WorldWideScience

Sample records for reactor conditions invited

  1. Nonperturbative measurement of the local magnetic field using pulsed polarimetry for fusion reactor conditions (invited)

    International Nuclear Information System (INIS)

    Smith, Roger J.

    2008-01-01

    A novel diagnostic technique for the remote and nonperturbative sensing of the local magnetic field in reactor relevant plasmas is presented. Pulsed polarimetry [Patent No. 12/150,169 (pending)] combines optical scattering with the Faraday effect. The polarimetric light detection and ranging (LIDAR)-like diagnostic has the potential to be a local B pol diagnostic on ITER and can achieve spatial resolutions of millimeters on high energy density (HED) plasmas using existing lasers. The pulsed polarimetry method is based on nonlocal measurements and subtle effects are introduced that are not present in either cw polarimetry or Thomson scattering LIDAR. Important features include the capability of simultaneously measuring local T e , n e , and B || along the line of sight, a resiliency to refractive effects, a short measurement duration providing near instantaneous data in time, and location for real-time feedback and control of magnetohydrodynamic (MHD) instabilities and the realization of a widely applicable internal magnetic field diagnostic for the magnetic fusion energy program. The technique improves for higher n e B || product and higher n e and is well suited for diagnosing the transient plasmas in the HED program. Larger devices such as ITER and DEMO are also better suited to the technique, allowing longer pulse lengths and thereby relaxing key technology constraints making pulsed polarimetry a valuable asset for next step devices. The pulsed polarimetry technique is clarified by way of illustration on the ITER tokamak and plasmas within the magnetized target fusion program within present technological means.

  2. Nonperturbative measurement of the local magnetic field using pulsed polarimetry for fusion reactor conditions (invited).

    Science.gov (United States)

    Smith, Roger J

    2008-10-01

    A novel diagnostic technique for the remote and nonperturbative sensing of the local magnetic field in reactor relevant plasmas is presented. Pulsed polarimetry [Patent No. 12/150,169 (pending)] combines optical scattering with the Faraday effect. The polarimetric light detection and ranging (LIDAR)-like diagnostic has the potential to be a local B(pol) diagnostic on ITER and can achieve spatial resolutions of millimeters on high energy density (HED) plasmas using existing lasers. The pulsed polarimetry method is based on nonlocal measurements and subtle effects are introduced that are not present in either cw polarimetry or Thomson scattering LIDAR. Important features include the capability of simultaneously measuring local T(e), n(e), and B(parallel) along the line of sight, a resiliency to refractive effects, a short measurement duration providing near instantaneous data in time, and location for real-time feedback and control of magnetohydrodynamic (MHD) instabilities and the realization of a widely applicable internal magnetic field diagnostic for the magnetic fusion energy program. The technique improves for higher n(e)B(parallel) product and higher n(e) and is well suited for diagnosing the transient plasmas in the HED program. Larger devices such as ITER and DEMO are also better suited to the technique, allowing longer pulse lengths and thereby relaxing key technology constraints making pulsed polarimetry a valuable asset for next step devices. The pulsed polarimetry technique is clarified by way of illustration on the ITER tokamak and plasmas within the magnetized target fusion program within present technological means.

  3. Invitation

    CERN Multimedia

    2004-01-01

    Monday, 11 October Gala-Concert to celebrate 50 years of CERN Jack Liebeck (violin) and Katya Apekisheva (piano) Works by C. Debussy, S. Prokofiev, and L. van Beethoven CERN, Main Auditorium, 20:00 For CERN staff and their families With the assistance of PPARC (UK) Free tickets available at the CERN Press Office (50-1-048). Maximum four tickets per person. For more information on the performers, see: http://www.jackliebeck.com/CERN50/JL.htm http://www.jackliebeck.com/CERN50/KA.htm The "WIENER SCHUBERTBUND" celebrates CERN's 50th "DEUTSCHE MESSE" By Franz Schubert performed by the "WIENER SCHUBERTBUND" On the occasion of the Organization's Fiftieth Anniversary, Mr Hans-Walter Gérard Schober, Honorary Consul-General of Austria in Geneva, is pleased to invite all those working at CERN to a mass at which Franz Schubert's German Mass will be sung by the Vienna Schubert Choral Society. This renown men's choir, founded in 1863. Eminent composers, including Richard Strauss, Wilhelm Kinzel, a...

  4. Conditioning of nuclear reactor fuel

    International Nuclear Information System (INIS)

    1975-01-01

    A method of conditioning the fuel of a nuclear reactor core to minimize failure of the fuel cladding comprising increasing the fuel rod power to a desired maximum power level at a rate below a critical rate which would cause cladding damage is given. Such conditioning allows subsequent freedom of power changes below and up to said maximum power level with minimized danger of cladding damage. (Auth.)

  5. Invited talk on ageing management of boiling water reactors (BWRs)

    International Nuclear Information System (INIS)

    Prasad, Y.S.R.; Srinivasan, V.S.

    1994-01-01

    A nuclear power plant is built with a certain design life but by managing the operation of the plant with a well designed in-service inspection, repair and replacement programme of the equipment as required we will be able to extend the operation of the plant well beyond it's design life. This is also economically a paying proposition in view of the astronomical cost of construction of a new plant of equivalent capacity. In view of this, there is a growing trend the world over to study the ageing phenomena, especially in respect of nuclear power plant equipment and system which will contribute towards the continued operation of the nuclear power plants beyond their economic life which is fixed mainly to amortize the investments over a period. Tarapur Atomic Power Station (TAPS) which consists of 2 nos. of Boiling Water Reactor (BWRs) with the presently rated capacity of 160 MWe each has been operating for the past 24 years and is completing its 25th year of service by the year 1994 which was considered as its economic life and the plant depreciation as well as fuel supply agreement were based on this period of 25 years. I will be discussing about the available residual life which is much more than the above (25 years) and the studies we have undertaken in respect of the assessment of this residual life. (author). 2 tabs., 6 figs

  6. Invitation to the IAEA FR17 Conference on Fast Reactors, Russia, 2017

    International Nuclear Information System (INIS)

    Rachkov, V.

    2013-01-01

    Russian Federation has a large successful experience of peaceful application of nuclear power and it is a leader in development of fast reactors that permit to create sustainable and safe base for future world nuclear power. Russian Government approved in 2010 a Federal Target Program “Nuclear power technologies of a new generation for period of 2010-2015 and with outlook to 2020” aimed at creation of new technological platform for future nuclear power based on fast reactors. State Atomic Energy Corporation “Rosatom” supports activities of the IAEA on strengthening international collaboration on fast reactors and invites IAEA to host next FR17 Conference on FRs in Russia in 2017

  7. Power conditioning system for a nuclear reactor

    International Nuclear Information System (INIS)

    Higashigawa, Yuichi; Joge, Toshio.

    1981-01-01

    Purpose: To provide a power conditioning system for a BWR type reactor which has a function to be automatically operated within a range that the relationship between the heat power of the reactor and the electric power of an electric generator does not lose the safety of fuel by eliminating the unnecessary fluctuation of the power of the reactor. Constitution: A load request error signal fed from a conventional turbine control system to recirculation flow regulator is eliminated, and a reactor power conditioning system is newly provided, to which an electric generator power signal, a reactor average power area monitor signal and a load request signal are inputted. Thus, the load request signal is compared directly with the electric power of the electric generator, the recirculation flow rate is controlled by the compared result, and whether the correlation between the heat power of the reqctor and the electric power of the generator satisfies the correlation determined to prove the safety of fuel or not is checked. If this correlation is satisfied, the recirculation flow rate is merely automatically controlled. (Yoshino, Y.)

  8. Oklo natural reactors: geological and geochemical conditions

    International Nuclear Information System (INIS)

    Jakubick, A.T.; Church, W.

    1986-02-01

    Published as well as unpublished material on the Oklo natural reactors in Gabon was evaluated with regard to the long-term aspects of nuclear waste disposal. Even though the vast data base available at present can provide only a site specific description of the phenomenon, already this material gives relevant information on plutonium retention, metamictization, fission product release, hydrogeochemical stability and migration of fission products. Generalized conclusions applicable to other nuclear waste repository would require the quantitative reconstruction of t s coupled thermo-hydrologic-chemical processes. This could be achieved by studying the deviations in the 2 H/ 1 H and 18 O/ 16 O ratios of minerals at Oklo. A further generalization of the findings from Oklo could be realized by examining the newly-discovered reactor zone 10, which was active under very different thermal conditions than the other reactors. 205 refs

  9. Approximate computation of hydrothermal conditions of nuclear reactor spray ponds

    International Nuclear Information System (INIS)

    Yarkho, A.A.; Borshchev, V.A.

    1990-01-01

    An algorithm is presented for determining the evaporation numbers of nuclear reactor spray ponds which provide necessary reactor cooling during its normal operation under given meteorological conditions with account of restrictions on the cooled water temperature at the reactor entrance

  10. Power conditioning for space nuclear reactor systems

    Science.gov (United States)

    Berman, Baruch

    1987-01-01

    This paper addresses the power conditioning subsystem for both Stirling and Brayton conversion of space nuclear reactor systems. Included are the requirements summary, trade results related to subsystem implementation, subsystem description, voltage level versus weight, efficiency and operational integrity, components selection, and shielding considerations. The discussion is supported by pertinent circuit and block diagrams. Summary conclusions and recommendations derived from the above studies are included.

  11. Fast reactor safety and related physics. Volume I. Invited papers; panels; summary

    Energy Technology Data Exchange (ETDEWEB)

    1976-01-01

    Separate abstracts were prepared for each of the twenty invited papers included. The papers covered sessions on licensing aspects of safety design bases, safety of demonstration plants, safety aspects of large commercial fast breeders, and safety test facilities.

  12. Air conditioning device for reactor buildings

    International Nuclear Information System (INIS)

    Kikuchi, Shiro.

    1982-01-01

    Purpose: To decrease the opening areas of pipe lines for an air conditioning device at the portions passing through the shielding walls of a reactor building for a FBR type reactor, as well as reduce the size of the building. Constitution: Airs in the building for containing reactor are liquefied in an air liquefying mechanism. The liquefied airs are sent by way of pipe lines to each of evaporators, wherein each of the chambers are cooled because of latent heat of evaporation and evaporated airs are released to each of the chambers. The airs released to each of the chambers are collected into an exhaust chamber and sent by way of a duct to the air liquefying mechanism and liquefied again. Since the volume of the liquefied airs may be smaller than the amount conventionally required for usual cooled airs, the pipe lines passing through the shielding walls of the building can be of smaller diameter. This can decrease the opening areas of the pipe lines at the portions passing through the walls of the shieldings and, since the opening areas are smaller, the structure of the radiation shieldings can be simplified in these portions. Further, since the space of the pipe lines in the building is reduced extremely, the size of the building can be reduced. (Moriyama, K.)

  13. Studies in Phebus reactor of fuel behaviour upon LOCA conditions

    International Nuclear Information System (INIS)

    Manin, A.; Del Negro, R.; Reocreux, M.

    1980-09-01

    The fuel behaviour upon LOCA conditions is studied in an in-pile loop, in Phebus reactor. This paper presents: a short description of Phebus reactor; the current program (adjusting the thermohydraulic conditions in order to get cladding failure); the program developments (consequences involved by cladding failure); the fuel test conditions determination [fr

  14. Conditioning CANDU reactor wastes for disposal

    International Nuclear Information System (INIS)

    Beamer, N.V.; Bourns, W.T.; Buckley, L.P.; Speranzini, R.A.

    1981-12-01

    A Waste Treatment Centre (WTC) is being constructed at the Chalk River Nuclear Laboratories to develop and demonstrate processes for converting reactor wastes to a form suitable for disposal. The WTC contains a starved air incinerator for reducing the volume of combustible solid wastes, a reverse osmosis section for reducing the volume of liquid wastes and an immobilization section for incorporating the conditioned wastes in bitumen. The incinerator is commissioned on inactive waste: approximately 16.5 Mg of waste packaged in polyethylene bags has been incinerated in 17 burns. Average weight and volume reductions of 8.4:1 and 32:1, respectively, have been achieved. Construction of the reverse osmosis section of WTC is complete and inactive commissioning will begin in 1982 January. The reverse osmosis section was designed to process 30,000 m 3 /a of dilute radioactive waste. The incinerator ash and concentrated aqueous waste will be immobiblized in bitumen using a horizontal mixer and wiped-film evaporator. Results obtained during inactive commissioning of the incinerator are described along with recent results of laboratory programs directed at demonstrating the reverse osmosis and bituminization processes

  15. New reactor concepts for new generation of nuclear power plants: an overview, invited paper

    International Nuclear Information System (INIS)

    Vujic, J.; Greenspan, E.; Milosevic, M.

    2006-01-01

    The outlook for energy demand underscores the need to increase the share of nuclear energy production. Achieving the vision of sustainable growth of nuclear energy will require development of both advanced nuclear fuel cycles and next generation reactor technologies and advanced reprocessing and fuel treatment technologies. To achieve this vision, the US department of energy (DOE) has adopted new strategy, the Global Nuclear Energy Partnership (GNEP), which integrates earlier programs: the Generation IV Nuclear Energy Systems Initiative (Generation IV), Nuclear Hydrogen Initiative (NHI), and the Advanced Fuel Cycle Initiative (AFCI) with proliferation-resistant spent fuel reprocessing to minimize nuclear waste. Generation IV furthers this vision beyond previous energy systems, such as Generation III+, through incremental improvements in economic competitiveness, sustainability, development of passively safe systems, and breakthrough methods to reduce the routes of nuclear proliferation. This paper summarizes the main characteristics of the six most promising nuclear energy systems identified by the Generation IV Roadmap and reviews some Generation IV system designs for small-side proliferation resistant reactors being developed by University of California at Berkeley. (author)

  16. Effect of reactor conditions on MSIV-ATWS power level

    International Nuclear Information System (INIS)

    Diamond, D.J.

    1987-01-01

    In a boiling water reactor (BWR) when there is closure of the main steam isolation valves (MSIVs), the energy generated in the core will be transferred to the pressure suppression pool (PSP) via steam that flows out of the relief valves. The pool has limited capacity as a heat sink and hence, if there is no reactor trip [an anticipated transient without scram (ATWS) event], there is the possibility that the pool temperature may rise beyond acceptable limits. The present study was undertaken to determine how the initial reactor conditions affect the power level during an MSIV-ATWS event. The time of interest is the 20- to 30-min period when it is assumed that the reactor is in a quasi equilibrium condition with the water level and pressure fixed, natural circulation conditions and no control rod movement or significant boron in the core. The initial conditions of interest are the time of the cycle and the operating state

  17. Evaluation of neutron exposure conditions for the Buffalo Reactor

    International Nuclear Information System (INIS)

    Lippincott, E.P.; Kellogg, L.S.; McElroy, W.N.; Baldwin, C.A.

    1984-04-01

    The light water test reactor at the Nuclear Science and Technology Facility of the State University of New York at Buffalo is currently being used to irradiate specimens in in-core positions for NRC-sponsored metallurgical tests. It is important that the neutron exposures for these Buffalo tests be consistent with those determined for related irradiations in the BSR and ORR reactor at ORNL. Therefore, HEDL National Reactor Dosimetry Center dosimetry procedures and ORNL calculational procedures were used for an evaluation of typical test conditions

  18. Cable condition monitoring in a pressurized water reactor environment

    International Nuclear Information System (INIS)

    Al-Hussaini, T.J.

    1988-01-01

    Oconee Nuclear Station is the first nuclear plant designed, engineered and constructed by Duke Power Company. Even though the accelerated aging method was available to determine the life expectancy of the cable used in the reactor building, no natural aging data was available at that time. In order to be able to verify the condition of the reactor building cable over the life of the plant, an on-going cable monitoring plan was instituted. Various types of cable were selected to be monitored, and they were installed in cable life evaluation circuits in the reactor building. At five year intervals over the life of the plant, cable samples would be removed from these cable life evaluation circuits and tested to determine the effects of the reactor building environment on the integrity of the cable. A review of the cable life evaluation circuits and the results of the evaluation program to date is presented

  19. Historical civilian nuclear accident based Nuclear Reactor Condition Analyzer

    Science.gov (United States)

    McCoy, Kaylyn Marie

    There are significant challenges to successfully monitoring multiple processes within a nuclear reactor facility. The evidence for this observation can be seen in the historical civilian nuclear incidents that have occurred with similar initiating conditions and sequences of events. Because there is a current lack within the nuclear industry, with regards to the monitoring of internal sensors across multiple processes for patterns of failure, this study has developed a program that is directed at accomplishing that charge through an innovation that monitors these systems simultaneously. The inclusion of digital sensor technology within the nuclear industry has appreciably increased computer systems' capabilities to manipulate sensor signals, thus making the satisfaction of these monitoring challenges possible. One such manipulation to signal data has been explored in this study. The Nuclear Reactor Condition Analyzer (NRCA) program that has been developed for this research, with the assistance of the Nuclear Regulatory Commission's Graduate Fellowship, utilizes one-norm distance and kernel weighting equations to normalize all nuclear reactor parameters under the program's analysis. This normalization allows the program to set more consistent parameter value thresholds for a more simplified approach to analyzing the condition of the nuclear reactor under its scrutiny. The product of this research provides a means for the nuclear industry to implement a safety and monitoring program that can oversee the system parameters of a nuclear power reactor facility, like that of a nuclear power plant.

  20. Study of startup conditions of a pulsed annular reactor

    International Nuclear Information System (INIS)

    Silva, Mario Augusto Bezerra da

    2003-10-01

    A new concept of reactor, which combines features of pulsed and stationary reactors, was proposed so as to produce intense neutronic fluxes. Such a reactor, known as VICHFPR (Very Intense Continuous High Flux Pulsed Reactor), consists of a subcritical core with an annular geometry and pulsed by a rotating reflector which acts as a reactivity modulator as it produces a short pulse (approximately equal to 1 ms) of high intensity, guiding the region near the pulser to super-prompt critical state. This dissertation intends to analyze the startup conditions of a Pulsed Annular Reactor. The evolution of the neutron pulse intensity is analyzed when the reactivity modulator is brought upwards according to a helicoidal path from its initial position (far away from the core), when the multiplication factor has a subcritical value, up to the final position (near the core), in which a super-prompt critical state is reached. Part of the analysis is based on the variation of neutron reflection, which is a uniform function of the exit and reflection angles between the core and the modulator. It must be emphasized that this work is an approximation of the real situation. As the initial and final reactor parameters are known, a programming code in Fortran is worked out to provide the multiplication factor and the flux intensity evolution. According to the results obtained with this code, the conditions under which the modulator must be lifted up during the startup are established. Basically, these conditions are related to the analysis of the rising and the rotation velocities, the reflector saving and the initial distance between the reactor and the modulator. The Pulsed Annular Reactor startup was divided into three stages. Because of its negative reactivity in the first two stages, the neutron multiplication is not large, while the last one, having a positive reactivity, shows an intense multiplication as is usually expected when handling pulsed systems. This last stage is quite

  1. A study of passive safety conditions for fast reactor core

    International Nuclear Information System (INIS)

    Shimizu, Akinao

    1991-01-01

    A study has been made for passive safety conditions of fast reactor cores. Objective of the study is to develop a concept of a core with passive safety as well as a simple safety philosophy. A simple safety philosophy, which is wore easy to explain to the public, is needed to enhance the public acceptance for nuclear reactors. The present paper describes a conceptual plan of the study including the definition of the problem a method of approach and identification of tasks to be solved

  2. Helminth Egg Removal Capacity of UASB Reactors under Subtropical Conditions

    Directory of Open Access Journals (Sweden)

    Rosa-Elena Yaya-Beas

    2015-05-01

    Full Text Available This research was conducted to study the anaerobic sludge filtration capacity regarding helminth egg removal in upflow anaerobic sludge blanket (UASB reactors. Two 25 L lab-scale UASB reactors were operated at an ambient temperature which varied between 17.1 and 28.6 °C. Ascaris suum egg was selected as the model egg considering its similarity in terms of size and morphology to Ascaris lumbricoides, a human pathogen. Ascaris suum eggs were obtained from female parasites of infected pigs. The anaerobic sludge filtration capacity was performed applying upflow velocities between 0.09 and 0.68 m·h−1. Three sludge bed heights in the range of 0.30–0.40 m, 0.50–0.60 m and 0.60–0.70 m were applied. These sludge bed heights corresponded to 19%–25%, 31%–38% and 38%–44% of the total reactor height, respectively. Under the mentioned conditions, the average helminth egg removal efficiency was reciprocally correlated to the imposed upflow velocity. The studied lab-scale reactors reported an average helminth egg removal between 34%–100%, 30%–91% and 34%–56%, when the sludge bed in the UASB reactor was 19%–25%, 31%–38% and 38%–44% of the total reactor height, respectively. The decreased filtration capacity at increasing sludge bed heights might be likely related to biogas production and channeling formation. The average helminth egg removal efficiency in the control experiments performed without any sludge bed, by plain sedimentation, varied between 44% and 66%.

  3. Behaviour of molten reactor fuels under accident conditions

    International Nuclear Information System (INIS)

    Xavier Swamikannu, A.; Mathews, C.K.

    1980-01-01

    The behaviour of molten reactor fuels under accident conditions has received considerable importance in recent times. The chemical processes that occur in the molten state among the fuel, the clad components and the concrete of the containment building under the conditions of a core melt down accident in oxide fuelled reactors have been reviewed with the purpose of identifying areas of developmental work required to be performed to assess and minimize the consequences of such an accident. This includes the computation and estimation of vapour pressure of various gaseous species over the fuel, the clad and the coolant, providing of sacrificial materials in the concrete in order to protect the containment building in order to prevent release of radioactive gases into the atmosphere and understanding the distribution and chemical state of fission products in the molten fuel in order to provide for the effective removal of their decay heats. (auth.)

  4. Electrochemical noise measurements under pressurized water reactor conditions

    International Nuclear Information System (INIS)

    Van Nieuwenhove, R.

    2000-01-01

    Electrochemical potential noise measurements on sensitized stainless steel pressure tubes under pressurized water reactor (PWR) conditions were performed for the first time. Very short potential spikes, believed to be associated to crack initiation events, were detected when stressing the sample above the yield strength and increased in magnitude until the sample broke. Sudden increases of plastic deformation, as induced by an increased tube pressure, resulted in slower, high-amplitude potential transients, often accompanied by a reduction in noise level

  5. A model to predict moisture conditions in concrete reactor containments

    International Nuclear Information System (INIS)

    Ahs, M.; Nilsson, L.O.; Poyet, S.; L'Hostis, V.

    2015-01-01

    Moisture has an impact in many of the degradation mechanisms that appear in the structures of a nuclear power plant. Moisture conditions in a reactor containment wall have been simulated by using a hygro-thermal model of drying concrete. Methods to estimate the temperature dependency of the sorption isotherms and moisture transport properties is suggested and applied in the model. This temperature dependency is included as there is a temperature gradient present through the containment wall. The hygro-thermal model was applied on a full scale 3D model of a real reactor containment building and the concrete relative humidity has been computed at 4 different moments: 1, 10, 20 and 30 years. The results show that the major part of the concrete is not dried at all even after 30 years of operation. It is also clear that the temperature distribution inside the whole concrete volume is affected by the variable boundary conditions. It was concluded that the suggested hygro-thermal model was appropriate to use as a method to estimate the existing conditions in a PWR reactor containment wall

  6. Mixing Characteristics during Fuel Coolant Interaction under Reactor Submerged Conditions

    International Nuclear Information System (INIS)

    Hong, S. W.; Na, Y. S.; Hong, S. H.; Song, J. H.

    2014-01-01

    A molten material is injected into an interaction chamber by free gravitation fall. This type of fuel coolant interaction could happen to operating plants. However, the flooding of a reactor cavity is considered as SAM measures for new PWRs such as APR-1400 and AP1000 to assure the IVR of a core melt. In this case, a molten corium in a reactor is directly injected into water surrounding the reactor vessel without a free fall. KAERI has carried out fuel coolant interaction tests without a free fall using ZrO 2 and corium to simulate the reactor submerged conditions. There are four phases in a steam explosion. The first phase is a premixing phase. The premixing is described in the literature as follows: during penetration of melt into water, hydrodynamic instabilities, generated by the velocities and density differences as well as vapor production, induce fragmentation of the melt into particles; the particles fragment in turn into smaller particles until they reach a critical size such that the cohesive forces (surface tension) balance exactly the disruptive forces (inertial); and the molten core material temperature (>2500 K) is such that the mixing always occurs in the film boiling regime of the water: It is very important to qualify and quantify this phase because it gives the initial conditions for a steam explosion This paper mainly focuses on the observation of the premixing phase between a case with 1 m free fall and a case without a free fall to simulate submerged reactor condition. The premixing behavior between a 1m free fall case and reactor case submerged without a free fall is observed experimentally. The average velocity of the melt front passing through 1m water pool; - Case without a free fall: The average velocity of corium, 2.7m/s, is faster than ZrO 2 , 2.3m/s, in water. - Cases of with a 1 m free fall and without a free fall : The case without a free fall is about two times faster than a case with a 1 m free fall. Bubble characteristics; - Case

  7. Behaviour of gas cooled reactor fuel under accident conditions

    International Nuclear Information System (INIS)

    1991-11-01

    The Specialists Meeting on Behaviour of Gas Cooled Reactor Fuel under Accident Conditions was convened by the International Atomic Energy Agency on the recommendation of the International Working Group on Gas Cooled Reactors. The purpose of the meeting was to provide an international forum for the review of the development status and for the discussion on the behaviour of gas cooled reactor fuel under accident conditions and to identify areas in which additional research and development are still needed and where international co-operation would be beneficial for all involved parties. The meeting was attended by 45 participants from France, Germany, Japan, Switzerland, the Union of Soviet Socialists Republics, the United Kingdom, the United States of America, CEC and the IAEA. The meeting was subdivided into five technical sessions: Summary of Current Research and Development Programmes for Fuel; Fuel Manufacture and Quality Control; Safety Requirements; Modelling of Fission Product Release - Part I and Part II; Irradiation Testing/Operational Experience with Fuel Elements; Behaviour at Depressurization, Core Heat-up, Power Transients; Water/Steam Ingress - Part I and Part II. 22 papers were presented. A separate abstract was prepared for each of these papers. At the end of the meeting a round table discussion was held on Directions for Future R and D Work and International Co-operation. Refs, figs and tabs

  8. Wireless condition monitoring for the RA-6 research reactor

    International Nuclear Information System (INIS)

    Garcia Peyrano, O.; Calzeta, O.; Rico, N.; Damiani, H.; Coutsiers, E.

    1999-01-01

    The vibration laboratory at C.A.B. has a great experience with the analysis and diagnostic of symptoms of failures in the rotating equipment of the R-6 research reactor and in our longest NPP (CANDU 600 Mw), located in Embalse town, Cordoba City, Argentina. Objective: The standard condition monitoring instrumentation system were designed for large equipment operating under different environmental conditions and sensitivities. The signal processing is not flexible and the diagnostic is an expensive method for the small poll type research reactors. This papers describes the research and development which are related whit the new concept, cheaper and flexible condition monitoring instrumentation system. Implementing a vibration analysis measurements technique with a sensor inside (in the pool) of the nuclear reactor RA-6, and mainly based on fft signal processing, an extensive program for vibration source identification was done. Different nuclear power conditions were monitored as full power and in zero power, also. This zero power shows the best acoustical environmental, because the cooling pumps are stop, and the core is cooling by natural convection. Two sensors were mainly used as the detector's subsystem. One of these detectors was an accelerometer attached to the top of the fine control rod and the other one was a water resistant omnidirectional microphone which was located underwater at different distances from the nuclear core. All the signal measurement by this two sensors were recorded and then was processed. Both signal was acquired at the same time for correlation analysis purposes. The analysis was composed by a 'Spectral Dynamics SD380' connected to a P.C. with dedicated post processing software. On the other hand, some calibration and sensitivity comparison was done using an SKFCM40, dual channel data collector and analyzer. (author)

  9. Historical Review of Uncommanded Lateral-Directional Motions At Transonic Conditions (Invited)

    Science.gov (United States)

    Chambers, Joseph R.; Hall, Robert M.

    2003-01-01

    This paper presents the results of a survey of past experiences with uncommanded lateral-directional motions at transonic speeds during specific military aircraft programs. The effort was undertaken to provide qualitative and quantitative information on past airplane programs that might be of use to the participants in the joint NASA/Navy/Air Force Abrupt Wing Stall (AWS) Program. The AWS Program was initiated because of the experiences of the F/A-18E/F development program, during which unexpected, severe wing-drop motions were encountered by preproduction aircraft at transonic conditions. These motions were judged to be significantly degrading to the primary mission requirements of the aircraft. Although the problem was subsequently solved for the production version of the F/A-l8E/F, a high-level review panel emphasized the poor understanding of such phenomena and issued a strong recommendation to: Initiate a national research effort to thoroughly and systematically study the wing drop phenomena. A comprehensive, cooperative NASA/Navy/Air Force AWS Program was designed to respond to provide the required technology requirements. As part of the AWS Program, a work element was directed at a historical review of wing-drop experiences in past aircraft development programs at high subsonic and transonic speeds. In particular, information was requested regarding: specific aircraft configurations that exhibited uncommanded motions and the nature of the motions; geometric characteristics of the air- planes; flight conditions involved in occurrences; relevant data, including wind-tunnel, computational, and flight sources; figures of merit used for analyses; and approaches used to alleviate the problem. An attempt was also made to summarize some of the more important lessons learned from past experiences, and to recommend specific research efforts. In addition to providing technical information to assist the AWS research objectives, the study produced fundamental information

  10. Invited commentary: co-occurring health conditions among women living with profound life challenges.

    Science.gov (United States)

    Coughlin, Steven S

    2011-09-01

    People challenged by homelessness are living with several losses including the loss of a home, employment, economic security, health or well-being, and personal security. Assistance programs for people who are homeless consist of housing, emergency shelter, food services, employment assistance, peer support, medical care, and mental health services. An article by Riley et al. (Am J Epidemiol. 2011;175(5):515-522) appearing in this issue of the Journal examines the relation between basic subsistence needs and health outcomes in a cohort of 129 human immunodeficiency virus-infected women who were recruited from a probability sample of low-income hotels, homeless shelters, and free food programs in San Francisco, California. The results of their study underscore the importance of addressing subsistence needs and providing access to medical and psychological treatment for homeless and unstably housed women. In addition to subsistence needs, more attention should be given to comorbid psychiatric and medical conditions that occur among homeless women, including trauma-related disorders.

  11. Reactor coolant pump shaft seal behavior during blackout conditions

    International Nuclear Information System (INIS)

    Mings, W.J.

    1985-01-01

    The United States Nuclear Regulatory Commission has classified the problem of reactor coolant pump seal failures as an unresolved safety issue. This decision was made in large part due to experimental results obtained from a research program developed to study shaft seal performance during station blackout and reported in this paper. Testing and analysis indicated a potential for pump seal failure under postulated blackout conditions leading to a loss of primary coolant with a concomitant danger of core uncovery. The work to date has not answered all the concerns regarding shaft seal failure but it has helped scope the problem and focus future research needed to completely resolve this issue

  12. 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.

  13. Computation of reactor control rod drop time under accident conditions

    International Nuclear Information System (INIS)

    Dou Yikang; Yao Weida; Yang Renan; Jiang Nanyan

    1998-01-01

    The computational method of reactor control rod drop time under accident conditions lies mainly in establishing forced vibration equations for the components under action of outside forces on control rod driven line and motion equation for the control rod moving in vertical direction. The above two kinds of equations are connected by considering the impact effects between control rod and its outside components. Finite difference method is adopted to make discretization of the vibration equations and Wilson-θ method is applied to deal with the time history problem. The non-linearity caused by impact is iteratively treated with modified Newton method. Some experimental results are used to validate the validity and reliability of the computational method. Theoretical and experimental testing problems show that the computer program based on the computational method is applicable and reliable. The program can act as an effective tool of design by analysis and safety analysis for the relevant components

  14. Impact of VOC Composition and Reactor Conditions on the Aging of Biomass Cookstove Emissions in an Oxidation Flow Reactor

    Science.gov (United States)

    Oxidation flow reactor (OFR) experiments in our lab have explored secondary organic aerosol (SOA) production during photochemical aging of emissions from cookstoves used by billions in developing countries. Previous experiments, conducted with red oak fuel under conditions of hig...

  15. System for stress corrosion conditions tests on PWR reactors

    International Nuclear Information System (INIS)

    Castro, Andre Cesar de Jesus

    2007-01-01

    The study of environmentally assisted cracking (EAC) involves the consideration and evaluation of the inherent compatibility between a material and the environment under conditions of either applied or residual stress. EAC is a critical problem because equipment, components and structure are subject to the influence of mechanical stress, water environment of different composition, temperature and different material history. Testing for resistance to EAC is one of the most effective ways to determine the interrelationships among this variables on the process of EAC. Up to now, several experimental techniques have been developed worldwide, which address different aspects of environmental caused damage. Constant loading of CT specimens test is a typical example of test, which is used for the estimation of parameters of stress corrosion cracking. To assess the initiation stages and kinetics of crack growth, the testing facility should allow active loading of specimens in the environment that is close to the actual operation conditions of assessed component. This paper presents a testing facility for stress corrosion cracking to be installed at CDTN, which was designed and developed at CDTN. The facility is used to carry out constant load tests under simulated PWR environment, where temperature, water pressure and chemistry are controlled, which are considered the most important factors in SCC. Also, the equipment operational conditions, its applications, and restrictions are presented. The system was developed to operate at temperature until 380 degree C and pressure until 180 bar. It consists in a autoclave stuck at a mechanical system, responsible of producing load , a water treatment station, and a data acquisition system. This testing facility allows the evaluation of cracking progress, especially at PWR reactor. (author) operational conditions. (author)

  16. Serviceability of rod ceramic fuel pins on motoring conditions of FTP or NEMF reactor

    International Nuclear Information System (INIS)

    Deryavko, I.I.

    2004-01-01

    The operation conditions of rod ceramic fuel pins in the running hydrogen-cooled technological canals of FTP or NEMF reactor on the motoring conditions are considered. The available postreactor researches of the fuel pins are presented and the additional postreactor researches of fuel pins, tested on this mode in IVG.1 and IRGIT reactors, are carried out. The fuel pins serviceability on motoring conditions of FTP or NEF reactor operation is concluded. (author)

  17. Bid invitations for nuclear power stations

    International Nuclear Information System (INIS)

    Zijl, N.A. van.

    1975-01-01

    Types of bid invitations, basic requirements on bid invitation documents, basic content of a turnkey bid invitation (bid invitation letter, instructions to the bidders, terms and conditions of the draft contract, technical specifications, site data and information), nuclear fuel procurement, differences turnkey - non-turnkey, legal, commerical, and technical matters concerning the contract document. (HP) [de

  18. Enhancement of Tokamak Fusion Test Reactor performance by lithium conditioning

    International Nuclear Information System (INIS)

    Mansfield, D.K.; Hill, K.W.; Strachan, J.D.; Bell, M.G.; Scott, S.D.; Budny, R.; Marmar, E.S.; Snipes, J.A.; Terry, J.L.; Batha, S.; Bell, R.E.; Bitter, M.; Bush, C.E.; Chang, Z.; Darrow, D.S.; Ernst, D.; Fredrickson, E.; Grek, B.; Herrmann, H.W.; Janos, A.; Jassby, D.L.; Jobes, F.C.; Johnson, D.W.; Johnson, L.C.; Levinton, F.M.; Mikkelsen, D.R.; Mueller, D.; Owens, D.K.; Park, H.; Ramsey, A.T.; Roquemore, A.L.; Skinner, C.H.; Stevenson, T.; Stratton, B.C.; Synakowski, E.; Taylor, G.; von Halle, A.; von Goeler, S.; Wong, K.L.; Zweben, S.J.

    1996-01-01

    Wall conditioning in the Tokamak Fusion Test Reactor (TFTR) [K. M. McGuire et al., Phys. Plasmas 2, 2176 (1995)] by injection of lithium pellets into the plasma has resulted in large improvements in deuterium endash tritium fusion power production (up to 10.7 MW), the Lawson triple product (up to 10 21 m -3 s keV), and energy confinement time (up to 330 ms). The maximum plasma current for access to high-performance supershots has been increased from 1.9 to 2.7 MA, leading to stable operation at plasma stored energy values greater than 5 MJ. The amount of lithium on the limiter and the effectiveness of its action are maximized through (1) distributing the Li over the limiter surface by injection of four Li pellets into Ohmic plasmas of increasing major and minor radius, and (2) injection of four Li pellets into the Ohmic phase of supershot discharges before neutral-beam heating is begun. copyright 1996 American Institute of Physics

  19. Operation characteristics and conditions of training reactor VR-1

    International Nuclear Information System (INIS)

    Matejka, K.; Kolros, A.; Polach, S.; Sklenka, L.

    1994-01-01

    The first 3 years of operation of the VR-1 training reactor are reviewed. This period includes its physical start-up (preparation, implementation, results) and operation development as far as the current operating configuration of the reactor core. The physical start-up was commenced using a reactor core referred to as AZ A1, whose physical parameters had been verified by calculation and whose configuration was based on data tested experimentally on the SR-0 reactor at Vochov. The next operating core, labelled AZ A2, was already prepared during the test operation of the VR-1 reactor. Its configuration was such that both of the main horizontal channels, radial and tangential, could be employed. The configuration that followed, AZ A3, was an intermediate step before testing the graphite side reflector. The current reactor core, labelled AZ A3 G, was obtained by supplementing the previous core with a one-sided graphite side reflector. (Z.S.). 2 tabs., 11 figs., 2 refs

  20. A simulation Model of the Reactor Hall Ventilation and air Conditioning Systems of ETRR-2

    International Nuclear Information System (INIS)

    Abd El-Rahman, M.F.

    2004-01-01

    Although the conceptual design for any system differs from one designer to another. each of them aims to achieve the function of the system required. the ventilation and air conditioning system of reactors hall is one of those systems that really differs but always dose its function for which it is designed. thus, ventilation and air conditioning in some reactor hall constitute only one system whereas in some other ones, they are separate systems. the Egypt Research Reactor-2 (ETRR-2)represents the second type. most studies conducted on ventilation and air conditioning simulation models either in traditional building or for research rectors show that those models were not designed similarly to the model of the hall of ETRR-2 in which ventilation and air conditioning constitute two separate systems.besides, those studies experimented on ventilation and air conditioning simulation models of reactor building predict the temperature and humidity inside these buildings at certain outside condition and it is difficult to predict when the outside conditions are changed . also those studies do not discuss the influences of reactor power changes. therefore, the present work deals with a computational study backed by infield experimental measurements of the performance of the ventilation and air conditioning systems of reactor hall during normal operation at different outside conditions as well as at different levels of reactor power

  1. Chemistry of fission product iodine under nuclear reactor accident conditions

    International Nuclear Information System (INIS)

    Malinauskas, A.P.; Bell, J.T.

    1986-01-01

    The radioisotopes of iodine are generally acknowledged to be the species whose release into the biosphere as a result of a nuclear reactor accident is of the greatest concern. In the course of its release, the fission product is subjected to differing chemical environments; these can alter the physicochemical form of the fission product and thus modify the manner and extent to which release occurs. Both the chemical environments which are characteristic of reactor accidents and their effect in determining physical and chemical form of fission product iodine have been studied extensively, and are reviewed in this report. 76 refs

  2. Fluidized-bed reactors processes and operating conditions

    CERN Document Server

    Yates, John G

    2016-01-01

    The fluidized-bed reactor is the centerpiece of industrial fluidization processes. This book focuses on the design and operation of fluidized beds in many different industrial processes, emphasizing the rationale for choosing fluidized beds for each particular process. The book starts with a brief history of fluidization from its inception in the 1940’s. The authors present both the fluid dynamics of gas-solid fluidized beds and the extensive experimental studies of operating systems and they set them in the context of operating processes that use fluid-bed reactors. Chemical engineering students and postdocs as well as practicing engineers will find great interest in this book.

  3. Pattern recognition application for surveillance of abnormal conditions in a nuclear reactor

    International Nuclear Information System (INIS)

    Pepelyshev, Yu.N.; Dzwinel, W.

    1990-01-01

    The system to monitor abnormal conditions in a nuclear reactor, based on the noise analysis of the reactor basic parameters such as power, temperature and coolant flow rate, has been developed. The pattern recognition techniques such as clustering, cluster analysis, feature selection and clusters visualization methods form the basis of the software. Apart from non-hierarchical clustering procedures applied earlier, the hierarchical one is recommended. The system application for IBR-2 Dubna reactor diagnostics is shown. 10 refs.; 6 figs

  4. A review of healthy condition for reactor operators of NPP

    International Nuclear Information System (INIS)

    Zhang Chi; Yang Di; Zhu Lixin; Zhou Limin

    2008-01-01

    Physical health is one of fundamental elements and the necessary qualification for holding operator license or senior operator license of a nuclear power plant issued by NNSA. Based upon requirements for healthy qualification of reactor operators in nuclear safety regulations and some new issued regulations, especially some involved national standard (GB), review direction, relevant contents and criteria were described, and some suggestions were provided in this paper. (authors)

  5. Effect of reactor finiteness on the boundary condition at the surface of a booster section

    International Nuclear Information System (INIS)

    Wassef, W.A.

    1982-01-01

    Effect of reactor finiteness on the boundary condition at the surface of an absorbing booster embedded in the reactor core is studied and formulated. The model used in these calculations depends on the Pl-Transport coupling technique. This method takes into consideration the rigorous neutron transport behavior inside the booster medium, while the Pl-approximation in the bulk of the scattering medium surrounding the booster which can be considered infinite in most practical applications. The neutron flux gradient parallel to the surface of the booster is considered. The geometrical configuration of the reactor core cross section is circular or rectangular. Finiteness of the reactor is introduced in the general formulation through its dimensions or buckling. Extensive numerical results are given to demonstrate the dependence of the boundary condition at the surface of the booster section on the reactor finiteness and the different physical parameters

  6. Safety analysis of switching between reductive and oxidative conditions in a reaction coupling reverse flow reactor.

    NARCIS (Netherlands)

    van Sint Annaland, M.; Kuipers, J.A.M.; van Swaaij, Willibrordus Petrus Maria

    2001-01-01

    A new reverse flow reactor is developed where endothermic reactants (propane dehydrogenation) and exothermic reactants (fuel combustion) are fed sequentially to a monolithic catalyst, while periodically alternating the inlet and outlet positions. Upon switching from reductive to oxidative conditions

  7. Thermohydraulic characteristics under some transient conditions of the Dalat Nuclear Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Huy, Ngo Quang; Khang, Ngo Phu; An, Tran Khac; Nghiem, Huynh Ton [Nuclear Research Inst., Da Lat (Viet Nam)

    1994-10-01

    Some experimental and theoretical thermal hydraulic characteristics of the Dalat Nuclear Research Reactor are presented, together with some general assessments, from a thermal hydraulic point of view, of its safety under transient conditions. (author). 3 refs., 9 figs., 7 tabs.

  8. Limits and conditions for permanent operation of the LVR-15 reactor

    International Nuclear Information System (INIS)

    1995-03-01

    The ''Limits and Conditions'' have been set up to demonstrate compliance with all nuclear safety requirements specified by the Decree of the Czechoslovak Atomic Energy Commission No. 9/1985, and to serve as a guide for the reactor operators to respond adequately to deviations from the required values in normal operation conditions. The publications is divided into the following chapters: (1) Safety limits of the LVR-15 reactor; (2) Adjusting the reactor protection and control system parameters; (3) Limiting conditions for normal operation modes; (4) Control requirements; and (5) reactor operation management. A document by the Institute for Research, Production, and Application of Radioisotopes specifying conditions for the acceptance of radioactive wastes to the central repository is appended. (P.A.). 7 tabs

  9. Characteristics of Flameless Combustion in 3D Highly Porous Reactors under Diesel Injection Conditions

    Directory of Open Access Journals (Sweden)

    M. Weclas

    2013-01-01

    Full Text Available The heat release process in a free volume combustion chamber and in porous reactors has been analyzed under Diesel engine-like conditions. The process has been investigated in a wide range of initial pressures and temperatures simulating engine conditions at the moment when fuel injection starts. The resulting pressure history in both porous reactors and in free volumes significantly depends on the initial pressure and temperature. At lower initial temperatures, the process in porous reactors is accelerated. Combustion in a porous reactor is characterized by heat accumulation in the solid phase of the porous structure and results in reduced pressure peaks and lowered combustion temperature. This depends on reactor heat capacity, pore density, specific surface area, pore structure, and heat transport properties. Characteristic modes of a heat release process in a two-dimensional field of initial pressure and temperature have been selected. There are three characteristic regions represented by a single- and multistep oxidation process (with two or three slopes in the reaction curve and characteristic delay time distribution has been selected in five characteristic ranges. There is a clear qualitative similarity of characteristic modes of the heat release process in a free volume and in porous reactors. A quantitative influence of porous reactor features (heat capacity, pore density, pore structure, specific surface area, and fuel distribution in the reactor volume has been clearly indicated.

  10. Air leakage analysis of research reactor HANARO building in typhoon condition for the nuclear emergency preparedness

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Goany Up; Lee, Hae Cho; Kim, Bong Seok; Kim, Jong Soo; Choi, Pyung Kyu [Dept. of Emergency Preparedness, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-12-15

    To find out the leak characteristic of research reactor 'HANARO' building in a typhoon condition MELCOR code which normally is used to simulate severe accident behavior in a nuclear power plant was used to simulate the leak rate of air and fission products from reactor hall after the shutdown of the ventilation system of HANARO reactor building. For the simulation, HANARO building was designed by MELCOR code and typhoon condition passed through Daejeon in 2012 was applied. It was found that the leak rate is 0.1%·day{sup -1} of air, 0.004%·day{sup -1} of noble gas and 3.7×10{sup -5}%·day{sup -1} of aerosol during typhoon passing. The air leak rate of 0.1%·day can be converted into 1.36 m{sup 3}·hr{sup -1} , but the design leak rate in HANARO safety analysis report was considered as 600 m3·hr{sup -1} under the condition of 20 m·sec{sup -1} wind speed outside of the building by typhoon. Most of fission products during the maximum hypothesis accident at HANARO reactor will be contained in the reactor hall, so the direct radiation by remained fission products in the reactor hall will be the most important factor in designing emergency preparedness for HANARO reactor.

  11. In-core materials testing under LWR conditions in the Halden reactor

    International Nuclear Information System (INIS)

    Bennett, P.J.; Hauso, E.; Hoegberg, N.W.; Karlsen, T.M.; McGrath, M.A.

    2002-01-01

    The Halden boiling water reactor (HBWR) has been in operation since 1958. It is a test reactor with a maximum power of 18 MW and is cooled and moderated by boiling heavy water, with a normal operating temperature of 230 C and a pressure of 34 bar. In the past 15 years increasing emphasis has been placed on materials testing, both of in-core structural materials and fuel claddings. These tests require representative light water reactor (LWR) conditions, which are achieved by housing the test rigs in pressure flasks that are positioned in fuel channels in the reactor and connected to dedicated water loops, in which boiling water reactor (BWR) or pressurised water reactor (PWR) conditions are simulated. Understanding of the in-core behaviour of fuel or reactor materials can be greatly improved by on-line measurements during power operation. The Halden Project has performed in-pile measurements for a period of over 35 years, beginning with fuel temperature measurements using thermocouples and use of differential transformers for measurement of fuel pellet or cladding dimensional changes and internal rod pressure. Experience gained over this period has been applied to on-line instrumentation for use in materials tests. This paper gives details of the systems used at Halden for materials testing under LWR conditions. The techniques used to provide on-line data are described and illustrative results are presented. (authors)

  12. In-core materials testing under LWR conditions in the Halden reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bennett, P.J.; Hauso, E.; Hoegberg, N.W.; Karlsen, T.M.; McGrath, M.A. [OECD Halden Reactor Project (Norway)

    2002-07-01

    The Halden boiling water reactor (HBWR) has been in operation since 1958. It is a test reactor with a maximum power of 18 MW and is cooled and moderated by boiling heavy water, with a normal operating temperature of 230 C and a pressure of 34 bar. In the past 15 years increasing emphasis has been placed on materials testing, both of in-core structural materials and fuel claddings. These tests require representative light water reactor (LWR) conditions, which are achieved by housing the test rigs in pressure flasks that are positioned in fuel channels in the reactor and connected to dedicated water loops, in which boiling water reactor (BWR) or pressurised water reactor (PWR) conditions are simulated. Understanding of the in-core behaviour of fuel or reactor materials can be greatly improved by on-line measurements during power operation. The Halden Project has performed in-pile measurements for a period of over 35 years, beginning with fuel temperature measurements using thermocouples and use of differential transformers for measurement of fuel pellet or cladding dimensional changes and internal rod pressure. Experience gained over this period has been applied to on-line instrumentation for use in materials tests. This paper gives details of the systems used at Halden for materials testing under LWR conditions. The techniques used to provide on-line data are described and illustrative results are presented. (authors)

  13. Air leakage analysis of research reactor HANARO building in typhoon condition for the nuclear emergency preparedness

    International Nuclear Information System (INIS)

    Lee, Goany Up; Lee, Hae Cho; Kim, Bong Seok; Kim, Jong Soo; Choi, Pyung Kyu

    2016-01-01

    To find out the leak characteristic of research reactor 'HANARO' building in a typhoon condition MELCOR code which normally is used to simulate severe accident behavior in a nuclear power plant was used to simulate the leak rate of air and fission products from reactor hall after the shutdown of the ventilation system of HANARO reactor building. For the simulation, HANARO building was designed by MELCOR code and typhoon condition passed through Daejeon in 2012 was applied. It was found that the leak rate is 0.1%·day -1 of air, 0.004%·day -1 of noble gas and 3.7×10 -5 %·day -1 of aerosol during typhoon passing. The air leak rate of 0.1%·day can be converted into 1.36 m 3 ·hr -1 , but the design leak rate in HANARO safety analysis report was considered as 600 m3·hr -1 under the condition of 20 m·sec -1 wind speed outside of the building by typhoon. Most of fission products during the maximum hypothesis accident at HANARO reactor will be contained in the reactor hall, so the direct radiation by remained fission products in the reactor hall will be the most important factor in designing emergency preparedness for HANARO reactor

  14. Heat transport and afterheat removal for gas cooled reactors under accident conditions

    International Nuclear Information System (INIS)

    2001-01-01

    The Co-ordinated Research Project (CRP) on Heat Transport and Afterheat Removal for Gas Cooled Reactors Under Accident Conditions was organized within the framework of the International Working Group on Gas Cooled Reactors (IWGGCR). This International Working Group serves as a forum for exchange of information on national programmes, provides advice to the IAEA on international co-operative activities in advanced technologies of gas cooled reactors (GCRs) and supports the conduct of these activities. Advanced GCR designs currently being developed are predicted to achieve a high degree of safety through reliance on inherent safety features. Such design features should permit the technical demonstration of exceptional public protection with significantly reduced emergency planning requirements. For advanced GCRs, this predicted high degree of safety largely derives from the ability of the ceramic coated fuel particles to retain the fission products under normal and accident conditions, the safe neutron physics behaviour of the core, the chemical stability of the core and the ability of the design to dissipate decay heat by natural heat transport mechanisms without reaching excessive temperatures. Prior to licensing and commercial deployment of advanced GCRs, these features must first be demonstrated under experimental conditions representing realistic reactor conditions, and the methods used to predict the performance of the fuel and reactor must be validated against these experimental data. Within this CRP, the participants addressed the inherent mechanisms for removal of decay heat from GCRs under accident conditions. The objective of this CRP was to establish sufficient experimental data at realistic conditions and validated analytical tools to confirm the predicted safe thermal response of advance gas cooled reactors during accidents. The scope includes experimental and analytical investigations of heat transport by natural convection conduction and thermal

  15. Development of reactor water level sensor for extreme conditions

    Energy Technology Data Exchange (ETDEWEB)

    Miura, K; Ogasawara, T [Sukegawa Electric Co., Ltd., Hitachi, Ibaraki (Japan); Shibata, Akira; Nakamura, Jinichi; Saito, Takashi; Tsuchiya, Kunihiko [Japan Atomic Energy Agency, Oarai Research and Development Center, Oarai, Ibaraki (Japan)

    2012-03-15

    In the Fukushima accident, measurement failure of water level was one of the most important factors which caused serious situation. The differential pressure type water level indicators are widely used in various place of nuclear power plant but after the accident of TMI-2, the need of other reliable method has been required. The BICOTH type and the TRICOTH type water level indicator for light water power reactors had been developed for in-pile water level indicator but currently those are not adopted to nuclear power plant. In this study, the development of new type water level indicator composed of thermocouple and heater is described. Demonstration test and characteristic evaluation of the water level indicator were performed and we had obtained satisfactory results. (author)

  16. Pressurized water reactor iodine spiking behavior under power transient conditions

    International Nuclear Information System (INIS)

    Ho, J.C.

    1992-01-01

    The most accepted theory explaining the cause of pressurized water reactor iodine spiking is steam formation and condensation in damaged fuel rods. The phase transformation of the primary coolant from water to steam and back again is believed to cause the iodine spiking phenomenon. But due to the complex nature of the phenomenon, a comprehensive model of the behavior has not yet been successfully developed. This paper presents a new model based on an empirical approach, which gives a first-order estimation of the peak iodine spiking magnitude. Based on the proposed iodine spiking model, it is apparent that it is feasible to derive a correlation using the plant operating data base to monitor and control the peak iodine spiking magnitude

  17. Reactor safety under design basis flood condition for inland sites

    International Nuclear Information System (INIS)

    Hajela, S.; Bajaj, S.S.; Samota, A.; Verma, U.S.P.; Warudkar, A.S.

    2002-01-01

    Full text: In June 1994, there was an incident of flooding at Kakrapar Atomic Power Station (KAPS) due to combination of heavy rains and mechanical failure in the operation of gates at the adjoining weir. An indepth review of the incident was carried out and a number of flood protection measures were recommended and were implemented at site. As part of this review, a safety analysis was also done to demonstrate reactor safety with a series of failures considered in the flood protection features. For each inland NPP site, as part of design, different flood scenarios are analysed to arrive at design basis flood (DBF) level. This level is estimated based on worst combination of heavy local precipitation, flooding in river, failure of upstream/downstream water control structures

  18. Dosimetry characterization of the Godiva Reactor under burst conditions

    Energy Technology Data Exchange (ETDEWEB)

    Hickman, D. P. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Heinrichs, D. P. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Hudson, R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Wong, C. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Ward, D. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Wilson, C. [Atomic Weapons Establishment (AWE), Berkshire (United Kingdom); Clark, L. [Atomic Weapons Establishment (AWE), Berkshire (United Kingdom); Trompier, F. [Inst. for Radiation Protection and Nuclear Safety, Fontenay-aux-Roses (France)

    2017-06-22

    A series of sixteen (16) burst irradiations were performed in May 2014, fifteen of which were part of an international collaboration to characterize the Godiva IV fast burst reactor at the National Criticality Experiments Research Center (NCERC). Godiva IV is a bare cylindrical assembly of approximately 65 kg of highly enriched uranium fuel (93.2% 235U metal alloyed with 1.5% molybdenum for strength) and is designed to perform controlled prompt critical excursions (Myers 2010, Goda 2013). Twelve of the irradiations were dedicated to neutron spectral measurements using a Bonner multiple sphere spectrometer. Three irradiations, with core temperature increases of 71.1°C, 136.9°C, and 229.9°C, were performed for generating comparative fluence data, establishing corrections for varying heights, testing linearity with burst temperature, and establishing gamma dose characteristics.

