector-current on a cell boundary are based on a set of detailed space-energy neutron flux distribution calculations with zero current across cell boundary and G calculations with linearly independent currents across the cell boundary. The equations for reaction rate matrices are formulated. Specific methods were developed for description of neutron migration in axial and radial directions. The methods for resonance level's approach for numerous high-energy resonances. On the basis of these approaches the theory, methods and computer codes were developed for 3D space-time react or problems including simulation of slow processes with fuel burn-up, control rod movements, Xe poisoning and fast transients depending on prompt and delayed neutrons. As a result reactors with several thousands of channels having non-uniform axial structure can be feasibly treated. (author)

  19. Simulation of hydrogen distribution in an Indian Nuclear Reactor Containment

    Energy Technology Data Exchange (ETDEWEB)

    Prabhudharwadkar, Deoras M. [Department of Mechanical Engineering, Indian Institute of Technology, Mumbai (India); Iyer, Kannan N., E-mail: kiyer@iitb.ac.i [Department of Mechanical Engineering, Indian Institute of Technology, Mumbai (India); Mohan, Nalini; Bajaj, Satinder S. [Nuclear Power Corporation of India Ltd., Mumbai (India); Markandeya, Suhas G. [Bhabha Atomic Research Centre, Trombay, Mumbai (India)

    2011-03-15

    Research highlights: This work addresses hydrogen dispersion in commercial nuclear reactor containment. The numerical tool used for simulation is first benchmarked with experimental data. Parametric results are then carried out for different release configurations. Results lead to the conclusion that the dispersal is buoyancy dominated. Also, the hydrogen concentration is high enough to demand mitigation devices. - Abstract: The management of hydrogen in a Nuclear Reactor Containment after LOCA (Loss Of Coolant Accident) is of practical importance to preserve the structural integrity of the containment. This paper presents the results of systematic work carried out using the commercial Computational Fluid Dynamics (CFD) software FLUENT to assess the concentration distribution of hydrogen in a typical Indian Nuclear Reactor Containment. In order to obtain an accurate estimate of hydrogen concentration distribution, a suitable model for turbulence closure is required to be selected. Using guidelines from the previous studies reported in the literature and a comparative simulation study using simple benchmark problems, the most suitable turbulence model for hydrogen mixing prediction was identified. Subsequently, unstructured meshes were generated to represent the containment of a typical Indian Nuclear Reactor. Analyses were carried out to quantify the hydrogen distribution for three cases. These were (1) Uniform injection of hydrogen for a given period of time at room temperature, (2) Time varying injection as has been computed from an accident analysis code, (3) Time varying injection (as used in case (2)) at a high temperature. A parametric exercise was also carried out in case (1) where the effect of various inlet orientations and locations on hydrogen distribution was studied. The results indicate that the process of hydrogen dispersal is buoyancy dominated. Further for typical injection rates encountered following LOCA, the dispersal is quite poor and most

  20. Preliminary design concepts of an advanced integral reactor

    International Nuclear Information System (INIS)

    An integral reactor on the basis of PWR technology is being conceptually developed at KAERI. Advanced technologies such as intrinsic and passive safety features are implemented in establishing the design concepts of the reactor to enhance the safety and performance. Research and development including laboratory-scale tests are concurrently underway for confirming the technical adoption of those concepts to the rector design. The power output of the reactor will be in the range of 100MWe to 600MWe which is relatively small compared to the existing loop type reactors. The detailed analysis to assure the design concepts is in progress. (author). 3 figs, 1 tab

  1. Systems study of tokamak fusion--fission reactors

    Energy Technology Data Exchange (ETDEWEB)

    Tenney, F.H.; Bathke, C.G.; Price, W.G. Jr.; Bohlke, W.H.; Mills, R.G.; Johnson, E.F.; Todd, A.M.M.; Buchanan, C.H.; Gralnick, S.L.

    1978-11-01

    This publication reports the results of a two to three year effort at a systematic analysis of a wide variety of tokamak-driven fissioning blanket reactors, i.e., fusion--fission hybrids. It addresses the quantitative problems of determining the economically most desirable mix of the two products: electric power and fissionable fuel and shows how the price of electric power can be minimized when subject to a variety of constraints. An attempt has been made to avoid restricting assumptions, and the result is an optimizing algorithm that operates in a six-dimensional parameter space. Comparisons are made on sets of as many as 100,000 distinct machine models, and the principal results of the study have been derived from the examination of several hundred thousand possible reactor configurations.

  2. Systems study of tokamak fusion--fission reactors

    International Nuclear Information System (INIS)

    This publication reports the results of a two to three year effort at a systematic analysis of a wide variety of tokamak-driven fissioning blanket reactors, i.e., fusion--fission hybrids. It addresses the quantitative problems of determining the economically most desirable mix of the two products: electric power and fissionable fuel and shows how the price of electric power can be minimized when subject to a variety of constraints. An attempt has been made to avoid restricting assumptions, and the result is an optimizing algorithm that operates in a six-dimensional parameter space. Comparisons are made on sets of as many as 100,000 distinct machine models, and the principal results of the study have been derived from the examination of several hundred thousand possible reactor configurations

  3. A study of reactor vessel integrity assessment

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Hoon [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Kim, Jong Kyung; Shin, Chang Ho; Seo, Bo Kyun [Hanyang Univ., Seoul (Korea, Republic of)

    1999-02-15

    The fast neutron fluence at the Reactor Pressure Vessel(RPV) of KNGR designed for 60 years lifetime was calculated by full-scope Monte Carlo simulation for reactor vessel integrity assessment. KNGR core geometry was modeled on a three-dimensional representation of the one-sixteenth of the reactor in-vessel component. Each fuel assemblies were modeled explicitly, and each fuel pins were axially divided into 5 segments. The maximum flux of 4.3 x 10{sup 10} neutrons/cm{sup 2}. sec at the RPV was obtained by tallying neutrons crossing the beltline of inner surface of the RPV.

  4. Preliminary Study for Conceptual Design of Advanced Long Life Small Modular Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tak, Taewoo; Choe, Jiwon; Jeong, Yongjin; Lee, Deokjung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Kim, T. K. [Argonne National Laboratory, Argonne (United States)

    2015-05-15

    As one of the non-water coolant Small-Modular Reactor (SMR) core concepts for use in the mid- to long-term, ANL has proposed a 100 MWe Advanced sodium-cooled Fast Reactor core concept (AFR-100) targeting a small grid, transportable from pre-licensed factories to the remote plant site for affordable supply. Various breed-and-burn core concepts have been proposed to extend the reactor cycle length, which includes CANDLE with a cigar-type depletion strategy, TerraPower reactors with fuel shuffling for effective breeding, et al. UNIST has also proposed an ultra-long cycle fast reactor (UCFR) core concept having the power rating of 1000 MWe. By adopting the breed-and-burn strategies, the UCFR core can maintain criticality for a targeting reactor lifetime of 60 years without refueling. The objective of this project is to develop an advanced long-life SMR core concept by adopting both the small modular design features of the AFR-100 and the long-life breed-and-burn concept of the UCFR. A conceptual design of long life small modular fast reactor is under development by adopting both the small modular design features of the AFR-100 and the long-life breed-and-burn concept of the UCFR. The feasibility of the long-life fast reactor concepts was reviewed to obtain the core design guidelines and the reactor design requirements of long life small modular fast reactor were proposed in this study.

  5. Studies and research concerning BNFP: converting reprocessing plant's fuel receiving and storage area to an away-from-reactor (AFR) storage facility. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Cottrell, Jim E.; Shallo, Frank A.; Musselwhite, E Larry; Wiedemann, George F.; Young, Moylen

    1979-09-01

    Converting a reprocessing plant's fuel receiving and storage station into an Away-From-Reactor storage facility is evaluated in this report. An engineering analysis is developed which includes (1) equipment modifications to the facility including the physical protection system, (2) planning schedules for licensing-related activities, and (3) cost estimates for implementing such a facility conversion. Storage capacities are evaluated using the presently available pools of the existing Barnwell Nuclear Fuel Plant-Fuel Receiving and Storage Station (BNFP-FRSS) as a model.

  6. Studies and research concerning BNFP: converting reprocessing plant's fuel receiving and storage area to an away-from-reactor (AFR) storage facility. Final report

    International Nuclear Information System (INIS)

    Converting a reprocessing plant's fuel receiving and storage station into an Away-From-Reactor storage facility is evaluated in this report. An engineering analysis is developed which includes (1) equipment modifications to the facility including the physical protection system, (2) planning schedules for licensing-related activities, and (3) cost estimates for implementing such a facility conversion. Storage capacities are evaluated using the presently available pools of the existing Barnwell Nuclear Fuel Plant-Fuel Receiving and Storage Station (BNFP-FRSS) as a model

  7. Study on Modeling Technology in Digital Reactor System

    Institute of Scientific and Technical Information of China (English)

    刘晓平; 罗月童; 童莉莉

    2004-01-01

    Modeling is the kernel part of a digital reactor system. As an extensible platform for reactor conceptual design, it is very important to study modeling technology and develop some kind of tools to speed up preparation of all classical computing models. This paper introduces the background of the project and basic conception of digital reactor. MCAM is taken as an example for modeling and its related technologies used are given. It is an interface program for MCNP geometry model developed by FDS team (ASIPP & HUT), and designed to run on windows system. MCAM aims at utilizing CAD technology to facilitate creation of MCNP geometry model. There have been two ways for MCAM to utilize CAD technology:(1) Making use of user interface technology in aid of generation of MCNP geometry model;(2) Making use of existing 3D CAD model to accelerate creation of MCNP geometry model. This paper gives an overview of MCAM's major function. At last, several examples are given to demonstrate MCAM's various capabilities.

  8. Experimental Study on Granular Sludge Cultivation Based on IC Reactor

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    [Objective]We aimed to discuss the optimal conditions of sludge granulation on the basis of IC reactor.[Method]By using the formulated glucose wastewater,we studied the rapid cultivation of granular sludge as well as its influencing factors,so as to discuss the optimal conditions of sludge granulation.[Result]Through the static culture outside of IC reactor and dynamic culture in IC reactor,granular sludge appeared within a training cycle,with particle size of 1.0-1.5 mm.In IC reactor,when COD concentration...

  9. Oak Ridge Tokamak experimental power reactor study scoping report

    Energy Technology Data Exchange (ETDEWEB)

    Roberts, M.

    1977-03-01

    This report presents the scoping studies performed as the initial part of the program to produce a conceptual design for a Tokamak Experimental Power Reactor (EPR). The EPR as considered in this study is to employ all systems necessary for significant electric power production at continuous high duty cycle operation; it is presently scheduled to be the final technological step before a Demonstration Reactor Plant (Demo). The scoping study tasks begin with an exploration and identification of principal problem areas and then concentrate on consideration and evaluation of alternate design choices for each of the following major systems: Plasma Engineering and Physics, Nuclear, Electromagnetics, Neutral Beam Injection, and Tritium Handling. In addition, consideration has been given to the integration of these systems and requirements arising out of their incorporation into an EPR. One intent of this study is to document the paths explored in search of the appropriate EPR characteristics. To satisfy this intent, the explorations are presented in chart form outlining possible options in key areas with extensive supporting footnotes. An important result of the scoping study has been the development and definition of an EPR reference design to serve as (1) a common focus for the continuing design study and (2) a guide for associated development programs. In addition, the study has identified research and development requirements essential to facilitate the successful conceptual design, construction, and operation of an EPR.

  10. Neutronics Study of the KANUTER Space Propulsion Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Venneri, Paolo; Nam, Seung Hyun; Kim, Yonghee [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-05-15

    The Korea Advanced Nuclear Thermal Engine Rocket (KANUTER) has been developed at the Korea Advanced Institute of Science and Technology (KAIST). This space propulsion system is unique in that it implements a HEU fuel with a thermal spectrum system. This allows the system to be designed with a minimal amount of fissile material and an incredibly small and light system. This then allows the implementation of the system in a cluster format which enables redundancy and easy scalability for different mission requirements. This combination of low fissile content, compact size, and thermalized spectrum contribute to an interesting and novel behavior of the reactor system. The two codes were both used for the burn up calculations in order to verify their validity while the static calculations and characterization of the core were done principally with MCNPX. The KANUTER space propulsion reactor is in the process of being characterized and improved. Its basic neutronic characteristics have been studied, and its behavior over time has been identified. It has been shown that this reactor will have difficulty operating as hoped in a bimodal configuration where it is able to provide both propulsion and power throughout mission to Mars. The reason for this has been identified as Xe{sup 135}, and it is believed that a possible solution to this issue does exist, either in the form of an appropriately designed neutron spectrum or the building in of sufficient excess reactivity.

  11. Conceptual Design Study of JSFR (2) - Reactor System

    International Nuclear Information System (INIS)

    Several innovative technologies are adopted in the JSFR design to meet the high level requirements for economic competitiveness in the design requirements. The cost-down approaches for JSFR are as follows. In order to reduce the amount of structural materials, the diameter of the reactor vessel shall be minimized and the reactor internal structures shall be simplified. The reduction of the reactor vessel diameter is achieved by adopting a advanced refueling system and the hot reactor vessel with high temperature wall. The flow velocity in the reactor upper plenum increases because the diameter of the reactor vessel is decreased. As the result, the coolant flow field in reactor upper plenum is severe. The optimization of the coolant flow field in the reactor upper plenum was carried out for prevention the cover gas entrainment and the vortex cavitations at the hot leg intake. In addition, structural integrities for seismic loadings and thermal loadings were evaluated because the design seismic loading was highly increased and the vessel wall is directly exposed to the thermal transients of the upper plenum. This paper describes the characteristics and the results of the design study of the reactor system. (author)

  12. Introducing an ILS methodology into research reactors

    International Nuclear Information System (INIS)

    Integrated Logistics Support (ILS) is the managerial organisation that co-ordinates the activities of many disciplines to develop the supporting resources (training, staffing, designing aids, equipment removal routes, etc) required by technologically complex systems. The application of an ILS methodology in defence projects is described in several places, but it is infrequently illustrated for other areas; therefore the present paper deals with applying this approach to research reactors under design or already in operation. Although better results are obtained when applied since the very beginning of a project, it can be applied successfully in facilities already in operation to improve their capability in a cost-effective way. In applying this methodology, the key objectives shall be previously identified in order to tailor the whole approach. Generally in high power multipurpose reactors, obtaining maximum profit at the lowest possible cost without reducing the safety levels are key issues, while in others the goal is to minimise drawbacks like spurious shutdowns, low quality experimental results or even to reduce staff dose to ALARA values. These items need to be quantified for establishing a system status base line in order to trace the process evolution. Thereafter, specific logistics analyses should be performed in the different areas composing the system. RAMS (Reliability, Availability, Maintainability and Supportability), Manning, Training Needs, Supplying Needs are some examples of these special logistic assessments. The following paragraphs summarise the different areas, encompassed by this ILS methodology. Plant design is influenced focussing the designers? attention on the objectives already identified. Careful design reviews are performed only in an early design stage, being useless a later application. In this paper is presented a methodology including appropriate tools for ensuring the designers abide to ILS issues and key objectives through the

  13. Reactor core conversion studies of Ghana: Research Reactor-1 and proposal for addition of safety rod

    International Nuclear Information System (INIS)

    The inclusion of an additional safety rod in conjunction with a core conversion study of Ghana Research Reactor-1 (GHARR-1) was carried out using neutronics, thermal hydraulics and burnup codes. The study is based on a recommendation by Integrated Safety Assessment for Research Reactors (INSARP) mission to incorporate a safety rod to the reactor safety system as well as the need to replace the reactor fuel with LEU. Conversion from one fuel type to another requires a complete re-evaluation of the safety analysis. Changes to the reactivity worth, shutdown margin, power density and material properties must be taken into account, and appropriate modifications made. Neutronics analysis including burnup was studied followed by thermal hydraulics analyses which comprise steady state and transients. Four computer codes were used for the analysis; MCNP, REBUS, PLTEP and PARET. The neutronics analysis revealed that the LEU core must be operated at 34 Kw in order to attain the flux of 1.0E12 n/cm2.s as the nominal flux of the HEU core. The auxiliary safety rod placed at a modified irradiation site gives a better worth than the cadmium capsules. For core excess reactivity of 4 mk, 348 fuel pins would be appropriate for the GHARR-1 LEU core. Results indicate that flux level of 1.0E12 n/cm2.s in the inner irradiation channel will not be compromised, if the power of the LEU core is increased to 34 kW. The GHARR-1 core using LEU-U02-12.5% fuel can be operated for 23 shim cycles, with cycles length 2.5 years, for over 57 years at the 17 kW power level. All 23 LEU cycles meet the ∼ 4.0 mk excess reactivity required at the beginning of cycle . For comparison, the MNSR HEU reference core can also be operated for 23 shim cycles, but with a cycle length of 2.0 years for just over 46 years at 15.0kW power level. It is observed that the GHARR-1 core with LEU UO2 fuel enriched to 12.5% and a power level of 34 kW can be operated ∼25% longer than the current HEU core operated at 30 k

  14. Nuclear Power from Fission Reactors. An Introduction.

    Science.gov (United States)

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

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

  15. Essential metal depletion in an anaerobic reactor.

    NARCIS (Netherlands)

    Osuna, M.B.; Iza, J.M.; Zandvoort, M.H.; Lens, P.N.L.

    2003-01-01

    The effect of the absence of trace elements on the conversion of a mixture of volatile fatty acids by a distillery anaerobic granular sludge was investigated. Two UASB reactors were operated under identical operational conditions except for the influent trace metal concentrations, during 140 days. E

  16. A photochemical reactor for studies of atmospheric chemistry

    DEFF Research Database (Denmark)

    Nilsson, Elna Johanna Kristina; Eskebjerg, Carsten; Johnson, Matthew Stanley

    2009-01-01

    A photochemical reactor for studies of atmospheric kinetics and spectroscopy has been built at the Copenhagen Center for Atmospheric Research. The reactor consists of a vacuum FTIR spectrometer coupled to a 100 L quartz cylinder by multipass optics mounted on electropolished stainless steel end...

  17. Exposure mode study to xenon-133 in a reactor building

    International Nuclear Information System (INIS)

    The work described in this thesis focuses on the external and internal dose assessment to xenon-133. During the nuclear reactor operation, fission products and radioactive inert gases, as 133Xe, are generated and might be responsible for the exposure of workers in case of clad defect. Particle Monte Carlo transport code is adapted in radioprotection to quantify dosimetric quantities. The study of exposure to xenon-133 is conducted by using Monte-Carlo simulations based on GEANT4, an anthropomorphic phantom, a realistic geometry of the reactor building, and compartmental models. The external exposure inside a reactor building is conducted with a realistic and conservative exposure scenario. The effective dose rate and the eye lens equivalent dose rate are determined by Monte-Carlo simulations. Due to the particular emission spectrum of xenon-133, the equivalent dose rate to the lens of eyes is discussed in the light of expected new eye dose limits. The internal exposure occurs while xenon-133 is inhaled. The lungs are firstly exposed by inhalation, and their equivalent dose rate is obtained by Monte-Carlo simulations. A biokinetic model is used to evaluate the internal exposure to xenon-133. This thesis gives us a better understanding to the dosimetric quantities related to external and internal exposure to xenon-133. Moreover the impacts of the dosimetric changes are studied on the current and future dosimetric limits. The dosimetric quantities are lower than the current and future dosimetric limits. (author)

  18. The inner containment of an EPR trademark pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ostermann, Dirk; Krumb, Christian; Wienand, Burkhard [AREVA GmbH, Offenbach (Germany)

    2014-08-15

    On February 12, 2014 the containment pressure and subsequent leak tightness tests on the containment of the Finnish Olkiluoto 3 EPR trademark reactor building were completed successfully. The containment of an EPR trademark pressurized water reactor consists of an outer containment to protect the reactor building against external hazards (such as airplane crash) and of an inner containment that is subjected to internal overpressure and high temperature in case of internal accidents. The current paper gives an overview of the containment structure, the design criteria, the validation by analyses and experiments and the containment pressure test.

  19. The study meeting report on the undermoderated spectrum reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nakajima, Nobuya; Ochiai, Masaaki [eds.

    1998-09-01

    The interest to the high converter or in the breeder is rising as the research and the development of the light water-type nuclear reactor in future. A study session about the undermoderated spectrum reactor of the Japan Atomic Energy Research Institute (JAERI) sponsorship was held in March, 1998 4, on the 5th. This report is the contents of the study session. The study session began with the basis lecture to entitle to be `The expectations to the undermoderated core study` almost. Next, the review of the high conversion-type core study about PWR and BWR was reported. As the undermoderated spectrum MOX core study, the latest situation of (1) the development of the supercritical pressure water reactor, (2) the development of RBWR, (3) the development of the advanced fuel cycle by BWR and (4) the development of the pressurized water-type breeder were reported from the university and the maker. As also the study present situation and the plan in future in JAERI, there was an explanation about the design study of the undermoderated spectrum core and the actinide research facility. The panel discussion lastly, to entitle to be `Undermoderated MOX core research and development of the future and the technical issues` was done. There was an opinion about the way of carrying forward concerned research and development, the acceptability of the society, the view of the future, the cooperation of the electric power or the desire to JAERI and there was wide inquiry replying. The 9 of the presented papers are indexed individually. (J.P.N.)

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

    Science.gov (United States)

    Carlson, Eric D.; Gast, Alice P.

    1998-01-01

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

  1. Design study of plant system for the fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    This report describes design study results of the FER plant system. The purpose of this study is to have an image of the FER plant system as a whole by designing major auxiliary systems, reactor building and maintenance and radwaste desposal systems. The major auxiliary systems include tritium, cooling, evacuation and fueling systems. For these each systems, flowdiagrams are studied and designs of devices and pipings are conducted. In the reactor building design, layout of the above auxiliary systems in the building is studied with careful zoning concept by the radiation level. Structural integrity of the reactor building is also studied including seismic analysis. In the design of the maintenance and radwaste system flowdiagram of failed reactor components is developed and transfer vehicles and buildings are designed. Finally assuming JAERI Naka site as the reactor site layout of the whole FER plant system is developed. (author)

  2. Numerical study on seismic response of the reactor coolant pump in Advanced Passive Pressurized Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    De, Cheng, E-mail: 0100209064@sjtu.edu.cn; Zhen-Qiang, Yao, E-mail: zqyaosjtu@gmail.com; Ya-bo, Xue; Hong, Shen

    2014-10-15

    Highlights: • An artificial accelerogram of the specified SSE is generated. • A dynamic FE model of the RCP in AP1000 (with gyroscopic and FSI effects) is developed. • The displacement, force, moment and stress in the RCP during the earthquake are summarized. - Abstract: The reactor coolant pump in the Advanced Passive Pressurized Water Reactor is a kind of nuclear canned-motor pump. The pump is classified as Seismic Category I, which must function normally during the Safe Shutdown Earthquake. When the nuclear power plant is located in seismically active region, the seismic response of the reactor coolant pump may become very important for the safety assessment of the whole nuclear power plant. In this article, an artificial accelerogram is generated. The response spectrum of the artificial accelerogram fits well with the design acceleration spectrum of the Safe Shutdown Earthquake. By applying the finite element modeling method, the dynamic finite element models of the rotor and stator in the reactor coolant pump are created separately. The rotor and stator are coupled by the journal bearings and the annular flow between the rotor and stator. Then the whole dynamic model of the reactor coolant pump is developed. Time domain analysis which uses the improved state-space Newmark method of a direct time integration scheme is carried out to investigate the response of the reactor coolant pump under the horizontal seismic load. The results show that the reactor coolant pump responds differently in the direction of the seismic load and in the perpendicular direction. During the Safe Shutdown Earthquake, the displacement response, the shear force, the moment and the journal bearing reaction forces in the reactor coolant pump are analyzed.

  3. An introduction to the engineering of fast nuclear reactors

    CERN Document Server

    Judd, Anthony M

    2014-01-01

    An invaluable resource for both graduate-level engineering students and practising nuclear engineers who want to expand their knowledge of fast nuclear reactors, the reactors of the future! This book is a concise yet comprehensive introduction to all aspects of fast reactor engineering. It covers topics including neutron physics; neutron flux spectra; flux distribution; Doppler and coolant temperature coefficients; the performance of ceramic and metal fuels under irradiation, structural changes, and fission-product migration; the effects of irradiation and corrosion on structural materials, irradiation swelling; heat transfer in the reactor core and its effect on core design; coolants including sodium and lead-bismuth alloy; coolant circuits; pumps; heat exchangers and steam generators; and plant control. The book includes new discussions on lead-alloy and gas coolants, metal fuel, the use of reactors to consume radioactive waste, and accelerator-driven subcritical systems.

  4. Advanced nuclear reactor systems - an Indian perspective

    International Nuclear Information System (INIS)

    The Indian nuclear power programme envisages use of closed nuclear fuel cycle and thorium utilisation as its mainstay for its sustainable growth. The current levels of deployment of nuclear energy in India need to be multiplied nearly hundred fold to reach levels of electricity generation that would facilitate the country to achieve energy independence as well as a developed status. The Indian thorium based nuclear energy systems are being developed to achieve sustainability in respect of fuel resource along with enhanced safety and reduced waste generation. Advanced Heavy Water Reactor and its variants have been designed to meet these objectives. The Indian High Temperature Reactor programme also envisages use of thorium-based fuel with advanced levels of passive safety features. (author)

  5. Transverse flow reactor studies of the dynamics of radical reactions

    Energy Technology Data Exchange (ETDEWEB)

    Macdonald, R.G. [Argonne National Laboratory, IL (United States)

    1993-12-01

    Radical reactions are in important in combustion chemistry; however, little state-specific information is available for these reactions. A new apparatus has been constructed to measure the dynamics of radical reactions. The unique feature of this apparatus is a transverse flow reactor in which an atom or radical of known concentration will be produced by pulsed laser photolysis of an appropriate precursor molecule. The time dependence of individual quantum states or products and/or reactants will be followed by rapid infrared laser absorption spectroscopy. The reaction H + O{sub 2} {yields} OH + O will be studied.

  6. Studies on the transient operation and stability of fast reactors

    International Nuclear Information System (INIS)

    These studies form part of the general programme of perfecting calculation methods for fast reactors. The basic formulae are given for the layouts used, i.e. the classic kinetic and thermal exchange equations, etc. A description is then given of the digital computer methods employed for studying the stable functioning of the reactor and of the methods used for transient operation studies. Finally, some examples of application are discussed and a comparison is made with parallel studies on the same subject. (author)

  7. Conceptual study of a straight field line mirror hybrid reactor

    International Nuclear Information System (INIS)

    A hybrid reactor based on the straight field line mirror (SFLM) with magnetic expanders at the ends is proposed as a compact device for transmutation of nuclear waste and power production. Compared to a fusion reactor, plasma confinement demands can be relaxed if there is a strong energy multiplication by the fission reactions, i.e. Qr=Pfission/Pfusion>>1. The values of Qr is primarily restricted by fission reactor safety requirements. For the SFLM, computations suggest that values of Qr ranging up to 150 are consistent with reactor safety. In a mirror hybrid device with Qr >100, the lower bound on the electron temperature for power production can then be estimated to be around 400 eV, which may be achievable for a mirror machine. The SFLM with its quadrupolar stabilizing fields does not rely on plasma flow into the expanders for MHD stability, and a scenario with plasma density depletion in the expanders is a possibility to increase the electron temperature. Efficient power production is predicted with a fusion Q = 0.15 and an electron temperature around 500 eV. A fusion power of 10 MW could then be amplified to 1.5 GW fission power in a compact 25 m long hybrid mirror machine. Beneficial features are that all sensitive equipment can be located outside the neutron rich region and a steady state power production seems possible. Self circulation of the lead coolant, which is useful for heat removal if coolant pumps cease to operate, could be arranged by orienting the magnetic axis vertically. Results from studies on plasma equilibrium and stability, coil designing, RF heating and neutron computations are presented.

  8. A preliminary conceptual design study for Korean fusion DEMO reactor

    International Nuclear Information System (INIS)

    Highlights: ► Perform a preliminary conceptual study for a steady-state Korean DEMO reactor. ► Present design guidelines and requirements of Korean DEMO reactor. ► Present a preliminary design of TF (toroidal field) and CS (central solenoid) magnet. ► Present a preliminary result of the radial build scheme of Korean DEMO reactor. -- Abstract: As the ITER is being constructed, there is a growing anticipation for an earlier realization of fusion energy, so called fast-track approach. Korean strategy for fusion energy can be regarded as a fast-track approach and one special concept discussed in this paper is a two-stage development plan. At first, a steady-state Korean DEMO Reactor (K-DEMO) is designed not only to demonstrate a net electricity generation and a self-sustained tritium cycle, but also to be used as a component test facility. Then, at its second stage, a major upgrade is carried out by replacing in-vessel components in order to show a net electric generation on the order of 300 MWe and the competitiveness in cost of electricity (COE). The major radius is designed to be just below 6.5 m, considering practical engineering feasibilities. By using high performance Nb3Sn-based superconducting cable currently available, high magnetic field at the plasma center above 8 T can be achieved. A design concept for TF magnets and radial builds for the K-DEMO considering a vertical maintenance scheme, are presented together with preliminary design parameters

  9. Binary breeder reactor: an option for Brazilian energy future

    International Nuclear Information System (INIS)

    To assure a continued supply of electric energy beyond a few decades from now, developmemnt of fast breeder reactors is a necessity. Binary fueled LMFBRs combine an improvement in the inherent safety of fast reactors and an efficient use of the abundant thorium. A nuclear system that starts with PWRs and gradually shifts to a FBR system or to a FBR-PWR symbiotic system appears to be the most reasonable one for Brazil. Natural uranium requirements and possible sequences of reactor introductions are discussed for some postulated Brazilian situations. A permanent system of approx. 100 GWe capacity can be established based on the estimated resource of natural uranium. (Author)

  10. Binary breeder reactor an option for Brazilian energy future

    International Nuclear Information System (INIS)

    To assure a continued supply of electric energy beyond a few decades from now, development of fast breeder reactors is a necessity. Binary fueled LMFBRs combine an improvement in the inherent safety of fast reactors and an efficient use of the abundant thorium. A nuclear system that starts with PWRs and gradually shifts to a FBR system or to a FBR-PWR symbiotic system appears to be the most resonable one for Brazil. Natural uranium requirements and possible sequences of reactor introductions are discussed for some postulated Brazilian situations. A permanent system of approximatelly 100 GWe capacity can be established based on the estimated resource of natural uranium. (Author)

  11. Steady-state spheromak reactor studies

    International Nuclear Information System (INIS)

    After summarizing the essential elements of a gun-sustained spheromak, the potential for a steady-state is explored by means of a comprehensive physics/engineering/costing model. A range of cost-optimized reactor design points is presented, and the sensitivity of cost to key physics, engineering, and operational variables is reported

  12. NGNP Reactor Coolant Chemistry Control Study

    Energy Technology Data Exchange (ETDEWEB)

    Brian Castle

    2010-11-01

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

  13. Development of an underwater AUT system for reactor walls

    International Nuclear Information System (INIS)

    KAERI(Korea Atomic Energy Research Institute) developed the KSNP(Korea Standard Nuclear Power Plant) in 1984. It was designed to generate 100MKw of electric power. The first KSNP was Ulchin Unit 3 constructed by Kepco(Korea Electric Power Corporation) in 1998. Korea has 6 KSNPs now. These NPPs have pressurized water reactors. It must stand a 150-160 air pressure and 300 degrees centigrade heat. If there are some defects in the reactor, these conditions may cause serious accidents such as a loss of national electric power and human lives. The reactor is made of carbon steel. It consists of a head, a body and a bottom head. There are welding areas on the body and bottom head. These welding areas are the weak points of the pressurized water reactor. The regular maintenance procedures for the nuclear power plant safety instruments are executed during the overhaul period every fourteen months in a KSNP. The duration of an overhaul is 3 weeks. The reactor inspection is executed based on an international standard code such as the ASME(American Society of Mechanical Engineers) code. The UT inspection method is adapted for a reactor welding area inspection. It must be executed in radioactive water because contaminated water can not be moved to on other place. It takes a long time to execute this inspection by the traditional equipment. We developed an automated and compact system to inspect the KSNP reactor welding areas

  14. An overview of future sustainable nuclear power reactors

    Directory of Open Access Journals (Sweden)

    Andreas Poullikkas

    2013-01-01

    Full Text Available In this paper an overview of the current and future nuclear power reactor technologies is carried out. In particular, the nuclear technology is described and the classification of the current and future nuclear reactors according to their generation is provided. The analysis has shown that generation II reactors currently in operation all around the world lack significantly in safety precautions and are prone to loss of coolant accident (LOCA. In contrast, generation III reactors, which are an evolution of generation II reactors, incorporate passive or inherent safety features that require no active controls or operational intervention to avoid accidents in the event of malfunction, and may rely on gravity, natural convection or resistance to high temperatures. Today, partly due to the high capital cost of large power reactors generating electricity and partly due to the consideration of public perception, there is a shift towards the development of smaller units. These may be built independently or as modules in a larger complex, with capacity added incrementally as required. Small reactors most importantly benefit from reduced capital costs, simpler units and the ability to produce power away from main grid systems. These factors combined with the ability of a nuclear power plant to use process heat for co-generation, make the small reactors an attractive option. Generally, modern small reactors for power generation are expected to have greater simplicity of design, economy of mass production and reduced installation costs. Many are also designed for a high level of passive or inherent safety in the event of malfunction. Generation III+ designs are generally extensions of the generation III concept, which include advanced passive safety features. These designs can maintain the safe state without the use of any active control components. Generation IV reactors, which are future designs that are currently under research and development, will

  15. An Investigation into Water Chemistry in Primary Coolant Circuit of an Advanced Boiling Water Reactor

    International Nuclear Information System (INIS)

    To ensure operation safety, an optimization on the coolant chemistry in the primary coolant circuit of a nuclear reactor is essential no matter what type or generation the reactor belongs to. For a better understanding toward the water chemistry in an advanced boiling water reactor (ABWR), such as the one being constructed in the northern part of Taiwan, and for a safer operation of this ABWR, we conducted a proactive, thorough water chemistry analysis prior to the completion of this reactor in this study. A numerical simulation model for water chemistry analyses in ABWRs has been developed, based upon the core technology we established in the past. This core technology for water chemistry modeling is basically an integration of water radiolysis, thermal-hydraulics, and reactor physics. The model, by the name of DEMACE-ABWR, is an improved version of the original DEMACE model and was used for radiolysis and water chemistry prediction in the Longmen ABWR in Taiwan. Predicted results pertinent to the water chemistry variation and the corrosion behavior of structure materials in the primary coolant circuit of this ABWR under rated-power operation were reported in this paper. (authors)

  16. Reactors licensing: proposal of an integrated quality and environment regulatory structure for nuclear research reactors in Brazil

    International Nuclear Information System (INIS)

    A new integrated regulatory structure based on quality and integrated issues has been proposed to be implemented on the licensing process of nuclear research reactors in Brazil. The study starts with a literature review about the licensing process in several countries, all of them members of the International Atomic Energy Agency. After this phase it is performed a comparative study with the Brazilian licensing process to identify good practices (positive aspects), the gaps on it and to propose an approach of an integrated quality and environmental management system, in order to contribute with a new licensing process scheme in Brazil. The literature review considered the following research nuclear reactors: Jules-Horowitz and OSIRIS (France), Hanaro (Korea), Maples 1 and 2 (Canada), OPAL (Australia), Pallas (Holand), ETRR-2 (Egypt) and IEA-R1 (Brazil). The current nuclear research reactors licensing process in Brazil is conducted by two regulatory bodies: the Brazilian National Nuclear Energy Commission (CNEN) and the Brazilian Institute of Environment and Renewable Natural Resources (IBAMA). CNEN is responsible by nuclear issues, while IBAMA by environmental one. To support the study it was applied a questionnaire and interviews based on the current regulatory structure to four nuclear research reactors in Brazil. Nowadays, the nuclear research reactor’s licensing process, in Brazil, has six phases and the environmental licensing process has three phases. A correlation study among these phases leads to a proposal of a new quality and environmental integrated licensing structure with four harmonized phases, hence reducing potential delays in this process. (author)

  17. Study of plutonium recycling physics in light water reactors

    International Nuclear Information System (INIS)

    A stock of plutonium from the reprocessing of thermal neutron reactor fuel is likely to appear in the next few years. The use of this plutonium as fuel replacing 235U in thermal reactors is probably more interesting than simple stock-piling storage: immobilization of a capital which moreover would deteriorate by radioactive decay of isotope 241 also fissile and present to an appreciable extend in plutonium from reprocessing (half-life 15 years); recycling, on the other hand, will supply energy without complete degradation of the stock for fast neutron reactor loads, the burned matter having been partially renewed by conversion; furthermore the use of plutonium will meet the needs created by a temporary pressure on the naturel and/or enriched uranium market. For these two reasons the recycling of plutonium in thermal neutron reactors is being considered seriously today. The present work is confined to neutronic aspects and centres mainly on pressurized water-moderated reactors, the most highly developed at present in France. Four aspects of the problem are examined: 1. the physics of a plutonium-recycling reactor special features of neutronic phenomena with respect to the 'conventional' scheme of the 235U burning reactor; 2. calculation of a plutonium-recycling reactor: adaptation of standard methods; 3. qualification of these calculations from the viewpoint of both data and inevitable approximations; 4. the fuel cycle and particularly the equivalence of fissile matters

  18. Safety aspects of an inertial confinement fusion reactor

    International Nuclear Information System (INIS)

    Releases into the environment of radioactive materials contained in heavy ion fusion (HIF) reactor plants must be prevented by similar safety design concepts as they are applied to present fission converter (e.g. LWR's) and breeder reactors (LMFBR's). This study is intended to identify significant safety aspects of inertial confinement fusion power plant concepts and to relate them to the more familliar basis of knowledge about the safety and the hazards of other advanced nuclear power reactor systems such as the LMFBR. Needs for safety related research and development specifically for inertial confinement fusion are pointed out. (orig./GG)

  19. Reactor-accelerator coupled experiments (RACE): A feasibility study at TAMU

    International Nuclear Information System (INIS)

    A series of accelerator driven system (ADS) experiments are being planned to conduct demonstration and benchmark studies involving a nuclear reactor in a subcritical condition coupled to an accelerator driven neutron source. These experiments are being planned to use various levels of criticality and various power levels allowing for kinetics evaluation with and without temperature feedback. The reactor fuel to be used in these experiments is TRIGA reactor fuel. These reactor fuels are inherently safe and the reactor systems involved have large coolant capacities to ensure safe operation. The cores to be considered will be fully-instrumented and will allow for detailed and accurate data on core power levels, temperatures, and neutron fluxes during the experiments. A feasibility study is currently in progress at Texas A and M Univ. (TAMU) to determine the viability and capability of these experiments. Preliminary results of this feasibility study are given. (authors)

  20. A Study of Reactor Neutrino Monitoring at Experimental Fast Reactor JOYO

    CERN Document Server

    Furuta, H; Hara, T; Haruna, T; Ishihara, N; Ishitsuka, M; Ito, C; Katsumata, M; Kawasaki, T; Konno, T; Kuze, M; Maeda, J; Matsubara, T; Miyata, H; Nagasaka, Y; Nitta, K; Sakamoto, Y; Suekane, F; Sumiyoshi, T; Tabata, H; Takamatsu, M; Tamura, N

    2011-01-01

    We carried out a study of neutrino detection at the experimental fast reactor JOYO using a 0.76 tons gadolinium loaded liquid scintillator detector. The detector was set up on the ground level at 24.3m from the JOYO reactor core of 140MW thermal power. The measured neutrino event rate from reactor on-off comparison was 1.11\\pm1.24(stat.)\\pm0.46(syst.)events/day. Although the statistical significance of the measurement was not enough, the background in such a compact detector at the ground level was studied in detail and MC simulation was found to describe the data well. A study for improvement of the detector for future such experiments is also shown.

  1. Influence of long-term thermal aging on the microstructural evolution of nuclear reactor pressure vessel materials: An atom probe study

    Energy Technology Data Exchange (ETDEWEB)

    Pareige, P.; Russell, K.F.; Stoller, R.E.; Miller, M.K. [Oak Ridge National Lab., TN (United States)

    1998-03-01

    Atom probe field ion microscopy (APFIM) investigations of the microstructure of unaged (as-fabricated) and long-term thermally aged ({approximately} 100,000 h at 280 C) surveillance materials from commercial reactor pressure vessel steels were performed. This combination of materials and conditions permitted the investigation of potential thermal-aging effects. This microstructural study focused on the quantification of the compositions of the matrix and carbides. The APFIM results indicate that there was no significant microstructural evolution after a long-term thermal exposure in weld, plate, or forging materials. The matrix depletion of copper that was observed in weld materials was consistent with the copper concentration in the matrix after the stress-relief heat treatment. The compositions of cementite carbides aged for 100,000 h were compared with the Thermocalc{trademark} prediction. The APFIM comparisons of materials under these conditions are consistent with the measured change in mechanical properties such as the Charpy transition temperature.

  2. An overview of inertial fusion reactor design

    International Nuclear Information System (INIS)

    Recent progress in the conceptual design of inertial fusion reaction chambers and power plants is reviewed. A discussion of expected operating parameters and a brief historical perspective are provided to organize the rich array of chamber and driver concepts. The technical feasibility of several reaction chamber concepts is discussed, along with technical issues that require future analysis, experiment, and development. Where these chambers have been integrated into a power plant design, the characteristics are described. Finally, requirements on the future development of inertial fusion reactor technology are discussed

  3. A Simple Parallel Photochemical Reactor for Photodecomposition Studies

    Science.gov (United States)

    Xiaobo Chen; Halasz, Sarah M.; Giles, Eric C.; Mankus, Jessica V.; Johnson, Joseph C.; Burda, Clemens

    2006-01-01

    A simple and useful parallel photochemical reactor intended to study the photodecomposition of dyes using semiconductor photocatalysis is presented. The photochemical reactions are followed through time-dependent changes in the ground-state absorption spectra of the dyes.