  19. Heat transfer study of a submerged reactor channel under boil-off condition

    Energy Technology Data Exchange (ETDEWEB)

    Mukhopadhyay, Deb [Bhabha Atomic Research Centre, Mumbai (India). Reactor Safety Div.; Sahoo, P.K. [Indian Institute of Technology, Roorkee (India). Dept. of Mechanical and Industrial Engineering; Ghosh, A.K. [Bhabha Atomic Research Centre, Mumbai (India). Health, Safety and Environment Group

    2012-12-15

    Experiments have been carried out to study the heatup behavior of a single segmented reactor channel for Pressurized Heavy Water Reactor under submerged, partially submerged and exposed conditions. This situation may arise from a severe accident scenario of Pressurised Heavy Water Reactors where full or segmented reactor channels are likely to be disassembled and form a submerged debris bed. An assembly of electrical heater rod, simulating fuel bundle and channel components like Pressure Tube and Calandria Tube constitutes the segmented reactor channel. Heatup of this assembly is observed with respect to different water levels ranging from full submergence to totally exposed and power levels of 6-8 kW, typical to decay power level. It has been observed from the set of experiment that fuel bundle local dry out followed by heatup does not happen till the bundle is partially submerged. Temperature excursion of the bundle is evident when the bundle is exposed to steam-air environment. (orig.)

  20. Scram reliability under seismic conditions at the Experimental Breeder Reactor II

    International Nuclear Information System (INIS)

    Roglans, J.; Wang, C.Y.; Hill, D.J.

    1993-01-01

    A Probabilistic Risk Assessment of the Experimental Breeder Reactor II has recently been completed. Seismic events are among the external initiating events included in the assessment. As part of the seismic PRA a detailed study has been performed of the ability to shutdown the reactor under seismic conditions. A comprehensive finite element model of the EBR-II control rod drive system has been used to analyze the control rod system response when subjected to input seismic accelerators. The results indicate the control rod drive system has a high seismic capacity. The estimated seismic fragility for the overall reactor shutdown system is dominated by the primary tank failure

  1. Seismic verification of the Italian PEC fast reactor and effects of seismic conditions on the design

    International Nuclear Information System (INIS)

    Martelli, A.; Cecchini, F.; Masoni, P.; Maresca, G.; Castoldi, A.

    1988-01-01

    This paper deals with the aseismic design features of the Italian PEC fast reactor and the effects of seismic conditions on the reactor design. More precisely, after some notes on the main plant features, the paper reports on the design earthquakes adopted, the seismic monitoring procedures and the related actions, the design requirements, criteria and methods, and also provides a brief summary of the main research and development studies performed in support of design analysis. For the above-mentioned items, comparisons with the other fast reactors of the European Community countries are presented. Furthermore, the paper stresses the design modifications adopted to guarantee PEC seismic safety

  2. Uranium redistribution under oxidizing conditions in Oklo natural reactor zone 2, Gabon

    International Nuclear Information System (INIS)

    Isobe, H.; Ohnuki, T.; Murakami, T.; Gauthier-Lafaye, F.

    1995-01-01

    This mineralogical study was completed to elucidate the relationships between uranium distribution and alteration products of the host rock of natural reactor zone clays just below the reactor core. Uraninite is preserved without any alteration in the reactor core. Uranium minerals are found to be present in the fractures in the reactor zone clays associated with iron-mineral veins, galena and Ti-bearing minerals. Uranium, for which the phases could not be identified, occurs in iron-mineral veins and the iron-mineral rim of pyrite grains in the reactor zone clays. Uranium is not associated with granular iron minerals occurring in the illite matrix of the reactor zone clays. The degree of crystallinity and uranium content of the three iron-bearing alteration products suggest that they formed under different conditions; the granular iron minerals, under alteration conditions where uranium was not mobilized while the iron-mineral veins and the iron-mineral rim of pyrite, under conditions in which uranium is mobilized after the formation of the granular iron minerals

  3. Conditioning spent fuels from research nuclear reactor in ceramic dies

    International Nuclear Information System (INIS)

    Russo, D.O; Rodriguez, D.S; Mateos, P; Heredia, A; Sangilippo, M; Sterba, M

    2002-01-01

    The problem of immobilizing nuclear wastes is a complex one and is vitally important in the nuclear fuels cycle. In the case of spent elements from research reactors, the presence of large amounts of aluminum makes the procedure more complex and, therefore, onerous. There are various alternatives proposed for processing these materials. Two methods were studied in the Nuclear Materials Division for obtaining, as a final product, a vitreous block that could be place definitively in a geological repository. The processes are briefly, as follows: 1.By mechanical and chemical processes eliminating all the exterior aluminum from the fuel plates and then placing the product which we will call 'meat' (with some additional treatment and mixing with the amount needed to produce a natural uranium compound or weakened by decreasing the isotope enrichment in U-235) in a vitreous matrix. 2.Mechanically eliminate the aluminum from the exterior frame (as shown below) by shearing and cutting off the sectors containing only the Al, but leaving the rest of the aluminum, a big part of which is still present (4511.03), then doing the same procedure as in the case above: mixing with a natural uranium compound or weakening and vitrifying this mixture. In both cases, the vitrification can be carried out by fusion as well as by sintering. Given that these methods imply a big increase in volume together with a big mass of uranium and an even bigger amount of glass we decided to study an alternative. The proposed process involves synthesizing the mixtures obtained from the pre-treatment of the fuel plates (as described later) with natural isotope uranium oxide in order to obtain a block with the appropriate properties for its final disposal in a deep geological repository (CW)

  4. Biodegradation of Perchlorate in Laboratory Reactors Under Different Environmental Conditions

    Science.gov (United States)

    2010-07-01

    21 Figure 8. Initial and final mass of perchlorate, chloride, and chlorate ...is the soluble anion associated with the solid salts of ammonium, potassium , and sodium perchlorate. Large-scale production of ammonium perchlorate...ions. Most perchlorate-respiring microorganisms are capable of functioning under varying environmental conditions and use oxygen, nitrate, and chlorate

  5. Conditions of job performance quality during nuclear reactor maintenance

    International Nuclear Information System (INIS)

    Babos, K.; Takacs, J.

    1991-01-01

    The education and training programs of the maintenance personnel at the Paks Nuclear Power Plant, Hungary are described. In addition to well trained personnel, the conditions of job quality involves the preparation of job by proper check lists and availability of documentation, and the inspection and checking of the accomplishment of job tasks. Such an inspection procedure is proposed. (R.P.) 2 figs

  6. The dynamic pressure measurements of the nuclear reactor coolant for condition-based maintenance of the reactor

    International Nuclear Information System (INIS)

    Es-Saheb, M.H.H.

    1990-01-01

    The condition-based maintenance of the nuclear reactor, by monitoring and measuring the instantaneous dynamic pressure distribution of the coolant (water) impact on the solid surfaces of the reactor during operation is presented. The behaviour of water domes (jets) produced by underwater explosions of small changes of P.E.T.N. at various depths in two different size cylindrical containers, which simulate the nuclear reactor, is investigated. Water surface domes (jets) from the underwater explosions are photographed. Depending on the depth of the charge, curved and flat top jets of up to 455 mm diameter and impact speeds of up to 70 m/sec. are observed. The instabilities in the dome surfaces are observed and the instantaneous profiles are analysed. It is found that, in all cases tested, the maximum pressure takes place at the center of the jet and could reach up to 3.0 times the on-dimensional impact pressure value. The use of their measurements, as online monitoring for condition-based maintenance and design-out maintenance is discussed. 18 refs

  7. A New Application of Support Vector Machine Method: Condition Monitoring and Analysis of Reactor Coolant Pump

    International Nuclear Information System (INIS)

    Meng Qinghu; Meng Qingfeng; Feng Wuwei

    2012-01-01

    Fukushima nuclear power plant accident caused huge losses and pollution and it showed that the reactor coolant pump is very important in a nuclear power plant. Therefore, to keep the safety and reliability, the condition of the coolant pump needs to be online condition monitored and fault analyzed. In this paper, condition monitoring and analysis based on support vector machine (SVM) is proposed. This method is just to aim at the small sample studies such as reactor coolant pump. Both experiment data and field data are analyzed. In order to eliminate the noise and useless frequency, these data are disposed through a multi-band FIR filter. After that, a fault feature selection method based on principal component analysis is proposed. The related variable quantity is changed into unrelated variable quantity, and the dimension is descended. Then the SVM method is used to separate different fault characteristics. Firstly, this method is used as a two-kind classifier to separate each two different running conditions. Then the SVM is used as a multiple classifier to separate all of the different condition types. The SVM could separate these conditions successfully. After that, software based on SVM was designed for reactor coolant pump condition analysis. This software is installed on the reactor plant control system of Qinshan nuclear power plant in China. It could monitor the online data and find the pump mechanical fault automatically.

  8. Generalized saddle point condition for ignition in a tokamak reactor with temperature and density profiles

    International Nuclear Information System (INIS)

    Mitari, O.; Hirose, A.; Skarsgard, H.M.

    1989-01-01

    In this paper, the concept of a generalized ignition contour map, is extended to the realistic case of a plasma with temperature and density profiles in order to study access to ignition in a tokamak reactor. The generalized saddle point is found to lie between the Lawson and ignition conditions. If the height of the operation path with Goldston L-mode scaling is higher than the generalized saddle point, a reactor can reach ignition with this scaling for the case with no confinement degradation effect due to alpha-particle heating. In this sense, the saddle point given in a general form is a new criterion for reaching ignition. Peaking the profiles for the plasma temperature and density can lower the height of the generalized saddle point and help a reactor to reach ignition. With this in mind, the authors can judge whether next-generation tokamaks, such as Compact Ignition Tokamak, Tokamak Ignition/Burn Experimental Reactor, Next European Torus, Fusion Experimental Reactor, International Tokamak Reactor, and AC Tokamak Reactor, can reach ignition with realistic profile parameters and an L-mode scaling law

  9. What's down below? Current and potential future applications of geophysical techniques to identify subsurface permafrost conditions (Invited)

    Science.gov (United States)

    Douglas, T. A.; Bjella, K.; Campbell, S. W.

    2013-12-01

    For infrastructure design, operations, and maintenance requirements in the North the ability to accurately and efficiently detect the presence (or absence) of ground ice in permafrost terrains is a serious challenge. Ground ice features including ice wedges, thermokarst cave-ice, and segregation ice are present in a variety of spatial scales and patterns. Currently, most engineering applications use borehole logging and sampling to extrapolate conditions at the point scale. However, there is high risk of over or under estimating the presence of frozen or unfrozen features when relying on borehole information alone. In addition, boreholes are costly, especially for planning linear structures like roads or runways. Predicted climate warming will provide further challenges for infrastructure development and transportation operations where permafrost degradation occurs. Accurately identifying the subsurface character in permafrost terrains will allow engineers and planners to cost effectively create novel infrastructure designs to withstand the changing environment. There is thus a great need for a low cost rapidly deployable, spatially extensive means of 'measuring' subsurface conditions. Geophysical measurements, both terrestrial and airborne, have strong potential to revolutionize our way of mapping subsurface conditions. Many studies in continuous and discontinuous permafrost have used geophysical measurements to identify discrete features and repeatable patterns in the subsurface. The most common measurements include galvanic and capacitive coupled resistivity, ground penetrating radar, and multi frequency electromagnetic induction techniques. Each of these measurements has strengths, weaknesses, and limitations. By combining horizontal geophysical measurements, downhole geophysics, multispectral remote sensing images, LiDAR measurements, and soil and vegetation mapping we can start to assemble a holistic view of how surface conditions and standoff measurements

  10. The influence of fast reactor emergency conditions upon fuel element performance

    International Nuclear Information System (INIS)

    Bagdasarov, Yu.E.; Buksha, Yu.K.; Zabudko, L.M.; Likhachev, Yu.I.

    1985-01-01

    Fuel-pin cladding is one of the most important protective barriers preventing the release and propagation of radioactive contamination. By now the calculated determination of fast-reactor fuel-element performance under stationary conditions has been considered in detail but the investigation of the influence of emergency conditions has been given less attention. Under emergency conditions of the fast reactor operation there arise short-duration excesses of rated parameters (temperature, energy release, etc.) which are confined within tolerable limits with the use of the safety system. Some features of the sodium-cooled fast reactors (small mean prompt-neutron lifetime, relatively weak reactivity feedback, etc.) complicate the work of safety systems. Therefore, the tolerable deviations of parameters should be carefully validated

  11. Optimisation of Shift Reactor Operating Conditions to Maximise Hydrogen Production

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez, J. M.; Marano, M.; Ruiz, E.

    2011-07-28

    This report compiles the results of the work conducted by CIEMAT for Task 6.5 Shift reaction of the FLEXGAS project Near Zero Emission Advanced Fluidized Bed Gasification, which has been carried out with financial support from the Research Fund for Coal and Steel, RFCR-CT-2007-00005. The activity of an iron-chromium-based catalyst for the water gas shift reaction is studied. Results about WGS experiments conducted by CIEMAT on laboratory scale under different operating conditions are presented. The influence on the activity of the catalyst of main operating parameters- temperature, pressure, excess steam, and space velocity and gas composition - is evaluated and discussed. (Author) 19 refs.

  12. Studies of reactor waste conditioning and disposal at CRNL

    International Nuclear Information System (INIS)

    Beamer, N.V.; Bourne, W.T.; Buckely, L.P.; Pettipas, W.H.; Burrill, K.A.; Dixon, D.F.; Charlesworth, D.H.

    1982-09-01

    This report is a compilation of five papers presented at the Second Annual Meeting of the Canadian Nuclear Society in Ottawa, 1981 June. These papers describe recent progress in studies being conducted at the Chalk River Nuclear Laboratories related to the permanent disposal of low-and intermediate-level wastes arising in the Canadian nuclear industry. The principal topics discussed include waste processing by incineration, ultrafiltration and reverse osmosis, immobilization in bitumen and glass, and also the strategy for disposal of the conditioned wastes

  13. The moisture conditions of nuclear reactor concrete containment walls - an example for a BWR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nilsson, L.O.; Johansson, P. [Lund Institute of Technology, Laboratory of Building Materials, PO Box 118, 221 00 Lund (Sweden)

    2006-07-01

    A method is presented on how to quantify the moisture conditions of nuclear concrete containment walls. The method is based on first quantifying the boundary conditions at the outer and inner surfaces and then describing the moisture fixation and moisture transport within the concrete wall. The temperature and humidity conditions of the outdoor air and of the air close to the wall surfaces are monitored for a period of time and the vapour contents in the different points are compared. From the differences between the vapour contents the sources of moisture are identified and quantified. The previous and future climatic conditions are then predicted. An example is given for the conditions in the containment walls at Barsebaeck nuclear power plant, where moisture measurements have been performed in situ and on samples taken from the walls. (authors)

  14. Effect of redox conditions on pharmaceutical loss during biological wastewater treatment using sequencing batch reactors

    Energy Technology Data Exchange (ETDEWEB)

    Stadler, Lauren B., E-mail: lstadler@umich.edu [Department of Civil and Environmental Engineering, University of Michigan, 1351 Beal Avenue, EWRE, Ann Arbor, MI 48109 (United States); Su, Lijuan, E-mail: lijuansu@buffalo.edu [Department of Chemistry, University at Buffalo, State University of New York, Buffalo, NY 14260 (United States); Moline, Christopher J., E-mail: christopher.moline@hdrinc.com [Department of Civil and Environmental Engineering, University of Michigan, 1351 Beal Avenue, EWRE, Ann Arbor, MI 48109 (United States); Ernstoff, Alexi S., E-mail: alexer@dtu.dk [Department of Civil and Environmental Engineering, University of Michigan, 1351 Beal Avenue, EWRE, Ann Arbor, MI 48109 (United States); Aga, Diana S., E-mail: dianaaga@buffalo.edu [Department of Chemistry, University at Buffalo, State University of New York, Buffalo, NY 14260 (United States); Love, Nancy G., E-mail: nglove@umich.edu [Department of Civil and Environmental Engineering, University of Michigan, 1351 Beal Avenue, EWRE, Ann Arbor, MI 48109 (United States)

    2015-01-23

    Highlights: • Pharmaceutical fate was studied in SBRs operated at different redox conditions. • Stable carbon oxidation and nitrification occurred under microaerobic conditions. • Losses of atenolol and trimethoprim were highest under fully aerobic conditions. • Loss of sulfamethoxazole was highest under microaerobic conditions. • Deconjugation occurred during treatment to form sulfamethoxazole and desvenlafaxine. - Abstract: We lack a clear understanding of how wastewater treatment plant (WWTP) process parameters, such as redox environment, impact pharmaceutical fate. WWTPs increasingly install more advanced aeration control systems to save energy and achieve better nutrient removal performance. The impact of redox condition, and specifically the use of microaerobic (low dissolved oxygen) treatment, is poorly understood. In this study, the fate of a mixture of pharmaceuticals and several of their transformation products present in the primary effluent of a local WWTP was assessed in sequencing batch reactors operated under different redox conditions: fully aerobic, anoxic/aerobic, and microaerobic (DO concentration ≈0.3 mg/L). Among the pharmaceuticals that were tracked during this study (atenolol, trimethoprim, sulfamethoxazole, desvenlafaxine, venlafaxine, and phenytoin), overall loss varied between them and between redox environments. Losses of atenolol and trimethoprim were highest in the aerobic reactor; sulfamethoxazole loss was highest in the microaerobic reactors; and phenytoin was recalcitrant in all reactors. Transformation products of sulfamethoxazole and desvenlafaxine resulted in the reformation of their parent compounds during treatment. The results suggest that transformation products must be accounted for when assessing removal efficiencies and that redox environment influences the degree of pharmaceutical loss.

  15. Effect of redox conditions on pharmaceutical loss during biological wastewater treatment using sequencing batch reactors

    International Nuclear Information System (INIS)

    Stadler, Lauren B.; Su, Lijuan; Moline, Christopher J.; Ernstoff, Alexi S.; Aga, Diana S.; Love, Nancy G.

    2015-01-01

    Highlights: • Pharmaceutical fate was studied in SBRs operated at different redox conditions. • Stable carbon oxidation and nitrification occurred under microaerobic conditions. • Losses of atenolol and trimethoprim were highest under fully aerobic conditions. • Loss of sulfamethoxazole was highest under microaerobic conditions. • Deconjugation occurred during treatment to form sulfamethoxazole and desvenlafaxine. - Abstract: We lack a clear understanding of how wastewater treatment plant (WWTP) process parameters, such as redox environment, impact pharmaceutical fate. WWTPs increasingly install more advanced aeration control systems to save energy and achieve better nutrient removal performance. The impact of redox condition, and specifically the use of microaerobic (low dissolved oxygen) treatment, is poorly understood. In this study, the fate of a mixture of pharmaceuticals and several of their transformation products present in the primary effluent of a local WWTP was assessed in sequencing batch reactors operated under different redox conditions: fully aerobic, anoxic/aerobic, and microaerobic (DO concentration ≈0.3 mg/L). Among the pharmaceuticals that were tracked during this study (atenolol, trimethoprim, sulfamethoxazole, desvenlafaxine, venlafaxine, and phenytoin), overall loss varied between them and between redox environments. Losses of atenolol and trimethoprim were highest in the aerobic reactor; sulfamethoxazole loss was highest in the microaerobic reactors; and phenytoin was recalcitrant in all reactors. Transformation products of sulfamethoxazole and desvenlafaxine resulted in the reformation of their parent compounds during treatment. The results suggest that transformation products must be accounted for when assessing removal efficiencies and that redox environment influences the degree of pharmaceutical loss

  16. Synchrotron in-situ deformation experiments of perovskite + (Mg,Fe)O aggregates under shallow lower mantle conditions (Invited)

    Science.gov (United States)

    Girard, J.; Amulele, G.; Farla, R. J.; Liu, Z.; Mohiuddin, A.; Karato, S.

    2013-12-01

    Experimental studies on rheological properties of mantle's minerals are crucial to understand the dynamics of Earth's interior, but direct experimental studies under the relevant lower mantle conditions are challenging. Most of the earlier studies were performed at lower mantle pressures but low temperatures using DAC (diamond anvil cell) (e.g., Merkel et al., 2003)), and in DAC experiments strain-rate and stress are unknown. Some previous studies were carried out under high pressures and high temperatures (e.g, Cordier et al., 2004) , but quantitative results on rheological behaviour of said minerals were not obtained. Here we present the results of the first in-situ deformation experiments of perovskite + (Mg,Fe)O (Pv + fp) aggregates using RDA (rotational Drickamer apparatus). The RDA has a better support for the anvils at high pressure than the more commonly used D-DIA apparatus and hence we can reach higher pressures and temperatures than the D-DIA. We have recently made new modifications to the cell assembly to reach the lower mantle conditions with less interference in X-ray diffraction patterns by the surrounding materials. The starting material was ringwoodite synthesized using a multi-anvil. In-situ deformation experiments were then carried at pressure up to 28 GPa (calculated from thermal EOS of Pt) and estimated temperatures up to 2200 K using RDA. Under these conditions, ringwoodite transformed to Pv + fp. We subsequently deformed the sample between strain rates of 10-4 to 10-5 s-1. Stress and strain were measured in-situ using X-ray synchrotron beam. The recovered sample analyses show evidence of perovskite+(Mg,Fe)O microstructure (Fig. 1). The radial X-ray diffraction data are being analysed to determine the stress levels of two minerals. Also microstructures of deformed specimens are studied to understand the deformation mechanisms and strain partitioning. The results will contribute towards our understanding of the rheological properties of the

  17. Reactor internals design/analysis for normal, upset, and faulted conditions

    International Nuclear Information System (INIS)

    Burke, F.R.

    1977-06-01

    The analytical procedures used by Babcock and Wilcox to demonstrate the structural integrity of the 205-FA reactor internals are described. Analytical results are presented and compared to ASME Code allowable limits for Normal, Upset, and Faulted conditions. The particular faulted condition considered is a simultaneous loss-of-coolant accident and safe shutdown earthquake. The operating basis earthquake is addressed as an Upset condition

  18. Operational limits and conditions and operating procedures for research reactors. Safety guide

    International Nuclear Information System (INIS)

    2008-01-01

    This publication provides practical guidance on all important aspects of developing, formulating and presenting the operational limits and conditions as well as the operating procedures for research reactors. It covers the concept of operational limits and conditions, their content, and the responsibilities of the operating organization with respect to their establishment, modification, documentation and compliance. The guidance also covers the training of operating personnel on performing periodic testing, established by the operational limits and conditions, and operating procedures

  19. Opinion about difficulties of RA reactor operation under conditions of high activity of the heavy water system - Annex 2

    International Nuclear Information System (INIS)

    Nikolic, M.

    1963-01-01

    It was concluded that reactor the reactor operation is very dangerous for the reactor installation as well as safety of the staff under conditions of heavy water increased activity. Two fundamental arguments in favour of this conclusion are: insufficient possibility of reactor components inspection during maintenance and operation in the future period; difficulties in prevention of accidents that could occur is equally dangerous for the reactor facility and the environment. Cleaning and decontamination of the complete heavy water system is needed before the reactor operation starts in order to avoid possible failures or accidental events [sr

  20. The early warning system of landslides and sediment runoffs using meteorological condition including rainfall-soil moisture index (Invited)

    Science.gov (United States)

    Kubota, T.; Silva, I. C.; Hasnawir, H.

    2009-12-01

    The research including observation of rain, soil moisture content and sediment discharge is conducted on a torrent in northern Kyushu whose geology consists of Paleozoic metamorphic rocks (mainly schist) and whose vegetation consists of mainly Japanese cypress and cedar. Soil depth is approximately 50cm in average and permeability k is 0.1~0.01 order. With data obtained by the observation for more than 4 years, standard rainfalls of warning and evacuation against the sudden sediment runoffs are analyzed. Then, the result was compared with the ones in Nuevo Leon Mexico (geology of schist, slate, k=0.01~0.001 order) and in southern Sulawesi Island Indonesia (volcanic geology, k=0.001~0.0001 order). Hitherto, various methods were proposed to analyze the warning critical standard for landslide disaster or large sediment discharge. In this study, we employed Hirano's element slope runoff theory, the Self Organized Criticality Assumption (SOC), and the Elementary Catastrophe Theory (ETC) to analyze the data, although the soil moisture fluctuation, meteorological condition such as upper air wind and dew point depression, the rainfall-soil moisture index provided by Japan Meteorological Agency was considered. The last one is a cutting edge technology based on the tank model calculation of soil moisture content combined with short term rainfall prediction which is a product of numerical simulation using radar image advection analysis compensated with surface rain data and with orographic rain effect. In Hirano's theory, we can describe the critical rain Rc and rain intensity Ric as following equation. Q/A/M/ cosθ = Ri ∫(r*cosθ)dt = Ri*R (1) ∴ Ric*Rc = C (2) Here, Q: sediment runoff or debris flow discharge, A: watershed area, M: function concerning with sediment deposit features on the upstream torrents or slopes (porosity, torrent bed slope gradient, sediment accumulation length and depth, cohesion), t: time, θ: torrent bed or hillside slope gradient, r: instant

  1. Possibilities of crack mouth opening displacement (CMOD) measurement under boiling and pressurized water reactor conditions

    International Nuclear Information System (INIS)

    Ehling, W.

    1984-01-01

    Fracture mechanics investigations carried out so far in laboratory conditions cover only part of the material stresses, as effects which occur in nuclear powerstations, in particular, such as corrosion and radioactive radiation are largely left out of account. Therefore experiments including these effects were recently carried out in autoclaves, test rigs simulating reactors (HRD experimental plant) and in experimental reactors. An important parameter of experimental fracture mechanics is the measurement of crack opening displacement (COD). The crack opening is measured with socalled clip gauges (transmitters based on strain gauges, which convert mechanical deformation of springs into electrical signals) on standard samples in the laboratory. It was therefore sensible to use these high temperature strain gauges (HTD) for the development of a measuring system for travel for pressurized water and boiling water reactor conditions. (orig.) [de

  2. Criticality safety studies involved in actions to improve conditions for storing 'RA' research reactor spent fuel

    International Nuclear Information System (INIS)

    Matausek, M.; Marinkovic, N.

    1998-01-01

    A project has recently been initiated by the VINCA Institute of Nuclear Sciences to improve conditions in the spent fuel storage pool at the 6.5 MW research reactor RA, as well as to consider transferring this spent fuel into a new dry storage facility built for the purpose. Since quantity and contents of fissile material in the spent fuel storage at the RA reactor are such that possibility of criticality accident can not be a priori excluded, according to standards and regulations for handling fissile material outside a reactor, before any action is undertaken subcriticality should be proven under normal, as well as under credible abnormal conditions. To perform this task, comprehensive nuclear criticality safety studies had to be performed. (author)

  3. Sensitivity of South American tropical climate to Last Glacial Maximum boundary conditions: focus on teleconnections with tropics and extratropics (Invited)

    Science.gov (United States)

    Khodri, M.; Kageyama, M.; Roche, D. M.

    2009-12-01

    dynamical equilibrium due to both reduced GHG and land ice sheets. We also show that the overall tropical Pacific circulation response to land ice albedo alone consists in a substantial thermo-dynamical stabilisation of the equatorial atmosphere. The upper troposphere warming spreading out from South East Central Pacific, analogous to the atmosphere response to El-Niño conditions, results in enhanced rainfall over Nordeste and Southeastern Brazil. Such tropics-tropics teleconnection is essential to explain the moistening of the southern tropics, amplifying thereby the influence of the extratropical atmosphere on the LGM tropical climate.

  4. Thermalydraulic processes in the reactor coolant system of a BWR under severe accident conditions

    International Nuclear Information System (INIS)

    Hodge, S.A.

    1990-01-01

    Boiling water reactors (BWRs) incorporate many unique structural features that make their expected response under severe accident conditions very different from that predicted in the case of pressurized water reactor accident sequences. Automatic main steam isolation valve (MIV) closure as the vessel water level approaches the top of the core would cause reactor vessel isolation while automatic recirculation pump trip would limit the in-vessel flows to those characteristic of natural circulation (as disturbed by vessel relief valve actuation). This paper provides a discussion of the BWR control blade, channel box, core plate, control rod guide tube, and reactor vessel safety relief valve (SRV) configuration and the effects of these structural components upon thermal hydraulic processes within the reactor vessel under severe accident conditions. The dominant BWR severe accident sequences as determined by probabilistic risk assessment are described and the expected timing of events for the unmitigated short-term station blackout severe accident sequence at the Peach Bottom atomic power station is presented

  5. Catalytic effect of different reactor materials under subcritical water conditions: decarboxylation of cysteic acid into taurine

    Science.gov (United States)

    Faisal, M.

    2018-03-01

    In order to understand the influence of reactor materials on the catalytic effect for a particular reaction, the decomposition of cysteic acid from Ni/Fe-based alloy reactors under subcritical water conditions was examined. Experiments were carried out in three batch reactors made of Inconel 625, Hastelloy C-22 and SUS 316 over temperatures of 200 to 300 °C. The highest amount of eluted metals was found for SUS 316. The results demonstrated that reactor materials contribute to the resulting product. Under the tested conditions, cysteic acid decomposes readily with SUS 316. However, the Ni-based materials (Inconel 625 and Hastelloy C-22) show better resistance to metal elution. It was found that among the materials used in this work, SUS 316 gave the highest reaction rate constant of 0.1934 s‑1. The same results were obtained at temperatures of 260 and 300 °C. Investigation of the Arrhenius activation energy revealed that the highest activation energy was for Hastelloy C-22 (109 kJ/mol), followed by Inconel 625 (90 kJ/mol) and SUS 316 (70 kJ/mol). The decomposition rate of cysteic acid was found to follow the results for the trend of the eluted metals. Therefore, it can be concluded that the decomposition of cysteic acid was catalyzed by the elution of heavy metals from the surface of the reactor. The highest amount of taurine from the decarboxylation of cysteic acid was obtained from SUS 316.

  6. Test of safety injection supply by diesel generator under reactor vessel closed condition

    International Nuclear Information System (INIS)

    Zhang Hao; Bi Fengchuan; Che Junxia; Zhang Jianwen; Yang Bo

    2014-01-01

    The paper studied that the test of diesel generator full load take-up under the condition of actual safety injection and reactor vessel closed in Ningde nuclear project unit l. It is proved that test result accorded with design criteria, meanwhile, the test was removed from the key path of project schedule, which cut a huge cost. (authors)

  7. Temperature conditions of foundation plates under nuclear power plant reactor compartments

    International Nuclear Information System (INIS)

    Ehsaulov, S.L.

    1990-01-01

    Method for calculation of temperature conditions for foundation plates under reactor compartments located in the main building, used in construction of the second stage of the Kostroma nuclear power plant, is considered. The obtained calculation data can be used for determining the most suitable period of concrete placement, composition, initial temperature, manufacturing technology and ways of delivery of concrete mixture

  8. Effects of ocean conditions upon the passive residual heat removal system (PRHRS) of ship reactor

    International Nuclear Information System (INIS)

    Su Guanghui; Zhang Jinling; Guo Yujun; Qiu Suizheng; Yu Zhenwan; Jia Dounan

    1996-01-01

    The authors investigate the influence of ocean conditions (heaving, listing, rolling) on the natural circulation flow and the ability of heat transfer of the ship reactor's PRHRS, and develops a mathematical model. A program, MISAP 02, is compiled with the structured FORTRAN 77 using the advanced Gear method. the program is used to calculate the above influence. The results show that the ocean conditions have some effects on the natural circulation flow and the ability of heat transfer

  9. On the influence of fusion reactor conditions on optical properties of metallic plasma-viewing mirrors

    International Nuclear Information System (INIS)

    Voitsenya, V.S.; Gritsyna, V.I.; Konovalov, V.G.; Ruzhitskij, V.V.; Shapoval, A.N.; Orlinskij, D.V.

    1997-01-01

    This paper presents the results of imitation experiments concerning the effects of fusion reactor conditions on the properties of mirrors made of stainless steel, copper and beryllium. The neutron irradiation was imitated using MeV energy range ions. To imitate the effects of charge exchange atoms (CXA) bombardment, keV energy range D + and He + ions were used. From the data obtained it was concluded that not only the reflectivity but also the resistance to CXA sputtering have to be taken into account when choosing the materials for the first mirrors of a fusion reactor. (orig.)

  10. Development of RETRAN-03/MOV code for thermal-hydraulic analysis of nuclear reactor under moving conditions

    International Nuclear Information System (INIS)

    Kim, Hak Jae; Park, Goon Cherl

    1996-01-01

    Nuclear ship reactors have several; features different from land-based PWR's. Especially, effects of ship motions on reactor thermal-hydraulics and good load following capability for abrupt load changes are essential characteristics of nuclear ship reactors. This study modified the RETRAN-03 to analyze the thermal-hydraulic transients under three-dimensional ship motions, named RETRAN-03/MOV in order to apply to future marine reactors. First Japanese nuclear ship MUTSU reactor have been analyzed under various ship motions to verify this code. Calculations have been performed under rolling,heaving and stationary inclination conditions during normal operation. Also, the natural circulation has been analyzed, which can provide the decay heat removed to ensure the passive safety of marine reactors. As results, typical thermal-hydraulic characteristics of marine reactors such as flow rate oscillations and S/G water level oscillations have been successfully simulated at various conditions. 7 refs., 11 figs. (author)

  11. Effect of redox conditions on pharmaceutical loss during biological wastewater treatment using sequencing batch reactors

    DEFF Research Database (Denmark)

    Stadler, Lauren B.; Su, Lijuan; Moline, Christopher J.

    2015-01-01

    We lack a clear understanding of how wastewater treatment plant (WWTP) process parameters, such as redox environment, impact pharmaceutical fate. WWTPs increasingly install more advanced aeration control systems to save energy and achieve better nutrient removal performance. The impact of redox...... under different redox conditions: fully aerobic, anoxic/aerobic, and microaerobic (DO concentration ≈0.3 mg/L). Among the pharmaceuticals that were tracked during this study (atenolol, trimethoprim, sulfamethoxazole, desvenlafaxine, venlafaxine, and phenytoin), overall loss varied between them...... and between redox environments. Losses of atenolol and trimethoprim were highest in the aerobic reactor; sulfamethoxazole loss was highest in the microaerobic reactors; and phenytoin was recalcitrant in all reactors. Transformation products of sulfamethoxazole and desvenlafaxine resulted in the reformation...

  12. Small power reactor projects in the United States of America and Canada. Information gathered as a result of invitations from Member States

    International Nuclear Information System (INIS)

    1962-01-01

    As part of its activities in connection with the development of nuclear power, and in response to the resolutions adopted by the General Conference, the Agency has been undertaking a continuing study of the technology and economics of small and medium sized power reactors, particularly with reference to the needs of the less-developed countries. This report summarizes the information gathered on the small power reactor projects in the United States of America and Canada, as a result of the opportunity afforded by these Member States to the Agency. It may be recalled that, at the third regular session of the General Conference, the United States Government offered to provide the Agency with relevant technical and economic data on several small power reactor projects of its Atomic Energy Commission. The Agency accepted the offer and since June 1960 it has sent one or two staff members at approximately six-monthly intervals to follow the development of nine power reactor projects in the United States which represent six different reactor systems. Last year, the Agency issued a report summarizing the information obtained through their visits and study of available published literature. The present document, which should be read in conjunction with that document, brings the information up to date and provides additional information on certain phases of the projects already discussed in the last report. Three more power reactor projects are also dealt with, namely the experimental gas-cooled reactor (EGCR), the high temperature gas-cooled reactor (HTGR) and the Hallam nuclear power facility (HNPF). Early in 1962, the Canadian Government expressed its willingness to make available to the Agency relevant information on the NPD and CANDU projects. The coverage of the NPD reactor is based upon the published information supplied by AECL of Canada and the visit by one of the staff members to the NPD site. The Agency wishes to acknowledge with thanks the co-operation extended

  13. The effect of steam cycle conditions upon the economics and design of a sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Philpott, E.F.; Pounder, F.; Willby, C.R.

    1978-01-01

    The paper studies the effect of variation of steam and feedwater conditions upon the economics, design and layout of a sodium-cooled fast reactor. The parameters investigated are steam temperature and pressure, feedwater temperature, and boiler recirculation ratio. The paper also includes an assessment of the effects of associating the fast reactor with saturated steam cycle conditions. (author)

  14. Adaptation of continuous biogas reactors operating under wet fermentation conditions to dry conditions with corn stover as substrate.

    Science.gov (United States)

    Kakuk, Balázs; Kovács, Kornél L; Szuhaj, Márk; Rákhely, Gábor; Bagi, Zoltán

    2017-08-01

    Corn stover (CS) is the agricultural by-product of maize cultivation. Due to its high abundance and high energy content it is a promising substrate for the bioenergy sector. However, it is currently neglected in industrial scale biogas plants, because of its slow decomposition and hydrophobic character. To assess the maximum biomethane potential of CS, long-term batch fermentations were carried out with various substrate concentrations and particle sizes for 72 days. In separate experiments we adapted the biogas producing microbial community in wet fermentation arrangement first to the lignocellulosic substrate, in Continuous Stirred Tank Reactor (CSTR), then subsequently, by continuously elevating the feed-in concentration, to dry conditions in solid state fermenters (SS-AD). In the batch tests, the CSTR experiment, the daily substrate loading was gradually increased from 1 to 2 g vs /L/day until the system produced signs of overloading. Then the biomass was transferred to SS-AD reactors and the adaptation process was studied. Although the specific methane yields were lower in the SS-AD arrangement (177 mL CH 4 /g vs in CSTR vs. 105 mL in SS-AD), the benefits of process operational parameters, i.e. lower energy consumption, smaller reactor volume, digestate amount generated and simpler configuration, may compensate the somewhat lower yield. Copyright © 2017 The Authors. Published by Elsevier Ltd.. All rights reserved.

  15. Role of plasma enhanced atomic layer deposition reactor wall conditions on radical and ion substrate fluxes

    Energy Technology Data Exchange (ETDEWEB)

    Sowa, Mark J., E-mail: msowa@ultratech.com [Ultratech/Cambridge NanoTech, 130 Turner Street, Building 2, Waltham, Massachusetts 02453 (United States)

    2014-01-15

    Chamber wall conditions, such as wall temperature and film deposits, have long been known to influence plasma source performance on thin film processing equipment. Plasma physical characteristics depend on conductive/insulating properties of chamber walls. Radical fluxes depend on plasma characteristics as well as wall recombination rates, which can be wall material and temperature dependent. Variations in substrate delivery of plasma generated species (radicals, ions, etc.) impact the resulting etch or deposition process resulting in process drift. Plasma enhanced atomic layer deposition is known to depend strongly on substrate radical flux, but film properties can be influenced by other plasma generated phenomena, such as ion bombardment. In this paper, the chamber wall conditions on a plasma enhanced atomic layer deposition process are investigated. The downstream oxygen radical and ion fluxes from an inductively coupled plasma source are indirectly monitored in temperature controlled (25–190 °C) stainless steel and quartz reactors over a range of oxygen flow rates. Etch rates of a photoresist coated quartz crystal microbalance are used to study the oxygen radical flux dependence on reactor characteristics. Plasma density estimates from Langmuir probe ion saturation current measurements are used to study the ion flux dependence on reactor characteristics. Reactor temperature was not found to impact radical and ion fluxes substantially. Radical and ion fluxes were higher for quartz walls compared to stainless steel walls over all oxygen flow rates considered. The radical flux to ion flux ratio is likely to be a critical parameter for the deposition of consistent film properties. Reactor wall material, gas flow rate/pressure, and distance from the plasma source all impact the radical to ion flux ratio. These results indicate maintaining chamber wall conditions will be important for delivering consistent results from plasma enhanced atomic layer deposition

  16. Conditioning for definitive storage of radioactive graphite bricks from reactor decommissioning

    International Nuclear Information System (INIS)

    Costes, J.R.; Koch, C.; Tassigny, C. de; Vidal, H.; Raymond, A.

    1990-01-01

    The decommissioning of gas-graphite reactors in the EC (e.g. French UNGGs, British Magnox reactors and AGRs, and reactors in Spain and in Italy) will produce large amounts of graphite bricks. This graphite cannot be accepted without particular conditioning by the existing shallow land disposal sites. The aim of the study is to examine the behaviour of graphite waste and to develop a conditioning technique which makes this waste acceptable for shallow land disposal sites. 18 kg of graphite core samples with an outside diameter of 74 mm were removed from the G2 gas-cooled reactor at Marcoule. Their radioactivity is highly dependent on the position of the graphite bricks inside the reactor. Measured results indicate an activity range of 100-400 MBq/kg with 90% Tritium, 5% 14 C, 3% 60 Co, 1.5% 63 Ni. Repeated porosity analyses showed that open porosity ranging from 0 to 100 μm exceeded 23 vol% in the graphite. Water penetration kinetics were investigated in unimpregnated graphite and resulted in impregnation by water of 50-90% of the open porosity. Preliminary lixiviation tests on the crude samples showed quick lixidegree of Cs (several per cent) and of 60 Co, and 133 Ba at a lesser degree. The proposed conditioning technique does not involve a simple coating but true impregnation by a tar-epoxy mixture. The bricks recovered intact from the core by robot services will be placed one by one inside a cylindrical metallic container. But this container may corrode and the bricks may become fragmented in the future, the normally porous graphite will be unaffected by leaching since it is proved that all pores larger than 0.1 μm will be filled with the tar-epoxy mixture. This is a true long-term waste packaging concept. The very simple technology required for industrial implementation is discussed

  17. Reactor

    International Nuclear Information System (INIS)

    Toyama, Masahiro; Kasai, Shigeo.

    1978-01-01

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

  18. Tools and techniques for ageing predictions in nuclear reactors through condition monitoring

    International Nuclear Information System (INIS)

    Verma, R.M.P.

    1994-01-01

    To operate the nuclear reactors beyond their design predicted life is gaining importance because of huge replacement and decommissioning costs. But experience shows that nuclear plant safety and reliability may decline in the later years of plant life due to ageing degradation. Ageing of nuclear plant components, structures and systems, if unmitigated reduces their safety margins provided in the design and thus increases risks to public health and safety. These safety margins must be monitored throughout plant service life including any extended life. Condition monitoring of nuclear reactor components/equipment and systems can be done to study the effect of ageing, status of safety margins and effect of corrective and mitigating actions taken. The tools and techniques of condition monitoring are also important in failure trending, predictive maintenance, evaluation of scheduled maintenance, in mitigation of ageing, life extension and reliability studies. (author). 1 fig., 1 annexure

  19. Long- and short-term trends in vessel conditioning of TFTR [Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    LaMarche, P.H.; Dylla, H.F.; Bell, M.G.

    1986-10-01

    We have investigated trends in the conditioning of the Tokamak Fusion Test Reactor (TFTR) vacuum vessel during the May 1984 to April 1985 run period. The initial conditioning of the vessel, consisting of glow discharge cleaning (GDC) and pulse discharge cleaning (PDC) in concert with a 150 0 C vessel bakeout, is necessary to assure plasma operation after atmospheric venting. A long-term conditioning process, ascribed to limiter conditioning, effectively improves operational conditions during the course of the run. Over several thousand high power plasma discharges, the improvement was documented by using standard parameter (fiducial) plasma discharges. Several techniques demonstrated short-term improvements in vessel conditioning during this time period, including: Cr gettering and programming the plasma position relative to the limiter contact area

  20. Reactor

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  1. Influence of geometrical parameters of the VVER-1000 reactor construction elements to internals irradiation conditions

    Directory of Open Access Journals (Sweden)

    О. M. Pugach

    2015-07-01

    Full Text Available Investigations to determine the influences of geometrical parameters of the calculational VVER-1000 reactor model to the results of internal irradiation condition determination are carried out. It is shown that the values of appropriate sensitivity matrix elements are not dependent on a height coordinate for any core level, but there is their azimuthal dependence. Maximum possible relative biases of neutron fluence due to inexact knowledge of internal geometrical parameters are obtained for the baffle and the barrel.

  2. Effect of decontamination on oxidation of austenitic stainless steel in reactor conditions

    International Nuclear Information System (INIS)

    Starkman, T.

    1984-07-01

    Austenitic stainless steels were oxidized in static autoclaves in light water reactor conditions. After the autoclave treatments the specimens were decontaminated with the aid of alkaline potassium permanganate (AP) and oxalic and citric acid (CITROX) as well as electrochemically in H 3 PO 4 . Alternating oxidation and decontamination tests were performed. An elemental analysis of the surfaces of the specimens was carried out by electron spectroscopy. Changes in structures and thicknesses of the oxide layers were observed. (author)

  3. Propagation of cracks by stress corrosion in conditions of BWR type reactor

    International Nuclear Information System (INIS)

    Merino C, F.J.; Fuentes C, P.

    2004-01-01

    In this work, the obtained results when applying the Hydrogen Chemistry to a test tube type Compact Tension (CT), built in austenitic stainless steel 304l, simulating the conditions to those that it operates a Boiling Water Reactor (BWR), temperature 288 C and pressure of 8 MPa are presented. With the application of this water chemistry, seeks to be proven the diminution of the crack propagation speed. (Author)

  4. Development of a diagnostic system for identifying accident conditions in a reactor

    International Nuclear Information System (INIS)

    Santhosh; Gera, B.; Kumar, Mithilesh; Thangamani, I.; Prasad, Hari; Srivastava, A.; Dutta, Anu; Sharma, Pavan K.; Majumdar, P.; Verma, V.; Mukhopadhyay, D.; Ganju, Sunil; Chatterjee, B.; Sanyasi Rao, V.V.S.; Lele, H.G.; Ghosh, A.K.

    2009-07-01

    This report describes a methodology for identification of accident conditions in a nuclear reactor from the signals available to the operator. A large database of such signals is generated through analyses - for core, containment, environmental dispersion and radiological dose to train a computer code based on an Artificial Neural Networks (ANNs). At present, in the prediction mode, information on LOCA (location and size of break), status of availability of ECCS, and expected doses can be predicted well for a 220 MWe PHWR. (author)

  5. Failure strains and proposed limit strains for an reactor pressure vessel under severe accident conditions

    International Nuclear Information System (INIS)

    Krieg, R.

    2005-01-01

    The local failure strains of essential design elements of a reactor vessel are investigated. The size influence of the structure is of special interest. Typical severe accident conditions including elevated temperatures and dynamic loads are considered. The main part of work consists of test families with specimens under uniaxial and biaxial load. Within one test family the specimen geometry and the load conditions are similar, but the size is varied up to reactor dimensions. Special attention is given to geometries with a hole or a notch causing non-uniform stress and strain distributions typical for the reactor vessel. A key problem is to determine the local failure strain. Here suitable methods had to be developed including the so-called 'vanishing gap method', and the 'forging die method'. They are based on post-test geometrical measurements of the fracture surfaces and reconstructions of the related strain fields using finite element models. The results indicate that stresses versus dimensionless deformations are approximately size independent up to failure for specimens of similar geometry under similar load conditions. Local failure strains could be determined. The values are rather high and size dependent. Statistical evaluation allow the proposal of limit strains which are also size dependent. If these limit strains are not exceeded, the structures will not fracture

  6. Advanced development and operating experience with a canned motor pump under pressurized water reactor conditions

    International Nuclear Information System (INIS)

    Dittmer, H.; Reymann, A.; Seibig, B.; Reinecker, H.

    1988-01-01

    At the research reactor FRG-2, Geesthacht, an irradiation device is in operation for testing defective light-water-reactor (LWR) test fuel rods under pressurized water reactor conditions (320 0 C, 160 bar). The requirements to the canned motor pump for cooling water circulation: medium: Demineralized water, operating temperature 320 0 C, operating pressure 155 bar, radiation field of the reactor, integration in the irradiation capsule, helium leak rate -6 mbar.dm 3 .s -1 , minimum working life 3000 hours, were high and caused difficulties in the acquisition of this component. First test runs with supplied pumps showed that the desired working life could not be achieved. The results of the development steps, the test runs, and the performance in service show that for our range of applications, the best combination of materials for the radial bearings is silicon-infiltrated SiC (8% free Si) against the same material. These bearings allowed a good working life for the pump to be achieved. (orig./GL) [de

  7. A matrix formalism to solve interface condition equations in a reactor system

    Energy Technology Data Exchange (ETDEWEB)

    Matausek, M V [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

    1970-05-15

    When a nuclear reactor or a reactor lattice cell is treated by an approximate procedure to solve the neutron transport equation, as the last computational step often appears a problem of solving systems of algebraic equations stating the interface and boundary conditions for the neutron flux moments. These systems have usually the coefficient matrices of the block-bi diagonal type, containing thus a large number of zero elements. In the present report it is shown how such a system can be solved efficiently accounting for all the zero elements both in the coefficient matrix and in the free term vector. The procedure is presented here for the case of multigroup P{sub 3} calculation of neutron flux distribution in a cylindrical reactor lattice cell. Compared with the standard gaussian elimination method, this procedure is more advantageous both in respect to the number of operations needed to solve a given problem and in respect to the computer memory storage requirements. A similar formalism can also be applied for other approximate methods, for instance for multigroup diffusion treatment of a multi zone reactor. (author)

  8. VALIDATION OF SIMBAT-PWR USING STANDARD CODE OF COBRA-EN ON REACTOR TRANSIENT CONDITION

    Directory of Open Access Journals (Sweden)

    Muhammad Darwis Isnaini

    2016-03-01

    Full Text Available The validation of Pressurized Water Reactor typed Nuclear Power Plant simulator developed by BATAN (SIMBAT-PWR using standard code of COBRA-EN on reactor transient condition has been done. The development of SIMBAT-PWR has accomplished several neutronics and thermal-hydraulic calculation modules. Therefore, the validation of the simulator is needed, especially in transient reactor operation condition. The research purpose is for characterizing the thermal-hydraulic parameters of PWR1000 core, which be able to be applied or as a comparison in developing the SIMBAT-PWR. The validation involves the calculation of the thermal-hydraulic parameters using COBRA-EN code. Furthermore, the calculation schemes are based on COBRA-EN with fixed material properties and dynamic properties that calculated by MATPRO subroutine (COBRA-EN+MATPRO for reactor condition of startup, power rise and power fluctuation from nominal to over power. The comparison of the temperature distribution at nominal 100% power shows that the fuel centerline temperature calculated by SIMBAT-PWR has 8.76% higher result than COBRA-EN result and 7.70% lower result than COBRA-EN+MATPRO. In general, SIMBAT-PWR calculation results on fuel temperature distribution are mostly between COBRA-EN and COBRA-EN+MATPRO results. The deviations of the fuel centerline, fuel surface, inner and outer cladding as well as coolant bulk temperature in the SIMBAT-PWR and the COBRA-EN calculation, are due to the value difference of the gap heat transfer coefficient and the cladding thermal conductivity.

  9. Study on vortex cavitation in a compact fast reactor. Effects of system pressure on inception condition

    International Nuclear Information System (INIS)

    Hiroyuki Sato; Toshiki Ezure; Hideki Kamide

    2005-01-01

    A compact sodium reactor is designed as a commercialized fast reactor cycle system. A 1/10 scaled water experiment was performed to optimize flow in an upper plenum of the reactor vessel, because of high flow velocity resulted from the compacted vessel. In the experiment, vortex cavitation was found at the hot leg inlet because of high velocity in the hot leg pipe (9.4m/s in the design). To evaluate cavitation inception condition of the commercialized reactor, we use the cavitation number k in order to consider the difference of system pressures (0.1MPa in the experiment and 0.3MPa in the design). The minimum pressure at the vortex center will depend on vortex core radius (size of forced vortex region). It is related to axial velocity gradient and fluid viscosity in theory of the Burger's stretched vortex model. We carried out a basic water experiment to investigate the influence of system pressure and fluid viscosity on the vortex cavitation. The cavitation number at the inception of vortex cavitation slightly increased according to the increase of the system pressure. It means that the vortex cavitation occurs easily under higher pressure condition as compared with the similar condition of cavitation number with lower pressure. However the increase was less than 30% when the system pressure was varied from 0.1 to 0.3MPa. The influence of fluid viscosity was examined by change of fluid temperature. Velocity distribution around the vortex was also measured to see the structure of vortex. (authors)

  10. Generation of net electric power with a tokamak reactor under foreseeable physical and engineering conditions

    International Nuclear Information System (INIS)

    Hiwatari, R.; Asaoka, Y.; Okano, K.; Yoshida, T.; Tomabechi, K.

    2004-01-01

    This study reveals for the first time the plasma performance required for a tokamak reactor to generate net electric power under foreseeable engineering conditions. It was found that the reference plasma performance of the ITER inductive operation mode with β N = 1.8, HH = 1.0, andf nGW 0.85 had sufficient potential to achieve the electric break-even condition (net electric power P e net = 0MW) under the following engineering conditions: machine major radius 6.5m ≤ R p ≤ 8.5m, the maximum magnetic field on TF coils B tmax = 16 T, thermal efficiency η e 30%, and NBI system efficiency η NBI = 50%. The key parameters used in demonstrating net electric power generation in tokamak reactors are β N and fη GW . ≥ 3.0 is required for P e net ∼ 600MW with fusion power P f ∼ 3000MW. On the other hand, fη GW ≥ 1.0 is inevitable to demonstrate net electric power generation, if high temperatures, such as average temperatures of T ave > 16 keV, cannot be selected for the reactor design. To apply these results to the design of a tokamak reactor for demonstrating net electric power generation, the plasma performance diagrams on the Q vs P f (energy multiplication factor vs fusion power) space for several major radii (i.e. 6.5, 7.5, and 8.5 m) were depicted. From these figures, we see that a design with a major radius R p ∼ 7.5m seems preferable for demonstrating net electric power generation when one aims at early realization of fusion energy. (author)

  11. Two new techniques for the remote evaluation of reactor steel condition - microscopic removal and surface examination

    International Nuclear Information System (INIS)

    Clayton, R.

    Much reactor inspection work involves an assessment of the condition of structural steel. This paper reviews two different techniques which provide information for such an assessment. The first - micro-sample removal (for the measurement of surface oxide thickness and chemical composition) - requires contact with the steel surface, whereas the second - a 'teach and learn' photographic technique (in which a special photogrammatic camera is used to obtain high-quality close-up photographs, to assess surface condition and corrosion growth) can obtain surface information on inaccessible components. (author)

  12. A structural evaluation of the Shippingport reactor pressure vessel for transport impact conditions

    International Nuclear Information System (INIS)

    Witte, M.C.; Chou, C.K.

    1989-01-01

    The Shippingport Atomic Power Station in Shippingport, Pennsylvania, is being decommissioned and dismantled. This government-leased property will be returned, in a radiologically safe condition, to its owner. All radioactive material is being removed from the Shippingport Station and transported for burial to the DOE Hanford Reservation in Richland, Washington. The reactor pressure vessel (RPV) will be transported by barge to Hanford. This paper describes an evaluation of the structural response of the RPV to the normal and accident impact test conditions as required by the Code of Federal Regulations. 3 refs., 5 figs., 3 tabs

  13. Multiregion, multigroup collision probability method with white boundary condition for light water reactor thermalization calculations

    International Nuclear Information System (INIS)

    Ozgener, B.; Ozgener, H.A.

    2005-01-01

    A multiregion, multigroup collision probability method with white boundary condition is developed for thermalization calculations of light water moderated reactors. Hydrogen scatterings are treated by Nelkin's kernel while scatterings from other nuclei are assumed to obey the free-gas scattering kernel. The isotropic return (white) boundary condition is applied directly by using the appropriate collision probabilities. Comparisons with alternate numerical methods show the validity of the present formulation. Comparisons with some experimental results indicate that the present formulation is capable of calculating disadvantage factors which are closer to the experimental results than alternative methods

  14. Environmentally-Assisted Cracking of Low-Alloy Reactor Pressure Vessel Steels under Boiling Water Reactor Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Seifert, H.P.; Ritter, S

    2002-02-01

    The present report summarizes the experimental work performed by PSI on the environmentally-assisted cracking (EAC) of low-alloy steels (LAS) in the frame of the RIKORR-project during the period from January 2000 to August 2001. Within this project, the EAC crack growth behaviour of different low-alloy reactor pressure vessel (RPV) steels, weld filler and weld heat-affected zone materials is investigated under simulated transient and steady-state BWR/NWC power operation conditions. The EAC crack growth behaviour of different low-alloy RPV steels was characterized by slow rising load (SRL) / low-frequency corrosion fatigue (LFCF) and constant load tests with pre-cracked fracture mechanics specimens in oxygenated high-temperature water at temperatures of either 288, 250, 200 or 150 C. These tests revealed the following important interim results: Under low-flow and highly oxidizing (ECP >= 100 mV SHE) conditions, the ASME XI 'wet' reference fatigue crack growth curve could be significantly exceeded by cyclic fatigue loading at low frequencies (<0.001 Hz), at high and low load-ratios R, and by ripple loading near to DKth fatigue thresholds. The BWR VIP 60 SCC disposition lines may be significantly or slightly exceeded (even in steels with a low sulphur content) in the case of small load fluctuations at high load ratios (ripple loading) or at intermediate temperatures (200 -250 C) in RPV materials, which show a distinct susceptibility to dynamic strain ageing (DSA). (author)

  15. Study On Safety Analysis Of PWR Reactor Core In Transient And Severe Accident Conditions

    International Nuclear Information System (INIS)

    Le Dai Dien; Hoang Minh Giang; Nguyen Thi Thanh Thuy; Nguyen Thi Tu Oanh; Le Thi Thu; Pham Tuan Nam; Tran Van Trung; Le Van Hong; Vo Thi Huong

    2014-01-01

    The cooperation research project on the Study on Safety Analysis of PWR Reactor Core in Transient and Severe Accident Conditions between Institute for Nuclear Science and Technology (INST), VINATOM and Korean Atomic Energy Research Institute (KAERI), Korea has been setup to strengthen the capability of researches in nuclear safety not only in mastering the methods and computer codes, but also in qualifying of young researchers in the field of nuclear safety analysis. Through the studies on the using of thermal hydraulics computer codes like RELAP5, COBRA, FLUENT and CFX the thermal hydraulics research group has made progress in the research including problems for safety analysis of APR1400 nuclear reactor, PIRT methodologies and sub-channel analysis. The study of severe accidents has been started by using MELCOR in collaboration with KAERI experts and the training on the fundamental phenomena occurred in postulated severe accident. For Vietnam side, VVER-1000 nuclear reactor is also intensively studied. The design of core catcher, reactor containment and severe accident management are the main tasks concerning VVER technology. The research results are presented in the 9 th National Conference on Mechanics, Ha Noi, December 8-9, 2012, the 10 th National Conference on Nuclear Science and Technology, Vung Tau, August 14-15, 2013, as well as published in the journal of Nuclear Science and Technology, Vietnam Nuclear Society and other journals. The skills and experience from using computer codes like RELAP5, MELCOR, ANSYS and COBRA in nuclear safety analysis are improved with the nuclear reactors APR1400, Westinghouse 4 loop PWR and especially the VVER-1000 chosen for the specific studies. During cooperation research project, man power and capability of Nuclear Safety center of INST have been strengthen. Three masters were graduated, 2 researchers are engaging in Ph.D course at Hanoi University of Science and Technology and University of Science and Technology, Korea

  16. Thermal Hydraulics Analysis for the 3MW TRIGA MARK-II Research Reactor Under Transient Condition

    International Nuclear Information System (INIS)

    Huda, M.Q.; Bhuiyan, S.I.; Mondal, M.A.W.

    1996-12-01

    Some important thermal hydraulic parameters of the 3 MW TRIGA MARK-II research reactor operating under transient condition were investigated using two computer codes PULTRI and TEMPUL. Major transient parameters, such as, peak power and prompt energy released after pulse, maximum fuel and coolant temperature, surface heat flux, time and radial distribution of temperature within fuel element after pulse, fuel, fuel-cladding gap width variation, etc. were computer and compared with the experimental and operational values as reported in the safety Analysis Report (SAR). It was observed that pulsing of the reactor inserting an excess reactivity of $2.00 shoots the reactor power level to 854.353 MW compared to an experimental value of 852 MW; the maximum fuel temperature corresponding to this peak power was found to be 846.76 o C which is much less than the limiting maximum value of fuel temperature of 1150 0 C as reported in SAR. During a pulse if the film boiling occurs for a peak adiabatic fuel temperature of 1000 o C, the calculated outer cladding wall temperature was observed to be 702.39 0 C compared to a value of 760 o C reported in SAR under the same condition. The investigated other results were also found to be in good agreement with the values reported in the SAR. 16 refs., 22 figs. (author)

  17. Power distributions in a nuclear reactor core under various operating conditions and their safety implications

    International Nuclear Information System (INIS)

    Yang, C.Y.; Cho, N.Z.

    1988-01-01

    A nonlinear reactor model has been developed considering typical feedback effects such as the moderator and fuel temperatures, xenon, and soluble boron. The resulting equation belongs to a class of nonlinear boundary value problems, and it was numerically shown using the arc-length continuation method that the nonlinear feedback equation may have multiple steady-state solutions due to existence of branch points for a range of parameters and feedback coefficients which are determined by various design and operating conditions. Solution diagrams of three distinct types were obtained depending on the values of feedback constants and stability was characterized for each solution. The starting point of instability (first limit point) in the solution diagram decreases as the temperature coefficients become large or the boron concentration increases, which means decrease of stability. When solution diagrams which are divided into stable and unstable regions are obtained for a wide range of feedback coefficients, we can choose the range of the reactor design and operating conditions such that the reactor does not encounter unstable situations

  18. Study on natural circulation flow under reactor cavity flooding condition in advanced PWRs

    International Nuclear Information System (INIS)

    Tao Jun; Yang Jiang; Cao Jianhua; Lu Xianghui; Guo Dingqing

    2015-01-01

    Cavity flooding is an important severe accident management measure for the in-vessel retention of a degraded core by external reactor vessel cooling in advanced PWRs. A code simulation study on the natural circulation flow in the gap between the reactor vessel wall and insulation material under cavity flooding condition is performed by using a detailed mechanistic thermal-hydraulic code package RELAP 5. By simulating of an experiment carried out for studying the natural circulation flow for APR1400 shows that the code is applicable for analyzing the circulation flow under this condition. The analysis results show that heat removal capacity of the natural circulation flow in AP1000 is sufficient to prevent thermal failure of the reactor vessel under bounding heat load. Several conclusions can be drawn from the sensitivity analysis. Larger coolant inlet area induced larger natural circulation flow rate. The outlet should be large enough and should not be submerged by the cavity water to vent the steam-water mixture. In the implementation of cavity flooding, the flooding water level should be high enough to provide sufficient natural circulation driven force. (authors)

  19. Formation of metabolites during biodegradation of linear alkylbenzene sulfonate in an upflow anaerobic sludge bed reactor under thermophilic conditions

    DEFF Research Database (Denmark)

    Mogensen, Anders Skibsted; Ahring, Birgitte Kiær

    2002-01-01

    Biodegradation of linear alkylbenzene sulfonate (LAS) was shown in an upflow anaerobic sludge blanket reactor under thermophilic conditions. The reactor was inoculated with granular biomass and fed with a synthetic medium and 3 mumol/L of a mixture of LAS with alkylchain length of 10 to 13 carbon...

  20. Modifications in the operational conditions of the IEA-R1 reactor under continuous 48 hours operation

    International Nuclear Information System (INIS)

    Moreira, Joao Manoel Losada; Frajndlich, Roberto

    1995-01-01

    This work shows the required changes in the IEA-R1 reactor for operation at 2 Mw, 48 hours continuously. The principal technical change regards the operating conditions of the reactor, namely, the required excess reactivity which now will amount to 4800 pcm in order to compensate the Xe poisoning at equilibrium at 2 Mw. (author). 6 refs, 1 fig, 1 tab

  1. Buckling resistance calculation of Guide Thimbles for the mechanical design of fuel assembly type PWR under normal reactor operating conditions

    International Nuclear Information System (INIS)

    Cruz, C.B.L.

    1990-01-01

    The calculations demonstrate the fulfillment of one of the mechanical design criteria for the Fuel Assembly Structure under normal reactor operating conditions. The calculations of stresses in the Guide Thimbles are performed with the aid of the program ANSYS. This paper contains program parameters and modelling of a typical Fuel Assembly for a Reactor similar to ANGRA II. (author)

  2. Bioremediation of endosulfan contaminated soil and water-Optimization of operating conditions in laboratory scale reactors

    International Nuclear Information System (INIS)

    Kumar, Mathava; Philip, Ligy

    2006-01-01

    A mixed bacterial culture consisted of Staphylococcus sp., Bacillus circulans-I and -II has been enriched from contaminated soil collected from the vicinity of an endosulfan processing industry. The degradation of endosulfan by mixed bacterial culture was studied in aerobic and facultative anaerobic conditions via batch experiments with an initial endosulfan concentration of 50 mg/L. After 3 weeks of incubation, mixed bacterial culture was able to degrade 71.58 ± 0.2% and 75.88 ± 0.2% of endosulfan in aerobic and facultative anaerobic conditions, respectively. The addition of external carbon (dextrose) increased the endosulfan degradation in both the conditions. The optimal dextrose concentration and inoculum size was estimated as 1 g/L and 75 mg/L, respectively. The pH of the system has significant effect on endosulfan degradation. The degradation of alpha endosulfan was more compared to beta endosulfan in all the experiments. Endosulfan biodegradation in soil was evaluated by miniature and bench scale soil reactors. The soils used for the biodegradation experiments were identified as clayey soil (CL, lean clay with sand), red soil (GM, silty gravel with sand), sandy soil (SM, silty sand with gravel) and composted soil (PT, peat) as per ASTM (American society for testing and materials) standards. Endosulfan degradation efficiency in miniature soil reactors were in the order of sandy soil followed by red soil, composted soil and clayey soil in both aerobic and anaerobic conditions. In bench scale soil reactors, endosulfan degradation was observed more in the bottom layers. After 4 weeks, maximum endosulfan degradation efficiency of 95.48 ± 0.17% was observed in red soil reactor where as in composted soil-I (moisture 38 ± 1%) and composted soil-II (moisture 45 ± 1%) it was 96.03 ± 0.23% and 94.84 ± 0.19%, respectively. The high moisture content in compost soil reactor-II increased the endosulfan concentration in the leachate. Known intermediate metabolites of

  3. Analysis of effects of calandria tube uncovery under severe accident conditions in CANDU reactors

    International Nuclear Information System (INIS)

    Rogers, J.T.; Currie, T.C.; Atkinson, J.C.; Dick, R.

    1983-01-01

    A study is being undertaken for the Atomic Energy Control Board to assess the thermal and hydraulic behaviour of CANDU reactor cores under accident conditions more severe than those normally considered in the licensing process. In this paper, we consider the effects on a coolant channel of the uncovery of a calandria tube by moderator boil-off following a LOCA in a Bruce reactor unit in which emergency cooling is ineffective and the moderator heat sink is impaired by the failure of the moderator cooling system. Calandria tube uncovery and its immediate consequences, as described here, constitute only one part of the entire accident sequence. Other aspects of this sequence as well as results of the analysis of the other accident sequences studied will be described in the final report on the project and in later papers

  4. Effect of inlet conditions on the performance of a palladium membrane reactor

    International Nuclear Information System (INIS)

    Birdsell, S.A.; Willms, R.S.; Arzu, P.; Costello, A.

    1997-10-01

    Palladium membrane reactors (PMR) will be used to remove tritium and other hydrogen isotopes from impurities, such as tritiated methane and tritiated water, in the exhaust of the International Thermonuclear Experimental Reactor. In addition to fusion-fuel processing, the PMR system can be used to recover tritium from tritiated waste water. This paper investigates the effect of inlet conditions on the performance of a PMR. A set of experiments were run to determine, independently, the effect of inlet compositions and residence time on performance. Also, the experiments were designed to determine if the injected form of hydrogen (CH 4 or H 2 O) effects performance. Results show that the PMR operates at optimal hydrogen recovery with a broad range of inlet compositions and performance is shown to increase with increased residence time. PMR performance is shown to be independent of whether hydrogen is injected in the form of CH 4 or H 2 O

  5. Stability criteria and critical runway conditions of propylene glycol manufacture in a continuous stirred tank reactor

    Directory of Open Access Journals (Sweden)

    Miguel Ángel Gómez

    2015-05-01

    Full Text Available Here, a new method for the analysis of the steady state and the safety operational conditions of the hydrolysis of propylene oxide with excess of water, in a Continuous Stirred Tank Reactor (CSTR, was developed. For industrial operational typical values, at first, the generated and removed heat balances were examined. Next, the effect of coolant fluid temperature in the critical ignition and extinction temperatures (TCI and TCE, respectively was analyzed. The influence of the heat exchange parameter (hS on coolant and critical temperatures was also studied. Finally, the steady state operation areas were defined. The existence of multiple stable states was recognized when the heat exchange parameter was in the range 6.636 < hS kJ/(min.K < 11.125. Unstable operation area was located between the TCI and TCE values, restricting the reactor operation area to the low stable temperatures.

  6. Representation of boundary conditions in thermal reactor global analysis by diffusion theory employing finite difference approximation

    International Nuclear Information System (INIS)

    Paul, O.P.K.

    1978-01-01

    An approach to simulate the flux vanishing boundary condition in solving the two group coupled neutron diffusion equations in three dimensions (x, y, z) employed to calculate the flux distribution and keff of the reactor is summarised. This is of particular interest when the flux vanishing boundary in x, y, z directions is not an integral multiple of the mesh spacings in these directions. The method assumes the flux to be negative, hypothetically at the mesh points lying outside the boundary and thus the finite difference formalism for Laplacian operator, taking into account six neighbours of a mesh point in a square mesh arrangement, is expressed in a general form so as to account for the boundary mesh points of the system. This approach has been incorporated in a three dimensional diffusion code similar to TAPPS23 and has been used for IRT-2000 reactor and the results are quite satisfactory. (author)

  7. Role of Passive Safety Features in Prevention And Mitigation of Severe Plant Conditions in Indian Advanced Heavy Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jain, Vikas; Nayak, A.; Dhiman, M.; Kulkarni, P. P.; Vijayan, P. K.; Vaze, K. K. [Bhabha Atomic Research Centre, Mumbai (India)

    2013-10-15

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

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

    Directory of Open Access Journals (Sweden)

    VIKAS JAIN

    2013-10-01

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

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  10. Leaching Studies on ACR-1000{sup R} Fuel Under Reactor Operating Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Sunder, S. [Atomic Energy of Canada Limited, Fuel and Fuel Channel Safety Branch, Chalk River, Ontario, K0J 1J0 (Canada)

    2009-06-15

    ACR-1000{sup R} is the latest nuclear power reactor being developed by AECL. The ACR-1000 fuel uses a modified CANFLEX{sup R} fuel bundle that contains low-enriched uranium and pellets of burnable neutron absorbers (BNA) in a central element. Dysprosium and gadolinium are used as the burnable neutron absorbers and are present as oxides in a 'fully-stabilized' zirconia matrix. The BNA material in the centre element is designed to limit the coolant void reactivity of the reactor core during postulated loss-of-coolant accidents. As part of the ACR-1000 fuel development, the stability of the BNA material under conditions associated with defects of the Zircaloy sheathing of the BNA central element has been investigated. The results of these tests can be used to demonstrate the phase stability and leaching behaviour of the ACR-1000 fuel under reactor operating conditions. The samples were disks, about 3-4 mm thick, obtained from BNA pellets and Candu fuel (natural uranium UO{sub 2}) pellets (the UO{sub 2} measurements provide a reference point). Leaching tests were carried out in light water at 325 deg. C, above the maximum coolant temperature in an ACR-1000 fuel channel during normal operating conditions (319 deg. C). This temperature also bounds the maximum operating temperature for the current Candu reactors (311 deg. C). The initial pH of the solution (measured at room temperature) used in the leaching tests was 10.3. The leach rates were determined by monitoring the amount of metals leached into solutions. Leaching tests were also carried out with BNA pellet samples in the presence of Zr-2.5%Nb pressure tube coupons to determine the effects, if any, of the presence of pressure tube material on leach rates. Other leaching tests with BNA pellet samples and UO{sub 2} pellets were conducted at 80 deg. C to study the effects of temperature on the leach rates. The temperature of 80 deg. C was selected as representative of typical shutdown temperatures

  11. Current activities on improving storage conditions of the research reactor RA spent fuel - Part II

    International Nuclear Information System (INIS)

    Matausek, M.V.; Kopecni, M.; Vukadin, Z.; Plecas, I.; Pavlovic, R.; Sotic, O.; Marinkovic, N.

    1998-01-01

    To minimize further corrosion and preserve integrity of aluminum barrels and the stainless steel channel-type containers that were found to contain leaking spent fuel, actions to improve conditions in the existing spent fuel storage pool at the RA research reactor were initiated. Technology was elaborated and equipment was produced and applied for removal of sludge and other debris from the bottom of the pool, filtration of the pool water, sludge conditioning in cement matrix and disposal at the low and medium waste repository at VINCA site. More sophisticated operations are to be performed together with foreign experts. Safety measures and precautions were determined. Subcriticality was proved under normal and/or possible abnormal conditions. (author)

  12. Study of behavior of cermet fuel elements on IGR reactor under RIA type accident condition

    International Nuclear Information System (INIS)

    Vasil'ev, Yu.S.; Vurim, A.D.; Koltyshev, S.M.; Pakhnits, V.A.; Tukhvatulin, Sh.T.; Popov, V.V.; Ryzhkov, A.N.

    1996-01-01

    In 1993 December in IGR reactor of Inst. of Atomic Energy of National Nuclear Center of Republic of Kazakstan the second batch of in-pile testing of perspective cermet fuel elements under the condition, simulating RIA type accident was conducted. In the second batch of testing during eight start-ups 10 cermet fuel elements were examined. Among which 8 of monolith type and 2 fuel elements with false jacket beside cladding (FJF), as well as, 6 standard fuel elements of WWER-1000 type reactor with dioxide fuel were tested. 2 fuel elements - cermet and standard were placed into capsule filled with water. To measure energy release for the each start-up two fission monitor and inside core control gauge were placed. In all the start-ups operation mode of IGR was neutron pulse. Power of fuel element kept changing from 151 to 336 k W; energy release was 38-93 kJ/gr m 235 U; maximum temperature of cermet fuel was 1943-2173 K, of dioxide fuel - 1923-2843 K. The testing has demonstrated that operability of cermet fuel elements under reactivity accident condition with pulse width of 0,2 s is, at least, not less that operability of dioxide fuel elements, through advantages of cermet fuel under these conditions are revealed to the least extent

  13. Reactor condition monitoring and singularity detection via wavelet and use of entropy in Monte Carlo calculation

    International Nuclear Information System (INIS)

    Kim, Ok Joo

    2007-02-01

    Wavelet theory was applied to detect the singularity in reactor power signal. Compared to Fourier transform, wavelet transform has localization properties in space and frequency. Therefore, by wavelet transform after de-noising, singular points can be found easily. To demonstrate this, we generated reactor power signals using a HANARO (a Korean multi-purpose research reactor) dynamics model consisting of 39 nonlinear differential equations and Gaussian noise. We applied wavelet transform decomposition and de-noising procedures to these signals. It was effective to detect the singular events such as sudden reactivity change and abrupt intrinsic property changes. Thus this method could be profitably utilized in a real-time system for automatic event recognition (e.g., reactor condition monitoring). In addition, using the wavelet de-noising concept, variance reduction of Monte Carlo result was tried. To get correct solution in Monte Carlo calculation, small uncertainty is required and it is quite time-consuming on a computer. Instead of long-time calculation in the Monte Carlo code (MCNP), wavelet de-noising can be performed to get small uncertainties. We applied this idea to MCNP results of k eff and fission source. Variance was reduced somewhat while the average value is kept constant. In MCNP criticality calculation, initial guess for the fission distribution is used and it could give contamination to solution. To avoid this situation, sufficient number of initial generations should be discarded, and they are called inactive cycles. Convergence check can give guildeline to determine when we should start the active cycles. Various entropy functions are tried to check the convergence of fission distribution. Some entropy functions reflect the convergence behavior of fission distribution well. Entropy could be a powerful method to determine inactive/active cycles in MCNP calculation

  14. Analysis of radionuclide dispersion at normal condition for AEC 1000 MW reactor power

    International Nuclear Information System (INIS)

    Sri Kuntjoro

    2010-01-01

    Analysis for radionuclide dispersion for the Atomic Energy Agency (AEC) 3,568 MWth Power Reactor, equal to the 1,000 MWe at normal condition has been done. Analysis was done for two piles that is separated by 500 m distance and angle of 90° one to other. Initial pace in doing the analysis is to determine reactors source term using ORIGEN2 and EMERALD NORMAL. computer code program. ORIGEN2 applied to determine radionuclide inventory emerged in the reactor. Hereinafter, by using Emerald Normal Computer code is calculated source term reaching the reactor stack. To analyze dose received by population is done by using PC-CREAM computer code. Calculation done for one and two PLTN attached in site candidate of plants. The result showed is that the highest radionuclide release for one PLTN is at 1 km distance and to 9 th zone toward ( 19.25° ) and for two PLTN is at 1 km distance and to 10 th zone toward (21.75° ). Radionuclide which up to population through two pathways that are foodstuff and inhalation. To foodstuff comes from radionuclide I 131 , and the biggest passed from milk product with 53.40 % for one and also two PLTN For inhalation pathway the highest radionuclide contribution come from Kr 85m is about 53.80 %. The highest total dose received by population is at 1 Km distance received by baby that is 4.10 µSi and 11.26 µSi for one and two PLTN respectively. Those result are very small compared to the maximum permission dose to population issued by regulatory body that is equal to 1 mSi. (author)

  15. Features, present condition of development and future scope on the high temperature gas reactor as an innovative one

    International Nuclear Information System (INIS)

    Shiozawa, Shusaku

    2001-01-01

    The high temperature gas reactor has some features without previous reactors such as high temperature capable of taking-out, high specific safety, feasibility adaptable to versatile fuel cycle, and so on. Then, it is expected to be an innovative reactor to contribute to diversification of energy supply and expansion of energy application field. In Japan, under the HTTR (high temperature engineering test reactor) plan, construction of HTTR, which is the first high temperature gas reactor in Japan, was finished and its output upgrading test has been promoted. And, on the HTTR plan, together with promotion of full power operation, reactor performance tests, safety proof test, and so on, it is planned to carry out study on application of the high temperature heat such as hydrogen production and so on to aim to practise establishment and upgrading of technologies on high temperature gas reactor in Japan. Here were introduced features and present condition of development of the high temperature gas reactor as an innovative type reactor and described role and future scope in Japan. (G.K.)

  16. Characterization, treatment and conditioning of radioactive graphite from decommissioning of nuclear reactors

    International Nuclear Information System (INIS)

    2006-09-01

    Graphite has been used as a moderator and reflector of neutrons in more than 100 nuclear power plants and in many research and plutonium-production reactors. It is used primarily as a neutron reflector or neutron moderator, although graphite is also used for other features of reactor cores, such as fuel sleeves. Many of the graphite-moderated reactors are now quite old, with some already shutdown. Therefore radioactive graphite dismantling and the management of radioactive graphite waste are becoming an increasingly important issue for a number of IAEA Member States. Worldwide, there are more than 230 000 tonnes of radioactive graphite which will eventually need to be managed as radioactive waste. Proper management of radioactive graphite waste requires complex planning and the implementation of several interrelated operations. There are two basic options for graphite waste management: (1) packaging of non-conditioned graphite waste with subsequent direct disposal of the waste packages, and (2) conditioning of graphite waste (principally either by incineration or calcination) with separate disposal of any waste products produced, such as incinerator ash. In both cases, the specific properties of graphite - such as Wigner energy, graphite dust explosibility, and radioactive gases released from waste graphite - have a potential impact on the safety of radioactive graphite waste management and need to be carefully considered. Radioactive graphite waste management is not specifically addressed in IAEA publications. Only general and limited information is available in publications dealing with decommissioning of nuclear reactors. This report provides a comprehensive discussion of radioactive graphite waste characterization, handling, conditioning and disposal throughout the operating and decommissioning life cycle. The first draft report was prepared at a meeting on 23-27 February 1998. A technical meeting (TM) was held in October 1999 in coincidence with the Seminar on

  17. Coolability of degraded core under reflooding conditions in Nordic boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lindholm, I; Pekkarinen, E [VTT Energy, Espoo (Finland); Nilsson, L [Studsvik EcoSafe AB, Nykoeping (Sweden); Sjoevall, H [Teollisuuden Voima Oy, Olkiluoto (Finland)

    1995-09-01

    Present work is part of the first phase of subproject RAK-2.1 of the new Nordic Co-operative Reactor Safety Program, NKS. The first phase comprises reflooding calculations for the boiling water reactors (BWRs) TVO I/II in Finland and Forsmark 3 in Sweden, as a continuation of earlier severe accident analyses which were made in the SIK-2 project. The objective of the core reflooding studies is to evaluate when and how the core is still coolable with water and what are the probable consequences of water cooling. In the following phase of the RAK-2.1 project, recriticality studies will be performed. Conditions for recriticality might occur if control rods have melted away with the fuel rods intact in a shape that critical conditions can be created in reflooding with insufficiently borated water. Core coolability was investigated for two reference plants, TVO I/II and Forsmark 3. The selected accident cases were anticipated station blackout with or without successful depressurization of reactor coolant system (RCS). The effects of the recovery of emergency core cooling (ECC) were studied by varying the starting time of core reflooding. The start of ECC systems were assigned to reaching a maximum cladding temperature: 1400 K, 1600 K, 1800 K and 2000 K in the core. Cases with coolant injection through the downcomer were studied for TVO I/II and both downcomer injection and core top spray were investigated for Forsmark 3. Calculations with three different computer codes: MAAP 4, MELCOR 1.8.3 and SCDA/RELAP5/MOD 3.1 for the basis for the presented reflooding studies. Presently, and experimental programme on core reflooding phenomena has been started in Kernforschungszentrum Karlsruhe in QUENCH test facility. (EG) 17 refs.

  18. WATER-GAS SHIFT KINETICS OVER IRON OXIDE CATALYSTS AT MEMBRANE REACTOR CONDITIONS; A

    International Nuclear Information System (INIS)

    Carl R.F. Lund

    2001-01-01

    This report covers the second year of a project investigating water-gas shift catalysts for use in membrane reactors. It has been established that a simple iron high temperature shift catalyst becomes ineffective in a membrane reactor because the reaction rate is severely inhibited by the build-up of the product CO(sub 2). During the past year, an improved microkinetic model for water-gas shift over iron oxide was developed. Its principal advantage over prior models is that it displays the correct asymptotic behavior at all temperatures and pressures as the composition approaches equilibrium. This model has been used to explore whether it might be possible to improve the performance of iron high temperature shift catalysts under conditions of high CO(sub 2) partial pressure. The model predicts that weakening the surface oxygen bond strength by less than 5% should lead to higher catalytic activity as well as resistance to rate inhibition at higher CO(sub 2) partial pressures. Two promoted iron high temperature shift catalysts were studied. Ceria and copper were each studied as promoters since there were indications in the literature that they might weaken the surface oxygen bond strength. Ceria was found to be ineffective as a promoter, but preliminary results with copper promoted FeCr high temperature shift catalyst show it to be much more resistant to rate inhibition by high levels of CO(sub 2). Finally, the performance of sulfided CoMo/Al(sub 2)O(sub 3) catalysts under conditions of high CO(sub 2) partial pressure was simulated using an available microkinetic model for water-gas shift over this catalyst. The model suggests that this catalyst might be quite effective in a medium temperature water-gas shift membrane reactor, provided that the membrane was resistant to the H(sub 2)S that is required in the feed

  19. Stabilization of burn conditions in a thermonuclear reactor using artificial neural networks

    Science.gov (United States)

    Vitela, Javier E.; Martinell, Julio J.

    1998-02-01

    In this work we develop an artificial neural network (ANN) for the feedback stabilization of a thermonuclear reactor at nearly ignited burn conditions. A volume-averaged zero-dimensional nonlinear model is used to represent the time evolution of the electron density, the relative density of alpha particles and the temperature of the plasma, where a particular scaling law for the energy confinement time previously used by other authors, was adopted. The control actions include the concurrent modulation of the D-T refuelling rate, the injection of a neutral He-4 beam and an auxiliary heating power modulation, which are constrained to take values within a maximum and minimum levels. For this purpose a feedforward multilayer artificial neural network with sigmoidal activation function is trained using a back-propagation through-time technique. Numerical examples are used to illustrate the behaviour of the resulting ANN-dynamical system configuration. It is concluded that the resulting ANN can successfully stabilize the nonlinear model of the thermonuclear reactor at nearly ignited conditions for temperature and density departures significantly far from their nominal operating values. The NN-dynamical system configuration is shown to be robust with respect to the thermalization time of the alpha particles for perturbations within the region used to train the NN.

  20. Stabilization of burn conditions in a thermonuclear reactor using artificial neural networks

    International Nuclear Information System (INIS)

    Vitela, J.E.; Martinell, J.J.

    1998-01-01

    In this work we develop an artificial neural network (ANN) for the feedback stabilization of a thermonuclear reactor at nearly ignited burn conditions. A volume-averaged zero-dimensional nonlinear model is used to represent the time evolution of the electron density, the relative density of alpha particles and the temperature of the plasma, where a particular scaling law for the energy confinement time previously used by other authors, was adopted. The control actions include the concurrent modulation of the D-T refuelling rate, the injection of a neutral He-4 beam and an auxiliary heating power modulation, which are constrained to take values within a maximum and minimum levels. For this purpose a feedforward multilayer artificial neural network with sigmoidal activation function is trained using a back-propagation through-time technique. Numerical examples are used to illustrate the behaviour of the resulting ANN-dynamical system configuration. It is concluded that the resulting ANN can successfully stabilize the nonlinear model of the thermonuclear reactor at nearly ignited conditions for temperature and density departures significantly far from their nominal operating values. The NN-dynamical system configuration is shown to be robust with respect to the thermalization time of the alpha particles for perturbations within the region used to train the NN. (author)

  1. Decay heat removal and transient analysis in accidental conditions in the EFIT reactor

    International Nuclear Information System (INIS)

    Bandini, G.; Meloni, P.; Polidori, M.; Casamirra, M.; Castiglia, F.; Giardina, M.

    2007-01-01

    The development of a conceptual design of an industrial scale transmutation facility (EFIT) of several 100 MW thermal power based on Accelerator Driven System (ADS) is addressed in the frame of the European EUROTRANS Integral Project. In normal operation, the core power of EFIT reactor is removed through steam generators by four secondary loops fed by water. A safety-related Decay Heat Removal (DHR) system provided with four independent inherently safe loops is installed in the primary vessel to remove the decay heat by natural convection circulation under accidental conditions which lead to the Loss of Heat Sink (LOHS). In order to confirm the adequacy of the adopted solution for decay heat removal in accidental conditions, some multi-D analyses have been carried out with the SIMMER-III code. The results of the SIMMER-III code have been then used to support the RELAP5 1-D representation of the natural circulation flow paths in the reactor vessel. Finally, the thermal-hydraulic RELAP5 code has been employed for the analysis of LOHS accidental scenarios. (author)

  2. Decay Heat Removal and Transient Analysis in Accidental Conditions in the EFIT Reactor

    Directory of Open Access Journals (Sweden)

    Giacomino Bandini

    2008-01-01

    Full Text Available The development of a conceptual design of an industrial-scale transmutation facility (EFIT of several 100 MW thermal power based on accelerator-driven system (ADS is addressed in the frame of the European EUROTRANS Integral Project. In normal operation, the core power of EFIT reactor is removed through steam generators by four secondary loops fed by water. A safety-related decay heat removal (DHR system provided with four independent inherently safe loops is installed in the primary vessel to remove the decay heat by natural convection circulation under accidental conditions which are caused by a loss-of-heat sink (LOHS. In order to confirm the adequacy of the adopted solution for decay heat removal in accidental conditions, some multi-D analyses have been carried out with the SIMMER-III code. The results of the SIMMER-III code have been then used to support the RELAP5 1D representation of the natural circulation flow paths in the reactor vessel. Finally, the thermal-hydraulic RELAP5 code has been employed for the analysis of LOHS accidental scenarios.

  3. Technical Needs for Enhancing Risk Monitors with Equipment Condition Assessment for Advanced Small Modular Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Coble, Jamie B.; Coles, Garill A.; Ramuhalli, Pradeep; Meyer, Ryan M.; Berglin, Eric J.; Wootan, David W.; Mitchell, Mark R.

    2013-04-04

    Advanced small modular reactors (aSMRs) can provide the United States with a safe, sustainable, and carbon-neutral energy source. The controllable day-to-day costs of aSMRs are expected to be dominated by operation and maintenance costs. Health and condition assessment coupled with online risk monitors can potentially enhance affordability of aSMRs through optimized operational planning and maintenance scheduling. Currently deployed risk monitors are an extension of probabilistic risk assessment (PRA). For complex engineered systems like nuclear power plants, PRA systematically combines event likelihoods and the probability of failure (POF) of key components, so that when combined with the magnitude of possible adverse consequences to determine risk. Traditional PRA uses population-based POF information to estimate the average plant risk over time. Currently, most nuclear power plants have a PRA that reflects the as-operated, as-modified plant; this model is updated periodically, typically once a year. Risk monitors expand on living PRA by incorporating changes in the day-by-day plant operation and configuration (e.g., changes in equipment availability, operating regime, environmental conditions). However, population-based POF (or population- and time-based POF) is still used to populate fault trees. Health monitoring techniques can be used to establish condition indicators and monitoring capabilities that indicate the component-specific POF at a desired point in time (or over a desired period), which can then be incorporated in the risk monitor to provide a more accurate estimate of the plant risk in different configurations. This is particularly important for active systems, structures, and components (SSCs) proposed for use in aSMR designs. These SSCs may differ significantly from those used in the operating fleet of light-water reactors (or even in LWR-based SMR designs). Additionally, the operating characteristics of aSMRs can present significantly different

  4. Scram characteristics of the control rods of a pressurized water reactor under seismic conditions

    International Nuclear Information System (INIS)

    Fujita, Katsuhisa; Shinohara, Yoshikazu; Nakatogawa, Tetsuto; Nanbu, Kiyoshi; Nomura, Tomonori.

    1987-01-01

    Control rod drop verification experiments of a pressurized water reactor under seismic conditions are performed to confirm the insertion function of control rods into a core. To evaluate these tests, computer simulations are performed. A fuel assembly, control rods, guide tube and other associated structures are immersed in a water tank, and shaken by four hydraulic shakers. The scram time of control rods under seismic conditions was measured, and confirmed to meet the scram function. Moreover, vibrational response characteristics of core structures and dropping behavior of control rods in consideration of collisions are calculated by using a finite difference method. The behavior of the dropping control rods and the scram time obtained by the computer simulation show a very good agreement with the verification experimental results. (author)

  5. Online calibration method for condition monitoring of nuclear reactor instrumentations based on electrical signature analysis

    International Nuclear Information System (INIS)

    Syaiful Bakhri

    2013-01-01

    Electrical signature analysis currently becomes an alternative in condition monitoring in nuclear power plants not only for stationary components such as sensors, measurement and instrumentation channels, and other components but also for dynamic components such as electric motors, pumps, generator or actuators. In order to guarantee the accuracy, the calibration of monitoring system is a necessary which practically is performed offline, under limited schedules and certain tight procedures. This research aims to introduce online calibration technique for electrical signature condition monitoring in order that the accuracy can be maintained continuously which in turn increases the reactor safety as a whole. The research was performed step by stepin detail from the conventional technique, online calibration using baseline information and online calibration using differential gain adjustment. Online calibration based on differential gain adjustment provides better results than other techniques even tough under extreme gain insertion as well as external disturbances such as supply voltages. (author)

  6. Scale analysis of decay heat removal system between HTR-10 and HTR-PM reactors under accidental conditions

    International Nuclear Information System (INIS)

    Roberto, Thiago D.; Alvim, Antonio C.M.

    2017-01-01

    The 10 MW high-temperature gas-cooled test module (HTR-10) is a graphite-moderated and helium-cooled pebble bed reactor prototype that was designed to demonstrate the technical and safety feasibility of this type of reactor project under normal and accidental conditions. In addition, one of the systems responsible for ensuring the safe operation of this type of reactor is the passive decay heat removal system (DHRS), which operates using passive heat removal processes. A demonstration of the heat removal capacity of the DHRS under accidental conditions was analyzed based on a benchmark problem for design-based accidents on an HTR-10, i.e., the pressurized loss of forced cooling (PLOFC) described in technical reports produced by the International Atomic Energy Agency. In fact, the HTR-10 is also a proof-of-concept reactor for the high-temperature gas-cooled reactor pebble-bed module (HTR-PM), which generates approximately 25 times more heat than the HTR-10, with a thermal power of 250 MW, thereby requiring a DHRS with a higher system capacity. Thus, because an HTR-10 is a prototype reactor for an HTR-PM, a scaling analysis of the heat transfer process from the reactor to the DHRS was carried out between the HTR-10 and HTR-PM systems to verify the distortions of scale and the differences between the main dimensionless numbers from the two projects. (author)

  7. Scale analysis of decay heat removal system between HTR-10 and HTR-PM reactors under accidental conditions

    Energy Technology Data Exchange (ETDEWEB)

    Roberto, Thiago D.; Alvim, Antonio C.M. [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear; Lapa, Celso M.F., E-mail: thiagodbtr@gmail.com, E-mail: lapa@ien.gov.br, E-mail: alvim@nuclear.ufrj.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2017-07-01

    The 10 MW high-temperature gas-cooled test module (HTR-10) is a graphite-moderated and helium-cooled pebble bed reactor prototype that was designed to demonstrate the technical and safety feasibility of this type of reactor project under normal and accidental conditions. In addition, one of the systems responsible for ensuring the safe operation of this type of reactor is the passive decay heat removal system (DHRS), which operates using passive heat removal processes. A demonstration of the heat removal capacity of the DHRS under accidental conditions was analyzed based on a benchmark problem for design-based accidents on an HTR-10, i.e., the pressurized loss of forced cooling (PLOFC) described in technical reports produced by the International Atomic Energy Agency. In fact, the HTR-10 is also a proof-of-concept reactor for the high-temperature gas-cooled reactor pebble-bed module (HTR-PM), which generates approximately 25 times more heat than the HTR-10, with a thermal power of 250 MW, thereby requiring a DHRS with a higher system capacity. Thus, because an HTR-10 is a prototype reactor for an HTR-PM, a scaling analysis of the heat transfer process from the reactor to the DHRS was carried out between the HTR-10 and HTR-PM systems to verify the distortions of scale and the differences between the main dimensionless numbers from the two projects. (author)

  8. Reactors

    DEFF Research Database (Denmark)

    Shah, Vivek; Vaz Salles, Marcos António

    2018-01-01

    The requirements for OLTP database systems are becoming ever more demanding. Domains such as finance and computer games increasingly mandate that developers be able to encode complex application logic and control transaction latencies in in-memory databases. At the same time, infrastructure...... engineers in these domains need to experiment with and deploy OLTP database architectures that ensure application scalability and maximize resource utilization in modern machines. In this paper, we propose a relational actor programming model for in-memory databases as a novel, holistic approach towards......-level function calls. In contrast to classic transactional models, however, reactors allow developers to take advantage of intra-transaction parallelism and state encapsulation in their applications to reduce latency and improve locality. Moreover, reactors enable a new degree of flexibility in database...

  9. Whole-Pin Furnace system: An experimental facility for studying irradiated fuel pin behavior under potential reactor accident conditions

    International Nuclear Information System (INIS)

    Liu, Y.Y.; Tsai, H.C.; Donahue, D.A.; Pushis, D.O.; Savoie, F.E.; Holland, J.W.; Wright, A.E.; August, C.; Bailey, J.L.; Patterson, D.R.

    1990-05-01

    The whole-pin furnace system is a new in-cell experimental facility constructed to investigate how irradiated fuel pins may fail under potential reactor accident conditions. Extensive checkouts have demonstrated excellent performance in remote operation, temperature control, pin breach detection, and fission gas handling. The system is currently being used in testing of EBIR-II-irradiated Integral Fast Reactor (IFR) metal fuel pins; future testing will include EBR-II-irradiated mixed-oxide fuel pins. 7 refs., 4 figs

  10. Cracks propagation by stress corrosion cracking in conditions of Boiling Water Reactor (BWR)

    International Nuclear Information System (INIS)

    Fuentes C, P.

    2003-01-01

    This work presents the results of the assays carried out in the Laboratory of Hot Cells of the National Institute of Nuclear Research (ININ) to a type test tube Compact Tension (CT), built in steel austenitic stainless type 304L, simulating those conditions those that it operates a Boiling Water Reactor (BWR), at temperature 288 C and pressure of 8 MPa, to determine the speed to which the cracks spread in this material that is of the one that different components of a reactor are made, among those that it highlights the reactor core vessel. The application of the Hydrogen Chemistry of the Water is presented (HWC) that is one alternative to diminish the corrosion effect low stress in the component, this is gets controlling the quantity of oxygen and of hydrogen as well as the conductivity of the water. The rehearsal is made following the principles of the Mechanics of Elastic Lineal Fracture (LEFM) that considers a crack of defined size with little plastic deformation in the tip of this; the measurement of crack advance is continued with the technique of potential drop of direct current of alternating signal, this is contained inside the standard Astm E-647 (Method of Test Standard for the Measurement of Speed of Growth of Crack by fatigue) that is the one that indicates us as carrying out this test. The specifications that should complete the test tubes that are rehearsed as for their dimensions, it forms, finish and determination of mechanical properties (tenacity to the fracture mainly) they are contained inside the norm Astm E-399, the one which it is also based on the principles of the fracture mechanics. The obtained results were part of a database to be compared with those of other rehearsals under different conditions, Normal Chemistry of the Water (NWC) and it dilutes with high content of O 2 ; to determine the conditions that slow more the phenomena of stress corrosion cracking, as well as the effectiveness of the used chemistry and of the method of

  11. Modeling of Microstructure Evolution in Austenitic Stainless Steels Irradiated Under Light Water Reactor Conditions

    International Nuclear Information System (INIS)

    Gan, J.; Stoller, R.E.; Was, G.S.

    1998-01-01

    A model for the development of microstructure during irradiation in fast reactors has been adapted for light water reactor (LWR) irradiation conditions (275 approximately 325 C, up to approximately10 dpa). The original model was based on the rate-theory, and included descriptions of the evolution of both dislocation loops and cavities. The model was modified by introducing in-cascade interstitial clustering, a term to account for the dose dependence of this clustering, and mobility of interstitial clusters. The purpose of this work was to understand microstructural development under LWR irradiation with a focus on loop nucleation and saturation of loop density. It was demonstrated that in-cascade interstitial clustering dominates loop nucleation in neutron irradiation in LWRS. Furthermore it was shown that the dose dependence of in-cascade interstitial clustering is needed to account for saturation behavior as commonly observed. Both quasi-steady-state (QSS) and non-steady-state (NSS) solutions to the rate equations were obtained. The difference between QSS and NSS treatments in the calculation of defect concentration is reduced at LWR temperature when in-cascade interstitial clustering dominates loop nucleation. The mobility of interstitial clusters was also investigated and its impact on loop density is to reduce the nucleation term. The ultimate goal of this study is to combine the evolution of microstructure and microchemistry together to account for the radiation damage in austenitic stainless steels

  12. Return momentum effect on reactor coolant water level distribution during mid-loop conditions

    International Nuclear Information System (INIS)

    Seo, Jae Kwang; Yang, Jae Young; Park, Goon Cherl

    2001-01-01

    An accurate prediction of the Reactor Coolant System( RCS) water level is of importance in the determination of the allowable operating range to ensure safety during mid-loop operations. However, complex hydrualic phenomena induced by the Shutdown Cooling System (SCS) return momentum causes different water levels from those in the loop where the water level indicators are located. This was apparently observed at the pre-core cold hydro test of the Younggwang Nuclear Unit 3 (YGN 3) in Korea. In this study, in order to analytically understand the effect of the SCS return momentum on the RCS water level distribution, a model using a one-dimensional momentum and energy conservation for cylindrical channel, hydraulic jump in operating cold leg, water level build-up at the Reactor Vessel (RV) inlet nozzle, Bernoulli constant in downcomer region, and total water volume conservation has been developed. The model predicts the RCS water levels at various RCS locations during the mid-loop conditions and the calculation results were compared with the test data. The analysis shows that the hydraulic jump in the operating cold legs, in conjuction with the pressure drop throughout the RCS, is the main cause creating the water level differences at various RCS locations. The prediction results provide good explanations for the test data and show the significant effect of the SCS return momentum on the RCS water levels

  13. Upflow anaerobic sludge reactors for the treatment of combined industrial effluent in subtropical conditions: a comparison between UASB and UASF reactors

    International Nuclear Information System (INIS)

    Yasar, A.; Ahmad, N.; Chaudhry, M.N.; Sarwar, M.; Masood, T.; Yaqub, A.

    2005-01-01

    The performance of anaerobic biological process is heavily process conditions dependent. In this study, an attempt has been made to investigate the influence of process conditions like temperature, sludge age and hydraulic retention time (HRT) on the efficiency of an upflow anaerobic sludge blanket (UASB) reactor and upflow anaerobic sludge filter (UASF) to treat combined industrial wastewater. Reactors were operated at easing ambient temperatures (38, 30, 20 and 14 deg. C) and correspondingly increasing sludge ages (60, 90, 120 and 150 days). At temperature 38 deg. C and sludge age of 60 days, UASF showed better performance than VASE reactor. This mainly due to the enhanced filtration through well-graded sand filter and fairly good biological activity in UASF. At this stage, lack of sludge granulation in VASE reactor resulted in poor biological activity; hence, relatively poor performance. At temperatures 30 and 20 deg. C with sludge ages of 90 and 120 days, respectively, UASB gave better results than UASF. The reason was rapid biological degradation due to proper sludge granulation and favorable temperature. At temperature 14 deg. C, a substantial decrease in the efficiency of UASB reactor as compared to the UASF was evident. Drop in efficiency was because of inhabitation of methanogenic bacteria and liquidation of sludge granules. These factors mounted to a decrease in biological activity, stoppage as production and an increase in total suspended solids (TSS) in the effluent. The influence of hydraulic retention time (ranging between 3-12 hours at an increment of 3 hours) on the removal efficiency of both UASB and UASF was not significant. At favorable temperature (20 to 30 deg. C) and sludge age (90 to 120 days) UASB reactor appeared to be more efficient than UASF.(author)

  14. Analysis of Reactor Pressurized Thermal Shock Conditions Considering Upgrading of Systems Important to Safety

    International Nuclear Information System (INIS)

    Mazurok, A.S; Vyshemirskyij, M.P.

    2015-01-01

    The paper analyzes conditions of pressurized thermal shock on the reactor pressure vessel taking into account upgrading of the emergency core cooling system and primary overpressure protection system. For representative accident scenarios, calculation and comparative analysis was carried out. These scenarios include a small leak from the hot leg and PRZ SV stuck opening with re closure after 3600 sec and 3 SG heat transfer tube rupture. The efficiency of mass flow control by valves on the pump head (emergency core cooling systems) and cold overpressure protection (primary overpressure protection system) was analyzed. The thermal hydraulic model for RELAP5/Mod3.2 code with detailed downcomer (DC) model and changes in accordance with upgrades was used for calculations. Detailed (realistic) modeling of piping and equipment was performed. The upgrades prevent excessive primary cooling and, consequently, help to preserve the RPV integrity and to avoid the formation of a through crack, which can lead to a severe accident

  15. Vaporization of low-volatile fission products under severe CANDU reactor accident conditions

    International Nuclear Information System (INIS)

    Lewis, B.J.; Corse, B.J.; Thompson, W.T.; Kaye, M.H.; Iglesias, F.C.; Elder, P.; Dickson, R.; Liu, Z.

    1997-01-01

    An analytical model has been developed to describe the release behaviour of low-volatile fission products from uranium dioxide fuel under severe reactor accident conditions. The effect of the oxygen potential on the chemical form and volatility of fission products is determined by Gibbs-energy minimization. The release kinetics are calculated according to the rate-controlling step of diffusional transport in the fuel matrix or fission product vaporization from the fuel surface. The effect of fuel volatilization (i.e., matrix stripping) on the release behaviour is also considered. The model has been compared to data from an out-of-pile annealing experiment performed in steam at the Chalk River Laboratories. (author)

  16. Chemical conditions in the repository for low- and intermediate-level reactor waste

    International Nuclear Information System (INIS)

    Snellman, M.; Uotila, H.

    1984-01-01

    The chemical conditions in the proposed repositories for low- and intermediate-level reactor waste at Haestholmen (IVO) and Olkiluoto (TVO) have been discussed with respect to materials introduced into the repository, their possible long-term changes and interaction with groundwater flowing into the repository. The main possible groundwater-rock interactions have been discussed, as well as the role of micro-organisms, organic acids and colloids in the estimation of the barrier integrity. Experimental and theoretical studies have been performed on the basis of the natural groundwater compositions expected at the repository sites. Main emphasis is put on the chemical parameters which might influence the integrity of the different barriers in the repository as well as on the parameters which might effect the release and transport of radionuclides from the repository

  17. MCCI study for Pressurized Heavy Water Reactor under hypothetical accident condition

    International Nuclear Information System (INIS)

    Verma, Vishnu; Mukhopadhyay, Deb; Chatterjee, B.; Singh, R.K.; Vaze, K.K.

    2011-01-01

    In case of severe core damage accident in Pressurized Heavy Water Reactor (PHWR), large amount of molten corium is expected to come out into the calandria vault due to failure of calandria vessel. Molten corium at high temperature is sufficient to decompose and ablate concrete. Such attack could fail CV by basement penetration. Since containment is ultimate barrier for activity release. The Molten Core Concrete Interaction (MCCI) of the resulting pool of debris with the concrete has been identified as an important part of the accident sequence. MCCI Analysis has been carried out for PHWR for a hypothetical accident condition where total core material is considered to be relocated in calandria vault. Concrete ablation rate in vertical and radial direction is evaluated for rectangular geometry using MEDICIS module of ASTEC Code. Amount of gases released during MCCI is also evaluated. (author)

  18. Limits and conditions for continuous operation of WWR-S reactor

    International Nuclear Information System (INIS)

    Pittermann, P.; Listik, E.

    1979-02-01

    The fundamental technological and nuclear characteristics of the WWR-S reactor, safety limits and concepts of technical surveillance with particular attention to radiation safety of staff and of neighbouring population are outlined. The rules are mandatory for the reactor staff and for the users. The material is part of safety documentation for the WWR-S reactor. (author)

  19. A novel ion transport membrane reactor for fundamental investigations of oxygen permeation and oxy-combustion under reactive flow conditions

    KAUST Repository

    Kirchen, Patrick

    2013-01-01

    Ion transport membrane (ITM) reactors present an attractive technology for combined air separation and fuel conversion in applications such as syngas production, oxidative coupling or oxy-combustion, with the promise of lower capital and operating costs, as well higher product selectivities than traditional technologies. The oxygen permeation rate through a given ITM is defined by the membrane temperature and oxygen chemical potential difference across it. Both of these parameters can be strongly influenced by thermochemical reactions occurring in the vicinity of the membrane, though in the literature they are often characterized in terms of the well mixed product stream at the reactor exit. This work presents the development of a novel ITM reactor for the fundamental investigation of the coupling between fuel conversion and oxygen permeation under well defined fluid dynamic and thermodynamic conditions, including provisions for spatially resolved, in-situ investigations. A planar, finite gap stagnation flow reactor with optical and probe access to the reaction zone is used to facilitate in-situ measurements and cross-validation with detailed numerical simulations. Using this novel reactor, baseline measurements are presented to elucidate the impact of the sweep gas fuel (CH4) fraction on the oxygen permeation and fuel conversion. In addition, the difference between well-mixed gas compositions measured at the reactor outlet and those measured in the vicinity of the membrane surface are discussed, demonstrating the unique utility of the reactor. © 2012 The Combustion Institute. Published by Elsevier Inc. All rights reserved.

  20. Thermal stress analysis of reactor containment building considering severe weather condition

    International Nuclear Information System (INIS)

    Lee, Yun; Kim, Yun-Yong; Hyun, Jung-Hwan; Kim, Do-Gyeum

    2014-01-01

    Highlights: • We examine that through-wall crack risk in cold weather is high. • It is predicted that cracking in concrete wall will not happen in hot region. • Cracking due to hydration heat can be controlled by appropriate curing condition. • Temperature differences between inner and outer face is relatively small in hot weather. - Abstract: Prediction of concrete cracking due to hydration heat in mass concrete such as reactor containment building (RCB) in nuclear power plant is a crucial issue in construction site. In this study, the numerical analysis for heat transfer and stress development is performed for the containment wall in RCB by considering the severe weather conditions. Finally, concrete cracking risk in hot and cold weather is discussed based on analysis results. In analyses considering severe weather conditions, it is found that the through-wall cracking risk in cold weather is high due to the abrupt temperature difference between inside concrete and the ambient air in cold region. In hot weather, temperature differences between inner and outer face is relatively small, and accordingly the relevant cracking risk is relatively low in contrast with cold weather

  1. Cleaning and air conditioning device for atmosphere in thermonuclear reactor chamber

    International Nuclear Information System (INIS)

    Ishida, Seiji.

    1993-01-01

    The device of the present invention removes tritium efficiently and attains ventilation and conditioning of a great amount of air flow. That is, there are disposed a humidity separator, a filter, a heater, a catalyst filled layer, a water jetting type humidifying heat insulation cooler and a cooler in this order from an inlet side (upstream) of contaminated room atmospheric gases. The catalyst filled layer, etc. are incorporated integrally into the ventilation air conditioning facility for ventilating air in the chamber of the thermonuclear reactor, to clean a tritium atmosphere at the same time. Accordingly, the device is made compact as a whole. A limit for the air flow rate owing to the use of the conventional catalyst tower and adsorbing tower is eliminated. Then a ventilating air conditioning for a great flow rate can be attained. Tritium is removed by cooling and dehumidification without using any adsorbent. Accordingly, an adsorbing tower is no more necessary and conventional regeneration operation is not required. As a result, space for installation is reduced, the system is simplified and the cost for construction and facility can be reduced. (I.S.)