  4. Modal analysis of an ECC duct for APR+ reactor barrel

    International Nuclear Information System (INIS)

    Advanced Power Reactor Plus (APR+) provides four Direct Vessel Injection (DVI) ducts on the reactor barrel to enhance the performance of Emergency Core Cooling System (ECCS). Several studies on safety analysis have verified the excellent performance of the DVI duct. In this study, from the viewpoint of mechanical integrity, modal analyses of two full-scaled DVI ducts have been presented; both numerical analysis and modal tests have been performed in air and water. It was found that the numerical simulation and modal test coincide with each other. The DVI duct is a thin shell of 5 mm thickness, so that harmonic responses to RCP blade passing frequencies should be checked. The dominant passing frequencies are known to be 20, 40, 60, 120 and 240 Hz. In addition, an interesting thing in this study is that added mass effect by coolant seems to be so significant that the natural frequency of the ducts under water could be considerably low as compared with those in air; the natural frequency under water is 60 % lower than that in air. (author)

  5. Study of a multi-beam accelerator driven thorium reactor

    International Nuclear Information System (INIS)

    The primary advantages that accelerator driven systems have over critical reactors are: (1) Greater flexibility regarding the composition and placement of fissile, fertile, or fission product waste within the blanket surrounding the target, and (2) Potentially enhanced safety brought about by operating at a sufficiently low value of the multiplication factor to preclude reactivity induced events. The control of the power production can be achieved by vary the accelerator beam current. Furthermore, once the beam is shut off the system shuts down. The primary difference between the operation of an accelerator driven system and a critical system is the issue of beam interruptions of the accelerator. These beam interruptions impose thermo-mechanical loads on the fuel and mechanical components not found in critical systems. Studies have been performed to estimate an acceptable number of trips, and the value is significantly less stringent than had been previously estimated. The number of acceptable beam interruptions is a function of the length of the interruption and the mission of the system. Thus, for demonstration type systems and interruption durations of 1sec 5mins 2500/yr and 50/yr are deemed acceptable. However, for industrial scale power generation without energy storage type systems and interruption durations of t 5mins, the acceptable number of interruptions are 25000, 2500, 250, and 3 respectively. However, it has also been concluded that further development is required to reduce the number of trips. It is with this in mind that the following study was undertaken. The primary focus of this study will be the merit of a multi-beam target system, which allows for multiple spallation sources within the target/blanket assembly. In this manner it is possible to ameliorate the effects of sudden accelerator beam interruption on the surrounding reactor, since the remaining beams will still be supplying source neutrons. The proton beam will be assumed to have an

  6. Parametric study of the Tormac fusion reactor concept

    International Nuclear Information System (INIS)

    A preliminary but comprehensive power balance for the D-T Tormac magnetic fusion reactor concept is examined parametrically in order to establish general scaling relationships, tradeoffs, and constraints. The results are based on the simplifying assumptions of steady-state operation, a homogeneous plasma, and ideal thin-sheath, mirror-like confinement. Crucial physics uncertainties requiring further theoretical and experimental research attention are identified. Representative reactor physics operating points are generated to illustrate anticipated Tormac reactor embodiments. This study should be considered preliminary to a more detailed physics and technology modeling effort and is intended only to scope and identify possible operating points, parametric sensitivities, and potential physics/technological problems

  7. The TITAN Reversed-Field Pinch Reactor: Design-point determination and parametric studies

    International Nuclear Information System (INIS)

    The multi-institutional TITAN study has examined the physics, technology, safety, and economics issues associated with the operation of a Reversed-Field Pinch (RFP) magnetic fusion reactor at high power density. A comprehensive system and trade study have been conducted as an integral and ongoing part of the reactor assessment. Attractive design points emerging from these parametric studies are subjected to more detailed analysis and design integration, the results of which are used to refine the parametric systems model. The design points and tradeoffs for two TITAN/RFP reactor embodiments are discussed. 14 refs

  8. Study of enzymatic reactors with microencapsulated lipase. Doctoral thesis. Estudo de reactores enzimaticos com lipase microencapsulada

    Energy Technology Data Exchange (ETDEWEB)

    de Franca Teixeira dos Prazeres, D.M.

    1992-10-01

    The work reports the development of a membrane reactor for the hydrolysis of triglycerides catalyzed by lipase B from Chromobacterium viscosum in AOT/isooctane reversed miceller system. In a preliminary phase the potential of the organic system was evaluated with comparative studies on the activity and stability of lipase B in aqueous media (emulsion) and in the alternative miceller media. A tubular ceramic membrane reactor with recirculation was selected for the olive oil hydrolysis in a reversed miceller system. The influence of the hydration degree, recirculation rate, AOT, olive oil and lipase concentration in the operation of the reactor were investigated in a batch mode. The hydration degree was identified as a critical parameter due to its influence in the separation process and in the kinetics of the reaction.

  9. Calculation of fundamental parameters for the dynamical study of TRIGA-3-Salazar reactor (Mixed reactor core)

    International Nuclear Information System (INIS)

    Kinetic parameters for dynamic study of two different configurations, 8 and 9, both with standard fuel, 20% enrichment and Flip (Fuel Life Improvement Program with 70% enrichment) fuel, for TRIGA Mark-III reactor from Mexico Nuclear Center, are obtained. A calculation method using both WIMS-D4 and DTF-IV and DAC1 was established, to decide which of those two configurations has the best safety and operational conditions. Validation of this methodology is done by calculate those parameters for a reactor core with new standard fuel. Configuration 9 is recommended to be use. (Author)

  10. Feasibility study of potential LEU fuels for generic MNSR reactor

    International Nuclear Information System (INIS)

    A feasibility study was performed for a generic Miniature Source Reactor (MNSR) reactor to identify potential LEU fuels that could be used for conversion of these reactors. The model that was used and analysis results obtained with the current HEU fuel are first described. Potential LEU fuels are then listed. The study results were used to identify fuels that may be suitable for use in MNSR conversions, the qualification status of these fuels, and potential manufacturers. Five LEU fuels and core designs were identified that match the excess reactivity of the HEU core. Conversion of MNSR reactors to these fuels would be technically feasible if all safety requirements are shown to be satisfied, the fuels are qualified and licensed, and manufacturers are available to fabricate the fuels at a reasonable price. (author)

  11. Specifications of an accelerator for the soliton hybrid reactor (RHYS)

    International Nuclear Information System (INIS)

    The soliton hybrid reactor is a concept of an Accelerator Driven System, with a design insuring a long lifetime without core interventions. Soliton reactors and 'candle reactors' have been proposed in order to use reactors for very long periods without reprocessing or enrichment. In this paper, we present the concept of hybrid soliton reactor. In this system, the constant displacement of the beam during the 30 years lifetime implies lower constraints on the window compared to other accelerator-driven systems. During its lifetime, the reactor can present constant profiles in chemical and isotopic composition and in power production if the beam of protons is maintained within certain limits. This is the soliton-like behavior. Using a mathematical analysis of the problem, we have shown that the solution of the equations presents a solitary wave behavior which is stable if the accelerator intensity is sufficiently low or if the velocity of the neutron source is sufficiently high. We have simulated these two behaviors with a GEANT III Monte Carlo program: a soliton behavior, and a one which may become unstable when the intensity is too large or the velocity too small. These preliminary considerations allow us to describe some specifications concerning an accelerator which can drive such a system

  12. Hydraulic characterization of an activated sludge reactor with recycling system by tracer experiment and analytical models.

    Science.gov (United States)

    Sánchez, F; Viedma, A; Kaiser, A S

    2016-09-15

    Fluid dynamic behaviour plays an important role in wastewater treatment. An efficient treatment requires the inexistence of certain hydraulic problems such as dead zones or short-circuiting flows. Residence time distribution (RTD) analysis is an excellent technique for detecting these inefficiencies. However, many wastewater treatment installations include water or sludge recycling systems, which prevent us from carrying out a conventional tracer pulse experiment to obtain the RTD curve of the installation. This paper develops an RTD analysis of an activated sludge reactor with recycling system. A tracer experiment in the reactor is carried out. Three analytical models, derived from the conventional pulse model, are proposed to obtain the RTD curve of the reactor. An analysis of the results is made, studying which model is the most suitable for each situation. This paper is useful to analyse the hydraulic efficiency of reactors with recycling systems.

  13. Hydraulic characterization of an activated sludge reactor with recycling system by tracer experiment and analytical models.

    Science.gov (United States)

    Sánchez, F; Viedma, A; Kaiser, A S

    2016-09-15

    Fluid dynamic behaviour plays an important role in wastewater treatment. An efficient treatment requires the inexistence of certain hydraulic problems such as dead zones or short-circuiting flows. Residence time distribution (RTD) analysis is an excellent technique for detecting these inefficiencies. However, many wastewater treatment installations include water or sludge recycling systems, which prevent us from carrying out a conventional tracer pulse experiment to obtain the RTD curve of the installation. This paper develops an RTD analysis of an activated sludge reactor with recycling system. A tracer experiment in the reactor is carried out. Three analytical models, derived from the conventional pulse model, are proposed to obtain the RTD curve of the reactor. An analysis of the results is made, studying which model is the most suitable for each situation. This paper is useful to analyse the hydraulic efficiency of reactors with recycling systems. PMID:27288672

  14. An analysis of reactor transient response for boiling water reactor ATWS events

    International Nuclear Information System (INIS)

    Numerical simulations of BWR (boiling water reactor) dynamic response under ATWS (anticipated transient without scram) conditions are presented for the case where the reactor is operated at natural circulation conditions. In non-isolation events, reactor stability is strongly influenced by the degree of core inlet subcooling. At normal water level and pressure, instabilities develop if core-inlet subcooling exceeds a critical value of ∼21 Btu/lbm. A sensitivity study with regard to the steam separator pressure - loss coefficient, however, indicates that system stability is strongly dependent on the magnitude of this parameter which suggests a significant degree of uncertainty In the results. Under isolation conditions at rated pressure, stability is significantly enhanced by rapid pressure fluctuations generated through cycling of safety/relief valves. Large-amplitude instabilities develop, however, in depressurization events, and SRV cycling no longer stabilizes the system. In a simulated depressurization to ∼500 psia, prompt critical excursions occurred, and oscillation amplitudes reached 1000% of rated power. Implications of the Present Study: With the exception of guidance to avoid SRV cycling, these preliminary results have provided further support for the validity of this response strategy. SABRE calculations have shown that the reactor is probably slightly unstable in natural circulation operation, but relief valve cycling prevents the occurrence of instabilities at or near design pressure. Thus, unstable operation should not be a concern when boron injection and HPCI are available and depressurization is unnecessary. The reactor water level for injection flow corresponding to HPCI operation has been shown to be acceptable and consistent with earlier estimates based on the NSAC results; however, condensation effects have a significant influence on the equilibrium reactor water level. In addition, the SABRE results reinforce PP and L's concerns that

  15. Feasibility study on small modular reactors for modern microgrids

    International Nuclear Information System (INIS)

    Microgrid is a solution of conventional power grid problem and offer sustainable decentralized power system. Microgrid with modern distributed energy resources (DER) could play an important role to alleviate dependency on the main electricity grid. Distributed energy resource comprises wind turbine, solar photovoltaic, diesel generator, gas engine, micro turbine, fuel cells, etc.Due to the gap between typical loads and supply within microgrid, larger scale energy generation could provide a possible solution to balance power demand and supply. Feasibility study of Small Nuclear Power Plant, such as Small Modular reactor (SMR), within microgrids could be achieved via different cases. To achieve the target, a comprehensive feasibility study is conducted on microgrid with SMR through electricity generation profiles, geographical and environmental assessment, as well as cost analysis using simulation practices and data analysis.Also potency of SMRs is analyzed. Parameters and Key Performance Indicators (KPIs) could be analyzed to achieve feasible solution of microgrids with small modular reactor (SMR) to improve the overall microgrid performance.The study shows that SMR could be a feasible solution if microgrid parameters are selected properly. (author)

  16. Nordic study on reactor waste. Technical part 1 and 2

    International Nuclear Information System (INIS)

    An important part of the Nordic studies on system- and safety analysis of the management of low and medium level radioactive waste from nuclear power plants, is the safety analysis of a Reference System. This reference system was established within the study and is described in this Technical Part 1. The reference system covers waste management Schemes that are potential possibilities in either one of the four participating Nordic countries. The reference system is based on: a power reactor system consisting of 6 BWR's of 500 MWe each, operated simultaneously over the same 30 year period, and deep bed granular ion exchange resin wastes from the Reactor Water Clean-Up System (RWCS and powdered ion exchange resin from the Spent Fuel Pool Cleanup System (SFPCS)). Both waste types are supposed to be solidified by mixing with cement and bitumen. Two basic types of containers are considered. Standard 200 liter steel drums and specially made cubicreinforced concrete moulds with a net volume of 1 m3. The Nordic study assumes temporary storage of the solidified waste for a maximum of 50 years before the waste is transferred to the disposal site. Transportation of the waste from the storage facilitiy to the disposal site will be by road or sea. Three different disposal facilities are considered: Shallow land burial, near surface concrete bunker, and rock cavern with about 30 m granite cover. (EG)

  17. Feasibility study on small modular reactors for modern microgrids

    Energy Technology Data Exchange (ETDEWEB)

    Islam, R.; Gabbar, H.A., E-mail: hossam.gabbar@uoit.ca [Univ. of Ontario Inst. of Technology, Faculty of Energy Systems and Nuclear Science, Oshawa, Ontario (Canada)

    2013-07-01

    Microgrid is a solution of conventional power grid problem and offer sustainable decentralized power system. Microgrid with modern distributed energy resources (DER) could play an important role to alleviate dependency on the main electricity grid. Distributed energy resource comprises wind turbine, solar photovoltaic, diesel generator, gas engine, micro turbine, fuel cells, etc.Due to the gap between typical loads and supply within microgrid, larger scale energy generation could provide a possible solution to balance power demand and supply. Feasibility study of Small Nuclear Power Plant, such as Small Modular reactor (SMR), within microgrids could be achieved via different cases. To achieve the target, a comprehensive feasibility study is conducted on microgrid with SMR through electricity generation profiles, geographical and environmental assessment, as well as cost analysis using simulation practices and data analysis.Also potency of SMRs is analyzed. Parameters and Key Performance Indicators (KPIs) could be analyzed to achieve feasible solution of microgrids with small modular reactor (SMR) to improve the overall microgrid performance.The study shows that SMR could be a feasible solution if microgrid parameters are selected properly. (author)

  18. Aging study of boiling water reactor high pressure injection systems

    Energy Technology Data Exchange (ETDEWEB)

    Conley, D.A.; Edson, J.L.; Fineman, C.F. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)

    1995-03-01

    The purpose of high pressure injection systems is to maintain an adequate coolant level in reactor pressure vessels, so that the fuel cladding temperature does not exceed 1,200{degrees}C (2,200{degrees}F), and to permit plant shutdown during a variety of design basis loss-of-coolant accidents. This report presents the results of a study on aging performed for high pressure injection systems of boiling water reactor plants in the United States. The purpose of the study was to identify and evaluate the effects of aging and the effectiveness of testing and maintenance in detecting and mitigating aging degradation. Guidelines from the United States Nuclear Regulatory Commission`s Nuclear Plant Aging Research Program were used in performing the aging study. Review and analysis of the failures reported in databases such as Nuclear Power Experience, Licensee Event Reports, and the Nuclear Plant Reliability Data System, along with plant-specific maintenance records databases, are included in this report to provide the information required to identify aging stressors, failure modes, and failure causes. Several probabilistic risk assessments were reviewed to identify risk-significant components in high pressure injection systems. Testing, maintenance, specific safety issues, and codes and standards are also discussed.

  19. Parametric study on effect of break size during LOCA on thermal hydraulic conditions in an indian pressurized heavy water reactor (220 MWe)

    Energy Technology Data Exchange (ETDEWEB)

    Rao, G.S.; Gupta, S.K.; Raj, V.V. [Bhabha Atomic Research Centre, Mumbai (India)

    1999-07-01

    Loss Of Coolant Accident (LOCA) in a Pressurized Heavy Water Reactor (PHWR) leads to coolant expulsion in a primary heat transport system resulting in depressurization and possible core voiding. This results in deterioration of cooling conditions in reactor channels and increase in power before reactor shutdown, leading to higher fuel temperatures. Coolant expulsion rates during LOCA are dictated by critical flow conditions governed by initial plant conditions prior to the accident, break geometry, location of break, etc. In addition the PHWRs have positive void-coefficient of reactivity for coolant resulting in reactor power rise in earlier part of LOCA, when the stored heat of the fuel has yet not been removed. If, in addition, heat transfer to the coolant drops sharply very high fuel surface temperatures are expected. The paper describes analyses carried out for three different break sizes. (author)

  20. Dynamic Response Testing in an Electrically Heated Reactor Test Facility

    Science.gov (United States)

    Bragg-Sitton, Shannon M.; Morton, T. J.

    2006-01-01

    Non-nuclear testing can be a valuable tool in development of a space nuclear power or propulsion system. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and full nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics, and assess potential design improvements at a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE-100a heat pipe cooled, electrically heated reactor and heat exchanger hardware, utilizing a one-group solution to the point kinetics equations to simulate the expected neutronic response of the system (Bragg-Sitton, 2005). The current paper applies the same testing methodology to a direct drive gas cooled reactor system, demonstrating the applicability of the testing methodology to any reactor type and demonstrating the variation in system response characteristics in different reactor concepts. In each testing application, core power transients were controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. Although both system designs utilize a fast spectrum reactor, the method of cooling the reactor differs significantly, leading to a variable system response that can be demonstrated and assessed in a non-nuclear test facility.

  1. SOLASE conceptual laser fusion reactor study

    International Nuclear Information System (INIS)

    A conceptual laser fusion reactor for electric power, SOLASE, has been designed. The SOLASE design utilizes a 1 MJ, 6.7% efficient laser to implode 20 fusion targets per second. The target gain is 150 and produces a net electrical power of 1000 MW. The reactor cavity is spherical with a 6 m radius. The first wall is graphite and has a neutron wall loading of 5 MW/m2. It is protected from the target debris by low pressure xenon gas that is introduced into the cavity. The blanket structure is a honeycombed graphite composite. The tritium breeding and heat transport medium is Li2O in the form of pellets that flow through the blanket. The tritium breeding ration is 1.34. Temperature decoupling of the graphite structure and the Li2O coolant enables the structure to operate at temperatures that minimize radiation damage effects. The graphite blanket is replaced every year but exhibits low levels of radioactivity so that limited hands on maintenance is possible two weeks after shutdown, thus facilitating rapid replacement

  2. Fusion reactor design studies: standard accounts for cost estimates

    International Nuclear Information System (INIS)

    The fusion reactor design studies--standard accounts for cost estimates provides a common format from which to assess the economic character of magnetically confined fusion reactor design concepts. The format will aid designers in the preparation of design concept costs estimates and also provide policymakers with a tool to assist in appraising which design concept may be economically promising. The format sets forth a categorization and accounting procedure to be used when estimating fusion reactor busbar energy cost that can be easily and consistently applied. Reasons for developing the procedure, explanations of the procedure, justifications for assumptions made in the procedure, and the applicability of the procedure are described in this document. Adherence to the format when evaluating prospective fusion reactor design concepts will result in the identification of the more promising design concepts thus enabling the fusion power alternatives with better economic potential to be quickly and efficiently developed

  3. Study of power peak migration due to insertion of control bars in a PWR reactor

    International Nuclear Information System (INIS)

    This paper aims to present a study on the power distribution behavior in a PWR reactor, considering the intensity and the migration of power peaks as is the insertion of control rods in the core banks. For this, the study of the diffusion of neutrons in the reactor was adopted by computer simulation that uses the finite difference method for numerically solving the neutron diffusion equation to two energy groups in steady state and in symmetry of a fourth quarter core. We decided to add the EPRI-9R 3D benchmark thermal-hydraulic parameters of a typical power PWR. With a new configuration for the reactor, the positions of the control rods banks were also modified. Due to the new positioning of these banks in the reactor, there was intense power gradients, favoring the occurrence of critical situations and logically unconventional for operation of a nuclear reactor. However, these facts have led interesting times for the study on the power distribution behavior in the reactor, showing axial migration of power peaks and mainly the effect of the geometry of the core on the latter. Based on the distribution of power was evident the increase of the power in elements located in the central region of the reactor core and, concomitantly, the reduction in elements of its periphery. Of course, the behavior exhibited by the simulated reactor is not in agreement with that expected in an actual reactor, where the insertion of control rods banks should lead to reduced power throughout the core as evenly as possible, avoiding sharp power peaks, standardizing the burning fuel, controlling reactivity deviations and acting in reactor shutdown

  4. The reformation of liquid hydrocarbons in an aqueous discharge reactor

    KAUST Repository

    Zhang, Xuming

    2015-04-21

    We present an aqueous discharge reactor for the reformation of liquid hydrocarbons. To increase a dielectric constant of a liquid medium, we added distilled water to iso-octane and n-dodecane. As expected, we found decreased discharge onset voltage and increased discharge power with increased water content. Results using optical emission spectroscopy identified OH radicals and O atoms as the predominant oxidative reactive species with the addition of water. Enriched CH radicals were also visualized, evidencing the existence of cascade carbon-carbon cleavage and dehydrogenation processes in the aqueous discharge. The gaseous product consisted primarily of hydrogen, carbon monoxide, and unsaturated hydrocarbons. The composition of the product was readily adjustable by varying the volume of water added, which demonstrated a significant difference in composition with respect to the tested liquid hydrocarbon. In this study, we found no presence of CO2 emissions or the contamination of the reactor by solid carbon deposition. These findings offer a new approach to the reforming processes of liquid hydrocarbons and provide a novel concept for the design of a practical and compact plasma reformer. © 2015 IOP Publishing Ltd.

  5. Review of light--water reactor safety studies. Volume 3 of health and safety impacts of nuclear, geothermal, and fossil-fuel electric generation in California

    International Nuclear Information System (INIS)

    This report summarizes and compares important studies of light-water nuclear reactor safety, emphasizing the Nuclear Regulatory Commission's Reactor Safety Study, work on risk assessment funded by the Electric Power Research Institute, and the Report of the American Physical Society study group on light-water reactor safety. These reports treat risk assessment for nuclear power plants and provide an introduction to the basic issues in reactor safety and the needs of the reactor safety research program. Earlier studies are treated more briefly. The report includes comments on the Reactor Safety Study. The manner in which these studies may be used and alterations which would increase their utility are discussed

  6. Design study of electrostatically plugged cusp fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dolan, T.J.

    1976-11-01

    This study concentrates on the following aspects of an electrostatically plugged cusp reactor that will be different from other fusion reactor designs: the coil geometry and structural supports, high voltage electrodes, plasma parameters, power balance, and operating cycle. Assuming the electron density distribution in the anodes to have a characteristic width of two electron Larmor radii, which is consistent with present experimental results, the theory predicts that a device with a magnetic field strength, B = 8 T sustained solely by electron beam injection at 300 kV will have a power gain ratio, Q, of about 5. A toroidal multipole cusp configuration with six cusps was selected for the present design, based on a study of the ratio of plasma volume to coil volume. Coil forces are sustained by cryogenic trusses between like coils, fiberglass compression columns, and room temperature hoops. Radiation collimators in front of the high voltage electrodes greatly reduce the radiation impinging on the cathodes, helping to avoid breakdown and to prolong insulator life. The operating cycle consists of a startup period of about 20 s, followed by a fusion burn period lasting about 200 s (limited by impurity buildup) and a 20-s flushing period.

  7. Nuclear reactor system study for NASA/JPL. Final report

    International Nuclear Information System (INIS)

    Reactor shielding and safety studies and heat pipe development work undertaken for the Jet Propulsion Laboratory during the period March 1, 1981 to October 30, 1981 are described. Monte Carlo calculations of gamma and neutron shield configurations show that substantial weight penalties are incurred if exposures at 25 m to neutrons and gammas must be limited to 1012 nvt and 106 rad, instead of the 1013 nvt and 107 rad values used earlier. For a 1.6 MW/sub t/ reactor, the required shield weight increases from 400 to 815 kg. Water immersion criticality calculations have been extended to study the effect of water in fuel void spaces as well as in the core heat pipes. These show that the insertion into the core of eight blades of B4C with a mass totaling 2.5 kg will guarantee subcriticality. The design, fabrication procedure, and testing of a 4-m-long molybdenum/lithium heat pipe are described. It appears that an excess of oxygen in the wick prevented the attainment of expected performance capability

  8. Operators of nuclear reactors. A study about their professional identity

    International Nuclear Information System (INIS)

    This research studies the process of identity formation/transformation in a group of workers at a Research Institute, which deals with nuclear energy for peaceful uses. It had as a referential the work developed by Antonio da Costa Ciampa (1995). This approach conceives identity as metamorphosis and it is empirically expressed by characters. The research was candied out at the IPEN-CNEN/SP with the oldest reactor operators at work, with open interviews mainly. At the end of the analysis, it was observed that the team, in spite of its grievance and complaints, is still united and operates the reactor with great responsibility. This fact can be proved since no accidents have happened in the last forty years. Nevertheless, the group misses the community recognition for the importance of its work. This was interpreted as the search for an emancipatory feeling towards the development of a collective identity. The lack of this feeling tends to produce an identity with a mere instrumental rationality, suggesting that the organization identity policy must be reviewed. (author)

  9. Pressurized-water reactor internals aging degradation study. Phase 1

    Energy Technology Data Exchange (ETDEWEB)

    Luk, K.H. [Oak Ridge National Lab., TN (United States)

    1993-09-01

    This report documents the results of a Phase I study on the effects of aging degradations on pr internals. Primary stressers for internals an generated by the primary coolant flow in the they include unsteady hydrodynamic forces and pump-generated pressure pulsations. Other stressors are applied loads, manufacturing processes, impurities in the coolant and exposures to fast neutron fluxes. A survey of reported aging-related failure information indicates that fatigue, stress corrosion cracking (SCC) and mechanical wear are the three major aging-related degradation mechanisms for PWR internals. Significant reported failures include thermal shield flow-induced vibration problems, SCC in guide tube support pins and core support structure bolts, fatigue-induced core baffle water-jet impingement problems and excess wear in flux thimbles. Many of the reported problems have been resolved by accepted engineering practices. Uncertainties remain in the assessment of long-term neutron irradiation effects and environmental factors in high-cycle fatigue failures. Reactor internals are examined by visual inspections and the technique is access limited. Improved inspection methods, especially one with an early failure detection capability, can enhance the safety and efficiency of reactor operations.

  10. A programming language study and an implementation of a simulation system with formal derivation. Application to nuclear reactors, control systems and elecronic networks

    International Nuclear Information System (INIS)

    Physical systems simulation requires a lot of information about the controlled process, and the mathematical approach must be appropriate. On the other hand, the parameters describing most systems components are nonlinear and time dependent. Moreover the differential equations describing them are 'stiff' equations of high order. The scope of the study is the description of the NEPTUNIX language and the differential equations. Most of the algorithms used, and the programs implementing these algorithms, are dealt with. Examples of nuclear reactors and mechanical processes simulation are investigated. NEPTUNIX handles a given mathematical description of a continuous system such as: f (x, x(.), t) = 0. Even more, symbolic derivation is performed automatically in order to compute the jacobian associated with the system, requisite for the numerical integration. So, for large systems the manual method for computer the jacobian and the classical method of differentiation are avoided, the former being tiresome and consuming of human time and the latter being costly in run time. The jacobian evaluated in this way is dealt with, by the approach of sparse matrices. Every element of this matrix is assigned a type attribute to improve time execution. Moreover, this is done only once, for a physical system which is described by a mathematical model which topology is invariant. The results of this process are sayed on a suitable device ready for performing repeated simulations

  11. ANSL-V: ENDF/B-V based multigroup cross-section libraries for Advanced Neutron Source (ANS) reactor studies. Supplement 1

    International Nuclear Information System (INIS)

    The original ANSL-V cross-section libraries (ORNL-6618) were developed over a period of several years for the physics analysis of the ANS reactor, with little thought toward including the materials commonly needed for shielding applications. Materials commonly used for shielding applications include calcium barium, sulfur, phosphorous, and bismuth. These materials, as well as 6Li, 7Li, and the naturally occurring isotopes of hafnium, have been added to the ANSL-V libraries. The gamma-ray production and gamma-ray interaction cross sections were completely regenerated for the ANSL-V 99n/44g library which did not exist previously. The MALOCS module was used to collapse the 99n/44g coupled library to the 39n/44g broad- group library. COMET was used to renormalize the two-dimensional (2- D) neutron matrix sums to agree with the one-dimensional (1-D) averaged values. The FRESH module was used to adjust the thermal scattering matrices on the 99n/44g and 39n/44g ANSL-V libraries. PERFUME was used to correct the original XLACS Legendre polynomial fits to produce acceptable distributions. The final ANSL-V 99n/44g and 39n/44g cross-section libraries were both checked by running RADE. The AIM module was used to convert the master cross-section libraries from binary coded decimal to binary format (or vice versa)

  12. ANSL-V: ENDF/B-V based multigroup cross-section libraries for Advanced Neutron Source (ANS) reactor studies. Supplement 1

    Energy Technology Data Exchange (ETDEWEB)

    Wright, R.Q.; Renier, J.P.; Bucholz, J.A.

    1995-08-01

    The original ANSL-V cross-section libraries (ORNL-6618) were developed over a period of several years for the physics analysis of the ANS reactor, with little thought toward including the materials commonly needed for shielding applications. Materials commonly used for shielding applications include calcium barium, sulfur, phosphorous, and bismuth. These materials, as well as {sup 6}Li, {sup 7}Li, and the naturally occurring isotopes of hafnium, have been added to the ANSL-V libraries. The gamma-ray production and gamma-ray interaction cross sections were completely regenerated for the ANSL-V 99n/44g library which did not exist previously. The MALOCS module was used to collapse the 99n/44g coupled library to the 39n/44g broad- group library. COMET was used to renormalize the two-dimensional (2- D) neutron matrix sums to agree with the one-dimensional (1-D) averaged values. The FRESH module was used to adjust the thermal scattering matrices on the 99n/44g and 39n/44g ANSL-V libraries. PERFUME was used to correct the original XLACS Legendre polynomial fits to produce acceptable distributions. The final ANSL-V 99n/44g and 39n/44g cross-section libraries were both checked by running RADE. The AIM module was used to convert the master cross-section libraries from binary coded decimal to binary format (or vice versa).

  13. Study of the characteristics of methanol synthesis in a recirculation slurry reactor - a novel three-phase synthesis reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hu, L.H.; Wang, X.J.; Yu, G.S.; Wang, F.C.; Yu, Z.H. [Institute of Clean Coal Technology, East China University of Science and Technology, Shanghai 200237 (China)

    2008-01-15

    A process feasibility analysis on the liquid phase methanol synthesis (LPMeOH trademark) process was performed in a recirculation slurry reactor (RSR). In the three-phase RSR system, a fine catalyst is slurried in the paraffin and this catalyst slurry is continuously recirculated through the nozzle from the slurry sector to the entrained sector by a pump. The syngas is fed concurrently with the downward flow of slurry to form the methanol product. A laboratory scale mini-pilot plant version of a recirculation slurry reactor system was successfully designed and built to carry out process engineering research, and in addition, an identical cold model was built to measure the mass transfer coefficient in the recirculation slurry reactor. The effects of operating conditions, including temperature, pressure, gas flow rate and catalyst slurry recirculation flow rate on the productivity of methanol were studied. This experimental data helps the scale-up and commercialization of the methanol synthesis process in recirculation slurry reactors. (Abstract Copyright [2008], Wiley Periodicals, Inc.)

  14. Studies of a larger fuel bundle for the ABWR improved evolutionary reactor

    International Nuclear Information System (INIS)

    Studies for an Improved Evolutionary Reactor (IER) based on the Advanced Boiling Water Reactor (ABWR) were initiated in 1990. The author summarizes the current status of the core and fuel design. A core and fuel design based on a BWR K-lattice fuel bundle with a pitch larger than the conventional BWR fuel bundle pitch is under investigation. The core and fuel design has potential for improved core design flexibility and improved reactor transient response. Furthermore, the large fuel bundle, coupled with a functional control rod layout, can achieve improvement of operation and maintenance, as well as improvement of overall plant economy

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-02-01

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

  16. Radiation Hydrodynamic Parameter Study of Inertial Fusion Energy Reactor Chambers

    Science.gov (United States)

    Sacks, Ryan; Moses, Gregory

    2014-10-01

    Inertial fusion energy reactors present great promise for the future as they are capable of providing baseline power with no carbon footprint. Simulation work regarding the chamber response and first wall insult is performed with the 1-D radiation hydrodynamics code BUCKY. Simulation with differing chamber parameters are implemented to study the effect of gas fill, gas mixtures and chamber radii. Xenon and argon gases are of particular interest as shielding for the first wall due to their high opacity values and ready availability. Mixing of the two gases is an attempt to engineer a gas cocktail to provide the maximum amount of shielding with the least amount of cost. A parameter study of different chamber radii shows a consistent relationship with that of first wall temperature (~1/r2) and overpressure (~1/r3). This work is performed under collaboration with Lawrence Livermore National Laboratory.

  17. Multivariable robust control of an integrated nuclear power reactor

    OpenAIRE

    Etchepareborda A.; Flury C.A.

    2002-01-01

    The design of the main control system of the CAREM nuclear power plant is presented. This plant is an inherently safe low-power nuclear reactor with natural convection on the primary coolant circuit and is self-pressurized with a steam dome on the top of the pressure vessel (PV). It is an integrated reactor as the whole primary coolant circuit is within the PV. The primary circuit transports the heat to the secondary circuit through once-through steam generators (SG). There is a feedwater val...

  18. Study on hydrodynamically induced dryout and post dryout important to heavy water reactors

    International Nuclear Information System (INIS)

    Recently, the safety of low pressure liquid cooled nuclear reactors has become a very important issue with reference to the operation of the heavy water reactors at Savannah River Plant. Under accident conditions such as loss-of-flow or loss-of-coolant, these reactors typically encounter unstable two-phase flow which may lead to the occurrence of dryout and subsequent fuel failure. An analytical study using the one-dimensional drift flux model was carried out to investigate the two-phase flow instability for Westinghouse Savannah River Site reactor. The analysis indicates that the first and higher order instabilities exist in the possible transient operational conditions. The instabilities are encountered at higher heat fluxes or lower flow rates. The subcooling has a stabilizing effect except at very low subcooling. An experimental loop has been designed and constructed. A study was conducted on the CHF induced by various flow instabilities. Details of this test loop are presented

  19. Feasibility study on the thorium fueled boiling water breeder reactor

    International Nuclear Information System (INIS)

    The feasibility of (Th,U)O 2 fueled, boiling water breeder reactor based on conventional BWR technology has been studied. In order to determine the potential use of water cooled thorium reactor as a competitive breeder, this study evaluated criticality, breeding and void reactivity coefficient in response to changes made in MFR and fissile enrichments. The result of the study shows that while using light water as moderator, low moderator to fuel volume ratio (MFR=0.5), it was possible to breed fissile fuel in negative void reactivity condition. However the burnup value was lower than the value of the current LWR. On the other hand, heavy water cooled reactor shows relatively wider feasible breeding region, which lead into possibility of designing a core having better neutronic and economic performance than light water with negative void reactivity coefficient. (authors)

  20. Design review and hazop studies for stable salt reactor

    International Nuclear Information System (INIS)

    Atkins has been assisting Moltex Energy in carrying out a review of their Stable Salt Reactor (SSR) (formerly the Simple Molten Salt Reactor (SMSR)), which they believe has many benefits over other nuclear reactor systems. Their current design is for a 2.5 GW thermal output reactor, supplying a superheated steam turbine that could provide an electrical output of 1GW. Our first task was to carry out an assessment of the SSR concept against relevant UK Regulatory Requirements, using data and information that was provided by Moltex Energy. Moltex Energy updated their design using the output from this review. The next activity was to carry out a Hazard and Operability (HAZOP) review to identify issues that may not have been apparent in the original design and the UK Regulatory Requirements assessment. For the HAZOP, we brought together a Process Engineer, Fuel Route and Mechanical Handling Specialist, Safety Case Engineer, Reactor Chemist, Decommissioning and Waste Management Specialist. The HAZOP review gave rise to the generation of a list of key Structures, Systems and Components (SSCs) that would be necessary in a fully designed reactor system, which were discussed and described in a report. This report provided the scope and assumptions that were used as the basis for the costing estimate, using expertise provided in-house by Faithful and Gould. Our paper will discuss the processes used in more detail, identify how these processes increased the knowledge and design concepts to ascertain the SSCs in more detail. Discuss the costing approach and the use of a three-point capital cost estimate, modelled using Monte Carlo simulation, to provide a cost/uncertainty distribution profile. Discussion on how Atkins can assist with the Indian reactor programme using a similar approach detailed above

  1. An evolutionary approach to advanced water cooled reactors

    International Nuclear Information System (INIS)

    Based on the result of the Feasibility Study undertaken since 1991, Indonesia may enter in the new nuclear era by introduction of several Nuclear Power Plants in our energy supply system. Requirements for the future NPP's are developed in two step approach. First step is for the immediate future that is the next 50 years where the system will be dominated by A-LWR's/A-PHWR's and the second step is for the time period beyond 50 years in which new reactor systems may start to dominate. The integral reactor concept provides a revolutionary improvements in terms of conceptual and safety. However, it creates a new set of complex machinery and operational problems of its own. The paper concerns with a brief description of nuclear technology status in Indonesia and a qualitative assessment of integral reactor concept. (author)

  2. Nuclear reactor decommissioning: an analysis of the regulatory environments

    International Nuclear Information System (INIS)

    The purpose of this study is to highlight some of the current and likely regulations that will significantly affect the costs, technical alternatives and financing schemes for reactor decommissioning encountered by electric utilities and their customers. The paper includes a general review of the decommissioning literature, as well as information on specific regulations at the federal, state, and utility levels. Available estimated costs for the decommissioning of individual reactors are also presented. Finally, classification of the specific policies into common trends and practices among the various regulatory bodies is used to examine more general regulatory environments and their potential financial implications

  3. Computational Fluid Dynamics Study of Channel Geometric Effect for Fischer-Tropsch Microchannel Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Na, Jonggeol; Jung, Ikhwan; Kshetrimayum, Krishnadash S.; Park, Seongho; Park, Chansaem; Han, Chonghun [Seoul National University, Seoul (Korea, Republic of)

    2014-12-15

    Driven by both environmental and economic reasons, the development of small to medium scale GTL(gas-to-liquid) process for offshore applications and for utilizing other stranded or associated gas has recently been studied increasingly. Microchannel GTL reactors have been preferred over the conventional GTL reactors for such applications, due to its compactness, and additional advantages of small heat and mass transfer distance desired for high heat transfer performance and reactor conversion. In this work, multi-microchannel reactor was simulated by using commercial CFD code, ANSYS FLUENT, to study the geometric effect of the microchannels on the heat transfer phenomena. A heat generation curve was first calculated by modeling a Fischer-Tropsch reaction in a single-microchannel reactor model using Matlab-ASPEN integration platform. The calculated heat generation curve was implemented to the CFD model. Four design variables based on the microchannel geometry namely coolant channel width, coolant channel height, coolant channel to process channel distance, and coolant channel to coolant channel distance, were selected for calculating three dependent variables namely, heat flux, maximum temperature of coolant channel, and maximum temperature of process channel. The simulation results were visualized to understand the effects of the design variables on the dependent variables. Heat flux and maximum temperature of cooling channel and process channel were found to be increasing when coolant channel width and height were decreased. Coolant channel to process channel distance was found to have no effect on the heat transfer phenomena. Finally, total heat flux was found to be increasing and maximum coolant channel temperature to be decreasing when coolant channel to coolant channel distance was decreased. Using the qualitative trend revealed from the present study, an appropriate process channel and coolant channel geometry along with the distance between the adjacent

  4. Experimental studies of irradiated and hydrogen implantation damaged reactor steels

    Science.gov (United States)

    Slugeň, Vladimír; Pecko, Stanislav; Sojak, Stanislav

    2016-01-01

    Radiation degradation of nuclear materials can be experimentally simulated via ion implantation. In our case, German reactor pressure vessel (RPV) steels were studied by positron annihilation lifetime spectroscopy (PALS). This unique non-destructive method can be effectively applied for the evaluation of microstructural changes and for the analysis of degradation of reactor steels due to neutron irradiation and proton implantation. Studied specimens of German reactor pressure vessel steels are originally from CARINA/CARISMA program. Eight specimens were measured in as-received state and two specimens were irradiated by neutrons in German experimental reactor VAK (Versuchsatomkraftwerk Kahl) in the 1980s. One of the specimens which was in as-received and neutron irradiated condition was also used for simulation of neutron damage by hydrogen nuclei implantation. Defects with the size of about 1-2 vacancies with relatively small contribution (with intensity on the level of 20-40 %) were observed in "as-received" steels. A significant increase in the size of the induced defects due to neutron damage was observed in the irradiated specimens resulting in 2-3 vacancies. The size and intensity of defects reached a similar level as in the specimens irradiated in the nuclear reactor due to the implantation of hydrogen ions with energies of 100 keV (up to the depth <500 nm).

  5. Configurational analysis of an EBT reactor in various magnetic geometries

    Energy Technology Data Exchange (ETDEWEB)

    Owen, L.W.; Uckan, N.A.