  2. The effect of code user and boundary conditions on RELAP calculations of MTR research reactor transient scenarios

    Directory of Open Access Journals (Sweden)

    Khedr Ahmed

    2005-01-01

    Full Text Available The safety evaluation of nuclear power and re search reactors is a very important step before their construction and during their operation. This evaluation based on the best estimate calculations requires qualified codes qualified users, and qualified nodalizations. The effect of code users on the RELAP5 results during the analysis of loss of flow transient in MTR research reactors is presented in this pa per. To clarify this effect, two nodalizations for research reactor different in the simulation of the open water surface boundary conditions of the reactor pool have been used. Very different results are obtained with few choices for code users. The core natural circulation flow with the be ginning of core boiling doesn't stop but in creases. The in creasing in the natural circulation flow shifts out the boiling from the core and the clad temperature decreases be low the local saturation temperature.

  3. A fracture mechanics method of evaluating structural integrity of a reactor vessel due to thermal shock effects following LOCA condition

    International Nuclear Information System (INIS)

    Ramani, D.T.

    1977-01-01

    The importance of knowledge of structural integrity of a reactor vessel due to thermal shock effects, is related to safety and operational requirements in assessing the adequacy and flawless functioing of the nuclear power systems. Followig a loss-of-coolant accident (LOCA) condition the integrity of the reactor vessel due to a sudden thermal shock induced by actuation of emergency core cooling system (ECCS), must be maintained to ensure safe and orderly shutdown of the reactor and its components. The paper encompasses criteria underlaying a fracture mechanics method of analysis to evaluate structural integrity of a typical 950 MWe PWR vessel as a result of very drastic changes in thermal and mechanical stress levels in the reactor vessel wall. The main object of this investigation therefore consists in assessing the capability of a PWR vessel to withstand the most critical thermal shock without inpairing its ability to conserve vital coolant owing to probable crack propagation. (Auth.)

  4. Evaluation of permeable and non-permeable tritium in normal condition in a fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Marta, V; Manuel, P J [Instituto de Fusion Nuclear (DENIM)/ETSII, Universidad Politecnica Madrid (UPM) (Spain); Sedano Luis, A [Ministerio de Educacion y Ciencia, Ciemat (Spain)], E-mail: marta@denim.upm.es

    2008-05-15

    The tritium cycle, technologies of process and control of the tritium in the plant will constitute a fraction of the environmental impact of the first generation of DT fusion reactors. The efforts of conceptual development of the tritium cycle are centered in the Internal Regenerator Cycle. The tritium could be recovered from a flow of He gas, or directly from solid breeder. The limits of transfers to the atmosphere are assumed {approx} 1 gr-T/a ({approx}20 Ci/a) (without species distinction). In the case of ITER, for example, we have global demands of control of 5 orders of magnitude have been demonstrated at experimental level. The transfer limits determine the key parameters in tritium Cycle (HT, HTO, as dominant, and T2, T2O as marginal). Presently, the transfer from the cycle to the environment is assumed through the exchange system of the power plant (primary to secondary). That transport is due to the permeation through HT, T2, or leakage to the coolant in the primary system. It is key the chemical optimization in the primary system, that needs to be reanalyzed in terms of radiological impact both for permeable, HT, T2, and non-permeable HTO, T2O. It is necessary considered the pathway of tritium from the reactor to the atmosphere, these processes are modelled adequately. Results of the assessments were early and chronic doses which have been evaluated for the Most Exposed Individual at particular distance bands from the release point. The impact evaluations will be performed with the computational tools (NORMTRI), besides national regulatory models, internationally accepted computer these code for dosimetric evaluations of tritiated effluents in operational conditions.

  5. Reactor

    International Nuclear Information System (INIS)

    Fujibayashi, Toru.

    1976-01-01

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

  6. Propagation of cracks by stress corrosion in conditions of BWR type reactor; Propagacion de grietas por corrosion bajo esfuerzo en condiciones de reactor de agua en ebullicion (BWR)

    Energy Technology Data Exchange (ETDEWEB)

    Merino C, F.J. [ININ, 52045 Estado de Mexico (Mexico); Fuentes C, P. [ITT, Metepec, Estado de Mexico (Mexico)]. E-mail: fjmc@nuclear.inin.mx

    2004-07-01

    In this work, the obtained results when applying the Hydrogen Chemistry to a test tube type Compact Tension (CT), built in austenitic stainless steel 304l, simulating the conditions to those that it operates a Boiling Water Reactor (BWR), temperature 288 C and pressure of 8 MPa are presented. With the application of this water chemistry, seeks to be proven the diminution of the crack propagation speed. (Author)

  7. Water reactor fuel behaviour and fission products release in off-normal and accident conditions

    International Nuclear Information System (INIS)

    1987-09-01

    The present meeting was scheduled by the International Atomic Energy Agency upon the proposal of the Members of the International Working Group on Water Reactor Fuel Performance and Technology and held at the IAEA Headquarters in Vienna from 10 to 13 November 1986. Thirty participants from 17 countries and an international organization attended the meeting. Eighteen papers were presented from 13 countries and one international organization. The meeting was composed of four sessions and covered subjects related to: physico-chemical properties of core materials under off-normal conditions, and their interactions up to and after melt-down (5 papers); core materials deformation, relocation and core coolability under (severe) accident conditions (4 papers); fission products release: including experience, mechanisms and modelling (5 papers); power plant experience (4 papers). A separate abstract was prepared for each of these 18 papers. Four working groups covering the above-mentioned topics were held to discuss the present status of the knowledge and to develop recommendations for future activities in this field. Refs, figs and tabs

  8. Study of startup conditions of a pulsed annular reactor; Estudo das reacoes de partida de um reator anelar pulsado

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Mario Augusto Bezerra da

    2003-10-15

    A new concept of reactor, which combines features of pulsed and stationary reactors, was proposed so as to produce intense neutronic fluxes. Such a reactor, known as VICHFPR (Very Intense Continuous High Flux Pulsed Reactor), consists of a subcritical core with an annular geometry and pulsed by a rotating reflector which acts as a reactivity modulator as it produces a short pulse (approximately equal to 1 ms) of high intensity, guiding the region near the pulser to super-prompt critical state. This dissertation intends to analyze the startup conditions of a Pulsed Annular Reactor. The evolution of the neutron pulse intensity is analyzed when the reactivity modulator is brought upwards according to a helicoidal path from its initial position (far away from the core), when the multiplication factor has a subcritical value, up to the final position (near the core), in which a super-prompt critical state is reached. Part of the analysis is based on the variation of neutron reflection, which is a uniform function of the exit and reflection angles between the core and the modulator. It must be emphasized that this work is an approximation of the real situation. As the initial and final reactor parameters are known, a programming code in Fortran is worked out to provide the multiplication factor and the flux intensity evolution. According to the results obtained with this code, the conditions under which the modulator must be lifted up during the startup are established. Basically, these conditions are related to the analysis of the rising and the rotation velocities, the reflector saving and the initial distance between the reactor and the modulator. The Pulsed Annular Reactor startup was divided into three stages. Because of its negative reactivity in the first two stages, the neutron multiplication is not large, while the last one, having a positive reactivity, shows an intense multiplication as is usually expected when handling pulsed systems. This last stage is quite

  9. Biogas Upgrading via Hydrogenotrophic Methanogenesis in Two-Stage Continuous Stirred Tank Reactors at Mesophilic and Thermophilic Conditions

    DEFF Research Database (Denmark)

    Bassani, Ilaria; Kougias, Panagiotis; Treu, Laura

    2015-01-01

    This study proposes an innovative setup composed by two stage reactors to achieve biogas upgrading coupling the CO2 in the biogas with external H2 and subsequent conversion into CH4 by hydrogenotrophic methanogenesis. In this configuration, the biogas produced in the first reactor was transferred...... production and CO2 conversion was recorded. The consequent increase of pH did not inhibit the process indicating adaptation of microorganisms to higher pH levels. The effects of H2 on the microbial community were studied using high-throughput Illumina random sequences and full-length 16S rRNA genes extracted...... to the second one, where H2 was injected. This configuration was tested at both mesophilic and thermophilic conditions. After H2 addition, the produced biogas was upgraded to average CH4 content of 89% in the mesophilic reactor and 85% in the thermophilic. At thermophilic conditions, a higher efficiency of CH4...

  10. Crack growth behaviour of low-alloy steels for pressure boundary components under transient light water reactor operating conditions - CASTOC, Part I: BWR/NWC conditions

    International Nuclear Information System (INIS)

    Ritter, S.; Seifert, H.P.; Devrient, B.; Roth, A.; Ehrnsten, U.; Ernestova, M.; Zamboch, M.; Foehl, J.; Weissenberg, T.; Gomez-Briceno, D.; Lapena, J.

    2004-01-01

    One of the ageing phenomena of pressure boundary components of light water reactors (LWR) is environmentally-assisted cracking (EAC). The project CASTOC (5. Framework Programme of the EU) was launched September 2000 with six European partners and terminated August 2003. It was focused in particular on the EAC behaviour of low-alloy steels (LAS) and to some extent to weld metal, heat affected zone and the influence of an austenitic cladding. The main objective was directed to the clarification of EAC crack growth behaviour/mechanism of LAS in high-temperature water under steady-state power operation (constant load) and transient operating conditions (e.g., start-up/shut-down, transients in water chemistry and load). Autoclave tests were performed with Western and Russian type reactor pressure vessel steels under simulated boiling water reactor (BWR)/normal water chemistry (NWC) and pressurised water reactor (VVER) conditions. The investigations were performed with fracture mechanics specimens of different sizes and geometries. The applied loading comprised cyclic loads, static loads and load spectra where the static load was periodically interrupted by partial unloading. With regard to water chemistry, the oxygen content (VVER) and impurities of sulphate and chlorides (BWR) were varied beyond allowable limits for continuous operation. The current paper summarises the most important crack growth results obtained under simulated BWR/NWC conditions. The results are discussed in the context of the current crack growth rate curves in the corresponding nuclear codes. (authors)

  11. Crack growth behaviour of low-alloy steels for pressure boundary components under transient light water reactor operating conditions - CASTOC, Part I: BWR/NWC conditions

    Energy Technology Data Exchange (ETDEWEB)

    Ritter, S.; Seifert, H.P. [Paul Scherrer Institute, PSI, Villigen (Switzerland); Devrient, B.; Roth, A. [Framatome ANP GmbH, Erlangen (Germany); Ehrnsten, U. [VTT Industrial Systems, Espoo (Finland); Ernestova, M.; Zamboch, M. [Nuclear Research Institute, NRI, Rez (Czech Republic); Foehl, J.; Weissenberg, T. [Staatliche Materialpruefungsanstalt, MPA, Stuttgart (Germany); Gomez-Briceno, D.; Lapena, J. [Centro de Investigaciones Energeticas Medioambientales y Tecnologicas, CIEMAT, Madrid (Spain)

    2004-07-01

    One of the ageing phenomena of pressure boundary components of light water reactors (LWR) is environmentally-assisted cracking (EAC). The project CASTOC (5. Framework Programme of the EU) was launched September 2000 with six European partners and terminated August 2003. It was focused in particular on the EAC behaviour of low-alloy steels (LAS) and to some extent to weld metal, heat affected zone and the influence of an austenitic cladding. The main objective was directed to the clarification of EAC crack growth behaviour/mechanism of LAS in high-temperature water under steady-state power operation (constant load) and transient operating conditions (e.g., start-up/shut-down, transients in water chemistry and load). Autoclave tests were performed with Western and Russian type reactor pressure vessel steels under simulated boiling water reactor (BWR)/normal water chemistry (NWC) and pressurised water reactor (VVER) conditions. The investigations were performed with fracture mechanics specimens of different sizes and geometries. The applied loading comprised cyclic loads, static loads and load spectra where the static load was periodically interrupted by partial unloading. With regard to water chemistry, the oxygen content (VVER) and impurities of sulphate and chlorides (BWR) were varied beyond allowable limits for continuous operation. The current paper summarises the most important crack growth results obtained under simulated BWR/NWC conditions. The results are discussed in the context of the current crack growth rate curves in the corresponding nuclear codes. (authors)

  12. Effect of process operating conditions in the biomass torrefaction: A simulation study using one-dimensional reactor and process model

    International Nuclear Information System (INIS)

    Park, Chansaem; Zahid, Umer; Lee, Sangho; Han, Chonghun

    2015-01-01

    Torrefaction reactor model is required for the development of reactor and process design for biomass torrefaction. In this study, a one-dimensional reactor model is developed based on the kinetic model describing volatiles components and solid evolution and the existing thermochemical model considering the heat and mass balance. The developed reactor model used the temperature and flow rate of the recycled gas as the practical manipulated variables instead of the torrefaction temperature. The temperature profiles of the gas and solid phase were generated, depending on the practical thermal conditions, using developed model. Moreover, the effect of each selected operating variables on the parameters of the torrefaction process and the effect of whole operating variables with particular energy yield were analyzed. Through the results of sensitivity analysis, it is shown that the residence time insignificantly influenced the energy yield when the flow rate of recycled gas is low. Moreover, higher temperature of recycled gas with low flow rate and residence time produces the attractive properties, including HHV and grindability, of torrefied biomass when the energy yield is specified. - Highlights: • A one-dimensional reactor model for biomass torrefaction is developed considering the heat and mass balance. • The developed reactor model uses the temperature and flow rate of the recycled gas as the practical manipulated variables. • The effect of operating variables on the parameters of the torrefaction process is analyzed. • The results of sensitivity analysis represent notable discussions which were not done by the previous researches

  13. Thermo-hydraulic characteristics of ship propulsion reactor in the conditions of ship motions and safety assessment

    International Nuclear Information System (INIS)

    Kobayashi, Michiyuki; Murata, Hiroyuki; Sawada, Kenichi; Inasaka, Fujio; Aya, Izuo; Shiozaki, Koki

    1999-01-01

    By inputting the experimental data, information and others on thermo-hydraulic characteristics of integrated ship propulsion reactor accumulated hitherto by the Ship Research Institute and some recent cooperation results into the nuclear ship engineering simulation system, it was conducted not only to contribute an improvement study on next ship reactor by executing general analysis and evaluation on motion characteristics under ship body motion conditions, safety at accidents, and others of the integrated ship reactor but also to investigate and prepare some measures to apply fundamental experiment results based on obtained here information to safety countermeasure of the nuclear ships. In 1997 fiscal year, on safety of the integrated ship propulsion reactor loading nuclear ship, by adding experimental data on unstable flow analysis and information on all around of the analysis to general data base fundamental program, development to intellectual data base program was intended; on effect of pulsation flow on thermo-hydraulic characteristics of ship propulsion reactor; after pulsation flow visualization experiment, experimental equipment was reconstructed into heat transfer type to conduct numerical analysis of pulsation flow by confirming validity of numerical analysis code under comparison with the visualization experiment results; and on thermo-hydraulic behavior in storage container at accident of active safety type ship propulsion reactor; a flashing vibration test using new apparatus finished on its higher pressurization at last fiscal year to examine effects of each parameter such as radius and length of exhausting nozzle and pool water temperature. (G.K.)

  14. Investigation of heat transfer in liquid-metal flows under fusion-reactor conditions

    Energy Technology Data Exchange (ETDEWEB)

    Poddubnyi, I. I., E-mail: poddubnyyii@nikiet.ru [Joint Stock Company Dollezhal Research and Development Institute of Power Engineering (JSC NIKIET) (Russian Federation); Pyatnitskaya, N. Yu.; Razuvanov, N. G.; Sviridov, V. G.; Sviridov, E. V. [Russian Academy of Science, Joint Institute of High Temperatures (Russian Federation); Leshukov, A. Yu. [Joint Stock Company Dollezhal Research and Development Institute of Power Engineering (JSC NIKIET) (Russian Federation); Aleskovskiy, K. V. [National Research University Moscow Power Engineering Institute (MPEI) (Russian Federation); Obukhov, D. M. [Joint Stock Company Efremov Institute of Electrophysical Apparatus (Russian Federation)

    2016-12-15

    The effect discovered in studying a downward liquid-metal flow in vertical pipe and in a channel of rectangular cross section in, respectively, a transverse and a coplanar magnetic field is analyzed. In test blanket modules (TBM), which are prototypes of a blanket for a demonstration fusion reactor (DEMO) and which are intended for experimental investigations at the International Thermonuclear Experimental Reactor (ITER), liquid metals are assumed to fulfil simultaneously the functions of (i) a tritium breeder, (ii) a coolant, and (iii) neutron moderator and multiplier. This approach to testing experimentally design solutions is motivated by plans to employ, in the majority of the currently developed DEMO blanket projects, liquid metals pumped through pipes and/or rectangular channels in a transvers magnetic field. At the present time, experiments that would directly simulate liquid-metal flows under conditions of ITER TBM and/or DEMO blanket operation (irradiation with thermonuclear neutrons, a cyclic temperature regime, and a magnetic-field strength of about 4 to 10 T) are not implementable for want of equipment that could reproduce simultaneously the aforementioned effects exerted by thermonuclear plasmas. This is the reason why use is made of an iterative approach to experimentally estimating the performance of design solutions for liquid-metal channels via simulating one or simultaneously two of the aforementioned factors. Therefore, the investigations reported in the present article are of considerable topical interest. The respective experiments were performed on the basis of the mercury magneto hydrodynamic (MHD) loop that is included in the structure of the MPEI—JIHT MHD experimental facility. Temperature fields were measured under conditions of two- and one-sided heating, and data on averaged-temperature fields, distributions of the wall temperature, and statistical fluctuation features were obtained. A substantial effect of counter thermo gravitational

  15. Analysis of metallic fuel pin behaviors under transient conditions of liquid metal reactors

    International Nuclear Information System (INIS)

    Nam, Cheol; Kwon, Hyoung Mun; Hwang, Woan

    1999-02-01

    Transient behavior of metallic fuel pins in liquid metal reactor is quite different to that in steady state conditions. Even in transient conditions, the fuel may behave differently depending on its accident situation and/or accident sequence. This report describes and identifies the possible and hypothetical transient events at the aspects of fuel pin behavior. Furthermore, the transient experiments on HT9 clad metallic fuel have been analyzed, and then failure assessments are performed based on accident classes. As a result, the failure mechanism of coolant-related accidents, such as LOF, is mainly due to plenum pressure and cladding thinning caused by eutectic penetration. In the reactivity-related accidents, such as TOP, the reason to cladding failure is believed to be the fuel swelling as well as plenum pressure. The probabilistic Weibull analysis is performed to evaluate the failure behavior of HT9 clad-metallic fuel pin on coolant related accidents.The Weibull failure function is derived as a function of cladding CDF. Using the function, a sample calculation for the ULOF accident of EBR-II fuel is performed, and the results indicate that failure probability is less the 0.3%. Further discussion on failure criteria of accident condition is provided. Finally, it is introduced the state-of-arts for developing computer codes of reactivity-related fuel pin behavior. The development efforts for a simple model to predict transient fuel swelling is described, and the preliminary calculation results compared to hot pressing test results in literature.This model is currently under development, and it is recommended in the future that the transient swelling model will be combined with the cladding model and the additional development for post-failure behavior of fuel pin is required. (Author). 36 refs., 9 tabs., 18 figs

  16. 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 ...

  17. Design report on the guide box-reactivity and safety control plates for MPR reactor under normal operation conditions

    International Nuclear Information System (INIS)

    Markiewicz, M.

    1999-01-01

    The reactivity control system for the MPR reactor (Multi Purpose Reactor) is a critical component regarding safety, it must ensure a fast shut down, maintaining the reactor in subcritical condition under normal or accidental operation condition. For this purpose, this core component must be designed to maintain its operating capacity during all the residence time and under any foreseen operation condition. The mechanical design of control plates and guide boxes must comply with structural integrity, maintaining its geometric and dimensional stability within the pre-established limits to prevent interferences with other core components. For this, the heat generation effect, mechanical loads and environment and irradiation effects were evaluated during the mechanical design. The reactivity control system is composed of guide boxes, manufactured from Aluminium alloy, located between the fuel elements, and control absorber plates of Ag-In-Cd alloy hermetically enclosed by a cladding of stainless steel sliding inside de guide boxes. The upward-downward movement is transmitted by a rod from the motion device located at the reactor lower part. The design requirements, criteria and limits were established to fulfill with the normal and abnormal operation conditions. The design verifications were performed by analytical method, estimating the guide box and control plates residence time. The result of the analysis performed, shows that the design of the reactivity control system and the material selected, are appropriate to fulfill the functional requirements, with no failures attributed to the mechanical design. (author)

  18. Application of macro-cellular SiC reactor to diesel engine-like injection and combustion conditions

    Science.gov (United States)

    Cypris, Weclas, M.; Greil, P.; Schlier, L. M.; Travitzky, N.; Zhang, W.

    2012-05-01

    One of novel combustion technologies for low emissions and highly efficient internal combustion engines is combustion in porous reactors (PM). The heat release process inside combustion reactor is homogeneous and flameless resulting in a nearly zero emissions level. Such combustion process, however is non-stationary, is performed under high pressure with requirement of mixture formation directly inside the combustion reactor (high pressure fuel injection). Reactor heat capacity resulting in lowering of combustion temperature as well as internal heat recuperation during the engine cycle changes the thermodynamic conditions of the process as compared to conventional engine. For the present investigations a macro-cellular lattice structure based on silicon carbide (non-foam structure) with 600 vertical cylindrical struts was fabricated and applied to engine-like combustion conditions (combustion chamber). The lattice design with a high porosity > 80% was shaped by indirect three-dimensional printing of a SiC powder mixed with a dextrin binder which also serves as a carbon precursor. In order to perform detailed investigations on low-and high-temperature oxidation processes in porous reactors under engine-like conditions, a special combustion chamber has been built and equipped with a Diesel common-rail injection system. This system simulates the thermodynamic conditions at the time instance of injection onset (corresponding to the nearly TDC of compression in a real engine). Overall analysis of oxidation processes (for variable initial pressure, temperature and air excess ratio) for free Diesel spray combustion and for combustion in porous reactor allows selection of three regions representing different characteristics of the oxidation process represented by a single-step and multi-step reactions Another characteristic feature of investigated processes is reaction delay time. There are five characteristic regions to be selected according to the delay time (t) duration

  19. Modelling the chemical behaviour of tellurium species in the reactor pressure vessel and the reactor cooling system under severe accident conditions

    International Nuclear Information System (INIS)

    Alonso, A.; Gonzalez, C.

    1991-07-01

    This state of the art report contains information on the behaviour of tellurium and its compounds in the reactor pressure vessel and the reactor coolant system under light water reactor severe accident conditions. To characterise tellurium behaviour, it is necessary the previous knowledge of the species of tellurium released from the core, and simultaneity of its release with that of other materials which can alter the transport, for instance, control rod and structural materials. Release and transport experiments have been reviewed along with the models implemented in the codes which are used in the international community: TRAPMELT, RAFT, VICTORIA and SOPHIE. From the experiments, it can be concluded that other species different to Te 2 , such as tin telluride and cesium telluride, may be released from the fuel. That is why they must be considered in the transport phenomena. There is also experimental evidence of the strong interaction of Te 2 with Inconel 600 and stainless steel of the pipe walls and structures, however this strong interaction is in competition with the interaction of tellurium with aerosols, which under severe accident conditions may represent an area greater than that of the primary system. It is for the absence of significant tellurium species in the transport models, and also for the interaction of tellurium with aerosols, for which some codes show the greatest deficiencies

  20. Tracking of fuel particles after pin failure in nominal, loss-of-flow and shutdown conditions in the MYRRHA reactor

    Energy Technology Data Exchange (ETDEWEB)

    Buckingham, Sophia; Planquart, Philippe [von Karman Institute, Chaussée de Waterloo 72, B-1640 Rhode-St-Genèse (Belgium); Van Tichelen, Katrien [SCK- CEN, Boeretang 200, 2400 Mol (Belgium)

    2017-02-15

    Highlights: • Quantification of the design and safety of the MYRRHA reactor in the event of a pin failure. • Simulation of different accident scenarios in both forced and natural convection regime. • The accumulation areas at the free-surface in case of the least dense particles depend on the flow regime. • The densest particles form an important deposit at the bottom of the vessel. • Further study of the risk of core blockage requires a detailed model of the core. - Abstract: This work on fuel dispersion aims at quantifying the design and safety of the MYRRHA nuclear reactor. A number of accidents leading to the release of a secondary phase into the primary coolant loop are investigated. Among these scenarios, an incident leading to the failure of one or more of the fuel pins is simulated while the reactor is operating in nominal conditions, but also in natural convection regime either during accident transients such as loss-of-flow or during the normal shut-down of the reactor. Two single-phase CFD models of the MYRRHA reactor are constructed in ANSYS Fluent to represent the reactor in nominal and natural convection conditions. An Euler–Lagrange approach with one-way coupling is used for the flow and particle tracking. Firstly, a steady state RANS solution is obtained for each of the three conditions. Secondly, the particles are released downstream from the core outlet and particle distributions are provided over the coolant circuit. Their size and density are defined such that test cases represent potential extremes that may occur. Analysis of the results highlights different particle behaviors, depending essentially on gravity forces and kinematic effects. Statistical distributions highlight potential accumulation regions that may form at the free-surfaces, on top of the upper diaphragm plate or at the bottom of the vessel. These results help to localize regions of fuel accumulation in order to provide insight for development of strategies for

  1. Scaling the Shear-flow Stabilized Z-pinch to Reactor Conditions

    Science.gov (United States)

    McLean, H. S.; Schmidt, A.; Shumlak, U.; Nelson, B. A.; Golingo, R. P.; Cleveau, E.

    2015-11-01

    We present a conceptual design along with scaling calculations for a pulsed fusion reactor based on the shear-flow-stabilized Z-pinch device. Experiments performed on the ZaP device, at the University of Washington, have demonstrated stable operation for durations of 20 usec at ~100kA discharge current for pinches that are ~1 cm in diameter and 100 cm long. The inverse of the pinch diameter and plasma energy density scale strongly with pinch current and calculations show that maintaining stabilization durations of ~7 usec for increased discharge current (~15x) in a shortened pinch (10 cm) results in a pinch diameter of ~200 um and plasma conditions that approach those needed to support significant fusion burn and energy gain (Ti ~ 30keV, density ~ 3e26/m3, ntau ~1.4e20 sec/m3). Compelling features of the concept include operation at modest discharge current (1.5 MA) and voltage (40kV) along with direct adoption of liquid metals for at least one electrode--technological capabilities that have been proven in existing, commercial, pulse power devices such as large ignitrons. LLNL-ABS-674920. This work performed under the auspices of the U.S. Department of Energy ARPAe ALPHA Program by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344.

  2. Pressurized thermal shock. Thermo-hydraulic conditions in the CNA-I reactor pressure vessel

    International Nuclear Information System (INIS)

    Ventura, Mirta A.; Rosso, Ricardo D.

    2002-01-01

    In this paper we analyze several reports issued by the Utility (Nucleo Electrica S.A.) and related to Reactor Pressure Vessel (RPV) phenomena in the CNA-I Nuclear Power Plant. These analyses are aimed at obtaining conclusions and establishing criteria ensuring the RPV integrity. Special attention was given to the effects ECCS cold-water injection at the RPV down-comer leading to pressurized thermal shock scenarios. The results deal with hypothetical primary system pipe breaks of different sizes, the inadvertent opening of the pressurizer safety valve, the double guillotine break of a live steam line in the containment and the inadvertent actuation pressurizer heaters. Modeling conditions were setup to represent experiments performed at the UPTF, under the hypothesis that they are representative of those that, hypothetically, may occur at the CNA-I. No system scaling analysis was performed, so this assertion and the inferred conclusions are no fully justified, at least in principle. The above mentioned studies, indicate that the RPV internal wall surface temperature will be nearly 40 degree. It was concluded that they allowed a better approximation of PTS phenomena in the RPV of the CNA-I. Special emphasis was made on the influence of the ECCS systems on the attained RPV wall temperature, particularly the low-pressure TJ water injection system. Some conservative hypothesis made, are discussed in this report. (author)

  3. Numerical Study of Pollutant Emissions in a Jet Stirred Reactor under Elevated Pressure Lean Premixed Conditions

    Directory of Open Access Journals (Sweden)

    Karim Mazaheri

    2016-01-01

    Full Text Available Numerical study of pollutant emissions (NO and CO in a Jet Stirred Reactor (JSR combustor for methane oxidation under Elevated Pressure Lean Premixed (EPLP conditions is presented. A Detailed Flow-field Simplified Chemistry (DFSC method, a low computational cost method, is employed for predicting NO and CO concentrations. Reynolds Averaged Navier Stokes (RANS equations with species transport equations are solved. Improved-coefficient five-step global mechanisms derived from a new evolutionary-based approach were taken as combustion kinetics. For modeling turbulent flow field, Reynolds Stress Model (RSM, and for turbulence chemistry interactions, finite rate-Eddy dissipation model are employed. Effects of pressure (3, 6.5 bars and inlet temperature (408–573 K over a range of residence time (1.49–3.97 ms are numerically examined. A good agreement between the numerical and experimental distribution of NO and CO was found. The effect of decreasing the operating pressure on NO generation is much more than the effect of increase in the inlet temperature.

  4. Continuous aryl alcohol oxidase production under growth-limited conditions using a trickle bed reactor.

    Science.gov (United States)

    Pardo-Planas, Oscar; Atiyeh, Hasan K; Prade, Rolf A; Müller, Michael; Wilkins, Mark R

    2018-05-01

    An A. nidulans strain with a pyridoxine marker was used for continuous production of aryl alcohol oxidase (AAO) in a trickle bed reactor (TBR). Modified medium with reduced zinc, no copper, and 5 g/L ascorbic acid that reduced melanin production and increased AAO productivity under growth limited conditions was used. Two air flow rates, 0.11 L/min (0.1 vvm) and 1.1 L/min (1.0 vvm) were tested. More melanin formation and reduced protein productivity were observed with air flow rate of 1.1 L/min. Three random packings were used as support for the fungus inside the TBR column, two of which were hydrophobic and one which was hydrophilic, and three different dilution rates were tested. The use of GEA BCN 030 hydrophobic packing resulted in greater AAO yield and productivity than the other packings. Increasing dilution rates favored melanin formation and citric, lactic and succinic acid accumulation, which decreased AAO yield and productivity. Copyright © 2018 Elsevier Ltd. All rights reserved.

  5. Validation of the ASSERT subchannel code for MAPLE-X10 reactor conditions

    International Nuclear Information System (INIS)

    Carver, M.B.; Kiteley, J.C.; Junop, S.V.; Wasilewicz, J.F.

    1993-01-01

    The ASSERT subchannel analysis code has been developed specifically to model flow and phase distributions within CANDU fuel channels. Recently, ASSERT has been adapted for use in simulating the MAPLE-X10 reactor. ASSERT uses an advanced drift-flux model, which permits the phases to have unequal velocities and unequal temperatures (UVUT), and thus can model non-equilibrium effects such as phase separation tendencies and subcooled boiling. Modelling subcooled boiling accurately is particularly important for MAPLE-X10. This paper briefly summarizes the non-equilibrium model used in the ASSERT code, the equations used to represent these models, and the algorithms used to solve the equations numerically. Very few modifications to the ASSERT models were needed to address MAPLE conditions. These centered on the manner in which finned fuel rods are treated, and they are discussed in the paper. The paper also gives results from validation exercises, in which the ASSERT code predictions of subcooled boiling void-fraction and critical heat flux were compared to experiments using MAPLE-X10 finned fuel elements in annuli and various bundles. 18 refs., 13 figs., 3 tabs

  6. Manufacturing and testing in reactor relevant conditions of brazed plasma facing components of the ITER divertor

    International Nuclear Information System (INIS)

    Bisio, M.; Branca, V.; Marco, M. Di; Federici, A.; Grattarola, M.; Gualco, G.; Guarnone, P.; Luconi, U.; Merola, M.; Ozzano, C.; Pasquale, G.; Poggi, P.; Rizzo, S.; Varone, F.

    2005-01-01

    A fabrication route based on brazing technology has been developed for the realization of the high heat flux components for the ITER vertical target and Dome-Liner. The divertor vertical target is armoured with carbon fiber reinforced carbon and tungsten in the lower straight part and in the upper curved part, respectively. The armour material is joined to heat sinks made of precipitation hardened copper-chromium-zirconium alloy. The plasma facing units of the dome component are based on a tungsten flat tile design with hypervapotron cooling. An innovative brazing technique based on the addition of carbon fibers to the active brazing alloy, developed by Ansaldo Ricerche for applications in the field of the energy production, has been used for the carbon fiber composite to copper joint to reduce residual stresses. The tungsten-copper joint has been realized by direct casting. A proper brazing thermal cycle has been studied to guarantee the required mechanical properties of the precipitation hardened alloy after brazing. The fabrication route of plasma facing components for the ITER vertical target and dome based on the brazing technology has been proved by means of thermal fatigue tests performed on mock-ups in reactor relevant conditions

  7. Monitoring transitory profiles of leachate humic substances in landfill aeration reactors in mesophilic and thermophilic conditions

    Energy Technology Data Exchange (ETDEWEB)

    Tong, Huanhuan [Residues and Resource Reclamation Centre, Nanyang Environment and Water Research Institute, Nanyang Technological University, 1 Cleantech Loop, CleanTech One, 637141 (Singapore); School of Civil and Environmental Engineering, Nanyang Technological University, 50 Nanyang Avenue, 639798 (Singapore); Yin, Ke; Ge, Liya; Giannis, Apostolos [Residues and Resource Reclamation Centre, Nanyang Environment and Water Research Institute, Nanyang Technological University, 1 Cleantech Loop, CleanTech One, 637141 (Singapore); Chuan, Valerie W.L. [School of Civil and Environmental Engineering, Nanyang Technological University, 50 Nanyang Avenue, 639798 (Singapore); Wang, Jing-Yuan, E-mail: JYWANG@ntu.edu.sg [Residues and Resource Reclamation Centre, Nanyang Environment and Water Research Institute, Nanyang Technological University, 1 Cleantech Loop, CleanTech One, 637141 (Singapore); School of Civil and Environmental Engineering, Nanyang Technological University, 50 Nanyang Avenue, 639798 (Singapore)

    2015-04-28

    Highlights: • Polymerization and condensation of humic substances (HS) were enhanced by aeration. • Carboxylic group was enriched in HS by aeration presenting improved hydrophilicity. • Mobility of humic acid, as a result was enhanced by aeration especially in young landfill. • Waste age plays an important role in leachate management during aeration. - Abstract: The presence of humic substances (HS) in landfill leachate is of great interest because of their structural stability and potential toxicity. This study examined the effects of temperature and waste age on the transformation of HS during in situ aeration of bioreactor landfills. By establishing aerobic conditions, dissolved organic carbon (DOC) rapidly accumulated in the bioreactor leachate. Fractional analysis showed that the elevated concentration of humic acids (HAs) was primarily responsible for the increment of leachate strength. Further structural characterization indicated that the molecular weight (MW) and aromacity of HS were enhanced by aeration in conjunction with thermophilic temperature. Interestingly, elevation of HAs concentration was not observed in the aeration reactor with a prolonged waste age, as the mobility of HAs was lowered by the high MW derived from extended waste age. Based on these results, aeration may be more favorable in aged landfills, since dissolution of HAs could be minimized by the evolution to larger MW compared to young landfills. Moreover, increased operation temperature during aeration likely offers benefits for the rapid maturation of HS.

  8. Enrichment of acetogenic bacteria in high rate anaerobic reactors under mesophilic and thermophilic conditions.

    Science.gov (United States)

    Ryan, P; Forbes, C; McHugh, S; O'Reilly, C; Fleming, G T A; Colleran, E

    2010-07-01

    The objective of the current study was to expand the knowledge of the role of acetogenic Bacteria in high rate anaerobic digesters. To this end, acetogens were enriched by supplying a variety of acetogenic growth supportive substrates to two laboratory scale high rate upflow anaerobic sludge bed (UASB) reactors operated at 37 degrees C (R1) and 55 degrees C (R2). The reactors were initially fed a glucose/acetate influent. Having achieved high operational performance and granular sludge development and activity, both reactors were changed to homoacetogenic bacterial substrates on day 373 of the trial. The reactors were initially fed with sodium vanillate as a sole substrate. Although % COD removal indicated that the 55 degrees C reactor out performed the 37 degrees C reactor, effluent acetate levels from R2 were generally higher than from R1, reaching values as high as 5023 mg l(-1). Homoacetogenic activity in both reactors was confirmed on day 419 by specific acetogenic activity (SAA) measurement, with higher values obtained for R2 than R1. Sodium formate was introduced as sole substrate to both reactors on day 464. It was found that formate supported acetogenic activity at both temperatures. By the end of the trial, no specific methanogenic activity (SMA) was observed against acetate and propionate indicating that the methane produced was solely by hydrogenotrophic Archaea. Higher SMA and SAA values against H(2)/CO(2) suggested development of a formate utilising acetogenic population growing in syntrophy with hydrogenotrophic methanogens. Throughout the formate trial, the mesophilic reactor performed better overall than the thermophilic reactor. Copyright 2010 Elsevier Ltd. All rights reserved.

  9. High Burnup Fuel Behaviour under LOCA Conditions as Observed in Halden Reactor Experiments

    International Nuclear Information System (INIS)

    Kolstad, E.; Wiesenack, W.; Oberlander, B.; Tverberg, T.

    2013-01-01

    In the context of assessing the validity of safety criteria for loss of coolant accidents with high burnup fuel, the OECD Halden Reactor Project has implemented an integral in-pile LOCA test series. In this series, fuel fragmentation and relocation, axial gas communication in high burnup rods as affected by gap closure and fuel- clad bonding, and secondary cladding oxidation and hydriding are of major interest. In addition, the data are being used for code validation as well as model development and verification. So far, nine tests with irradiated fuel segments (burnup 40-92 MW.d.kg -1 ) from PWR, BWR and VVER commercial nuclear power plants have been carried out. The in-pile measurements and the PIE results show a good repeatability of the experiments. The paper describes the experimental setup as well as the principal features and main results of these tests. Fuel fragmentation and relocation have occurred to varying degrees in these tests. The paper compares the conditions leading to the presence or absence of fuel fragmentation, e.g., burnup and loss of constraint. Axial gas flow is an important driving force for clad ballooning, fuel relocation and fuel expulsion. The experiments have provided evidence that such gas flow can be impeded in high burnup fuel with a potential impact on the ballooning and fuel dispersal. Although the results of the Halden LOCA tests are, to some extent, amplified by conditions and features deliberately introduced into the test series, the fuel behaviour identified in the Halden tests has an impact on the safety assessment of high burnup fuel and should give rise to improvements of the predictive capabilities of LOCA modelling codes. (author)

  10. Hydrothermal Testing of K Basin Sludge and N Reactor Fuel at Sludge Treatment Project Operating Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Delegard, Calvin H.; Schmidt, Andrew J.; Thornton, Brenda M.

    2007-03-30

    The Sludge Treatment Project (STP), managed for the U. S. DOE by Fluor Hanford (FH), was created to design and operate a process to eliminate uranium metal from K Basin sludge prior to packaging for Waste Isolation Pilot Plant (WIPP). The STP process uses high temperature liquid water to accelerate the reaction, produce uranium dioxide from the uranium metal, and safely discharge the hydrogen. Under nominal process conditions, the sludge will be heated in pressurized water at 185°C for as long as 72 hours to assure the complete reaction (corrosion) of up to 0.25-inch diameter uranium metal pieces. Under contract to FH, the Pacific Northwest National Laboratory (PNNL) conducted bench-scale testing of the STP hydrothermal process in November and December 2006. Five tests (~50 ml each) were conducted in sealed, un-agitated reaction vessels under the hydrothermal conditions (e.g., 7 to 72 h at 185°C) of the STP corrosion process using radioactive sludge samples collected from the K East Basin and particles/coupons of N Reactor fuel also taken from the K Basins. The tests were designed to evaluate and understand the chemical changes that may be occurring and the effects that any changes would have on sludge rheological properties. The tests were not designed to evaluate engineering aspects of the process. The hydrothermal treatment affected the chemical and physical properties of the sludge. In each test, significant uranium compound phase changes were identified, resulting from dehydration and chemical reduction reactions. Physical properties of the sludge were significantly altered from their initial, as-settled sludge values, including, shear strength, settled density, weight percent water, and gas retention.

  11. Serpentine tube heat transfer characteristic under accident condition in gas cooled reactors

    International Nuclear Information System (INIS)

    Abouhadra, D.S.; Byrne, J.E.

    2004-01-01

    In nuclear reactors of the Magnox or advanced gas Cooled type, serpentine tubing is used in some designs to generate steam in a once through arrangement. The calculation of accident conditions using two phase flow codes requires knowledge of the heat transfer behavior of the boiler steam side. A series of experiments to study the blowdown characteristics of a typical serpentine boiler section was devised in order to validate the MARTHA section of the MACE code used by nuclear Electric. The tests were carried out on the Thermal Hydraulics Experimental Research Assembly (THERA) loop at Manchester University. The Thermal Hydraulic Experimental Research Assembly was designed to operate with pressures up to 180 bar and temperatures of 450degC. The geometry and dimensions of this test section were similar to part of a gas cooled reactor boiler of the Hinkley Point design. Blowdown from a pressure of 60 bar with subcoolings of 5degC, 50degC, 100degC formed the main part of the programme. A set of tests was conducted using discharge orifices of different sizes to produce depressurization times from 30 s to 10 mins, and in a few cases, the duration of blowdown approached 1 hour. These times were defined using the criterion of blowdown end as a final pressure of 10% of the initial pressure. Pressures, wall and fluid temperatures were all measured at average time intervals of 1.1s during the excursion and an inventory of the remaining water content in the serpentine was taken when the blowdown ended. Some tests were also conducted at an initial pressure of 30 bar. The results obtained show interesting stratification effects for the relatively fast discharge, with substantial wall circumferential temperature variations. For these tests, a relatively small water inventory remained after blowdown. The discharge characteristics of the serpentine in terms of orifice size have been mapped, and tests at 30 bar show the equivalence in terms of orifice size have been mapped

  12. Effect of conditions of air-lift type reactor work on cadmium adsorption

    International Nuclear Information System (INIS)

    Filipkowska, Urszula; Szymczyk, Paula Szymczyk; Kuczajowska-Zadrozna, Malgorzata; Joezwiak, Tomasz

    2015-01-01

    We investigated cadmium sorption by activated sludge immobilized in 1.5% sodium alginate with 0.5% polyvinyl alcohol. Experiments were conducted in an air-lift type reactor at the constant concentration of biosorbent reaching 5 d.m./dm 3 , at three flow rates: 0.1, 0.25 and 0.5 V/h, and at three concentrations of the inflowing cadmium solution: 10, 25 and 50mg/dm 3 . Analyses determined adsorption capacity of activated sludge immobilized in alginate as well as reactor's work time depending on flow rate and initial concentration of the solution. Results achieved were described with the use of Thomas model. The highest adsorption capacity of the sorbent (determined from the Thomas model), i.e., 200.2mg/g d.m. was obtained at inflowing solution concentration of 50mg/dm 3 and flow rate of 0.1V/h, whereas the lowest one reached 53.69mg/g d.m. at the respective values of 10mg/dm 3 and 0.1 V/h. Analyses were also carried out to determine the degree of biosorbent adsorption capacity utilization at the assumed effectiveness of cadmium removal - at the breakthrough point (C=0.05*C 0 ) and at adsorption capacity depletion point (C−0.9*C0). The study demonstrated that the effectiveness of adsorption capacity utilization was influenced by both the concentration and flow rate of the inflowing solution. The highest degree of sorbent capacity utilization was noted at inflowing solution concentration of 50mg/dm 3 and flow rate of 0.1 V/h, whereas the lowest one at the respective values of 10mg/dm 3 and 0.1 V/h. The course of the process under dynamic conditions was evaluated using coefficients of tangent inclination - a, at point C/C 0 =1/2. A distinct tendency was demonstrated in changes of tangent slope a as affected by the initial concentration of cadmium and flow rate of the solution. The highest values of a coefficient were achieved at the flow rate of 0.1 V/h and initial cadmium concentration of 50mg/dm 3 .

  13. Model experiments on simulation of the WWER water-chemical conditions at loop facilities of the MIR reactor

    International Nuclear Information System (INIS)

    Benderskaya, O.S.; Zotov, E.A.; Kuprienko, V.A.; Ovchinnikov, V.A.

    1999-01-01

    The experiments on simulation of the WWER type reactors water-chemical conditions have been started at the State Scientific Center RIAR. These experiments are being conducted at the multi-loop research MIR reactor at the PVK-2 loop facility. The dosage stand was created. It allows introduction of boric acid, potassium and lithium hydroxides, ammonia solutions and gaseous hydrogen. Corrosion tests of the Russian E-635 and E-110 alloys are being conducted at the PVK-2 loop under the WWER water-chemical conditions. If necessary, fuel elements are periodically extracted from the reactor to perform visual examination, to measure their length, diameter, to remove the deposits from the claddings, to measure the burnup and to distribute the fission products over the fuel element by gamma-spectrometry. The chemical analytical 'on line' equipment produced by the ORBISPHERE Laboratory (Switzerland) will be commissioned in the nearest future to measure concentration of the dissolved hydrogen and oxygen as well as pH and specific conductivity. The objective of the report is to familiarize the participants of the IAEA Technical Committee with the capabilities of performing the model water-chemical experiments under the MIR reactor loop facility conditions. (author)

  14. A study on nonlinear behavior of reactor containment structures during ultimate accident condition(I)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sun Hoon; Kim, Young Jin; Park, Joo Yeon [Youngdong Univ., Yeongdong (Korea, Republic of)] (and others)

    2003-03-15

    In this study, the following scope and contents are established for first year's study of determining ultimate pressure capacity of CANDU-type reactor containment. State-of-arts on the prediction of the ultimate pressure capacity of prestressed concrete reactor containment. Comparative study on structural characteristics and analysis model of CANDU-type reactor containment. State-of-arts on evaluation method of the ultimate pressure capacity of prestressed concrete reactor containment. Enhancement of evaluation method of the ultimate pressure capacity for PWR containment structure. In order to determine a realistic lower bound of a typical reactor containment structural capacity for internal pressure, modelling techniques and analytical investigation to predict its non-linear behavior up to ultimate capacity are required. Especially, the in-depth evaluation of modeling technique and analysis procedure for determining ultimate pressure capacity of CANDU-type reactor containment is required. Therefore, modelling techniques and analytical investigation to predict its non-linear behavior up to ultimate pressure capacity of CANDU-type reactor containment for internal pressure will be suggested in this study.

  15. A study on nonlinear behavior of reactor containment structures during ultimate accident condition(I)

    International Nuclear Information System (INIS)

    Kim, Sun Hoon; Kim, Young Jin; Park, Joo Yeon

    2003-03-01

    In this study, the following scope and contents are established for first year's study of determining ultimate pressure capacity of CANDU-type reactor containment. State-of-arts on the prediction of the ultimate pressure capacity of prestressed concrete reactor containment. Comparative study on structural characteristics and analysis model of CANDU-type reactor containment. State-of-arts on evaluation method of the ultimate pressure capacity of prestressed concrete reactor containment. Enhancement of evaluation method of the ultimate pressure capacity for PWR containment structure. In order to determine a realistic lower bound of a typical reactor containment structural capacity for internal pressure, modelling techniques and analytical investigation to predict its non-linear behavior up to ultimate capacity are required. Especially, the in-depth evaluation of modeling technique and analysis procedure for determining ultimate pressure capacity of CANDU-type reactor containment is required. Therefore, modelling techniques and analytical investigation to predict its non-linear behavior up to ultimate pressure capacity of CANDU-type reactor containment for internal pressure will be suggested in this study

  16. Proceedings of the twenty-fourth water reactor safety information meeting. Volume 2: Reactor pressure vessel embrittlement and thermal annealing; Reactor vessel lower head integrity; Evaluation and projection of steam generator tube condition and integrity

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)

    1997-02-01

    This three-volume report contains papers presented at the Twenty-Fourth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 21--23, 1996. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Czech Republic, Finland, France, Japan, Norway, Russia and United Kingdom. This volume is divided into the following sections: reactor pressure vessel embrittlement and thermal annealing; reactor vessel lower head integrity; and evaluation and projection of steam generator tube condition and integrity. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

  17. Proceedings of the twenty-fourth water reactor safety information meeting. Volume 2: Reactor pressure vessel embrittlement and thermal annealing; Reactor vessel lower head integrity; Evaluation and projection of steam generator tube condition and integrity

    International Nuclear Information System (INIS)

    Monteleone, S.

    1997-02-01

    This three-volume report contains papers presented at the Twenty-Fourth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 21--23, 1996. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Czech Republic, Finland, France, Japan, Norway, Russia and United Kingdom. This volume is divided into the following sections: reactor pressure vessel embrittlement and thermal annealing; reactor vessel lower head integrity; and evaluation and projection of steam generator tube condition and integrity. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database

  18. Robust stabilization of burn conditions in subignited fusion reactors using artificial neural networks

    International Nuclear Information System (INIS)

    Vitela, E. Javier; Martinell, J. Julio

    2000-01-01

    In this work it is shown that robust burn control in long pulse operations of thermonuclear reactors can be successfully achieved with artificial neural networks. The results reported here correspond to a volume averaged zero-dimensional nonlinear model of a subignited fusion device using the design parameters of the tokamak EDA-ITER group. A Radial Basis Neural Network (RBNN) was trained to provide feedback stabilization at a fixed operating point independently of any particular scaling law that the reactor confinement time may follow. A numerically simulated transient is used to illustrate the stabilization capabilities of the resulting RBNN when the reactor follows an ELMy scaling law corrupted with Gaussian noise. (author)

  19. Corrosion fatigue initiation and short crack growth behaviour of austenitic stainless steels under light water reactor conditions

    International Nuclear Information System (INIS)

    Seifert, H.P.; Ritter, S.; Leber, H.J.

    2012-01-01

    Highlights: ► Corrosion fatigue in austenitic stainless steels under light water reactor conditions. ► Identification of major parameters of influence on initiation and short crack growth. ► Critical system conditions for environmental reduction of fatigue initiation life. ► Comparison with the environmental factor (F env ) approach. - Abstract: The corrosion fatigue initiation and short crack growth behaviour of different wrought low-carbon and stabilised austenitic stainless steels was characterised under simulated boiling water reactor and pressurised water reactor primary water conditions by cyclic fatigue tests with sharply notched fracture mechanics specimens. The special emphasis was placed to the behaviour at low corrosion potentials and, in particular, to hydrogen water chemistry conditions. The major parameter effects and critical conjoint threshold conditions, which result in relevant environmental reduction and acceleration of fatigue initiation life and subsequent short crack growth, respectively, are discussed and summarised. The observed corrosion fatigue behaviour is compared with the fatigue evaluation procedures in codes and regulatory guidelines.

  20. EFFECTS OF REACTOR CONDITIONS ON ELECTROCHEMICAL DECHLORINATION OF TRICHLOROETHYLENE USING GRANULAR-GRAPHITE ELECTRODE.

    Science.gov (United States)

    Trichloroethylene (TCE) was electrochemically dechlorinated in aqueous environments using granular graphite cathode in a mixed reactor. Effects of pH, current, electrolyte type, and flow rate on TCE dechlorination rate were evaluated. TCE dechlorination rate constant and gas pr...

  1. Decay heat removal and heat transfer under normal and accident conditions in gas cooled reactors

    International Nuclear Information System (INIS)

    1994-08-01

    The meeting was convened by the International Atomic Energy Agency on the recommendation of the IAEA's International Working Group on Gas Cooled Reactors. It was attended by participants from China, France, Germany, Japan, Poland, the Russian Federation, Switzerland, the United Kingdom and the United States of America. The meeting was chaired by Prof. Dr. K. Kugeler and Prof. Dr. E. Hicken, Directors of the Institute for Safety Research Technology of the KFA Research Center, and covered the following: Design and licensing requirements for gas cooled reactors; concepts for decay heat removal in modern gas cooled reactors; analytical methods for predictions of thermal response, accuracy of predictions; experimental data for validation of predictive methods - operational experience from gas cooled reactors and experimental data from test facilities. Refs, figs and tabs

  2. Core management, operational limits and conditions and safety aspects of the Australian High Flux Reactor (HIFAR)

    International Nuclear Information System (INIS)

    Town, S.L.

    1997-01-01

    HIFAR is a DIDO class reactor which commenced routine operation at approximately 10 MW in 1960. It is principally used for production of medical radio-isotopes, scientific research using neutron scattering facilities and irradiation of silicon ingots for the electronics industry. A detailed description of the core, including fuel types, is presented. Details are given of the current fuel management program HIFUEL and the experimental measurements associated with reactor physics analysis of HIFAR are discussed. (author)

  3. Crack growth behaviour of low-alloy steels for pressure boundary components under transient light water reactor operating conditions - CASTOC, Part II: WWER conditions

    Energy Technology Data Exchange (ETDEWEB)

    Ernestova, M.; Zamboch, M. [Nuclear Research Institute, NRI, Rez (Czech Republic); Devrient, B.; Roth, A. [Framatome ANP GmbH, Erlangen (Germany); Ehrnsten, U. [VTT Industrial Systems, Espoo (Finland); Foehl, J.; Weissenberg, T. [Staatliche Materialpruefungsanstalt, MPA, Stuttgart (Germany); Gomez-Briceno, D.; Lapena, J. [Centro de Investigaciones Energeticas Medioambientales y Tecnologicas, CIEMAT, Madrid (Spain); Ritter, S.; Seifert, H.P. [Paul Scherrer Institute, PSI, Villigen (Switzerland)

    2004-07-01

    One of the ageing phenomena of pressure boundary components of light water reactors (LWRs) is environmentally-assisted cracking (EAC). The project CASTOC (5. Framework Programme of the EU) was launched September 2000 with six European partners and terminated August 2003. It focused in particular on the EAC behaviour of low-alloy steels (LAS) and to some extent to weld metal, heat affected zone and the influence of an austenitic cladding. The main objective was directed to the clarification of crack growth behavior of LAS in high-temperature water due to EAC under constant load (steady-state power operation), to study the effect of transient conditions (during operation or start-up/shut-down of a plant) using their impact on time-based and cycle-based crack growth rates and to a more detailed understanding of the acting mechanisms. Autoclave tests were performed with Western and Russian type reactor pressure vessel steels under simulated boiling water reactor (BWR)/normal water chemistry (NWC) and pressurized water reactor (WWER) conditions. The investigations were performed with fracture mechanics specimens of different sizes and geometries. The applied loading comprised cyclic loads, static loads and load spectra where the static load was periodically interrupted by partial unloading. With regard to water chemistry, the oxygen content (WWER) and impurities of sulphate and chlorides (BWR) were varied beyond allowable limits for continuous operation. The current paper summarizes the most important crack growth results obtained under simulated WWER conditions. The influence of oxygen content and the effect of specimen size (C(T)25 versus C(T)50 specimens) on the crack growth rates are shown. The results are discussed in the context of the current crack growth rate curves in the corresponding nuclear codes. (authors)

  4. Crack growth behaviour of low-alloy steels for pressure boundary components under transient light water reactor operating conditions - CASTOC, Part II: WWER conditions

    International Nuclear Information System (INIS)

    Ernestova, M.; Zamboch, M.; Devrient, B.; Roth, A.; Ehrnsten, U.; Foehl, J.; Weissenberg, T.; Gomez-Briceno, D.; Lapena, J.; Ritter, S.; Seifert, H.P.

    2004-01-01

    One of the ageing phenomena of pressure boundary components of light water reactors (LWRs) is environmentally-assisted cracking (EAC). The project CASTOC (5. Framework Programme of the EU) was launched September 2000 with six European partners and terminated August 2003. It focused in particular on the EAC behaviour of low-alloy steels (LAS) and to some extent to weld metal, heat affected zone and the influence of an austenitic cladding. The main objective was directed to the clarification of crack growth behavior of LAS in high-temperature water due to EAC under constant load (steady-state power operation), to study the effect of transient conditions (during operation or start-up/shut-down of a plant) using their impact on time-based and cycle-based crack growth rates and to a more detailed understanding of the acting mechanisms. Autoclave tests were performed with Western and Russian type reactor pressure vessel steels under simulated boiling water reactor (BWR)/normal water chemistry (NWC) and pressurized water reactor (WWER) conditions. The investigations were performed with fracture mechanics specimens of different sizes and geometries. The applied loading comprised cyclic loads, static loads and load spectra where the static load was periodically interrupted by partial unloading. With regard to water chemistry, the oxygen content (WWER) and impurities of sulphate and chlorides (BWR) were varied beyond allowable limits for continuous operation. The current paper summarizes the most important crack growth results obtained under simulated WWER conditions. The influence of oxygen content and the effect of specimen size (C(T)25 versus C(T)50 specimens) on the crack growth rates are shown. The results are discussed in the context of the current crack growth rate curves in the corresponding nuclear codes. (authors)

  5. Thermodynamic Modelling of Fe-Cr-Ni-Spinel Formation at the Light-Water Reactor Conditions

    International Nuclear Information System (INIS)

    Kurepin, V. A.; Kulik, D. A.; Hitpold, A.; Nicolet, M.

    2002-03-01

    In the light water reactors (LWR), the neutron activation and transport of corrosion products is of concern in the context of minimizing the radiation doses received by the personnel during maintenance works. A practically useful model for transport and deposition of the stainless steel corrosion products in LWR can only be based on an improved understanding of chemical processes, in particular, on the attainment of equilibrium in this hydrothermal system, which can be described by means of a thermodynamic solid-solution -aqueous-solution (SSAS) model. In this contribution, a new thermodynamic model for a Fe-Cr-Ni multi-component spinel solid solutions was developed that considers thermodynamic consequences of cation interactions in both spinel sub-Iattices. The obtained standard thermodynamic properties of two ferrite and two chromite end-members and their mixing parameters at 90 bar pressure and 290 *c temperature predict a large miscibility gap between (Fe,Ni) chromite and (Fe,Ni) ferrite phases. Together with the SUPCRT92-98 thermo- dynamic database for aqueous species, the 'spinel' thermodynamic dataset was applied to modeling oxidation of austenitic stainless steel in hydrothermal water at 290*C and 90 bar using the Gibbs energy minimization (GEM) algorithm, implemented in the GEMS-PSI code. Firstly, the equilibrium compositions of steel oxidation products were modelIed as function of oxygen fugacity .fO 2 by incremental additions of O 2 in H 2 O-free system Cr-Fe- Ni-O. Secondly, oxidation of corrosion products in the Fe-Cr-Ni-O-H aquatic system was modelIed at different initial solid/water ratios. It is demonstrated that in the transition region from hydrogen regime to oxygen regime, the most significant changes in composition of two spinel-oxide phases (chromite and ferrite) and hematite must take place. Under more reduced conditions, the Fe-rich ferrite (magnetite) and Ni-poor chromite phases co-exist at equilibrium with a metal Ni phase, maintaining

  6. Bioremediation of polychlorinated-p-dioxins/dibenzofurans contaminated soil using simulated compost-amended landfill reactors under hypoxic conditions

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Wei-Yu; Wu, Jer-Horng, E-mail: enewujh@mail.ncku.edu.tw; Lin, Shih-Chiang; Chang, Juu-En

    2016-07-15

    Highlights: • We developed a new hypoxic reactor system for remediating PCDD/Fs. • We demonstrated effects of compost on the degradation of PCDD/Fs. • We uncovered microbial compositions and dynamics during the degradation of PCDD/Fs. - Abstract: Compost-amended landfill reactors were developed to reduce polychlorinated-p-dioxins and dibenzofurans (PCDD/Fs) in contaminated soils. By periodically recirculating leachate and suppling oxygen, the online monitoring of the oxidation reduction potential confirmed that the reactors were maintained under hypoxic conditions, with redox levels constantly fluctuating between −400 and +80 mV. The subsequent reactor operation demonstrated that PCDD/F degradation in soil could be facilitated by amending compost originating from the cow manure and waste sludge and that the degradation might be affected by the availability of easily degradable substrates in the soil and compost. The pyrosequencing analysis of V4/V5 regions of bacterial 16S rRNA genes suggested that species richness of the soil microbial community was increased by a factor of 1.37–1.61. Although the bacterial community varied with the compost origin and changed markedly during reactor operation, it was dominated by Alphaproteobacteria, Gammaproteobacteria, Actinobacteria, and Firmicutes. The aerotolerant anaerobic Sedimentibacter and Propionibacterium spp., and the uncultured Chloroflexi group could be temporarily induced to a high abundance by amending the cow manure compost; the bacterial growths were associated with the rapid degradation of PCDD/Fs. Overall, the novel bioremediation method for PCDD/F-contaminated soils using hypoxic conditions was effective, simple, energy saving, and thus easily practicable.

  7. Bioremediation of polychlorinated-p-dioxins/dibenzofurans contaminated soil using simulated compost-amended landfill reactors under hypoxic conditions

    International Nuclear Information System (INIS)

    Chen, Wei-Yu; Wu, Jer-Horng; Lin, Shih-Chiang; Chang, Juu-En

    2016-01-01

    Highlights: • We developed a new hypoxic reactor system for remediating PCDD/Fs. • We demonstrated effects of compost on the degradation of PCDD/Fs. • We uncovered microbial compositions and dynamics during the degradation of PCDD/Fs. - Abstract: Compost-amended landfill reactors were developed to reduce polychlorinated-p-dioxins and dibenzofurans (PCDD/Fs) in contaminated soils. By periodically recirculating leachate and suppling oxygen, the online monitoring of the oxidation reduction potential confirmed that the reactors were maintained under hypoxic conditions, with redox levels constantly fluctuating between −400 and +80 mV. The subsequent reactor operation demonstrated that PCDD/F degradation in soil could be facilitated by amending compost originating from the cow manure and waste sludge and that the degradation might be affected by the availability of easily degradable substrates in the soil and compost. The pyrosequencing analysis of V4/V5 regions of bacterial 16S rRNA genes suggested that species richness of the soil microbial community was increased by a factor of 1.37–1.61. Although the bacterial community varied with the compost origin and changed markedly during reactor operation, it was dominated by Alphaproteobacteria, Gammaproteobacteria, Actinobacteria, and Firmicutes. The aerotolerant anaerobic Sedimentibacter and Propionibacterium spp., and the uncultured Chloroflexi group could be temporarily induced to a high abundance by amending the cow manure compost; the bacterial growths were associated with the rapid degradation of PCDD/Fs. Overall, the novel bioremediation method for PCDD/F-contaminated soils using hypoxic conditions was effective, simple, energy saving, and thus easily practicable.

  8. Analysis of core and core barrel heat-up under conditions simulating severe reactor accidents

    International Nuclear Information System (INIS)

    Chellaiah, S.; Viskanta, R.; Ranganathan, P.; Anand, N.K.

    1987-01-01

    This paper reports on the development of a model for estimating the temperature distributions in the reactor core, core barrel, thermal shield and reactor pressure vessel of a PWR during an undercooling transient. A number of numerical calculations simulating the core uncovering of the TMI-2 reactor and the subsequent heat-up of the core have been performed. The results of the calculations show that the exothermic heat release due to Zircaloy oxidation contributes to the sharp heat-up of the core. However, the core barrel temperature rise which is driven by the temperature increase of the edge of the core (e.g., the core baffle) is very modest. The maximum temperature of the core barrel never exceeded 610 K (at a system pressure of 68 bar) after a 75 minute simulation following the start of core uncovering

  9. The Canadian development program for conditioning CANDU reactor wastes for disposal

    International Nuclear Information System (INIS)

    Charlesworth, D.H.; Bourns, W.T.; Buckley, L.P.

    1978-07-01

    Currently, radioactive wastes arising from the operation of Canadian nuclear reactors are placed in interim storage in concrete containment structures except for gaseous and liquid wastes containing small amounts of radioactivity which are dispersed. With the objective of replacing storage by permanent disposal, a program is underway to develop and demonstrate an integrated process for converting all reactor wastes to a stable, leach-resistant form which will immobilize the radionuclides in the waste repository. The major tool for this development is a Waste Treatment Centre, now being constructed at Chalk River Nuclear Laboratories, which will combine reverse-osmosis, incineration, evaporation and bituminizing processes. (author)

  10. Research reactors - an overview

    International Nuclear Information System (INIS)

    West, C.D.

    1997-01-01

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

  11. Design of a proteus lattice representative of a burnt and fresh fuel interface at power conditions in light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hursin, M.; Perret, G. [Paul Scherrer Institut (PSI), 5232 Villigen (Switzerland)

    2012-07-01

    The research program LIFE (Large-scale Irradiated Fuel Experiment) between PSI and Swissnuclear has been started in 2006 to study the interaction between large sets of burnt and fresh fuel pins in conditions representative of power light water reactors. Reactor physics parameters such as flux ratios and reaction rate distributions ({sup 235}U and {sup 238}U fissions and {sup 238}U capture) are calculated to estimate an appropriate arrangement of burnt and fresh fuel pins within the central element of the test zone of the zero-power research reactor PROTEUS. The arrangement should minimize the number of burnt fuel pins to ease fuel handling and reduce costs, whilst guaranteeing that the neutron spectrum in both burnt and fresh fuel regions and at their interface is representative of a large uniform array of burnt and fresh pins in the same moderation conditions. First results are encouraging, showing that the burnt/fresh fuel interface is well represented with a 6 x 6 bundle of burnt pins. The second part of the project involves the use of TSUNAMI, CASMO-4E and DAKOTA to perform parametric and optimization studies on the PROTEUS lattice by varying its pitch (P) and fraction of D{sub 2}O in moderator (F{sub D2O}) to be as representative as possible of a power light water reactor core at hot full power conditions at beginning of cycle (BOC). The parameters P and F{sub D2O} that best represent a PWR at BOC are 1.36 cm and 5% respectively. (authors)

  12. Cellulase production by Trichoderma harzianum in static and mixed solid-state fermentation reactors under nonaseptic conditions

    Energy Technology Data Exchange (ETDEWEB)

    Deschamps, F.; Giuliano, C.; Asther, M.; Huet, M.C.; Roussos, S.

    1985-09-01

    Cellulase production from lignocellulosic materials was studied in solid-state cultivation by both static and mixed techniques under nonaseptic conditions. The effects of fermentation conditions, such as moisture content, pH, temperature, and aeration, on cellulase production by Trichoderma harzianum using a mixture of wheat straw (80%) and bran (20%) were investigated. With a moisture content of 74% and a pH of 5.8, 18 IU filter paper activity and 198 IU endoglucanase activity/g initial substrate content were obtained in 66 hours. The extension from static column cultivation to stirred tank reactor of 65 l capacity gave similar yields of cellulase.

  13. Hydrolysis and degradation of filtrated organic particulates in a biofilm reactor under anoxic and aerobic conditions

    DEFF Research Database (Denmark)

    Janning, K.F.; Mesterton, K.; Harremoës, P.

    1997-01-01

    Two experiments were performed in order to investigate the anoxic and the aerobic degradation of filtrated organic matter in a biofilter. In submerged lab: scale reactors with Biocarbone media as filter material, accumulated particulate organic matter from pre-settled wastewater served as the only...

  14. Development of reactivity feedback effect measurement techniques under sub-critical condition in fast reactors

    International Nuclear Information System (INIS)

    Kitano, A.; Nishi, H.; Suzuki, T.; Okajima, S.; Kanemoto, S.

    2012-01-01

    The first-of-a-kind reactor has been licensed by a safety examination of the plant design based on the measured data in precedent mock-up experiments. The validity of the safety design can be confirmed without a mock-up experiment, if the reactor feed-back characteristics can be measured before operation, with the constructed reactor itself. The 'Synthesis Method', a systematic and sophisticated method of sub-criticality measurement, is proposed in this work to ensure the safety margin before operation. The 'Synthesis Method' is based on the modified source multiplication method (MSM) combined with the noise analysis method to measure the reference sub-criticality level for MSM. A numerical simulation for the control-rod reactivity worth and the isothermal feed-back reactivity was conducted for typical fast reactors of 100 MWe-size, 300 MWe-size, 750 MWe-size, and 1500 MWe-size to investigate the applicability of Synthesis Method. The number of neutron detectors and their positions necessary for the measurement were investigated for both methods of MSM and the noise analysis by a series of parametric survey calculations. As a result, it was suggested that a neutron detector located above the core center and three or more neutron detectors located above the radial blanket region enable the measurement of sub-criticality within 10% uncertainty from -$0.5 to -$2 and within 15% uncertainty for the deeper sub-criticality. (authors)

  15. Biofilm formation on membranes used for membrane aerated biological reactors, under different stress conditions

    International Nuclear Information System (INIS)

    Andrade-Molinar, C.; Ballinas-Casarrubias, M. L.; Solis-Martinez, F. J.; Rivera-Chavira, B. E.; Cuevas-Rodirguez, G.; Nevarez-Moorillon, G. V.

    2009-01-01

    Bacterial biofilm play an important role in wastewater treatment processes, and have been optimized in the membrane aerated biofilm reactors (MABR). In MABR, a hydrophobic membrane is used as support for the formation of biofilm, and supplements enough aeration to assure an aerobic process. (Author)

  16. Thermodynamic Modelling of Fe-Cr-Ni-Spinel Formation at the Light-Water Reactor Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Kurepin, V.A.; Kulik, D.A.; Hitpold, A.; Nicolet, M

    2002-03-01

    In the light water reactors (LWR), the neutron activation and transport of corrosion products is of concern in the context of minimizing the radiation doses received by the personnel during maintenance works. A practically useful model for transport and deposition of the stainless steel corrosion products in LWR can only be based on an improved understanding of chemical processes, in particular, on the attainment of equilibrium in this hydrothermal system, which can be described by means of a thermodynamic solid-solution -aqueous-solution (SSAS) model. In this contribution, a new thermodynamic model for a Fe-Cr-Ni multi-component spinel solid solutions was developed that considers thermodynamic consequences of cation interactions in both spinel sub-Iattices. The obtained standard thermodynamic properties of two ferrite and two chromite end-members and their mixing parameters at 90 bar pressure and 290 *c temperature predict a large miscibility gap between (Fe,Ni) chromite and (Fe,Ni) ferrite phases. Together with the SUPCRT92-98 thermo- dynamic database for aqueous species, the 'spinel' thermodynamic dataset was applied to modeling oxidation of austenitic stainless steel in hydrothermal water at 290*C and 90 bar using the Gibbs energy minimization (GEM) algorithm, implemented in the GEMS-PSI code. Firstly, the equilibrium compositions of steel oxidation products were modelIed as function of oxygen fugacity .fO{sub 2} by incremental additions of O{sub 2} in H{sub 2}O-free system Cr-Fe- Ni-O. Secondly, oxidation of corrosion products in the Fe-Cr-Ni-O-H aquatic system was modelIed at different initial solid/water ratios. It is demonstrated that in the transition region from hydrogen regime to oxygen regime, the most significant changes in composition of two spinel-oxide phases (chromite and ferrite) and hematite must take place. Under more reduced conditions, the Fe-rich ferrite (magnetite) and Ni-poor chromite phases co-exist at equilibrium with a metal Ni

  17. Dynamic analysis of the reactor building for soft (Kozloduy) and hard (Temelin) soil conditions and different seismic loading

    International Nuclear Information System (INIS)

    Krutzik, N.

    1995-01-01

    Analyses were conducted for the reactor building to determine the dynamic responses of the coupled system, soil and structure and the forces in the characteristic structural members. This report summarizes the results of structural dynamic analyses derived for soft and hard soil conditions by the modal time history method using substructure models as well as (for soft soil conditions) in the frequency domain using complex (coupled) models of the soil and the structure. The mathematical model of the reactor building is represented as a lumped mass beam model. The capabilities of the soil were represented by means of global frequency independent springs and dampers (substructure models) or by an appropriate final element model. The results of the above-mentioned analysis presented in this report comprise in particular the maximum values of accelerations, displacements and internal forces as well as the acceleration response spectra for the relevant building regions. The time domain (modal time history) calculations were performed for real soil conditions which corresponds to the site Kozloduy (soft) and Temelin (hard). As seismic input data the corresponding free-field data here been used. The dynamic response obtained for the soft-soil conditions using both type of (substructure and complex) models were compared and demonstrated in one plot. Similar comparison were performed for the results obtained for soft and hard soil conditions

  18. Experimental and Computational Study of the Hydrodynamics of Trickle Bed Flow Reactor Operating Under Different Pressure Conditions

    Science.gov (United States)

    Rabbani, S.; Ben Salem, I.; Nadeem, H.; Kurnia, J. C.; Shamim, T.; Sassi, M.

    2014-12-01

    Pressure drop estimation and prediction of liquid holdup play a crucial role in design and operation of trickle bed reactors. Experiments are performed for Light Gas Oil (LGO)-nitrogen system in ambient temperature conditions in an industrial pilot plant with reactor height 0.79 m and diameter of 0.0183 m and pressure ranging from atmospheric to 10 bars. It was found that pressure drop increased with increase in system pressure, superficial gas velocity and superficial liquid velocity. It was demonstrated in the experiments that liquid holdup of the system increases with the increase in superficial liquid velocity and tends to decrease with increase in superficial gas velocity which is in good agreement with existing literature. Similar conditions were also simulated using CFD-software FLUENT. The Volume of Fluid (VoF) technique was employed in combination with "discrete particle approach" and results were compared with that of experiments. The overall pressure drop results were compared with the different available models and a new comprehensive model was proposed to predict the pressure drop in Trickle Bed Flow Reactor.

  19. Cracks propagation by stress corrosion cracking in conditions of Boiling Water Reactor (BWR); Propagacion de grietas por corrosion bajo esfuerzo en condiciones de reactor de agua hirviente (BWR)

    Energy Technology Data Exchange (ETDEWEB)

    Fuentes C, P

    2003-07-01

    This work presents the results of the assays carried out in the Laboratory of Hot Cells of the National Institute of Nuclear Research (ININ) to a type test tube Compact Tension (CT), built in steel austenitic stainless type 304L, simulating those conditions those that it operates a Boiling Water Reactor (BWR), at temperature 288 C and pressure of 8 MPa, to determine the speed to which the cracks spread in this material that is of the one that different components of a reactor are made, among those that it highlights the reactor core vessel. The application of the Hydrogen Chemistry of the Water is presented (HWC) that is one alternative to diminish the corrosion effect low stress in the component, this is gets controlling the quantity of oxygen and of hydrogen as well as the conductivity of the water. The rehearsal is made following the principles of the Mechanics of Elastic Lineal Fracture (LEFM) that considers a crack of defined size with little plastic deformation in the tip of this; the measurement of crack advance is continued with the technique of potential drop of direct current of alternating signal, this is contained inside the standard Astm E-647 (Method of Test Standard for the Measurement of Speed of Growth of Crack by fatigue) that is the one that indicates us as carrying out this test. The specifications that should complete the test tubes that are rehearsed as for their dimensions, it forms, finish and determination of mechanical properties (tenacity to the fracture mainly) they are contained inside the norm Astm E-399, the one which it is also based on the principles of the fracture mechanics. The obtained results were part of a database to be compared with those of other rehearsals under different conditions, Normal Chemistry of the Water (NWC) and it dilutes with high content of O{sub 2}; to determine the conditions that slow more the phenomena of stress corrosion cracking, as well as the effectiveness of the used chemistry and of the method of

  20. Development of TPNCIRC code for Evaluation of Two-Phase Natural Circulation Flow Performance under External Reactor Vessel Cooling Conditions

    International Nuclear Information System (INIS)

    Choi, A-Reum; Song, Hyuk-Jin; Park, Jong-Woon

    2015-01-01

    During a severe accident, corium is relocated to the lower head of the nuclear reactor pressure vessel (RPV). Design concept of retaining the corium inside a nuclear reactor pressure vessel (RPV) through external cooling under hypothetical core melting accidents is called external reactor vessel cooling (ERVC). In this respect, validated two-phase natural circulation flow (TPNC) model is necessary to determine the adequacy of the ERVC design and operating conditions such as inlet area, form losses, gap distance, riser length and coolant conditions. The most important model generally characterizing the TPNC are void fraction and two-phase friction factors. Typical experimental and analytical studies to be referred to on two-phase circulation flow characteristics are those by Reyes, Gartia et al. based on Vijayan et al., Nayak et al. and Dubey et al. In the present paper, two-phase natural circulation (TPNC) flow characteristics under external reactor vessel cooling (ERVC) conditions are studied using two existing TPNC flow models of Reyes and Gartia et al. incorporating more improved void fraction and two-phase friction models. These models and correlations are integrated into a computer program, TPNCIRC, which can handle candidate ERVC design parameters, such as inlet, riser and downcomer flow lengths and areas, gap size between reactor vessel and surrounding insulations, minor loss factors and operating parameters of decay power, pressure and subcooling. Accuracy of the TPNCIRC program is investigated with respect to the flow rate and void fractions for existing measured data from a general experiment and ULPU specifically designed for the AP1000 in-vessel retention. Also, the effect of some important design parameters are examined for the experimental and plant conditions. Using the flow models and correlations are integrated into a computer program, TPNCIRC, a number of correlations have been examined. This seems coming from the differences of void fractions

  1. Thermal-hydraulics of wave propagation and pressure distribution under hypothetical steam explosion conditions in the ANS reactor

    Energy Technology Data Exchange (ETDEWEB)

    Taleyarkhan, R.P.; Georgevich, V.; N-Valenit, S.; Kim, S.H. [Oak Ridge National Lab., TN (United States)

    1995-09-01

    This paper describes salient aspects of the modeling and analysis framework for evaluation of dynamic loads, wave propagation, and pressure distributions (under hypothetical steam explosion conditions) around key structural boundaries of the Advanced Neutron Source (ANS) reactor core region. A staged approach was followed, using simple thermodynamic models for bounding loads and the CTH code for evaluating realistic estimates in a staged multidimensional framework. Effects of nodalization, melt dispersal into coolant during explosion, single versus multidirectional dissipation, energy level of melt, and rate of energy deposition into coolant were studied. The importance of capturing multidimensional effects that simultaneously account for fluid-structural interactions was demonstrated. As opposed to using bounding loads from thermodynamic evaluations, it was revealed that the ANS reactor system will not be vulnerable to vertically generated missiles that threaten containment if realistic estimates of energetics are used (from CTH calculations for thermally generated steam explosions without significant aluminum ignition).

  2. Biological Phosphorus Release and Uptake Under Alternating Anaerobic and Anoxic Conditions In a Fixed-Film Reactor

    DEFF Research Database (Denmark)

    Kerrn-Jespersen, Jens Peter; Henze, Mogens; Strube, Rune

    1994-01-01

    Biological phosphorus removal was investigated in a fixed-film reactor with alternating anaerobic and anoxic conditions. The tests showed that biological phosphorus removal can be obtained in a fixed-film reactor with nitrate as oxidising agent. In the anaerobic period, 0.52 mg of PO4-P...... was released per mg of acetate taken up on an average. In the anoxic period, 2.0 mg of PO4-P was taken up per mg of NO3-N reduced on an average. The relationship between potassium released and phosphate released in the anaerobic phase was determined to be 0.37 mg K/mg P, while the relationship between...

  3. Magnetic studies of archaeological obsidian: Variability of eruptive conditions within obsidian flows is key to high-resolution artifact sourcing (Invited)

    Science.gov (United States)

    Feinberg, J. M.; Frahm, E.; Muth, M.

    2013-12-01

    Previous studies have endeavored to use petrophysical traits of obsidian, particularly its magnetic properties, as an alternative to conventional geochemical sourcing, one of the greatest successes in archaeological science. Magnetic approaches, however, have not seen widespread application due to their mixed success. In a time when geochemical analyses can be conducted non-destructively, in the field, and in a minute or two, magnetic measurements of obsidian must offer novel archaeological insights to be worthwhile, not merely act as a less successful version of geochemistry. To this end, we report the findings of a large-scale study of obsidian magnetism, which includes 912 geological obsidian specimens and 97 artifacts measured for six simple magnetic parameters. Based on these results, we propose, rather than using magnetic properties to source artifacts to a particular obsidian flow (inter-flow sourcing), these properties are best used to differentiate quarrying sites within an individual flow (intra-flow sourcing). The magnetic properties within an individual flow are highly variable, due to the fact that a single flow experiences a wide array of cooling rates, absolute temperatures, viscosities, deformation, and oxidation. These conditions affect the concentrations, compositions, size distributions, shapes, and spatial arrangements of magnetic grains within an obsidian specimen and, thus, its intrinsic magnetic properties. This variability decreases dramatically at spatial scales of individual outcrops, and decreases even further at scales of hand samples. Thus, magnetic data appear to shift the scale of obsidian sourcing from flows to quarries and, in turn, enable new insights into raw-material procurement strategies, group mobility, lithic technology, and the organization of space and production. From a geologic perspective, the magnetic variability of obsidian can be broadly interpreted within the context of the igneous processes that were active during

  4. Fatigue crack growth characteristics of nitrogen-alloyed type 347 stainless under operating conditions of a pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Min, Ki Deuk; Hong, Seok Min; Kim, Dae Whan; Lee, Bong Sang [Korea Atomic Energy Research Institute, Nuclear Materials Safety Research Division, Daejeon (Korea, Republic of); Kim, Seon Jin [Hanyang University, Division of materials science and engineering, Seoul (Korea, Republic of)

    2017-06-15

    The fatigue crack growth behavior of Type 347 (S347) and Type 347N (S347N) stainless steel was evaluated under the operating conditions of a pressurized water reactor (PWR). These two materials showed different fatigue crack growth rates (FCGRs) according to the changes in dissolved oxygen content and frequency. Under the simulated PWR conditions for normal operation, the FCGR of S347N was lower than that of S347 and insensitive to the changes in PWR water conditions. The higher yield strength and better corrosion resistance of the nitrogen-alloyed Type 347 stainless steel might be a main cause of slower FCGR and more stable properties against changes in environmental conditions.

  5. A new shape reproduction method based on the Cauchy-condition surface for real-time tokamak reactor control

    International Nuclear Information System (INIS)

    Kurihara, K.

    2000-01-01

    A new shape reproduction method is investigated on the basis of an applied mathematical approach. An analytically exact solution of Maxwell's equations in a static current field yields an (boundary) integral equation. In application of this equation to tokamak plasma shape reproduction, it is made clear that a Cauchy condition (both Dirichlet and Neumann conditions) on a hypothetical surface is necessarily identified. To calculate the Cauchy condition using magnetic sensor signals, conversion to numerical formulation of this method is conducted. Then, reproduction errors by this method are evaluated through two numerical tests: The first test uses ideal signals produced from a full equilibrium code in the JT-60 geometry, and the second test uses actual sensor signals in JT-60 experiments. In addition, it is shown that positioning and shape of the Cauchy condition surface is insensitive to reproduction error. Finally, this method is clarified to have preferable features for real-time tokamak reactor control

  6. Modelling of plate-out under gas-cooled reactor (GCR) accident conditions

    International Nuclear Information System (INIS)

    Taig, A.R.

    1981-01-01

    The importance of plate-out in mitigating consequences of gas-cooled reactor accidents, and its place in assessing these consequences, are discussed. The data requirements of a plate-out modelling program are discussed, and a brief description is given of parallel work programs on thermal/hydraulic reactor behaviour and fuel modelling, both of which will provide inputs to the plate-out program under development. The representation of a GCR system used in SRD studies is presented, and the equations governing iodine adsorption, desorption and transport round the circuit are derived. The status of SRD's plate-out program is described, and the type of sensitivity studies to be undertaken with the partially-developed computer program in order to identify the most useful lines for future research is discussed. (author)

  7. VICTORIA: A mechanistic model of radionuclide behavior in the reactor coolant system under severe accident conditions

    International Nuclear Information System (INIS)

    Heams, T.J.; Williams, D.A.; Johns, N.A.; Mason, A.; Bixler, N.E.; Grimley, A.J.; Wheatley, C.J.; Dickson, L.W.; Osborn-Lee, I.; Domagala, P.; Zawadzki, S.; Rest, J.; Alexander, C.A.; Lee, R.Y.

    1992-12-01

    The VICTORIA model of radionuclide behavior in the reactor coolant system (RCS) of a light water reactor during a severe accident is described. It has been developed by the USNRC to define the radionuclide phenomena and processes that must be considered in systems-level models used for integrated analyses of severe accident source terms. The VICTORIA code, based upon this model, predicts fission product release from the fuel, chemical reactions involving fission products, vapor and aerosol behavior, and fission product decay heating. Also included is a detailed description of how the model is implemented in VICTORIA, the numerical algorithms used, and the correlations and thermochemical data necessary for determining a solution. A description of the code structure, input and output, and a sample problem are provided

  8. Radiation conditions at the training IRT-2000 and IR-100 reactors

    International Nuclear Information System (INIS)

    Fedorin, Eh.V.; Bronshtejn, I.Eh; Martynov, Yu.N.; Chistyakov, N.I.

    1978-01-01

    The experience is reviewed of radiation hygiene surveys and radiation safety provision during instructional processes on two training and research nuclear reactors of the IRT-2000 type (No. 1 and No. 2) and on an IR-200 reactor. From an analysis of individual dosimetry data the conclusion is made that the trainees and personnel are exposed mainly to external gamma-radiation and also, to a minor degree, to thermal neutrons and beta-radiation. It has been found that a high level of radiation safety is ensured on the training and research so that research and instruction activities are conducted at annual levels of exposure substantially lower than 0.5 rem in the case of trainees and 5 rem in the case of personnel

  9. An Invitation to Mathematics

    CERN Document Server

    Schleicher, Dierk

    2011-01-01

    This "Invitation to Mathematics" consists of 14 contributions, many from the world's leading mathematicians, which introduce the readers to exciting aspects of current mathematical research. The contributions are as varied as the personalities of active mathematicians, but together they show mathematics as a rich and lively field of research. The contributions are written for interested students at the age of transition between high school and university who know high school mathematics and perhaps competition mathematics and who want to find out what current research mathematics is

  10. Thermofluid experiments for Fusion Reactor Safety. Visualization of exchange flows through breaches of a vacuum vessel in a fusion reactor under the LOVA condition

    International Nuclear Information System (INIS)

    Fujii, Sadao; Shibazaki, Hiroaki; Takase, Kazuyuki; Kunugi, Tomoaki.

    1997-01-01

    Exchange flow rates through breaches of a vacuum vessel in a fusion reactor under the LOVA (Loss of VAcuum event) conditions were measured quantitatively by using a preliminary LOVA apparatus and exchange flow patterns over the breach were visualized qualitatively by smoke. Velocity distributions in the exchange flows were predicted from the observed flow patterns by using the correlation method in the flow visualization procedures. Mean velocities calculated from the predicted velocity distributions at the outside of the breach were in good agreement with the LOVA experimental results when the exchange flow velocities were low. It was found that the present flow visualization and the image processing system might be an useful procedure to evaluate the exchange flow rates. (author)

  11. Development, application and also modern condition of the calculated program Imitator of a reactor

    International Nuclear Information System (INIS)

    Aver'yanova, S.P.; Kovel', A.I.; Mamichev, V.V.; Filimonov, P.E.

    2008-01-01

    Features of the calculated program Imitator of a reactor (IR) for WWER-1000 operation simulation are discussed. It is noted that IR application at NPP provides for the project program (BIPR-7) on-line working. This offers a new means, on the one hand, for the efficient prediction and information support of operator, on the other hand, for the verification and development of calculated scheme and neutron-physical model of the WWER-1000 projection program [ru

  12. Experimental conditions for determination of the neutrino mass hierarchy with reactor antineutrinos

    Directory of Open Access Journals (Sweden)

    Myoung Youl Pac

    2016-01-01

    Full Text Available This article reports the optimized experimental requirements to determine neutrino mass hierarchy using electron antineutrinos (ν¯e generated in a nuclear reactor. The features of the neutrino mass hierarchy can be extracted from the |Δm312| and |Δm322| oscillations by applying the Fourier sine and cosine transforms to the L/E spectrum. To determine the neutrino mass hierarchy above 90% probability, the requirements on the energy resolution as a function of the baseline are studied at sin2⁡2θ13=0.1. If the energy resolution of the neutrino detector is less than 0.04/Eν and the determination probability obtained from Bayes' theorem is above 90%, the detector needs to be located around 48–53 km from the reactor(s to measure the energy spectrum of ν¯e. These results will be helpful for setting up an experiment to determine the neutrino mass hierarchy, which is an important problem in neutrino physics.

  13. Molecular analysis of the biomass of a fluidized bed reactor treating synthetic vinasse at anaerobic and micro-aerobic conditions.

    Science.gov (United States)

    Rodríguez, Elisa; Lopes, Alexandre; Fdz-Polanco, María; Stams, Alfons J M; García-Encina, Pedro A

    2012-03-01

    The microbial communities (Bacteria and Archaea) established in an anaerobic fluidized bed reactor used to treat synthetic vinasse (betaine, glucose, acetate, propionate, and butyrate) were characterized by denaturing gradient gel electrophoresis (DGGE) and phylogenetic analysis. This study was focused on the competitive and syntrophic interactions between the different microbial groups at varying influent substrate to sulfate ratios of 8, 4, and 2 and anaerobic or micro-aerobic conditions. Acetogens detected along the anaerobic phases at substrate to sulfate ratios of 8 and 4 seemed to be mainly involved in the fermentation of glucose and betaine, but they were substituted by other sugar or betaine degraders after oxygen application. Typical fatty acid degraders that grow in syntrophy with methanogens were not detected during the entire reactor run. Likely, sugar and betaine degraders outnumbered them in the DGGE analysis. The detected sulfate-reducing bacteria (SRB) belonged to the hydrogen-utilizing Desulfovibrio. The introduction of oxygen led to the formation of elemental sulfur (S(0)) and probably other sulfur compounds by sulfide-oxidizing bacteria (γ-Proteobacteria). It is likely that the sulfur intermediates produced from sulfide oxidation were used by SRB and other microorganisms as electron acceptors, as was supported by the detection of the sulfur respiring Wolinella succinogenes. Within the Archaea population, members of Methanomethylovorans and Methanosaeta were detected throughout the entire reactor operation. Hydrogenotrophic methanogens mainly belonging to the genus Methanobacterium were detected at the highest substrate to sulfate ratio but rapidly disappeared by increasing the sulfate concentration.

  14. Evaluation of the RSG-GAS Alpha-Beta Aerosol Contaminant Monitor Performance Under Reactor Operation Condition

    International Nuclear Information System (INIS)

    Hartoyo, Unggul; Setiawanto, Anto; Sumarno, Yulius

    2000-01-01

    Analysis to evaluate the RSG-GAS alpha-beta aerosol contaminant monitor performance was done. The high potential radiation working area such as in RSG-GAS is important to monitored for personal safety. Further it is necessary to assure that the system monitor is reliable enough under normal conditions as well as emergency condition. The method uses in this analysis are monitoring and comparing with the standard source. The standard course indicator and panel in main control room indicate that the result is 1 x 110 exp 9 Ci/m exp 3. Based on data monitor observation, the RSG-GAS alpha-beta aerosol contaminant monitor system under reactor operation condition has a good enough performance

  15. Influence of coolant pH on corrosion of 6061 aluminum under reactor heat transfer conditions

    International Nuclear Information System (INIS)

    Pawel, S.J.; Felde, D.K.; Pawel, R.E.

    1995-10-01

    To support the design of the Advanced Neutron Source (ANS), an experimental program was conducted wherein aluminum alloy specimens were exposed at high heat fluxes to high-velocity aqueous coolants in a corrosion test loop. The aluminum alloys selected for exposure were candidate fuel cladding materials, and the loop system was constructed to emulate the primary coolant system for the proposed ANS reactor. One major result of this program has been the generation of an experimental database defining oxide film growth on 6061 aluminum alloy cladding. Additionally, a data correlation was developed from the database to permit the prediction of film growth for any reasonable thermal-hydraulic excursion. This capability was utilized effectively during the conceptual design stages of the reactor. During the course of this research, it became clear that the kinetics of film growth on the aluminum alloy specimens were sensitively dependent on the chemistry of the aqueous coolant and that relatively small deviations from the intended pH 5 operational level resulted in unexpectedly large changes in the corrosion behavior. Examination of the kinetic influences and the details of the film morphology suggested that a mechanism involving mass transport from other parts of the test loop was involved. Such a mechanism would also be expected to be active in the operating reactor. This report emphasizes the results of experiments that best illustrate the influence of the nonthermal-hydraulic parameters on film growth and presents data to show that comparatively small variations in pH near 5.0 invoke a sensitive response. Simply, for operation in the temperature and heat flux range appropriate for the ANS studies, coolant pH levels from 4.5 to 4.9 produced significantly less film growth than those from pH 5.1 to 6. A mechanism for this behavior based on the concept of treating the entire loop as an active corrosion system is presented

  16. Program requirements to determine and relate fuel damage and failure thresholds to anticipated conditions in pressurized water reactors

    International Nuclear Information System (INIS)

    Loyd, R.F.; Croucher, D.W.

    1980-03-01

    Anticipated transients, licensing criteria, and damage mechanisms for PWR fuel rods are reviewed. Potential mechanistic fuel rod damage limits for PWRs are discussed. An expermental program to be conducted out-of-pile and in the Engineering Test Reactor (ETR) to generate a safety data base to define mechanistic fuel damage and failure thresholds and to relate these thresholds to the thermal-hydraulic and power conditions in a PWR is proposed. The requirements for performing the tests are outlined. Analytical support requirements are defined

  17. Electrochemical measurements and modeling predictions in boiling water reactors under various operating conditions

    International Nuclear Information System (INIS)

    Indig, M.E.

    1991-01-01

    One important issue for providing life extension to operating boiling water nuclear reactors (BWRs) is the control of stress corrosion cracking in all sections of the primary coolant circuit. This paper links experimental and theoretical methods that provide understanding and measurements of the critical parameter, the electrochemical potential (ECP), and its application to determining crack growth rate among and within the family of BWRs. Measurement of in-core ECP required the development of a new family of radiation-resistant sensors. With these sensors, ECPs were measured in the core and piping of two operating BWRs. Concurrent crack growth measurements were used to benchmark a crack growth prediction algorithm with measured ECPs

  18. Study on light water reactor fuel behavior under reactivity initiated accident condition in TREAT

    International Nuclear Information System (INIS)

    Ohnishi, Nobuaki; Ishijima, Kiyomi; Ochiai, Masaaki; Tanzawa, Sadamitsu; Uemura, Mutsumi

    1981-05-01

    This report reviews the results of the fuel failure experiments performed in TREAT in the U.S.A. simulating Reactivity Initiated Accidents. One of the main purposes of the TREAT experiments is the study of the fuel failure behavior, and the other is the study of the molten fuel-water coolant interaction and the consequent hydrogen behavior. This report mainly shows the results of the TREAT experiments studying the fuel failure behavior in Light Water Reactor, and then it describes the fuel failure threshold and the fuel failure mechanism, considering the results of the photographic experiments of the fuel failure behavior with transparent capsules. (author)

  19. Combined conditioning in the high-temperature experimental nuclear reactor (AVR) at Juelich

    International Nuclear Information System (INIS)

    Nieder, R.; Vey, K.; Ivens, G.

    1984-01-01

    The high temperature experimental nuclear reactor (AVR) is the first nuclear power plant in which combined cycle operation has been introduced. The water-steam cycle has been operated for about 15 years according to the alkali method of working with ammonia and hydrazine. The VGB-guidelines have been adhered to througout. Since January 1983 cobined cycle operation has been employed, and in this process a pH-value of about 8.5 and an oxygen concentration of about 200 μg/kg in the feedwater have been used. A distinct reduction of tritium concentration in the water-steam cycle was the outstanding new result. (orig.) [de

  20. Flow patterns in a slurry-bubble-column reactor under reaction conditions

    Energy Technology Data Exchange (ETDEWEB)

    Toselane, B.A.; Brown, D.M.; Zou, B.S.; Dudukovic, M.P. [Washington Univ., St. Louis, MO (United States)

    1995-12-31

    The gas and liquid radioactive tracer response curves obtained in an industrial bubble column reactor of height to diameter ratio of 10 are analyzed and the suitability of the axial dispersion model for interpretation of the results is discussed. The relationship between the tracer concentration distribution and measured detector response of the soluble gas tracer (Ar-41) is possibly dominated by the dissolved gas. The one dimensional axial dispersion model cannot match all the experimental observations well and the flow pattern of the undissolved gas cannot be determined with certainty.

  1. By invitation only

    CERN Multimedia

    Antonella Del Rosso

    2011-01-01

    The Hôtel Métropole in Brussels is a legendary conference venue that will ring a bell to all physicists. 100 years after the first meeting of the Conseil de Physique Solvay, a representative group of the world’s most eminent physicists met again in October this year in the Excelsior Room to discuss “The Theory of the Quantum World”. Three members of CERN's Theory Group were invited to participate. Gian Giudice, one of them, shares with us his thoughts and impressions about this exclusive conference.   Attendees at the 25th Solvay Conference on Physics, Brussels. Photo provided by "International Solvay Institutes". “In the tradition of the Solvay Conferences, this is a discussion-oriented meeting with few talks by rapporteurs.” Thus reads the Introduction to the Scientific Programme of the Solvay Conference on Physics. In the Conference programme, the rapporteurs speak for only 30 minutes,...

  2. The ReactorSTM: Atomically resolved scanning tunneling microscopy under high-pressure, high-temperature catalytic reaction conditions

    Energy Technology Data Exchange (ETDEWEB)

    Herbschleb, C. T.; Tuijn, P. C. van der; Roobol, S. B.; Navarro, V.; Bakker, J. W.; Liu, Q.; Stoltz, D.; Cañas-Ventura, M. E.; Verdoes, G.; Spronsen, M. A. van; Bergman, M.; Crama, L.; Taminiau, I.; Frenken, J. W. M., E-mail: frenken@physics.leidenuniv.nl [Huygens-Kamerlingh Onnes Laboratory, Leiden University, P.O. box 9504, 2300 RA Leiden (Netherlands); Ofitserov, A.; Baarle, G. J. C. van [Leiden Probe Microscopy B.V., J.H. Oortweg 21, 2333 CH Leiden (Netherlands)

    2014-08-15

    To enable atomic-scale observations of model catalysts under conditions approaching those used by the chemical industry, we have developed a second generation, high-pressure, high-temperature scanning tunneling microscope (STM): the ReactorSTM. It consists of a compact STM scanner, of which the tip extends into a 0.5 ml reactor flow-cell, that is housed in a ultra-high vacuum (UHV) system. The STM can be operated from UHV to 6 bars and from room temperature up to 600 K. A gas mixing and analysis system optimized for fast response times allows us to directly correlate the surface structure observed by STM with reactivity measurements from a mass spectrometer. The in situ STM experiments can be combined with ex situ UHV sample preparation and analysis techniques, including ion bombardment, thin film deposition, low-energy electron diffraction and x-ray photoelectron spectroscopy. The performance of the instrument is demonstrated by atomically resolved images of Au(111) and atom-row resolution on Pt(110), both under high-pressure and high-temperature conditions.

  3. Albedo boundary conditions for global calculations of thermal nuclear reactors with the model of discrete ordinates to two energy groups

    International Nuclear Information System (INIS)

    Nunes, Carlos Eduardo de Araujo

    2011-01-01

    As neutron fission events do not take place in the non-multiplying regions of nuclear reactors, e.g., moderator, reflector, and structural core, these regions do not generate power and the computational efficiency of nuclear reactor global calculations can hence be improved by eliminating the explicit numerical calculations within the non-multiplying regions around the active domain. Discussed here is the computational efficiency of approximate discrete ordinates (SN) albedo boundary conditions for two-energy group eigenvalue problems in X, Y geometry. Albedo, the Latin word for w hiteness , was originally defined as the fraction of incident light reflected diffusely by a surface. This Latin word has remained the usual scientific term in astronomy and in this dissertation this concept is extended for the reflection of neutrons. The non-standard SN albedo substitutes approximately the reflector region around the active domain, as we neglect the transverse leakage terms within the non-multiplying reflector. Should the problem have no transverse leakage terms, i.e., one dimensional slab geometry, then the offered albedo boundary conditions are exact. By computational efficiency we mean analyzing the accuracy of the numerical results versus the CPU execution time of each run for a given model problem. Numerical results to two 1/4 symmetric test problems are shown to illustrate this efficiency analysis. (author)

  4. Eastern Sources of Invitational Education.

    Science.gov (United States)

    Ryback, David

    1993-01-01

    Presents historical perspective suggesting that invitational theory shares many beliefs with ancient Eastern philosophies. Submits that teachers and other educators who embrace the invitational perspective may benefit from an understanding of Eastern principles. Briefly describes Buddhism, Confucianism, Taoism, Zen Buddhism, and their relevance to…

  5. Radiolysis of cesium iodide solutions in conditions prevailing in a pressurized water reactor severe accident

    International Nuclear Information System (INIS)

    Lucas, M.

    1988-01-01

    Measurements were made of I/sub 2/ formed when aqueous cesium iodide (CsI) solutions were exposed to two temperatures, 43 and 95 0 C, with irradiation. Iodine partition coefficients were obtained from the experiments. The parameters varied were dose, CsI concentration, and Cs/sub 2/CO/sub 3/ concentration, in the presence of air-carbon dioxide and air-carbon dioxide-hydrogen mixtures, to provide information to calculate the form in which iodine released from fuel as CsI in a reactor accident might reach the environment. In a series of experiments, a two-compartment cell was used to trap the gaseous iodine produced. In this case, it was found that the quantity of gaseous iodine produced increased approximately linearly with the dose (at the dose rate used)

  6. The study of flow resistance in nuclear reactor Maria under coolant boiling condition

    International Nuclear Information System (INIS)

    Czerski, P.

    1999-01-01

    This study describes an analysis of experiments carried out in the WIW-300 installation located in the Institute of Atomic Energy (Swierk, Poland). The flow, simulated in the annular gap of test section, was similar to the flow in Maria reactor fuel channel. Experimental character of the work lead to the conclusions related to the physical nature of the hydrodynamic phenomena investigated as well as to the practical aspects of future research. A hypothesis defining a cause of pressure changes was formulated and specific problems related to the mathematical model were defined. The analysis shows that hydrodynamic phenomena studies are of basic significance for the prediction of burnout effects and that heat exchange is very often determined by local phenomena. All described observations are the base for further research on thermodynamic aspects of investigated phenomena. (author)

  7. Measurements of combined neutron and photon fluxes for the accurate characterization of the future Jules Horowitz irradiation reactor experimental conditions

    International Nuclear Information System (INIS)

    Fourmentel, D.

    2013-01-01

    A new Material Testing Reactor (MTR), the Jules Horowitz Reactor (JHR), is under construction at the CEA Cadarache (French Alternatives Energies and Atomic Energy Commission). From 2016 this new MTR will be a new facility for the nuclear research on materials and fuels. The quality of the experiments to be conducted in this reactor is largely linked to the good knowledge of the irradiation conditions. Since 2009, a new research program called IN-CORE1 'Instrumentation for Nuclear radiations and Calorimetry Online in Reactor' is under progress between CEA and Aix-Marseille University in the framework of a joint laboratory LIMMEX2. This program aims to improve knowledge of the neutron and photon fluxes in the RJH core, with one hand, an innovative instrumentation performing mapping of experimental locations, and on the other hand by coupling neutron flux, photon flux and thermal measurements. Neutron flux expected in the JHR core is about 10 15 n.cm -2 .s -1 and nuclear heating up to 20 W.g -1 for a nominal power of 100 MWth. One of the challenges is to identify sensors able to measure such fluxes in JHR experimental conditions and to determine how to analyse the signals delivered by these sensors with the most appropriate methods. The thesis is part of this ambitious program and aims to study the potential and the interest of the combination of radiation measurements in the prospect of a better assessment of the levels of neutron flux, gamma radiation and nuclear heating in the JHR experimental locations. The first step of IN-CORE program is to develop and operate an instrumented device called CARMEN-1 adapted to the mapping of the OSIRIS reactor, then to develop a second version called CARMEN-2 dedicated to experiments in the JHR core, especially for its start-up. This experiment was the opportunity to test all the radiation sensors which could meet the needs of JHR, including recently developed sensors. Reference neutron measurements are performed by activation

  8. Calibration of hydrodynamic behavior and biokinetics for TOC removal modeling in biofilm reactors under different hydraulic conditions.

    Science.gov (United States)

    Zeng, Ming; Soric, Audrey; Roche, Nicolas

    2013-09-01

    In this study, total organic carbon (TOC) biodegradation was simulated by GPS-X software in biofilm reactors with carriers of plastic rings and glass beads under different hydraulic conditions. Hydrodynamic model by retention time distribution and biokinetic measurement by in-situ batch test served as two significant parts of model calibration. Experimental results showed that TOC removal efficiency was stable in both media due to the enough height of column, although the actual hydraulic volume changed during the variation of hydraulic condition. Simulated TOC removal efficiencies were close to experimental ones with low theil inequality coefficient values (below 0.15). Compared with glass beads, more TOC was removed in the filter with plastic rings due to the larger actual hydraulic volume and lower half saturation coefficient in spite of its lower maximum specific growth rate of biofilm, which highlighted the importance of calibrating hydrodynamic behavior and biokinetics. Copyright © 2013 Elsevier Ltd. All rights reserved.

  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)

    Pourcelot, L.

    1997-01-01

    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. Presentation and comparison of experimental critical heat flux data at conditions prototypical of light water small modular reactors

    Energy Technology Data Exchange (ETDEWEB)

    Greenwood, M.S., E-mail: 1greenwoodms@ornl.gov; Duarte, J.P.; Corradini, M.

    2017-06-15

    Highlights: • Low mass flux and moderate to high pressure CHF experimental results are presented. • Facility uses chopped-cosine heater profile in a 2 × 2 square bundle geometry. • The EPRI, CISE-GE, and W-3 CHF correlations provide reasonable average CHF prediction. • Neural network analysis predicts experimental data and demonstrates utility of method. - Abstract: The critical heat flux (CHF) is a two-phase flow phenomenon which rapidly decreases the efficiency of the heat transfer performance at a heated surface. This phenomenon is one of the limiting criteria in the design and operation of light water reactors. Deviations of operating parameters greatly alters the CHF condition and must be experimentally determined for any new parameters such as those proposed in small modular reactors (SMR) (e.g. moderate to high pressure and low mass fluxes). Current open literature provides too little data for functional use at the proposed conditions of prototypical SMRs. This paper presents a brief summary of CHF data acquired from an experimental facility at the University of Wisconsin-Madison designed and built to study CHF at high pressure and low mass flux ranges in a 2 × 2 chopped cosine rod bundle prototypical of conceptual SMR designs. The experimental CHF test inlet conditions range from pressures of 8–16 MPa, mass fluxes of 500–1600 kg/m2 s, and inlet water subcooling from 250 to 650 kJ/kg. The experimental data is also compared against several accepted prediction methods whose application ranges are most similar to the test conditions.

  11. France - About the possibility to stop the 19 French reactors which are over 30 years of service. Overview of conditions for phasing out nuclear under condition

    International Nuclear Information System (INIS)

    Lenoir, Yves

    2011-03-01

    After having set a 30 year limit for a nuclear reactor life, the author analyses the associated consequences in terms of loss of production (while taking the load factor of the concerned reactors into account). He indicates the French production of different resources (nuclear, conventional thermal, hydraulic, renewable) and the French energy consumption, assesses the avoided production due to the shutting down of the 19 reactors, outlines that the present reactor fleet is under-exploited, and addresses the peak issue

  12. Performance evaluation of a completely stirred anaerobic reactor treating pig manure at a low range of mesophilic conditions

    International Nuclear Information System (INIS)

    Guo, Jianbin; Dong, Renjie; Clemens, Joachim; Wang, Wei

    2013-01-01

    Highlights: • The biogas process can run stably at 20 °C at extremely low OLR after long-term acclimation of bacteria. • A biogas plant running at 28 °C seems as efficient as that operated at 38 °C at low OLR of 1.3 g ODM L −1 d −1 . • Lower temperature operation is inadvisable for the commercial biogas plant running at rather high OLR. • The estimated sludge yield at 28 °C is higher than that at 38 °C. - Abstract: Many Chinese biogas plants run in the lower range of mesophilic conditions. This study evaluated the performance of a completely stirred anaerobic reactor treating pig manure at different temperatures (20, 28 and 38 °C). The start-up phase of the reactor at 20 °C was very long and extremely poor performance was observed with increasing organic loading rate (OLR). At an OLR of 4.3 g ODM L −1 d −1 , methane production at 28 °C was comparable (3% less) with that at 38 °C, but the risk of acidification was high at 28 °C. At low OLR (1.3 g ODM L −1 d −1 ), the biogas process appeared stable at 28 °C and gave same methane yields as compared to the reactor operating at 38 °C. The estimated sludge yield at 28 °C was 0.065 g VSS g −1 COD removed, which was higher than that at 38 °C (0.016 g VSS g −1 COD removed )

  13. Atmospheric pressure flow reactor: Gas phase chemical kinetics under tropospheric conditions without wall effects

    Science.gov (United States)

    Koontz, Steven L. (Inventor); Davis, Dennis D. (Inventor)

    1991-01-01

    A flow reactor for simulating the interaction in the troposphere is set forth. A first reactant mixed with a carrier gas is delivered from a pump and flows through a duct having louvers therein. The louvers straighten out the flow, reduce turbulence and provide laminar flow discharge from the duct. A second reactant delivered from a source through a pump is input into the flowing stream, the second reactant being diffused through a plurality of small diffusion tubes to avoid disturbing the laminar flow. The commingled first and second reactants in the carrier gas are then directed along an elongated duct where the walls are spaced away from the flow of reactants to avoid wall interference, disturbance or turbulence arising from the walls. A probe connected with a measuring device can be inserted through various sampling ports in the second duct to complete measurements of the first and second reactants and the product of their reaction at selected XYZ locations relative to the flowing system.

  14. CONCIDERATION OF FOUNDATION AND SEISMIC CONDITIONS OF AREA IN ANALYSIS OF SEISMIC RESISTANCE OF REACTOR COMPARTMENT

    Directory of Open Access Journals (Sweden)

    SEDIN V. L.

    2015-11-01

    Full Text Available Problem statement. Providing of safe exploitation of nuclear power plants, as well as a safety of staff and environment is a very important problem. A distinct feature of this problem is a necessity to provide not only a strength of structures, but also a safe functioning of all systems that control nuclear process. In particular, the influence of earthquake should be considered on constructions of buildings and structures of nuclear and thermal power plant, taking into account soil-structure interaction. According to IAEA’s SSD-9 recommendations, a risk of vibration of soil should be analyzed for each NPP connected with earthquakes soil that means researches, including general, detailed and microseismic zoning of the area works. One of the distinctive features of the considered problem is an evaluation of the seismicity of area and getting the response spectrum on the free surface. Purpose. Determination of seismic resistance of buildings of high category of safety with the example of the reactor compartment of Zaporoghskaya NPP including the soil structure interaction. Conclusion The seismicity assessment of the area and obtaining of response specters on free surface was made during research and analysis of seismic resistance of buildings of high category of safety including the effects of foundation and structures. The method of modeling of the equivalent dynamic characteristics of the base was considered during the research in seismic impacts.