    1980-01-01

    Optimization of vacuum field particle confinement in an ELMO Bumpy Torus (EBT) reactor has been considered. Several methods of improving the efficient utilization of magnetic fields and the particle confinement characteristics of a reactor have been analyzed. These include the use of (1) magnets with a large mirror ratio, (2) high field Nb/sub 3/Sn or Nb/sub 3/Sn/NbTi hybrid mirror coils, (3) split-wedge mirror coils, (4) aspect ratio enhancement (ARE) coils, and (5) recently developed field symmetrizing (SYM) coils. Of these, particle drift orbits and three-dimensional tensor pressure equilibrium calculations have shown that the ARE and SYM coils used in conjunction with high field magnets offer the most promise of good plasma performance in a smaller size (up to 50%) EBT reactor. The relative merits of each magnetic configuration are discussed, and the design characteristics are given.

  6. An implicit solution framework for reactor fuel performance simulation

    International Nuclear Information System (INIS)

    The simulation of nuclear reactor fuel performance involves complex thermomechanical processes between fuel pellets, made of fissile material, and the protective cladding that surrounds the pellets. An important design goal for a fuel is to maximize the life of the cladding thereby allowing the fuel to remain in the reactor for a longer period of time to achieve higher degrees of burnup. This presentation examines various mathematical and computational issues that impact the modeling of the thermomechanical response of reactor fuel, and are thus important to the development of INL's fuel performance analysis code, BISON. The code employs advanced methods for solving coupled partial differential equation systems that describe multidimensional fuel thermomechanics, heat generation, and transport within the fuel

  7. Entrained Flow Reactor Study of KCl Capture by Solid Additives

    DEFF Research Database (Denmark)

    Wang, Guoliang; Jensen, Peter Arendt; Wu, Hao;

    been proved to be very promising additives and havereceived extensive studies during the past decades. However, mostprevious studies were carried out in fixed-bed reactors where the reaction conditions are obviously different from that in suspension fired boilers.Detailed knowledge on the reaction...

  8. An analysis of reactor transient response for boiling water reactor ATWS events

    International Nuclear Information System (INIS)

    Numerical simulations of BWR (boiling water reactor) dynamic response under ATWS (anticipated transient without scram) conditions are presented for the case where the reactor is operated at natural circulation conditions. In non-isolation events, reactor stability is strongly influenced by the degree of core inlet subcooling. At normal water level and pressure, instabilities develop if core-inlet subcooling exceeds a critical value of ∼ 21 Btu/Ibm. A sensitivity study with regard to the steam separator pressure-loss coefficient, however, indicates that system stability is strongly dependent on the magnitude of this parameter which suggests a significant degree of uncertainty in the results. Under isolation conditions at rated pressure, stability is significantly enhanced by rapid pressure fluctuations generated through cycling of safety/relief valves. Large-amplitude instabilities develop, however, in depressurization events, and SRV cycling no longer stabilizes the system. In a simulated depressurization to ∼ 500 psia, prompt critical excursions occurred, and oscillation amplitudes reached 1000% of rated power

  9. Pressurized water reactor thorium fuel cycle studies

    International Nuclear Information System (INIS)

    The use of a thorium fuel cycle in a PWR is studied. The thorium has no fissile isotope and a fissile nuclide must be added to the thorium fuel. This nuclide can be uranium 235, plutonium 239 or uranium 233. In this work we have kept the fuel assembly geometry and the control rod system of an usual PWR. Cell calculations showed that the moderation ratio of an usual PWR can be used with uranium 235 and plutonium 239 fuels. But this moderation ratio must be decreased and accordingly the pumping power must be increased in the case of a uranium 233 fuel. The three fuels can be controlled with soluble boron. The power distribution inside an assembly agrees with the safety rules and the worth of the control rods is sufficient. To be interesting the thorium fuels must be recycled. Because the activity and the residual power are higher for a thorium fuel than for a uranium fuel the shielding of the shipping casks and storage pools must be increased. The Uranium 235-Thorium fuel is the best even if it needs expensive enrichment work. With this type of fuel more natural uranium is saved. The thorium fuel would become very interesting if we observe again in the future an increase of the uranium cost

  10. Modelling of NO destruction in a low-pressure reactor by an Ar plasma jet: species abundances in the reactor

    Science.gov (United States)

    Kutasi, Kinga

    2011-03-01

    The destruction of NO molecules by an Ar plasma jet in a low-pressure (0.2 Torr) reactor is investigated by means of a 3D hydrodynamic model. The density distribution of species created through molecular kinetics triggered by the collision of Ar+ with NO is calculated, showing that in the case of the most abundant species a quasi-homogeneous density distribution builds up in a large part of the reactor. The conversion of NO into stable O2 and N2 molecules is followed under different plasma jet conditions and NO gas flows, and the effect of N2 addition on NO destruction is studied. It is shown that in the present system the reproduction of NO molecules on the surface through surface-assisted recombination of N and O atoms becomes impossible due to the fast disappearance of N atoms in the jet's inlet vicinity.

  11. An approach to model reactor core nodalization for deterministic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Salim, Mohd Faiz, E-mail: mohdfaizs@tnb.com.my; Samsudin, Mohd Rafie, E-mail: rafies@tnb.com.my [Nuclear Energy Department, Regulatory Economics & Planning Division, Tenaga Nasional Berhad (Malaysia); Mamat Ibrahim, Mohd Rizal, E-mail: m-rizal@nuclearmalaysia.gov.my [Prototypes & Plant Development Center, Malaysian Nuclear Agency (Malaysia); Roslan, Ridha, E-mail: ridha@aelb.gov.my; Sadri, Abd Aziz [Nuclear Installation Divisions, Atomic Energy Licensing Board (Malaysia); Farid, Mohd Fairus Abd [Reactor Technology Center, Malaysian Nuclear Agency (Malaysia)

    2016-01-22

    Adopting good nodalization strategy is essential to produce an accurate and high quality input model for Deterministic Safety Analysis (DSA) using System Thermal-Hydraulic (SYS-TH) computer code. The purpose of such analysis is to demonstrate the compliance against regulatory requirements and to verify the behavior of the reactor during normal and accident conditions as it was originally designed. Numerous studies in the past have been devoted to the development of the nodalization strategy for small research reactor (e.g. 250kW) up to the bigger research reactor (e.g. 30MW). As such, this paper aims to discuss the state-of-arts thermal hydraulics channel to be employed in the nodalization for RTP-TRIGA Research Reactor specifically for the reactor core. At present, the required thermal-hydraulic parameters for reactor core, such as core geometrical data (length, coolant flow area, hydraulic diameters, and axial power profile) and material properties (including the UZrH{sub 1.6}, stainless steel clad, graphite reflector) have been collected, analyzed and consolidated in the Reference Database of RTP using standardized methodology, mainly derived from the available technical documentations. Based on the available information in the database, assumptions made on the nodalization approach and calculations performed will be discussed and presented. The development and identification of the thermal hydraulics channel for the reactor core will be implemented during the SYS-TH calculation using RELAP5-3D{sup ®} computer code. This activity presented in this paper is part of the development of overall nodalization description for RTP-TRIGA Research Reactor under the IAEA Norwegian Extra-Budgetary Programme (NOKEBP) mentoring project on Expertise Development through the Analysis of Reactor Thermal-Hydraulics for Malaysia, denoted as EARTH-M.

  12. An approach to model reactor core nodalization for deterministic safety analysis

    Science.gov (United States)

    Salim, Mohd Faiz; Samsudin, Mohd Rafie; Mamat @ Ibrahim, Mohd Rizal; Roslan, Ridha; Sadri, Abd Aziz; Farid, Mohd Fairus Abd

    2016-01-01

    Adopting good nodalization strategy is essential to produce an accurate and high quality input model for Deterministic Safety Analysis (DSA) using System Thermal-Hydraulic (SYS-TH) computer code. The purpose of such analysis is to demonstrate the compliance against regulatory requirements and to verify the behavior of the reactor during normal and accident conditions as it was originally designed. Numerous studies in the past have been devoted to the development of the nodalization strategy for small research reactor (e.g. 250kW) up to the bigger research reactor (e.g. 30MW). As such, this paper aims to discuss the state-of-arts thermal hydraulics channel to be employed in the nodalization for RTP-TRIGA Research Reactor specifically for the reactor core. At present, the required thermal-hydraulic parameters for reactor core, such as core geometrical data (length, coolant flow area, hydraulic diameters, and axial power profile) and material properties (including the UZrH1.6, stainless steel clad, graphite reflector) have been collected, analyzed and consolidated in the Reference Database of RTP using standardized methodology, mainly derived from the available technical documentations. Based on the available information in the database, assumptions made on the nodalization approach and calculations performed will be discussed and presented. The development and identification of the thermal hydraulics channel for the reactor core will be implemented during the SYS-TH calculation using RELAP5-3D® computer code. This activity presented in this paper is part of the development of overall nodalization description for RTP-TRIGA Research Reactor under the IAEA Norwegian Extra-Budgetary Programme (NOKEBP) mentoring project on Expertise Development through the Analysis of Reactor Thermal-Hydraulics for Malaysia, denoted as EARTH-M.

  13. An approach to model reactor core nodalization for deterministic safety analysis

    International Nuclear Information System (INIS)

    Adopting good nodalization strategy is essential to produce an accurate and high quality input model for Deterministic Safety Analysis (DSA) using System Thermal-Hydraulic (SYS-TH) computer code. The purpose of such analysis is to demonstrate the compliance against regulatory requirements and to verify the behavior of the reactor during normal and accident conditions as it was originally designed. Numerous studies in the past have been devoted to the development of the nodalization strategy for small research reactor (e.g. 250kW) up to the bigger research reactor (e.g. 30MW). As such, this paper aims to discuss the state-of-arts thermal hydraulics channel to be employed in the nodalization for RTP-TRIGA Research Reactor specifically for the reactor core. At present, the required thermal-hydraulic parameters for reactor core, such as core geometrical data (length, coolant flow area, hydraulic diameters, and axial power profile) and material properties (including the UZrH1.6, stainless steel clad, graphite reflector) have been collected, analyzed and consolidated in the Reference Database of RTP using standardized methodology, mainly derived from the available technical documentations. Based on the available information in the database, assumptions made on the nodalization approach and calculations performed will be discussed and presented. The development and identification of the thermal hydraulics channel for the reactor core will be implemented during the SYS-TH calculation using RELAP5-3D® computer code. This activity presented in this paper is part of the development of overall nodalization description for RTP-TRIGA Research Reactor under the IAEA Norwegian Extra-Budgetary Programme (NOKEBP) mentoring project on Expertise Development through the Analysis of Reactor Thermal-Hydraulics for Malaysia, denoted as EARTH-M

  14. Study on the Export Strategies for Research Reactors

    International Nuclear Information System (INIS)

    Key strategic considerations taken into account should be based on understanding in the forecasts of demand and supply balance as well as the missions of research reactor for customers. For timely arrival at the competition, it may be advantageous to categorize the potential customers into 3 groups, the developed, the developing and the underdeveloped countries in respect of nuclear technology, and to be ready for the group-wise reference designs of the key reactor systems. Customizing the design to specific owner's requirements can advance from one of these reference designs when competition starts. To mobilize this approach effectively, it is useful to establish an integral project and technology management system earlier. This system will function as an important success factor for international research reactor business, because it makes easy to accommodate customer requirements and to achieve the design-to-cost.

  15. Study on the Export Strategies for Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Oh, S. K.; Lee, Y. J.; Ham, T. K.; Hong, S. T.; Kim, J. H. [Ajou University, Suwon (Korea, Republic of)

    2008-12-15

    Key strategic considerations taken into account should be based on understanding in the forecasts of demand and supply balance as well as the missions of research reactor for customers. For timely arrival at the competition, it may be advantageous to categorize the potential customers into 3 groups, the developed, the developing and the underdeveloped countries in respect of nuclear technology, and to be ready for the group-wise reference designs of the key reactor systems. Customizing the design to specific owner's requirements can advance from one of these reference designs when competition starts. To mobilize this approach effectively, it is useful to establish an integral project and technology management system earlier. This system will function as an important success factor for international research reactor business, because it makes easy to accommodate customer requirements and to achieve the design-to-cost.

  16. An analysis of CDTN performance in the reactors technology area

    International Nuclear Information System (INIS)

    The author makes an analysis of CDTN (Centro de Desenvolvimento da Tecnologia Nuclear) performance in the reactors technology area, showing difficulties and failures, but emphasizing the particular competence and capacity acquired in this area, as for example: the capacity in codes and methods are of neutronic calculations and nuclear projects, experimental thermohydraulic program, tests services in components and the others. (C.M.)

  17. An overview of the Daya Bay Reactor Neutrino Experiment

    CERN Document Server

    Cao, Jun

    2016-01-01

    The Daya Bay Reactor Neutrino Experiment discovered an unexpectedly large neutrino oscillation related to the mixing angle $\\theta_{13}$ in 2012. This finding paved the way to the next generation of neutrino oscillation experiments. In this article, we review the history, featured design, and scientific results of Daya Bay. Prospects of the experiment are also described.

  18. An overview of the Daya Bay reactor neutrino experiment

    Science.gov (United States)

    Cao, Jun; Luk, Kam-Biu

    2016-07-01

    The Daya Bay Reactor Neutrino Experiment discovered an unexpectedly large neutrino oscillation related to the mixing angle θ13 in 2012. This finding paved the way to the next generation of neutrino oscillation experiments. In this article, we review the history, featured design, and scientific results of Daya Bay. Prospects of the experiment are also described.

  19. Methane production in an UASB reactor operated under periodic mesophilic-thermophilic conditions.

    Science.gov (United States)

    Bourque, J-S; Guiot, S R; Tartakovsky, B

    2008-08-15

    Methane production was studied in a laboratory-scale 10 L anaerobic upflow sludge bed (UASB) reactor with periodic variations of the reactor temperature. On a daily basis the temperature was varied between 35 and 45 degrees C or 35 and 55 degrees C with a heating period of 6 h. Each temperature increase was accompanied by an increase in methane production and a decrease in the concentration of soluble organic matter in the effluent. In comparison to a reactor operated at 35 degrees C, a net increase in methane production of up to 22% was observed. Batch activity tests demonstrated a tolerance of mesophilic methanogenic populations to short-term, 2-6 h, temperature increases, although activity of acetoclastic methanogens decreased after 6 h exposure to a temperature of 55 degrees C. 16S sequencing of DGGE bands revealed proliferation of temperature-tolerant Methanospirillum hungatii sp. in the reactor.

  20. A Boiling-Potassium Fluoride Reactor for an Artificial-Gravity NEP Vehicle

    Science.gov (United States)

    Sorensen, Kirk; Juhasz, Albert

    2007-01-01

    Several years ago a rotating manned spacecraft employing nuclear-electric propulsion was examined for Mars exploration. The reactor and its power conversion system essentially served as the counter-mass to an inflatable manned module. A solid-core boiling potassium reactor based on the MPRE concept of the 1960s was baselined in that study. This paper proposes the use of a liquid-fluoride reactor, employing direct boiling of potassium in the core, as a means to overcome some of the residual issues with the MPRE reactor concept. Several other improvements to the rotating Mars vehicle are proposed as well, such as Canfield joints to enable the electric engines to track the inertial thrust vector during rotation, and innovative "cold-ion" engine technologies to improve engine performance.

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

    Directory of Open Access Journals (Sweden)

    Curtisha D. Travis

    2013-08-01

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

  2. CFD Simulation of an Anaerobic Membrane BioReactor (AnMBR) to Treat Industrial Wastewater

    OpenAIRE

    Laura C. Zuluaga; Luz N. Naranjo; Jan Svojitka; Thomas Wintgens; Manuel Rodriguez; Nicolas Ratkovich

    2015-01-01

    A Computational Fluid Dynamics (CFD) simulation has been developed for an Anaerobic Membrane BioReactor (AnMBR) to treat industrial wastewater. As the process consists of a side-stream MBR, two separate simulations were created: (i) reactor and (ii) membrane. Different cases were conducted for each one, so the surrounding temperature and the total suspended solids (TSS) concentration were checked. For the reactor, the most important aspects to consider were the dead zones and the mixing, wher...

  3. Research and development studies on the seismic behaviour of the PEC fast reactor

    International Nuclear Information System (INIS)

    As introduction to the meeting, this paper provides an overview on the extensive research and development studies performed by ENEA, in co-operation with ANSALDO and ISMES, in the framework of the seismic verification of the Italian PEC fast reactor. The purpose is also to stress the reasons why a wide-ranging experimental programme and detailed numerical analysis, validated on the test results, have been performed for the PEC reactor building and the main vessel. Thus, after some notes on the high levels of the design earthquakes adopted for PEC and the important features of fast reactors in general and PEC as a specific case (making it particularly sensitive to seismic excitations), the paper presents the studies performed for the reactor-block, the core and the shutdown system, summarizing their main features and showing some of the main results. Furthermore, the non-negligible feed-backs of the seismic studies on the reactor-block design are recalled, and the needs of checking seismic design analysis of the main vessel and the reactor building are explained. The on-site experimental programme and the related numerical analysis concerning the main vessel and the reactor building are also shortly described: however, specific papers will present more details on these studies, and will also stress the usefulness of the on-site tests performed on the reactor building for the optimization of the PEC seismic monitoring system. Finally, the Italian lecture invited to this meeting will provide an overview on the state-of-the-art on on-site testing and seismic monitoring in Italy, stressing the perspective of adopting methodologies similar to those used for PEC, for nuclear power plants in general. (author)

  4. A mathematical model for multiple hydrogeneration reactions in a continuous stirred three phase slurry reactor with an evaporating solvent

    NARCIS (Netherlands)

    Janssen, H.J.; Westerterp, K.R.; Vos, J.

    1992-01-01

    An experimental study of the catalytic hydorgenation of 2,4-dinitrotoluene (DNT) in a mini-installation with a continuously operated stirred three-phase slurry reactor and an evaporating solvent is discussed. Some characteristic properties of the reactor system and the influence of the operating par

  5. Isotopic studies relative to the Oklo natural fission reactors

    International Nuclear Information System (INIS)

    It has been clearly demonstrated that natural fission reactors operated about 2 109 years ago, in rich uranium one deposits of the Oklo mine in the Republique of Gabon. Six reactions zones have been identified in which approximately six tons of 235U were consumed and the same amount of fission products deposited in the ground. These fission products, their filiation isotopes and nuclei formed from neutron captures are precious tracers, which now can be analysed on well localized samples, to obtain informations on the stability in soil of such elements and data on the nuclear parameters and characteristics of the nuclear reactors. The studies which have been developed at Saclay concern several aspects of this phenomenon: the migrations of fission products, the age of the nuclear reaction, the date of the uranium deposit and the temperature of the reaction zones during the operation of the reactors

  6. A study on the decommissioning of research reactor

    International Nuclear Information System (INIS)

    As the result of study on decommissioning, discussion has made and data have been collected about experiences, plannings, and techniques for decommissioning through visit to GA and JAERI. GA supplied our Research Reactor No. 1 and No. 2, and JAERI made a memorial museum after dicommissioning of JRR-1 and is dismentling JPDR now. Also many kinds of documents are collected and arranged such as documents related to TRIGA reactor dicommissioning, 30 kinds of documents including decommissioning plan, technical criteria and related regulatory, and 1,200 kinds of facility description data. (Author)

  7. Dynamic Response Testing in an Electrically Heated Reactor Test Facility

    Science.gov (United States)

    Bragg-Sitton, Shannon M.; Morton, T. J.

    2006-01-01

    Non-nuclear testing can be a valuable tool in the development of a space nuclear power or propulsion system. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and fueled nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics, and assess potential design improvements at a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE-100a heat pipe (HP) cooled, electrically heated reactor and heat exchanger hardware, utilizing a one-group solution to the point kinetics equations to simulate the expected neutronic response of the system. Reactivity feedback calculations were then based on a bulk reactivity feedback coefficient and measured average core temperature. This paper presents preliminary results from similar dynamic testing of a direct drive gas cooled reactor system (DDG), demonstrating the applicability of the testing methodology to any reactor type and demonstrating the variation in system response characteristics in different reactor concepts. Although the HP and DDG designs both utilize a fast spectrum reactor, the method of cooling the reactor differs significantly, leading to a variable system response that can be demonstrated and assessed in a non-nuclear test facility. Planned system upgrades to allow implementation of higher fidelity dynamic testing are also discussed. Proposed DDG

  8. Low temperature oxidation of hydrocarbons using an electrochemical reactor

    DEFF Research Database (Denmark)

    Ippolito, Davide

    high propene conversion at open circuit voltage together high rate enhancement ratio and faradaic efficiency values at low temperatures (300-350 °C). Although some stability problems affected the performance of multiple infiltrated Ce0.9Gd0.1O1.95 on LSM/CGO backbone, the strong activation of LSM upon......This study investigated the use of a ceramic porous electrochemical reactor for the deep oxidation of propene. Two electrode composites, La0.85Sr0.15MnO3±d/Ce0.9Gd0.1O1.95 (LSM/CGO) and La0.85Sr0.15FeMnO3/Ce0.9Gd0.1O1.95 (LSF/CGO), were produced in a 5 single cells stacked configuration and used...... as backbone for the infiltration of materials able to modify the electrochemical and catalytic activity of the reactor. The catalytic activity of the reactor and the effect of polarization on the propene oxidation have been studied by gas analysis with a gas chromatograph (GC), during the reactor polarization...

  9. System Study: Reactor Core Isolation Cooling 1998-2014

    Energy Technology Data Exchange (ETDEWEB)

    Schroeder, John Alton [Idaho National Lab. (INL), Idaho Falls, ID (United States). Risk Assessment and Management Services Dept.

    2015-12-01

    This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2014 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10 year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant trends were identified in the RCIC results.

  10. System Study: Reactor Core Isolation Cooling 1998–2012

    Energy Technology Data Exchange (ETDEWEB)

    T. E. Wierman

    2013-10-01

    This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2012 for selected components were obtained from the Equipment Performance and Information Exchange (EPIX). The unreliability results are trended for the most recent 10 year period while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing trend was identified in the HPCI results. Statistically significant decreasing trends were identified for RCIC start-only and 8-hour trends.

  11. System Study: Reactor Core Isolation Cooling 1998–2013

    Energy Technology Data Exchange (ETDEWEB)

    Schroeder, John Alton [Idaho National Lab. (INL), Idaho Falls, ID (United States). Risk Assessment and Management Services Dept.

    2015-01-31

    This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2013 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10-year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant trends were identified in the RCIC results.

  12. Treatment of domestic wastewater in an up-flow anaerobic sludge blanket reactor followed by moving bed biofilm reactor

    NARCIS (Netherlands)

    Tawfik, A.; El-Gohary, F.; Temmink, B.G.

    2010-01-01

    The performance of a laboratory-scale sewage treatment system composed of an up-flow anaerobic sludge blanket (UASB) reactor and a moving bed biofilm reactor (MBBR) at a temperature of (22-35 A degrees C) was evaluated. The entire treatment system was operated at different hydraulic retention times

  13. Experimental studies of fission properties utilized in reactor design

    International Nuclear Information System (INIS)

    Experimental studies of fission properties utilized in reactor design. A programme of experimental studies of fission parameters useful in reactor design is described including the following: (a) The periods and yields of delayed-neutron groups emitted following the neutron-induced fission of Pu241 are measured. Evidence for systematic isotopic dependence of delayed-neutron yields is presented. An experimental investigation of the relation between the time behaviour of delayed-neutron emission and the energy of the incident neutron inducing fission is described. (b) The cross-section for the inducing, of fission in Am243, Pu242 and Pu241 with neutrons in the energy range 0.030 to 1.8 MeV is measured. Emphasis is placed upon the detailed dependence of the fission cross-section on the incident-neutron energy. The absolute values of the cross-sections are given to a precision of ∼25%. (c) Detailed results of a measurement of the Pu241 fission-neutron spectrum are given, including the spectral shape and average fission-neutron energy. Techniques and methods of measuring prompt-fission-neutron spectra are described. (d) The dependence of #-v# (the average number of neutrons emitted per fission) of U235 on the incident neutron energy is measured from 100 keV to 1.6 MeV. #-v# of U238 and other fissile isotopes is compared to #-v# of U235 (thermal). The relative precision of the measurements is #>approx#1.2%. (author)

  14. Development of an automated core model for nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mosteller, R.D.

    1998-12-31

    This is the final report of a three-year, Laboratory Directed Research and Development (LDRD) project at the Los Alamos National Laboratory (LANL). The objective of this project was to develop an automated package of computer codes that can model the steady-state behavior of nuclear-reactor cores of various designs. As an added benefit, data produced for steady-state analysis also can be used as input to the TRAC transient-analysis code for subsequent safety analysis of the reactor at any point in its operating lifetime. The basic capability to perform steady-state reactor-core analysis already existed in the combination of the HELIOS lattice-physics code and the NESTLE advanced nodal code. In this project, the automated package was completed by (1) obtaining cross-section libraries for HELIOS, (2) validating HELIOS by comparing its predictions to results from critical experiments and from the MCNP Monte Carlo code, (3) validating NESTLE by comparing its predictions to results from numerical benchmarks and to measured data from operating reactors, and (4) developing a linkage code to transform HELIOS output into NESTLE input.

  15. Kartini Research Reactor prospective studies for neutron scattering application

    Energy Technology Data Exchange (ETDEWEB)

    Widarto [Yogyakarta Nuclear Research Center, BATAN (Indonesia)

    1999-10-01

    The Kartini Research Reactor (KRR) is located in Yogyakarta Nuclear Research Center, Yogyakarta - Indonesia. The reactor is operated for 100 kW thermal power used for research, experiments and training of nuclear technology. There are 4 beam ports and 1 column thermal are available at the reactor. Those beam ports have thermal neutron flux around 10{sup 7} n/cm{sup 2}s each other and used for sub critical assembly, neutron radiography studies and Neutron Activation Analysis (NAA). Design of neutron collimator has been done for piercing radial beam port and the calculation result of collimated neutron flux is around 10{sup 9} n/cm{sup 2}s. This paper describes experiment facilities and parameters of the Kartini research reactor, and further more the prospective studies for neutron scattering application. The purpose of this paper is to optimize in utilization of the beam ports facilities and enhance the manpower specialty. The special characteristic of the beam ports and preliminary studies, pre activities regarding with neutron scattering studies for KKR is presented. (author)

  16. Magnetic enzyme reactors for isolation and study of heterogeneous glycoproteins

    Energy Technology Data Exchange (ETDEWEB)

    Korecka, Lucie [Department of Analytical Chemistry, University of Pardubice, Namesti Cs. Legii 565, 532 10 Pardubice (Czech Republic)]. E-mail: lucie.korecka@upce.cz; Jezova, Jana [Department of Analytical Chemistry, University of Pardubice, Namesti Cs. Legii 565, 532 10 Pardubice (Czech Republic); Bilkova, Zuzana [Department of Biological and Biochemical Sciences, University of Pardubice, Namesti Cs. Legii 565, 532 10 Pardubice (Czech Republic); Benes, Milan [Institute of Macromolecular Chemistry, Academy of Sciences of the Czech Republic, Heyrovskeho Namesti 2, 162 06 Prague (Czech Republic); Horak, Daniel [Institute of Macromolecular Chemistry, Academy of Sciences of the Czech Republic, Heyrovskeho Namesti 2, 162 06 Prague (Czech Republic); Hradcova, Olga [Department of Biological and Biochemical Sciences, University of Pardubice, Namesti Cs. Legii 565, 532 10 Pardubice (Czech Republic); Slovakova, Marcela [Department of Biological and Biochemical Sciences, University of Pardubice, Namesti Cs. Legii 565, 532 10 Pardubice (Czech Republic); Laboratoire Physicochimie Curie, UMR 168 CNRS/Institute Curie, Paris Cedex 05 (France); Viovy, Jean-Louis [Laboratoire Physicochimie Curie, UMR 168 CNRS/Institute Curie, Paris Cedex 05 (France)

    2005-05-15

    The newly developed magnetic micro- and nanoparticles with defined hydrophobicity and porosity were used for the preparation of magnetic enzyme reactors. Magnetic particles with immobilized proteolytic enzymes trypsin, chymotrypsin and papain and with enzyme neuraminidase were used to study the structure of heterogeneous glycoproteins. Factors such as the type of carrier, immobilization procedure, operational and storage stability, and experimental conditions were optimized.

  17. Experimental study of neutrino oscillations at a fission reactor

    International Nuclear Information System (INIS)

    The energy spectrum of neutrinos from a fission reactor was studied with the aim of gaining information on neutrino oscillations. The well shielded detector was set up at a fixed position of 8.76 m from the point-like core of the Laue-Langevin reactor in an antineutrino flux of 9.8 x 1011cm-2s-1. The target protons in the reaction antiνsub(e)p → e+n were provided by liquid scintillation counters (total volume of 377l) which also served as positron detectors. The product neutrons moderated in the scintillator were detected by 3He wire chambers. A coincidence signature was required between the prompt positron and the delayed neutron events. The positron energy resolution was 18% FWHM at 0.91 MeV. The signal-to-background ratio was better than one to one between 2 MeV and 6 MeV positron energy. At a counting rate of 1.58 counts per hour, 4890+-180 neutrino induced events were detected. The shape of the measured positron spectrum was analyzed in terms of the parameters Δ2 and sin2 2theta for two-neutrino oscillations. The experimental data are consistent with no oscillations. An upper limit of 0.15 eV2 (90% c.l.) for the mass-squared differences Δ2 of the neutrinos was obtained, assuming maximum mixing of the two neutrino states. The ratio of the measured to the expected integral yield of positrons assuming no oscillations was determined to be ∫Ysub(exp)/∫Ysub(th) = 0.955+-0.035 (statistical)+-0.110 (systematic)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1978-07-15

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

  19. Conceptual Study of Transmutation Reactor Based on LAR Tokamak Fusion Neutron Source

    International Nuclear Information System (INIS)

    A compact tokamak reactor concept as a 14 MeV neutron source is desirable from an economic viewpoint for a fusion-driven transmutation reactor. For the optimal design of a reactor, a radial build of reactor components has to be determined by considering the plasma physics and engineering constraints which interrelate various reactor components. In a transmutation reactor, design of blanket and shield play a key role in determining the size of a reactor: the blanket should produce enough tritium for tritium self-sufficiency, the transmutation rate of waste has to be maximized, and the shield should provide sufficient protection for the superconducting toroidal field (TF) coil. To determine the radial build of the blanket and the shield, not only a radiation transport analysis but also a burn-up calculation were coupled with the system analysis and it allowed the self-consistent determination of the design parameters of a transmutation reactor. For neutronic optimization of the blanket and the shield, the quantities such as the tritium breeding ratio (TBR), nuclear heating, radiation damage to the toroidal field coil have to be calculated and burn-up rates of Li, actinides and fission products have to be calculated. Thus the neutronic analysis need to be coupled in the system analysis. In most of the previous system studies, neutronic calculation and plasma analysis are performed separately, so blanket and shield size was determined independently from the reactor size. In this work, to account for the interrelation of blanket and shield with the other components of a reactor system, we coupled the system analysis with one-dimensional neutronic calculation to determine the reactor parameters in self consistent manner. LAR (Low Aspect Ratio) tokamak plasma has the potential of high β operation with high bootstrap current fractions. In the LAR tokamak reactor, the radial build of TF coil(TFC) and the shield play a key role in determining the size of a reactor since it

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-11-01

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

  1. Physics and Control Assessment of AN 850 Mw(e) - Leu-Candu Reactor.

    Science.gov (United States)

    Barbone, Michelangelo

    The physics and control assessment of an 850 MW(e) Low Enriched Uranium CANDU Pressurized Heavy Water (LEU -CANDU-PHW) reactor constitute the major objective of this thesis. The use of Low Enriched Uranium fuel in the present CANDU nuclear power generating stations is recognized as economically beneficial due to reduced fuelling costs. The LEU fuel cycle is also recognized as a stepping stone to transit from the present CANDU-PHW once-through natural Uranium cycle to advanced cycles such as those based on Plutonium recycle, once-through Th + U-235 cycle, Thorium with Uranium recycle and net U-235 feed, Thorium with Uranium recycle and Plutonium feed. However, although the use of Low Enriched Uranium in the present CANDU-PHW reactor has economic advantages, and it would act as a technical bridge between the present cycle and advanced cycles, technical problems in different areas of reactor physics and fuel management were anticipated. The present thesis research work adresses the areas of reactor physics, fuel management, and control (in particular, the spatial control of large CANDU-PHW reactors). The main conclusions that have been drawn following these studies are as follows: (1) The Low Enriched Uranium Cycle is feasible in a CANDU-PHW reactor of present design and provided that: (a) The enrichment is kept relatively low (that is, about 1% instead of 0.711%); (b) the number of bundles to be replaced at every refuelling operation is about one-half that of the natural Uranium fuel case; (c) The channels are refuelled in the same direction as the coolant. (2) The response of an LEU-CANDU-PHW reactor to reactivity perturbation such as single- and two-channel refuelling operation, shim transient, shutdown-start-up transient with enrichment levels of 0.9% and 1.2% is essentially very similar {provided that certain conditions in (1) are respected} to that of the natural uranium reactor core case without any reactor reoptimization. The general behaviour of the reactor

  2. An atmospheric pressure flow reactor: Gas phase kinetics and mechanism in tropospheric conditions without wall effects

    Science.gov (United States)

    Koontz, Steven L.; Davis, Dennis D.; Hansen, Merrill

    1988-01-01

    A new type of gas phase flow reactor, designed to permit the study of gas phase reactions near 1 atm of pressure, is described. A general solution to the flow/diffusion/reaction equations describing reactor performance under pseudo-first-order kinetic conditions is presented along with a discussion of critical reactor parameters and reactor limitations. The results of numerical simulations of the reactions of ozone with monomethylhydrazine and hydrazine are discussed, and performance data from a prototype flow reactor are presented.

  3. The U.S.-Russian joint studies on using power reactors to disposition surplus weapons plutonium as spent fuel

    International Nuclear Information System (INIS)

    In 1996, the US and the Russian Federation completed an initial joint study of the candidate options for the disposition of surplus weapons plutonium in both countries. The options included long term storage, immobilization of the plutonium in glass or ceramic for geologic disposal, and the conversion of weapons plutonium to spent fuel in power reactors. For the latter option, the US is only considering the use of existing light water reactors (LWRs) with no new reactor construction for plutonium disposition, or the use of Canadian deuterium uranium (CANDU) heavy water reactors. While Russia advocates building new reactors, the cost is high, and the continuing joint study of the Russian options is considering only the use of existing VVER-1000 LWRs in Russia and possibly Ukraine, the existing BN-60O fast neutron reactor at the Beloyarsk Nuclear Power Plant in Russia, or the use of the Canadian CANDU reactors. Six of the seven existing VVER-1000 reactors in Russia and the eleven VVER-1000 reactors in Ukraine are all of recent vintage and can be converted to use partial MOX cores. These existing VVER-1000 reactors are capable of converting almost 300 kg of surplus weapons plutonium to spent fuel each year with minimum nuclear power plant modifications. Higher core loads may be achievable in future years

  4. The U.S.-Russian joint studies on using power reactors to disposition surplus weapon plutonium as spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Chebeskov, A.; Kalashnikov, A. [State Scientific Center, Obninsk (Russian Federation). Inst. of Physics and Power Engineering; Bevard, B.; Moses, D. [Oak Ridge National Lab., TN (United States); Pavlovichev, A. [State Scientific Center, Moscow (Russian Federation). Kurchatov Inst.

    1997-09-01

    In 1996, the US and the Russian Federation completed an initial joint study of the candidate options for the disposition of surplus weapons plutonium in both countries. The options included long term storage, immobilization of the plutonium in glass or ceramic for geologic disposal, and the conversion of weapons plutonium to spent fuel in power reactors. For the latter option, the US is only considering the use of existing light water reactors (LWRs) with no new reactor construction for plutonium disposition, or the use of Canadian deuterium uranium (CANDU) heavy water reactors. While Russia advocates building new reactors, the cost is high, and the continuing joint study of the Russian options is considering only the use of existing VVER-1000 LWRs in Russia and possibly Ukraine, the existing BN-60O fast neutron reactor at the Beloyarsk Nuclear Power Plant in Russia, or the use of the Canadian CANDU reactors. Six of the seven existing VVER-1000 reactors in Russia and the eleven VVER-1000 reactors in Ukraine are all of recent vintage and can be converted to use partial MOX cores. These existing VVER-1000 reactors are capable of converting almost 300 kg of surplus weapons plutonium to spent fuel each year with minimum nuclear power plant modifications. Higher core loads may be achievable in future years.

  5. Safety benefits from CANDU reactor replacement. A case study

    International Nuclear Information System (INIS)

    Both total core replacement and core retubing have been used in the CANDU industry. For future plant refurbishments, based on experience both in new construction and in recent refurbishments, the concept of total core replacement has been revisited. This builds on practices for replacement of other large plant equipment like boilers. The Bruce CANDU reactors, with their local shield tanks built around the Calandria and containment closely located around that Calandria Shield Tank Assembly (CSTA), are believed to be good candidates for core replacement. A structured process was used to design a replacement CSTA suitable for Bruce A use. The work started with a study of opportunities for safety enhancements in the core. This progressed into design studies and related design assist safety analysis on the reactor. A key element of the work involved consideration of how verified features from later CANDU designs, and from our new reactor design work, could be tailored to fit this replacement core. The replacement reactor core brings in structural improvements in both calandria and end shield, and safety improvements like the natural circulation enhancing moderator cooling layout and further optimized reactivity layouts to improve shutdown system performance. Bruce Power are currently studying the business implications of this and retube techniques as part of preparation for future refurbishments. The work explained in this paper is in the context of the safety related changes and the work to choose and quantify them. (author)

  6. Safety benefits from CANDU reactor replacement - a case study

    International Nuclear Information System (INIS)

    Both total core replacement and core retubing have been used in the CANDU industry. For future plant refurbishments, based on experience both in new construction and in recent refurbishments, the concept of total core replacement has been revisited. This builds on practices for replacement of other large plant equipment like boilers. The Bruce CANDU reactors, with their local shield tanks built around the Calandria and containment closely located around that Calandria Shield Tank Assembly (CSTA), are believed to be good candidates for core replacement. A structured process was used to design a replacement CSTA suitable for Bruce A use. The work started with a study of opportunities for safety enhancements in the core. This progressed into design studies and related design assist safety analysis on the reactor. A key element of the work involved consideration of how verified features from later CANDU designs, and from our new reactor design work, could be tailored to fit this replacement core. The replacement reactor core brings in structural improvements in both calandria and end shield, and safety improvements like the natural circulation enhancing moderator cooling layout and further optimized reactivity layouts to improve shutdown system performance. Bruce Power are currently studying the business implications of this and retube techniques as part of preparation for future refurbishments. The work explained in this paper is in the context of the safety related changes and the work to choose and quantify them. (author)

  7. Small and medium reactors I. status and prospects Report by an Expert Group

    International Nuclear Information System (INIS)

    Until recently the thrust of power reactor development has been to take advantage of the economies of scale, which has led to the deployment of reactor of over 1 000MWe. However, there is now considerable interest arising in small reactor types. Small and Medium-sized Reactor (SMRs) are being designed in several countries for three purposes: electric power production, heat generation (both for industrial process heat and space heating) and cogeneration of both heat and electric power. These designs, in some cases, have evolved from larger reactors currently used for power production; in other cases, more radical design changes have been introduced. The Nuclear Development Committee of the NEA believed it was timely to prepare an analysis of the role of these newer reactor concepts in OECD countries, in relation to both electricity and heat production. In the context of current concerns over the fuel sources and technologies that will be available to provide reliable low-cost energy with minimal environmental impact, it is desirable for high-level decision makers in government and industry to have an objective view of the state of development of these reactor concepts, their potential to open up new applications, their likely costs, and steps to be taken before they are deployed commercially. This study, published under the responsibilities of the Secretary General of the OECD, has been prepared by experts nominated by the Agency's Member countries. The report does not necessarily represent the views of those countries nor of other participating organisations. Volume 2 of the report is a technical supplement which contains information on User Requirements as well as descriptions of various reactor projects

  8. The Jules Horowitz Reactor : A new high Performances European MTR (Material Testing Reactor) with modern experimental capacities : Toward an International User Facility

    International Nuclear Information System (INIS)

    The Jules Horowitz Reactor (JHR) is a new Material Testing Reactor (MTR) currently under construction at CEA Cadarache research centre in the south of France. It will be a major Research facility in support to the development and the qualification of materials and fuels under irradiation with sizes and environment conditions relevant for nuclear power plants in order to optimise and demonstrate safe operations of existing power reactors as well as to support future reactor design. It will represent also an important Research Infrastructure for scientific studies dealing with material and fuel behaviour under irradiation. The JHR will contribute also to secure the production of radioisotope for medical application. This is a key public health stake. The construction of JHR which was started in 2007 is on-going. The first operation is planned before the end of this decade.The design of the reactor will provide an essential facility supporting the programs for the nuclear energy for the next 50 years. JHR is designed to provide high neutron flux (enhancing the maximum available today in MTRs), to run highly instrumented experiments to support advanced modelling giving prediction beyond experimental points, and to operate experimental devices giving environment conditions (pressure, temperature, flux, coolant chemistry, ···) relevant for water reactors, for gas cooled thermal or fast reactors, for sodium fast reactors, ···So, the reactor will perform R and D programs for the optimization of the present generation of NPP, support the development of the next generation of NPP (mainly LWR) and also offer irradiation possibilities for future reactors. In parallel to the construction of the reactor, the preparation of an international community around JHR is continuing; this is an important topic as building and gathering a strong international community in support to MTR experiments is a key-issue for the R and D in nuclear energy field. Consequently, CEA is

  9. Summary discussion: An integrated advanced tokamak reactor

    International Nuclear Information System (INIS)

    The tokamak concept improvement workshop addressed a wide range of issues involved in the development of a more attractive tokamak. The agenda for the workshop progressed from a general discussion of the long-range energy context (with the objective being the identification of a set of criteria and ''figures of merit'' for measuring the attractiveness of a tokamak concept) to particular opportunities for the improvement of the tokamak concept. The discussions concluded with a compilation of research program elements leading to an improved tokamak concept

  10. System and method for air temperature control in an oxygen transport membrane based reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, Sean M

    2016-09-27

    A system and method for air temperature control in an oxygen transport membrane based reactor is provided. The system and method involves introducing a specific quantity of cooling air or trim air in between stages in a multistage oxygen transport membrane based reactor or furnace to maintain generally consistent surface temperatures of the oxygen transport membrane elements and associated reactors. The associated reactors may include reforming reactors, boilers or process gas heaters.

  11. Flibe Use in Fusion Reactors - An Initial Safety Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Cadwallader, Lee Charles; Longhurst, Glen Reed

    1999-04-01

    This report is an initial effort to identify and evaluate safety issues associated with the use of Flibe (LiF-BeF2) as a molten salt coolant for nuclear fusion power plant applications. Flibe experience in the Molten Salt Reactor Experiment is briefly reviewed. Safety issues identified include chemical toxicity, radiological issues resulting from neutron activation, and the operational concerns of handling a high temperature coolant. Beryllium compounds and fluorine pose be toxicological concerns. Some controls to protect workers are discussed. Since Flibe has been handled safely in other applications, its hazards appear to be manageable. Some safety issues that require further study are pointed out. Flibe salt interaction with strong magnetic fields should be investigated. Evolution of Flibe constituents and activation products at high temperature (i.e., will Fluorine release as a gas or remain in the molten salt) is an issue. Aerosol and tritium release from a Flibe spill requires study, as does neutronics analysis to characterize radiological doses. Tritium migration from Flibe into the cooling system is also a safety concern. Investigation of these issues will help determine the extent to which Flibe shows promise as a fusion power plant coolant or plasma-facing material.

  12. Flibe use in fusion reactors -- An initial safety assessment

    Energy Technology Data Exchange (ETDEWEB)

    Cadwallader, L.C.; Longhurst, G.R.

    1999-03-01

    This report is an initial effort to identify and evaluate safety issues associated with the use of Flibe (LiF-BeF{sub 2}) as a molten salt coolant for nuclear fusion power plant applications. Flibe experience in the Molten Salt Reactor Experiment is briefly reviewed. Safety issues identified include chemical toxicity, radiological issues resulting from neutron activation, and the operational concerns of handling a high temperature coolant. Beryllium compounds and fluorine pose be toxicological concerns. Some controls to protect workers are discussed. Since Flibe has been handled safely in other applications, its hazards appear to be manageable. Some safety issues that require further study are pointed out. Flibe salt interaction with strong magnetic fields should be investigated. Evolution of Flibe constituents and activation products at high temperature (i.e., will Fluorine release as a gas or remain in the molten salt) is an issue. Aerosol and tritium release from a Flibe spill requires study, as does neutronics analysis to characterize radiological doses. Tritium migration from Flibe into the cooling system is also a safety concern. Investigation of these issues will help determine the extent to which Flibe shows promise as a fusion power plant coolant or plasma-facing material.

  13. Flibe use in fusion reactors: An initial safety assessment

    International Nuclear Information System (INIS)

    This report is an initial effort to identify and evaluate safety issues associated with the use of Flibe (LiF-BeF2) as a molten salt coolant for nuclear fusion power plant applications. Flibe experience in the Molten Salt Reactor Experiment is briefly reviewed. Safety issues identified include chemical toxicity, radiological issues resulting from neutron activation, and the operational concerns of handling a high temperature coolant. Beryllium compounds and fluorine pose be toxicological concerns. Some controls to protect workers are discussed. Since Flibe has been handled safely in other applications, its hazards appear to be manageable. Some safety issues that require further study are pointed out. Flibe salt interaction with strong magnetic fields should be investigated. Evolution of Flibe constituents and activation products at high temperature (i.e., will Fluorine release as a gas or remain in the molten salt) is an issue. Aerosol and tritium release from a Flibe spill requires study, as does neutronics analysis to characterize radiological doses. Tritium migration from Flibe into the cooling system is also a safety concern. Investigation of these issues will help determine the extent to which Flibe shows promise as a fusion power plant coolant or plasma-facing material

  14. Analytical Study of High Concentration PCB Paint at the Heavy Water Components Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lowry, N.J.

    1998-10-21

    This report provides results of an analytical study of high concentration PCB paint in a shutdown nuclear test reactor located at the US Department of Energy's Savannah River Site (SRS). The study was designed to obtain data relevant for an evaluation of potential hazards associated with the use of and exposure to such paints.

  15. Analytical Study of High Concentration PCB Paint at the Heavy Water Components Test Reactor

    International Nuclear Information System (INIS)

    This report provides results of an analytical study of high concentration PCB paint in a shutdown nuclear test reactor located at the US Department of Energy's Savannah River Site (SRS). The study was designed to obtain data relevant for an evaluation of potential hazards associated with the use of and exposure to such paints

  16. Reactor Technology Options Study for Near-Term Deployment of GNEP Grid-Appropriate Reactors

    International Nuclear Information System (INIS)

    World energy demand is projected to significantly increase over the coming decades. The International Energy Agency projects that electricity demand will increase 50% by 2015 and double by 2030, with most of the increase coming in developing countries as they experience double-digit rates of economic growth and seek to improve their standards of living. Energy is the necessary driver for human development, and the demand for energy in these countries will be met using whatever production technologies are available. Recognizing this inevitable energy demand and its implications for the United States, the U.S. National Security Strategy has proposed the Global Nuclear Energy Partnership (GNEP) to work with other nations to develop and deploy advanced nuclear recycling and reactor technologies. This initiative will help provide reliable, emission-free energy with less of the waste burden of older technologies and without making available separated plutonium that could be used by rogue states or terrorists for nuclear weapons. These new technologies will make possible a dramatic expansion of safe, clean nuclear energy to help meet the growing global energy demand. In other words, GNEP seeks to create an international regime to support large-scale growth in the worldwide use of nuclear energy without increasing the risk of nuclear weapon proliferation. This global expansion of nuclear power is strategically important to the United States for several reasons, including the following: (1) National security, by reducing the competition and potential for conflict over increasingly scarce fossil energy resources; (2) Economic security, by helping maintain stable prices for nonrenewable resources such as oil, gas, and coal; (3) Environmental security, by replacing or off-setting large-scale burning of greenhouse gas-emitting fuels for electricity production; and (4) Regaining technical leadership, through deployment of innovative U.S. technology-based reactors. Fully meeting

  17. Reactor Technology Options Study for Near-Term Deployment of GNEP Grid-Appropriate Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ingersoll, Daniel T [ORNL; Poore III, Willis P [ORNL

    2007-09-01

    World energy demand is projected to significantly increase over the coming decades. The International Energy Agency projects that electricity demand will increase 50% by 2015 and double by 2030, with most of the increase coming in developing countries as they experience double-digit rates of economic growth and seek to improve their standards of living. Energy is the necessary driver for human development, and the demand for energy in these countries will be met using whatever production technologies are available. Recognizing this inevitable energy demand and its implications for the United States, the U.S. National Security Strategy has proposed the Global Nuclear Energy Partnership (GNEP) to work with other nations to develop and deploy advanced nuclear recycling and reactor technologies. This initiative will help provide reliable, emission-free energy with less of the waste burden of older technologies and without making available separated plutonium that could be used by rogue states or terrorists for nuclear weapons. These new technologies will make possible a dramatic expansion of safe, clean nuclear energy to help meet the growing global energy demand. In other words, GNEP seeks to create an international regime to support large-scale growth in the worldwide use of nuclear energy without increasing the risk of nuclear weapon proliferation. This global expansion of nuclear power is strategically important to the United States for several reasons, including the following: (1) National security, by reducing the competition and potential for conflict over increasingly scarce fossil energy resources; (2) Economic security, by helping maintain stable prices for nonrenewable resources such as oil, gas, and coal; (3) Environmental security, by replacing or off-setting large-scale burning of greenhouse gas-emitting fuels for electricity production; and (4) Regaining technical leadership, through deployment of innovative U.S. technology-based reactors. Fully meeting

  18. HBGS (hydrate based gas separation) process for carbon dioxide capture employing an unstirred reactor with cyclopentane

    International Nuclear Information System (INIS)

    The effect of CP (cyclopentane) as a promoter/additive, in the HBGS (hydrate based gas separation) process for pre-combustion gas mixture was investigated by employing an unstirred reactor configuration. Gas uptake measurements were performed at two different temperatures (275.7 K and 285.7 K) and at an experimental pressure of 6.0 MPa to determine the kinetics of hydrate formation. Experiments were conducted with three different volumes (7.5, 15 and 22 ml) of CP and based on induction time and the rate of hydrate growth, 15 ml of CP was determined to be the optimal volume for carbon dioxide capture at 6.0 MPa and 275.7 K. In addition, the effect of a kinetic promoter, SDS (sodium dodecyl sulfate), was investigated. Surprisingly, no improvement in kinetic performance was observed at 6.0 MPa and 275.7 K in the presence of SDS and CP. From the study, it was found that at the optimal 15 ml CP (CP layer thickness of 1.8 mm), the average composition of carbon dioxide in the hydrate phase was 90.36 mol% with a separation factor of 17.82. Furthermore, the unstirred reactor also yielded better kinetic performance over the stirred tank reactor with the unstirred reactor having a 2.28 times higher average gas uptake. - Highlights: • HBGS process for pre-combustion capture in an unstirred reactor is presented. • Effect of cyclopentane as a thermodynamic promoter and sodium dodecyl sulfate as a kinetic promoter is investigated. • Cyclopentane significantly reduces the operating conditions and improves the kinetics for the HBGS process. • In this study, kinetic performance in an unstirred reactor is better than stirred tank reactor

  19. Study of Pu consumption in Advanced Light Water Reactors

    International Nuclear Information System (INIS)

    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

  20. Compact reactor studies: Technical progress report, 1 December 1986 through 30 November 1987

    International Nuclear Information System (INIS)

    This status report describes activities in two task areas: the Titan Reversed-field-Pinch Reactor (RFPR) study and Engineering Test Reactor (ETR) studies. In the Titan study, an investigation has been made of the integrated-blanket-coil concept. This paper describes the thermal hydraulics, wall loading, and divertor coil and fast wall design. The ETR studies have focused on breeding blanket/breeding shield design options. A description of water chemistry and corrosion effects is provided. Uncertainties in the calculational model for the dose and dose limit in the polyimide insulator are also expressed. (FI)

  1. Conceptual design study of fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    Since 1980 the design study has been conducted at JAERI for the Fusion Experimental Reactor (FER) which has been proposed to be the next machine to JT-60 in the Japanese long term program of fusion reactor development. During two years from 1984 to 1985 FER concept was reviewed and redesigned. This report is the summary of the results obtained in the review and redesign activities in 1984 and 85. In the first year FER concept was discussed again and its frame work was reestablished. According to the new frame work the major reactor components of FER were designed. In the second year the whole plant system design including plant layout plan was conducted as well as the more detailed design analysis of the reactor conponents. The newly established frame for FER design is as follows: 1) Plasma : Self-ignition. 2) Operation scenario : Quasi-steady state operation with long burn pulse. 3) Neutron fluence on the first wall : 0.3 MWY/M2. 4) Blanket : Non-tritium breeding blanket with test modules for breeding blanket development. 5) Magnets : Superconducting Magnets. (author)

  2. A model based on equations of kinetics to study nitrogen dioxide behavior within a plasma discharge reactor.

    Science.gov (United States)

    Abedi-Varaki, Mehdi; Ganjovi, Alireza; Shojaei, Fahimeh; Hassani, Zahra

    2015-01-01

    In this work, a zero-dimensional kinetics model is used to study the temporal behavior of different species such as charged particles, radicals and excited states inside a Dielectric Barrier Discharge plasma reactor. It is shown that, the reactor significantly reduces the concentration of nitrogen monoxide as an environmental pollutant. After a drastic increase, a decrease in the concentration of the NO2 molecules inside the reactor is seen. Nitrogen monoxide molecules with a very low concentration are produced inside the reactor and its quick conversion to other products is proved. The obtained results are compared with the existing experimental and simulation findings, whenever possible.

  3. Research Nuclear Reactors

    International Nuclear Information System (INIS)

    Published in English and in French, this large report first proposes an overview of the use and history of research nuclear reactors. It discusses their definition, and presents the various types of research reactors which can be either related to nuclear power (critical mock-ups, material test reactors, safety test reactors, training reactors, prototypes), or to research (basic research, industry, health), or to specific particle physics phenomena (neutron diffraction, isotope production, neutron activation, neutron radiography, semiconductor doping). It reports the history of the French research reactors by distinguishing the first atomic pile (ZOE), and the activities and achievements during the fifties, the sixties and the seventies. It also addresses the development of instrumentation for research reactors (neutron, thermal, mechanical and fission gas release measurements). The other parts of the report concern the validation of neutronics calculations for different reactors (the EOLE water critical mock-up, the MASURCA air critical mock-up dedicated to fast neutron reactor study, the MINERVE water critical mock-up, the CALIBAN pulsed research reactor), the testing of materials under irradiation (OSIRIS reactor, laboratories associated with research reactors, the Jules Horowitz reactor and its experimental programs and related devices, irradiation of materials with ion beams), the investigation of accident situations (on the CABRI, Phebus, Silene and Jules Horowitz reactors). The last part proposes a worldwide overview of research reactors

  4. Performances and microbial features of an aerobic packed-bed biofilm reactor developed to post-treat an olive mill effluent from an anaerobic GAC reactor

    Directory of Open Access Journals (Sweden)

    Marchetti Leonardo

    2006-04-01

    , enriched significantly in the biofilter throughout the treatment. Conclusion The silica-bead packed bed biofilm reactor developed and characterized in this study was able to significantly decontaminate anaerobically digested OMWs. Therefore, the application of an integrated anaerobic-aerobic process resulted in an improved system for valorization and decontamination of OMWs.

  5. Enhanced situation awareness and decision making for an intelligent reconfigurable reactor power controller

    International Nuclear Information System (INIS)

    A Learning Automata based intelligent reconfigurable controller has been adapted for use as a reactor power controller to achieve improved reactor temperature performance. The intelligent reconfigurable controller is capable of enforcing either a classical or an optimal reactor power controller based on control performance feedback. Four control performance evaluation measures: dynamically estimated average quadratic temperature error, power, rod reactivity and rod reactivity rate were developed to provide feedback to the control decision component of the intelligent reconfigurable controller. Fuzzy Logic and Neural Network controllers have been studied for inclusion in the bank of controllers that form the intermediate level of an enhanced intelligent reconfigurable reactor power controller (IRRPC). The increased number of alternatives available to the supervisory level of the IRRPC requires enhanced situation awareness. Additional performance measures have been designed and a method for synthesizing them into a single indication of the overall performance of the currently enforced reactor power controller has been conceptualized. Modification of the reward/penalty scheme implemented in the existing IRRPC to increase the quality of the supervisory level decision process has been studied. The logogen model of human memory (Morton, 1969) and individual controller design information could be used to allocate reward to the most appropriate controller. Methods for allocating supervisory level attention were also studied with the goal of maximizing learning rate

  6. Enhanced situation awareness and decision making for an intelligent reconfigurable reactor power controller

    Energy Technology Data Exchange (ETDEWEB)

    Kenney, S.J.; Edwards, R.M. [Pennsylvania State Univ., University Park, PA (United States). Dept. of Nuclear Engineering

    1996-07-01

    A Learning Automata based intelligent reconfigurable controller has been adapted for use as a reactor power controller to achieve improved reactor temperature performance. The intelligent reconfigurable controller is capable of enforcing either a classical or an optimal reactor power controller based on control performance feedback. Four control performance evaluation measures: dynamically estimated average quadratic temperature error, power, rod reactivity and rod reactivity rate were developed to provide feedback to the control decision component of the intelligent reconfigurable controller. Fuzzy Logic and Neural Network controllers have been studied for inclusion in the bank of controllers that form the intermediate level of an enhanced intelligent reconfigurable reactor power controller (IRRPC). The increased number of alternatives available to the supervisory level of the IRRPC requires enhanced situation awareness. Additional performance measures have been designed and a method for synthesizing them into a single indication of the overall performance of the currently enforced reactor power controller has been conceptualized. Modification of the reward/penalty scheme implemented in the existing IRRPC to increase the quality of the supervisory level decision process has been studied. The logogen model of human memory (Morton, 1969) and individual controller design information could be used to allocate reward to the most appropriate controller. Methods for allocating supervisory level attention were also studied with the goal of maximizing learning rate.

  7. Fast reactor development strategy targets study in China

    International Nuclear Information System (INIS)

    China is a big developing Country who needs a huge energy resources and a rapid growing rate. Considering energy resources limited and environment issues it is sure that the nuclear energy will be becoming one of the main energy resources. The Government has decided to develop the nuclear power capacity to 40 GW in 2020. It is envisaged that it will reach to 240 GW in 2050. It is stimulate us to consider conscientiously the development of the fast breeder reactor's and related closed nuclear fuel cycle by the limitation of Uranium resources and uncertainties of international Uranium market. Followings are the proposed strategic targets of fast reactor development in China. (1) To realize the operation of commercial fast breeder reactors with an unit size of 800-900 MWe and one site-multi reactors in 2030. (2) To develop the nuclear power capacity to 240 GW in 2050. (3) To replace step by step the fossil fuel utilization in large scale by nuclear energy beyond 2050. (authors)

  8. Comparative end-plug study for tandem mirror reactors

    International Nuclear Information System (INIS)

    A comparative evaluation was made of several end plug configurations for tandem mirror fusion reactors with thermal barriers. The axicell configuration has been selected for further study and will be the basis for a detailed conceptual design study to be carried out over the next two years. The axicell end plug has a simple mirror cell produced by two circular coils followed by a transition coil and a yin-yang pair, which provides for MHD stability

  9. Design study of blanket structure for tokamak experimental fusion reactor

    International Nuclear Information System (INIS)

    Design study of the blanket structure for JAERI Experimental Fusion Reactor (JXFR) has been carried out. Studied here were fabrication and testing of the blanket structure (blanket cells, blanket rings, piping and blanket modules), assembly and disassembly of the blanket module, and monitering and testing technique. Problems in design and fabrication of the blanket structure could be revealed. Research and development problems for the future were also disclosed. (author)

  10. A Compact Gas-Cooled Fast Reactor with an Ultra-Long Fuel Cycle

    Directory of Open Access Journals (Sweden)

    Hangbok Choi

    2013-01-01

    Full Text Available In an attempt to allow nuclear power to reach its full economic potential, General Atomics is developing the Energy Multiplier Module (EM2, which is a compact gas-cooled fast reactor (GFR. The EM2 augments its fissile fuel load with fertile materials to enhance an ultra-long fuel cycle based on a “convert-and-burn” core design which converts fertile material to fissile fuel and burns it in situ over a 30-year core life without fuel supplementation or shuffling. A series of reactor physics trade studies were conducted and a baseline core was developed under the specific physics design requirements of the long-life small reactor. The EM2 core performance was assessed for operation time, fuel burnup, excess reactivity, peak power density, uranium utilization, etc., and it was confirmed that an ultra-long fuel cycle core is feasible if the conversion is enough to produce fissile material and maintain criticality, the amount of matrix material is minimized not to soften the neutron spectrum, and the reactor core size is optimized to minimize the neutron loss. This study has shown the feasibility, from the reactor physics standpoint, of a compact GFR that can meet the objectives of ultra-long fuel cycle, factory-fabrication, and excellent fuel utilization.

  11. Supercell Depletion Studies for Prismatic High Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    J. Ortensi

    2012-10-01

    The traditional two-step method of analysis is not accurate enough to represent the neutronic effects present in the prismatic high temperature reactor concept. The long range coupling of the various regions in high temperature reactors poses a set of challenges that are not seen in either LWRs or fast reactors. Unlike LWRs, which exhibit large, localized effects, the dominant effects in PMRs are, for the most part, distributed over larger regions, but with lower magnitude. The 1-D in-line treatment currently used in pebble bed reactor analysis is not sufficient because of the 2-D nature of the prismatic blocks. Considerable challenges exist in the modeling of blocks in the vicinity of reflectors, which, for current small modular reactor designs with thin annular cores, include the majority of the blocks. Additional challenges involve the treatment of burnable poisons, operational and shutdown control rods. The use of a large domain for cross section preparation provides a better representation of the neutron spectrum, enables the proper modeling of BPs and CRs, allows the calculation of generalized equivalence theory parameters, and generates a relative power distribution that can be used in compact power reconstruction. The purpose of this paper is to quantify the effects of the reflector, burnable poison, and operational control rods on an LEU design and to delineate an analysis approach for the Idaho National Laboratory. This work concludes that the use of supercells should capture these long-range effects in the preparation of cross sections and along with a set of triangular meshes to treat BPs, and CRs a high fidelity neutronics computation is attainable.

  12. Experimental study on thermal stratification in a reactor hot plenum of a Japanese demonstration LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Koga, Tomonari [Central Research Inst. of Electric Power Industry, Abiko, Chiba (Japan). Abiko Research Lab.; Yamamoto, K.; Takakuwa, M.; Kajiwara, H.; Watanabe, O.; Akamatsu, K.

    1997-12-31

    Thermal stratification which occurs in a reactor hot plenum after reactor trip has been regarded as one of the most serious phenomena in the thermal-hydraulics of LMFBR. Using a 1/8th scale water model, an experimental study has been conducted to estimate the thermal stratification for a Japanese demonstration LMFBR (DFBR). In the present study, reactor trip was simulated by changing the core outlet temperature with maintaining a constant flow rate. Temperature distribution was measured during the transient and detailed phenomena have been acquired in the study. A severe density interface on structural integrity occurs in a hot plenum under the thermal stratification. Experimental results for temperature gradient and rising speed of the density interface were estimated based on a similarity rule so that an actual condition in the DFBR could be fully discerned. (author)

  13. Study on axial offset oscillation for WWER-1000 reactor by using WWER-1000 simulator

    International Nuclear Information System (INIS)

    In the operation of thermal neutron reactors, it is known that the spatial xenon oscillations arise frequently. The nature of these oscillations is that almost of power concentrate just at a small region in the reactor core volume. The characteristic parameter for the axial power distribution is axial offset. In this subject, the cause of axial offset oscillation and its characteristics are studied. We investigate axial offset oscillation in begin of fuel cycle (BOC) and end of fuel cycle (EOC) of loading 1 enrichment and loading 5 enrichment for WWER-1000 reactor, using WWER-1000 reactor simulation program that was originally developed by IAEA. The results are: (1) the formation of axial offset oscillation is due to periodic deviation from an equilibrium distribution of iodine, xenon and neutron flux density between the upper half and the lower half of the reactor core, when the control group number 1 is inserted into the bottom of reactor core; (2) regarding the same fuel enrichment, in BOC, offset oscillates with larger amplitude and slower damping than in EOC. On the other hand, in higher fuel enrichment, offset oscillates with smaller amplitude and quicker damping than in lower fuel enrichment. (author)

  14. Small and medium power reactors: project initiation study, Phase 1

    International Nuclear Information System (INIS)

    In conformity with the Agency's promotional role in the peaceful uses of nuclear energy, IAEA has provided, over the past 20 years, assistance to Member States, particularly developing countries, in planning for the introduction of nuclear power plants in the Small and Medium range (SMPR). However these efforts did not produce any significant results in the market introduction of these reactors, due to various factors. In 1983 the Agency launched a new SMPR Project Initiation Study with the objective of surveying the available designs, examining the major factors influencing the decision-making processes in Developing Countries and thereby arriving at an estimate of the potential market. Two questionnaires were used to obtain information from possible suppliers and prospective buyers. The Nuclear Energy Agency of OECD assisted in making a study of the potential market in industrialized countries. The information gained during the study and discussed during a Technical Committee Meeting on SMPRs held in Vienna in March 1985, along with the contribution by OECD-NEA is embodied in the present report

  15. An Axial Dispersion Model for Evaporating Bubble Column Reactor

    Institute of Scientific and Technical Information of China (English)

    谢刚; 李希

    2004-01-01

    Evaporating bubble column reactor (EBCR) is a kind of aerated reactor in which the reaction heat is removed by the evaporation of volatile reaction mixture. In this paper, a mathematical model that accounts for the gas-liquid exothermic reaction and axial dispersions of both gas and liquid phase is employed to study the performance of EBCR for the process of p-xylene(PX) oxidation. The computational results show that there are remarkable concentration and temperature gradients in EBCR for high ratio of height to diameter (H/DT). The temperature is lower at the bottom of column and higher at the top, due to rapid evaporation induced by the feed gas near the bottom. The concentration profiles in the gas phase are more nonuniform than those (except PX) in the liquid phase, which causes more solvent burning consumption at high H/DT ratio. For p-xylene oxidation, theo ptimal H/DT is around 5.

  16. Study on Fuels and Materials of Reactor

    Institute of Scientific and Technical Information of China (English)

    1996-01-01

    2.1 Effect of N+ Ions Implanted on Electrochemical Property of 316 s.s.,Super-pure Aluminium and 6061 Aluminium Alloy Ma Yan Lu Haolin Yang Qifa Yang Hongguang 316 s.s., super-pure aluminium and 6061 aluminium alloy are three materials widely used in engineering. For improving some of their mechanical properties, such as self-welding, friction and fretting resistence properties, surface processing is employed sometimes, for example, ion implantation, thermal spray coating, nitridation. Ion implantation is an advanced method of material surface treatment. This report describes the effect of corrosion resistance on three materials above mentioned by N+ ions implanted. Firstly, the samples were implanted with 160 keV N+ ion and the total dose was 15×1017 cm-2. The method of controled voltage was used to measure the electrochemical property. The corrosive medium was 0.5 mol·L-1 H2SO4 and reference electrode was calomel electrode. The anodic polarization curves of materials implanted and unimplanted were obtained by Model 350A corrosion measuring instrument(Fig. 1, 2,3).

  17. Modelling of an ASR countercurrent pyrolysis reactor with nonlinear kinetics

    Energy Technology Data Exchange (ETDEWEB)

    Chiarioni, A.; Reverberi, A.P.; Dovi, V.G. [Universita degli Studi di Genova (Italy). Dipartimento di Ingegneria Chimica e di Processo ' G.B. Bonino' ; El-Shaarawi, A.H. [National Water Research Institute, Burlington, Ont. (Canada)

    2003-10-01

    The main objective of this work is focused on the modelling of a steady-state reactor where an automotive shredder residue (ASR) is subject to pyrolysis. The gas and solid temperature inside the reactor and the relevant density profiles of both phases are simulated for fixed values of the geometry of the apparatus and a lumped kinetic model is adopted to take into account the high heterogeneity of the ASR material. The key elements for the simulation are the inlet solid temperature and the outlet gas temperature. The problem is modelled by a system of first-order boundary-value ordinary differential equations and it is solved by means of a relaxation technique owing to the nonlinearities contained in the chemical kinetic expression. (author)

  18. An Innovative Reactor Technology to Improve Indoor Air Quality

    Energy Technology Data Exchange (ETDEWEB)

    Rempel, Jane [TIAX LLC., Lexington, MA (United States)

    2013-03-30

    As residential buildings achieve tighter envelopes in order to minimize energy used for space heating and cooling, accumulation of indoor air pollutants such as volatile organic compounds (VOCs), becomes a major concern causing poor air quality and increased health risks. Current VOC removal methods include sorbents, ultraviolet photocatalytic oxidation (UVPCO), and increased ventilation, but these methods do not capture or destroy all VOCs or are prohibitively expensive to implement. TIAX's objective in this program was to develop a new VOC removal technology for residential buildings. This novel air purification technology is based on an innovative reactor and light source design along with UVPCO properties of the chosen catalyst to purify indoor air and enhance indoor air quality (IAQ). During the program we designed, fabricated and tested a prototype air purifier to demonstrate its feasibility and effectiveness. We also measured kinetics of VOC destruction on photocatalysts, providing deep insight into reactor design.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-12-01

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

  20. Conceptual design study of fusion experimental reactor (FY86 FER)

    International Nuclear Information System (INIS)

    This report summarizes the FER magnet design which was conducted last year (1986). Main objective of the new FER design is to have better cost performance of the machine. The physics assumptions are reviewed to reduce risks. Optimization of the physics design and improvements of the engineering design have been done without changing missions of the device. After a preliminary investigation for the optimization and improvements, six FER concepts have been developed to establish the improved design point, and have been studied in more detail. In the magnet design, the improvements of superconducting magnet design were mainly investigated to reduce the reactor size. A normal conductor was studied as an alternative option for appling to the special poloidal field coils that were located on the interior to the toroidal field coils. Some improvements were made on the superconducting magnet design. Based on the preliminary investigation, the magnet design specifications have been modified somewhat. The conceptual design of the magnet system components have been done for the candidate FER concepts. (author)

  1. Reactor physics studies for assessment of tramp uranium methods

    International Nuclear Information System (INIS)

    This paper presents calculation studies towards validation of a methodology for estimations of the tramp uranium mass from water chemistry measurements. Particular emphasis is given to verify, from a reactor physics point of view, the justification basis for the so-called 'Pu-based model' versus the 'U-based model' as a key assumption for the methodology. The computational studies are carried out for a typical BWR fuel assembly with CASMO-5M and MCNPX. By approximating the evolution of fissile nuclides and the fraction of 235U fissions to total fissions in different zones of a fuel rod, including tramp uranium on the clad surface, it is found that Pu gives the dominant contribution to fissions for tramp uranium after an irradiation on the outer clad surface of at least one cycle in a BWR. Thus, the use of the so-called Pu model for the determination of the tramp uranium mass (this means in particular using the yields for 239Pu fission) appears justified in the cases considered. On that basis, replacing the older U model by a Pu model is recommended. (authors)

  2. Reactor physics studies for assessment of tramp uranium methods

    Energy Technology Data Exchange (ETDEWEB)

    Grimm, P.; Vasiliev, A.; Wieselquist, W.; Ferroukhi, H. [Paul Scherrer Institut, CH 5232 Villigen (Switzerland); Ledergerber, G. [Kernkraftwerk Leibstadt AG, CH 5325 Leibstadt (Switzerland)

    2012-07-01

    This paper presents calculation studies towards validation of a methodology for estimations of the tramp uranium mass from water chemistry measurements. Particular emphasis is given to verify, from a reactor physics point of view, the justification basis for the so-called 'Pu-based model' versus the 'U-based model' as a key assumption for the methodology. The computational studies are carried out for a typical BWR fuel assembly with CASMO-5M and MCNPX. By approximating the evolution of fissile nuclides and the fraction of {sup 235}U fissions to total fissions in different zones of a fuel rod, including tramp uranium on the clad surface, it is found that Pu gives the dominant contribution to fissions for tramp uranium after an irradiation on the outer clad surface of at least one cycle in a BWR. Thus, the use of the so-called Pu model for the determination of the tramp uranium mass (this means in particular using the yields for {sup 239}Pu fission) appears justified in the cases considered. On that basis, replacing the older U model by a Pu model is recommended. (authors)

  3. Study on a decay heat removal system of light water reactors using air coolers

    International Nuclear Information System (INIS)

    In the present work, a passive decay heat removal system for light water reactors (LWRs) based on a new concept is studied referring to an air cooling system (ACS) of the fast breeder reactor Monju. The present study will contribute to the reduction of severe accident risks of nuclear power plants. In this system, a blower for an air cooler (AC) is operated using the rotation of a small steam turbine by generated steam in order to cool heat transfer tubes by forced convection of air. The purpose of the present work is to investigate the plant transient caused by a station blackout (SBO) using the plant system code NETFLOW++ and decay heat removal characteristics. A calculation model is the Advanced Boiling Water Reactor (ABWR) in Japan. (author)

  4. Theoretical study for ICRF sustained LHD type p-{sup 11}B reactor

    Energy Technology Data Exchange (ETDEWEB)

    Watanabe, Tsuguhiro (ed.)

    2003-04-01

    This is a summary of the workshop on 'Theoretical Study for ICRF Sustained LHD Type p-{sup 11}B Reactor' held in National Institute for Fusion Science (NIFS) on July 25, 2002. In the workshop, study of LHD type D-{sup 3}He reactor is also reported. A review concerning the advanced nuclear fusion fuels is also attached. This review was reported at the workshop of last year. The development of the p-{sup 11}B reactor research which uses the LHD magnetic field configuration has been briefly summarized in section 1. In section 2, an integrated report on advanced nuclear fusion fuels is given. Ignition conditions in a D-{sup 3}He helical reactor are summarized in section 3. 0-dimensional particle and power balance equations are solved numerically assuming the ISS95 confinement law including a confinement factor ({gamma}{sub HH}). It is shown that high average beta plasma confinement, a large confinement factor ({gamma}{sub HH} > 3) and the hot ion mode (T{sub i}/T{sub e} > 1.4) are necessary to achieve the ignition of the D-{sup 3}He helical reactor. Characteristics of ICRF sustained p-{sup 11}B reactor are analyzed in section 4. The nuclear fusion reaction rate < {sigma}{upsilon} > is derived assuming a quasilinear plateau distribution function (QPDF) for protons, and an ignition condition of p-{sup 11}B reactor is shown to be possible. The 3 of the presented papers are indexed individually. (J.P.N.)

  5. Inertial Fusion Energy reactor design studies: Prometheus-L, Prometheus-H

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-03-01

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

  7. Report of the APS Neutrino Study Reactor Working Group

    Energy Technology Data Exchange (ETDEWEB)

    Abouzaid, E.; Anderson, K.; Barenboim, G.; Berger, B.; Blucher, E.; Bolton, T.; Choubey, S.; Conrad, J.; Formaggio, J.; Freedman, S.; Finely, D.; Fisher, P.; Fujikawa, B.; Gai, M.; Goodman, M.; de Goueva, A.; Hadley, N.; Hahn, R.; Horton-Smith, G.; Kadel, R.; Kayser, B.; Heeger, K.; Klein, J.; Learned, J.; Lindner, M.; Link, J.; Luk, K.-B.; McKeown, R.; Mocioiu, I.; Mohapatra, R.; Naples, D.; Peng, J.; Petcov, S.; Pilcher, J.; Rapidis, P.; Reyna, D.; Shaevitz, M.; Shrock, R.; Stanton, N.; Stefanski, R.; Yamamoto, R.; Worcester, M.

    2004-10-28

    . If {theta}{sub 13} turns out to be smaller than these values, one will need other strategies for getting to the physics. Thus, an unambiguous reactor measurement of {theta}{sub 13} is an important ingredient in planning the strategy for the future neutrino program.

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

    Science.gov (United States)

    Brown, Nicholas Robert

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

  9. Purification of bioethanol effluent in an UASB reactor system with simultaneous biogas formation

    DEFF Research Database (Denmark)

    Torry-Smith, Mads Peter; Sommer, Peter; Ahring, Birgitte Kiær

    2003-01-01

    In this study, the prospect of using an Upflow Anaerobic Sludge Blanket (UASB) reactor for detoxification of process water derived from bioethanol production has been investigated. The bioethanol effluent (BEE) originated from wet oxidized wheat straw fermented by Saccharomyces cerevisiae and The...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-12-01

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

  11. Conceptual design study of Hyb-WT as fusion–fission hybrid reactor for waste transmutation

    International Nuclear Information System (INIS)

    Highlights: • Conceptual design study of fusion-fission hybrid reactor for waste transmutation. • MCNPX and MONTEBURNS are compared for transmutation performance of WT-Hyb. • Detailed neutronic performance of final optimized Hyb-WT design is analyzed. • A new tube-in-duct core design is implemented and compared with pin type design. • Study shows many aspects of hybrid reactor even though scope was limited to neutronic analysis. - Abstract: This study proposes a conceptual design of a hybrid reactor for waste transmutation (Hyb-WT). The design of Hyb-WT is based on a low-power tokamak (less than 150 MWt) and an annular ring-shaped reactor core with metal fuel (TRU 60 w/o, Zr 40 w/o) and a fission product (FP) zone. The computational code systems MONTEBURNS and MCNPX2.6 are investigated for their suitability in evaluating the performance of Hyb-WT. The overall design performance of the proposed reactor is determined by considering pin-type and tube-in-duct core designs. The objective of such consideration is to explore the possibilities for enhanced transmutation with reduced wall loading from fusion neutrons and reduced transuranic (TRU) inventory. TRU and FP depletion is analyzed by calculating waste transmutation ratio, mass burned per full power year (in units of kg/fpy), and support ratio. The radio toxicity analysis of TRUs and FPs is performed by calculating the percentage of toxicity reduction in TRU and FP over a burn cycle

  12. Collective study of plans and feature of the reactor for medical usage

    International Nuclear Information System (INIS)

    In order to construct the reactor for medical usage comparative studies of irradiating apparatus were performed, and plans to construct medical reactors were constructed by 20 groups consisted of universities, institutes, and companies. As for facilities, a research for TRIGA type reactor, combination of a reactor and an accelerator, and problems in constructing a reactor were investigated. Examinations, with regard to flux, were carried out from the view point of flux variation due to absorber and monitoring thermal neutron dose, while irradiating boron. Some physical problems of neutron detector, neutron source, and preparing enriched isotopes of 10B were also studied. Analysis of boron was developed by utilization of α autoradiography, synthesis of Na210B12H11SH, and enrichment of 10B. In the field of biomedical science, application of neutron capture method to cerebral tumors, histo-immunological study of the normal brain by enzyme antibody method, and selective radiotherapy of malignant skin tumors were examined using animals. Radiotherapy by neutron capture was carried out to the patients with various tumors, and the remote anesthetization was also tried. (Nakanishi, T.)

  13. Kinetic evaluation and process performance of an upflow anaerobic filter reactor degrading terephthalic acid.

    Science.gov (United States)

    Davutluoglu, Orkun I; Seckin, Galip

    2014-01-01

    The anaerobic degradation of terephthalic acid (TA) as the sole organic carbon source was studied in an upflow anaerobic filter (UAF) reactor. The reactor was seeded with biomass obtained from a full-scale upflow anaerobic sludge bed (UASB) reactor and was used to treat wastewater from a petrochemical facility producing dimethyl terephthalate. The UAF reactor was operated for 252 d with a constant hydraulic retention time of 24 h, and the organic loading rate (OLR) was gradually increased from 1 to 10 g-chemical oxygen demand (COD)/L d. After a lag period of approximately 40 d, the COD removal efficiency increased exponentially and high removal rate values (≈90%) were obtained, except for at highest OLR (10 g-COD/L d). The high removal rates and the robustness of the reactor performance could be attributed to the formation of biofilm as well as granular sludge. The methane production rates (0.22 to 2.15 L/d) correlated well with the removed OLRs (0.3 to 6.8 g-COD/L d) during the various phases of treatment, indicating that the main mechanism of TA degradation occurs via methanogenic reactions. The average methane content of the produced biogas was 70.3%. The modified Stover-Kincannon model was found to be applicable for the anaerobic degradation of TA in UAFs (Umax = 64.5, KB = 69.1 g-COD/L d and Ymax = 0.27 L-CH4/g-CODremoved). These results suggest that UAF reactors are among the most effective reactor configurations for the anaerobic degradation of TA.

  14. High temperature gas-cooled reactor: gas turbine application study

    International Nuclear Information System (INIS)

    The high-temperature capability of the High-Temperature Gas-Cooled Reactor (HTGR) is a distinguishing characteristic which has long been recognized as significant both within the US and within foreign nuclear energy programs. This high-temperature capability of the HTGR concept leads to increased efficiency in conventional applications and, in addition, makes possible a number of unique applications in both electrical generation and industrial process heat. In particular, coupling the HTGR nuclear heat source to the Brayton (gas turbine) Cycle offers significant potential benefits to operating utilities. This HTGR-GT Application Study documents the effort to evaluate the appropriateness of the HTGR-GT as an HTGR Lead Project. The scope of this effort included evaluation of the HTGR-GT technology, evaluation of potential HTGR-GT markets, assessment of the economics of commercial HTGR-GT plants, and evaluation of the program and expenditures necessary to establish HTGR-GT technology through the completion of the Lead Project

  15. Design study of toroidal magnets for tokamak experimental power reactors

    International Nuclear Information System (INIS)

    This report contains the results of a six-month study of superconducting toroidal field coils for a Tokamak Experimental Power Reactor to be built in the late 1980s. The designs are for 8 T and 12 T maximum magnetic field at the superconducting winding. At each field level two main concepts were generated; one in which each of the 16 coils comprising the system has an individual vacuum vessel and the other in which all the coils are contained in a single vacuum vessel. The coils have a D shape and have openings of 11.25 m x 7.5 m for the 8 T coils and 10.2 m x 6.8 m for the 12 T coils. All the designs utilize rectangular cabled conductor made from copper stabilized Niobium Titanium composite which operates at 4.2 K for the 8 T design and at 2.5 K for the 12 T design. Manufacturing procedures, processes and schedule estimates are also discussed

  16. A study of reactor-neutron-induced reactions

    International Nuclear Information System (INIS)

    Cross sections of the second neutron capture in the double-neutron-capture process were studied in the irradiation of 26Mg, 64Ni, 93Nb, and 164Dy with reactor neutrons. Weak radioactivities produced were determined with sufficient accuracy by taking advantage of high resolution Ge(Li) detectors. The neutron spectrum at the irradiating site was defined with Westcott's epithermal index determined by the use of the cadmium ratio method. The isomer cross-section ratio was obtained with the 94Nb target. The statistical theory was found to be applicable to such an unstable or odd-odd nuclide as 94Nb. Cross sections of isomers of 165Dy were determined. The capture cross section of the ground state nuclei was found greater than that of the metastable state nuclei in this case. This differs from the other four isomers observed so far. Cross sections of 28Al(n, p)28Mg and 58Ni(n, 2n)57Ni reactions with fission neutrons were also measured. A simple formula was proposed for the systematics of (n, 2n) reaction cross sections, which enabled prediction of the cross section more accurately than the previously proposed formulae. (author)

  17. High temperature gas-cooled reactor: gas turbine application study

    Energy Technology Data Exchange (ETDEWEB)

    1980-12-01

    The high-temperature capability of the High-Temperature Gas-Cooled Reactor (HTGR) is a distinguishing characteristic which has long been recognized as significant both within the US and within foreign nuclear energy programs. This high-temperature capability of the HTGR concept leads to increased efficiency in conventional applications and, in addition, makes possible a number of unique applications in both electrical generation and industrial process heat. In particular, coupling the HTGR nuclear heat source to the Brayton (gas turbine) Cycle offers significant potential benefits to operating utilities. This HTGR-GT Application Study documents the effort to evaluate the appropriateness of the HTGR-GT as an HTGR Lead Project. The scope of this effort included evaluation of the HTGR-GT technology, evaluation of potential HTGR-GT markets, assessment of the economics of commercial HTGR-GT plants, and evaluation of the program and expenditures necessary to establish HTGR-GT technology through the completion of the Lead Project.