  15. Results from integral tests of single reformer tubes under simulated nuclear reactor conditions

    International Nuclear Information System (INIS)

    Decken, C.B. von der; Fedders, H.; Harth, R.; Hoehlein, B.; Riensche, E.

    1980-01-01

    The possibility of supplying high temperature heat from a HTGR for process application is being investigated at some places in the world. In all programmes or projects existing with respect to this application, the endothermic steam reforming of methane is one main step in the transmission of heat produced by nuclear fission to different chemical processes. The KFA is involved in the two German projects PNP - Prototypanlage Nukleare Prozesswaerme (Prototype-plant Nuclear Process-heat), and NFE -Nukleare Fernenergie (Long Distance Energy Transport). In a HTGR, helium generally serves as reactor coolant. It transports the heat from the core to the different components which take over this heat for various purposes. In case of arranging a steam reformer in the helium circuit, it is necessary for economic reasons to reach very high temperatures. In the two German projects mentioned above, the helium temperature at HTGR core outlet is determined to 950 0 C. Thus the main design data for a steam reformer supplied by heat from a HTGR are maximum helium temperature 950 0 C, helium pressure 40 bar. By an extensive utilization of the available advanced conventional steam reforming technology, the helium heated steam reformer design is using normal steam reforming tubes arranged in compact bundles

  16. Functional properties of poly(tetrafluoroethylene) (PTFE) gasket working in nuclear reactor conditions

    Science.gov (United States)

    Wyszkowska, Edyta; Leśniak, Magdalena; Kurpaska, Lukasz; Prokopowicz, Rafal; Jozwik, Iwona; Sitarz, Maciej; Jagielski, Jacek

    2018-04-01

    In this study structural and nanomechanical properties of polytetrafluoroethylene (PTFE) used as a gasket in the nuclear reactor have been deeply investigated. In order to reveal structural changes caused by long-term pressure, temperature and irradiation (possibly neutron and gamma), methods such as SEM, X-ray diffraction and Raman Spectroscopy have been used. Nanomechanical properties such as Young Modulus and hardness were investigated by means of the nanoindentation technique. Presented study confirmed the influence of working (radiative) environment on the functional properties of PTFE. The results of Raman spectroscopy and X-ray diffraction techniques revealed shift of the major band positions and band intensities increase. Moreover, changes of hardness and Young Modulus values of the irradiated material with respect to the virgin specimen have been recorded. This phenomenon can be attributed to the modifications in crystallinity of the material. Presented work suggest that morphology of the irradiated material altered from well-ordered parallel fibers to more dense and thicker ones.

  17. INVITATION FROM THE UBS

    CERN Multimedia

    UBS SA

    2000-01-01

    Via its Web pages (http://www.ubs.com), the UBS offers you direct access to all its telebanking services, namely:information relating to accounts and custody accounts (balances, debits and credits);stock market information and placing of orders;domestic and foreign payment transactions.These supplementary and inexpensive banking services mean that you no longer have to go to the bank to obtain the information you require and are not therefore limited by bank opening times.The UBS will be holding two sessions on the CERN site to present this modern product on:25.4.2000 (in French),26.4.2000 (in English).The agenda will be as follows:12.00 to 12.40 p.m.Presentation of the product;12.40 to 13.10 p.m.Question-and-answer session;13.10 to 14.00 p.m.Buffet and discussions.To enable us to meet your requirements and to expedite matters, we invite you to register for one of these sessions by sending an e-mail or letter to:pierre.guyenon@ubs.com before 7.4.2000 indicating:your surname and first name, the account number(...

  18. Flow Components in a NaK Test Loop Designed to Simulate Conditions in a Nuclear Surface Power Reactor

    Science.gov (United States)

    Polzin, Kurt A.; Godfroy, Thomas J.

    2008-01-01

    A test loop using NaK as the working fluid is presently in use to study material compatibility effects on various components that comprise a possible nuclear reactor design for use on the lunar surface. A DC electromagnetic (EM) pump has been designed and implemented as a means of actively controlling the NaK flow rate through the system and an EM flow sensor is employed to monitor the developed flow rate. These components allow for the matching of the flow rate conditions in test loops with those that would be found in a full-scale surface-power reactor. The design and operating characteristics of the EM pump and flow sensor are presented. In the EM pump, current is applied to a set of electrodes to produce a Lorentz body force in the fluid. A measurement of the induced voltage (back-EMF) in the flow sensor provides the means of monitoring flow rate. Both components are compact, employing high magnetic field strength neodymium magnets thermally coupled to a water-cooled housing. A vacuum gap limits the heat transferred from the high temperature NaK tube to the magnets and a magnetically-permeable material completes the magnetic circuit. The pump is designed to produce a pressure rise of 5 psi, and the flow sensor's predicted output is roughly 20 mV at the loop's nominal flow rate of 0.5 GPM.

  19. Heat transfer tests conducted on full-scale model, to investigate cooling conditions of EL.3 experimental reactor

    International Nuclear Information System (INIS)

    Raievski, R.; Bousquet, M.; Braudeau, M.; Milliat, M.

    1958-01-01

    For such high heat flux density as is released in the channels of EL3 reactor (2.10 6 kcal/m 2 h on the hottest point) cooling conditions have proved to be satisfactory, that is free from nucleate boiling. The arrangements provided for these tests and the technique used for measurements (of temperature particularly) are specified. Two fields have been investigated: in the former (forced convection without nucleate boiling) a good agreement is found with Colburn's formula. The influence of the ratio L/D is pointed out. The latter field is of forced convection with beginning of nucleate boiling; there the observed raise of the transfer coefficient has been shown occurring with some delay. (author) [fr

  20. Development of empirical models for performance evaluation of UASB reactors treating poultry manure wastewater under different operational conditions

    International Nuclear Information System (INIS)

    Yetilmezsoy, Kaan; Sakar, Suleyman

    2008-01-01

    A nonlinear modeling study was carried out to evaluate the performance of UASB reactors treating poultry manure wastewater under different organic and hydraulic loading conditions. Two identical pilot scale up-flow anaerobic sludge blanket (UASB) reactors (15.7 L) were run at mesophilic conditions (30-35 deg. C) in a temperature-controlled environment with three hydraulic retention times (θ) of 15.7, 12 and 8.0 days. Imposed volumetric organic loading rates (L V ) ranged from 0.65 to 4.257 kg COD/(m 3 day). The pH of the feed varied between 6.68 and 7.82. The hydraulic loading rates (L H ) were controlled between 0.105 and 0.21 m 3 /(m 2 day). The daily biogas production rates ranged between 4.2 and 29.4 L/day. High volumetric COD removal rates (R V ) ranging from 0.546 to 3.779 kg COD removed /(m 3 day) were achieved. On the basis of experimental results, two empirical models having a satisfactory correlation coefficient of about 0.9954 and 0.9416 were developed to predict daily biogas production (Q g ) and effluent COD concentration (S e ), respectively. Findings of this modeling study showed that optimal COD removals ranging from 86.3% to 90.6% were predicted with HRTs of 7.9, 9.5, 11.2, 12.6, 13.7 and 14.3 days, and L V of 1.27, 1.58, 1.78, 1.99, 2.20 and 2.45 kg COD/(m 3 day) for the corresponding influent substrate concentrations (S i ) of 10,000, 15,000, 20,000, 25,000, 30,000 and 35,000 mg/L, respectively

  1. Optimization of operation conditions for the startup of aerobic granular sludge reactors biologically removing carbon, nitrogen, and phosphorous.

    Science.gov (United States)

    Lochmatter, Samuel; Holliger, Christof

    2014-08-01

    The transformation of conventional flocculent sludge to aerobic granular sludge (AGS) biologically removing carbon, nitrogen and phosphorus (COD, N, P) is still a main challenge in startup of AGS sequencing batch reactors (AGS-SBRs). On the one hand a rapid granulation is desired, on the other hand good biological nutrient removal capacities have to be maintained. So far, several operation parameters have been studied separately, which makes it difficult to compare their impacts. We investigated seven operation parameters in parallel by applying a Plackett-Burman experimental design approach with the aim to propose an optimized startup strategy. Five out of the seven tested parameters had a significant impact on the startup duration. The conditions identified to allow a rapid startup of AGS-SBRs with good nutrient removal performances were (i) alternation of high and low dissolved oxygen phases during aeration, (ii) a settling strategy avoiding too high biomass washout during the first weeks of reactor operation, (iii) adaptation of the contaminant load in the early stage of the startup in order to ensure that all soluble COD was consumed before the beginning of the aeration phase, (iv) a temperature of 20 °C, and (v) a neutral pH. Under such conditions, it took less than 30 days to produce granular sludge with high removal performances for COD, N, and P. A control run using this optimized startup strategy produced again AGS with good nutrient removal performances within four weeks and the system was stable during the additional operation period of more than 50 days. Copyright © 2014 Elsevier Ltd. All rights reserved.

  2. INITIATORS AND TRIGGERING CONDITIONS FOR ADAPTIVE AUTOMATION IN ADVANCED SMALL MODULAR REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Katya L Le Blanc; Johanna h Oxstrand

    2014-04-01

    It is anticipated that Advanced Small Modular Reactors (AdvSMRs) will employ high degrees of automation. High levels of automation can enhance system performance, but often at the cost of reduced human performance. Automation can lead to human out-of the loop issues, unbalanced workload, complacency, and other problems if it is not designed properly. Researchers have proposed adaptive automation (defined as dynamic or flexible allocation of functions) as a way to get the benefits of higher levels of automation without the human performance costs. Adaptive automation has the potential to balance operator workload and enhance operator situation awareness by allocating functions to the operators in a way that is sensitive to overall workload and capabilities at the time of operation. However, there still a number of questions regarding how to effectively design adaptive automation to achieve that potential. One of those questions is related to how to initiate (or trigger) a shift in automation in order to provide maximal sensitivity to operator needs without introducing undesirable consequences (such as unpredictable mode changes). Several triggering mechanisms for shifts in adaptive automation have been proposed including: operator initiated, critical events, performance-based, physiological measurement, model-based, and hybrid methods. As part of a larger project to develop design guidance for human-automation collaboration in AdvSMRs, researchers at Idaho National Laboratory have investigated the effectiveness and applicability of each of these triggering mechanisms in the context of AdvSMR. Researchers reviewed the empirical literature on adaptive automation and assessed each triggering mechanism based on the human-system performance consequences of employing that mechanism. Researchers also assessed the practicality and feasibility of using the mechanism in the context of an AdvSMR control room. Results indicate that there are tradeoffs associated with each

  3. Characterization of noble metals deposits and oxides in conditions of BWR reactors

    International Nuclear Information System (INIS)

    Arganis J, C.R.; Aguilar T, J.A.; Contreras R, A.

    2008-01-01

    The oxides deposited on steel 304l under normal chemistry conditions (NWC) and hydrogen chemistry (HWC) with presence of Zn, being that the first ones present hexagonal oxides of Hematite and the second bipyramidal crystals possibly Magnetite with traces of Zn. Deposits of Pt on the oxidized surfaces under NWC conditions were obtained, being glasses from 2 to 4 μm and Pt-Rh deposits were obtained on the oxidized surfaces in presence of Zn, by its size its were not possible to observe them by scanning electron microscopy. The kinetics of the surfaces of Hematite and of the deposits of Pt it was measured by means of the Tafel extrapolation technique, being proven the catalytic effect of the Pt, even in sizes of 11 μm. (Author)

  4. The radiological consequences of notional accidental releases of radioactivity from fast breeder reactors: the influence of the meteorological conditions

    International Nuclear Information System (INIS)

    Hemming, C.R.; Hallam, J.; Kelly, G.N.

    1980-04-01

    The radiological consequences of a wide range of notional accidental releases from a 1300 MW(e) LMFBR (liquid Metal-cooled Fast Breeder Reactor) were assessed in a previous study by the National Radiological Protection Board. In the present study the sensitivity of the predicted consequences to the meteorological conditions was investigated. The influence of the wind direction, and hence the distribution of the exposed population, and of atmospheric stability at the time of the release are considered separately. Conclusions are reached on the precision required in specifying atmospheric stability conditions to estimate reliably the distribution of possible consequences following a given release of activity. The probability distributions of consequences for selected releases from two locations are evaluated taking account of the frequency distribution of wind direction and atmospheric stability at each location. The variation is large and demonstrates the importance of taking account of the whole range of meteorological conditions when assessing the risk presented by an accidental release of activity. (author)

  5. Evaluations of two-phase natural circulation flow induced in the reactor vessel annular gap under ERVC conditions

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Kwang Soon, E-mail: tomo@kaeri.re.kr [Korea Atomic Energy Research Institute, 1045 Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Cheung, Fan-Bill [The Pennsylvania State University, University Park, PA 16802 (United States); Park, Rae Joon; Kim, Sang Baik [Korea Atomic Energy Research Institute, 1045 Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)

    2012-12-15

    Highlights: Black-Right-Pointing-Pointer Two-phase natural circulation flow induced in insulation gap was investigated. Black-Right-Pointing-Pointer Half-scaled non-heating experiments were performed to evaluate flow behavior. Black-Right-Pointing-Pointer The loop-integrated momentum equation was formulated and solved asymptotically. Black-Right-Pointing-Pointer First-order approximate solution was obtained and agreed with experimental data. - Abstract: The process of two-phase natural circulation flow induced in the annular gap between the reactor vessel and the insulation under external reactor vessel cooling conditions was investigated experimentally and analytically in this study. HERMES-HALF experiments were performed to observe and quantify the induced two-phase natural circulation flow in the annular gap. A half-scaled non-heating experimental facility was designed by utilizing the results of a scaling analysis to simulate the APR1400 reactor and its insulation system. The behavior of the boiling-induced two-phase natural circulation flow in the annular gap was observed, and the liquid mass flow rates driven by the natural circulation loop and the void fraction distribution were measured. Direct flow visualization revealed that choking would occur under certain flow conditions in the minimum gap region near the shear keys. Specifically, large recirculation flows were observed in the minimum gap region for large air injection rates and small outlet areas. Under such conditions, the injected air could not pass through the minimum gap region, resulting in the occurrence of choking near the minimum gap with a periodical air back flow being generated. Therefore, a design modification of the minimum gap region needs to be done to facilitate steam venting and to prevent choking from occurring. To complement the HERMES-HALF experimental effort, an analytical study of the dependence of the induced natural circulation mass flow rate on the inlet area and the

  6. The behaviour of water-cooled reactor fuel rods in steady state and transient conditions

    International Nuclear Information System (INIS)

    Strupczewski, A.; Marks, P.

    1997-01-01

    In this report, the results of temperature field and filling gas pressure calculations by means of contemporary calculational models for a WWER-440 and WWER-1000 type fuel rod at low and high burnup operating under steady-state conditions are presented. A review of in-core temperature and pressure measurements for various types of LWR fuel is also included. Basing on calculational and collected measured data, the behaviour of fuel cladding during large and small break LOCA, is estimated with special emphasis on their oxidation and failure resistance. (author)

  7. Calculated and experimental definition of neutron-physical and temperature conditions of material testing in the SM reactor

    International Nuclear Information System (INIS)

    Toporova, V.G.; Pimenov, V.V.

    2004-01-01

    Full text: Reactor material science is one of the main scientific directions of the RIAR activities. Particularly, a wide range of materials and products testing under irradiation is performed in reactor facility SM (RF SM). To solve the tasks specified in the technical specification for an experiment, previously, the test conditions are chosen. At the minimum a space-energy distribution of neutrons and heating rate in the materials under test are important as well as temperature conditions of irradiation. The up-to-date software and libraries of nuclear data allow modeling of neutron-material interaction processes to a considerable degree of details and also obtaining a true neutron distribution by calculation methods. As a result of a great scope of work on verification, a calculation model, developed on the basis of a package of applied software MCU (option MCU-4/SM22) and analogue Monte-Carlo method, is widely used at RIAR. The MCU geometric module makes it possible to model the SM core and reflector in three-dimensional geometry with sufficient accuracy and to describe all elements of the channel structure and irradiation device with specimens. The calculation model of RF SM is tested using the results of activation experiments performed in its critical assembly, geometric parameters and structural materials of which correspond completely with the prototype. The difference in the calculated and experimental values is less than 2.5%. Possibilities of the calculated estimation of operating temperature conditions of absorbing elements under irradiation should be considered separately. As the conducted calculations and their analysis show, to define the fuel column temperature correctly, one needs reliable data on thermal-physical parameters of materials, especially ceramic ones, such as titanium, dysprosium or boron carbide. This is very important for boron carbide-absorbing elements for actually all their operation parameters (such as: gas release, swelling

  8. Analysis of passive moderator cooling system of Candu-6A reactor at emergency condition

    International Nuclear Information System (INIS)

    Umar, Efrizon; Subki, M. Hadid; Vecchiarelli, Jack

    2001-01-01

    Analysis of passive moderator cooling system subject to in-core LOCA with no emergency core cooling injection has been done. In this study, the new model of passive moderator system has been tested for emergency conditions and CATHENA code Mod-3.5b/Rev1 is used to calculate some parameters of this passive moderator cooling system. This result of simulation show that the proposed moderator cooling system have given satisfactory result, especially for the case with 0.7 m riser diameter and the number of heat exchanger tubes 8100. For PEWS tank containing 3000 m3 of light water initially at 30 0C and a 3641 m2 moderator heat exchanger, the average long-term heat removed rate balances the moderator heat load and the flow through the passive moderator loop remains stable for over 72 hours with no saturated boiling in the calandria and flow instabilities do not develop during long-term period

  9. A simplified dynamic analysis for reactor piping systems under blowdown conditions

    International Nuclear Information System (INIS)

    Chen, M.M.

    1975-01-01

    In the design of pipelines in a nuclear power plant for blowdown conditions, is it customary to conduct dynamic analysis of the piping system to obtain the responses and the resulting stresses. Calculations are repeated for each design modification in piping geometry or supporting system until the design codes are met. The numerical calculations are, in general, very costly and time consuming. Until now, there have been no simple means for calculating the dynamic responses for the design. The proposed method reduces the dynamic calculation to a quasi-static one, and can be beneficially used for the preliminary design. The method is followed by a complete dynamical analysis to improve the final results. The new formulations greatly simplify the numerical computation and provide design guides. When used to design a given piping system, the method saved approximately one order of magnitude of computer time. The approach can also be used for other types of structures

  10. The Behavior of Corrosion Products in Sampling Systems under Boiling Water Reactor Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Hermansson, Hans-Peter

    1977-08-15

    A high pressure loop has been used to simulate sampling systems employed under BWR conditions. The reliability of the sampling method was studied in a series of six test runs. A variety of parameters that are thought to influence the reliability of the sampling was investigated. These included piping geometry, water oxygen content, flow, temperature and temperature gradients. Amongst other things the results indicate that the loss by deposition of iron containing corrosion products does not exceed 50 %; this figure is only influenced to a minor extent by the above mentioned parameters. The major part of the corrosion products thus deposited is found along the first few meters of the piping and cooler coil. A moderate prolongation of a pipe which is already relatively long should thus be incapable of producing a major influence on the sampling error

  11. Microcomputer based program for predicting heat transfer under reactor accident conditions. Volume I

    International Nuclear Information System (INIS)

    Cheng, S.C.; Groeneveld, D.C.; Leung, L.K.H.; Wong, Y.L.; Nguyen, C.

    1987-07-01

    A microcomputer based program called Heat Transfer Prediction Software (HTPS) has been developed. It calculates the heat transfer for the tube and bundle geometries for steady state and transient conditions. This program is capable of providing the best estimated of the hot pin temperatures during slow transients for 37- and 28-element CANDU type fuel bundles. The program is designed for an IBM-PC AT/XT (or IBM-PC compatible computer) equipped with a Math Co-processor. The following input parameters are required: pressure, mass flux, hydraulic diameter, and quality. For the steady state case, the critical heat flux (CHF), the critical heat flux temperature, the minimum film boiling temperature, and the minimum film boiling heat flux are the primary outputs. With either the surface heat flux or wall temperature specified, the program determines the heat transfer regime and calculates the surface heat flux, wall temperatures and heat transfer coefficient. For the slow transient case, the pressure, mass flux, quality, and volumetric heat generation rate are the time dependent input parameters required to calculate the hot pin sheath temperatures and surface heat fluxes. A simple routine for generating properties has been developed for light water to support the above program. It contains correlations that have been verified for pressures ranging from 0.6kPa to 30 MPa, and temperatures up to 1100 degrees Celcius. The thermodynamic and transport properties that can be generated from this routine are: density, specific volume, enthalpy, specific heat capacity, conductivity, viscosity, surface tension and Prandtl number for saturated liquid, saturated vapour, subcooled liquid for superheated vapour. A software for predicting flow regime has also been developed. It determines the flow pattern at specific flow conditions, and provides a correction factor for calculating the CHF during partially stratified horizontal flow. The technical bases for the program and its

  12. Microcomputer based program for predicting heat transfer under reactor accident conditions. Volume II

    International Nuclear Information System (INIS)

    Cheng, S.C.; Groeneveld, D.C.; Leung, L.K.H.; Wong, Y.L.; Nguyen, C.

    1987-07-01

    A microcomputer based program called Heat Transfer Prediction Software (HTPS) has been developed. It calculates the heat transfer for tube and bundle geometries for steady state and transient conditions. This program is capable of providing the best estimated of the hot pin temperatures during slow transients for 37- and 28-element CANDU type fuel bundles. The program is designed for an IBM-PC AT/XT (or IBM-PC compatible computer) equipped with a Math Co-processor. The following input parameters are required: pressure, mass flux, hydraulic diameter, and quality. For the steady state case, the critical heat flux (CHF), the critical heat flux temperature, the minimum film boiling temperature, and the minimum film boiling heat flux are the primary outputs. With either the surface heat flux or wall temperature specified, the program determines the heat transfer regime and calculates the surface heat flux, wall temperature and heat transfer coefficient. For the slow transient case, the pressure, mass flux, quality, and volumetric heat generation rate are the time dependent input parameters are required to calculate the hot pin sheath temperatures and surface heat fluxes. A simple routine for generating properties has been developed for light water to support the above program. It contains correlations that have been verified for pressures ranging from 0.6kPa to 30 MPa, and temperatures up to 1100 degrees Celcius. The thermodynamic and transport properties that can be generated from this routine are: density, specific volume, enthalpy, specific heat capacity, conductivity, viscosity, surface tension and Prandtle number for saturated liquid, saturated vapour, subcooled liquid of superheated vapour. A software for predicting flow regime has also been developed. It determines the flow pattern at specific flow conditions, and provides a correction factor for calculating the CHF during partially stratified horizontal flow. The technical bases for the program and its structure

  13. Environmental-assisted fatigue in austenitic stainless steels under light water reactor conditions

    International Nuclear Information System (INIS)

    Seifert, H.P.; Ritter, S.; Spaetig, P.

    2015-01-01

    The environmental-assisted fatigue (EAF) initiation and subsequent short crack growth behaviour of different austenitic stainless steels were characterised under simulated BWR/HWC and primary PWR conditions by cyclic fatigue tests with sharply notched fracture mechanics specimens. After a brief summary overview on the previous PSI observations, an update with new and preliminary results about the effect of pH, dissolved hydrogen, load ratio/mean stress, long static load hold times and load sequences is given in this paper. At low electrochemical corrosion potentials (ECP), the physical EAF initiation life moderately decreases with increasing dissolved hydrogen content and decreasing pH. Both parameters have little effect on the subsequent short EAF crack growth within the investigated range. Notch strain amplitude thresholds for environmental effects on physical EAF crack initiation decrease with increasing load ratio and mean stress. At small notch strain amplitudes, the effect of mean stress is more pronounced in BWR/HWC environment than in air and predicted by typical fatigue life mean stress corrections. Under certain loading conditions, long static load hold times result in an increase of the physical EAF initiation life, which saturates for very long hold times. On the other hand, little effect of hold times on subsequent stationary short EAF crack growth rates is observed. The physical EAF initiation life under load sequence loading in high-temperature water may be moderately shorter or significantly longer than predicted by a linear damage accumulation rule and corresponding constant load amplitude tests depending on the load history. (authors)

  14. Removal of pharmaceuticals from wastewater by electrochemical oxidation using cylindrical flow reactor and optimization of treatment conditions.

    Science.gov (United States)

    Babu, B Ramesh; Venkatesan, P; Kanimozhi, R; Basha, C Ahmed

    2009-08-01

    This paper examines the use of electrooxidation for treatment of wastewater obtained from a pharmaceutical industry. The wastewater primarily contained Gentamicin and Dexamethasone. With NaCl as supporting electrolyte, the effluent was treated in a cylindrical flow reactor in continuous (single pass) mode under various current densities (2-5 A/dm2) and flow rates (10-40 L/h). By cyclic voltammetric (CV) analysis, the optimum condition for maximum redox reaction was determined. The efficiency of chemical oxygen demand (COD) reduction and power consumption were studied for different operating conditions. From the results it was observed that maximum COD reduction of about 85.56% was obtained at a flow rate of 10 L/h with an applied current density of 4 A/dm2. FT-IR spectra studies showed that during electrooxidation, the intensities of characteristic functional groups such as N-H, O-H were reduced and some new peaks also started to appear. Probable theory, reaction mechanism and modeling were proposed for the oxidation of pharmaceutical effluent. The experimental results demonstrated that electrooxidation treatment was very effective and capable of elevating the quality of treated wastewater to the reuse standard prescribed for pharmaceutical industries.

  15. Kinetics of devolatilization and oxidation of a pulverized biomass in an entrained flow reactor under realistic combustion conditions

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez, Santiago [LITEC-CSIC (Spanish Council for Scientific Research), Maria de Luna 10, 50018 Zaragoza (Spain); Remacha, Pilar; Ballester, Javier [LITEC-CSIC (Spanish Council for Scientific Research), Maria de Luna 10, 50018 Zaragoza (Spain); Fluid Mechanics Group, University of Zaragoza, Maria de Luna 3, 50018 Zaragoza (Spain); Ballesteros, Juan C.; Gimenez, Antonio [ENDESA GENERACION, S.A., Ribera del Loira 60, 28042 Madrid (Spain)

    2008-03-15

    In this paper the results of a complete set of devolatilization and combustion experiments performed with pulverized ({proportional_to}500 {mu}m) biomass in an entrained flow reactor under realistic combustion conditions are presented. The data obtained are used to derive the kinetic parameters that best fit the observed behaviors, according to a simple model of particle combustion (one-step devolatilization, apparent oxidation kinetics, thermally thin particles). The model is found to adequately reproduce the experimental trends regarding both volatile release and char oxidation rates for the range of particle sizes and combustion conditions explored. The experimental and numerical procedures, similar to those recently proposed for the combustion of pulverized coal [J. Ballester, S. Jimenez, Combust. Flame 142 (2005) 210-222], have been designed to derive the parameters required for the analysis of biomass combustion in practical pulverized fuel configurations and allow a reliable characterization of any finely pulverized biomass. Additionally, the results of a limited study on the release rate of nitrogen from the biomass particle along combustion are shown. (author)

  16. Economic and environmental performance of oil transesterification in supercritical methanol at different reaction conditions: Experimental study with a batch reactor

    International Nuclear Information System (INIS)

    Tomic, Milan; Micic, Radoslav; Kiss, Ferenc; Dedovic, Nebojsa; Simikic, Mirko

    2015-01-01

    Highlights: • Influence of reaction parameters on FAME yields has been investigated. • The highest yield (93%) was achieved after 15 min at 350 °C and 12 MPa. • Models which predict with high certainty yields at different reaction conditions. • Economic and environmental performance of supercritical transesterification. • The lowest costs and impacts are always achieved at the highest yields. - Abstract: This study aims to investigate the influence of various reaction parameters (temperatures, working pressures and reaction time) on biodiesel yields and environmental and economic performance of rapeseed oil transesterification in supercritical methanol. Experiments were carried out in a laboratory-scale batch reactor. Results were statistically analysed and multiple regression models which describe and predict biodiesel yields with high certainty at different reaction conditions were provided. The highest biodiesel yield (93 wt%) was achieved at 350 °C and 12 MPa after 15 min of reaction. The lowest direct costs and life cycle environmental impacts (in terms of GHG emissions and fossil energy demand) are achieved at the highest yield due to the lowest oil consumption per unit of biodiesel produced. The results of sensitivity analysis showed that even at significantly lower oil feedstock prices this observation stands firm

  17. The problems of material pairs under the specific conditions of high-temperature reactors

    International Nuclear Information System (INIS)

    Schoening, J.

    1981-01-01

    This dissertation gives a detailed descirption of the problems of friction and solid lubrication of slide and ball bearings under HTR conditions. The first part is concerned with an analysis of established approaches for determining the friction factor and the development of a mathematical model on this basis, taking particular account of the following parameters. (1) Surface characteristics (peak-to-valley height, pressure angles of roughness elevations) and, (2) material properties of the friction partner (ductility, brittleness, crystal structure). The formula for friction factor thus arrived at incorporates the microscopic and sub-microscopic influence of the surface grain structure. In addition, the effects of elastic and plastic properties at the region of contact was studied. The derived equation is used to characterise frictional properties and enables a suitable and economic friction pair to be selected. The second part deals with the application of solid lubrication and the effect of the solid lubricant film on the mechanicsm and kinematics of ball bearings. In the final part of this investigation an evaluation of experimental results is carried out on the basis of the preceding theoretical studies in order to confirm the theoretical approach and to provide information of practical use on limits of application, cost details and aids to construction. (orig./IHOE) [de

  18. Aerosol resuspension in the reactor cooling system of LWR's under severe accident conditions

    International Nuclear Information System (INIS)

    Alonso, A.; Bolado, R.; Hontanon, E.

    1991-07-01

    Aerosol resuspension from the pipes of the RCS under severe accident conditions happens when the carrier gas flow is turbulent. The origin of such phenomenon seems to be the existence of turbulent bursts in the neighbourhood of the pipe wall. These bursts are of random nature, in space and time. Three theoretical models have been found in available literature; those are: Cleaver and Yates', RESUS and Reeks' models. The first two of them are force balance models, in which particle detachment is supposed whenever aerodynamic lift or drag forces, respectively exceed adhesive forces, and the third one is an energy balance model in which resuspension happens when particle vibrational energy exceeds adhesive potential. From experimental evidence it seems that the studied phenomenon is a force balance problem and RESUS seems to be the most appropriate to it, among the available ones. Small-scale experiments have shown, as main parameters affecting resuspension, the Reynolds number of the flow, aerosol composition and initial loading per unit of area. Moreover, the resuspension rate decreases with time in all experiments where temporal measurements were taken

  19. Thermo-hydraulic characteristics of ship propulsion reactor in the conditions of ship motions and safety assessment

    International Nuclear Information System (INIS)

    Kobayashi, Michiyuki; Aya, Izuo; Inasaka, Fujio; Murata, Hiroyuki; Odano, Naoteru; Shiozaki, Koki

    1998-01-01

    A research project from 1995-1999 had a plan to make experimental studies on (1) safety of nuclear ship loaded with an integral ship propulsion reactor (2) effects of pulsating flow on the thermo-hydraulic characteristics of ship propulsion reactor and (3) thermo-hydraulic behaviors of the reactor container at the time of accident in a passively safe ship propulsion reactor. Development of a data base for ship propulsion reactor was attempted using previous experimental data on the thermo-hydraulic characteristics of the reactor in the institute in addition to the present results aiming to make general analytical evaluation for the safety of the engineering-simulation system for nuclear ship. A general data base was obtained by integrating the data list and the analytical program for static characteristics. A test equipment which allows to visualize the pulsating flow was produced and visualization experiments have started. (M.N.)

  20. Strength analysis of fast gas cooled reactor fuel element in conditions of fuel-cladding interraction and non-uniform azimuthal heating

    International Nuclear Information System (INIS)

    Kulikov, I.S.; Tverkovkin, B.E.

    1984-01-01

    The technique and the PRORT mathematical program in FORTRAN language for determining mechanical properties of a fuel element with motionless fuel-cladding interaction taking into account circular temperature non-uniformity in gas-cooled fast reactor conditions are proposed. The calculation results of the fuel element of dissociating gas cooled fast reactor are presented for seven cross-sections over the height of the core. The obtained data testify to appreciable swelling of Cr16Ni15Mo3Nb steel fuel cladding in the conditions of dissociating gas cooled fast reactor through the allowance for the effect of stresses on this essential parameter shows, that its value is lower in comparison with swelling, wherein stresses are not taken into account

  1. Conditioning of radioactive aluminium generated by the VVR-S Nuclear Reactor Decommissioning Laboratory Inactive Tests

    International Nuclear Information System (INIS)

    Nicu, M.; Ionascu, L.; Turcau, C.; Dragolici, F.; Rotarescu, G.

    2015-01-01

    Aluminium is a reactive amphoteric metal, readily forming a protective oxide layer on contact with air or water. However, as the oxides are amphoteric, aluminium is not resistant to corrosion in acidic and alkaline conditions, because the protective films dissolve. As a consequence radioactive waste containing bulk aluminium alloys can not be embedded in Ordinary Portland Cement (OPC). A potential encapsulating material for the radioactive aluminium is potassium magnesium phosphate (MKP). This paper presents the characterization results obtained from analyzing the potential magnesium phosphate formulations and assesses its potential to reduce the corrosion of aluminium. A series of experiments have been performed. The main conclusions of the paper are as follows. First, the pH values of magnesium phosphate formulation investigated increased gradually over the test duration, with pH measurement ranging from 8.1 - 9.1, indicating lower values compared with the reference composite OPC (pH ∼ 13). The reduction of pH is an important controlling factor for the corrosion of aluminium. Secondly, according to XRD, the hardened magnesium phosphate matrix is polycrystalline and the main reaction product of magnesium phosphate cement formulations was confirmed as MgKPO 4 -6H 2 O, which was found to dominate the crystalline phase composition. Thirdly, the compressive strengths obtained for magnesium phosphate matrices investigated are included in the accepted limits for the embedding matrix with cement (above 5 N/mm 2 ). And fourthly, the corrosion of metallic aluminium in magnesium phosphate matrix is markedly reduced in comparison with the composite OPC

  2. Characterization of the planktonic microbiome in upflow anaerobic sludge blanket reactors during adaptation of mesophilic methanogenic granules to thermophilic operational conditions

    DEFF Research Database (Denmark)

    Zhu, Xinyu; Treu, Laura; Kougias, Panagiotis

    2017-01-01

    Upflow anaerobic sludge blanket (UASB) technology refers to reactor technology where granules, i.e. self-immobilised microbial associations, are the biological catalysts involved in the anaerobic digestion process. During the start-up period, UASB reactors operate at relatively long HRT and there......Upflow anaerobic sludge blanket (UASB) technology refers to reactor technology where granules, i.e. self-immobilised microbial associations, are the biological catalysts involved in the anaerobic digestion process. During the start-up period, UASB reactors operate at relatively long HRT...... and therefore the liquid phase of the reactor becomes a favourable environment for microbial growth. The current study aimed to elucidate the dynamicity of the suspended microbial community in UASB reactors, during the transition from mesophilic to thermophilic conditions. High throughput 16S rRNA amplicon...... sequencing was used to characterize the taxonomic composition of the microbiome. The results showed that the microbial community was mainly composed by hydrolytic and fermentative bacteria. Results revealed relevant shifts in the microbial community composition, which is mainly determined by the operational...

  3. Development of empirical models for performance evaluation of UASB reactors treating poultry manure wastewater under different operational conditions

    Energy Technology Data Exchange (ETDEWEB)

    Yetilmezsoy, Kaan [Department of Environmental Engineering, Yildiz Technical University, 34349 Yildiz, Besiktas, Istanbul (Turkey)], E-mail: yetilmez@yildiz.edu.tr; Sakar, Suleyman [Department of Environmental Engineering, Yildiz Technical University, 34349 Yildiz, Besiktas, Istanbul (Turkey)

    2008-05-01

    A nonlinear modeling study was carried out to evaluate the performance of UASB reactors treating poultry manure wastewater under different organic and hydraulic loading conditions. Two identical pilot scale up-flow anaerobic sludge blanket (UASB) reactors (15.7 L) were run at mesophilic conditions (30-35 deg. C) in a temperature-controlled environment with three hydraulic retention times ({theta}) of 15.7, 12 and 8.0 days. Imposed volumetric organic loading rates (L{sub V}) ranged from 0.65 to 4.257 kg COD/(m{sup 3} day). The pH of the feed varied between 6.68 and 7.82. The hydraulic loading rates (L{sub H}) were controlled between 0.105 and 0.21 m{sup 3}/(m{sup 2} day). The daily biogas production rates ranged between 4.2 and 29.4 L/day. High volumetric COD removal rates (R{sub V}) ranging from 0.546 to 3.779 kg COD{sub removed}/(m{sup 3} day) were achieved. On the basis of experimental results, two empirical models having a satisfactory correlation coefficient of about 0.9954 and 0.9416 were developed to predict daily biogas production (Q{sub g}) and effluent COD concentration (S{sub e}), respectively. Findings of this modeling study showed that optimal COD removals ranging from 86.3% to 90.6% were predicted with HRTs of 7.9, 9.5, 11.2, 12.6, 13.7 and 14.3 days, and L{sub V} of 1.27, 1.58, 1.78, 1.99, 2.20 and 2.45 kg COD/(m{sup 3} day) for the corresponding influent substrate concentrations (S{sub i}) of 10,000, 15,000, 20,000, 25,000, 30,000 and 35,000 mg/L, respectively.

  4. Crack growth behaviour of low alloy steels for pressure boundary components under transient light water reactor operating conditions (CASTOC)

    International Nuclear Information System (INIS)

    Foehl, J.; Weissenberg, T.; Gomez-Briceno, D.; Lapena, J.; Ernestova, M.; Zamboch, M.; Seifert, H.P.; Ritter, S.; Roth, A.; Devrient, B.; Ehrnsten, U.

    2004-01-01

    The CASTOC project addresses environmentally assisted cracking (EAC) phenomena in low alloy steels used for pressure boundary components in both Western type boiling water reactors (BWR) and Russian type pressurised water reactors (VVER). It comprises the four work packages (WP): inter-laboratory comparison test (WP1); EAC behaviour under static load (WP2), EAC behaviour under cyclic load and load transients (WP3); evaluation of the results with regard to their relevance for components in practice (WP4). The use of sophisticated test facilities and measurement techniques for the on-line detection of crack advances have provided a more detailed understanding of the mechanisms of environmentally assisted cracking and provided quantitative data of crack growth rates as a function of loading events and time, respectively. The effect of several major parameters controlling EAC was investigated with particular emphasis on the transferability of the results to components in service. The obtained crack growth rate data were reflected on literature data and on commonly applied prediction curves as presented in the appropriate Code. At relevant stress intensity factors it could be shown that immediate cessation of growing cracks occurs after changing from cyclic to static load in high purity oxygenated BWR water and oxygen-free VVER water corresponding to steady state operation conditions. Susceptibility to environmentally assisted cracking under static load was observed for a heat affected zone material in oxygenated high purity water and also in base materials during a chloride transient representing BWR water condition below Action Level 1 of the EPRI Water Chemistry Guidelines according to the lectrical conductivity of the water but in the range of Action Level 2 according to the content of chlorides. Time based crack growth was also observed in one Russian type base material in oxygenated VVER water and in one Western type base material in oxygenated high purity BWR

  5. Thermal hydraulic conditions inducing incipient cracking in the 900 MWe unit 93 D reactor coolant pump shafts

    International Nuclear Information System (INIS)

    Bore, C.

    1995-01-01

    From 1987, 900 MWe plant operating feedback revealed cracking in the lower part of the reactor coolant pump shafts, beneath the thermal ring. Metallurgical examinations established that this was due to a thermal fatigue phenomenon known as thermal crazing, occurring after a large number of cycles. Analysis of thermal hydraulic conditions initiating the cracks does not allow exact quantification of the thermal load inducing cracking. Only qualitative analyses are thus possible, the first of which, undertaken by the pump manufacturer, Jeumont Industrie, showed that the cracks could not be due to the major transients (stop-start, injection cut-off), which were too few in number. Another explanation was then put forward: the thermal ring, shrunk onto the shaft it is required to protect against thermal shocks, loosens to allow an alternating downflow of cold water from the shaft seals and an upflow of hot water from the primary system. However, approximate calculations showed that the flow involved would be too slight to initiate the cracking observed. A more stringent analysis undertaken with the 2D flow analysis code MELODIE subsequently refuted the possibility of alternating flows beneath the ring establishing that only a hot water upflow occurred due to a 'viscosity pump' phenomenon. Crack initiation was finally considered to be due to flowrate variations beneath the ring, with the associated temperature fluctuations. This flowrate fluctuation could be due to an unidentified transient phenomenon or to a variation in pump operating conditions. This analysis of the hydraulic conditions initiating the cracks disregards shaft surface residual stresses. These are tensile stresses and show that loads less penalizing than those initially retained could cause incipient cracking. Thermal ring modifications to reduce these risks were proposed and implemented. In addition, final metallurgical treatment of the shafts was altered and implemented. In addition, final metallurgical

  6. Nuclear reactor

    International Nuclear Information System (INIS)

    Garabedian, G.

    1988-01-01

    A liquid reactor is described comprising: (a) a reactor vessel having a core; (b) one or more satellite tanks; (c) pump means in the satellite tank; (d) heat exchanger means in the satellite tank; (e) an upper liquid metal conduit extending between the reactor vessel and the satellite tank; (f) a lower liquid metal duct extending between the reactor vessel and satellite tanks the upper liquid metal conduit and the lower liquid metal duct being arranged to permit free circulation of liquid metal between the reactor vessel core and the satellite tank by convective flow of liquid metal; (g) a separate sealed common containment vessel around the reactor vessel, conduits and satellite tanks; (h) the satellite tank having space for a volume of liquid metal that is sufficient to dampen temperature transients resulting from abnormal operating conditions

  7. Accident conditions analysis of spent fuel storage pool RA research reactor in Vinca; Analiza udesnih stanja u odlagalistu isluzenog goriva istrazivackog rektora RA u Vinci

    Energy Technology Data Exchange (ETDEWEB)

    Jovic, V; Jovic, L [Institute of Nuclear Sciences VINCA, Belgrade (Serbia and Montenegro)

    2000-07-01

    Based on Safety analysis of the spent fuel pool RA research reactor in Vinca, conditions and possibilities accident sequences in present configuration storage facility are considered (author) [Serbo-Croat] Na osnovu Analize sigurnosti odlagalista isluzenog goriva istrazivackog reaktora RA u Vinci razmatraju se uslovi i mogucnosti pojave udesnih stanja u postojecoj konfiguraciji odlagalista (author)

  8. PC-version of RAM6-code for calculation of parameters of the effective logarithmic boundary condition at the absorbent rod surface in reactor

    International Nuclear Information System (INIS)

    Le Van Ngoc; Ngo Dang Nhan

    1990-01-01

    The RAM-6 code for calculation of parameters of the effective logarithmic boundary condition at the absorbent rod surface in reactor is suitably modofied to work on IBM PC, the instructions for its usage are presented and capabilities of the personal cpmputer oriented RAM-6 code are demonstrated. (author). 4 refs, 5 tabs, 2 figs

  9. Computer simulations of a 1/5-scale experiment of a Mark I boiler water reactor pressure-suppression system under hypothetical LOCA conditions

    International Nuclear Information System (INIS)

    Edwards, L.L.

    1978-01-01

    The CHAMP computer code was employed to simulate a plane-geometry cross section of a Mark I boiling water reactor toroidal pressure suppression system air discharge experiment under hypothetical loss-of-coolant accident conditions. The experiments were performed at the Lawrence Livermore Laboratory on a 1 / 5 -scale model of the Peach Bottom Nuclear Power Plant

  10. Study on liquid-metal MHD power generation system with two-phase natural circulation. Applicability to fast reactor conditions

    International Nuclear Information System (INIS)

    Saito, Masaki

    2001-03-01

    the buoyancy in the two-phase natural circulation. The second series of the experiments with water and nitrogen gas were performed to demonstrate the effects of the mixing conditions of liquid and gas on the reduction of the gas-liquid slip ratio, which is essential to improve the cycle efficiency. It was confirmed that the intermittent mixing of liquid and gas effectively reduces the gas-liquid slip ratio, which results were also confirmed in the experiments with Pb-Bi (Density: 10453 Kg/m 3 ) and nitrogen gas. Based on the present experiments, the cycle analyses performed for the applicability of the present system to the next generation Fast Reactor showed, for example, the possibility of the cycle efficiency of 45% in the conditions of the maximum system temperature of 800 K and maximum system pressure of 3 MPa. (author)

  11. Biohydrogen production from glucose in upflow biofilm reactors with plastic carriers under extreme thermophilic conditions (70(degree)C)

    DEFF Research Database (Denmark)

    Zheng, H.; Zeng, Raymond Jianxiong; Angelidaki, Irini

    2008-01-01

    Biohydrogen could efficiently be produced in glucose-fed biofilm reactors filled with plastic carriers and operated at 70°C. Batch experiments were, in addition, conducted to enrich and cultivate glucose-fed extremethermophilic hydrogen producing microorganisms from a biohydrogen CSTR reactor fed...

  12. Specific schedule conditions for the formation of personnel of A or B category working in nuclear facilities. Option nuclear reactor

    CERN Document Server

    Int. At. Energy Agency, Wien

    2002-01-01

    This document describes the specific dispositions relative to the nuclear reactor domain, for the formation to the conventional and radiation risks prevention of personnel of A or B category working in nuclear facilities. The application domain, the applicable documents, the liability, the specificity of the nuclear reactor and of the retraining, the Passerelle formation, are presented. (A.L.B.)

  13. Behavior of U3Si2 Fuel and FeCrAl Cladding under Normal Operating and Accident Reactor Conditions

    International Nuclear Information System (INIS)

    Gamble, Kyle Allan Lawrence; Hales, Jason Dean; Barani, Tommaso; Pizzocri, Davide; Pastore, Giovanni

    2016-01-01

    As part of the Department of Energy's Nuclear Energy Advanced Modeling and Simulation program, an Accident Tolerant Fuel High Impact Problem was initiated at the beginning of fiscal year 2015 to investigate the behavior of \\usi~fuel and iron-chromium-aluminum (FeCrAl) claddings under normal operating and accident reactor conditions. The High Impact Problem was created in response to the United States Department of Energy's renewed interest in accident tolerant materials after the events that occurred at the Fukushima Daiichi Nuclear Power Plant in 2011. The High Impact Problem is a multinational laboratory and university collaborative research effort between Idaho National Laboratory, Los Alamos National Laboratory, Argonne National Laboratory, and the University of Tennessee, Knoxville. This report primarily focuses on the engineering scale research in fiscal year 2016 with brief summaries of the lower length scale developments in the areas of density functional theory, cluster dynamics, rate theory, and phase field being presented.

  14. Fuel-coolant interaction visualization test for in-vessel corium retention external reactor vessel cooling (IVR-ERVC) condition

    Energy Technology Data Exchange (ETDEWEB)

    Na, Young Su; Hong, Seong Ho; Song, Jin Ho; Hong, Seong Wan [Severe Accident and PHWR Safety Research Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-12-15

    A visualization test of the fuel-coolant interaction in the Test for Real cOrium Interaction with water (TROI) test facility was carried out. To experimentally simulate the In-Vessel corium Retention (IVR)- External Reactor Vessel Cooling (ERVC) conditions, prototypic corium was released directly into the coolant water without a free fall in a gas phase before making contact with the coolant. Corium (34.39 kg) consisting of uranium oxide and zirconium oxide with a weight ratio of 8:2 was superheated, and 22.54 kg of the 34.39 kg corium was passed through water contained in a transparent interaction vessel. An image of the corium jet behavior in the coolant was taken by a high-speed camera every millisecond. Thermocouple junctions installed in the vertical direction of the coolant were cut sequentially by the falling corium jet. It was clearly observed that the visualization image of the corium jet taken during the fuel-coolant interaction corresponded with the temperature variations in the direction of the falling melt. The corium penetrated through the coolant, and the jet leading edge velocity was 2.0 m/s. Debris smaller than 1 mm was 15% of the total weight of the debris collected after a fuel-coolant interaction test, and the mass median diameter was 2.9 mm.

  15. Basic visualization experiments on eutectic reaction of boron carbide and stainless steel under sodium-cooled fast reactor conditions

    International Nuclear Information System (INIS)

    Yamano, Hidemasa; Suzuki, Tohru; Kamiyama, Kenji; Kudo, Isamu

    2016-01-01

    This paper describes basic visualization experiments on eutectic reaction and relocation of boron carbide (B 4 C) and stainless steel (SS) under a high temperature condition exceeding 1500degC as well as the importance of such behaviors in molten core during a core disruptive accident in a Generation-IV sodium-cooled fast reactor (750 MWe class) designed in Japan. At first, a reactivity history was calculated using an exact perturbation calculation tool taking into account expected behaviors. This calculation indicated the importance of a relocation behavior of the B 4 C-SS eutectic because its behavior has a large uncertainty in the reactivity history. To clarify this behavior, basic experiments were carried out by visualizing the reaction of a B 4 C pellet contacted with molten SS in a high temperature-heating furnace. The experiments have shown the eutectic reaction visualization as well as freezing and relocation of the B 4 C-SS eutectic in upper part of the solidified test piece due to the density separation. (author)

  16. Changes in bacterial community structure correlate with initial operating conditions of a field-scale denitrifying fluidized bed reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, C. [Miami Univ., Oxford, OH (United States). Dept. of Microbiology; Wu, W.M. [Stanford Univ., CA (United States). Dept. of Civil and Environmental Engineering; Gentry, T.J. [Oak Ridge National Lab., TN (US). Environmental Sciences Div.] (and others)

    2006-08-15

    High levels of nitrate are present in groundwater migrating from the former waste disposal ponds at the Y-12 National Security Complex in Oak Ridge, TN. A field-scale denitrifying fluidized bed reactor (FBR) was designed, constructed, and operated with ethanol as an electron donor for the removal of nitrate. After inoculation, biofilms developed on the granular activated carbon particles. Changes in the bacterial community of the FBR were evaluated with clone libraries (n=500 partial sequences) of the small-subunit rRNA gene for samples taken over a 4-month start-up period. Early phases of start-up operation were characterized by a period of selection, followed by low diversity and predominance by Azoarcus-like sequences. Possible explanations were high pH and nutrient limitations. After amelioration of these conditions, diversification increased rapidly, with the appearance of Dechloromonas, Pseudomonas, and Hydrogenophaga sequences. Changes in NO{sub 3}, SO{sub 4}, and pH also likely contributed to shifts in community composition. The detection of sulfate-reducing-bacteria-like sequences closely related to Desulfovibrio and Desulfuromonas in the FBR have important implications for downstream applications at the field site. (orig.)

  17. Results of VVER fuel rods tests in the MIR.M1 reactor under power cycling conditions

    International Nuclear Information System (INIS)

    Burukin, A.; Izhutov, A.; Ovchinnikov, V.; Kalygin, V.; Markov, D.; Pimenov, Y.; Novikov, V.; Medvedev, A.; Nesterov, B.

    2011-01-01

    The paper presents the main results of the 50 ... 60 MWd/kgU burnup VVER fuel rods tests performed in the MIR.M1 reactor loop facilities under power cycling. The non-destructive PIE results are presented as well. A series of experiments was performed, including overall measurement of fuel rod parameters test, in one of which 300 cycles were done. Irradiation under power cycling conditions and PIE of high-burnup VVER fuel rods showed the following: 1) all fuel rods claddings preserved their integrity under irradiation at linear heat rate (LHR) higher than the NPP operating one; 2) experimental data were obtained on the axial and radial cladding strain and fission gas release (FGR) from 50 ... 60 MWd/kgU burnup VVER-440 and VVER-1000 fuel rods as well as on the kinetics of the change in these parameters and fuel temperature under the power cycling; 3) non-destructive PIE results are in a satisfactory correlation with the data obtained by means of in-pile measurement gages during irradiation. (authors)

  18. Thermal–mechanical stress analysis of pressurized water reactor pressure vessel with/without a preexisting crack under grid load following conditions

    Energy Technology Data Exchange (ETDEWEB)

    Mohanty, Subhasish, E-mail: smohanty@anl.gov; Soppet, William K.; Majumdar, Saurin; Natesan, Krishnamurti

    2016-12-15

    Highlights: • Use of intermittent renewable-energy source in power grid is becoming a trend. • Gird load-following can leads to variable power demand from Nuclear power plant. • Reactor components can be stressed differently under gird load-following mode. • Estimation of stress–strain state under grid load-following condition is essential. - Abstract: In this paper, we present thermal–mechanical stress analysis of a pressurized water reactor pressure vessel and its hot-leg and cold-leg nozzles. Results are presented from thermal and thermal–mechanical stress analysis under reactor heat-up, cool-down, and grid load-following conditions. Analysis results are given with and without the presence of preexisting crack in the reactor nozzle (axial crack in hot leg nozzle). From the model results it is found that the stress–strain states are significantly higher in case of presence of crack than without crack. The stress–strain state under grid load following condition are more realistic compared to the stress–strain state estimated assuming simplified transients.

  19. Thermal–mechanical stress analysis of pressurized water reactor pressure vessel with/without a preexisting crack under grid load following conditions

    International Nuclear Information System (INIS)

    Mohanty, Subhasish; Soppet, William K.; Majumdar, Saurin; Natesan, Krishnamurti

    2016-01-01

    Highlights: • Use of intermittent renewable-energy source in power grid is becoming a trend. • Gird load-following can leads to variable power demand from Nuclear power plant. • Reactor components can be stressed differently under gird load-following mode. • Estimation of stress–strain state under grid load-following condition is essential. - Abstract: In this paper, we present thermal–mechanical stress analysis of a pressurized water reactor pressure vessel and its hot-leg and cold-leg nozzles. Results are presented from thermal and thermal–mechanical stress analysis under reactor heat-up, cool-down, and grid load-following conditions. Analysis results are given with and without the presence of preexisting crack in the reactor nozzle (axial crack in hot leg nozzle). From the model results it is found that the stress–strain states are significantly higher in case of presence of crack than without crack. The stress–strain state under grid load following condition are more realistic compared to the stress–strain state estimated assuming simplified transients.

  20. Description of the blowdown test facility COG program on in-reactor fission product release, transport, and deposition under severe accident conditions

    International Nuclear Information System (INIS)

    Fehrenbach, P.J.; Wood, J.C.