  18. Feasibility study of advanced fuel burning nuclear reactors

    International Nuclear Information System (INIS)

    An investigation has been conducted to determine both physics, engineering and economic aspects of fusion power reactors based on magnetic confinement and on burning advanced fuels (AFs). DT burning Tokamaks are taken as reference concept. We show that the attractive features of advanced fuels, in particular of neutronlean proton-based AFs, can be combined, in appropriately designed AF reactors (high beta), with power densities comparable to or even higher than those achievable in DT Tokamaks. Moreover we identify physical requirements which would assure Q values well above unity. As an example a semi-open confinement scheme is analyzed based on a self-consistent plasma calculation. We find that a mirror, even if only ''semi-open'' as a result of strong diamagnetism, can barely be expected to achieve high Q values. Therefore confinement schemes such as compact tori, multipole surmacs etc. may be required to burn AFs. We conclude that the economics of AF reactors, as determined by the nuclear boiler power density, may be superior to that of DT-rectors if low recirculating power fractions can be obtained by appropriate plasma tayloring (high fractional transfer of fusion power to ions required). A more detailed investigation is suggested for proton-based fuel cycles. (orig.)

  19. Study and optimization of a low power plasma reactor for the deposition of ZnO doped and undoped with photovoltaic properties from an aqueous solution

    OpenAIRE

    Ma, Alexandre

    2015-01-01

    This work is part of the Research and Development of Photovoltaic. The aim was to study, develop and optimize a new deposition plasma process for the elaboration of zinc oxide thin layers (ZnO) as the window layer in Cu(In,Ga)Se2 solar cells of. The particularity of this process is to quickly realize oxide layers (≥ 0.6 nm/s) from an aqueous solution of non-toxic precursors, interacting in the form of droplets, with the plasma. The feasibility of the ZnO deposition by the low power plasma rea...

  20. STUDY OF A SEQUENCING BATCH REACTOR TREATING WASTEWATER FROM AN INDUSTRY OF SOFT DRINKS = ESTUDO DE UM REATOR DE LODOS ATIVADOS POR BATELADA PARA O TRATAMENTO DOS EFLUENTES DE UMA INDÚSTRIA DE REFRIGERANTE

    Directory of Open Access Journals (Sweden)

    Rafael Brito de Moura

    2009-01-01

    Full Text Available This work aimed to study a pilot scale sequencing batch reactor in order to enhance its performance treating wastewater from a soft drink industry. Initially, 5 liters of anaerobic sludge from an UASB (Upflow Anaerobic Sludge Blanket reactor were inoculated in the reactor, which was acclimatized to the new conditions during two months. Conducted microscopy examination was developed to observe the microorganisms present in the sludge. After this stage, experimental procedure was divided in six phases with different times of cycles ranging from 8 to 18 hours, divided in aeration, stir and settle. As a result, higher removal percentages of COD were obtained in the 10 hours cycle, reaching an average of 89.8%. The cycle that had higher nitrogen removals was of 14 hours, with mean removal of 78.8%. In all cycles tested, the presence of nitrate in treated effluent was not detected, characterizing biological denitrification. = Esse trabalho teve como objetivo estudar o funcionamento de um reator sequencial por batelada em escala piloto, de forma a aumentar seu rendimento no tratamento de águas residuárias provenientes de uma indústria de refrigerantes. Inicialmente, foram inoculados 5 litros de lodo anaeróbio proveniente de um reator UASB (Upflow Anaerobic Sludge Blanket, sendo este aclimatado para o novo sistema durante dois meses. Foram realizados exames microscópicos do lodo para observar os microrganismos presentes nele. Após essa etapa, o período experimental foi dividido em seis fases de operação com tempos de ciclo variando entre 8 e 18 horas, divididos em aeração, agitação e sedimentação. Como resultado, obteve-se uma maior remoção de DQO para o ciclo de 10 horas, alcançando um percentual médio de 89,8%. O ciclo que apresentou maior remoção de nitrogênio foi o de 14 horas, com remoção de 78,8%. Em todos os ciclos testados, não houve a presença de nitrato no efluente tratado, caracterizando a desnitrificação.

  1. Dominant bacteria correlated with elimination of sludge in an innovative reactor

    Institute of Scientific and Technical Information of China (English)

    Shanshan Lin; Ying Wang; Jifang Lin; Xinrui Wang; Huili Gong

    2009-01-01

    This study was conducted to identify bacteria responsible for the elimination of sludge in an innovative gravel contact oxidation reactor (GCOR) and a conventional activated sludge reactor (ASR),Fluorescent in situ hybridization revealed that α-,β-,γ-and ε-Proteobacteria and Bacteroidetes were present in both reactors.In the GCOR,γ-Proteobacteria accounted for 13% of the biofilm biomass on the carrier,while α-,ε-and β-Proteobacteria accounted for 12%.However,the predominant bacteria in the pores of the carrier in the GCOR were Bacteroidetes (18%),ε-Proteobacteria (13%),β-and α-Proteobacteria (10%) and γ-Proteobacteria (9%).Conversely,β-Proteobacteria (18%),Bacteroidetes (13%),α-and γ-Proteobacteria (12%) and ε-Proteobacteria (5%) dominated the sediment community of the ASR.

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

    Directory of Open Access Journals (Sweden)

    Shin Jae Sun

    2016-01-01

    Full Text Available GTL (Gas-to-liquid process is becoming an attractive technology which can produce liquid petroleum products using natural gas. As a part of preliminary design of GTL-FPSO application, process simulation analysis for conceptual design and optimization of reformers and F-T reactors are performed in GTL-FPSO applications by implementing the user made subroutine programs of kinetic equations into PRO/II PROVISION simulator. As for the F-T reactors, Plug Flow Reactor (PFR model is used with detailed kinetics equations over two different Fe based catalysts (Fe-Cu-K and K/Fe-Cu-Al. Dry reformer is also studied with Plug Flow Reactor (PFR model. In this study, simulation results are compared with available experimental data and found well agreed with experimental data for both reformer and FT reactor. The Peng-Robinson equation of state is also used to calculate the vapor phase non-idealities and vapor-liquid equilibrium. The optimum operating conditions and process simulation analysis are also presented.

  3. MATHEMATICAL MODELLING OF METHANE STEAM REFORMING IN A MEMBRANE REACTOR: AN ISOTHERMIC MODEL

    OpenAIRE

    ASSAF E.M.; JESUS C.D.F.; J.M. ASSAF

    1998-01-01

    A mathematical modelling of one-dimensional, stationary and isothermic membrane reactor for methane steam reforming was developed to compare the maximum yield for methane conversion in this reactor with that in a conventional fixed-bed reactor. Fick's first law was used to describe the mechanism of hydrogen permeation. The variables studied include: reaction temperature, hydrogen feed flow rate and membrane thickness. The results show that the membrane reactor presents a higher methane conver...

  4. MATHEMATICAL MODELLING OF METHANE STEAM REFORMING IN A MEMBRANE REACTOR : AN ISOTHERMIC MODEL

    OpenAIRE

    ASSAF E.M.; JESUS C.D.F.; J.M. ASSAF

    1998-01-01

    A mathematical modelling of one-dimensional, stationary and isothermic membrane reactor for methane steam reforming was developed to compare the maximum yield for methane conversion in this reactor with that in a conventional fixed-bed reactor. Fick's first law was used to describe the mechanism of hydrogen permeation. The variables studied include: reaction temperature, hydrogen feed flow rate and membrane thickness. The results show that the membrane reactor presents a higher methane conver...

  5. Studies on modelling of bubble driven flows in chemical reactors

    Energy Technology Data Exchange (ETDEWEB)

    Grevskott, Sverre

    1997-12-31

    Multiphase reactors are widely used in the process industry, especially in the petrochemical industry. They very often are characterized by very good thermal control and high heat transfer coefficients against heating and cooling surfaces. This thesis first reviews recent advances in bubble column modelling, focusing on the fundamental flow equations, drag forces, transversal forces and added mass forces. The mathematical equations for the bubble column reactor are developed, using an Eulerian description for the continuous and dispersed phase in tensor notation. Conservation equations for mass, momentum, energy and chemical species are given, and the k-{epsilon} and Rice-Geary models for turbulence are described. The different algebraic solvers used in the model are described, as are relaxation procedures. Simulation results are presented and compared with experimental values. Attention is focused on the modelling of void fractions and gas velocities in the column. The energy conservation equation has been included in the bubble column model in order to model temperature distributions in a heated reactor. The conservation equation of chemical species has been included to simulate absorption of CO{sub 2}. Simulated axial and radial mass fraction profiles for CO{sub 2} in the gas phase are compared with measured values. Simulations of the dynamic behaviour of the column are also presented. 189 refs., 124 figs., 1 tab.

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

    NARCIS (Netherlands)

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

    2010-01-01

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

  7. Comparative study of pulsed and steady-state tokamak reactor burn cycles

    International Nuclear Information System (INIS)

    Four distinct operating modes have been proposed for tokamaks. Our study focuses on capital costs and lifetime limitations of reactor subsystems in an attempt to quantify sensitivity to pulsed operation. Major problem areas considered include: thermal fatigue on first wall, limiter/divertor; thermal energy storage; fatigue in pulsed poloidal field coils; out-of-plant fatigue and eddy current heating in toroidal field coils; electric power supply costs; and noninductive driver costs. We assume a high availability and low cost of energy will be mandatory for a commercial fusion reactor, and we characterize improvements in physics and engineering which will help achieve these goals for different burn cycles

  8. Studies of the restructuring of fast breeder test reactor fuel by out-of-pile simulation

    International Nuclear Information System (INIS)

    The fast breeder test reactor (FBTR) at Kalpakkam, India, currently employs a mixed carbide of uranium and plutonium with a Pu/(Pu + U) ratio of 0.70 as fuel. The behavior of this fuel in a thermal gradient is investigated. An out-of-pile simulation facility is designed, set up, and commissioned. Experiments are conducted on FBTR fuel pellets to study the restructuring of the fuel at various levels of linear power and its cracking behavior in a thermal gradient. The results are discussed in terms of their significance for reactor operation

  9. Mössbauer study of EUROFER and VVER steel reactor materials

    Science.gov (United States)

    Kuzmann, E.; Horváth, Á.; Alves, L.; Silva, J. F.; Gomes, U.; Souza, C.; Homonnay, Z.

    2013-04-01

    57Fe Mössbauer spectroscopy and X-ray diffractometry were used to study EUROFER or VVER ferritic reactor steels mechanically alloyed with TaC or NbC. Significant changes were found in the Mössbauer spectra and in the corresponding hyperfine field distributions between the ball milled pure steel and that alloyed with TaC or NbC. Spectral differences were also found in the case of use of same carbides with different origin, too. The observed spectral changes as an effect of ball milling of the reactor material steels with carbides can be associated with change in short range order of the constituents of steel.

  10. Entrained Flow Reactor Study of K-Capture by Solid Additives

    DEFF Research Database (Denmark)

    Wang, Guoliang; Jensen, Peter Arendt; Wu, Hao;

    2016-01-01

    A method to simulate the reaction between gaseous K-species and solid additives, at suspension fired conditions has been developed, using an entrained flow reactor (EFR). A water slurry containing solid additives (kaolin or coal fly ash) and KCl, is injected into the EFR and the solid products...... of additives, rose when increasing the molar ratio of K/(Al+Si) in the reactants. A change of the reaction temperature, from 1100 °C to 1450 °C, did not significantly influence the extent of the reaction, which is in contradiction to the trend observed in previous fixed-bed reactor studies. The method using...

  11. Reactor safety study. An assessment of accident risks in U. S. commercial nuclear power plants. Appendices III and IV. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    1975-10-01

    The items listed below summarize the detail sections which follow: a listing of definitions and a discussion of the general treatment of data within the random variable approach as utilized by the study; a tabulation of the assessed data base containing failure classifications, final assessed ranges utilized in quantification and reference source values considered in determining the ranges; a discussion of nuclear power plant experience that was used to validate the data assessment by testing its applicability as well as to check on the adequacy of the model to incorporate typical real incidents; an expanded presentation of the data assessment giving information on applicability considerations; a discussion of test and maintenance data including comparisons of models with experience data; and special topics, including assessments required for the initiating event probabilities and human error data and modeling.

  12. An integral CFD approach for the thermal simulation of the PBMR reactor unit

    OpenAIRE

    Kleingeld, Marius; Janse van Rensburg, Jacobus Johannes

    2011-01-01

    A CFD method was developed to conduct integral thermal reactor analysis for the complete Reactor Unit of the Pebble Bed Modular Reactor (Pty) Ltd (PBMR). The requirement was however also to include very detailed aspects such as leakage and bypass flow paths through the reflector blocks and sleeves. The aim was therefore to investigate the influence of leakage and bypass flow on the thermal performance of the Reactor Unit in an integral fashion. The focus of this paper is to discuss the method...

  13. Long-term studies in COD elimination and nitrification in an overcongested packed-bed reactor (biofilter); Langzeituntersuchungen zur CSB-Elimination und Nitrifikation in einem ueberstauten Festbettreaktor (Biofilter)

    Energy Technology Data Exchange (ETDEWEB)

    Engelhart, M.; Dichtl, N. [Technische Univ. Braunschweig (Germany). Inst. fuer Siedlungswasserwirtschaft

    1999-07-01

    On a semi-technical scale, two process combinations were tested for their suitability for COD elimination and nitrification in combination with an overcongested packed-bed reactor (biofilter). (orig.) [German] Im halbtechnischen Massstab wurden zwei Verfahrenskombinationen unter Einbeziehung eines ueberstauten Festbettreaktors (Biofilter) auf ihre Tauglichkeit zur CSB-Elimination und Nitrifikation untersucht. (orig.)

  14. Oxygen concentration distribution in an airlift loop reactor

    Institute of Scientific and Technical Information of China (English)

    李国庆; 杨守志; 蔡昭铃; 陈家镛

    1995-01-01

    Oxygen, concentration distributions of the liquid and gas phases along the axial direction of an airlift loop reactor have been calculated for various gas superficial velocities and oxygen consumption rates with water and CMC solutions respectively by applying the axial backmixing model to the riser and the downcomer and the complete mixing model for the separator. The results show that the dissolved oxygen concentration is zero at the bottom part of the downcomer when the rate of dissolved oxygen consumption by microorganisms is very high.

  15. Thermo-magnetic systems for space nuclear reactors an introduction

    CERN Document Server

    Maidana, Carlos O

    2014-01-01

    Introduces the reader to engineering magnetohydrodynamics applications and presents a comprehensive guide of how to approach different problems found in this multidisciplinary field. An introduction to engineering magnetohydrodynamics, this brief focuses heavily on the design of thermo-magnetic systems for liquid metals, with emphasis on the design of electromagnetic annular linear induction pumps for space nuclear reactors. Alloy systems that are liquid at room temperature have a high degree of thermal conductivity far superior to ordinary non-metallic liquids. This results in their use for

  16. System analysis study for Korean fusion DEMO reactor

    International Nuclear Information System (INIS)

    Highlights: ► A conceptual design study for a steady-state K-DEMO has been initiated. ► The major radius is designed to be below 6.5 m, considering engineering feasibilities. ► Magnetic field at the plasma center around 8 T is achieved by using Nb3Sn technology. ► Feasibility of near-future DEMO reactor is studied with a system analysis code. ► A net electric generation on the order of 300 MWe can be achieved below the βN of 5. -- Abstract: A conceptual design study for a steady-state Korean fusion DEMO reactor (K-DEMO) has been initiated. Two peculiar features need to be noted. First, the major radius is designed to be just below 6.5 m, considering practical engineering feasibilities. But still, high magnetic field at the plasma center around 8 T is expected to be achieved by using current state-of-the-art high performance Nb3Sn strand technology. Second, a two-stage development plan is being considered. In the first stage, K-DEMO will demonstrate a net electricity generation but will also act as a component test facility. Then, after a major upgrade, K-DEMO is expected to show a net electric generation on the order of 300 MWe and the competitiveness in cost of electricity (COE). Feasibility of such a practical, near-future demonstration reactor is studied in this paper, based on a zero dimensional system analysis code study. It was shown that a net electric generation on the order of 300 MWe can be achieved below the optimistic βN limit of 5. The elongation of K-DEMO is around 1.8 with single null configuration. Detailed optimization process and the resultant various plasma parameters are described

  17. Design study and R and D progress on Japan sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    This paper describes the progress of the design study and research and development (R and D) for the Japan Sodium-cooled Fast Reactor (JSFR) implemented in the 'Fast Reactor Cycle Technology Development (FaCT)' project. A sodium-cooled fast reactor with an electric power of 1,500 MWe is targeted for commercialization at around 2050, and a demonstration reactor assuming a power output from 500 to 750 MWe is planned to start operation at around 2025. R and D on innovative technologies to achieve economic competitiveness and enhance reliability and safety is carried out for the commercialization. A compact reactor vessel without a vessel wall cooling system is pursued in consideration of the wall thickness enough to resist the severest seismic condition. A two-loop cooling system with shortened high-chromium steel piping is a crucial feature, and studies on the hydraulics in the pipe elbow and the fabrication capability of the pipes are being carried out. A double-walled straight tube steam generator is investigated to enhance the reliability against sodium/water reaction, and developmental works are progressing, including the thermal-hydraulic design and trial manufacturing for components. Self-Actuated Shutdown System (SASS) is being developed with safety analysis of the applicability for JSFR and experimental demonstration in the experimental fast reactor JOYO. An advanced fuel handling system is pursued to enhance economic performance. In parallel with considering the necessity of studies on alternative technologies, discussion on whether the innovative technologies can be adopted for JSFR is in progress to be finalized in 2010. (author)

  18. Study of the thermal decomposition of petrochemical sludge in a pilot plant reactor

    OpenAIRE

    Conesa Ferrer, Juan Antonio; Moltó Berenguer, Julia; Ariza, José; Ariza, María; García Barneto, Agustín

    2014-01-01

    The pyrolysis of a sludge produced in the waste water treatment plant of an oil refinery was studied in a pilot plant reactor provided with a system for condensation of semivolatile matter. The study comprises experiments at 350, 400, 470 and 530 °C in nitrogen atmosphere. Analysis of all the products obtained (gases, liquids and chars) are presented, with a thermogravimetric study of the char produced and analysis of main components of the liquid. In the temperature range studied, the compos...

  19. Utilization of reactors overseas for study of radiation effects in materials

    International Nuclear Information System (INIS)

    This paper presents the current status of utilization of overseas reactors for study of radiation effects in materials as joint-use research at International Research Center for Nuclear Materials Science, Institute for Materials Research (IMR), Tohoku University (IMR-Oarai Center). Japan Materials Testing Reactor (JMTR) which had been mainly utilized for material irradiation at IMR-Oarai Center since 1970 was shut down in August 2006. Since then an overseas reactor, BR2 at SCK/CEN in Belgium, has been used in order to continue our irradiation programs and perform advanced material irradiation in collaboration with SCK/CEN for development of new irradiation techniques in BR2. In this paper, the features of irradiation facilities/devices in BR2 such as CALLISTO, MISTRAL, ROBIN and BAMI and the irradiation programs in 2006-2008 with use of each facility/device are described. (author)

  20. An in situ sample environment reaction cell for spatially resolved x-ray absorption spectroscopy studies of powders and small structured reactors

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Chu; Gustafson, Johan; Merte, Lindsay R.; Evertsson, Jonas [Division of Synchrotron Radiation Research, Lund University, Box 118, SE-221 00 Lund (Sweden); Norén, Katarina; Carlson, Stefan; Svensson, Håkan [MAX IV Laboratory, Lund University, Box 118, SE-221 00 Lund (Sweden); Carlsson, Per-Anders [Competence Centre for Catalysis, Chalmers University of Technology, SE-412 96 Göteborg (Sweden)

    2015-03-15

    An easy-to-use sample environment reaction cell for X-ray based in situ studies of powders and small structured samples, e.g., powder, pellet, and monolith catalysts, is described. The design of the cell allows for flexible use of appropriate X-ray transparent windows, shielding the sample from ambient conditions, such that incident X-ray energies as low as 3 keV can be used. Thus, in situ X-ray absorption spectroscopy (XAS) measurements in either transmission or fluorescence mode are facilitated. Total gas flows up to about 500 ml{sub n}/min can be fed while the sample temperature is accurately controlled (at least) in the range of 25–500 °C. The gas feed is composed by a versatile gas-mixing system and the effluent gas flow composition is monitored with mass spectrometry (MS). These systems are described briefly. Results from simultaneous XAS/MS measurements during oxidation of carbon monoxide over a 4% Pt/Al{sub 2}O{sub 3} powder catalyst are used to illustrate the system performance in terms of transmission XAS. Also, 2.2% Pd/Al{sub 2}O{sub 3} and 2% Ag − Al{sub 2}O{sub 3} powder catalysts have been used to demonstrate X-ray absorption near-edge structure (XANES) spectroscopy in fluorescence mode. Further, a 2% Pt/Al{sub 2}O{sub 3} monolith catalyst was used ex situ for transmission XANES. The reaction cell opens for facile studies of structure-function relationships for model as well as realistic catalysts both in the form of powders, small pellets, and coated or extruded monoliths at near realistic conditions. The applicability of the cell for X-ray diffraction measurements is discussed.

  1. Langmuir probe diagnostic studies of pulsed hydrogen plasmas in planar microwave reactors

    OpenAIRE

    A. Rousseau(MSSL, Surrey, United Kingdom); Teboul, E.; Lang, N.; M. Hannemann; Röpcke, J.

    2002-01-01

    Langmuir probe techniques have been used to study time and spatially resolved electron densities and electron temperatures in pulse-modulated hydrogen discharges in two different planar microwave reactors (fmicrowave= 2.45 GHz, tpulse= 1 ms). The reactors are (i) a standing-wave radiative slotted waveguide reactor and (ii) a modified travelling-wave radiative slotted waveguide reactor, which generate relatively large plasmas over areas from about 350 cm^2 to 500 cm^2. The plasma properties of...

  2. Study of emergency-action levels for light water reactors

    International Nuclear Information System (INIS)

    An emergency action level (EAL) is an observation or judgment that forms the basis for declaring an emergency status at a nuclear generating facility. There are four graded emergency category classifications which indicate an increasing potential for offsite radiological impact. Each emergency category is normally associated with an implementation procedure that outlines the preplanned actions that the emergency director will undertake. Thus a transient which causes a system or parameter to reach an EAL will also cause a transition of the normal station organization to an emergency organization. This transition will include an augmentation of the basic shift staff in order to support the corrective and mitigative actions of the nuclear reactor operators. In this regard, the major purpose of EALs is to provide an early indication of potential problems. Ideally, the ensuing response of the emergency organization will prevent a propagation of errors or failures that could result in serious consequences

  3. Hydrodynamic Studies on a Trickle Bed Reactor for Foaming Liquids

    Directory of Open Access Journals (Sweden)

    Ajay Bansal

    2010-10-01

    Full Text Available Hydrodynamic studies of trickle bed reactors (TBRs are essential for the design and prediction of their performance. The hydrodynamic characteristics involving pressure drop and dynamic liquid saturation are greatly affected by the physical properties of the liquids. In the present study experiments have been carried out in a concurrent downflow air - liquid trickle bed reactor to investigate the dynamic liquid saturation and pressure drop for the water (non-foaming and 3% polyethylene glycol and 4% polyethylene glycol foaming liquids in the gas continuous regime (GCF and foaming pulsing regime (FP. In the GCF regime the dynamic liquid saturation was found to increase with increase in liquid flow rate for non-foaming and foaming liquids. While for 3% and 4% polyethylene glycol solutions the severe foaming was observed in the high interaction regime and the regime is referred to as foaming pulsing (FP regime. The decrease in dynamic liquid saturation followed by a sharp rise in the pressure drop was observed during transition from gas GCF to FP regime. However in the FP regime, a dip in the dynamic liquid saturation was observed. The pressure drop for foaming liquids is observed to be manifolds higher compared to non-foaming liquid in the GCF regime. ©2010 BCREC UNDIP. All rights reserved(Received: 16th January 2010, Revised: 10th February 2010, Accepted: 21st Feberuary 2010[How to Cite: R. Gupta, A. Bansal. (2010. Hydrodynamic Studies on a Trickle Bed Reactor for Foaming Liquids. Bulletin of Chemical Reaction Engineering & Catalysis, 5 (1: 31-37. doi:10.9767/bcrec.5.1.775.31-37][How to Link / DOI: http://dx.doi.org/10.9767/bcrec.5.1.775.31-37 ][Cited by: Scopus 1 |

  4. CFD Simulation of an Anaerobic Membrane BioReactor (AnMBR to Treat Industrial Wastewater

    Directory of Open Access Journals (Sweden)

    Laura C. Zuluaga

    2015-06-01

    Full Text Available A Computational Fluid Dynamics (CFD simulation has been developed for an Anaerobic Membrane BioReactor (AnMBR to treat industrial wastewater. As the process consists of a side-stream MBR, two separate simulations were created: (i reactor and (ii membrane. Different cases were conducted for each one, so the surrounding temperature and the total suspended solids (TSS concentration were checked. For the reactor, the most important aspects to consider were the dead zones and the mixing, whereas for the ceramic membrane, it was the shear stress over the membrane surface. Results show that the reactor's mixing process was adequate and that the membrane presented higher shear stress in the 'triangular' channel.

  5. START/TM: a study of start-up and fractional power operation of tandem mirror fusion reactors

    International Nuclear Information System (INIS)

    Start-up, shutdown and fractional power operation are important parts of power reactor operation. Special requirements for operation during these phases often place design constraints on key subsystems and can influence the fundamental design approach. This report presents investigations of these problems for tandem mirror fusion reactors (TMR's) and is referred to as the START/TM study. As a basis for the work, the MARS conceptual tandem mirror reactor design is used as the general reactor model. An overall framework is developed for start-up and fractional power increases to full power, applicable to any fusion reactor. Five phases are identified that include initial commissioning, cold or hot shutdown, system testing and plasma initiation to a standby mode, staged power increases, and rated power operation. Both general and specific constraints associated with these phases are identified and a plasma shutdown scenario is developed

  6. Conceptual design study of a scyllac fusion test reactor

    International Nuclear Information System (INIS)

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

  7. Conceptual design study of a scyllac fusion test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Thomassen, K.I. (comp.)

    1975-07-01

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

  8. Performance evaluation of cigarette filter rods as a biofilm carrier in an anaerobic moving bed biofilm reactor.

    Science.gov (United States)

    Sabzali, Ahmad; Nikaeen, Mahnaz; Bina, Bijan

    2012-01-01

    Biocarriers are an important component of anaerobic moving bed biofilm reactors (AMBBRs). In this study, the capability of cigarette filter rods (CFRs) as a biocarrier in an anaerobic moving bed biofilm reactor was evaluated. Two similar lab-scale anaerobic moving bed biofilm reactors were undertaken using Kaldnes-K3 plastic media and cigarette filter rods (wasted filters from tobacco factories) as biofilm attachment media for wastewater treatment. Organic substance and total posphours (TP) removal was investigated over 100 days. Synthetic wastewater was prepared with ordinary water and glucose as the main sources of carbon and energy, plus balanced macro- and micro-nutrients. Process performance was studied by increasing the organic loading rate (OLR) in the range of 1.6-4.5 kg COD/m3 x d. The COD average removal efficiency were 61.3% and 64.5% for AMBBR with cigarette filter rods (Reactor A) and AMBBR with Kaldnes plastic media (Reactor B), respectively. The results demonstrate that the performance of the AMBBR containing 0.25 litres of cigarette filters was comparable with a similar reactor containing 1.5 litres of Kaldnes plastic media. An average phosphorus removal of 67.7% and 72.9% was achieved by Reactors A and B, respectively.

  9. Development of studies on helium cooled fast reactors

    International Nuclear Information System (INIS)

    A necessity is shown of developing breeders with high reproductive properties. Helium cooled fast reactor is considered. The reactor performances, heating circuit with the use of a steam turbine unit in the secondary circuit is outlined. The reactor design and fuel assemblies are described

  10. Ethanol fermentation in an immobilized cell reactor using Saccharomyces cerevisiae.

    Science.gov (United States)

    Najafpour, Ghasem; Younesi, Habibollah; Syahidah Ku Ismail, Ku

    2004-05-01

    Fermentation of sugar by Saccharomyces cerevisiae, for production of ethanol in an immobilized cell reactor (ICR) was successfully carried out to improve the performance of the fermentation process. The fermentation set-up was comprised of a column packed with beads of immobilized cells. The immobilization of S. cerevisiae was simply performed by the enriched cells cultured media harvested at exponential growth phase. The fixed cell loaded ICR was carried out at initial stage of operation and the cell was entrapped by calcium alginate. The production of ethanol was steady after 24 h of operation. The concentration of ethanol was affected by the media flow rates and residence time distribution from 2 to 7 h. In addition, batch fermentation was carried out with 50 g/l glucose concentration. Subsequently, the ethanol productions and the reactor productivities of batch fermentation and immobilized cells were compared. In batch fermentation, sugar consumption and ethanol production obtained were 99.6% and 12.5% v/v after 27 h while in the ICR, 88.2% and 16.7% v/v were obtained with 6 h retention time. Nearly 5% ethanol production was achieved with high glucose concentration (150 g/l) at 6 h retention time. A yield of 38% was obtained with 150 g/l glucose. The yield was improved approximately 27% on ICR and a 24 h fermentation time was reduced to 7 h. The cell growth rate was based on the Monod rate equation. The kinetic constants (K(s) and mu(m)) of batch fermentation were 2.3 g/l and 0.35 g/lh, respectively. The maximum yield of biomass on substrate (Y(X-S)) and the maximum yield of product on substrate (Y(P-S)) in batch fermentations were 50.8% and 31.2% respectively. Productivity of the ICR were 1.3, 2.3, and 2.8 g/lh for 25, 35, 50 g/l of glucose concentration, respectively. The productivity of ethanol in batch fermentation with 50 g/l glucose was calculated as 0.29 g/lh. Maximum production of ethanol in ICR when compared to batch reactor has shown to increase

  11. Temperature effects studies in light water reactor lattices

    International Nuclear Information System (INIS)

    The CREOLE experiments performed in the EOLE critical facility located in the Nuclear Center of CADARACHE - CEA (UO2 and UO2-PuO2 lattice reactivity temperature coefficient continuous measurements between 200C and 3000C; integral measurements by boron equivalent effect in the moderator; water density effects measurements with the use of over cladding aluminium tubes to remove moderator) allow to get an interesting and complete information on the temperature effects in the light water reactor lattices. A very elaborated calcurated scheme using the transport theory and the APOLLO cross sections library, has been developed. The analysed results of the whole lot of experiments show that the discrepancy between theory and experiment strongly depends on the temperature range and on the type of lattices considered. The error is mainly linked with the thermal spectrum effects. A study on the temperature coefficient sensitivity to the different cell neutron parameters has shown that only the shapes of the 235U and 238U thermal cross sections have enough weight and uncertainty margins to explain the observed experimental/calculation bias. Instead of arbitrarily fitting the identified wrong data on the calculation of the reactivity temperature coefficient we have defined a procedure of modification of the cross sections based on the consideration of the basic nuclear data: resonance parameters and associated statistic laws. The implementation of this procedure has led to propose new thermal cross sections sets for 235U and 238U consistent with the uncertainty margins associated with the previously accepted values and with some experimental data

  12. nuclear emergency management system case study:- Egypt's second research reactor

    International Nuclear Information System (INIS)

    the response to a radiological accident is basically the same as the response to any accident involving hazardous material. provisions should be developed to identify potential radiological hazard and inform the public and emergency workers of the action they should take. radiological emergency plans provide an efficient and effective response operation that, should an emergency occur, will protect the health and safety of workers, responders, the public, and the environment . one of the most important aspects of managing a nuclear emergency is the ability to promptly and adequately estimate the consequences of an accident. because of the need for protective actions to be initiated promptly in order to be effective, nuclear accident assessment must make use of all information that is available to on-site and of - site organizations. the work done in this paper describes the overall organization, including its relationship to the overall structure, and responsibility of all internal organizational elements with emergency responsibilities, practical guidance and tools for accident assessment and a basic assessment capability needed in the event of a serious reactor accident.An emergency management intelligent system (EMIS) is developed to provide assistance in the situation of radiological accident. The EMIS can be applied to a broad spectrum of accidents at Egypt's second research reactor. Complete data analysis is given in case of loss of coolant accident (LOCA) including dose assessment for the public

  13. Mathematical modelling of methane steam reforming in a membrane reactor: an isothermal model

    Energy Technology Data Exchange (ETDEWEB)

    Assaf, E.M. [Sao Paulo Univ., Sao Carlos, SP (Brazil). Dept. de Fisico-Quimica; Jesus, C.D.F.; Assaf, J.M. [Sao Carlos Univ., SP (Brazil). Dept. de Engenharia Quimica

    1998-06-01

    A mathematical modelling of one-dimensional, stationary and isothermic membrane reactor for methane steam reforming was developed to compare the maximum yield for methane conversion in this reactor with that in a conventional fixed-bed reactor. Fick`s first law was used to describe the mechanism of hydrogen permeation. The variables studied include: reaction temperature, hydrogen feed flow rate and membrane thickness. The results show that the membrane reactor presents a higher methane conversion yield than the conventional fixed-bed reactor. (author) 16 refs., 5 figs., 1 tab.; e-mail: eassaf at iqsc.sc.usp.br; mansur at power.ufscar.br

  14. MATHEMATICAL MODELLING OF METHANE STEAM REFORMING IN A MEMBRANE REACTOR: AN ISOTHERMIC MODEL

    Directory of Open Access Journals (Sweden)

    E.M. ASSAF

    1998-06-01

    Full Text Available A mathematical modelling of one-dimensional, stationary and isothermic membrane reactor for methane steam reforming was developed to compare the maximum yield for methane conversion in this reactor with that in a conventional fixed-bed reactor. Fick's first law was used to describe the mechanism of hydrogen permeation. The variables studied include: reaction temperature, hydrogen feed flow rate and membrane thickness. The results show that the membrane reactor presents a higher methane conversion yield than the conventional fixed-bed reactor.

  15. Shut-off rod performance study for the Point Lepreau reactor

    International Nuclear Information System (INIS)

    Shut-Off Rods (SORs) are strong neutron absorbing reactivity devices which are inserted into the Pt. Lepreau reactor core to terminate the nuclear chain reaction. SORs are inserted automatically by Shut-Down System One (SDS-1), in the event any of its trip parameters reach the trip set point on 2 out of the 3 electronic trip channels. SOR performance requirements for the Pt. Lepreau reactor have recently been reviewed and are reported here. In addition, the impact of rod unavailability and slow insertion characteristics on overall SDS-1 capability have been assessed. The shut-off rod system for the Pt. Lepreau reactor consists of 28 vertical rods which re suspended normally by an electromagnetic clutch assembly at the top fo the reactor just outside the core. Upon receipt of a trip signal, on two out of three SDS-1 trip channels, the electromagnetic clutches de-energize and allow the SORs to fall (spring assisted) into the core. In assessing the operational performance of the SOR system the two key parameters are the number of SORs available for insertion at any given time and the rate at which the rods are inserted when they fall. During the operation of the plant the number of rods available to be inserted may change due to routine maintenance and testing or due to system impairments. The rate at which the rods drop may also change due to mechanical wear or malfunction of the clutch or suspension assembly. The purpose of this study was to define the operational performance envelope which clearly identifies the SOR performance required to ensure a safe reactor shut down for all anticipated reactivity transients. Thus the main impetus of the study was to provide a clear technical basis for establishing operating performance requirements from which operating impairment guidelines could be derived. These guidelines are necessary to provide instructions for appropriate and timely action to the reactor operator in the event of SOR performance degradation

  16. An emergency water injection system (EWIS) for future CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Marques, Andre L.F. [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), SP (Brazil). E-mail: momarques@uol.com.br; Todreas, Neil E.; Driscoll, Michael J. [Massachusetts Inst.of Tech., Cambridge, MA (United States). Nuclear Engineering Dept.

    2000-07-01

    This paper deals with the investigation of the feasibility and effectiveness of water injection into the annulus between the calandria tubes and the pressure tubes of CANDU reactors. The purpose is to provide an efficient decay heat removal process that avoids permanent deformation of pressure tubes severe accident conditions, such as loss of coolant accident (LOCA). The water injection may present the benefit of cost reduction and better actuation of other related safety systems. The experimental work was conducted at the Massachusetts Institute of Technology (MIT), in a setup that simulated, as close as possible, a CANDU bundle annular configuration, with heat fluxes on the order of 90 kW/m{sup 2}: the inner cylinder simulates the pressure tube and the outer tube represents the calandria tube. The experimental matrix had three dimensions: power level, annulus water level and boundary conditions. The results achieved overall heat transfer coefficients (U), which are comparable to those required (for nominal accident progression) to avoid pressure tube permanent deformation, considering current CANDU reactor data. Nonetheless, future work should be carried out to investigate the fluid dynamics such as blowdown behavior, in the peak bundle, and the system lay-out inside the containment to provide fast water injection. (author)

  17. An Economic Analysis of Generation IV Small Modular Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Stewart, J S; Lamont, A D; Rothwell, G S; Smith, C F; Greenspan, E; Brown, N; Barak, A

    2002-03-01

    This report examines some conditions necessary for Generation IV Small Modular Reactors (SMRs) to be competitive in the world energy market. The key areas that make nuclear reactors an attractive choice for investors are reviewed, and a cost model based on the ideal conditions is developed. Recommendations are then made based on the output of the cost model and on conditions and tactics that have proven successful in other industries. The Encapsulated Nuclear Heat Source (ENHS), a specific SMR design concept, is used to develop the cost model and complete the analysis because information about the ENHS design is readily available from the University of California at Berkeley Nuclear Engineering Department. However, the cost model can be used to analyze any of the current SMR designs being considered. On the basis of our analysis, we determined that the nuclear power industry can benefit from and SMRs can become competitive in the world energy market if a combination of standardization and simplification of orders, configuration, and production are implemented. This would require wholesale changes in the way SMRs are produced, manufactured and regulated, but nothing that other industries have not implemented and proven successful.

  18. Solid state laser driver for an ICF reactor

    Energy Technology Data Exchange (ETDEWEB)

    Krupke, W.F.

    1988-01-01

    A conceptual design is presented of the main power amplifier of a multi-beamline, multi-megawatt solid state ICF reactor driver. Simultaneous achievement of useful beam quality and high average power is achieved by a proper choice of amplifier geometry. An amplifier beamline consists of a sequence of face-pumped rectangular slab gain elements, oriented at the Brewster angle relative to the beamline axis, and cooled on their large faces by helium gas that is flowing subsonically. The infrared amplifier output radiation is shifted to an appropriately short wavelength (<500 nm) using nonlinear crystals that are also gas cooled. We project an overall driver efficiency >10% (including all flow cooling input power) when the amplifiers are pumped by efficient high-power AlGaAs semiconductor laser diode arrays. 11 refs., 3 figs., 7 tabs.

  19. Joint KAERI/VAEC pre-possibility study on a new research reactor for Vietnam

    Energy Technology Data Exchange (ETDEWEB)

    Park, Cheol; Lee, B. C.; Chae, H. T.; Kim, H.; Lee, C. S.; Choi, C. O.; Jun, B. J. [KAERI, Taejon (Korea, Republic of); Vien, Luong Ba; Dien, Nguyen Nhi [Vietnam Atomic Energy Commission, Hanoi (Viet Nam)

    2004-05-01

    Based on the agreement on the technical cooperation for nuclear technology between Korea and Vietnam, a KAERI/VAEC joint study on the pre-possibility of a new research reactor for Vietnam has been carried out in the research reactor area from Nov. 2003 to May 2004. In this report, the results of the pre-possibility study on a new research reactor are described. The report presents the necessity of a new research reactor in Vietnam, and the desired performance requirements of the new research reactor if necessary. The major design characteristics of some existing research reactors and those under planning were also reviewed and the main characteristics which should be considered in selecting a new multipurpose research reactor for Vietnam were drawn. Some recommendations on the considerations for the next step of the feasibility study such as the project formulation, manpower requirements and international co-operation were also briefly touched upon.

  20. Joint KAERI/VAEC pre-possibility study on a new research reactor for Vietnam

    International Nuclear Information System (INIS)

    Based on the agreement on the technical cooperation for nuclear technology between Korea and Vietnam, a KAERI/VAEC joint study on the pre-possibility of a new research reactor for Vietnam has been carried out in the research reactor area from Nov. 2003 to May 2004. In this report, the results of the pre-possibility study on a new research reactor are described. The report presents the necessity of a new research reactor in Vietnam, and the desired performance requirements of the new research reactor if necessary. The major design characteristics of some existing research reactors and those under planning were also reviewed and the main characteristics which should be considered in selecting a new multipurpose research reactor for Vietnam were drawn. Some recommendations on the considerations for the next step of the feasibility study such as the project formulation, manpower requirements and international co-operation were also briefly touched upon

  1. Study and Evaluation of Innovative Fuel Handling Systems for Sodium-Cooled Fast Reactors: Fuel Handling Route Optimization

    OpenAIRE

    Franck Dechelette; Franck Morin; Guy Laffont; Gilles Rodriguez; Emmanuel Sanseigne; Sébastien Christin; Xavier Mognot; Aurélien Morcillo

    2014-01-01

    International audience The research for technological improvement and innovation in sodium-cooled fast reactor is a matter of concern in fuel handling systems in a view to perform a better load factor of the reactor thanks to a quicker fuelling/defueling process. An optimized fuel handling route will also limit its investment cost. In that field, CEA has engaged some innovation study either of complete FHR or on the optimization of some specific components. This paper presents the study of...

  2. A study of temperature sensor location based on fractal analysis for cascade control schemes in tubular reactors

    DEFF Research Database (Denmark)

    Eduardo Ramirez-Castelan, Carlos; Moguel-Castañeda, Jazael; Puebla, Hector;

    2016-01-01

    Temperature sensor location for cascade control schemes in tubular reactors is still an open research problem. Several studies have pointed out that most temperature sensitive zones along the length of the reactor are suitable to this end. In this work, we have studied the problem of sensor...... location in a cascade control configuration using fractal analysis of time series obtained by random forcing of the jacket rector. A benchmark dispersion axial model displaying different temperature profiles is used to illustrate our findings....

  3. Study meeting on 'criteria for materials of nuclear fusion reactors'

    International Nuclear Information System (INIS)

    This study meeting was held on March 1 and 2, 1984, at the Institute of Plasma Physics, Nagoya University. Recently, the problems required for the materials of nuclear fusion reactors have become considerably clear. The problem of the high concentration damage due to 14 MeV neutrons and the problem of surface materials have been well known from the beginning, but moreover, the radioactivation of materials, the problem of safety, and the feasibility of remote operation related to it have become urgent problems. Besides, the plan of large scale facilities as the means of promoting research is one of the important themes. The research on materials must take part in the whole technological problems which enable the construction of actual nuclear fusion devices. This study meeting was held as a part of the R project of the Institute of Plasma Physics, Nagoya University, but it aimed at grasping the present status and discussing the future perspective of the materials of nuclear fusion reactors, and examining the criteria for nuclear fusion materials. The gists of 23 lectures presented at the meeting are collected in this report. (Kako, I.)

  4. Treatment of domestic wastewater in an up-flow anaerobic sludge blanket reactor followed by moving bed biofilm reactor

    OpenAIRE

    Tawfik, A.; El-Gohary, F.; Temmink, B.G.

    2010-01-01

    The performance of a laboratory-scale sewage treatment system composed of an up-flow anaerobic sludge blanket (UASB) reactor and a moving bed biofilm reactor (MBBR) at a temperature of (22-35 A degrees C) was evaluated. The entire treatment system was operated at different hydraulic retention times (HRT's) of 13.3, 10 and 5.0 h. An overall reduction of 80-86% for CODtotal; 51-73% for CODcolloidal and 20-55% for CODsoluble was found at a total HRT of 5-10 h, respectively. By prolonging the HRT...

  5. Status of reactor pressure vessel embrittlement study in Japan

    International Nuclear Information System (INIS)

    Since the construction of Japanese first commercial nuclear power plant in 1966, 52 nuclear power plants have been commissioned in Japan to commercial operation. Japanese first nuclear power plant has now been service for 30 years and the aging of nuclear power plants is steadily progressing in general. Under these circumstances, the Japan Power Engineering and Inspection Corporation (JAPEIC) is executing, under consignment by the Ministry of International Trade and Industry (MITI), the development and verification test programs for plant integrity evaluation technology by which nuclear power plant aging can be appropriately handled. This paper shows the outline of study dealing with embrittlement of RPV caused by neutron irradiation, as one of the activity of JAPEIC. The embrittlement of RPV caused by neutron irradiation is manifested as a shift of transition temperature and as a reduction in Upper Shelf Energy (USE). In JAPEIC, the study dealing with a shift of transition temperature was conducted in the ''Reactor Pressure Vessel Pressurized Thermal Shock Test Project (the PTS Project)'', and the study dealing with a reduction in USE has been conducted in the ''Nuclear Power Plant Life Management Technology (the PLIM Project)''. And the reconstitution technology of surveillance test specimen has been conducted in PLIM Project as one of the measures to improve monitoring above material characteristic changes. The integrity evaluation under the Pressurized Thermal Shock (PTS) events including the effect of neutron irradiation embrittlement was initiated in 1983 FY as the PTS Project and was completed in the 1991 FY. The study verified that plant integrity could be assured at not only the end of design life, but also an extended service life even when the severest PTS events were postulated. The PLIM Project, designed to develop and verify the integrity evaluation technology dealing with reduction of USE by neutron irradiation, was started in the 1996 FY as a 10

  6. A study on the irradiation embrittlement and recovery characteristics of light water reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Chi, Se Hwan; Hong, Jun Hwa; Lee, Bong Sang; Oh, Jong Myung; Song, Sook Hyang; Milan, Brumovsky [NRI Czech (Czech Republic)

    1999-03-01

    The neutron irradiation embrittlement phenomenon of light water RPV steels greatly affects the life span for safe operation of a reactor. Reliable evaluation and prediction of the embrittlement of RPV steels, especially of aged reactors, are of importance to the safe operation of a reactor. In addition, the thermal recovery of embrittled RPV has been recognized as an option for life extension. This study aimed to tracer/refine available technologies for embrittlement characterization and prediction, to prepare relevant materials for several domestic RPV steels of the embrittlement and recovery, and to find out possible remedy for steel property betterment. Small specimen test techniques, magnetic measurement techniques, and the Meechan and Brinkmann's recovery curve analysis method were examined/applied as the evaluation techniques. Results revealed a high irradiation sensitivity in YG 3 RPV steel. Further extended study may be urgently needed. Both the small specimen test technique for the direct determination of fracture toughness, and the magnetic measurement technique for embrittlement evaluation appeared to be continued for the technical improvement and data base preparation. Manufacturing process relevant to the heat treatment appeared to be improved in lowering the irradiation sensitivity of the steel. Further study is needed especially in applying the present techniques to the new structural materials under new irradiation environment of advanced reactors. (author)

  7. A study on the irradiation embrittlement and recovery characteristics of light water reactor pressure vessel steels

    International Nuclear Information System (INIS)

    The neutron irradiation embrittlement phenomenon of light water RPV steels greatly affects the life span for safe operation of a reactor. Reliable evaluation and prediction of the embrittlement of RPV steels, especially of aged reactors, are of importance to the safe operation of a reactor. In addition, the thermal recovery of embrittled RPV has been recognized as an option for life extension. This study aimed to tracer/refine available technologies for embrittlement characterization and prediction, to prepare relevant materials for several domestic RPV steels of the embrittlement and recovery, and to find out possible remedy for steel property betterment. Small specimen test techniques, magnetic measurement techniques, and the Meechan and Brinkmann's recovery curve analysis method were examined/applied as the evaluation techniques. Results revealed a high irradiation sensitivity in YG 3 RPV steel. Further extended study may be urgently needed. Both the small specimen test technique for the direct determination of fracture toughness, and the magnetic measurement technique for embrittlement evaluation appeared to be continued for the technical improvement and data base preparation. Manufacturing process relevant to the heat treatment appeared to be improved in lowering the irradiation sensitivity of the steel. Further study is needed especially in applying the present techniques to the new structural materials under new irradiation environment of advanced reactors. (author)

  8. Biological treatment of phenolic wastewater in an anaerobic continuous stirred tank reactor

    Directory of Open Access Journals (Sweden)

    Firozjaee Taghizade Tahere

    2013-01-01

    Full Text Available In the present study, an anaerobic continuous stirred tank reactor (ACSTR with consortium of mixed culture was operated continuously for a period of 110 days. The experiments were performed with three different hydraulic retention times and by varying initial phenol concentrations between 100 to 1000 mg/L. A maximum phenol removal was observed at a hydraulic retention time (HRT of 4 days, with an organic loading rate (OLR of 170.86 mg/L.d. At this condition, phenol removal rate of 89% was achieved. In addition, the chemical oxygen demand (COD removal corresponds to phenol removal. Additional operating parameters such as pH, MLSS and biogas production rate of the effluents were also measured. The present study provides valuable information to design an anaerobic ACSTR reactor for the biodegradation of phenolic wastewater.

  9. A preliminary study on the activation analysis for a conceptual K DEMO fusion reactor

    International Nuclear Information System (INIS)

    It is known that the radioactive materials of nuclear fusion reactor have a low radioactivity in comparison with fission reactors. However, in the case of a specific module which is directly facing the fusion plasma, its radioactivity is so high that the module is needed to control and decommission by the legal regulation. The typical modules which have such a characteristic are Blanket Module (BM) and Divertor (Dv). These modules are regularly substituted with new module when irradiated sufficiently, so an accurate assessment of neutron activation is mandatory for those components of the fusion reactors because of the environmental effect and safety advantages. In order to reduce the neutron activation effect, two primary considerations are required. One is the use of the materials which abate the radioactive effects, and the other is the development of effective designs for low neutron activation. Various research groups of the ITER member countries have been performing the researches to satisfy these requirements. As part of the international research trends on neutron activation fields, the Korean research group has been designing the demonstration fusion reactor, which is named K DEMO. In this study, the activation calculations of K DEMO were carried out, and then those calculation results were compared with the calculation results of ITER model. In addition to two main modules, the activation calculations for Vacuum Vessel (VV) were performed, because that component represents the containment integrity of the fusion reactor. Neutron flux distributions in the fusion reactor were provided by a MCNP calculation. The activation calculations were performed by FISPACT 2007 code. The calculated fluxes were employed to FISPACT for the activation calculation

  10. On an optimized neutron shielding for an advanced molten salt fast reactor design

    International Nuclear Information System (INIS)

    The molten salt reactor technology has gained renewed interest. In contrast to the historic molten salt reactors, the current projects are based on designing a molten salt fast reactor. Thus the shielding becomes significantly more challenging than in historic concepts. One very interesting and innovative result of the most recent EURATOM project on molten salt reactors – EVOL – is the fluid flow optimized design of the inner core vessel using curved blanket walls. The developed structure leads to a very uniform flow distribution. The design avoids all core internal structures. On the basis of this new geometry a model for neutron physics calculation is presented and applied for a shielding optimization. Based on these results an optimized shielding strategy is developed for the molten salt fast reactor to keep the fluence in the safety related outer vessel below expected limit values. A lifetime of 80 years can be assured, but the size of the core/blanket system has to be significantly increased and will finally be comparable to a sodium cooled fast reactor. The HELIOS results are verified against Monte-Carlo calculations with very satisfactory agreement for a deep penetration problem. (author)

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

    Science.gov (United States)

    Sekimoto, Hiroshi

    2014-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-09-30

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

  13. Introduction and synopsis of the TITAN reversed-field-pinch fusion-reactor study

    International Nuclear Information System (INIS)

    The TITAN reversed-field-pinch (RFP) fusion-reactor study aims to determine the technical feasibility and key developmental issues for an RFP fusion reactor operating at high power density; and to determine the potential economic, operational, safety and environmental features of high mass-power-density fusion-reactor systems. Mass power density (MPD) is defined as the ratio of net electric output to the mass of the fusion power core (FPC). The FPC includes the plasma chamber, first wall, blanket, shield, magnets, and related structure. Two different detailed designs TITAN-I and TITAN-II, have been produced. TITAN-I is a self-cooled lithium design with a vanadium-alloy structure. TITAN-II is a self-cooled aqueous loop-in-pool design with 9-C ferritic steel as the structural material. Both designs use RFP plasmas operating with essentially the same parameters. Both conceptual reactors are based on the DT fuel cycle, have a net electric output of about 1000 MWe, are compact, and have a high MPD of 800 kWe per tonne of FPC. The inherent physical characteristics of the RFP confinement concept make possible compact fusion reactors with such a high MPD. The TITAN designs would meet the U.S. criteria for the near-surface disposal of radioactive waste (Class C, 10CFR61) and would achieve a high Level of Safety Assurance with respect to FPC damage by decay afterheat and radioactivity release caused by accidents. Very importantly, a 'single-piece' FPC maintenance appears feasible for both designs. The design window for such compact RFP reactors would include machines with neutron wall loadings in the range of 10-20 MW/m2 with a shallow minimum COE at about 18 MW/m2. Even though operation at the lower end of this range of wall loading (10-12 MW/m2) is possible, and may be preferable, the TITAN study adopted the design point at the upper end (18 MW/m2) in order to quantify and assess the technical feasibility and physics limits for such high-MPD reactors. From this work, key

  14. Neutronic study of a nuclear reactor of fused salts; Estudio neutronico de un reactor nuclear de sales fundidas

    Energy Technology Data Exchange (ETDEWEB)

    Garcia B, F. B.; Francois L, J. L., E-mail: faviolabelen@gmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)

    2012-10-15

    The reactors of fused salts called Molten Salt Reactor have presented a resurgence of interest in the last decade, due to they have a versatility in particular to operate, either with a thermal or fast neutrons spectrum. The most active development was by the middle of 1950 and principles of 1970 in the Oak Ridge National Laboratory. In this work some developed models are presented particularly and studied with the help of the MCNPX code, for the development of the neutronic study of this reactor, starting of proposed models and from a simple and homogeneous geometry until other more complex models and approximate to more real cases. In particular the geometry conditions and criticality of each model were analyzed, the isotopic balance, as well as the concentrations of the salts and different assigned fuel types. (Author)

  15. Hydroxylamine as a potential reagent for dissolution off gas scrubbing in spent fuel reprocessing: kinetics of the iodine reduction. An example of similarity between the studies on the chemistry of iodine in reactor safety and in spent fuel reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Cau Dit Coumes, C.; Devisme, F. [Commissariat a l`Energie Atomique, CE/VRH, Bagnols-sur-Ceze (France); Vargas, S.; Chopin-Dumas, J. [Laboratoire d`Electrochimie Inorganique, ENSSPICAM, Marseille (France)

    1996-12-01

    Iodine, which can be released inside the containment building when an accident occurs, can be traced, in normal operating conditions, at the back end of the fuel cycle. Hydroxylamine has been selected as a reagent of potential interest to trap iodine in the dissolution off gas treatment. The kinetics of the reaction between hydroxylamine and iodine has been studied in a narrow range of pH (1-2), with hydroxylamine in excess (ratios of hydroxylamine to iodine initial concentrations varying from 2 to 40), at constant temperature (30{sup o}C) and ionic strength (0.1 mol/L). Spectrophotometry and voltametry have been coupled for analytical investigation. The problem of the rapid mixing of the reactants has been solved using a continuous reactor. Triiodide has been shown non reactive towards hydroxylamine. An initial rate law has been proposed, pointing out the first order of the reaction with respect to hydroxylamine and iodine, and the inhibitory effect of iodide and hydrogen ions. Nitrous acid has been identified as a transitory product. Nitrous oxide and nitrogen monoxide have been detected by gas chromatography, the ratio of the amounts of products formed depending on acidity. The complexity of the overall reaction has been ascribed to the competition of four reactions as previously proposed in the literature. (author) 8 figs., 1 tab., 13 refs.

  16. Hydroxylamine as a potential reagent for dissolution off gas scrubbing in spent fuel reprocessing: kinetics of the iodine reduction. An example of similarity between the studies on the chemistry of iodine in reactor safety and in spent fuel reprocessing

    International Nuclear Information System (INIS)

    Iodine, which can be released inside the containment building when an accident occurs, can be traced, in normal operating conditions, at the back end of the fuel cycle. Hydroxylamine has been selected as a reagent of potential interest to trap iodine in the dissolution off gas treatment. The kinetics of the reaction between hydroxylamine and iodine has been studied in a narrow range of pH (1-2), with hydroxylamine in excess (ratios of hydroxylamine to iodine initial concentrations varying from 2 to 40), at constant temperature (30oC) and ionic strength (0.1 mol/L). Spectrophotometry and voltametry have been coupled for analytical investigation. The problem of the rapid mixing of the reactants has been solved using a continuous reactor. Triiodide has been shown non reactive towards hydroxylamine. An initial rate law has been proposed, pointing out the first order of the reaction with respect to hydroxylamine and iodine, and the inhibitory effect of iodide and hydrogen ions. Nitrous acid has been identified as a transitory product. Nitrous oxide and nitrogen monoxide have been detected by gas chromatography, the ratio of the amounts of products formed depending on acidity. The complexity of the overall reaction has been ascribed to the competition of four reactions as previously proposed in the literature. (author) 8 figs., 1 tab., 13 refs

  17. Sulfide-oxidizing bacteria establishment in an innovative microaerobic reactor with an internal silicone membrane for sulfur recovery from wastewater.

    Science.gov (United States)

    Valdés, F; Camiloti, P R; Rodriguez, R P; Delforno, T P; Carrillo-Reyes, J; Zaiat, M; Jeison, D

    2016-06-01

    A novel bioreactor, employing a silicone membrane for microaeration, was studied for partial sulfide oxidation to elemental sulfur. The objective of this study was to assess the feasibility of using an internal silicone membrane reactor (ISMR) to treat dissolved sulfide and to characterize its microbial community. The ISMR is an effective system to eliminate sulfide produced in anaerobic reactors. Sulfide removal efficiencies reached 96 % in a combined anaerobic/microaerobic reactor and significant sulfate production did not occur. The oxygen transfer was strongly influenced by air pressure and flow. Pyrosequencing analysis indicated various sulfide-oxidizing bacteria (SOB) affiliated to the species Acidithiobacillus thiooxidans, Sulfuricurvum kujiense and Pseudomonas stutzeri attached to the membrane and also indicated similarity between the biomass deposited on the membrane wall and the biomass drawn from the material support, supported the establishment of SOB in an anaerobic sludge under microaerobic conditions. Furthermore, these results showed that the reactor configuration can develop SOB under microaerobic conditions and can improve and reestablish the sulfide conversion to elemental sulfur. PMID:27003697

  18. PLATO [Program for Load And Transient Operation]: An operator tool for load transients in PWRs [pressurized water reactors

    International Nuclear Information System (INIS)

    The axial stability of PWRs [pressurized water reactors] is rather poor. Power transients can introduce large oscillations which are sometimes difficult to handle. An axial model has been developed to study the controllability of a PWR in different power transients. After investigation with the XENOPT code, a new code PLATO [Program for Load And Transient Operation] with a user friendly interface and life reactor data, has been successfully implemented at the Doel power station

  19. BOR-60 reactor as an instrument for experimental substantiation of fuel rods for advanced NPPs

    International Nuclear Information System (INIS)

    . The principle task was to provide the required temperature conditions on specimens. This was achieved through the use of the thermal insulation gaps, intense cooling or additional heating at the expense of radiation energy release or fuel fission. As a result the reactor is used for testing various advanced types of fuel and structural materials at high thermal loads (100kW/m), temperatures (100 deg. C), burnups (33% h.a.) and fluences (1.8·1023cm-2 with E>0.1MeV). In case of necessity, the temperature can be stabilized by changing the thermal resistance in the heat transfer or heat removal intensification scheme using the liquid metal kept in the boiling condition. These units also provide the specified height and azimuthal temperature nonuniformity. The experimental facilities can be used for testing the fuel rods of up to 15 mm in diameter placed in different grids (triangular, square etc.) and environments (sodium, lithium, lead, various gases etc.). Due to the availability of the experimental facility for reprocessing of irradiated fuel and production of fast reactor fuel and fuel rods, the reactor is used for the experiments related the closed fuel cycle, such as testing of refabricated fuel with involvement of minor actinides and long-lived fission products into the fuel cycle. The BOR-60 demonstrated an effective operation as a MA burner as well as a power and weapon grade burner. This allows us to solve the important tasks of the nuclear power engineering, in particular to reduce the fuel cost and the quantity of radioactive waste and improve the environmental situation. The BOR-60 reactor has great experience in irradiation of oxide, metal, ceramic, carbide and nitride fuel compositions for reactors of different purposes, in particular for fast sodium reactors. Such fuel properties as regularities of gas release, shape change and structure formation were studied. The results obtained allowed us to substantiate the use of fuel rods for the BN-350 and BN-600

  20. FAFTRCS: an experiment in computerized reactor safety systems

    International Nuclear Information System (INIS)

    Nuclear Power Plant availability and reliability could be improved by the integration of computers into the control environment. However, computer-based systems are historically viewed as being unreliable. This places a burden upon the designer to demonstrate adequate reliability and availability for the computer. The complexity associated with computers coupled with the manual nature of these demonstrations results in a high cost which typically has been justified for critical applications only. This paper investigates a methodology for automating this process and discusses a project which intends to apply this methodology to design verification and validation for a control system which will be installed and tested in an actual reactor control environment. 7 refs., 4 figs., 1 tab

  1. Energy recovery from municipal solid waste in an anaerobic reactor.

    Science.gov (United States)

    Jeyapriya, S P; Saseetharan, M K

    2008-07-01

    Anaerobic digestion of municipal solid waste was carried out in the laboratory at room temperature to assess the bio-energy production from municipal solid waste (MSW) with high total solids content. The total biogas production from the municipal garbage was found to be 3.2 L in 120 days. The results from the biomethanation process showed that an increase in gas production was observed with increase in digestion period when the bioconversion parameters were found to be favorable for the production of gas. Changes in the parameters, such as pH, affected the production of gas significantly. Samples taken from the reactor at definite interval of time during the degradation process showed considerable reduction in total volatile solids, total carbon, total nitrogen and COD, etc. indicating the waste stabilization. PMID:19552079

  2. Mixing Characteristics and Bubble Behavior in an Airlift Internal Loop Reactor with Low Aspect Ratio

    Institute of Scientific and Technical Information of China (English)

    张伟鹏; 雍玉梅; 张广积; 杨超; 毛在砂

    2014-01-01

    The present study summarizes the results of macro-and micro-mixing characteristics in an airlift inter-nal loop reactor with low aspect ratio (H/D≤5) using the electrolytic tracer response technique and the method of parallel competing reactions respectively. The micro-mixing has never been investigated in airlift loop reactors. The dual-tip electrical conductivity probe technique is used for measurement of local bubble behavior in the reactor. The effects of several operating parameters and geometric variables are investigated. It is found that the increase in su-perficial gas velocity corresponds to the increase in energy input, liquid circulation velocity and shear rate, decreas-ing the macro-mixing time and segregation index. Moreover, it is shown that top clearance and draft diameter affect flow resistance. However, the bubble redistribution with a screen mesh on the perforated plate distributor for macro-mixing is insignificant. The top region with a high energy dissipation rate is a suitable location for feeding reactants. The analysis of present experimental data provides a valuable insight into the interaction between gas and liquid phases for mixing and improves the understanding of intrinsic roles of hydrodynamics upon the reactor de-sign and operating parameter selection.

  3. Experimental and modeling study of sulfur dioxide oxidation in packed-bed tubular reactor

    Directory of Open Access Journals (Sweden)

    Hanen NOURI

    2013-08-01

    Full Text Available The conversion of sulfur dioxide into sulfur trioxide is a reaction which interests not only the industry of sulfuric acid production but also the processes of pollution control of certain gas effluents containing SO2. This exothermic reaction needs a very good control of temperature, that's why it is led in the industry in a multistage converter with intermediate heat exchangers. Microreactors represent a good alternative for such reaction due to their intensification of mass and heat transfer and enhancement of temperature control. In this study, this reaction was conducted in a stainless steel tubular (4mm ID packed bed reactor using particles of vanadium pentoxide as catalyst at atmospheric pressure. Experiments were performed with different inlet SO2 concentration in 3-9% range and reaction temperature between 685-833K. We noticed that the conversion decreases with the amount of SO2 and increases with the temperature until an optimum, above this value the conversion drop according to the shape of the equilibrium curve. Controlling rate mechanism is studied by varying temperature. Pseudohomogeneous perfect plug flow is used to describe this small tubular reactor. Numerical simulations with MATLAB were performed to validate the experimental results. Good agreement between the model predictions and the experimental results is achieved. Fluid flow description inside the packed bed reactor was performed by using the free fluid and porous media flow model. This model was solved by the commercial software COMSOL Multiphysics. Velocity profile inside the reactor is theoretically obtained.

  4. Study on radioactive waste management scenarios in regular maintenance of a fusion reactor

    International Nuclear Information System (INIS)

    Low-level radioactive waste is generated in large amounts in the operation of a fusion reactor. For this reason, there are needs for the study of radioactive waste management scenarios, as well as the clarification of the function of waste handling facilities in the design phase. This paper describes the management scenarios with a focus on the radioactive waste generated at the time of scheduled maintenance of a nuclear fusion prototype reactor. Based on the temporal change of the residual heat and dose rate of the blanket and diverter, as the furnace equipment associated with induced radioactivity, management period was determined. At this time, the attenuation rate of dose rate and the like of each device are different. So, if maintenance cycle is established for each device and thus storage area is minimized, the control area can be optimized. Based on the 'principle for minimizing radioactive waste,' the reuse of devices is effective in reducing waste. So, in view of a commercial reactor, research and development is required for the establishment of reuse process under high-dose. Since the commitment to radioactive waste is considered to be an important factor in the future for the social acceptance of nuclear fusion reactor development, comprehensive study including the disposal of waste and the reuse of equipment is important. (A.O.)

  5. Studies of fragileness in steels of vessels of BWR reactors; Estudios de fragilizacion en aceros de vasija de reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Robles, E.F.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2003-07-01

    The structural materials with those that are manufactured the pressure vessels of the BWR reactors, suffer degradation in its mechanical properties mainly to the damage taken place by the fast neutrons (E > 1 MeV) coming from the reactor core. Its are experimentally studied those mechanisms of neutron damage in this material type, by means of the irradiation of steel vessel in experimental reactors to age them quickly. Alternatively it is simulated the neutron damage by means of irradiation of steel with heavy ions. In this work those are shown first results of the damage induced by irradiation from a similar steel to the vessel of a BWR reactor. The irradiation was carried out with fast neutrons (E > 1 MeV, fluence of 1.45 x 10{sup 18} n/cm{sup 2}) in the TRIGA MARK lll reactor and separately with Ni{sup +3} ions in a Tandetrom accelerator, E = 4.8 MeV and range of the ionic flow of 0.1 to 53 iones/A{sup 2}. (Author)

  6. Nuclear Properties of a Reactor Used for Nuclear Seeding of an MHD Plasma

    International Nuclear Information System (INIS)

    A study has been carried out to examine a number of engineering and reactor parameters associated with generation of ionization in a gas as it passes through a nuclear reactor. The reactor will also act as the supply of heat to raise the temperature and pressure of the gas, although not to such a degree as to cause thermal ionization, and the heat from fission will provide the motive force. The main results establish the interrelationships between reactor power density, production of ionization, rate of consumption of the seed material, rate of temperature rise, and general characteristics of a nuclear reactor when used in such an application. The materials studied were a combination of uranium-dioxide and graphite as a moderator, both chosen because of their excellent high-temperature characteristics. Helium was used as the coolant, and it was enriched in 3He. The geometry selected was that of uranium rods surrounded by a small gas- filled coolant gap spaced in a rectangular array in a medium of graphite. The parameters used, such as the enrichment of the helium gas in 3He, spacing between fuel rods and the diameter of fuel rods, and the gas channel dimensions, were all selected in a somewhat arbitrary, fashion. But their rearrangement to somewhat more desirable values, say from the standpoint of heat transfer, does not significantly affect the overall results obtained in this study. Effective densities, temperatures and pressures were evaluated on the basis of a set of expected operating conditions. Two-group reactor theory was used to evaluate the criticality of this system, and the same model was used to evaluate the effect on multiplication or stability of a potential loss in 3He coolant-poison. This same two-group model also yields information about both the flux levels and the neutron spectra which one can expect in such a system. Using the currently obtainable power densities, calculations were carried out to determine the density of ionization that one can

  7. An in-depth study on oscillation peculiarities by step-input of reactivity and frequency of helium fan's transducer of 10 MW high temperature gas-cooled reactor and its control strategy

    International Nuclear Information System (INIS)

    In order to know more deeply about dynamic characteristics of 10 MW High Temperature Gas-Cooled Reactor (HTR-10), the dynamic math model of HTR-10 is established. On the basis of this model, the paper simulates and studies on the main dynamic processes of HTR-10, especially the oscillation peculiarities that are very different from PWR. The conclusions are: the main reason of oscillation caused by step reactivity disturbing is that the special structure of HTR-10's fuel component causes fuel granule's temperature to change slower than nuclear power; the main reasons of oscillation caused by step-input frequency disturbing of helium fan's transducer are the structure, flux and pressure peculiarities of HTR-10 primary loop, and synchronous vibration between HTR-10 primary and secondary loop. In order to make reactor running smoother and safety, the multi-layer harmony control strategy is adopted. This paper also discusses the design outlines of the under layer controllers. (author)

  8. Study of gas-solid contact in an ultra-rapid reactor for cumene catalytic cracking; Etude du contact gaz-solide dans un reacteur a co-courant descendant par la mise en oeuvre du craquage catalytique du cumene

    Energy Technology Data Exchange (ETDEWEB)

    Bayle, J.

    1996-11-05

    Few studies have been carried out on the notion of gas-solid contact in ultra-rapid reactors. Both gas and solid move in the reactor and the contact can be directly estimated when using a chemical reaction such as cumene cracking. It`s a pure and light feedstock whose kinetics can be determined in a fixed bed. The study was carried out on a downflow ultra-rapid reactor (ID = 20 mm, length = 1 m) at the University of Western Ontario. It proved that the quench and the ultra-rapid separation of gas and solid must be carefully designed in the pilot plant. Cumene conversion dropped when reducing gas-solid contact, which led to push the temperature over 550 deg. C and increase the cat/oil ratio at 25 working at solid mass fluxes below 85 kg/m{sup 2}.s. Change of selectivity at very short residence time were also observed due to deactivation effects. Experiments made by Roques (1994) with phosphorescent pigments on the Residence Time Distribution of solids gave Hydrodynamic data on a cold flow copy of the pilot plant. Experiments made on packed bed gave kinetic data on the cracking of cumene. These data were combined to optimize a mono dimensional plug flow model for cumene cracking. (author)

  9. The applications of research reactors. Report of an advisory group meeting

    International Nuclear Information System (INIS)

    Owners and operators of many research reactors are finding that their facilities are not being utilized as fully as they might wish. Perhaps the original mission of the reactor has been accomplished or a particular analysis is now performed better in other ways. In addition, the fact that a research reactor exists and is available does not guarantee that users will come seeking to take advantage of the facility. Therefore, many research reactor owners and operators recognize that there is a need to develop a strategic plan for long term sustainability, including the 'marketing' of their facilities. An important first element in writing a strategic plan is to evaluate the current and potential capabilities of the reactor. The purpose of this document is to assist in such an evaluation by providing some factual and advisory information with respect to all of the current applications of research reactors. By reference to this text, each facility owner and operator will be able to assess whether or not a new application is feasible with the reactor, and what will be required to develop capability in that application. Applications fall into four broad categories: human resource development, irradiations, extracted beam work and testing. The human resource category includes public information, training and education and can be accomplished by any reactor. Irradiation applications involves inserting material into the reactor to induce radioactivity for analytical purposes, to produce radioisotopes or to induce radiation damage effects. Almost all reactors can be utilized for some irradiation applications, but as the reactor flux gets higher the range of potential uses gets larger. Beam work usually includes using neutron beams outside of the reactor for a variety of analytical purposes. Because of the magnitude of the fluxes needed at some distance from the core, most beam work can only be performed by the intermediate and higher powered research reactors. Testing nuclear

  10. The use of experimental data in an MTR-type nuclear reactor safety analysis

    Science.gov (United States)

    Day, Simon E.

    Reactivity initiated accidents (RIAs) are a category of events required for research reactor safety analysis. A subset of this is unprotected RIAs in which mechanical systems or human intervention are not credited in the response of the system. Light-water cooled and moderated MTR-type ( i.e., aluminum-clad uranium plate fuel) reactors are self-limiting up to some reactivity insertion limit beyond which fuel damage occurs. This characteristic was studied in the Borax and Spert reactor tests of the 1950s and 1960s in the USA. This thesis considers the use of this experimental data in generic MTR-type reactor safety analysis. The approach presented herein is based on fundamental phenomenological understanding and uses correlations in the reactor test data with suitable account taken for differences in important system parameters. Specifically, a semi-empirical approach is used to quantify the relationship between the power, energy and temperature rise response of the system as well as parametric dependencies on void coefficient and the degree of subcooling. Secondary effects including the dependence on coolant flow are also examined. A rigorous curve fitting approach and error assessment is used to quantify the trends in the experimental data. In addition to the initial power burst stage of an unprotected transient, the longer term stability of the system is considered with a stylized treatment of characteristic power/temperature oscillations (chugging). A bridge from the HEU-based experimental data to the LEU fuel cycle is assessed and outlined based on existing simulation results presented in the literature. A cell-model based parametric study is included. The results are used to construct a practical safety analysis methodology for determining reactivity insertion safety limits for a light-water moderated and cooled MTR-type core.

  11. Nuclear reactor decommissioning: an analysis of the regulatory environments

    International Nuclear Information System (INIS)

    In the next several decades, the electric utility industry will be faced withthe retirement of 50,000 megawatts (mW) of nuclear capacity. Responsibility for the financial and technical burdens this activity entails has been delegated to the utilities operating the reactors. However, the operators will have to perform the tasks of reactor decommissioning within the regulatory environment dictated by federal, state and local regulations. The purpose of this study was to highlight some of the current and likely trends in regulations and regulatory practices that will significantly affect the costs, technical alternatives and financing schemes encountered by the electric utilities and their customers. To identify significant trends and practices among regulatory bodies and utilities, a reviw of these factors was undertaken at various levels in the regulatory hierarchy. The technical policies were examined in reference to their treatment of allowed technical modes, restoration of the plant site including any specific recognition of the residual radioactivity levels, and planning requirements. The financial policies were examined for specification of acceptable financing arrangements, mechanisms which adjust for changes in the important parameters used to establish the fund, tax and rate-base treatments of the payments to and earnings on the fund, and whether or not escalation and/or discounting were considered in the estimates of decommissioning costs. The attitudes of regulators toward financial risk, the tax treatment of the decommissioning fund, and the time distribution of the technical mode were found to have the greatest effect on the discounted revenue requirements. Under plausible assumptions, the cost of a highly restricted environment is about seven times that of the minimum revenue requirement environment for the plants that must be decommissioned in the next three decades

  12. Experimental and modelling evaluation of an ammonia-fuelled microchannel reactor for hydrogen generation / Steven Chiuta

    OpenAIRE

    Chiuta, Steven

    2015-01-01

    In this thesis, ammonia (NH3) decomposition was assessed as a fuel processing technology for producing on-demand hydrogen (H2) for portable and distributed fuel cell applications. This study was motivated by the present lack of infrastructure to generate H2 for proton exchange membrane (PEM) fuel cells. An overview of past and recent worldwide research activities in the development of reactor technologies for portable and distributed hydrogen generation via NH3 decomposition wa...

  13. Design of an Electrical Impedance Tomography Sensor for Flow Measurement in an Oscillatory Baffled Reactor

    Science.gov (United States)

    Vilar, G.; Williams, R. A.; Wang, M.

    2007-06-01

    In this paper, a new application of electrical impedance tomography (EIT) for an advanced on-line measurement is presented. This application involves the design, manufacture and adaptation of an EIT sensor for the measurement in an oscillatory baffled reactor (OBR). The main goal is to develop of a novel measurement and modeling method for control of the OBR. The reactor itself enables the production of water-in-oil/oil-in water emulsions along with the use of chemical reagents for a variety of manufacturing processes. Use of electrical tomography facilitates detailed measurement of the concentration and flow of components in the reactor. The paper reports on design philosophy of the EIT for this application that has not, to our knowledge, been reported previously.

  14. Study of fuel assemblies for the nuclear reactor GFR; Estudio de ensambles de combustible para el reactor nuclear GFR

    Energy Technology Data Exchange (ETDEWEB)

    Reyes R, R.; Martin del Campo M, C.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, Jiutepec, Morelos 62550 (Mexico)]. e-mail: ricarera@yahoo.com.mx

    2008-07-01

    In the present work the criticality calculations for two models of fuel assembly were realized to study the nuclear reactor cooled by gas (Gas Fast Reactor) of IV Generation. Model 1 is an assembly with hexagonal adjustment of fuel rods with reflector in the axial ends higher and lower, the coolant flows between the rods. Model 2 is an hexagonal assembly type block with spheres dispersion and cylindrical channels for where the coolant with reflector in the axial ends also flows. The materials selected for each component of the assemblies, should be resistant to the radiation of fast neutrons and high operation temperatures, for what in both models the following materials were chosen: a mixture of uranium carbide more plutonium for the fuel; a mixture of silicon carbide in different theoretical density percentages for structures and shieldings; helium gas like coolant and a zirconium carbide mixture like reflector, which fulfill the restrictions of being resistant to the high operation temperatures and means of irradiation. General considerations were taken, which are common parameters to both types of assemblies, like size and materials used in the different parts of each model of assembly. The criticality calculations were obtained with the help of the MCNPx code, based on the Monte Carlo method. It was realized a validation of the atomic density data of each component of the assemblies, to have the certainty of the proportionate values that they were introduced of correct way in the code. The results show that model 1 makes better use of the fissile material in a assembly that has the same dimensions externally. That is to say, that from the only considered viewpoint, the neutron one, model 1 is better than model 2. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1986-09-30

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

  16. An alternative solution for heavy liquid metal cooled reactors fuel assemblies

    International Nuclear Information System (INIS)

    Highlights: • A new fuel assembly locking system for heavy metal cooled reactor is proposed. • Neutronic, mechanical and thermal-hydraulic evaluations of the system behavior have been performed. • A comparison with other solutions has been presented. - Abstract: In the coming future, the electric energy production from nuclear power plants will be provided by both thermal reactors and fast reactors. In order to have a sustainable energy production through fission reactors, fast reactors should provide an increasing contribution to the total electricity production from nuclear power plants. Fast reactors have to achieve economic and technical targets of Generation IV. Among these reactors, Sodium cooled Fast Reactors (SFRs) and Lead cooled Fast Reactors (LFRs) have the greatest possibility to be developed as industrial power plants within few decades. Both SFRs and LFRs require a great R and D effort to overcome some open issues which affect the present designs (e.g. sodium-water reaction for the SFRs, erosion/corrosion for LFRs, etc.). The present paper is mainly focused on LFR fuel assembly (FA) design: issues linked with the high coolant density of lead or lead–bismuth eutectic cooled reactors have been investigated and an innovative solution for the core mechanical design is here proposed and analyzed. The solution, which foresees cylindrical fuel assemblies and exploits the buoyancy force due to the lead high density, allows to simplify the FAs locking system, to reduce their length and could lead to a more uniform neutron flux distribution

  17. An alternative solution for heavy liquid metal cooled reactors fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Vitale Di Maio, Damiano, E-mail: damiano.vitaledimaio@uniroma1.it [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Cretara, Luca; Giannetti, Fabio [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Peluso, Vincenzo [“ENEA”, Via Martiri di Monte Sole 4, 40129 Bologna (Italy); Gandini, Augusto [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Manni, Fabio [“SRS Engineering Design S.r.l.”, Vicolo delle Palle 25-25/b, 00186 Rome (Italy); Caruso, Gianfranco [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy)

    2014-10-15

    Highlights: • A new fuel assembly locking system for heavy metal cooled reactor is proposed. • Neutronic, mechanical and thermal-hydraulic evaluations of the system behavior have been performed. • A comparison with other solutions has been presented. - Abstract: In the coming future, the electric energy production from nuclear power plants will be provided by both thermal reactors and fast reactors. In order to have a sustainable energy production through fission reactors, fast reactors should provide an increasing contribution to the total electricity production from nuclear power plants. Fast reactors have to achieve economic and technical targets of Generation IV. Among these reactors, Sodium cooled Fast Reactors (SFRs) and Lead cooled Fast Reactors (LFRs) have the greatest possibility to be developed as industrial power plants within few decades. Both SFRs and LFRs require a great R and D effort to overcome some open issues which affect the present designs (e.g. sodium-water reaction for the SFRs, erosion/corrosion for LFRs, etc.). The present paper is mainly focused on LFR fuel assembly (FA) design: issues linked with the high coolant density of lead or lead–bismuth eutectic cooled reactors have been investigated and an innovative solution for the core mechanical design is here proposed and analyzed. The solution, which foresees cylindrical fuel assemblies and exploits the buoyancy force due to the lead high density, allows to simplify the FAs locking system, to reduce their length and could lead to a more uniform neutron flux distribution.