    1987-06-01

    Loss-of-coolant accidents with additional impairment of emergency cooling would probably result in high fuel temperatures leading to severe fuel damage (SFD) and significant fission product activity would then be transported along the PHTS to the break where a fraction of it would be released and transport under such conditions, there are many interacting and sometimes competing phenomena to consider. Laboratory simulations are being used to provide data on these individual phenomena, such as UO 2 oxidation and Zr-UO 2 interaction, from which mathematical models can be constructed. These are then combined into computer codes to include the interaction effects and assess the overall releases. In addition, in-reactor tests are the only source of data on release and transport of short-lived fission product nuclides, which are important in the consequence analysis of CANDU reactor accidents. Post-test decontamination of an in-reactor test facility also provides a unique opportunity to demonstrate techniques and obtain decontamination data relevant to post-accident rehabilitation of CANDU power reactors. Specialized facilities are required for in-reactor testing because of the extensive release of radioactive fission products and the high temperatures involved (up to 2500 degrees Celsius). To meet this need for the Canadian program, the Blowdown Test Facility (BTF) has been built in the NRU reactor at Chalk River. Between completion of construction in mid-1987 and the first Zircaloy-sheathed fuel test in fiscal year 1987/88, several commissioning tests are being performed. Similarly, extensive development work has been completed to permit application of instrumentation to irradiated fuel elements, and in support of post-test fuel assembly examination. A program of decontamination studies has also been developed to generate information relevant to post-accident decontamination of power reactors. The BTF shared cost test program funded by the COG High Temperature

  1. Large-scale experimental facility for emergency condition investigation of a new generation NPP WWER-640 reactor with passive safety systems

    International Nuclear Information System (INIS)

    Aniskevich, Y.N.; Vasilenko, V.A.; Zasukha, V.K.; Migrov, Y.A.; Khabensky, V.B.

    1997-01-01

    The creation of the large-scale integral experimental facility (KMS) is specified by the programme of the experimental investigations to justify the engineering decisions on the safety of the design of the new generation NPP with the reactor WWER-640. The construction of KMS in a full volume will allow to conduct experimental investigations of all physical phenomena and processes, practically, occurring during the accidents on the NPPs with the reactor of WWER type and including the heat - mass exchange processes with low rates of the coolant, which is typical during the utilization of the passive safety systems, process during the accidents with a large leak, and also the complex intercommunicated processes in the reactor unit, passive safety systems and in the containment with the condition of long-term heat removal to the final absorber. KMS is being constructed at the Research Institute of Technology (NITI), Sosnovy Bor, Leningrad region, Russia. (orig.)

  2. Large-scale experimental facility for emergency condition investigation of a new generation NPP WWER-640 reactor with passive safety systems

    Energy Technology Data Exchange (ETDEWEB)

    Aniskevich, Y.N.; Vasilenko, V.A.; Zasukha, V.K.; Migrov, Y.A.; Khabensky, V.B. [Research Inst. of Technology NITI (Russian Federation)

    1997-12-31

    The creation of the large-scale integral experimental facility (KMS) is specified by the programme of the experimental investigations to justify the engineering decisions on the safety of the design of the new generation NPP with the reactor WWER-640. The construction of KMS in a full volume will allow to conduct experimental investigations of all physical phenomena and processes, practically, occurring during the accidents on the NPPs with the reactor of WWER type and including the heat - mass exchange processes with low rates of the coolant, which is typical during the utilization of the passive safety systems, process during the accidents with a large leak, and also the complex intercommunicated processes in the reactor unit, passive safety systems and in the containment with the condition of long-term heat removal to the final absorber. KMS is being constructed at the Research Institute of Technology (NITI), Sosnovy Bor, Leningrad region, Russia. (orig.). 5 refs.

  3. Applicability of AWJ technique for dismantling reactor of the Fukushima Daiichi Nuclear Power Station. Cutting test of imitation of fuel debris and optimization of the cutting condition

    International Nuclear Information System (INIS)

    Maruyama, Shin-ichiro; Watatani, Satoshi

    2016-01-01

    Based on findings during recovery works that followed the accident at Three Mile Island Station 2, it is assumed that the reactor internals at the Fukushima Daiichi Nuclear Power Station (1F) have complex geometries intermixed with melted fuel and confined in limited spaces. Accordingly, abrasive water jet (AWJ) cutting method is considered to be a promising technique that can be safely and reasonably used for cutting and removing reactor internals. The authors conducted tests to examine the possibility of application and to solve the problems of this technique. In the tests imitation of fuel debris and optimization of the cutting condition is used. The test result made the measures for some of the associated issues clear, and demonstrated that AWJ cutting method is assumed as one of the promising techniques for removing reactor internals. (author)

  4. Large-scale experimental facility for emergency condition investigation of a new generation NPP WWER-640 reactor with passive safety systems

    Energy Technology Data Exchange (ETDEWEB)

    Aniskevich, Y N; Vasilenko, V A; Zasukha, V K; Migrov, Y A; Khabensky, V B [Research Inst. of Technology NITI (Russian Federation)

    1998-12-31

    The creation of the large-scale integral experimental facility (KMS) is specified by the programme of the experimental investigations to justify the engineering decisions on the safety of the design of the new generation NPP with the reactor WWER-640. The construction of KMS in a full volume will allow to conduct experimental investigations of all physical phenomena and processes, practically, occurring during the accidents on the NPPs with the reactor of WWER type and including the heat - mass exchange processes with low rates of the coolant, which is typical during the utilization of the passive safety systems, process during the accidents with a large leak, and also the complex intercommunicated processes in the reactor unit, passive safety systems and in the containment with the condition of long-term heat removal to the final absorber. KMS is being constructed at the Research Institute of Technology (NITI), Sosnovy Bor, Leningrad region, Russia. (orig.). 5 refs.

  5. Determining the microwave coupling and operational efficiencies of a microwave plasma assisted chemical vapor deposition reactor under high pressure diamond synthesis operating conditions

    Energy Technology Data Exchange (ETDEWEB)

    Nad, Shreya [Department of Electrical and Computer Engineering, Michigan State University, East Lansing, Michigan 48824 (United States); Department of Physics and Astronomy, Michigan State University, East Lansing, Michigan 48824 (United States); Gu, Yajun; Asmussen, Jes [Department of Electrical and Computer Engineering, Michigan State University, East Lansing, Michigan 48824 (United States)

    2015-07-15

    The microwave coupling efficiency of the 2.45 GHz, microwave plasma assisted diamond synthesis process is investigated by experimentally measuring the performance of a specific single mode excited, internally tuned microwave plasma reactor. Plasma reactor coupling efficiencies (η) > 90% are achieved over the entire 100–260 Torr pressure range and 1.5–2.4 kW input power diamond synthesis regime. When operating at a specific experimental operating condition, small additional internal tuning adjustments can be made to achieve η > 98%. When the plasma reactor has low empty cavity losses, i.e., the empty cavity quality factor is >1500, then overall microwave discharge coupling efficiencies (η{sub coup}) of >94% can be achieved. A large, safe, and efficient experimental operating regime is identified. Both substrate hot spots and the formation of microwave plasmoids are eliminated when operating within this regime. This investigation suggests that both the reactor design and the reactor process operation must be considered when attempting to lower diamond synthesis electrical energy costs while still enabling a very versatile and flexible operation performance.

  6. Numerical Analysis of The Effect of Hydrodynamics and Operating Conditions on Biodiesel Synthesis in a Rotor-Stator Spinning Disk Reactor

    Directory of Open Access Journals (Sweden)

    Wen Zhuqing

    2017-06-01

    Full Text Available A rotor-stator spinning disk reactor for intensified biodiesel synthesis is described and numerically simulated in the present research. The reactor consists of two flat disks, located coaxially and parallel to each other with a gap ranging from 0.1 mm to 0.2 mm between the disks. The upper disk is located on a rotating shaft while the lower disk is stationary. The feed liquids, triglycerides (TG and methanol are injected into the reactor from centres of rotating disk and stationary disk, respectively. Fluid hydrodynamics in the reactor for synthesis of biodiesel from TG and methanol in the presence of a sodium hydroxide catalyst are simulated, using convection-diffusion-reaction multicomponent transport model with the CFD software ANSYS©Fluent v. 13.0. Effect of operating conditions on TG conversion is particularly investigated. Simulation results indicate that there is occurrence of back flow close to the stator at the outlet zone. Small gap size and fast rotational speed generally help to intensify mixing among reagents, and consequently enhance TG conversion. However, increasing rotational speed of spinning disk leads to more backflow, which decreases TG conversion. Large flow rate of TG at inlet is not recommended as well because of the short mean residence time of reactants inside the reactor.

  7. A novel ion transport membrane reactor for fundamental investigations of oxygen permeation and oxy-combustion under reactive flow conditions

    KAUST Repository

    Kirchen, Patrick; Apo, Daniel J.; Hunt, Anton; Ghoniem, Ahmed F.

    2013-01-01

    Ion transport membrane (ITM) reactors present an attractive technology for combined air separation and fuel conversion in applications such as syngas production, oxidative coupling or oxy-combustion, with the promise of lower capital and operating

  8. Molecular analysis of the biomass of a fluidized bed reactor treating synthetic vinasse at anaerobic and micro-aerobic conditions

    NARCIS (Netherlands)

    Rodriguez, E.; Lopes, A.; Fdz-Polanco, M.; Stams, A.J.M.; Garcia Encina, P.A.

    2012-01-01

    The microbial communities (Bacteria and Archaea) established in an anaerobic fluidized bed reactor used to treat synthetic vinasse (betaine, glucose, acetate, propionate, and butyrate) were characterized by denaturing gradient gel electrophoresis (DGGE) and phylogenetic analysis. This study was

  9. Reactor core in FBR type reactor

    International Nuclear Information System (INIS)

    Masumi, Ryoji; Kawashima, Katsuyuki; Kurihara, Kunitoshi.

    1989-01-01

    In a reactor core in FBR type reactors, a portion of homogenous fuels constituting the homogenous reactor core is replaced with multi-region fuels in which the enrichment degree of fissile materials is lower nearer to the axial center. This enables to condition the composition such that a reactor core having neutron flux distribution either of a homogenous reactor core or a heterogenous reactor core has substantially identical reactivity. Accordingly, in the transfer from the homogenous reactor core to the axially heterogenous reactor core, the average reactivity in the reactor core is substantially equal in each of the cycles. Further, by replacing a portion of the homogenous fuels with a multi-region fuels, thereby increasing the heat generation near the axial center, it is possiable to reduce the linear power output in the regions above and below thereof and, in addition, to improve the thermal margin in the reactor core. (T.M.)

  10. Comparison of CSTR and UASB reactor configuration for the treatment of sulfate rich wastewaters under acidifying conditions

    NARCIS (Netherlands)

    Lopes, S.I.C.; Dreissen, C.; Capela, M.I.; Lens, P.N.L.

    2008-01-01

    The effects of lowering the operational pH from 6 to 5 on mesophilic (30 degrees C) sulfate reduction during the acidification of sucrose at an organic loading rate of 5 gCOD(1(reactor)d)(-1) and at a COD/SO42- ratio of 4 were evaluated in a CSTR and in a UASB reactor. The HRT was 24 h and 10 h,

  11. Invitational Education: Theory, Research and Practice

    Science.gov (United States)

    Haigh, Martin

    2011-01-01

    Invitational Theory argues that learning is enhanced when learners are positively encouraged or "invited" into the educational experience. Arising from perceptual and self-concept theory, Invitational Pedagogy is constructed on four principles: respect for people, trust, optimism and intentionality, and upon five pillars: people, places, policies,…

  12. Slurry reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kuerten, H; Zehner, P [BASF A.G., Ludwigshafen am Rhein (Germany, F.R.)

    1979-08-01

    Slurry reactors are designed on the basis of empirical data and model investigations. It is as yet not possible to calculate the flow behavior of such reactors. The swarm of gas bubbles and cluster formations of solid particles and their interaction in industrial reactors are not known. These effects control to a large extent the gas hold-up, the gas-liquid interface and, similarly as in bubble columns, the back-mixing of liquids and solids. These hydrodynamic problems are illustrated in slurry reactors which constructionally may be bubble columns, stirred tanks or jet loop reactors. The expected effects are predicted by means of tests with model systems modified to represent the conditions in industrial hydrogenation reactors. In his book 'Mass Transfer in Heterogeneous Catalysis' (1970) Satterfield complained of the lack of knowledge about the design of slurry reactors and hence of the impossible task of the engineer who has to design a plant according to accepted rules. There have been no fundamental changes since then. This paper presents the problems facing the engineer in designing slurry reactors, and shows new development trends.

  13. Investigation of conditions inside the reactor building annulus of a PWR plant of KONVOI type in case of severe accidents with increased containment leakages

    Energy Technology Data Exchange (ETDEWEB)

    Bakalov, Ivan [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Berlin (Germany); Sonnenkalb, Martin [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Koeln (Germany)

    2018-02-15

    Improvements of the implemented severe accident management (SAM) concepts have been done in all operating German NPPs after the Fukushima Daiichi accidents following recommendations of the German Reactor Safety Commission (RSK) and as a result of the stress test being performed. The efficiency of newly developed severe accident management guidelines (SAMG) for a PWR KONVOI reference plant related to the mitigation of challenging conditions inside the reactor building (RB) annulus due to increased containment leakages during severe accidents have been assessed. Based on two representative severe accident scenarios the releases of both hydrogen and radionuclides into the RB annulus have been predicted with different boundary conditions. The accident scenarios have been analysed without and with the impact of several SAM measures (already planned or proposed in addition), which turned out to be efficient to mitigate the consequences. The work was done within the frame of a research project financially supported by the Federal Ministry BMUB.

  14. Investigation of conditions inside the reactor building annulus of a PWR plant of KONVOI type in case of severe accidents with increased containment leakages

    International Nuclear Information System (INIS)

    Bakalov, Ivan; Sonnenkalb, Martin

    2018-01-01

    Improvements of the implemented severe accident management (SAM) concepts have been done in all operating German NPPs after the Fukushima Daiichi accidents following recommendations of the German Reactor Safety Commission (RSK) and as a result of the stress test being performed. The efficiency of newly developed severe accident management guidelines (SAMG) for a PWR KONVOI reference plant related to the mitigation of challenging conditions inside the reactor building (RB) annulus due to increased containment leakages during severe accidents have been assessed. Based on two representative severe accident scenarios the releases of both hydrogen and radionuclides into the RB annulus have been predicted with different boundary conditions. The accident scenarios have been analysed without and with the impact of several SAM measures (already planned or proposed in addition), which turned out to be efficient to mitigate the consequences. The work was done within the frame of a research project financially supported by the Federal Ministry BMUB.

  15. Construction of an external electrode for determination of electrochemical corrosion potential in normal operational conditions of an BWR type reactor for hot cells

    International Nuclear Information System (INIS)

    Aguilar T, J.A.; Rivera M, H.; Hernandez C, R.

    2001-01-01

    The behavior of the corrosion processes at high temperature requires of external devices that being capable to resist a temperature of 288 Centigrade and a pressure of 80 Kg/cm 2 , to give stable and reproducible results of some variable and resisting physically and chemically the radiation. The external electrode of Ag/AgCl fulfils all the requirements in the determination of the electrochemical corrosion potential under normal operational conditions of a BWR type reactor in hot cells. (Author)

  16. Research on reactor physics data

    International Nuclear Information System (INIS)

    1961-01-01

    In the early years of nuclear reactor research, each national program tended to develop its own reactor physics information. The Government of Norway proposed to the Agency the undertaking of a joint program in reactor physics utilizing the facilities and staff of its zero power reactor NORA then under construction. Following the approval by the Board of Governors in February, the Agency invited Member States to submit the names and qualifications of scientists they wished to suggest for the project. All the results and information gained through the program, which is expected to last about three years, will be placed at the disposal of the Agency's Member States

  17. Fluid structure interaction studies on acoustic load response of light water nuclear reactor core internals under blowdown condition

    International Nuclear Information System (INIS)

    Moses Lemuel Raj, G.; Singh, R.K.; Kushwaha, H.S.; Venkat Raj, V.

    1998-12-01

    Acoustic load evaluation within two phase medium and the related fluid-structure interaction analysis in case of Loss of Coolant Accidents (LOCA) for light water reactor systems is an important inter-disciplinary area. The present work highlights the development of a three-dimensional finite element code FLUSHEL to analyse LOCA induced depressurization problems for Pressurised Water Reactor (PWR) core barrel and Boiling Water Reactor (BWR) core shroud. With good comparison obtained between prediction made by the present code and the experimental results of HDR-PWR test problem, coupled fluid-structure interaction analysis of core shroud of Tarapur Atomic Power Station (TAPS) is presented for recirculation line break. It is shown that the acoustic load induced stresses in the core shroud are small and downcomer acoustic cavity modes are decoupled with the shell multi-lobe modes. Thus the structural integrity of TAPS core shroud for recirculation line break induced acoustic load is demonstrated. (author)

  18. Kinetic characterization for hemicellulose hydrolysis of corn stover in a dilute acid cycle spray flow-through reactor at moderate conditions

    International Nuclear Information System (INIS)

    Jin, Qiang; Zhang, Hongman; Yan, Lishi; Qu, Liang; Huang, He

    2011-01-01

    The kinetic characterization of hemicellulose hydrolysis of corn stover was investigated using a new reactor of dilute acid cycle spray flow-through (DCF) pretreatment. The primary purpose was to obtain kinetic data for hemicellulose hydrolysis with sulfuric acid concentrations (10-30 kg m -3 ) at relatively low temperatures (90-100 o C). A simplified kinetic model was used to describe its performance at moderate conditions. The results indicate that the rates of xylose formation and degradation are sensitive to flow rate, temperature and acid concentration. Moreover, the kinetic data of hemicellulose hydrolysis fit a first-order reaction model and the experimental data with actual acid concentration after accounting for the neutralization effect of the substrates at different temperatures. Over 90% of the xylose monomer yield and below 5.5% of degradation product (furfural) yield were observed in this reactor. Kinetic constants for hemicellulose hydrolysis models were analyzed by an Arrhenius-type equation, and the activation energy of xylose formation were 111.6 kJ mol -1 , and 95.7 kJ mol -1 for xylose degradation, respectively. -- Highlights: → Investigating a novel pretreatment reactor of dilute acid cycle spray flow-through. → Xylose yield is sensitive to flow rate, temperature and acid concentration. → Obtaining relatively higher xylose monomer yield and lower fermentation inhibitor. → Lumping hemicellulose and xylan oligmers together in the model is a valid way. → The kinetic model as a guide for reactor design, and operation strategy optimization.

  19. Invitation to View Method with Advantages and Disadvantages

    Directory of Open Access Journals (Sweden)

    Selçuk BUYRUKOĞLU

    2018-01-01

    Full Text Available Taxpayers and tax administrators sometimes may experience tax dispute. These disputes may arise from the taxpayers as well as from the tax administrators. In resolving tax disputes, tax administrators and tax payers are the first, usually administratively, solution. Because while the tax revenues of the state are possible entered into the state treasury, the taxpayers will benefit from the advantages of administrative solutions. The state and taxpayer herein will be the winner. Invitation to view is one of the administrative remedies for tax disputes. With this practice taxpayers who have caused tax evasion penalty are invited by tax administrators to make a declaration on the condition that the tax examination of the tax disputes is not started and the dispatch process has not been done to the discretion comission. In this study, advantages and disadvantages will be addressed after the definition, legal support, purpose, scope and mechanism of Invitation to view.

  20. A standalone decay heat removal device for the Gas-cooled Fast Reactor for intermediate to atmospheric pressure conditions

    Energy Technology Data Exchange (ETDEWEB)

    Epiney, A., E-mail: aaron@epiney.ch [Paul Scherrer Institute PSI, Villigen (Switzerland); Ecole Polytechnique Federale EPFL, Lausanne (Switzerland); Alpy, N., E-mail: nicolas.alpy@cea.fr [CEA, DEN, Service d' Etudes des Systemes Innovants, F-13108 Saint Paul Lez Durance (France); Mikityuk, K., E-mail: konstantin.mikityuk@psi.ch [Paul Scherrer Institute PSI, Villigen (Switzerland); Chawla, R., E-mail: rakesh.chawla@psi.ch [Paul Scherrer Institute PSI, Villigen (Switzerland); Ecole Polytechnique Federale EPFL, Lausanne (Switzerland)

    2012-01-15

    Highlights: Black-Right-Pointing-Pointer An analytical model predicting Brayton cycle off-design steady states, is developed. Black-Right-Pointing-Pointer The model is used to design an autonomous decay heat removal system for the GFR. Black-Right-Pointing-Pointer Predictions of the analytical model are verified using CATHARE. Black-Right-Pointing-Pointer CATHARE code is used to simulate a set of GFR safety depressurization transients using this device. Black-Right-Pointing-Pointer Convenient turbo-machine designs exist for the targeted autonomous decay heat removal for a wide pressure range. - Abstract: This paper reports a design study for a Brayton cycle machine, which would constitute a dedicated, standalone decay heat removal (DHR) device for the Generation IV Gas-cooled Fast Reactor (GFR). In comparison to the DHR reference strategy developed by the French Commissariat a l'Energie Atomique during the GFR pre-conceptual design phase (which was completed at the end of 2007), the salient feature of this alternative device would be to combine the energetic autonomy of the natural convection process - which is foreseen for operation at high and medium pressures - with the efficiency of the forced convection process which is foreseen for operation down to very low pressures. An analytical model, the so-called 'Brayton scoping model', is described first. This is based on simplified thermodynamic and aerodynamic equations, and was developed to highlight design choices. Two different machine designs are analyzed: a Brayton loop turbo-machine working with helium, and a second one working with nitrogen, since nitrogen is the heavy gas foreseen to be injected into the primary system to enhance the natural convection under loss-of-coolant-accident (LOCA) conditions. Simulations of the steady-state and transient behavior of the proposed device have then been carried out using the CATHARE code. These serve to confirm the insights obtained from usage of the

  1. Preliminary results of the BTF-104 experiment: an in-reactor test of fuel behaviour and fission-product release and transport under LOCA/LOECC conditions

    Energy Technology Data Exchange (ETDEWEB)

    Dickson, L W; Elder, P H; Devaal, J W; Irish, J D; Yamazaki, A R [Atomic Energy of Canada Ltd., Chalk River, ON (Canada)

    1996-12-31

    The BTF-104 experiment is one of a series of in-reactor tests being performed to measure fuel behaviour and fission-product release from nuclear fuel subjected to accident conditions. The primary objective of the BTF-104 experiment was to measure fission-product releases from a CANDU-sized fuel element under combined Loss-of-Coolant Accident (LOCA) and Loss-of-Emergency-Core-Cooling (LOECC) conditions at an average fuel temperature of about 1550 deg C. The preliminary results of the BTF-104 experiment are presented in this paper. (author). 6 refs., 12 figs.

  2. Vibration analysis of primary inlet pipe line during steady state and transient conditions of Pakistan research reactor-1

    International Nuclear Information System (INIS)

    Ayazuddin, S.K.; Qureshi, A.A.; Hayat, T.

    1997-11-01

    The Primary Water Inlet Pipeline (PW-IPL) is of stainless steel conveying demineralized water from hold-up tank to the reactor pool of Pakistan Research Reactor-1 (PARR-1). The section of the pipeline from heat exchangers to the valve pit is hanger supported in the pump room and the rest of the section from valve pit to the reactor pool is embedded. The PW-IPL is subjected to steady state and transient vibrations. The reactor pumps, which drive the coolant through various circuits mainly contribute the steady state vibrations, while transient vibrations arise due to instant closure of the check valve (water hammer). The ASME Boiler and Pressure Vessel code provides data about the acceptable limits of stresses related to the primary static stress due to steady state vibrations. However, due to complexity in the pipe structure, stresses related to the transient vibrations are neglected in the code. In this report attempt has been made to analyzed both steady state and transient vibrations of PW-IPL of PARR-1. Since, both the steady state and transient vibrations affect the hanger-supported section of the PW-IPL, therefore, it was selected for vibration test measurements. In the analysis vibration data was compared with the allowable limits and estimations of maximum pressure build-up, eflection, natural frequency, tensile and shear load on hanger support, and the ratio of maximum combine stress to the allowable load were made. (author)

  3. CHF experiments of tight pitch lattice rod bundles under PWR pressure condition for development of reduced moderation water reactor

    International Nuclear Information System (INIS)

    Araya, Fumimasa; Nakatsuka, Toru; Yoritsune, Tsutomu

    2002-10-01

    In order to improve plutonium utilization, design studies of reduced moderation water reactors which have hard neutron energy spectrum have been carried out at Division of Energy System Research of Japan Atomic Energy Research Institute (JAERI). At present, triangle, tight pitch lattice cores with about 1 mm gap width between fuel rods have been focused in the neutronic core design. Since a degradation of the heat removal from the fuel rods is worried, an evaluation of heat removal capability i.e. critical heat flux becomes one of important evaluation items in the feasibility study. However, any of published data base, which can be applicable to the evaluation on such narrow gap width cores, does not exist. Therefore, in the present study, in order to accumulate applicable data and to confirm applicability of an evaluation methodology of critical heat flux, basic experiments on the critical heat flux were performed using the test sections consisted of 7 heater rods bundles with the gap widths of 1.5, 1.0 and 0.6 mm under the PWR pressure conditions. The present report describes the experimental apparatus, experimental conditions and accumulated data. Analysis results of the data and the applicability of the evaluation methodology used for the design work are also discussed in this report. As the results of the experiment, it was found that the critical heat flux increased as the mass flux and the inlet subcooling increased. In the region of the mass flux less than about 2,000 kg/m 2 /s, the critical heat flux decreased as the gap width decreased. In the larger mass flux region, obvious trend of effects of the gap width on critical heat flux were not observed due to data scatterings. The flow-area-averaged thermal-equilibrium quality at the CHF position was in the higher ranges from 0.3 to 0.8 in the cases of gap widths of 1.0 and 0.6 mm, and 0.1 to 0.3 in the 1.5 mm case. Based on the experimental results such that the CHFs occurred in the higher quality range and

  4. 77 FR 1618 - Special Conditions: The Boeing Company, Model 767-300; Seats With Inflatable Lapbelts

    Science.gov (United States)

    2012-01-11

    ... energy absorption throughout that range of occupants. In addition, the following situations must be... these special conditions effective upon issuance. Comments Invited We invite interested people to take...

  5. Reactor network synthesis for isothermal conditions = Síntese de redes de reatores para condições isotérmicas

    Directory of Open Access Journals (Sweden)

    Lincoln Kotsuka da Silva

    2008-07-01

    Full Text Available In the present paper, a computational systematic procedure for isothermal Reactor Network Synthesis (RNS is presented. A superstructure of ideal CSTR and PFR reactors is proposed and the model is formulated as a constrained Nonlinear Programming (NLP problem. Complex reactions (series/parallel reactions are considered. The objective function is based on yield or selectivity, depending on the desired product, subject to different operational conditions. The problem constraints are mass balances in the reactorsand in the considered reactor network superstructure. A systematic computational procedure is proposed and a Genetic Algorithm (GA is developed to obtain the optimal reactor arrangement with the maximum yield or selectivity and minimum reactor volume. Results are as good as or better than those reported in the literature.No presentetrabalho apresenta-se um procedimento computacional para síntese de redes de reatores (SRR operando em condições isotérmicas. Uma superestrutura de rede de reatores formada por reatores ideais CSTR e PFR é proposta e o problema apresenta uma formulação de programação não linear (PNL. São consideradas reações complexas (série/paralelas. A função objetivo é baseada no rendimento ou na seletividade em relação ao produto desejado, sujeito a diferentes condições de operação. As restrições ao problema são provenientes dos balanços de massa e da configuração da superestrutura considerada.No procedimento computacional é proposto um Algoritmo Genético (AG para obtenção do arranjo ótimo de reatores com máximo rendimento ou seletividade com menor volume reacional. Os resultados obtidos são condizentes com os obtidos na literatura.

  6. The long-term behaviour of cemented research reactor waste under the geological disposal conditions of the Boom Clay Formation: results from leach experiments

    International Nuclear Information System (INIS)

    Sneyers, A.; Fays, J.; Iseghem, P. van

    2001-01-01

    The Belgian Nuclear Research Centre SCK-CEN has carried out a number of studies to evaluate the long-term behaviour of cemented research reactor waste under the geological disposal conditions of the Boom Clay Formation. Static leach experiments in synthetic clay water were performed on active samples of cemented research reactor waste. The leach experiments were carried out under anaerobic conditions at two testing temperatures (23 and 85 o C). Leach rates of seven radionuclides ( 60 Co, 90 Sr, 134 Cs, 137 Cs, 144 Ce, 154 Eu and 241 Am) were measured. Most investigated radionuclides are well retained within the cement matrix over a 280 days testing period. Results on the source term of radionuclides were complemented with data on the leaching behaviour of cement matrix constituents as Ca, Si, Al, Na, K, Mg and SO 4 as well as with data from performance assessment calculations and in situ tests. Despite limitations inherent to short-term experiments, combined results from these investigations indicate only limited interactions of disposed research reactor waste with the near field of a geological repository in clay. (author)

  7. The influence of ppb levels of chloride impurities on the stress corrosion crack growth behaviour of low-alloy steels under simulated boiling water reactor conditions

    International Nuclear Information System (INIS)

    Seifert, H.P.; Ritter, S.

    2016-01-01

    Highlights: • Chloride effects on SCC crack growth in RPV steels under boiling water reactor conditions. • ppb-levels of chloride may result in fast SCC in normal water chemistry environment. • Much higher chloride tolerance for SCC in hydrogen water chemistry environment. • Potential long-term (memory) effects after severe and prolonged temporary chloride transients. - Abstract: The effect of chloride on the stress corrosion crack (SCC) growth behaviour in low-alloy reactor pressure vessel steels was evaluated under simulated boiling water reactor conditions. In normal water chemistry environment, ppb-levels of chloride may result in fast SCC after rather short incubation periods of few hours. After moderate and short-term chloride transients, the SCC crack growth rates return to the same very low high-purity water values within few 100 h. Potential long-term (memory) effects on SCC crack growth cannot be excluded after severe and prolonged chloride transients. The chloride tolerance for SCC in hydrogen water chemistry environment is much higher.

  8. Safety analysis of increase in heat removal from reactor coolant system with inadvertent operation of passive residual heat removal at no load conditions

    Energy Technology Data Exchange (ETDEWEB)

    Shao, Ge; Cao, Xuewu [School of Mechanical and Engineering, Shanghai Jiao Tong University, Shanghai (China)

    2015-06-15

    The advanced passive pressurized water reactor (PWR) is being constructed in China and the passive residual heat removal (PRHR) system was designed to remove the decay heat. During accident scenarios with increase of heat removal from the primary coolant system, the actuation of the PRHR will enhance the cooldown of the primary coolant system. There is a risk of power excursion during the cooldown of the primary coolant system. Therefore, it is necessary to analyze the thermal hydraulic behavior of the reactor coolant system (RCS) at this condition. The advanced passive PWR model, including major components in the RCS, is built by SCDAP/RELAP5 code. The thermal hydraulic behavior of the core is studied for two typical accident sequences with PRHR actuation to investigate the core cooling capability with conservative assumptions, a main steam line break (MSLB) event and inadvertent opening of a steam generator (SG) safety valve event. The results show that the core is ultimately shut down by the boric acid solution delivered by Core Makeup Tank (CMT) injections. The effects of CMT boric acid concentration and the activation delay time on accident consequences are analyzed for MSLB, which shows that there is no consequential damage to the fuel or reactor coolant system in the selected conditions.

  9. The experimental testing of the long-term behaviour of cemented radioactive waste from nuclear research reactors in the geological disposal conditions of the boom clay

    International Nuclear Information System (INIS)

    Sneyers, A.; Marivoet, J.; Iseghem, P. van

    1998-01-01

    Liquid wastes, resulting from the reprocessing of spent nuclear fuel from the BR-2 Materials Testing Reactor, will be conditioned in a cement matrix at the dedicated cementation facility of UKAEA at Dounreay. In Belgium, the Boom clay formation is studied as a potential host rock for the final geological disposal of cemented research reactor waste. In view of evaluating the safety of disposal, laboratory leach experiments and in situ tests have been performed. Leach experiments in synthetic clay water indicate that the leach rates of calcium and silicium are relatively low compared to those of sodium and potassium. In situ experiments on inactive samples are performed in order to obtain information on the microchemical and mineralogical changes of the cemented waste in contact with the Boom clay. Finally, results from a preliminary performance assessment calculation suggest a non-negligible maximum dose rate of 5 10 -9 Sv/a for 129 I. (author)

  10. The experimental testing of the long-term behaviour of cemented radioactive waste from nuclear research reactors in the geological disposal conditions of the boom clay

    Energy Technology Data Exchange (ETDEWEB)

    Sneyers, A.; Marivoet, J.; Iseghem, P. van [SCK-CEN, B-2400 Mol (Belgium)

    1998-07-01

    Liquid wastes, resulting from the reprocessing of spent nuclear fuel from the BR-2 Materials Testing Reactor, will be conditioned in a cement matrix at the dedicated cementation facility of UKAEA at Dounreay. In Belgium, the Boom clay formation is studied as a potential host rock for the final geological disposal of cemented research reactor waste. In view of evaluating the safety of disposal, laboratory leach experiments and in situ tests have been performed. Leach experiments in synthetic clay water indicate that the leach rates of calcium and silicium are relatively low compared to those of sodium and potassium. In situ experiments on inactive samples are performed in order to obtain information on the microchemical and mineralogical changes of the cemented waste in contact with the Boom clay. Finally, results from a preliminary performance assessment calculation suggest a non-negligible maximum dose rate of 5 10{sup -9} Sv/a for {sup 129}I. (author)

  11. Inviting In the Private Sector

    Science.gov (United States)

    2002-10-01

    private sector . After all, corruption does not just involve government. Business people and lawyers and citizens pay the bribes, even as they condemn bribery. They should be invited to become part of the solution. But how? The first point to note is that business people and citizens know where corruption exists and how corrupt systems work. Citizens understand how bribery shapes the services they receive or don’t receive. Accountants know the illicit games played with audits and taxes. Lawyers understand corrupt legal practices. Business people know all about corrupt

  12. Biofiltration of a styrene/acetone vapor mixture in two reactor types under conditions of styrene overloading

    Directory of Open Access Journals (Sweden)

    Lubos Zapotocky

    2014-10-01

    Full Text Available This aim of study was to compare the performance of a biofilter (BF and trickle bed reactor (TBR under increased styrene loading with a constant acetone load, 2 gc/m3/h. At styrene loading rates up to 30 gc/m3/h, the BF showed higher styrene removal than TBR. However, the BF efficiency started to drop beyond this threshold loading and could never reach steady state, whereas the TBR continued to yield a 50% styrene removal. The acetone removal remained constant (93-98% in both the reactors at any styrene loading. Once the overloading was lifted, the BF recovered within 26 min, whereas the TBR efficiency bounced back only to 95%, gradually returning to complete removal only in 10 h.

  13. VICTORIA: A mechanistic model of radionuclide behavior in the reactor coolant system under severe accident conditions. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Heams, T J [Science Applications International Corp., Albuquerque, NM (United States); Williams, D A; Johns, N A; Mason, A [UKAEA, Winfrith, (England); Bixler, N E; Grimley, A J [Sandia National Labs., Albuquerque, NM (United States); Wheatley, C J [UKAEA, Culcheth (England); Dickson, L W [Atomic Energy of Canada Ltd., Chalk River, ON (Canada); Osborn-Lee, I [Oak Ridge National Lab., TN (United States); Domagala, P; Zawadzki, S; Rest, J [Argonne National Lab., IL (United States); Alexander, C A [Battelle, Columbus, OH (United States); Lee, R Y [Nuclear Regulatory Commission, Washington, DC (United States)

    1992-12-01

    The VICTORIA model of radionuclide behavior in the reactor coolant system (RCS) of a light water reactor during a severe accident is described. It has been developed by the USNRC to define the radionuclide phenomena and processes that must be considered in systems-level models used for integrated analyses of severe accident source terms. The VICTORIA code, based upon this model, predicts fission product release from the fuel, chemical reactions involving fission products, vapor and aerosol behavior, and fission product decay heating. Also included is a detailed description of how the model is implemented in VICTORIA, the numerical algorithms used, and the correlations and thermochemical data necessary for determining a solution. A description of the code structure, input and output, and a sample problem are provided.

  14. X-ray and pressure conditions on the first wall of a particle beam inertial confinement reactor

    International Nuclear Information System (INIS)

    Magelssen, G.R.

    1979-01-01

    Because of the presence of a chamber gas in a particle beam reactor cavity, nonneutron target debris created from thermonuclear burn will be modified or stopped before it reaches the first reactor wall. The resulting modified spectra and pulse lengths of the debris need to be calculated to determine first wall effects. Further, the cavity overpressure created by the momentum and energy exchange between the debris and gas must also be calculated to determine its effect. The purpose of this paper is to present results of the debris-background gas problem obtained with a one fluid, two temperature plasma hydrodynamic computer code model which includes multifrequency radiation transport. Spherical symmetry, ideal gas equation of state, and LTE for each radiation frequency group were assumed. The transport of debris ions was not included and all the debris energy was assumed to be in radiation. The calculated x-ray spectra and pulse lengths and the background overpressure are presented

  15. New measurement system for on line in core high-energy neutron flux monitoring in materials testing reactor conditions

    Science.gov (United States)

    Geslot, B.; Vermeeren, L.; Filliatre, P.; Lopez, A. Legrand; Barbot, L.; Jammes, C.; Bréaud, S.; Oriol, L.; Villard, J.-F.

    2011-03-01

    Flux monitoring is of great interest for experimental studies in material testing reactors. Nowadays, only the thermal neutron flux can be monitored on line, e.g., using fission chambers or self-powered neutron detectors. In the framework of the Joint Instrumentation Laboratory between SCK-CEN and CEA, we have developed a fast neutron detector system (FNDS) capable of measuring on line the local high-energy neutron flux in fission reactor core and reflector locations. FNDS is based on fission chambers measurements in Campbelling mode. The system consists of two detectors, one detector being mainly sensitive to fast neutrons and the other one to thermal neutrons. On line data processing uses the CEA depletion code DARWIN in order to disentangle fast and thermal neutrons components, taking into account the isotopic evolution of the fissile deposit. The first results of FNDS experimental test in the BR2 reactor are presented in this paper. Several fission chambers have been irradiated up to a fluence of about 7 × 1020 n/cm2. A good agreement (less than 10% discrepancy) was observed between FNDS fast flux estimation and reference flux measurement.

  16. New measurement system for on line in core high-energy neutron flux monitoring in materials testing reactor conditions

    Energy Technology Data Exchange (ETDEWEB)

    Geslot, B.; Filliatre, P.; Barbot, L.; Jammes, C.; Breaud, S.; Oriol, L.; Villard, J.-F. [CEA, DEN, Cadarache, SPEx/LDCI, F-13108 Saint-Paul-lez-Durance (France); Vermeeren, L. [SCK-CEN, Boeretang 200, B-2400 Mol (Belgium); Lopez, A. Legrand [CEA, DEN, Saclay, SIREN/LECSI, F-91400 Saclay (France)

    2011-03-15

    Flux monitoring is of great interest for experimental studies in material testing reactors. Nowadays, only the thermal neutron flux can be monitored on line, e.g., using fission chambers or self-powered neutron detectors. In the framework of the Joint Instrumentation Laboratory between SCK-CEN and CEA, we have developed a fast neutron detector system (FNDS) capable of measuring on line the local high-energy neutron flux in fission reactor core and reflector locations. FNDS is based on fission chambers measurements in Campbelling mode. The system consists of two detectors, one detector being mainly sensitive to fast neutrons and the other one to thermal neutrons. On line data processing uses the CEA depletion code DARWIN in order to disentangle fast and thermal neutrons components, taking into account the isotopic evolution of the fissile deposit. The first results of FNDS experimental test in the BR2 reactor are presented in this paper. Several fission chambers have been irradiated up to a fluence of about 7 x 10{sup 20} n/cm{sup 2}. A good agreement (less than 10% discrepancy) was observed between FNDS fast flux estimation and reference flux measurement.

  17. New measurement system for on line in core high-energy neutron flux monitoring in materials testing reactor conditions

    International Nuclear Information System (INIS)

    Geslot, B.; Filliatre, P.; Barbot, L.; Jammes, C.; Breaud, S.; Oriol, L.; Villard, J.-F.; Vermeeren, L.; Lopez, A. Legrand

    2011-01-01

    Flux monitoring is of great interest for experimental studies in material testing reactors. Nowadays, only the thermal neutron flux can be monitored on line, e.g., using fission chambers or self-powered neutron detectors. In the framework of the Joint Instrumentation Laboratory between SCK-CEN and CEA, we have developed a fast neutron detector system (FNDS) capable of measuring on line the local high-energy neutron flux in fission reactor core and reflector locations. FNDS is based on fission chambers measurements in Campbelling mode. The system consists of two detectors, one detector being mainly sensitive to fast neutrons and the other one to thermal neutrons. On line data processing uses the CEA depletion code DARWIN in order to disentangle fast and thermal neutrons components, taking into account the isotopic evolution of the fissile deposit. The first results of FNDS experimental test in the BR2 reactor are presented in this paper. Several fission chambers have been irradiated up to a fluence of about 7 x 10 20 n/cm 2 . A good agreement (less than 10% discrepancy) was observed between FNDS fast flux estimation and reference flux measurement.

  18. Program of assessment of mechanical and corrosion mechanical properties of reactor internals materials due to operation conditions in WWERs

    International Nuclear Information System (INIS)

    Ruscak, M.; Zamboch, M.

    1998-01-01

    Reactor internals are subject to three principle operation influences: neutron and gamma irradiation, mechanical stresses, both static and dynamic, and coolant chemistry. Several cases of damage have been reported in previous years in both boiling and pressure water reactors. They are linked with the term of irradiation assisted stress corrosion cracking as a possible damage mechanism. In WWERs, the principal material used for reactor internals is austenitic titanium stabilized stainless steel 08Kh18N10T, however high strength steels are used as well. To assess the changes of mechanical properties and to determine whether sensitivity to intergranular cracking can be increased by high neutron fluences, the experimental program has been started. The goal is to assure safe operation of the internals as well as life management for all planned operation period. The program consists of tests of material properties, both mechanical and corrosion-mechanical. Detailed neutron fluxes calculation as well as stress and deformation calculations are part of the assessment. Model of change will be proposed in order to plan inspections of the facility. In situ measurements of internals will be used to monitor exact status of structure during operation. Tensile specimens manufactured from both base metal and model weld joint have been irradiated to the total fluences of 3-20 dpa. Changes of mechanical properties are tested by the tensile test, stress corrosion cracking tests are performed in the autoclave with water loop and active loading. Operation temperature, pressure and water chemistry are chosen for the tests. (author)

  19. Demand-driven biogas production from sugar beet silage in a novel fixed bed disc reactor under mesophilic and thermophilic conditions.

    Science.gov (United States)

    Terboven, Christiane; Ramm, Patrice; Herrmann, Christiane

    2017-10-01

    A newly developed fixed bed disc reactor (FBDR) which combines biofilm formation on biofilm carriers and reactor agitation in one single system was assessed for its applicability to demand-driven biogas production by variable feeding of sugar beet silage. Five different feeding patterns were studied at an organic loading of 4g VS L -1 d -1 under mesophilic and thermophilic conditions. High methane yields of 449-462L N kg VS were reached. Feeding variable punctual loadings caused immediate response with 1.2- to 3.5-fold increase in biogas production rates within 15min. Although variable feeding did not induce process instability, a temporary decrease in pH-value and methane concentration below 40% occurred. Thermophilic temperature was advantageous as it resulted in a more rapid, higher methane production and less pronounced decrease in methane content after feeding. The FBDR was demonstrated to be well-suited for flexible biogas production, but further research and comparison with traditional reactor systems are required. Copyright © 2017 Elsevier Ltd. All rights reserved.

  20. Prokaryotic diversity and dynamics in a full-scale municipal solid waste anaerobic reactor from start-up to steady-state conditions.

    Science.gov (United States)

    Cardinali-Rezende, Juliana; Colturato, Luís F D B; Colturato, Thiago D B; Chartone-Souza, Edmar; Nascimento, Andréa M A; Sanz, José L

    2012-09-01

    The prokaryotic diversity of an anaerobic reactor for the treatment of municipal solid waste was investigated over the course of 2 years with the use of 16S rDNA-targeted molecular approaches. The fermentative Bacteroidetes and Firmicutes predominated, and Proteobacteria, Actinobacteria, Tenericutes and the candidate division WWE1 were also identified. Methane production was dominated by the hydrogenotrophic Methanomicrobiales (Methanoculleus sp.) and their syntrophic association with acetate-utilizing and propionate-oxidizing bacteria. qPCR demonstrated the predominance of the hydrogenotrophic over aceticlastic Methanosarcinaceae (Methanosarcina sp. and Methanimicrococcus sp.), and Methanosaetaceae (Methanosaeta sp.) were measured in low numbers in the reactor. According to the FISH and CARD-FISH analyses, Bacteria and Archaea accounted for 85% and 15% of the cells, respectively. Different cell counts for these domains were obtained by qPCR versus FISH analyses. The use of several molecular tools increases our knowledge of the prokaryotic community dynamics from start-up to steady-state conditions in a full-scale MSW reactor. Copyright © 2012 Elsevier Ltd. All rights reserved.

  1. Oxy-coal combustion in an entrained flow reactor: Application of specific char and volatile combustion and radiation models for oxy-firing conditions

    DEFF Research Database (Denmark)

    Álvarez, L.; Yin, Chungen; Riaza, J.

    2013-01-01

    The deployment of oxy-fuel combustion in utility boilers is one of the major options for CO2 capture. However, combustion under oxy-firing conditions differs from conventional air-firing combustion, e.g., in the aspect of radiative heat transfer, coal conversion and pollutants formation....... In this work, a numerical study on pulverised coal combustion was conducted to verify the applicability and accuracy of several sub-models refined for oxy-fuel conditions, e.g., gaseous radiative property model, gas-phase combustion mechanism and heterogeneous char reaction model. The sub-models were...... implemented in CFD (Computational Fluid Dynamics) simulations of combustion of three coals under air-firing and various oxy-firing (21-35% vol O2 in O2/CO2 mixture) conditions in an EFR (entrained flow reactor). The predicted coal burnouts and gaseous emissions were compared against experimental results...

  2. Effects of operating conditions on molten-salt electrorefining for zirconium recovery from irradiated Zircaloy-4 cladding of pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jaeyeong, E-mail: d486916@snu.ac.kr [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 151-742 (Korea, Republic of); Choi, Sungyeol [Korea Atomic Energy Research Institute, 1045 Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Sohn, Sungjune [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 151-742 (Korea, Republic of); Kim, Kwang-Rag [Korea Atomic Energy Research Institute, 1045 Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Hwang, Il Soon [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 151-742 (Korea, Republic of)

    2014-08-15

    Highlights: • Computational simulation on electrorefining of irradiated Zircaloy-4 cladding. • Composition of irradiated Zircaloy-4 cladding of pressurized water reactor. • Redox behavior of elements in irradiated Zircaloy cladding during electrorefining. • Effect of electrorefining operating conditions on decontamination factor. - Abstract: To reduce the final waste volume from used nuclear fuel assembly, it is significant to decontaminate irradiated cladding. Electrorefining in high temperature molten salt could be one of volume decontamination processes for the cladding. This study examines the effect of operating conditions on decontamination factor in electrorefining of irradiated Zircaloy-4 cladding of pressurized water reactor. One-dimensional time-dependent electrochemical reaction code, REFIN, was utilized for simulating irradiated cladding electrorefining. Composition of irradiated Zircaloy was estimated based on ORIGEN-2 and other literatures. Co and U were considered in electrorefining simulation with major elements of Zircaloy-4 to represent activation products and actinides penetrating into the cladding respectively. Total 240 cases of electrorefining are simulated including 8 diffusion boundary layer thicknesses, 10 concentrations of contaminated molten salt and 3 termination conditions. Decontamination factors for each case were evaluated and it is revealed that the radioactivity of Co-60 in recovered zirconium on cathode could decrease below the clearance level when initial concentration of chlorides except ZrCl{sub 4} is lower than 1 × 10{sup −11} weight fraction if electrorefining is finished before anode potential reaches −1.8 V (vs. Cl{sub 2}/Cl{sup −})

  3. On fission product retention in the core of the low powered high temperature reactor under accident conditions

    International Nuclear Information System (INIS)

    Bastek, H.

    1984-01-01

    In the core of the high temperature reactor the fuel element and the coated particles contained herein provide the safest enclosure for fission products. The complex process of fission product transport out of the particle kernel, through the particle coating and within the fuel element graphite is described in a simplified form by the Fick's diffusion. The effective diffusion coefficient is used for calculation. Starting from the existing ideas of fission product transport five burn-up and temperature-dependent diffusion coefficients for Cesium in (Th,U)O 2 -kernels are derived in this study. The results have been gained from several fuel element radiation experiments in recent years, which showed extreme variation in regard to burn-up, temperature cycle, neutron flux and operation time. Cs-137 release measurements from single particle kernels were present from all the experiments. Furthermore, annealing tests of AVR-fuel elements were analyzed. Heat-temperatur and heating-time, the fuel element burn-up in the AVR-reactor, as well as the measured Cs-137 inventory of the fuel elements before and after annealing, are included in the investigation as essential parameters. With the aid of the derived diffusion coeffizients and already present data sets the Cs-137 release of fuel elements into a small reactor core is investigated under unrestricted core heat-up. While the released Cs-137 is derived mainly from defective particles at accident temperatures up to 1600 0 C, the main part diffuses through the particle coating at higher accident temperatures. (orig./HP) [de

  4. Long-term ammonia removal in a coconut fiber-packed biofilter: analysis of N fractionation and reactor performance under steady-state and transient conditions.

    Science.gov (United States)

    Baquerizo, Guillermo; Maestre, Juan P; Machado, Vinicius C; Gamisans, Xavier; Gabriel, David

    2009-05-01

    A comprehensive study of long-term ammonia removal in a biofilter packed with coconut fiber is presented under both steady-state and transient conditions. Low and high ammonia loads were applied to the reactor by varying the inlet ammonia concentration from 90 to 260 ppm(v) and gas contact times ranging from 20 to 36 s. Gas samples and leachate measurements were periodically analyzed and used for characterizing biofilter performance in terms of removal efficiency (RE) and elimination capacity (EC). Also, N fractions in the leachate were quantified to both identify the experimental rates of nitritation and nitratation and to determine the N leachate distribution. Results showed stratification in the biofilter activity and, thus, most of the NH(3) removal was performed in the lower part of the reactor. An average EC of 0.5 kg N-NH(3)m(-3)d(-1) was obtained for the whole reactor with a maximum local average EC of 1.7 kg N-NH(3)m(-3)d(-1). Leachate analyses showed that a ratio of 1:1 of ammonium and nitrate ions in the leachate was obtained throughout steady-state operation at low ammonia loads with similar values for nitritation and nitratation rates. Low nitratation rates during high ammonia load periods occurred because large amounts of ammonium and nitrite accumulated in the packed bed, thus causing inhibition episodes on nitrite-oxidizing bacteria due to free ammonia accumulation. Mass balances showed that 50% of the ammonia fed to the reactor was oxidized to either nitrite or nitrate and the rest was recovered as ammonium indicating that sorption processes play a fundamental role in the treatment of ammonia by biofiltration.

  5. Natural fission reactors in the Franceville basin, Gabon: A review of the conditions and results of a open-quotes critical eventclose quotes in a geologic system

    International Nuclear Information System (INIS)

    Gauthier-Lafaye, F.; Holliger, P.; Blanc, P.L.

    1996-01-01

    Natural nuclear fission reactors are only known in two uranium deposits in the world, the Oklo and Bangombe deposits of the Franceville basin: Gabon. Since 1982, five new reactor zones have been discovered in these deposits and studied since 1989 in a cooperative European program. New geological, mineralogical, and geochemical studies have been carried out in order to understand the behavior of the actinides and fission products which have been stored in a geological environment for more than 2.0 Ga years. The Franceville basin and the uranium deposits remained geologically stable over a long period of time. Therefore, the sites of Oklo and Bangombe are well preserved. For the reactors, two main periods of actinide and radionuclides migration have been observed: during the criticality, under P-T conditions of 300 bars and 400-500 degrees C, respectively, and during a distention event which affected the Franceville basin 800 to 900 Ma ago and which was responsible for the intrusion of dolerite dikes close to the reactors. New isotopic analyses on uranium dioxides, clays, and phosphates allow us to determine their respective importance for the retention of fission products. The UO 2 matrix appears to be efficient at retaining most actinides and fission products such as REEs, Y, and Zr but not the volatile fission products (Cd, Cs, Xe, and Kr) nor Rb, Sr, and Ba. Some fissiogenic elements such as Mo, Tc, Ru, Rh, Pd, and Te could have formed metallic and oxide inclusion in the UO 2 matrix which are similar to those observed in artificial spent fuel. Clays and phosphate minerals also appear to have played a role in the retention of fissiogenic REEs and also of Pu. 82 refs., 21 figs., 12 tabs

  6. Unique rod lens/video system designed to observe flow conditions in emergency core coolant loops of pressurized water reactors

    International Nuclear Information System (INIS)

    Carter, G.W.

    1979-01-01

    Techniques and equipment are described which are used for video recordings of the single- and two-phase fluid flow tests conducted with the PKL Spool Piece Measurement System designed by Lawrence Livermore Laboratory and EG and G Inc. The instrumented spool piece provides valuable information on what would happen in pressurized water reactor emergency coolant loops should an accident or rupture result in loss of fluid. The complete closed-circuit television video system, including rod lens, light supply, and associated spool mounting fixtures, is discussed in detail. Photographic examples of test flows taken during actual spool piece system operation are shown

  7. Initial results from dissolution rate testing of N-Reactor spent fuel over a range of potential geologic repository aqueous conditions

    International Nuclear Information System (INIS)

    Gray, W.J.; Einziger, R.E.

    1998-04-01

    Hanford N-Reactor spent nuclear fuel (HSNF) may ultimately be placed in a geologic repository for permanent disposal. To determine whether the engineered barrier system that will be designed for emplacement of light-water-reactor (LWR) spent fuel will also suffice for HSNF, aqueous dissolution rate measurements were conducted on the HSNF. The purpose of these tests was to determine whether HSNF dissolves faster or slower than LWR spent fuel under some limited repository-relevant water chemistry conditions. The tests were conducted using a flowthrough method that allows the dissolution rate of the uranium matrix to be measured without interference by secondary precipitation reactions that would confuse interpretation of the results. Similar tests had been conducted earlier with LWR spent fuel, thereby allowing direct comparisons. Two distinct corrosion modes were observed during the course of these 12 tests. The first, Stage 1, involved no visible corrosion of the test specimen and produced no undissolved corrosion products. The second, Stage 2, resulted in both visible corrosion of the test specimen and left behind undissolved corrosion products. During Stage 1, the rate of dissolution could be readily determined because the dissolved uranium and associated fission products remained in solution where they could be quantitatively analyzed. The measured rates were much faster than has been observed for LWR spent fuel under all conditions tested to date when normalized to the exposed test specimen surface areas. Application of these results to repository conditions, however, requires some comparison of the physical conditions of the different fuels. The surface area of LWR fuel that could potentially be exposed to repository groundwater is estimated to be approximately 100 times greater than HSNF. Therefore, when compared on the basis of mass, which is more relevant to repository conditions, the HSNF and LWR spent fuel dissolve at similar rates

  8. An invitation to critical mathematics education

    DEFF Research Database (Denmark)

    Skovsmose, Ole

    An Invitation to Critical Mathematics Education deals with a range of crucial topics. Among these are students’ foreground, landscapes of investigation, and mathematics in action. The book is intended for a broad audience: educators, students, teachers, policy makers, anybody interested...... in the further development of mathematics education. The book discusses concerns and preoccupation. This way it provides an invitation into critical mathematics education....

  9. Generation IV reactors: reactor concepts

    International Nuclear Information System (INIS)

    Cardonnier, J.L.; Dumaz, P.; Antoni, O.; Arnoux, P.; Bergeron, A.; Renault, C.; Rimpault, G.; Delpech, M.; Garnier, J.C.; Anzieu, P.; Francois, G.; Lecomte, M.