  18. The status of studies on fast reactor core thermal hydraulics at PNC

    International Nuclear Information System (INIS)

    An outlook was addressed on investigative activities of the fast reactor core thermal-hydraulics at Power Reactor and Nuclear Fuel Development Corporation. Firstly, a computational modeling to predict flow field under natural circulation decay heat removal condition using multi-dimensional codes and its validation were presented. The validation was carried out through calculations of sodium experiments on an inter-subassembly heat transfer, a transient from forced to natural circulation and an inter-wrapper flow. Secondly, experimental and computational studies were expressed on local blockage with porous media in a fuel subassembly. Lastly, information was presented on an advanced computational code based on a subchannel analysis code. The code is under the development and extended to perform whole core simulation. (author)

  19. Blanket design study for a Commercial Tokamak Hybrid Reactor (CTHR)

    International Nuclear Information System (INIS)

    The results are presented of a study on two blanket design concepts for application in a Commercial Tokamak Hybrid Reactor (CTHR). Both blankets operate on the U-Pu cycle and are designed to achieve tritium self-sufficiency while maximizing the fissile fuel production within thermal and mechanical design constraints. The two blanket concepts that were evaluated were: (1) a UC fueled, stainless steel clad and structure, helium cooled blanket; and (2) a UO2 fueled, zircaloy clad, stainless steel structure, boiling water cooled blanket. Two different tritium breeding media, Li2O and LiH, were evaluated for use in both blanket concepts. The use of lead as a neutron multiplier or reflector and graphite as a reflector was also considered for both blankets

  20. Experimental Study on the Hydrodynamics and Mass Transfer Characteristics of Airlift Loop Reactors(ALR)

    Institute of Scientific and Technical Information of China (English)

    Tang Lixin; Han Pingfang; Lu Xiaoping

    2007-01-01

    The promoting effect of ultrasonic wave on the hydrodynamics and mass transfer characteristics of the airlift loop reactor was studied. The effect of the airlift reactor and ultrasonic wave on the reactor's gas holdup, liquid circulation velocity, mixing time and overall volumetric mass transfer coefficient respectively with and without the presence of ultrasonic wave is empathetically examined and compared. The experiment has proven that the incorporation of ultrasonic wave has no effect on the gas holdup but has the tendency to gradually decrease the liquid circulation velocity and increase the overall volumetric mass transfer coefficient; the effect on the mixing time is relatively complex. At low gas velocity, low powered ultrasonic wave promotes the radial mixing of fluid; with the increase of ultrasonic power, ultrasonic vibration obstructs the radial mixing of fluid. Therefore, there exists an optimal ultrasonic power. Moreover, the effect of ultrasonic wave on the mixing time gradually decreases with the increase of the superficial gas velocity. Correlations were also proposed for the hydrodynamics and mass transfer characteristics of the reactor.

  1. Performance comparison of a continuous-flow stirred-tank reactor and an anaerobic sequencing batch reactor for fermentative hydrogen production depending on substrate concentration.

    Science.gov (United States)

    Kim, S-H; Han, S-K; Shin, H-S

    2005-01-01

    This study was conducted to compare the performance of a continuous-flow stirred-tank reactor (CSTR) and an anaerobic sequencing batch reactor (ASBR) for fermentative hydrogen production at various substrate concentrations. Heat-treated anaerobic sludge was utilized as an inoculum, and hydraulic retention time (HRT) for each reactor was maintained at 12 h. At the influent sucrose concentration of 5 g COD/L, start-up was not successful in both reactors. The CSTR, which was started-up at 10 g COD/L, showed stable hydrogen production at the influent sucrose concentrations of 10-60 g COD/L during 203 days. Hydrogen production was dependent on substrate concentration, resulting in the highest performance at 30 g COD/L. At the lower substrate concentration, the hydrogen yield (based on hexose consumed) decreased with biomass reduction and changes in fermentation products. At the higher substrate concentration, substrate inhibition on biomass growth caused the decrease of carbohydrate degradation and hydrogen yield (based on hexose added). The ASBR showed higher biomass concentration and carbohydrate degradation efficiency than the CSTR, but hydrogen production in the ASBR was less effective than that in the CSTR at all the substrate concentrations.

  2. An automated boron management system for WWER-1000 nuclear reactors

    Directory of Open Access Journals (Sweden)

    Taisiya O. Tsiselskaya

    2015-03-01

    Full Text Available The article is devoted to the problem of creating a system of automated control with boron regulation for reactor WWER-1000 series. Using the boron regulation to control WWER-1000 allows to extend its maximum output operation period, ensuring the economic efficiency of the power unit, as well as to maintain the reactor facility within relevant safety limits that prevents from emergencies occurrence and development. The results of this problem solution, related to the process simulation, optimization and prediction, were used at further development of computer-integrated control system increasing the efficiency of decisions, taken by operational staff at reactor control.

  3. Research Reactors - An analysis with a focus on non-proliferation and export control

    International Nuclear Information System (INIS)

    The Swedish Defence Research Agency, FOI, has under contract work financed by the Swedish Radiation Safety Authority, performed a study on research reactors. The principle of a research reactor and its characteristics and uses are described in this report. The potential use of research reactors for plutonium production for nuclear weapons is also described and the parameters of importance for optimal plutonium production are identified. Research reactors, mainly heavy water or graphite moderated reactors, have been used by some countries to produce weapon-grade plutonium. To prevent nuclear weapons proliferation, the Nuclear Suppliers' Group has identified nuclear reactor equipment and technology that is of importance to export control. This equipment and technology has also been implemented in the EU-regulation 428/2009. The equipment and technology that can be used in nuclear reactor applications and possible indicators on a nuclear reactor in operation is described in the report. The differences between a power reactor and a research reactor concerning these areas are high-lighted

  4. Design studies of interaction processes between melt reactor core material composition, coolant and construction material

    International Nuclear Information System (INIS)

    This effort presents results of the design studies performed for correct conducting of out-of-pile experiments at the National Nuclear Center of the Republic of Kazakhstan facilities to study interaction processes between the melt fuel composition and core materials, which might take place during accidents at the nuclear power plants. The design methods are considered to determine electromagnetic parameters of the facility melting unit - an induction melting furnace, to determine temperature fields in furnace construction component materials during the experiment process and at the moment of melt discharge. The methods are presented to calculate temperature fields and thermal flows in experimental device modelling a reactor vessel bottom when it contacts with the melt fuel composition. The results of thermal electric and hydraulic calculations are presented validating the experimental device operability designed to study the interaction processes between the melt fuel composition and reactor vessel bottom in simulating a residual energy release in the melt by direct current transmitting

  5. Dosimetry requirements in support of an in-reactor radiation chemistry programme

    International Nuclear Information System (INIS)

    Reactors are now used as radiation sources in a wide range of studies. In some, the changes observed result from the interaction of the irradiated material with only one type of radiation present in the reactor; in others, a number of different radiations may produce the observed effect. In any investigation of the effects of radiation on matter, it is necessary to know the spectrum of the radiation used. This information should be available in the form of the variation of photon or particle flux density with photon or particle energy. The amount of change observed in an irradiated material will always be the product of the radiation flux density and an interaction function, integrated over the radiation spectrum. The nature of this interaction function will vary with the type of study being performed. The radiation spectra in reactors are complex in form and cover wide ranges of photon and particle energies. Given sufficient information about the fission reaction, and about the geometry and materials of construction of a reactor, it is possible, in principle, to calculate the spectra of the reactor radiations. It is then possible, in principle, to calculate the rates of production of change in irradiated materials. In this paper the conventions used in radiation chemistry are discussed. This is followed by a discussion of the various types of calorimeters which can be used for dosimetry, with some remarks about their limitations. The reasons for choosing an isothermal-type calorimeter for the work at the Atomic Energy Research Establishment (AERE) are given, followed by a description of the construction, characteristics, and calibration of the calorimeters in use. It is shown that under the proper conditions these calorimeters measure the ideal absorbed dose-rate, or kerma rate. By the use of three calorimeters, as described, containing respectively graphite, anthracene, and an empty can, measurements were made of the dose rates in graphite and in anthracene, and

  6. Analysis of thermal-hydraulics of a marine reactor in an oscillating acceleration field

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jae Hak; Park, Goon Cherl [Seoul National Univ., Seoul (Korea, Republic of)

    1996-07-01

    In this study the RETRAN-03 code was modified to analyze the thermal-hydraulic transients under three-dimensional ship motions for the application to the future marine reactors. First Japanese nuclear ship MUTSU reactor have been analyzed under various ship motions to verify this code. As results, typical thermal-hydraulic characteristics of marine reactors such as flow rate oscillations and S/G water level oscillations are successfully simulated at various conditions.

  7. Design and implementation of an intensified coprecipitation reactor for the treatment of liquid radioactive wastes

    International Nuclear Information System (INIS)

    The coprecipitation is a robust and inexpensive process for the treatment of important volumes of low and intermediate radioactive level liquid wastes. Its major inconvenient is the huge volume of sludge generated. The purpose of this work is to optimize the industrial coprecipitation continuous process by achieving the following objectives: - maximize the decontamination efficiency; - minimize the volume of sludge generated by the process; - reduce the treatment cost decreasing the installation volume. An innovative reactor with an infinite recycling ratio was therefore designed. It is a multifunctional reactor composed of two zones: a perfectly mixed precipitation zone and a classifier to perform liquid-solid separation. The experiments are focused on the coprecipitation of strontium by barium sulphate. The effluent containing sulphate ions and the barium nitrate solution are injected in the reaction zone where strontium and barium co-precipitate as sulphates. The produced solid phase is returned into the reaction zone by the classifier and goes out slowly from the reactor bottom with a residence time much higher than the liquid phase. This creates both a high concentration of solid phase in the reaction zone and a high efficiency of decontamination. The experimental conditions simulate the industrial effluents. The total treatment flow rate is 17 L/h, with an effluent flow rate of 16 L/h and a reactive flow rate of 1 L/h, hence a mean residence time of 10 minutes. In these experimental conditions, the molar ratio sulphate/barium after mixing corresponds to 4.9. These conditions are used in the reprocessing plant of La Hague. The decontamination factor reached in these experimental conditions is excellent: DF = 1500. The decontamination factor obtained with the classical continuous process is only equal to 60. Different process parameters are studied in order to optimize the reactor/classifier: residence time, barium nitrate flow rate and racking flow rate. The

  8. Shuffling strategy study of breeding-burning integrated fast reactor

    International Nuclear Information System (INIS)

    The breeding-burning integrated fast reactor uses burning assemblies to generate thermal power, meanwhile, converts 238U into 239Pu in the fertile assemblies. With periodical shuffling of assemblies, the reactor can maintain criticality for decades of years. To maintain long-term stability of the core reactivity, the core layout and shuffling strategy should balance the burning and the breeding of the assemblies. The scattered core layout and shuffling strategy ensures fast breeding of the fertile assemblies, and keeps stable core power distribution in whole life of the reactor. Moreover, at the end of the reactor life, the discharge burnups of different fuel assemblies are close to each other, which are about 250300 GW · d/t. This is important for breeding-burning integrated fast reactor to achieve very efficient utilization of uranium resource without reprocessing. (authors)

  9. Lead-cooled fast reactor (BREST) with an on-site fuel cycle

    International Nuclear Information System (INIS)

    Full text: Out of a great many of new power technologies, fission nuclear power is the only realistic way to stop the growth in extraction and combustion of fossil fuel. However, this is something to be achieved only through the nuclear power to have by the mid-21st century the capacity an order of magnitude as high as the current level. Nuclear power of such a scale will necessitate a new nuclear technology which is required to provide: transition to power with unlimited fuel resources; economically competitive nuclear power through reducing the cost of building and operating NPPs of a high inherent safety level with highly efficient utilization of fuel and generated heat; elimination of severe accidents with radioactive release which require evacuation of population, to be achieved primarily through combining inherent safety, passive protection features and impossible loss of lead coolant; an environmentally safe closed fuel cycle with in-pile combustion of minor actinides and radiation-equivalent disposal of radioactive waste; creation of nuclear proliferation barriers by way of eliminating uranium enrichment and plutonium separation facilities. These problems are solved in the BREST lead-cooled fast reactor. The initial stage plan is to build an NPP with a demonstration reactor and an on-site fuel cycle to verify designs, try out processes involving lead using as the coolant and study the behavior of the reactor and its systems and components in different modes, including resistance to anticipated operational occurrences and accidents simulated on the reactor. This will be followed by a rapid transition to a fast lead-cooled power reactor of 1200 MW(e) featuring two- circuit heat removal from the core to the turbine with supercritical steam parameters. The state of the activities to develop the NPP with lead-cooled fast reactors and an on-site fuel cycle is presented. A core with a moderate power rating has been considered. This will have a uranium

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

    DEFF Research Database (Denmark)

    2014-01-01

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

  11. Further optimization studies of experimental dynamic responses measured on the HTGC Dragon reactor

    International Nuclear Information System (INIS)

    This report considers some measurements made of the dynamics of the HTGC Dragon reactor and the optimization of a mathematical model which represents the reactor, by altering the parameters until a least squares fit between the experimental responses and the mathematical model is obtained. The experimental information was processed in various ways. The experimental response to an impulse, step or periodic sine wave change in reactivity was processed as an impulse, step or periodic sine wave response respectively and compared with a similar response from the model. In other studies the result of a binary cross correlation experiment (effectively an impulse response input) was processed as a frequency response and this experimental frequency response was compared with the frequency response from the mathematical model. It was possible therefore to compare the optimum values of parameters, obtained for different forms of perturbing signal and for different methods of processing and to relate the optima obtained to the problem of parameter estimation. (author)

  12. A comparative study of sequencing batch reactor and moving-bed sequencing batch reactor for piggery wastewater treatment

    Directory of Open Access Journals (Sweden)

    Kwannate Sombatsompop

    2011-06-01

    Full Text Available This research aims to comparatively study the efficiency of piggery wastewater treatment by the moving-bed sequencing batch reactor (moving-bed SBR system with held medium, and the conventional sequencing batch reactor (SBR system, by varying the organic load from 0.59 to 2.36 kgCOD/m3.d. The COD treatment efficiency of the SBR and moving-bed SBR was higher than 60% at an organic load of 0.59 kgCOD/m3.d and higher than 80% at the organic loads of 1.18-2.36 kgCOD/m3.d. The BOD removal efficiency was greater than 90% at high organic loads of 1.18-2.36 kgCOD/m3.d. The moving-bed SBR gave TKN removal efficiency of 86-93%, whereas the SBR system exhibited the removal efficiency of 75-87% at all organic loads. The amount of effluent suspended solids for SBR systems exceeded the piggery wastewater limit of 200 mg/L at the organic load of 2.36 kgCOD/m3.d while that for the moving-bed SBR system did not. When the organic load was increased, the moving-bed SBR system yielded better treatment efficiency than that of the SBR system. The wastewater treated by the moving-bed SBR system met the criteria of wastewater standard for pig farms at all organic loads, while that treated by the SBR system was not satisfactory at a high organic load of 2.36 kgCOD/m3.d.

  13. Development of an advanced antineutrino detector for reactor monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Classen, T., E-mail: classen2@llnl.gov [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Bernstein, A.; Bowden, N.S. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Cabrera-Palmer, B. [Sandia Livermore National Laboratories, Livermore, CA 94550 (United States); Ho, A.; Jonkmans, G. [Atomic Energy of Canada, Limited, Chalk River Laboratories, Chalk River, ON (Canada); Kogler, L.; Reyna, D. [Sandia Livermore National Laboratories, Livermore, CA 94550 (United States); Sur, B. [Atomic Energy of Canada, Limited, Chalk River Laboratories, Chalk River, ON (Canada)

    2015-01-21

    Here we present the development of a compact antineutrino detector for the purpose of nuclear reactor monitoring, improving upon a previously successful design. This paper will describe the design improvements of the detector which increases the antineutrino detection efficiency threefold over the previous effort. There are two main design improvements over previous generations of detectors for nuclear reactor monitoring: dual-ended optical readout and single volume detection mass. The dual-ended optical readout eliminates the need for fiducialization and increases the uniformity of the detector's optical response. The containment of the detection mass in a single active volume provides more target mass per detector footprint, a key design criteria for operating within a nuclear power plant. This technology could allow for real-time monitoring of the evolution of a nuclear reactor core, independent of reactor operator declarations of fuel inventories, and may be of interest to the safeguards community.

  14. Aerosol transport and behaviour in the context of nuclear reactor safety - an overview

    International Nuclear Information System (INIS)

    The phenomenon of aerosol behaviour in the context of nuclear reactor safety following a severe accident has generated considerable interest. A large number of studies, both theoretical and experimental have been carried out. Some of the representative studies on aerosol transport and behaviour within the containment of nuclear reactors are reviewed

  15. The use of an object oriented technique for fault diagnosis in nuclear reactors

    International Nuclear Information System (INIS)

    In recent years there has been an increased growth of interest in object oriented programming which is a new approach to software construction having wide application. The possibility of using object oriented programming to build a software package that will assist the nuclear reactor operator in diagnosing any faults or alarms in the Loss Of Fluid Test (LOFT), a small scale pressurised water reactor, reactor is investigated in this work. (author)

  16. Testing of an advanced thermochemical conversion reactor system

    Energy Technology Data Exchange (ETDEWEB)

    1990-01-01

    This report presents the results of work conducted by MTCI to verify and confirm experimentally the ability of the MTCI gasification process to effectively generate a high-quality, medium-Btu gas from a wider variety of feedstock and waste than that attainable in air-blown, direct gasification systems. The system's overall simplicity, due to the compact nature of the pulse combustor, and the high heat transfer rates attainable within the pulsating flow resonance tubes, provide a decided and near-term potential economic advantage for the MTCI indirect gasification system. The primary objective of this project was the design, construction, and testing of a Process Design Verification System for an indirectly heated, thermochemical fluid-bed reactor and a pulse combustor an an integrated system that can process alternative renewable sources of energy such as biomass, black liquor, municipal solid waste and waste hydrocarbons, including heavy oils into a useful product gas. The test objectives for the biomass portion of this program were to establish definitive performance data on biomass feedstocks covering a wide range of feedstock qualities and characteristics. The test objectives for the black liquor portion of this program were to verify the operation of the indirect gasifier on commercial black liquor containing 65 percent solids at several temperature levels and to characterize the bed carbon content, bed solids particle size and sulfur distribution as a function of gasification conditions. 6 refs., 59 figs., 29 tabs.

  17. The catalytic hydrogenation of 2,4-dinitrotoluene in a continuous stirred three-phase slurry reactor with an evaporting solvent

    NARCIS (Netherlands)

    Westerterp, K.R.; Janssen, H.J.; Kwast, van der H.J.

    1992-01-01

    An experimental study of the catalytic hydorgenation of 2,4-dinitrotoluene (DNT) in a mini-installation with a continuously operated stirred three-phase slurry reactor and an evaporating solvent is discussed. Some characteristic properties of the reactor system and the influence of the operating par

  18. 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.

  19. Monte Carlo calculation of neutron generation time in critical reactor and subcritical reactor with an external source

    International Nuclear Information System (INIS)

    The neutron generation time Λ plays an important role in the reactor kinetics. However, it is not straightforward nor standard in most continuous energy Monte Carlo codes which are able to calculate the prompt neutron lifetime lp directly. The difference between Λ and lp are sometimes very apparent. As very few delayed neutrons are produced in the reactor, they have little influence on Λ. Thus on the assumption that no delayed neutrons are produced in the system, the prompt kinetics equations for critical system and subcritical system with an external source are proposed. And then the equations are applied to calculating Λ with pulsed neutron technique using Monte Carlo. Only one fission neutron source is simulated with Monte Carlo in critical system while two neutron sources, including a fission source and an external source, are simulated for subcritical system. Calculations are performed on both critical benchmarks and subcritical system with an external source and the results are consistent with the reference values. (author)

  20. Antineutrino analysis for continuous monitoring of nuclear reactors: Sensitivity study

    Energy Technology Data Exchange (ETDEWEB)

    Stewart, Christopher; Erickson, Anna [Georgia Institute of Technology, Nuclear and Radiological Engineering, G. W. Woodruff School of Mechanical Engineering, 770 State St. NW, Atlanta, Georgia 30332 (United States)

    2015-10-28

    This paper explores the various contributors to uncertainty on predictions of the antineutrino source term which is used for reactor antineutrino experiments and is proposed as a safeguard mechanism for future reactor installations. The errors introduced during simulation of the reactor burnup cycle from variation in nuclear reaction cross sections, operating power, and other factors are combined with those from experimental and predicted antineutrino yields, resulting from fissions, evaluated, and compared. The most significant contributor to uncertainty on the reactor antineutrino source term when the reactor was modeled in 3D fidelity with assembly-level heterogeneity was found to be the uncertainty on the antineutrino yields. Using the reactor simulation uncertainty data, the dedicated observation of a rigorously modeled small, fast reactor by a few-ton near-field detector was estimated to offer reduction of uncertainty on antineutrino yields in the 3.0–6.5 MeV range to a few percent for the primary power-producing fuel isotopes, even with zero prior knowledge of the yields.

  1. Conceptual Study on Dismantling of CANDU Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Woo-Tae; Lee, Sang-Guk [KHNP-CRI, Daejeon (Korea, Republic of)

    2014-10-15

    In this paper, we reviewed 3D design model of the CANDU type reactor and suggested feasible cutting scheme. The structure of CANDU nuclear reactor, the calandria assembly was reviewed using 3-D CAD model for future decommissioning. Through the schematic diagram of CANDU nuclear power plant, we identified the differences between PWR and CANDU reactor assembly. Method of dismantling the fuel channels from the calandria assembly was suggested. Custom made cutter is recommended to cut all the fuel channels. The calandria vessel is recommended to be cut by band saw or plasma torch. After removal of the fuel channels, it was assumed that radiation level near the calandria vessel is not very high. For cutting of the end shields, various methods such as band saw, plasma torch, CAMC could be used. The choice of a specific method is largely dependent on radiological environment. Finally, method of cutting the embedment rings is considered. As we assume that operators could cut the rings without much radiation exposure, various industrial cutting methods are suggested to be applied. From the above reviews, we could conclude that decommissioning of CANDU reactor is relatively easy compared to that of PWR reactor. Technologies developed from PWR reactor decommissioning could be applied to CANDU reactor dismantling.

  2. Fusion reactor physics and technology studies: Final report, December 1, 1985 to June 15, 1986

    International Nuclear Information System (INIS)

    The work performed for ''Fusion Reactor Physics and Technology Studies'' was to participate with Lawrence Livermore National Laboratory (LLNL) and others in the design of the Minimars Tandem Mirror power reactor. This design activity was a collaborative effort with LLNL being responsible for the physics and program administration. The Fusion Technology Institute (FTI) of the University of Wisconsin was responsible for nuclear technology (central cell blanket/reflector/shield design, neutronics, activation, tritium issues, and end cell shielding), and the design of the central cell magnets. In addition, FTI provided assistance to LLNL on the plasma physics modeling of the plasma halo. An interim report on the Minimars design has been issued, and a final report covering all aspects of Minimars is in press. In addition, various reports and conference papers covering aspects of the design have been issued, and are listed in this report. A summary of the FTI contribution to the Minimars design study is presented

  3. Experimental study of thermohydraulic processes and gas distribution in a model of the containment shell of the AST-500 reactor

    International Nuclear Information System (INIS)

    Experiments were made on a setup consisting of a large-scale twin-assembly model of the primary circuit of an integral reactor and of a model of a containment shell which is a means for confining the outflow of coolant from the reactor. The large-scale model of an AST-500 reactor has vertical dimensions close to the actual dimensions and similar coefficients of hydraulic resistance and volume ratios of the principal elements the circuit with natural circulation. The model of the containment shell is a vertical cylindrical vessel with a size of 426 x 12 mm, a height of 9.78 m, and a volume of 1.24 m3. The volume scale of the reactor model and of the model of the containment shell is 1:170. The elements of the latter model are made from steel 20. The models of the reactor and of the containment shell are joined through two pipelines with a size of 57 x 3.5 mm and shut-off valves with a diameter of 50 mm mounted thereon. A total of 70 experiments were made to simulate leakage of the primary circuit of the integrated reactor and the outflow of coolant into the containment shell. The authors have provided detailed information on the large-scale model, have described the experimental conditions, and have reported on the main results of their study of the development of an accident involving the loss of coolant in the reactor-containment shell system. The present article reports on a study of the thermohydraulic processes and the gas distribution in the containment shell. Since the designs of the model and of the actual containment shell of the AST-500 reactor are not identical, the authors assume that the results reported can be used in appropriate computer programs describing the processes which occur in containment vessels of atomic power stations (containment shells, protective shells, sealed assemblies)

  4. Hydrodynamic modelling of flow patterns in a vortex reactor - Application to the mixing study

    Energy Technology Data Exchange (ETDEWEB)

    Zoppe, B. [CEA Valrho Marcoule DEN/DRCP/SCPS/LCA, BP 17171, 30207 Bagnols-sur-Ceze (France); CEA Grenoble DEN/DER/SSTH/LMDL, 38054 Grenoble Cedex 9 (France); Lebaigue, O.; Ducros, F. [CEA Grenoble DEN/DER/SSTH/LMDL, 38054 Grenoble Cedex 9 (France); Bertrand, M. [CEA Valrho Marcoule DEN/DRCP/SCPS/LCA, BP 17171, 30207 Bagnols-sur-Ceze (France)

    2008-07-01

    In the fuel reprocessing industry, an un-baffled magnetic rod-stirred multiphase reactor was developed for a precipitation operation. The flow generated in such a reactor is complex and the rotating agitator at the bottom of tank creates a vortex on the liquid surface. A Computational Fluid Dynamics (CFD) modelling is developed based on a Large Eddy Scale (LES) approach for turbulence effect simulation. The numerical simulations are performed in 3-dimensions using the Trio-U code developed at the Commissariat a l'Energie Atomique (Cea). The vortex study is based on an interface tracking method and the rotating magnetic rod is taken into account through a free IBC immersed boundary. The hydrodynamic modelling is in good agreement with Nagata's theory and will be validated from experimental data obtained by laser doppler velocimetry (LDV) measurements. (authors)

  5. The nuclear data, A key component for reactor studies, Overview of AREVA NP needs and applications

    Directory of Open Access Journals (Sweden)

    Ravaux Simon

    2016-01-01

    Full Text Available The quality of the nuclear data is essential for AREVA NP. Indeed, many AREVA NP activities such as reactor design, safety studies or reactor instrumentation use them as input data. So, the nuclear data can be considered as a key element for AREVA NP. REVA NP’s contribution in the improvement of the nuclear data consists in a joint effort with the CEA. It means a financing and a sharing of information which can give an orientation to the future research axis. The aim of this article is to present the industrial point of view from AREVA NP on the research on nuclear data. Several examples of collaborations with the CEA which have resulted in an improvement of the nuclear data are presented.

  6. A concept of an advanced inertia fusion reactor; TAKANAWA-I

    International Nuclear Information System (INIS)

    A concept of an advanced inertia fusion reactor: TAKANAWA-I is proposed. A pellet with DT ignitor and DD major fuel, Pb wet walls, C or SiC blocks for shielding, and SiC vessels in the water pool are employed. This reactor does not need blanckets for T breeding, since T is supplied through DD reaction, and has low induced radioactivities. These and a simple structure might give a hopeful prediction of economical and safe advantages and mitigate difficulties of reactor technologies, especially remote maintenance of the reactor. (author)

  7. Coupling of a System Analysis with a Neutronics Analysis for Conceptual Study of Fusion- Driven Transmutation Reactor

    International Nuclear Information System (INIS)

    A compact tokamak reactor concept as a 14 MeV neutron source is desirable from an economic viewpoint for a fusion-driven transmutation reactor. For the optimal design of a reactor, a radial build of reactor components has to be determined by considering the plasma physics and engineering constraints which interrelate various reactor components. In a transmutation reactor, design of blanket and shield play a key role in determining the size of a reactor; the blanket should produce enough tritium for tritium self-sufficiency, the transmutation rate of waste has to be maximized, and the shield should provide sufficient protection for the superconducting toroidal field (TF) coil. To determine the radial build of the blanket and the shield, not only a radiation transport analysis but also a burn-up calculation were coupled with the system analysis and it allowed the self-consistent determination of the design parameters of a transmutation reactor. For neutronic optimization of the blanket and the shield, the quantities such as the tritium breeding ratio (TBR), nuclear heating, radiation damage to the toroidal field coil have to be calculated and burn-up rates of Li, actinides and fission products have to be calculated. Thus the neutronic analysis need to be coupled in the system analysis. In most of the previous system studies, neutronic calculation and plasma analysis are performed separately, so blanket and shield size was determined independently from the reactor size. In this work, to account for the interrelation of blanket and shield with the other components of a reactor system, we coupled the system analysis with one-dimensional neutronic calculation to determine the reactor parameters in self consistent manner. LAR (Low Aspect Ratio) tokamak plasma has the potential of high β operation with high bootstrap current fractions. In the LAR tokamak reactor, the radial build of TF coil(TFC) and the shield play a key role in determining the size of a reactor since it

  8. Using activation method to measure neutron spectrum in an irradiation chamber of a research reactor

    International Nuclear Information System (INIS)

    Neutron spectrum should be measured before test samples are irradiated. Neutron spectrum in an irradiation chamber of a research reactor was measured by using activation method when the reactor is in normal operation under 2 MW. Sixteen kinds of non-fission foils (19 reaction channels) were selected, of which 10 were sensitive to thermal and intermediate energy regions, while the others were of different threshold energy and sensitive to fast energy regions. By measuring the foil radioactivity, the neutron spectrum was unfolded with the iterative methods SAND-II and MSIT. Finally, shielding corrections of group cross-section and main factors affecting the calculation accuracy were studied and the uncertainty of solution was analyzed using the Monte Carlo method in the process of SAND-II. (authors)

  9. Numerical study of effect of operating and design parameters for design of steam reforming reactor

    International Nuclear Information System (INIS)

    A numerical study on the design of a steam reforming reactor consisting of several reforming tubes and one burner is conducted with respect to various operating and design parameters such as GHSV (gas hourly space velocity), input heat capacity, catalyst layer length, and number of tubes. The calculation of the reforming reaction rate is coupled with a three-dimensional heat and mass transfer calculation. It is shown that a large temperature gradient exists in the reforming reactor, resulting in significant variation of the gas temperature and reaction rate along the reforming tube. The reduction of the catalyst layer length induces a decrease in H2 (hydrogen) concentration as well as pressure loss. An increased number of tubes leads to better system efficiency owing to the enhanced heat transfer to the reforming tube. Consequently, to improve the system efficiency and reduce the pressure loss, an increase in heat transfer area and decrease in catalyst layer length should be essential design considerations. - Highlights: • A numerical study for the design on a steam reforming reactor is investigated. • The gas concentrations change significantly with variations in the GHSV and input heat capacity. • The reduction of the catalyst layer length induces a decrease in H2 concentration as well as pressure loss. • An increased number of tubes leads to better system efficiency owing to the enhanced heat transfer to reforming tube

  10. RNA-protein binding kinetics in an automated microfluidic reactor.

    Science.gov (United States)

    Ridgeway, William K; Seitaridou, Effrosyni; Phillips, Rob; Williamson, James R

    2009-11-01

    Microfluidic chips can automate biochemical assays on the nanoliter scale, which is of considerable utility for RNA-protein binding reactions that would otherwise require large quantities of proteins. Unfortunately, complex reactions involving multiple reactants cannot be prepared in current microfluidic mixer designs, nor is investigation of long-time scale reactions possible. Here, a microfluidic 'Riboreactor' has been designed and constructed to facilitate the study of kinetics of RNA-protein complex formation over long time scales. With computer automation, the reactor can prepare binding reactions from any combination of eight reagents, and is optimized to monitor long reaction times. By integrating a two-photon microscope into the microfluidic platform, 5-nl reactions can be observed for longer than 1000 s with single-molecule sensitivity and negligible photobleaching. Using the Riboreactor, RNA-protein binding reactions with a fragment of the bacterial 30S ribosome were prepared in a fully automated fashion and binding rates were consistent with rates obtained from conventional assays. The microfluidic chip successfully combines automation, low sample consumption, ultra-sensitive fluorescence detection and a high degree of reproducibility. The chip should be able to probe complex reaction networks describing the assembly of large multicomponent RNPs such as the ribosome.

  11. Conceptual design study of fusion experimental reactor (FY86FER)

    International Nuclear Information System (INIS)

    This report describes the results of the capacity estimation for the electrical power system on the typical two candidates for the FER (Fusion Experimental Reactor) which were picked out through the process of '86 FER scoping studies. Main concern in the electrical systems is coil power supplies which have a capacity of about 1 GW, and this is dominated by poloidal coil power supplies. Then, studies to reduce the converter capacity are concentrated on the poloidal coil power system in relation to the sypplying poloidal flux at the initial phase of plasma ramp-up. A quench protection circuit was proposed on the toroidal coil power supply. On the position control power supply, a circuit with reasonable functions was proposed. Under these system studies, general specifications were determined and the capacity of each power supply unit was estimated. On the poloidal coil power supply system, the accumulated capacity of converters amounted to 885 MW for the one candidate and 782 MW for another. (author)

  12. Operational transparency: an advanced safeguards strategy for future on-load refuelled reactors

    International Nuclear Information System (INIS)

    The IAEA's system for tracking fuel movement in an on-load refuelled heavy-water reactor is robust, but an opportunity remains to exploit the wealth of data streaming from the reactor vault during operation and provide real-time, third-party monitoring of reactor status and history. This concept of Operational Transparency would require that large amounts of operational data be reduced in near-real time to a small subset of high-level information. Operational Transparency would enhance the IAEA's ability to monitor the state of the core to an unprecedented level. This paper provides an overview of the novel concept of Operational Transparency in heavy water reactors, using potential application to CANDU reactors as an example, and explores some of the technical challenges that will need to be solved for efficient implementation. (author)

  13. Study of reactor parameters of on critical systems, Phase I: Safety report for RB zero power reactor

    International Nuclear Information System (INIS)

    In addition to the safety analysis for the zero power RB reactor, this report contains a general description of the reactor, reactor components, auxiliary equipment and the reactor building. Reactor Rb has been reconstructed during 1961-1962 and supplied with new safety-control system as well as with a complete dosimetry instrumentation. Since RB reactor was constructed without shielding special attention is devoted to safety and protection of the staff performing experiments. Due to changed circumstances in the Institute ( start-up of the RA 7 MW power reactor) the role of the RB reactor was redefined

  14. Reactor safeguards

    CERN Document Server

    Russell, Charles R

    2013-01-01

    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

  15. An advanced frequency-domain code for boiling water reactor (BWR) stability analysis and design

    International Nuclear Information System (INIS)

    for a large reactor could reach 20 gigabytes) that it is not possible to load into RAM memory of an operating system with 32 bit architecture. A special procedure has been developed within the MATLAB environment to remove this memory limitation, and to invert such large matrices and finally obtain the reactor transfer functions that enable the study of system stability. Various applications of the present frequency-domain code to a typical BWR fuel assembly, a BWR core, and to a chemical reactor showed a good agreement with reference results. (author)

  16. Study and Analysis on Naphtha Catalytic Reforming Reactor Simulation

    Institute of Scientific and Technical Information of China (English)

    Liang Ke min; Song Yongji; Pan Shiwei

    2004-01-01

    A naphtha catalytic reforming unit with four reactors connected in series is analyzed. A physical model is proposed to describe the catalytic reforming radial flow reactor. Kinetics and thermodynamics equations are selected to describe the naphtha catalytic reforming reaction characteristics based on idealizing the complex naphtha mixture to represent the paraffin, naphthene, and aromatic groups with individual compounds. The simulation results based on above models agree very well with actual operating data of process unit.

  17. A study on naphtha catalytic reforming reactor simulation and analysis

    Institute of Scientific and Technical Information of China (English)

    LIANG Ke-min; GUO Hai-yan; PAN Shi-wei

    2005-01-01

    A naphtha catalytic reforming unit with four reactors in series is analyzed. A physical model is proposed to describe the catalytic reforming radial flow reactor. Kinetics and thermodynamics equations are selected to describe the naphtha catalytic reforming reactions characteristics based on idealizing the complex naphtha mixture by representing the paraffin, naphthene, and aromatic groups by single compounds. The simulation results based above models agree very well with actual operation unit data.

  18. Competitiveness of small-medium reactors. A probabilistic study on the economy of scale factor

    International Nuclear Information System (INIS)

    Smaller size reactors are able to play an important role in the worldwide nuclear renaissance. The major disadvantage of those new reactors - the unit size - would label the small-medium size reactors as not economically competitive with larger plants. But, the economy of scale law applies only if the designs are similar, which is not the case here, since the SMRs are designed with original and innovative solutions not accessible to large size reactors: the IRIS reactor is used as an example of small medium reactors (SMR), but the analyses and conclusions are applicable to the whole spectrum of SMRs. The aim of this paper is to present latest advances in the differential economical assessment of Generation Cost of SMRs compared to LRs. The international literature has started to present studies focused on the two major differential accounts of Levelized Unit Electricity Cost - Captial Costs ($/kWe) and Operation and Maintenance Costs ($/kWh) - providing deterministic values for the main cost drivers (i.e. economy of scale, multiple units, learning during construction, design characteristics and modular build, shorter construction time for CC, economy of scale location of the plant, number of units, capacity factor, learning by doing, plant obsolescence for O and M costs). Since the modern SMR market is in the early stages of development, it is necessary to consider also the uncertainties associated to current estimates of those cost drivers. When available, the uncertainty has been integrated in the Open Model assigning a probabilistic distribution to the input value of each cost driver. As Far as other cost drivers are concerned, parametric analyses are still under development and uncertainty analyses are not available: thus, conservative but realistic values for both of them have been assumed. Some reasonable future scenarios have been assumed, considering the private operator perspective for a single plant investment and postulating, among the others

  19. Study on the transient behaviours of MNSR reactor for control rod withdrawal

    International Nuclear Information System (INIS)

    The transient behaviours of Miniature Neutron Source Reactor MNSR are analyzed and calculated with the reactor thermohydraulics RETRAN-02 program and the reactor physics MARIA program. The obtained event sequence and consequence from the calculation are compared with the experiments. The effective resonance integral for study on Doppler effect is taken into account. The reactivity temperature coefficient weighting factors are computed. The transient parameters related to reactor power peaking, coolant inlet temperatures, outlet temperatures and coolant mass flow, etc. are computed and compared with the experimental results. (6 refs., 2 figs., 5 tabs.)

  20. Study on the Neutrino Oscillation with a Next Generation Medium-Baseline Reactor Experiment

    International Nuclear Information System (INIS)

    For over fifty years, reactor experiments have played an important role in neutrino physics, in both discoveries and precision measurements. One of the methods to verify the existence of neutrino is the observation of neutrino oscillation phenomena. Electron antineutrinos emitted from a reactor provide the measurement of the small mixing angle θ13, providing rich programs of neutrino properties, detector development, nuclear monitoring, and application. Using reactor neutrinos, future reactor neutrino experiments, more precise measurements of θ12,Δm122, and mass hierarchy will be explored. The precise measurement of θ13 would be crucial for measuring the CP violation parameters at accelerators. Therefore, reactor neutrino physics will assist in the complete understanding of the fundamental nature and implications of neutrino masses and mixing. In this paper, we investigated several characteristics of RENO-50, which is a future medium-baseline reactor neutrino oscillation experiment, by using the GloBES simulation package

  1. Microbial ureolysis in the seawater-catalysed urine phosphorus recovery system: Kinetic study and reactor verification.

    Science.gov (United States)

    Tang, Wen-Tao; Dai, Ji; Liu, Rulong; Chen, Guang-Hao

    2015-12-15

    Our previous study has confirmed the feasibility of using seawater as an economical precipitant for urine phosphorus (P) precipitation. However, we still understand very little about the ureolysis in the Seawater-based Urine Phosphorus Recovery (SUPR) system despite its being a crucial step for urine P recovery. In this study, batch experiments were conducted to investigate the kinetics of microbial ureolysis in the seawater-urine system. Indigenous bacteria from urine and seawater exhibited relatively low ureolytic activity, but they adapted quickly to the urine-seawater mixture during batch cultivation. During cultivation, both the abundance and specific ureolysis rate of the indigenous bacteria were greatly enhanced as confirmed by a biomass-dependent Michaelis-Menten model. The period for fully ureolysis was decreased from 180 h to 2.5 h after four cycles of cultivation. Based on the successful cultivation, a lab-scale SUPR reactor was set up to verify the fast ureolysis and efficient P recovery in the SUPR system. Nearly complete urine P removal was achieved in the reactor in 6 h without adding any chemicals. Terminal Restriction Fragment Length Polymorphism (TRFLP) analysis revealed that the predominant groups of bacteria in the SUPR reactor likely originated from seawater rather than urine. Moreover, batch tests confirmed the high ureolysis rates and high phosphorus removal efficiency induced by cultivated bacteria in the SUPR reactor under seawater-to-urine mixing ratios ranging from 1:1 to 9:1. This study has proved that the enrichment of indigenous bacteria in the SUPR system can lead to sufficient ureolytic activity for phosphate precipitation, thus providing an efficient and economical method for urine P recovery.

  2. Simple analysis of an External Vessel Cooling Thermosyphon for a Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    KALIMER has three different DHR systems: two non-safety grade systems and one safety grade system. The non-safety grade systems are an IRACS (Intermediate Reactor Auxiliary Cooling System) and a steam/feedwater system. The safety grade system is a PDRC (Passive Decay Heat Removal Circuit). In case of the foreign reactor designs, ABTR (Advanced Burner Test Reactor) has a DRACS (Direct Reactor Auxiliary Cooling System), a PFBR (Indian Prototype Fast Breeder Reactor) has an SGDHRS (Safety Grade Decay Heat Removal System), and an EFR (European Fast Reactor) has DRC (Direct Reactor Cooling). Those designs have advantage on relatively high decay heat removal capacity. However, larger vessel size due to subsidiary in-vessel structure and possible accident propagation to reactor induced by sodium fire. In this paper, an ex-vessel thermosyphon design was proposed for the removal of decay heat for an iSFR. The proposed ex-vessel thermosyphon was designed to remove decay heat in both transient cases and BDBA cases, such as vessel failure. Proper working fluid was selected based on thermodynamic properties and chemical stability. Mercury was chosen as the working fluid, and SUS 314 was used for the corresponding structure material. Possible chemical reactions and adverse effects from using the thermosyphon were inherently eliminated by the system layout. A model for a high-temperature thermosyphon and numerical algorithms were used for the analysis. As a result of the simulation, the thermosyphon design was optimized, and it showed sufficient DHR performance to maintain core integrity

  3. Azo dye removal in a membrane-free up-flow biocatalyzed electrolysis reactor coupled with an aerobic bio-contact oxidation reactor

    Energy Technology Data Exchange (ETDEWEB)

    Cui, Dan; Guo, Yu-Qi; Cheng, Hao-Yi; Liang, Bin; Kong, Fan-Ying [State Key Laboratory of Urban Water Resource and Environment, Harbin Institute of Technology, No. 202 Haihe Road, Harbin 150090 (China); Lee, Hyung-Sool [Department of Civil and Environmental Engineering, University of Waterloo, 200 University Avenue West Waterloo, Ontario, Canada N2L 3G1 (Canada); Wang, Ai-Jie, E-mail: waj0578@hit.edu.cn [State Key Laboratory of Urban Water Resource and Environment, Harbin Institute of Technology, No. 202 Haihe Road, Harbin 150090 (China)

    2012-11-15

    Highlights: Black-Right-Pointing-Pointer A membrane-free up-flow biocatalyzed electrolysis reactor coupled with an aerobic bio-contact oxidation reactor was developed. Black-Right-Pointing-Pointer Alizarin Yellow R as the mode of azo dyes was efficiently converted to p-phenylenediamine (PPD) and 5-aminosalicylic acid (5-ASA). Black-Right-Pointing-Pointer PPD and 5-ASA were further oxidized in a bio-contact oxidation reactor. Black-Right-Pointing-Pointer The mechanism of UBER for azo dye removal was discussed. - Abstract: Azo dyes that consist of a large quantity of dye wastewater are toxic and persistent to biodegradation, while they should be removed before being discharged to water body. In this study, Alizarin Yellow R (AYR) as a model azo dye was decolorized in a combined bio-system of membrane-free, continuous up-flow bio-catalyzed electrolysis reactor (UBER) and subsequent aerobic bio-contact oxidation reactor (ABOR). With the supply of external power source 0.5 V in the UBER, AYR decolorization efficiency increased up to 94.8 {+-} 1.5%. Products formation efficiencies of p-phenylenediamine (PPD) and 5-aminosalicylic acid (5-ASA) were above 90% and 60%, respectively. Electron recovery efficiency based on AYR removal in cathode zone was nearly 100% at HRTs longer than 6 h. Relatively high concentration of AYR accumulated at higher AYR loading rates (>780 g m{sup -3} d{sup -1}) likely inhibited acetate oxidation of anode-respiring bacteria on the anode, which decreased current density in the UBER; optimal AYR loading rate for the UBER was 680 g m{sup -3} d{sup -1} (HRT 2.5 h). The subsequent ABOR further improved effluent quality. Overall the Chroma decreased from 320 times to 80 times in the combined bio-system to meet the textile wastewater discharge standard II in China.

  4. Azo dye removal in a membrane-free up-flow biocatalyzed electrolysis reactor coupled with an aerobic bio-contact oxidation reactor

    International Nuclear Information System (INIS)

    Highlights: ► A membrane-free up-flow biocatalyzed electrolysis reactor coupled with an aerobic bio-contact oxidation reactor was developed. ► Alizarin Yellow R as the mode of azo dyes was efficiently converted to p-phenylenediamine (PPD) and 5-aminosalicylic acid (5-ASA). ► PPD and 5-ASA were further oxidized in a bio-contact oxidation reactor. ► The mechanism of UBER for azo dye removal was discussed. - Abstract: Azo dyes that consist of a large quantity of dye wastewater are toxic and persistent to biodegradation, while they should be removed before being discharged to water body. In this study, Alizarin Yellow R (AYR) as a model azo dye was decolorized in a combined bio-system of membrane-free, continuous up-flow bio-catalyzed electrolysis reactor (UBER) and subsequent aerobic bio-contact oxidation reactor (ABOR). With the supply of external power source 0.5 V in the UBER, AYR decolorization efficiency increased up to 94.8 ± 1.5%. Products formation efficiencies of p-phenylenediamine (PPD) and 5-aminosalicylic acid (5-ASA) were above 90% and 60%, respectively. Electron recovery efficiency based on AYR removal in cathode zone was nearly 100% at HRTs longer than 6 h. Relatively high concentration of AYR accumulated at higher AYR loading rates (>780 g m−3 d−1) likely inhibited acetate oxidation of anode-respiring bacteria on the anode, which decreased current density in the UBER; optimal AYR loading rate for the UBER was 680 g m−3 d−1 (HRT 2.5 h). The subsequent ABOR further improved effluent quality. Overall the Chroma decreased from 320 times to 80 times in the combined bio-system to meet the textile wastewater discharge standard II in China.

  5. 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

  6. Study of risk reduction by improving operation of reactor core isolation cooling system

    International Nuclear Information System (INIS)

    The Fukushima Daiichi nuclear power plant fell into a station blackout (SBO) due to the earthquake and tsunami in which most of the core cooling systems were disabled. In the units 2 and 3, water injection to the core was performed only by water injection system with turbine driven pumps. In particular, it is inferred from observed plant parameters that the reactor core isolation cooling system (RCIC) continued its operation much longer than it was originally expected (8 hours). Since the preparation of safety measures did not work, the reactor core damaged. With a view to reduce risk of station blackout events in a BWR by accident management, this study investigated the efficacy of operation procedures that takes advantage of RCIC which can be operated with only equipment inside reactor building and does not require an AC power source. The efficacy was assessed in this study by two steps. The first step is a thermal hydraulic analysis with the RETRAN3D code to estimate the potential extension of duration of core cooling by RCIC and the second step is the estimation of time required for recovery of off-site power from experiences at nuclear power stations under the 3.11 earthquake. This study showed that it is possible to implement more reliable measures for accident termination and to greatly reduce the risk of SBO by the installation of accident management measures with use of RCIC for extension of core cooling under SBO conditions. (author)

  7. Artificial neural network used in the study of sensitivities in IRIS reactor pressurizer

    International Nuclear Information System (INIS)

    In general, the technique of sensibility analysis studies the behavior of the ratio between the variation of output results and the variation of input parameters. This study performed in the reactor pressurizer, which is a component responsible for control of the pressure inside the vessel, has fundamental importance in designing the security of any concept of advanced reactor. Above all, for its feature of passive action of the pressurizer (there is no spray), this analysis becomes a necessary step for safety and performance of the plant. The direct method through code MODPRESS, which represents the pressurizer model of the International Reactor Innovative and Secure (IRIS), has required large computational effort. Unlike this method, Artificial Neural Networks (ANN), beyond faster, do not require a typical linear behavior of the system. Moreover, they can also use experimental data for their training and learning. If the ANN are satisfactory in this theoretical case may be used for future mapping and forecasting of the behavior of various phenomena in both plant operation and small-scale experiment to be installed in CRCN-NE. Based on the results obtained in this study, one can conclude that the artificial neural networks are presented as an alternative to MODPRESS code, as well as a great tool to calculate the sensitivity coefficient. (author)

  8. DOE plutonium disposition study: Analysis of existing ABB-CE Light Water Reactors for the disposition of weapons-grade plutonium

    International Nuclear Information System (INIS)

    Core reactivity and basic fuel management calculations were conducted on the selected reactors (with emphasis on the System 80 units as being the most desirable choice). Methods used were identical to those reported in the Evolutionary Reactor Report. From these calculations, the basic mission capability was assessed. The selected reactors were studied for modification, such as the addition of control rod nozzles to increase rod worth, and internals and control system modifications that might also be needed. Other system modifications studied included the use of enriched boric acid as soluble poison, and examination of the fuel pool capacities. The basic geometry and mechanical characteristics, materials and fabrication techniques of the fuel assemblies for the selected existing reactors are the same as for System 80+. There will be some differences in plutonium loading, according to the ability of the reactors to load MOX fuel. These differences are not expected to affect licensability or EPA requirements. Therefore, the fuel technology and fuel qualification sections provided in the Evolutionary Reactor Report apply to the existing reactors. An additional factor, in that the existing reactor availability presupposes the use of that reactor for the irradiation of Lead Test Assemblies, is discussed. The reactor operating and facility licenses for the operating plants were reviewed. Licensing strategies for each selected reactor were identified. The spent fuel pool for the selected reactors (Palo Verde) was reviewed for capacity and upgrade requirements. Reactor waste streams were identified and assessed in comparison to uranium fuel operations. Cost assessments and schedules for converting to plutonium disposition were estimated for some of the major modification items. Economic factors (incremental costs associated with using weapons plutonium) were listed and where possible under the scope of work, estimates were made

  9. Reduction of Worldwide Plutonium Inventories Using Conventional Reactors and Advanced Fuels: A Systems Study

    Energy Technology Data Exchange (ETDEWEB)

    Krakowski, R.A., Bathke, C.G.

    1997-12-31

    The potential for reducing plutonium inventories in the civilian nuclear fuel cycle through recycle in LWRs of a variety of mixed oxide forms is examined by means of a cost based plutonium flow systems model. This model emphasizes: (1) the minimization of separated plutonium; (2) the long term reduction of spent fuel plutonium; (3) the optimum utilization of uranium resources; and (4) the reduction of (relative) proliferation risks. This parametric systems study utilizes a globally aggregated, long term (approx. 100 years) nuclear energy model that interprets scenario consequences in terms of material inventories, energy costs, and relative proliferation risks associated with the civilian fuel cycle. The impact of introducing nonfertile fuels (NFF,e.g., plutonium oxide in an oxide matrix that contains no uranium) into conventional (LWR) reactors to reduce net plutonium generation, to increase plutonium burnup, and to reduce exo- reactor plutonium inventories also is examined.

  10. Energy Neutral Phosphate Fertilizer Production Using High Temperature Reactors: A Philippine Case Study

    International Nuclear Information System (INIS)

    The Philippines may profit from extracting uranium (U) from phosphoric acid during fertilizer production in a way that the recovered U can be beneficiated and taken as raw material for nuclear reactor fuel. Used in a high temperature reactor (HTR) that provides electricity and/or process heat for fertilizer processing and U extraction, energy-neutral fertilizer production, an idea first proposed by Haneklaus et al.,is possible. This paper presents a first case study of the concept regarding a representative phosphate fertilizer plant in the Philippines and exemplary HTR designs (HTR50S and GTHTR300C) developed by the Japan Atomic Energy Agency (JAEA). Three different arrangements (version I-III), ranging from basic electricity supply to overall power supply including on site hydrogen production for ammonia conversion, are introduced and discussed

  11. Positron annihilation study of Fe-ion irradiated reactor pressure vessel model alloys

    Science.gov (United States)

    Chen, L.; Li, Z. C.; Schut, H.; Sekimura, N.

    2016-01-01

    The degradation of reactor pressure vessel steels under irradiation, which results from the hardening and embrittlement caused by a high number density of nanometer scale damage, is of increasingly crucial concern for safe nuclear power plant operation and possible reactor lifetime prolongation. In this paper, the radiation damage in model alloys with increasing chemical complexity (Fe, Fe-Cu, Fe-Cu-Si, Fe-Cu-Ni and Fe-Cu-Ni-Mn) has been studied by Positron Annihilation Doppler Broadening spectroscopy after 1.5 MeV Fe-ion implantation at room temperature or high temperature (290 oC). It is found that the room temperature irradiation generally leads to the formation of vacancy-type defects in the Fe matrix. The high temperature irradiation exhibits an additional annealing effect for the radiation damage. Besides the Cu-rich clusters observed by the positron probe, the results show formation of vacancy-Mn complexes for implantation at low temperatures.

  12. Fabrication of First Chinese Made Reactor for Oxosvnthesis of Acetic Acid in Xi'an

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    @@ The first set of Chinese made reactor for oxo-synthesis of acetic acid has been fabricated by the Xi'an Nuclear Equipment Company,Ltd.This reactor has been transported to the site of equipment installation at the acetic acid production project owned by Shandong Yimeng Company,Ltd.,which has shattered the long-time precedent of relying upon imported equipment.

  13. Experimental study of coal topping process in a downer reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wang, J.G.; Lu, X.S.; Yao, J.X.; Lin, W.G.; Cui, L.J. [Chinese Academy of Science, Beijing (China). Inst. of Processing Engineering

    2005-02-02

    Experiments were carried out in a downer reactor integrated in a circulating fluidized bed combustor to examine the performance of the coal topping process. The effects of reaction temperature and coal particle size on the product distribution and their compositions were determined. The experimental results show that an increase in temperature will increase the yields of gas and liquid product, and the liquid yield decreases with the increase in coal particle size. The experiments exhibit an optimal condition for the liquid product. When the pyrolysis temperature is 660{sup o}C and coal particle size is less than 0.2-8 mm, the yield of light tar (hexane-soluble fraction) reaches 7.5 wt % (dry coal basis). The light tar is composed of acid groups (57.1 wt %), crude gasoline (aliphatics) (12.9 wt %), aromatics (21.4 wt %), and polar and basic groups (8.6 wt %). The experiments indicate that the coal topping process is a promising technology for partially converting coal into liquid fuels and fine chemicals.

  14. Study on an Automatic Knife Switch of New Type High-Voltage Small Current Reactor%新型高电压小电流电抗器自动刀闸的研究

    Institute of Scientific and Technical Information of China (English)

    许杰; 李世武; 孙伟

    2014-01-01

    针对当前电抗器投切刀闸存在的投切行程小、适用范围受限制、远程操控不方便等问题,研发了一种新型的高电压小电流电抗器自动刀闸,给出了自动刀闸结构图,分析了其工作原理。对该自动刀闸进行峰值耐受电流试验和机械操作试验,测试结果符合要求,有效通断率100%,提高了电抗器串并联自动化程度,易于远程控制且性能稳定,可广泛用于电力系统机构的各项试验。%Aiming at several problems such that the existing reactor knife switch is small in travel distance for knife making and breaking, limited in applicable range, not convenient in remote operating etc., this paper developed an automatic knife switch of new type high-voltage small current reactor and gave the structural diagram of the automatic knife switch, analyzing its working principle. The peak value withstand current and mechanical operating tests were carried out for the automatic knife switch. The test results are in conformity to the requirements, with 100% valid making-breaking rate, which raises the reactor serial and parallel connection automation degree, easy for remote control and stable in performance. The switch could be widely used in each test of electric power system organization.

  15. Innovative concept for an ultra-small nuclear thermal rocket utilizing a new moderated reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nam, Seung Hyun; Venneri, Paolo; Kim, Yong Hee; Lee, Jeong Ik; Chang, Soon Heung; Jeong, Yong Hoon [Dept. of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2015-10-15

    Although the harsh space environment imposes many severe challenges to space pioneers, space exploration is a realistic and profitable goal for long-term humanity survival. One of the viable and promising options to overcome the harsh environment of space is nuclear propulsion. Particularly, the Nuclear Thermal Rocket (NTR) is a leading candidate for near-term human missions to Mars and beyond due to its relatively high thrust and efficiency. Traditional NTR designs use typically high power reactors with fast or epithermal neutron spectrums to simplify core design and to maximize thrust. In parallel there are a series of new NTR designs with lower thrust and higher efficiency, designed to enhance mission versatility and safety through the use of redundant engines (when used in a clustered engine arrangement) for future commercialization. This paper proposes a new NTR design of the second design philosophy, Korea Advanced NUclear Thermal Engine Rocket (KANUTER), for future space applications. The KANUTER consists of an Extremely High Temperature Gas cooled Reactor (EHTGR) utilizing hydrogen propellant, a propulsion system, and an optional electricity generation system to provide propulsion as well as electricity generation. The innovatively small engine has the characteristics of high efficiency, being compact and lightweight, and bimodal capability. The notable characteristics result from the moderated EHTGR design, uniquely utilizing the integrated fuel element with an ultra heat-resistant carbide fuel, an efficient metal hydride moderator, protectively cooling channels and an individual pressure tube in an all-in-one package. The EHTGR can be bimodally operated in a propulsion mode of 100 MW{sub th} and an electricity generation mode of 100 kW{sub th}, equipped with a dynamic energy conversion system. To investigate the design features of the new reactor and to estimate referential engine performance, a preliminary design study in terms of neutronics and

  16. Análisis de Ruido en Reactores PWR

    OpenAIRE

    Bermejo Piñar, Juan Antonio

    2015-01-01

    Esta tesis se ha llevado a cabo persiguiendo dos objetivos principales: uno de ellos es el desarrollo y la aplicación de modelos para el mantenimiento predictivo de sensores en centrales nucleares, y el otro es profundizar en el entendimiento de los fenómenos que tienen influencia en el ruido de la señal de los detectores de neutrones de los reactores de agua a presión con ayuda de herramientas de simulación 3D. Para el desarrollo de los trabajos se ha contado con medidas de ruido de reactore...

  17. Experimental Studies on Assemblies 1 and 2 of the Fast Reactor FR-0. Part 2

    International Nuclear Information System (INIS)

    In a first part of this report, published as AE-195, an account was given of critical mass determinations and measurements of flux distribution and reaction ratios in the first assemblies of the fast zero power reactor FR0. This second part of the report deals with various investigations involving the measurement of reactivity. Control rod calibrations have been made using the positive period, the inverse multiplication, the rod drop and the pulsed source techniques, and show satisfactory agreement between the various methods. The reactivity worths of samples of different materials and different sizes have been measured at the core centre. Comparisons with perturbation calculations show that the regular and adjoint fluxes are well predicted in the central region of the core. The variation in the prompt neutron life-time with reactivity has been studied by means of the pulsed source and the Rossi-α techniques. Comparison with one region calculations reveals large discrepancies, indicating that this simple model is inadequate. Some investigations of streaming effects in an empty channel in the reactor and of interaction effects between channels have been made and are compared with theoretical estimates. Measurements of the reactivity worth of an air gap between the reactor halves and of the temperature coefficient are also described in the report. The work has been performed as a joint effort by AB Atomenergi and the Research Institute of National Defence

  18. Intermediate Heat Exchanger (IHX) Trade Study in Support of Advanced Fast Reactor-100 Development

    International Nuclear Information System (INIS)

    Reactor fabrication costs are closely linked to the commodities utilized in their construction. One large component of many pool-type sodium fast reactor designs that can significantly impact the overall size of the reactor vessel is the Intermediate Heat Exchanger (IHX). On this basis, a trade study was performed to evaluate four different types of novel IHX designs from the viewpoint of minimizing the impact on the size of the reactor. The four designs that were evaluated as part of this study were: i) conventional shell-with-straight tube design, ii) shell with twisted tube design, iii) Printed Circuit Heat Exchanger (PCHE), and iv) a spiral tube design. The approach was to develop simple thermal-hydraulics models in order to provide a technical basis for assessing the overall size of the units. Simple mechanical stress models were also developed so that component thicknesses could be calculated. This allowed material volumes and commodities for the various designs to be estimated. The results indicate that adopting a twisted vs. straight tube design can reduce commodities by ~ 30 %, and overall height of the units by roughly 30-40%. Aside from the choice of tube design, geometry also plays an important role in reducing commodity and space requirements; i.e., commodities for a kidney-shaped design were found to be ~10% less relative to the circular case, and the reduction is even more (~20%) for an annular design. If a kidney-shaped design is adopted, then the radial thickness of the vessel occluded by the IHXs can be reduced by 20%. Conversely, the reduction for annular designs approaches 75%. Height requirements are also reduced, and this is principally due to the smaller radius of the elliptical heads that are required in the radially thinned designs. Additional costs that may be incurred by fabricating a more complex design have not been factored into these analyses. (author)

  19. Jules Horowitz reactor. France development of an experimental loop integrating an optimized irradiation process

    International Nuclear Information System (INIS)

    Experimental reactors in the world enable researchers to meet the needs of industry and institutions not only by providing support to the existing nuclear infrastructure (Gen.2) but also by preparing the future generation (Gen.3, Gen.4) or even by responding to other needs as well (supports with fusion, medical applications). It is for this specific purpose that the Jules Horowitz Polyvalent Irradiation Reactor is now being built at the CEA Cadarache Research Center (located in the south of France). This Material Testing Reactor (MTR type) is designed to irradiate materials or fuel samples for various experimental tests. The reactor will also produce Mo99 radioelements that will supply 25% to 50% of current European needs. The goal of this paper is to describe a fuel irradiation loop, now under study, that will be designed to carry out power ramps tests to provide technical support to the Generation 2 and 3 nuclear power plants. In order to increase its irradiation capacity (2 to 3 per cycle), this loop takes into account the requirements that will lead to the optimization of all experimental processes in the facility (such as non-destructive examinations before and after the test, specific loading tools). All these considerations are being taken into account in order to offer to the customer's complete and optimized conditions in terms of experimental irradiations processes. (author)

  20. TIBER (Tokamak Ignition/Burn Experimental Reactor) II as a precursor to an international thermonuclear experimental reactor

    International Nuclear Information System (INIS)

    The Tokamak Ignition/Burn Experimental Reactor (TIBER) was pursued in the US as one option for an International Thermonuclear Experimental Reactor (ITER). This concept evolved from earlier work on the Tokamak Fusion Core Experiment (TFCX) to develop a small, ignited tokamak. While the copper-coil versions of TFCX became the short-pulsed, 1.23-m radius, Compact Ignition Tokamak (CIT), the superconducting TIBER with long pulse or steady state and a 2.6-m radius was considered for international collaboration. Recently the design was updated to TIBER II, to accommodate more conservative confinement scaling, double-poloidal divertors for impurity control, steady-state current drive, and nuclear testing. 18 refs., 1 fig

  1. Reactor physics studies in the GCFR Phase III critical assembly

    International Nuclear Information System (INIS)

    The third phase of the gas cooled fast reactor (GCFR) program, ZPR-9 Assembly 30, is based on a multi-zoned core of PuO2-UO2 with radial and axial blankets of UO2. Studies performed in this assembly will be compared to the previous phases of the GCFR program and will help to define parameters in this power-flattened demonstration plant-type core. Measurements in the Phase III program included small sample reactivity worths of various materials, central reaction rates and reaction rate distributions, absorption-to-fission ratios and the central point conversion ratio and the worth of steam entry into a small central zone. The reactivity change associated with the construction of a central pin zone in the core and axial blanket was measured. Reaction rate and steam entry measurements were repeated in the pin environment. Standard analysis methods using ENDF/B-IV data are described and the results are compared to measurements performed during the program

  2. Comparison study on cell calculation method of fast reactor

    International Nuclear Information System (INIS)

    Effective cross sections obtained by cell calculations are used in core calculations in current deterministic methods. Therefore, it is important to calculate the effective cross sections accurately and several methods have been proposed. In this study, some of the methods are compared to each other using a continuous energy Monte Carlo method as a reference. The result shows that the table look-up method used in Japan Nuclear Cycle Development Institute (JNC) sometimes has a difference over 10% in effective microscopic cross sections and be inferior to the sub-group method. The problem was overcome by introducing a new nuclear constant system developed in JNC, in which the ultra free energy group library is used. The system can also deal with resonance interaction effects between nuclides which are not able to be considered by other methods. In addition, a new method was proposed to calculate effective cross section accurately for power reactor fuel subassembly where the new nuclear constant system cannot be applied. This method uses the sub-group method and the ultra fine energy group collision probability method. The microscopic effective cross sections obtained by this method agree with the reference values within 5% difference. (author)

  3. Uses of reactor neutrons for studying the microcomposition of materials

    International Nuclear Information System (INIS)

    Reactor neutrons constitute excellents 'probes' for exploring and measuring a wide range both of minor and trace constituents in solids and liquids with high sensitivity because of their transparency in materials. Nondestructive neutron prompt-gamma analysis (PGA) utilizing either cold or thermal neutrons, such as at JRR-3M, is compared and contrasted to the more common (delayed) instrumental neutron activation analysis (INAA) and epithermal NAA. Clearly PGA offers high sensitivity for selected elements: B, H, Cd and REE's in suitable matrices, and is therefore, complementary to INAA which is not as useful for them, or for Ni, Sn, Fe, C or N. Recent INAA applications in our laboratory that demonstrate some of the uniqueness of neutron methods include use of epithermal neutrons for small biological specimens to measure Cd, K, As, Zn and, multielemental INAA for environmental pollution studies. The latter involves large data sets of multielemental concentrations which are subjected to statistical multivariant factor analysis to reveal unknown or unsuspected quantitative relationships among groups of trace constituents. These patterns, or 'factors' are shown to be uniquely related to pollution sources and can be utilized to compute the relative source contributions at a given receptor site. (author)

  4. Flow through reactors for organic chemistry: directly electrically heated tubular mini reactors as an enabling technology for organic synthesis

    Directory of Open Access Journals (Sweden)

    Ulrich Kunz

    2009-11-01

    Full Text Available Until recently traditional heating in organic chemistry has been done with oil heating baths or using electric heat exchangers. With the advent of microwave equipment, heating by microwaves was rapidly introduced as standard method in organic chemistry laboratories, mainly because of the convenient possibility to operate at high temperature accompanied by accelerated reaction rates. In the present contribution we discuss the method of heating small, continuously operated reactors by passing electric current directly through the reactor wall as an enabling technology in organic chemistry. The benefit of this method is that the heat is generated directly inside the reactor wall. By this means high heating rates comparable to microwave ovens can be reached but at much lower cost for the equipment. A tool for the comparison of microwave heating and traditional heating is provided. As an example kinetic data for the acid catalyzed hydrolysis of methyl formate were measured using this heating concept. The reaction is not only a suitable model but also one of industrial importance since this is the main production process for formic acid.

  5. Flow through reactors for organic chemistry: directly electrically heated tubular mini reactors as an enabling technology for organic synthesis.

    Science.gov (United States)

    Kunz, Ulrich; Turek, Thomas

    2009-11-30

    Until recently traditional heating in organic chemistry has been done with oil heating baths or using electric heat exchangers. With the advent of microwave equipment, heating by microwaves was rapidly introduced as standard method in organic chemistry laboratories, mainly because of the convenient possibility to operate at high temperature accompanied by accelerated reaction rates. In the present contribution we discuss the method of heating small, continuously operated reactors by passing electric current directly through the reactor wall as an enabling technology in organic chemistry. The benefit of this method is that the heat is generated directly inside the reactor wall. By this means high heating rates comparable to microwave ovens can be reached but at much lower cost for the equipment. A tool for the comparison of microwave heating and traditional heating is provided. As an example kinetic data for the acid catalyzed hydrolysis of methyl formate were measured using this heating concept. The reaction is not only a suitable model but also one of industrial importance since this is the main production process for formic acid.

  6. Flow through reactors for organic chemistry: directly electrically heated tubular mini reactors as an enabling technology for organic synthesis

    Science.gov (United States)

    Turek, Thomas

    2009-01-01

    Summary Until recently traditional heating in organic chemistry has been done with oil heating baths or using electric heat exchangers. With the advent of microwave equipment, heating by microwaves was rapidly introduced as standard method in organic chemistry laboratories, mainly because of the convenient possibility to operate at high temperature accompanied by accelerated reaction rates. In the present contribution we discuss the method of heating small, continuously operated reactors by passing electric current directly through the reactor wall as an enabling technology in organic chemistry. The benefit of this method is that the heat is generated directly inside the reactor wall. By this means high heating rates comparable to microwave ovens can be reached but at much lower cost for the equipment. A tool for the comparison of microwave heating and traditional heating is provided. As an example kinetic data for the acid catalyzed hydrolysis of methyl formate were measured using this heating concept. The reaction is not only a suitable model but also one of industrial importance since this is the main production process for formic acid. PMID:20300506

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-09-01

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

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

    International Nuclear Information System (INIS)

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

  9. Natural fission reactors from Gabon. Contribution to the study of the conditions of stability of a natural radioactive wastes storage site (2 Ga)

    International Nuclear Information System (INIS)

    The natural fission reactors of Oklo consists of a core of uraninite (60%) with fission products, embedded in a pure clay matrix. Thus, the aim of geological, mineral, and geochemical studies of the Oklo Reactors is to assess the behaviour of fission products in an artificial waste depository. Previous studies have shown that Reactor Zone 10, located in the Oklo mine, represents an example for an exceptional confinement of fission products since 2 Ga. In reactor Zone 9, located in Oklo open pit, migrations are more important. Reactor ZOne 13 was influenced by a thermal event due to a doleritic intrusion, located some twenty meters far away, one Ga years after fission reaction operations. In this study,we characterized temperature and redox conditions of fluids by using stable isotopes of uraninites and clays. Moreover mineralogical and chemical characteristics were defined. (author)

  10. Performance studies of a new core cooling monitor in a boiling water reactor

    International Nuclear Information System (INIS)

    Performance studies of a new type of core cooling monitors have been carried out in the Barsebaeck Nuclear Power Station during the operation periods 1988-10-04 to 1989-07-05, 1989-08-03 to 1990-09-05 and 1990-09-28 to 1991-07-04. The results showed that the monitors, which were placed inside the reactor core, are very sensitive to variations of the reactor operating conditions, and that 34 months of irradiation did not influence the signals from the monitors. Experiments were also carried out in a 160 bar loop, where sudden uncovers of the monitors were achieved by decreasing the liquid level of the coolant surrounding the monitors. The experiments included the pressures of 5, 20, 50, 70 and 155 bar, and the responses to uncover were in the ranges between 11 and 82 mV/sec or a total step change of 2 V at typical BWR conditions. This is of the order of two decades higher than the responses from monitors based on thermocouple readings. The monitors can be operated in two modes, the core cooling mode and the temperature mode. In the former mode the electrical current is 3-4 A, and in the latter mode, where the monitor actually serves as a thermometer, the current is in the order of 50-100 mA. In the laboratory the monitors have been studied for temperatures up to 1265 deg. C, which is very useful in case of a severe reactor accident. Thus, during such events the temperatures in the reactor core could be followed up to this level and the monitors could also be used to activate certain safety equipment. The function as well as the design of the instrument is verified in laboratory experiments, computer calculations and reactor tests and is now ready for implementation in the BWR instrumentation. In summary: 1. The proposed monitor can operate in two modes; the core cooling mode and the temperature mode. 2. Laboratory studies have shown that the responses to uncover are two decades higher than signals from monitors based on thermocouple readings. 3. No effects of

  11. Light a CANDLE. An innovative burnup strategy of nuclear reactors

    International Nuclear Information System (INIS)

    CANDLE is a new burnup strategy for nuclear reactors, which stands for Constant Axial Shape of Neutron Flux, Nuclide Densities and Power Shape During Life of Energy Production. When this candle-like burnup strategy is adopted, although the fuel is fixed in a reactor core, the burning region moves, at a speed proportionate to the power output, along the direction of the core axis without changing the spatial distribution of the number density of the nuclides, neutron flux, and power density. Excess reactivity is not necessary for burnup and the shape of the power distribution and core characteristics do not change with the progress of burnup. It is not necessary to use control rods for the control of the burnup. This booklet described the concept of the CANDLE burnup strategy with basic explanations of excess neutrons and its specific application to a high-temperature gas-cooled reactor and a fast reactor with excellent neutron economy. Supplementary issues concerning the initial core and high burnup were also referred. (T. Tanaka)

  12. Downer reactor: From fundamental study to industrial application

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, Y.; Wu, C.N.; Zhu, J.X.; Wei, F.; Jin, Y. [Tsinghua University, Beijing (China). Dept. of Chemical Engineering

    2008-04-15

    Downer reactor, in which gas and solids move downward co-currently, has unique features such as the plug-flow reactor performance and relatively uniform flow structure compared to other gas-solids fluidized bed reactors, e.g., bubbling bed, turbulent bed and riser. Downer is therefore acknowledged as a novel multiphase flow reactor with great potential in high-severity operated processes, such as the high temperature, ultra-short contact time reactions with the intermediates as the desired products. Typical process developments in industry have directed to (1) the new-generation refinery process for cracking of heavier feedstock to gasoline and light olefins (e.g., propylene) as by-products; and (2) coal pyrolysis in hydrogen plasma which opens up a direct means for producing acetylene, i.e., a new route to synthesize chemicals from a clean coal utilization process. This paper gives a comprehensive review on the development of fundamental research on downer reactors as well as the particular industrial demonstrations for the fluid catalytic cracking (FCC) of heavy oils and coal pyrolysis in thermal plasma.

  13. A Study on Dismantling of Westinghouse Type Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Woo-Tae; Lee, Sang-Guk [KHNP-CRI, Daejeon (Korea, Republic of)

    2014-10-15

    KHNP started a research project this year to develop a methodology to dismantle nuclear reactors and internals. In this paper, we reviewed 3D design model of the reactor and suggested feasible cutting scheme.. Using 3-D CAD model of Westinghouse type nuclear reactor and its internals, we reviewed possible options for disposal. Among various options of dismantling the nuclear reactor, plasma cutting was selected to be the best feasible and economical method. The upper internals could be segmented by using a band saw. It is relatively fast, and easily maintained. For cutting the lower internals, plasma torch was chosen to be the best efficient tool. Disassembling the baffle and the former plate by removing the baffle former bolts was also recommended for minimizing storage volume. When using plasma torch for cutting the reactor vessel and its internal, installation of a ventilation system for preventing pollution of atmosphere was recommended. For minimizing radiation exposure during the cutting operation, remotely controlled robotic tool was recommended to be used.

  14. A numerical study of boiling flow instability of a reactor thermosyphon system

    Energy Technology Data Exchange (ETDEWEB)

    Nayak, A.K.; Lathouwers, D.; Hagen, T.H.J.J. van der [Interfaculty Reactor Institute, Delft University of Technology, Mekelweg 15, 2629 JB Delft (Netherlands); Schrauwen, Frans; Molenaar, Peter; Rogers, Andrew [Shell Research and Technology Centre, Badhuisweg 3, 1031 CM Amsterdam (Netherlands)

    2006-04-01

    A numerical study has been carried out to investigate the boiling flow instability of a reactor thermosyphon system. The numerical model solves the conservation equations of mass, momentum and energy applicable to a two-fluid and three-field steam-water system using a finite difference technique. The computer code MONA was used for this purpose. The code was applied to the thermosyphon system of an EO (ethylene oxide) chemical reactor in which the heat released by a catalytic reaction is carried by boiling water under natural circulation conditions. The steady-state characteristics of the reactor thermosyphon system were predicted using the MONA code and conventional two-phase flow models in order to understand the model applicability for this type of thermosyphon system. The two-fluid model was found to predict the flow closest to the measured value of the plant. The stability behaviour of the thermosyphon system was investigated for a wide range of operating conditions. The effects of power, subcooling, riser length and riser diameter on the boiling flow instability were determined. The system was found to be unstable at higher power conditions which is typical for a Type II instability. However, with an increase in riser diameter, oscillations at low power were observed as well. These are classified as Type I instabilities. Stability maps were predicted for both Type I and Type II instabilities. Methods of improving the stability of the system are discussed. [Author].

  15. Fuel, Structural Material and Coolant for an Advanced Fast Micro-Reactor

    Science.gov (United States)

    Do Nascimento, J. A.; Duimarães, L. N. F.; Ono, S.

    The use of nuclear reactors in space, seabed or other Earth hostile environment in the future is a vision that some Brazilian nuclear researchers share. Currently, the USA, a leader in space exploration, has as long-term objectives the establishment of a permanent Moon base and to launch a manned mission to Mars. A nuclear micro-reactor is the power source chosen to provide energy for life support, electricity for systems, in these missions. A strategy to develop an advanced micro-reactor technologies may consider the current fast reactor technologies as back-up and the development of advanced fuel, structural and coolant materials. The next generation reactors (GEN-IV) for terrestrial applications will operate with high output temperature to allow advanced conversion cycle, such as Brayton, and hydrogen production, among others. The development of an advanced fast micro-reactor may create a synergy between the GEN-IV and space reactor technologies. Considering a set of basic requirements and materials properties this paper discusses the choice of advanced fuel, structural and coolant materials for a fast micro-reactor. The chosen candidate materials are: nitride, oxide as back-up, for fuel, lead, tin and gallium for coolant, ferritic MA-ODS and Mo alloys for core structures. The next step will be the neutronic and burnup evaluation of core concepts with this set of materials.

  16. HYDROGEN KINETICS LIMITATION OF AN AUTOTROPHIC SULPHATE REDUCTION REACTOR

    Directory of Open Access Journals (Sweden)

    CÉSAR SÁEZ-NAVARRETE

    2012-01-01

    Full Text Available El uso de sustratos inorgánicos podría reducir los costos y simplificar la operación de sistemas de tratamiento de aguas que utilizan bacterias reductoras de sulfato. Sin embargo, el uso de H2 como sustrato energético y la bioproducción de H2S podrían provocar limitaciones cinéticas. El objetivo de este estudio fue evaluar las condiciones en las que la capacidad de transferencia de masa de un bioreactor de reducción de sulfato, limita su cinética de reducción. La cinética del reactor fue obtenida monitoreando la presión del sistema en condiciones de no limitación por sulfato. Se concluyó que el diseño del bioreactor debería basarse en sus propiedades de transferencia. La tasa de consumo de H2 alcanzó un máximo de 10-4 M/min, para una tasa de reducción de sulfato de 3.4 g·L-1·d-1. Para evitar limitación por H2 se requirió un kLa de 1.48 min-1 a 1.2·109 cells/L (1.23·10-9 L·min-1·cell-1, valor relevante para propósitos de escalamiento.

  17. Passive Safety Systems in Advanced Water Cooled Reactors (AWCRS). Case Studies. A Report of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO)

    International Nuclear Information System (INIS)

    This report presents the results from the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) collaborative project (CP) on Advanced Water Cooled Reactor Case Studies in Support of Passive Safety Systems (AWCR), undertaken under the INPRO Programme Area C. INPRO was launched in 2000 - on the basis of a resolution of the IAEA General Conference (GC(44)/RES/21) - to ensure that nuclear energy is available in the 21st century in a sustainable manner, and it seeks to bring together all interested Member States to consider actions to achieve innovation. An important objective of nuclear energy system assessments is to identify 'gaps' in the various technologies and corresponding research and development (R and D) needs. This programme area fosters collaboration among INPRO Member States on selected innovative nuclear technologies to bridge technology gaps. Public concern about nuclear reactor safety has increased after the Fukushima Daiichi nuclear power plant accident caused by the loss of power to pump water for removing residual heat in the core. As a consequence, there has been an increasing interest in designing safety systems for new and advanced reactors that are passive in nature. Compared to active systems, passive safety features do not require operator intervention, active controls, or an external energy source. Passive systems rely only on physical phenomena such as natural circulation, thermal convection, gravity and self-pressurization. Passive safety features, therefore, are increasingly recognized as an essential component of the next-generation advanced reactors. A high level of safety and improved competitiveness are common goals for designing advanced nuclear power plants. Many of these systems incorporate several passive design concepts aimed at improving safety and reliability. The advantages of passive safety systems include simplicity, and avoidance of human intervention, external power or signals. For these reasons, most

  18. Study and Evaluation of Innovative Fuel Handling Systems for Sodium-Cooled Fast Reactors: Fuel Handling Route Optimization

    Directory of Open Access Journals (Sweden)

    Franck Dechelette

    2014-01-01

    Full Text Available The research for technological improvement and innovation in sodium-cooled fast reactor is a matter of concern in fuel handling systems in a view to perform a better load factor of the reactor thanks to a quicker fuelling/defueling process. An optimized fuel handling route will also limit its investment cost. In that field, CEA has engaged some innovation study either of complete FHR or on the optimization of some specific components. This paper presents the study of three SFR fuel handling route fully described and compared to a reference FHR option. In those three FHR, two use a gas corridor to transfer spent and fresh fuel assembly and the third uses two casks with a sodium pot to evacuate and load an assembly in parallel. All of them are designed for the ASTRID reactor (1500 MWth but can be extrapolated to power reactors and are compatible with the mutualisation of one FHS coupled with two reactors. These three concepts are then intercompared and evaluated with the reference FHR according to four criteria: performances, risk assessment, investment cost, and qualification time. This analysis reveals that the “mixed way” FHR presents interesting solutions mainly in terms of design simplicity and time reduction. Therefore its study will be pursued for ASTRID as an alternative option.

  19. A Conceptual Study of a Supercritical CO2-Cooled Micro Modular Reactor

    Directory of Open Access Journals (Sweden)

    Hwanyeal Yu

    2015-12-01

    Full Text Available A neutronics conceptual study of a supercritical CO2-cooled micro modular reactor (MMR has been performed in this work. The suggested MMR is an extremely compact and truck-transportable nuclear reactor. The thermal power of the MMR is 36.2 MWth and it is designed to have a 20-year lifetime without refueling. A salient feature of the MMR is that all the components including the generator are integrated in a small reactor vessel. For a minimal volume and long lifetime of the MMR core, a fast neutron spectrum is utilized in this work. To enhance neutron economy and maximize the fuel volume fraction in the core, a high-density uranium mono-nitride U15N fuel is used in the fast-spectrum MMR. Unlike the conventional supercritical CO2-cooled fast reactors, a replaceable fixed absorber (RFA is introduced in a unique way to minimize the excess reactivity and the power peaking factor of the core. For a compact core design, the drum-type control absorber is adopted as the primary reactivity control mechanism. In this study, the neutronics analyses and depletions have been performed by using the continuous energy Monte Carlo Serpent code with the evaluated nuclear data file ENDF/B-VII.1 Library. The MMR core is characterized in view of several important safety parameters such as control system worth, fuel temperature coefficient (FTC and coolant void reactivity (CVR, etc. In addition, a preliminary thermal-hydraulic analysis has also been performed for the hottest channel of the Korea Advanced Institute of Science and Technology (KAIST MMR.

  20. A study on future nuclear reactor technology and development strategy

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. Y.; Kim, S. H.; Sohn, D. S.; Suk, S. D.; Zee, S. K.; Yang, M. H.; Kim, H. J.; Park, W. S

    2000-12-01

    Development of nuclear reactor and fuel cycle technology for future is essential to meet the current issues such as enhancement of nuclear power reactor safety, economically competitive with gas turbine power generation, less production of radioactive waste, proliferation resistant fuel cycle, and public acceptance in consideration of lack of energy resources in the nuclear countries worldwide as well as in Korea. This report deals with as follows, 1) Review the world energy demand and supply perspective and analyse nature of energy and sustainable development to set-up nuclear policy in Korea 2) Recaptitulate the current long term nuclear R and D activities 3) Review nuclear R and D activities and programs of USA, Japan, France, Russia, international organizations such as IAEA, OECD/NEA 4) Recommend development directions of nuclear reactors and fuels.