    2003-01-01

    Liquid metal reactor concept looks promising because of its hard neutron spectrum. Sodium reactors benefit a large feedback experience in Japan and in France. Lead reactors have serious assets concerning safety but they require a great effort in technological research to overcome the corrosion issue and they lack a leader country to develop this innovative technology. In molten salt reactor concept, salt is both the nuclear fuel and the coolant fluid. The high exit temperature of the primary salt (700 Celsius degrees) allows a high energy efficiency (44%). Furthermore molten salts have interesting specificities concerning the transmutation of actinides: they are almost insensitive to irradiation damage, some salts can dissolve large quantities of actinides and they are compatible with most reprocessing processes based on pyro-chemistry. Supercritical water reactor concept is based on operating temperature and pressure conditions that infers water to be beyond its critical point. In this range water gets some useful characteristics: - boiling crisis is no more possible because liquid and vapour phase can not coexist, - a high heat transfer coefficient due to the low thermal conductivity of supercritical water, and - a high global energy efficiency due to the high temperature of water. Gas-cooled fast reactors combining hard neutron spectrum and closed fuel cycle open the way to a high valorization of natural uranium while minimizing ultimate radioactive wastes and proliferation risks. Very high temperature gas-cooled reactor concept is developed in the prospect of producing hydrogen from no-fossil fuels in large scale. This use implies a reactor producing helium over 1000 Celsius degrees. (A.C.)

  10. Upgrading of existing NPPs with 440 and 1000 MW WWER type pressurized water reactors for severe external loading conditions. Proceedings. Working material. V. 1, 2

    Energy Technology Data Exchange (ETDEWEB)

    Godoy, A; Gurpinar, A [eds.; International Atomic Energy Agency, Division of Nuclear Installation Safety, Vienna (Austria)

    1993-07-01

    The Seminar was intended to provide the opportunity for the exchange of updated information concerning the state-of-the-art related to structural safety of the WWER 440 and 1000 MW nuclear power plants operating or under construction in the Russian Federation and several Eastern European countries. Codes and standards, design assumptions, seismic upgrading and requalification activities, recent verification by experimental methods as well as the corresponding reanalysis and reevaluation procedures were included in the scope. Also presented and discussed were the results of recent studies of typical structures and components under seismic loading. The Seminar was divided into five working sessions. Each session consisted of several presentations by invited speakers and extensive discussions. The topic of discussions involved the following: standards used during design phase, input load definitions and criteria, evaluation of as-built conditions and walk down information, site-specific dynamic response of structures based on recent studies and evaluation of results, capacity evaluations of components and systems based on updated acceptance criteria and strengthening concepts, functional qualification of active mechanical and electrical components and systems.

  11. Upgrading of existing NPPs with 440 and 1000 MW WWER type pressurized water reactors for severe external loading conditions. Proceedings. Working material. V. 1, 2

    International Nuclear Information System (INIS)

    Godoy, A.; Gurpinar, A.

    1993-01-01

    The Seminar was intended to provide the opportunity for the exchange of updated information concerning the state-of-the-art related to structural safety of the WWER 440 and 1000 MW nuclear power plants operating or under construction in the Russian Federation and several Eastern European countries. Codes and standards, design assumptions, seismic upgrading and requalification activities, recent verification by experimental methods as well as the corresponding reanalysis and reevaluation procedures were included in the scope. Also presented and discussed were the results of recent studies of typical structures and components under seismic loading. The Seminar was divided into five working sessions. Each session consisted of several presentations by invited speakers and extensive discussions. The topic of discussions involved the following: standards used during design phase, input load definitions and criteria, evaluation of as-built conditions and walk down information, site-specific dynamic response of structures based on recent studies and evaluation of results, capacity evaluations of components and systems based on updated acceptance criteria and strengthening concepts, functional qualification of active mechanical and electrical components and systems

  12. Providing thermal-hydraulic boundary conditions to the reactor code TINTE through a Flownex-TINTE coupling - HTR2008-58110

    International Nuclear Information System (INIS)

    Marais, D.; Greyvenstein, G. P.

    2008-01-01

    TINTE is a well established reactor analysis code which models the transient behaviour of pebble bed reactor cores but it does not include the capabilities to model a power conversion unit (PCU). This raises the issue that TINTE cannot model full system transients. One way to overcome this problem is to supply TINTE with time-dependant thermal-hydraulic boundary conditions which are obtained from PCU simulations. This study investigates a method to provide boundary conditions for the nuclear code TINTE during full system transients. This was accomplished by creating a high level interface between the systems CFD code Flownex and TINTE. An indirect coupling method is explored whereby characteristics of the PCU are matched to characteristics of the nuclear core. This method eliminates the need to iterate between the two codes. A number of transients are simulated using the coupled code and then compared against stand-alone Flownex simulations. The coupling method introduces relatively small errors when reproducing mass flow, temperature and pressure in steady state analysis, but become more pronounced when dealing with fast thermal-hydraulic transients. Decreasing the maximum time step length of TINTE reduces this problem, but increases the computational time. Copyright ASME 2008. (authors)

  13. Conditioning of spent fuel assemblies from the Rossendorf RFR research reactor in transport and storage containers of the type CASTOR MTR 2

    International Nuclear Information System (INIS)

    Schneider, B.; Hofmann, G.

    1994-09-01

    Most of the spent fuel assemblies are temporarily stored in the flooded fuel ponds AB 1 and AB 2 of the RFR, and some are still in the reactor core. The conditioning task described here is part of the RFR spent fuel management concept and covers the safe emplacement of the spent fuel elements in the CASTOR MTR 2 shipping containers and the sealing of the containers in compliance with the nuclear licence issued for the conditioning task. The transfer of the spent fuel assemblies from the present wet storage conditions to the dry storage conditions in the CASTOR MTR 2 containers is done by a mobile manipulation equipment consisting essentially of the transfer sluice gate and a transfer container. Subsequent to conditioning, the shipping containers are to be transported to a licensed intermediate storage facility to await their transport to a national radwaste repository. The technical handling tools for the transfer and manipulation are briefly described, as well as the process steps involved, putting emphasis on the detailed description of processes and the accompanying time frame, so that the conditioning task can be incorporated into the work plan of the entire project. The report further presents the EDP concept established for the task, including the required data archivation and documentation. (orig.) [de

  14. Analysis of an out-of-pile experiment for materials redistribution under core disruptive accident condition of fast breeder reactors

    International Nuclear Information System (INIS)

    Sawada, Tetsuo; Ninokata, Hisashi; Shimizu, Akinao

    1995-01-01

    Calculation of one of the SIMBATH experiments was performed using the SIMMER-II code. The experiments were intended to simulate the fuel pin disintegration, the molten materials relocation and following materials redistribution that could occur during core disruptive accidents assumed in fast breeder reactors. The calculation by SIMMER-II showed that the incorporated step-wise fuel pin disintegration model and the modified particle jamming model were capable of reproducing the course of materials relocation within the identified ranges of the parameters which governed the blockages formation, i.e. the characteristic radius of solid particles jamming and/or sieving out in the flow and the effective particle viscosity. In particular the final materials redistribution calculated by SIMMER-II very well reproduced the experiment. This fact made it possible to interpret theoretically the mechanisms of flow blockages formation and related materials redistribution. (author)

  15. Optimal conditions and operational parameters for conversion of Robusta coffee residues in a continuous stirred tank reactor

    Energy Technology Data Exchange (ETDEWEB)

    Msambichaka, B L; Kivaisi, A K; Rubindamayugi, M S.T. [Univ. of Dar es Salaam, Applied Microbiology Unit (Tanzania, United Republic of)

    1998-12-31

    This experiment studied the possibility of optimizing anaerobic degradation, developing microbial adaptation and establishing long term process stability in a Continuous Stirred Tank Reactor (CSTR) running on Robusta coffee hulls as feed substrate. Decrease in lag phase and increase in methane production rate in batch culture experiment conducted before and after process stabilization of each operational phase in the CSTR clearly suggested that microbial adaptation to increasing coffee percentage composition was attained. Through gradual increase of coffee percentage composition, from 10% coffee, 2% VS, 20 days HRT and a 1 g VS/1/day loading rate to 80% coffee, 4.5% VS, 12 days HRT and a loading rate of 3 g VS/1/day the CSTR system was optimized at a maximum methane yield of 535 ml/g VS. Again it was possible to attain long term process stability at the above mentioned optimal operational parameters for a further 3 month period. (au)

  16. Optimal conditions and operational parameters for conversion of Robusta coffee residues in a continuous stirred tank reactor

    Energy Technology Data Exchange (ETDEWEB)

    Msambichaka, B.L.; Kivaisi, A.K.; Rubindamayugi, M.S.T. [Univ. of Dar es Salaam, Applied Microbiology Unit (Tanzania, United Republic of)

    1997-12-31

    This experiment studied the possibility of optimizing anaerobic degradation, developing microbial adaptation and establishing long term process stability in a Continuous Stirred Tank Reactor (CSTR) running on Robusta coffee hulls as feed substrate. Decrease in lag phase and increase in methane production rate in batch culture experiment conducted before and after process stabilization of each operational phase in the CSTR clearly suggested that microbial adaptation to increasing coffee percentage composition was attained. Through gradual increase of coffee percentage composition, from 10% coffee, 2% VS, 20 days HRT and a 1 g VS/1/day loading rate to 80% coffee, 4.5% VS, 12 days HRT and a loading rate of 3 g VS/1/day the CSTR system was optimized at a maximum methane yield of 535 ml/g VS. Again it was possible to attain long term process stability at the above mentioned optimal operational parameters for a further 3 month period. (au)

  17. Thermohydraulic behaviour of the hot channel in a PWR type reactor under loss-of-coolant accident conditions (LOCA)

    International Nuclear Information System (INIS)

    Costa, J.R.

    1978-12-01

    An analysis is done of the core behavior for a 1861 MW(th) pressurized water reactor with two coolant loops, during the blowdown phase of a double-ended cold leg rupture, between the main feedwater pump, and the pressure vessel. The analysis is done through a detailed thermohydraulic study of the hot pin channel with RELAP4/MOD 5 code, including the Evaluatin Model options. The problem is solved separately for two values of discharge coefficient (C sub(D)= 1,0 and 0,4). The results show that the maximum clad temperature is lower than the limit value for licensing purposes. Concerning clad material oxidation, the maximum value obtained is also under the limit of acceptance. (author) [pt

  18. Study on liquid-metal MHD power generation system with two-phase natural circulation. Applicability to fast reactor conditions

    International Nuclear Information System (INIS)

    Saito, Masaki

    2000-03-01

    Feasibility study of the liquid-metal MHD power generation system combined with the high-density two-phase natural circulation has been performed for the applicability to the simple, autonomic energy conversion system of the liquid-metal cooled fast reactor. The present system has many promising aspects not only in the energy conversion process, but also in safety and economical improvements of the liquid-metal cooled fast reactor. For example, the high cycle efficiency can be expected because of the similarity of the present cycle to the Ericsson cycle. Sodium-Water Interaction problem can be excluded by proper combination of the working fluids. As the economical feature, the present system is so simple that the liquid-metal main circular pump, the steam turbine generator, and even the steam generator can be excluded if the thermodynamic working fluid is injected directly into the high temperature liquid metal MHD working fluid. In addition, the present system has the potential to be applied to various heat sources including solar energy because of the high flexibility of the operation temperature. In the present paper, as the first step of the feasibility study, the cycle analyses were performed to examine the effects of the main system parameters on the fundamental characteristics of the system. It is found that the cycle efficiency of the present system is enough competitive with that of the conventional steam turbine system. It is, however, found that the cycle efficiency depends strongly on the gas-liquid slip ratio in the two-phase flow channel. As the conclusions, it is recommended to perform experimental study to obtain the fundamental data, such as the gas-liquid slip ratio in the high-density liquid-metal two-phase natural circulation. (author)

  19. Evaluation of the CO{sub 2} sequestration capacity for coal fly ash using a flow-through column reactor under ambient conditions

    Energy Technology Data Exchange (ETDEWEB)

    Jo, Ho Young, E-mail: hyjo@korea.ac.kr [Department of Earth and Environmental Sciences, Korea University, Anam-dong, Seongbuk-gu, Seoul 136-713 (Korea, Republic of); Ahn, Joon-Hoon; Jo, Hwanju [Department of Earth and Environmental Sciences, Korea University, Anam-dong, Seongbuk-gu, Seoul 136-713 (Korea, Republic of)

    2012-11-30

    Highlights: Black-Right-Pointing-Pointer A conceptual in-situ mineral carbonation method using a coal ash pond is proposed. Black-Right-Pointing-Pointer CO{sub 2} uptake occurred by carbonation reaction of CO{sub 2} with Ca{sup 2+} ions from coal fly ash. Black-Right-Pointing-Pointer The CO{sub 2} sequestration capacity was affected by the solid dosage. Black-Right-Pointing-Pointer Seawater can be used as a solvent for mineral carbonation of coal fly ash. - Abstract: An in-situ CO{sub 2} sequestration method using coal ash ponds located in coastal regions is proposed. The CO{sub 2} sequestration capacity of coal fly ash (CFA) by mineral carbonation was evaluated in a flow-through column reactor under various conditions (solid dosage: 100-330 g/L, CO{sub 2} flow rate: 20-80 mL/min, solvent type: deionized (DI) water, 1 M NH{sub 4}Cl solution, and seawater). The CO{sub 2} sequestration tests were conducted on CFA slurries using flow-through column reactors to simulate more realistic flow-through conditions. The CO{sub 2} sequestration capacity increased when the solid dosage was increased, whereas it was affected insignificantly by the CO{sub 2} flow rate. A 1 M NH{sub 4}Cl solution was the most effective solvent, but it was not significantly different from DI water or seawater. The CO{sub 2} sequestration capacity of CFA under the flow-through conditions was approximately 0.019 g CO{sub 2}/g CFA under the test conditions (solid dosage: 333 g/L, CO{sub 2} flow rate: 40 mL/min, and solvent: seawater).

  20. Evaluation of the CO2 sequestration capacity for coal fly ash using a flow-through column reactor under ambient conditions

    International Nuclear Information System (INIS)

    Jo, Ho Young; Ahn, Joon-Hoon; Jo, Hwanju

    2012-01-01

    Highlights: ► A conceptual in-situ mineral carbonation method using a coal ash pond is proposed. ► CO 2 uptake occurred by carbonation reaction of CO 2 with Ca 2+ ions from coal fly ash. ► The CO 2 sequestration capacity was affected by the solid dosage. ► Seawater can be used as a solvent for mineral carbonation of coal fly ash. - Abstract: An in-situ CO 2 sequestration method using coal ash ponds located in coastal regions is proposed. The CO 2 sequestration capacity of coal fly ash (CFA) by mineral carbonation was evaluated in a flow-through column reactor under various conditions (solid dosage: 100–330 g/L, CO 2 flow rate: 20–80 mL/min, solvent type: deionized (DI) water, 1 M NH 4 Cl solution, and seawater). The CO 2 sequestration tests were conducted on CFA slurries using flow-through column reactors to simulate more realistic flow-through conditions. The CO 2 sequestration capacity increased when the solid dosage was increased, whereas it was affected insignificantly by the CO 2 flow rate. A 1 M NH 4 Cl solution was the most effective solvent, but it was not significantly different from DI water or seawater. The CO 2 sequestration capacity of CFA under the flow-through conditions was approximately 0.019 g CO 2 /g CFA under the test conditions (solid dosage: 333 g/L, CO 2 flow rate: 40 mL/min, and solvent: seawater).

  1. Investigations to the potential of the high temperature reactor for steam power processes with highest steam conditions and comparison with according conventional power plants

    International Nuclear Information System (INIS)

    Mondry, M.

    1988-04-01

    Already in the fifties conventional power plants with high parameters of the live steam were built to improve the total efficiency. The power plant with the highest steam conditions in the Federal Republic of Germany has 300 bar pressure and 600deg C temperature. Because of high material costs and other problems power plants with such high conditions were not continued to be built. Standard conditions of today's power plants are in the order of 180-250 bar pressure and 535deg C temperature. As the high temperature reactor is partly built up in another way than a conventional power plant, the results regarding the high steam parameters are not transferable. Possibilities for the technical realization of determined HTR-specific components are introduced and discussed. Then different HTR-power plants with steam conditions up to 350 bar pressure and 650deg C temperature are projected. Economical considerations show that an HTR with higher steam parameters brings financial profits. Further efficiency increase, which is possible by the high steam conditions, is shortly presented. The work ends with a technical and economical comparison of corresponding conventional power plants. (orig./UA) [de

  2. Interdisciplinary study of the influence on effectiveness of catalytic hydrogen recombiners of operating conditions in the reactor containment

    International Nuclear Information System (INIS)

    Kelm, S.; Reinecke, E.A.; Schoppe, L.; Dornseiffer, J.; Leistner, F.; Juehe, S.

    2008-01-01

    At the Emsland nuclear power station, a total of 58 autocatalytic hydrogen recombiners were backfitted in 1999 as an additional measure of risk reduction in connection with major hydrogen releases after events going beyond the design basis. Annual in-service inspections after 2002 revealed that some of the catalyst sheets developed startup delays and marked evolutions of smoke and smell. Recombiners not meeting the inspection criterion were completely regenerated as a measure of precaution. A preventive study was conducted jointly with institutes of the Juelich Research Center and the Aachen Technical University to analyze the composition of the deposits, which was then compared with the chemical characteristics of potential sources in the reactor containment. At the same time, the influence on effectiveness of the catalyst sheets was examined. On the basis of a random evaluation of the in-service inspection logs of the past few years, representative samples were taken whose startup behavior and operating characteristics were studied in a test rig alongside chemical analyses so as to allow a correlation to be established between the analytical findings and the catalytic activity of the samples. The findings made allowed internal sources of the catalyst deposits to be excluded. The impurities are introduced with the outside air. As a consequence, the air ducts in the vicinity of the respective recombiners were inspected and optimization steps were taken in connection with in-service inspections and regeneration procedures. (orig.)

  3. Study on the behavior of unirradiated light water reactor fuel with iodine-127 under the reactivity initiated accident (RIA) conditions

    International Nuclear Information System (INIS)

    Sasajima, Hideo; Yanagisawa, Kazuaki; Kanazawa, Hiroyuki

    1988-07-01

    In a light water reactor fuel, a stress arised from pellet-cladding interaction (PCI) will have possibility to cause stress corrosion claddling (PCI failure) under an environment of corrosive fission product ; line iodine. A pulse irradiation experiment by NSRR was carried out to examine whether or not an unirradiated NSRR standard fuel rod in which 1.3 mg (33 x 10 -6 g/cm 2 ) of iodine was artificially filled could cause the PCI failure. Obtained results are: (1) The fuel rods with iodine did not fail both at deposited energy levels of 268 and 280 cal/g · UO 2 . On the other hand, the fuel rods without iodine failed at the same energy levels due to thinning of the cladding wall thickness. Within this experimental scope, PCI-failure did not occur on iodine filled fuel rods. (2) At a periphery of the fuel pellet of iodine filled rod, an uniform torus ring was formed. The torus ring consisted of an equi-axed large grains at 268 cal/g · UO 2 and a columnar ones at 280 cal/g · UO 2 . The torus ring was not formed in the fuel without iodine. (author)

  4. Layout of the manipulator-arm (boom) for the TFTR fusion reactor (Princeton, USA) under UHV-conditions

    International Nuclear Information System (INIS)

    Klaubert, J.

    1987-01-01

    This presentation shows the main criteria for the layout of the manipulator - arm and the antechamber - vessel of the TFTR - FUSION - REACTOR at Princeton University, PLASMA PHYSICS LABORATORY (USA). The main problem during layout of a manipulator system like the TFTR - Boom has been the limitation of the vertical deflections due to deadweight of the construction. The design problem is rather a deformation problem and a problem of stability than a stress problem. The way of optimizing the ratio between stiffness and deadweight is the most important part during the complete design - process. Additional earthquake requirements need further investigations for a satisfying layout (horizontal forces, weak-axis of moment of inertia). The details of the construction (welding, connections etc.) have to be designed in respect to UHV - requirements --> no holes and no fillet welds (outgasing - rate.) are allowed. All weldings have to be designed as bevel-welds. This manipulator system is designed for working in a plane system (two degrees of freedom). A manipulator system with the same operating capabilities in a three degree of freedom system needs larger cross sections for the different beam-elements than those of the discussed TFTR - BOOM

  5. Procedure for qualification of electric equipment installed in containments for pressurized water reactors subject to accident conditions

    International Nuclear Information System (INIS)

    1991-11-01

    This generic norm is usable for electrical equipment installed in containment building of PWR subject to accidental conditions. She defines the qualification methods and the general rules usable for the test specifications of qualification for these materials

  6. Development of core hot spot evaluation method for decay heat removal by natural circulation under transient conditions in sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Ohshima, Hiroyuki; Doda, Norihiro; Kamide, Hideki; Watanabe, Osamu; Ohkubo, Yoshiyuki

    2010-01-01

    Toward the commercialization of fast reactors, a design study of Japan Sodium-cooled Fast Reactor (JSFR) is being performed. In this design study, the adoption of decay heat removal system operated by fully natural circulation is being examined from viewpoints of economic competitiveness and passive safety. This paper describes a new evaluation method of core hot spot under transient conditions from forced to natural circulation operations that is necessary for confirming feasibility of the fully natural circulation decay heat removal system. The new method consists of three analysis steps in order to include effects of thermal hydraulic phenomena particular to the natural circulation decay heat removal, e.g., flow redistribution in fuel assemblies caused by buoyancy force, and therefore it enables more rational hot spot evaluation rather than conventional ones. This method was applied to a hot spot evaluation of loss-of-external-power event and the result was compared with those by conventional 1D and detailed 3D simulations. It was confirmed that the proposed method can estimate the hot spot with reasonable degree of conservativeness. (author)

  7. Behavior to the fracture of an AISI 304 stainless steel sensitized in BWR reactor conditions (288 degrees Centigrade and 80 Kg/cm2)

    International Nuclear Information System (INIS)

    Hernandez C, R.; Diaz S, A.; Garcia R, R.; Aguilar T, A.; Gachuz M, M.; Arganis J, C.; Merino C, J.

    1999-01-01

    It is a knew fact that ductility of a lot of structural alloys can be deteriorated by the environment effect which are exposed, and that their consequent embrittlement can put in doubt the safety of their functioning; such is the case of austenitic stainless steels used in internal components of the BWR type reactors which not only is subjected to the effect combined of the aggressive environment which surround it (pressure, temperature, corrosion potential, conductivity medium, local state of efforts, etc.), but also to the action of present neutron radiation, manifesting microstructural changes which are reflected in the augmentation of its susceptibility to the intergranular cracking, phenomena generally known as IASCC ''Irradiation Assisted Stress Corrosion Cracking''. Once appeared the cracking in the material, the useful life of a component is limited by the rapidity to growth of these cracking, making necessary evaluations which can to predict its behavior, therefore the present work shows the preliminary results for determining the behavior to the fracture of an AISI 304 stainless steel sensitized, in a dynamic recirculation circuit which allows to simulate the operation conditions of a BWR reactor (288 Centigrade and 80 kg/cm 2 ). (Author)

  8. Understanding of the operation behaviour of a Passive Autocatalytic Recombiner (PAR) for hydrogen mitigation in realistic containment conditions during a severe Light Water nuclear Reactor (LWR) accident

    International Nuclear Information System (INIS)

    Payot, Frédéric; Reinecke, Ernst-Arndt; Morfin, Franck; Sabroux, Jean-Christophe; Meynet, Nicolas; Bentaib, Ahmed; March, Philippe; Zeyen, Roland

    2012-01-01

    Highlights: ► Recombineur operation in the presence of fission products (severe accident conditions). ► Operation of catalysts in the integral and small-scale tests. ► The catalyst performance was observed by measuring the coupon temperature increase. ► The experimental observations were corroborated by numerical calculations (SPARK code). - Abstract: In the context of hydrogen risk mitigation in nuclear power plants (NPPs), experimental studies of a possible poisoning of Passive Autocatalytic Recombiners (PARs) by fission products (FPs) and aerosols released during a core meltdown accident were mainly conducted in the past with non-radioactive fission product surrogates (e.g., in the H2PAR facility at Cadarache, France). The decision was taken in 1997 to complete these studies by a test in the Phébus facility, a research nuclear reactor also at Cadarache: it was a rare opportunity to expose catalyst samples to an atmosphere as representative as possible of a core meltdown accident, containing gaseous fission products and aerosols released during the degradation of an actual irradiated nuclear fuel bundle. Before testing in Phébus during the FPT3 experiment, reference and qualification tests were performed in the H2PAR facility using the same samples — the so-called “coupons” — and coupons holder to check that the apparatus was functional and correctly designed for avoiding to tamper with the thermal-hydraulics and chemical conditions in the Phébus containment. The correct operation of catalysts was checked by measuring the surface temperature increase of the coupons due to the exothermic reaction between hydrogen and oxygen. After the Phébus FPT3 test (November 2004), REKO-1 tests were initiated at Jülich, Germany, to confirm the discrepancy in coupons temperature observed in Phébus FPT3 and H2PAR PHEB-03 tests, and to study the operation behaviour of PARs. Besides, before REKO-1 tests, a first interpretation of H2PAR and Phébus experiments

  9. Chaboche-based cyclic material hardening models for 316 SS–316 SS weld under in-air and pressurized water reactor water conditions

    Energy Technology Data Exchange (ETDEWEB)

    Mohanty, Subhasish, E-mail: smohanty@anl.gov; Soppet, William K.; Majumdar, Saurindranath; Natesan, Krishnamurti

    2016-08-15

    Highlights: • 316 SS–316 SS weld cyclically harden/soften while undergoing fatigue loading. • Cyclic hardening/softening creates cycle dependent stress-strain curves. • This necessitate to estimate the cycle dependence of material properties. • Cyclic evolution of Chaboche parameters are estimated under different conditions. - Abstract: This paper discusses a material hardening models for welds made from 316 stainless steel (SS) to 316 SS. The model parameters were estimated from the strain-versus-stress curves obtained from tensile and fatigue tests conducted under different conditions (air at room temperature, air at 300 °C, and primary loop water conditions for a pressurized water reactor). These data were used to check the fatigue cycle dependency of the material hardening parameters (yield stress, parameters related to Chaboche-based linear and nonlinear kinematic hardening models, etc.). The details of the experimental results, material hardening models, and associated calculated results are published in an Argonne report (ANL/LWRS-15/2). This paper summarizes the reported material parameters for 316 SS–316 SS welds and their dependency on fatigue cycles and other test conditions.

  10. The optimum operating conditions of the phased double-rotor facility at the ET-RR-1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Naguib, K.; Habib, N.; Wahba, M.; Kilany, M.; Adib, M. [National Research Centre, Cairo (Egypt). Reactor and Neutron Physics Dept.

    1997-02-07

    A pulsed neutron polyenergetic thermal beam at ET-RR-1 is produced by a phased double-rotor facility. One of the rotors has two diametrically opposite curved slots, while the second is designed to operate as a rotating collimator. The dimensions of the phased rotating collimator are selected to match the curved slot rotor. The calculated collimator transmissions at different operating conditions are found to be in good agreement with the experimental ones. The optimum operating conditions of the double-rotor facility are deduced. The calculations were carried out using a computer program RCOL. The RCOL was designed in FORTRAN-77 to operate on PCs. (author).

  11. The optimum operating conditions of the phased double-rotor facility at the ET-RR-1 reactor

    International Nuclear Information System (INIS)

    Naguib, K.; Habib, N.; Wahba, M.; Kilany, M.; Adib, M.

    1997-01-01

    A pulsed neutron polyenergetic thermal beam at ET-RR-1 is produced by a phased double-rotor facility. One of the rotors has two diametrically opposite curved slots, while the second is designed to operate as a rotating collimator. The dimensions of the phased rotating collimator are selected to match the curved slot rotor. The calculated collimator transmissions at different operating conditions are found to be in good agreement with the experimental ones. The optimum operating conditions of the double-rotor facility are deduced. The calculations were carried out using a computer program RCOL. The RCOL was designed in FORTRAN-77 to operate on PCs. (author)

  12. Development of technology for the in-service inspection of reactor pressure vessel metal condition using a magnetic method

    International Nuclear Information System (INIS)

    Bakirov, M. B.; Zabruskov, N.Y; Massoud, J.P.

    2002-01-01

    The opportunity to perform the inspection of condition of base metal and metal of welded joints of cladded vessels of PWR by non destructive methods is shown. The technique for on-site specimen-free testing is offered on the basis of sharing a kinetic hardness method and magnetic method. The results of studies of magnetic and mechanical properties of vessel steels in various condition after irradiation and thermal processing are submitted. It is shown, that the magnetic properties (first of all coercive force) are sensitive to change of structure of a material. (authors)

  13. Eliciting Information on Sensitive Matters Without Inviting ...

    Indian Academy of Sciences (India)

    Eliciting Information on Sensitive Matters. Without Inviting Respondents' ... methods based on Randomized Response tech- niques. ... while collecting data on some sensitive issues are well ..... Suppose there is an association of professionals.

  14. Non-Linear Response to Periodic Forcing of Methane-Air Global and Detailed Kinetics in Continuous Stirred Tank Reactors Close to Extinction Conditions

    Directory of Open Access Journals (Sweden)

    Francesco Saverio Marra

    2015-09-01

    Full Text Available This paper focus on the behavior of a continuous stirred tank reactor (CSTR subject to perturbations of finite amplitude and frequency. Two main objectives are pursued: to determine the extinction line in the equivalence ratio (φ - residence time (τ plane, fixed the thermodynamic state conditions; and to characterize the response of the chemical system to periodic forcing of the residence time. Transient simulations of combustion of methane with air, using both global single-step and detailed chemical kinetic mechanisms, have been conducted and the corresponding asymptotic solutions analyzed. Results indicate very different dynamical behaviors, posing the issue of a proper choice of the kinetic scheme for the numerical study of combustion oscillations.

  15. Safety aspects of the cleaning and conditioning of radioactive sludge from spent fuel storage pool on 'RA' Research reactor in the Vinca Institute

    International Nuclear Information System (INIS)

    Pavlovic, R; Pavlovic, S.; Plecas, I.

    1999-01-01

    Spent fuel elements from nuclear reactors in the Vinca Institute have been temporary stored in water filled storage pool. Due to the fact that the water in the spent fuel elements storage pool have not been purified for a long time, all metallic components submerged in the water have been hardly corroded and significant amount of the sludge has been settled on the bottom of the pool. As a first step in improving spent fuel elements storage conditions and slowing down corrosion in the storage spent fuel elements pool we have decided to remove the sludge from the bottom of the pool. Although not high, but slightly radioactive, this sludge had to be treated as radioactive waste material. Some safety aspects and radiation protection measures in the process of the spent fuel storage pool cleaning are presented in this paper

  16. The Effect of Influent Characteristics and Operational Conditions over the Performance and Microbial Community Structure of Partial Nitritation Reactors

    Directory of Open Access Journals (Sweden)

    Alejandro Rodriguez-Sanchez

    2014-06-01

    Full Text Available Nitrogen is a main contaminant of wastewater worldwide. Novel processes for nitrogen removal have been developed over the last several decades. One of these is the partial nitritation process. This process includes the oxidation of ammonium to nitrite without the generation of nitrate. The partial nitritation process has several advantages over traditional nitrification-denitrification processes for nitrogen removal from wastewaters. In addition, partial nitritation is required for anammox elimination of nitrogen from wastewater. Partial nitritation is affected by operational conditions and substances present in the influent, such as quinolone antibiotics. In this review, the impact that several operational conditions, such as temperature, pH, dissolved oxygen concentration, hydraulic retention time and solids retention time, have over the partial nitritation process is covered. The effect of quinolone antibiotics and other emerging contaminants are discussed. Finally, future perspectives for the partial nitritation process are commented upon.

  17. Metallographic examination of (uth) O2 and UO2 fuel tested in power ramp conditions in triga reactor

    International Nuclear Information System (INIS)

    Ioncescu, M.; Uta, O.

    2015-01-01

    The purpose of this paper is to determine the behavior of two fuel experimental elements (EC1 and EC2), by destructive post-irradiation examination. The fuel elements were mounted inside a pattern port, one in extension of the other and irradiated in power ramp conditions in order to check their behavior. Fuel element 1 (EC1) contains (UTh)O''2 pellet, and other one (EC2) UO''2 pellet. The results of destructive post-irradiation examination are evidenced by metallographic and ceramographic analyses. The data obtained from the post-irradiation examinations are used, first to confirm the security, reliability and nuclear fuel performance, and second, for the development of CANDU fuel. The results obtained by destructive examinations regarding the integrity, sheath hydrating and oxidation as well as the structural modifications are typical for fuel elements tested in power ramp conditions. (authors)

  18. Effectiveness of solar disinfection using batch reactors with non-imaging aluminium reflectors under real conditions: Natural well-water and solar light.

    Science.gov (United States)

    Navntoft, C; Ubomba-Jaswa, E; McGuigan, K G; Fernández-Ibáñez, P

    2008-12-11

    Inactivation kinetics are reported for suspensions of Escherichia coli in well-water using compound parabolic collector (CPC) mirrors to enhance the efficiency of solar disinfection (SODIS) for batch reactors under real, solar radiation (cloudy and cloudless) conditions. On clear days, the system with CPC reflectors achieved complete inactivation (more than 5-log unit reduction in bacterial population to below the detection limit of 4CFU/mL) one hour sooner than the system fitted with no CPC. On cloudy days, only systems fitted with CPCs achieved complete inactivation. Degradation of the mirrors under field conditions was also evaluated. The reflectivity of CPC systems that had been in use outdoors for at least 3 years deteriorated in a non-homogeneous fashion. Reflectivity values for these older systems were found to vary between 27% and 72% compared to uniform values of 87% for new CPC systems. The use of CPC has been proven to be a good technological enhancement to inactivate bacteria under real conditions in clear and cloudy days. A comparison between enhancing optics and thermal effect is also discussed.

  19. Energies and media nr 26. Conditions for the nuclear sector. Nuclear reactors for developing countries? Fast reactors: why not to make a fuss about sodium. Depleted uranium: a godsend, not a waste

    International Nuclear Information System (INIS)

    2008-06-01

    After a brief article of the actuality in the nuclear sector (commissioning of new reactors in the USA, investments allowed in the UK for the construction of nuclear reactors, boost on the Chinese program, perspectives in South Africa), this issue proposes a discussion of the problems raised by providing nuclear reactors to developing countries (problems of safety and security, risk of proliferation, important issues on expertise development and personnel training, international controls, regulations and cooperation). Other topics are briefly addressed: the risk associated with the use of sodium in fast neutron reactors, the use and interest of depleted uranium as a resource

  20. Reactor core and initially loaded reactor core of nuclear reactor

    International Nuclear Information System (INIS)

    Koyama, Jun-ichi; Aoyama, Motoo.

    1989-01-01

    In BWR type reactors, improvement for the reactor shutdown margin is an important characteristic condition togehter with power distribution flattening . However, in the reactor core at high burnup degree, the reactor shutdown margin is different depending on the radial position of the reactor core. That is , the reactor shutdown margin is smaller in the outer peripheral region than in the central region of the reactor core. In view of the above, the reactor core is divided radially into a central region and as outer region. The amount of fissionable material of first fuel assemblies newly loaded in the outer region is made less than the amount of the fissionable material of second fuel assemblies newly loaded in the central region, to thereby improve the reactor shutdown margin in the outer region. Further, the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower portion of the first fuel assemblies is made smaller than the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower region of the second fuel assemblies, to thereby obtain a sufficient thermal margin in the central region. (K.M.)

  1. Regulations for RA reactor operation

    International Nuclear Information System (INIS)

    1980-09-01

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

  2. 11 CFR 100.137 - Invitations, food, and beverages.

    Science.gov (United States)

    2010-01-01

    ... 11 Federal Elections 1 2010-01-01 2010-01-01 false Invitations, food, and beverages. 100.137...) Exceptions to Expenditures § 100.137 Invitations, food, and beverages. The cost of invitations, food, and... invitations, food and beverages provided by the individual on behalf of the candidate does not exceed $1,000...

  3. 11 CFR 100.77 - Invitations, food, and beverages.

    Science.gov (United States)

    2010-01-01

    ... 11 Federal Elections 1 2010-01-01 2010-01-01 false Invitations, food, and beverages. 100.77...) Exceptions to Contributions § 100.77 Invitations, food, and beverages. The cost of invitations, food and... invitations, food and beverages provided by the individual on behalf of the candidate does not exceed $1,000...

  4. Nuclear reactor safety systems

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  5. Guidance of reactor operators and TSC personnel with the severe accident management guidance under shutdown and low power conditions

    International Nuclear Information System (INIS)

    Van Haesendonck, M.F.; Prior, R.P.

    2000-01-01

    The Westinghouse Owners Group Severe Accident Management Guidance (WOG SAMG) was developed between 1991 and 1994. The primary goals for severe accident management that form the basis of the WOG SAMG are to terminate any radioactive releases to the environment; to prevent failure of any containment fission product boundary and to return the plant to a controlled stable condition. The WOG SAMG is primarily a TSC tool for mitigation of low probability core damage events. The philosophy is that control room operators should remain focused on the prevention of core damage, whereas the TSC personnel should concentrate on the mitigation of the severe accident. The symptom based package is built up as a structured process for choosing appropriate actions based on actual plant conditions. No detailed knowledge of severe accident phenomena is required. The scope of the WOG SAMG is limited to severe accidents resulting from initiating events occurring during full power operation. However, a number of studies such as the EdF EPS 1300 Probabilistic Safety Assessment (PSA), the shutdown Probabilistic Risk Assessment (PRA) for Surry, the BERA shutdown PRA for Beznau, the EPRI/ Westinghouse ORAM methodology etc. have shown that the frequency of core damage (a severe accident) during shutdown and low power operation can be of the same order of magnitude as for full power operation. The at-power SAMG is viewed as the resolution of the severe accident issue. Similarly, it is expected that as shutdown PRAs mature, the final resolution of the severe accident issue will lie in SAMG for low power and shutdown operation. Therefore in resolution of this issue, Westinghouse has developed the Shutdown Severe Accident Management Guidance (SSAMG) which gives guidance for both control room and TSC personnel to mitigate a severe accident under shutdown or low power conditions. In the last few years, many LWR plants have been implementing SAMG. In the US, all plants have developed SAMG, and many

  6. Experimental Investigation on the Effects of Coolant Concentration on Sub-Cooled Boiling and Crud Deposition on Reactor Cladding at Prototypical PWR Operating Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Schultis, J., Kenneth; Fenton, Donald, L.

    2006-10-20

    Increasing demand for energy necessitates nuclear power units to increase power limits. This implies significant changes in the design of the core of the nuclear power units, therefore providing better performance and safety in operations. A major hindrance to the increase of nuclear reactor performance especially in Pressurized Deionized water Reactors (PWR) is Axial Offset Anomaly (AOA)--the unexpected change in the core axial power distribution during operation from the predicted distribution. This problem is thought to be occur because of precipitation and deposition of lithiated compounds like boric acid (H{sub 2}BO{sub 3}) and lithium metaborate (LiBO{sub 2}) on the fuel rod cladding. Deposited boron absorbs neutrons thereby affecting the total power distribution inside the reactor. AOA is thought to occur when there is sufficient build-up of crud deposits on the cladding during subcooled nucleate boiling. Predicting AOA is difficult as there is very little information regarding the heat and mass transfer during subcooled nucleate boiling. An experimental investigation was conducted to study the heat transfer characteristics during subcooled nucleate boiling at prototypical PWR conditions. Pool boiling tests were conducted with varying concentrations of lithium metaborate (LiBO{sub 2}) and boric acid (H{sub 2}BO{sub 3}) solutions in deionized water. The experimental data collected includes the effect of coolant concentration, subcooling, system pressure and heat flux on pool the boiling heat transfer coefficient. The analysis of particulate deposits formed on the fuel cladding surface during subcooled nucleate boiling was also performed. The results indicate that the pool boiling heat transfer coefficient degrades in the presence of boric acid and lithium metaborate compared to pure deionized water due to lesser nucleation. The pool boiling heat transfer coefficients decreased by about 24% for 5000 ppm concentrated boric acid solution and by 27% for 5000 ppm

  7. Modeling and analysis of thermal-hydraulic response of uranium-aluminum reactor fuel plates under transient heatup conditions

    Energy Technology Data Exchange (ETDEWEB)

    Navarro-Valenti, S.; Kim, S.H.; Georgevich, V. [Oak Ridge National Lab., TN (United States)] [and others

    1995-09-01

    The purpose of this paper is to describe the analysis performed to predict the thermal behavior of fuel miniplates under rapid transient heatup conditions. The possibility of explosive boiling was considered, and it was concluded that the heating rates are not large enough for explosive boiling to occur. However, transient boiling effects were pronounced. Because of the complexity of transient pool boiling and the unavailability of experimental data for the situations studied, an approximation was made that predicted the data very well within the uncertainties present. If pool boiling from the miniplates had been assumed to be steady during the heating pulse, the experimental data would have been greatly overestimated. This fact demonstrates the importance of considering the transient nature of heat transfer in the analysis of reactivity excursion accidents. An additional contribution of the present work is that it provided data on highly subcooled steady nulceate boiling from the cooling portion of the thermocouple traces.

  8. One piece reactor removal

    International Nuclear Information System (INIS)

    Chia, Wei-Min; Wang, Song-Feng

    1993-01-01

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

  9. Fission reactors and materials

    International Nuclear Information System (INIS)

    Frost, B.R.T.

    1981-12-01

    The American-designed boiling water reactor and pressurized water reactor dominate the designs currently in use and under construction worldwide. As in all energy systems, materials problems have appeared during service; these include stress-corrosion of stainless steel pipes and heat exchangers and questions regarding crack behavior in pressure vessels. To obtain the maximum potential energy from our limited uranium supplies is is essential to develop the fast breeder reactor. The materials in these reactors are subjected to higher temperatures and neutron fluxes but lower pressures than in the water reactors. The performance required of the fuel elements is more arduous in the breeder than in water reactors. Extensive materials programs are in progress in test reactors and in large test rigs to ensure that materials will be available to meet these conditions

  10. Nuclear reactors

    International Nuclear Information System (INIS)

    Barre, Bertrand

    2015-10-01

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

  11. Experimental studies of the influence of fuel properties and operational conditions on stoking when combusting fuels in a fixed-bed reactor

    Energy Technology Data Exchange (ETDEWEB)

    Arias, Fabiana; Kolb, T.; Seifert, H.; Gehrmann, Hans-Joachim [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany). Inst. for Technical Chemistry (ITC)

    2013-09-01

    Besides from knowledge about pollutant emission, knowledge of the combustion behavior of fuels plays a major role in the operation and optimization of combustion plants for waste and biomass. If the fuel is exchanged partly or completely in existing or newly designed grate-type combustion plants, adaptation of technical parameters is usually based on purely empirical studies. In the KLEAA fixed-bed reactor of KIT, Institute for Technical Chemistry (ITC), quantitative data on the combustion behavior can be determined from experimental investigations on the laboratory scale. Based on the characteristics obtained, the combustion behavior on a continuous grate can be estimated, This estimation is based on the assumption that no back mixing of the fuel occurs on the grate. Depending on the type of grate, however, stoking and back mixing play an important role. To improve the quality of the characteristics determined in KLEAA and enhance their transferability to the continuous process, it is necessary to determine the influence of fuel properties and operation conditions on stoking. Work is aimed at further developing the characteristics model taking into account a stoking factor describing the combustion behavior of a non-stoked fixed bed compared to a stoked fixed bed. The main task is to make a systematic study of the major parameters influencing stoking (e.g. stroke length, stroke frequency, geometry of the stoking unit, and fuel properties) in a fixed-bed reactor. The results shall be presented in the form of a semi-empirical equation. It is recommended to first study a model fuel, whose fuel properties are defined exactly and can be adjusted variably. Then, a stoking factor shall be derived from the studies. Possibly, a dimension analysis may be helpful. Finally, the results obtained are to be verified for residue-derived fuel. (orig.)

  12. Reactor building

    International Nuclear Information System (INIS)

    Ebata, Sakae.

    1990-01-01

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

  13. Electrochemical behaviour of stainless steel under radiation and exposed to representative chemistry in pressurised water reactor conditions

    International Nuclear Information System (INIS)

    Wang, Mi

    2013-01-01

    The dissertation focuses on the behaviour of stainless steel under irradiation and exposed to primary PWR conditions. The electrochemical potential of austenitic 316L stainless steel and the environmental parameters (hydrogen pressure, temperature, etc.,) have been measured continuously at high temperature (HT) and high pressure (HP) under irradiation, using a unique experimental HTHP working cell. Two sources of irradiation, proton and electron beams, have been employed in the study. A high similarity of electrochemical behaviour under both types of irradiations has been observed: (i) an oxidative potential response under irradiation (few tens of milli-volts); (ii) an increase in the hydrogen pressure reduces the oxidative potential response; (iii) a synergetic effect of thermal ageing and fluence leading to a decrease of the oxidative response under irradiation. The observations of the oxide film showed that without irradiation, metallic nickel in the inner and outer oxide films has been observed under a high hydrogen pressure. Under irradiation, um scale cavities (pits) have been observed in the strongly electron irradiated oxide film formed on 316L stainless steel. These defects are induced by the effect of irradiation of the passive film and water radiolysis. It is also shown that water radiolysis influences the PWR water chemistry by making it become a stronger oxidant at the oxide/solution interface. As a result, the release of metallic cations is increased and a-Fe 2 O 3 hematite has been observed on the irradiated outer oxide film where cavities were formed. (author) [fr

  14. Assessment on the Reactor Containment Cooling Capability of Kori Unit 1 Under LOCA Conditions with Loss of Offsite Power

    International Nuclear Information System (INIS)

    Lee, Jin Yong; Park, Jong Woon; Kim, Hyeong Taek

    2006-01-01

    The fan cooler system is designed to remove heat from containment under postulated accident conditions. During a postulated LOCA concurrent with a Loss of Offsite Power (LOOP), the Component Cooling Water (CCW) pumps that supply cooling water to the fan cooler and the fan that supplies containment air to the fan cooler will temporarily lose power. Then, the high temperature steam in the containment atmosphere will pass over the fan cooler tubing without forced cooling water flow. In that case, boiling may occur in the fan cooler tubes causing steam bubbles to form and pass into the attached CCW piping creating steam voids. Prior to the CCW pumps restart, the presence of steam and subcooled water can induce the potential for water hammer. As the CCW pumps restart, the accumulated steam condenses and the pumped water can produce a water hammer when the void closes. The hydrodynamic loads caused by such a water hammer event could challenge the integrity and the function of the fan cooler and associated CCW system. With respect to this phenomena, the United States Nuclear Regulatory Commission (USNRC) issued the Generic Letter (GL) 96-06, which requests an assessment of the possibility of boiling and water hammer in the cooling water system. The objectives of this study are to develop a analysis method for predicting the thermal hydraulic status of containment fan cooler and then to assess the containment fan cooler of Kori Unit 1 using the developed model under a LOCA with LOOP

  15. Food Security Under Shifting Economic, Demographic, and Climatic Conditions (Invited)

    Science.gov (United States)

    Naylor, R. L.

    2013-12-01

    Global demand for food, feed, and fuel will continue to rise in a more populous and affluent world. Meeting this demand in the future will become increasingly challenging with global climate change; when production shocks stemming from climate variability are added to the new mean climate state, food markets could become more volatile. This talk will focus on the interacting market effects of demand and supply for major food commodities, with an eye on climate-related supply trends and shocks. Lessons from historical patterns of climate variability (e.g., ENSO and its global teleconnections) will be used to infer potential food security outcomes in the event of abrupt changes in the mean climate state. Domestic food and trade policy responses to crop output and price volatility in key producing and consuming nations, such as export bans and import tariffs, will be discussed as a potentially major destabilizing force, underscoring the important influence of uncertainty in achieving--or failing to achieve--food security.

  16. A study on framework for assessing limiting condition for operation(LCOs) quantitatively in a CANDU reactor

    International Nuclear Information System (INIS)

    Kang, Kyung Min

    2003-02-01

    Limiting conditions for operations define the allowed outage times and the actions to be taken if the repair cannot be completed within the AOT. Typically the actions required are plant shutdown. In situations where the risk associated with the action, i.e., the risk of plant shutdown given a failure in the safety system, may be substantial, a strategy is needed to control the risk implications. In this study the strategy is evaluated quantitatively using a tool of system dynamics. The strategic actions associated with LCO needs to be assessed dynamically. System dynamics technique is easy enough to be applied for quantitative assessment of LCO, where the system dynamics is an important factor in evaluating operational modes of nuclear power plants. The review on the improvements of overly conservative technical specification has been performed in this study. The VENSIM tool has been applied to evaluate quantitatively the LCO of the auxiliary feed water systems in the Wol-Sung nuclear power plant. A value of core damage frequency in PSA is used as risk measure. The analysis of both full power operation and shutdown operation has been compared for the value of the CDF. This study shows a plot of LCO full power operating and shutdown risks in term of core damage probability for failure of AFWS (auxiliary feedwater system). Obtaining a lower risk level in a stable mode, compared to the LCO operation alternative, is the principal motivation of going to full power operational mode or shutdown operational mode. A time dependent framework developed in this study has been applied to assess the LCO of the example problem and it is shown that it is very flexible in that it can be applied to assess LCO quantitatively under any operational context of the technical specifications. This study contributes risk informed regulation to enhance and optimize the LCO of previous technical specifications for CANDU through developing framework in application proposed in this methodology

  17. Pressurised water reactor operation

    International Nuclear Information System (INIS)

    Birnie, S.; Lamonby, J.K.

    1987-01-01

    The operation of a pressurized water reactor (PWR) is described with respect to the procedure for a unit start-up. The systems details and numerical data are for a four loop PWR station of the design proposed for Sizewell-'B', United Kingdom. A description is given of: the initial conditions, filling the reactor coolant system (RCS), heat-up and pressurisation of the RCS, secondary system preparations, reactor start-up, and reactivity control at power. (UK)

  18. RA Reactor applications, Annex A

    International Nuclear Information System (INIS)

    Cupac, S.; Vukadin, Z.

    2000-01-01

    Full text: In 2000 Ra reactor was not operated. New instrumentation is not complete, without it, it is not possible to think about reactor start-up. Since 1985, when reactor operation was forbidden, there are 480 fuel elements left in 48 fuel channels in the reactor core. Heavy water was removed from the reactor core because of the repair of the heavy water pumps in 1986. The old instrumentation was removed. Eleven years after being left to its own destiny, it would be difficult to imagine that anybody would think of reactor restart without examining the state of reactor vessel and other vital reactor components. Maintaining the reactor under existing conditions without final decision about restart or permanent shutdown is destructive for this nuclear facility. The existing state that pertains for more than 10 years would have only one result, destruction of the RA reactor [sr

  19. RA Reactor applications, Annex A

    International Nuclear Information System (INIS)

    Cupac, S.; Vukadin, Z.

    1998-01-01

    Full text: In 1998 Ra reactor was not operated. New instrumentation is not complete, without it, it is not possible to think about reactor start-up. Since 1985, when reactor operation was forbidden, there are 480 fuel elements left in 48 fuel channels in the reactor core. Heavy water was removed from the reactor core because of the repair of the heavy water pumps in 1986. The old instrumentation was removed. Eleven years after being left to its own destiny, it would be difficult to imagine that anybody would think of reactor restart without examining the state of reactor vessel and other vital reactor components. Maintaining the reactor under existing conditions without final decision about restart or permanent shutdown is destructive for this nuclear facility. The existing state that pertains for more than 10 years would have only one result, destruction of the RA reactor [sr

  20. RA Reactor applications, Annex A

    International Nuclear Information System (INIS)

    Cupac, S.; Vukadin, Z.

    1996-01-01

    Full text: In 2000 Ra reactor was not operated. New instrumentation is not complete, without it, it is not possible to think about reactor start-up. Since 1985, when reactor operation was forbidden, there are 480 fuel elements left in 48 fuel channels in the reactor core. Heavy water was removed from the reactor core because of the repair of the heavy water pumps in 1986. The old instrumentation was removed. Eleven years after being left to its own destiny, it would be difficult to imagine that anybody would think of reactor restart without examining the state of reactor vessel and other vital reactor components. Maintaining the reactor under existing conditions without final decision about restart or permanent shutdown is destructive for this nuclear facility. The existing state that pertains for more than 10 years would have only one result, destruction of the RA reactor [sr

  1. Opinion about difficulties of RA reactor operation under conditions of high activity of the heavy water system - Annex 2; Prilog 2 - Misljenje o teskocama eksploatacije reaktora RA u uslovima visoke aktivnosti teskovodnog sistema

    Energy Technology Data Exchange (ETDEWEB)

    Nikolic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1963-12-15

    It was concluded that reactor the reactor operation is very dangerous for the reactor installation as well as safety of the staff under conditions of heavy water increased activity. Two fundamental arguments in favour of this conclusion are: insufficient possibility of reactor components inspection during maintenance and operation in the future period; difficulties in prevention of accidents that could occur is equally dangerous for the reactor facility and the environment. Cleaning and decontamination of the complete heavy water system is needed before the reactor operation starts in order to avoid possible failures or accidental events. [Serbo-Croat] Zakljuceno je da je eksploatacija reakora u uslovima postojece aktivnosti teskovodnog sistema veoma opasna po sam reaktor i po personal. Dva osnovna razloga u prilog ovog zakljucka su: nedovoljna mogucnost kontrole ispravnosti svih elemenata reaktora u toku remonta i ekspolatacije u predstojecem periodu; teskoca borbe sa udesima, koji bi se eventualno dogodili podjednako je opasna po instalaciju i okolinu. Pre no sto se nastavi sa daljom ekspolatacijom reaktora potrebno ciscenje, dekontaminacija sistema teske vode kako bi se izbegla moguca ostecenja ili akcidentalne situacije.

  2. Trench reactor: an overview

    International Nuclear Information System (INIS)

    Spinrad, B.I.; Rohach, A.F.; Razzaque, M.M.; Sankoorikal, J.T.; Schmidt, R.S.; Lofshult, J.; Ramin, T.; Sokmen, N.; Lin, L.C.

    1988-01-01

    Recent fast, sodium-cooled reactor designs reflect new conditions. In nuclear energy these conditions are (a) emphasis on maintainability and operability, (b) design for more transparent safety, and (c) a surplus of uranium and enrichment availability that eases concerns about light water reactor fueling costs. In utility practice the demand is for less capital exposure, short construction time, smaller new unit sizes, and low capital cost. The PRISM, SAFR, and integral fast reactor (IFR) concepts are responses to these conditions. Fast reactors will not soon be deployed commercially, so more radical designs can be considered. The trench reactor is the product of such thinking. Its concepts are intended as contributions to the literature, which may be picked up by one of the existing programs or used in a new experimental project. The trench reactor is a thin-slab, pool-type reactor operated at very low power density and- for sodium-modest temperature. The thin slab is repeated in the sodium tank and the reactor core. The low power density permits a longer than conventional core height and a large-diameter fuel pin. Control is by borated steel slabs that can be lowered between the core and lateral sodium reflector. Shutdown is by semaphore slabs that can be swung into place just outside the control slabs. The paper presents major characteristics of the trench reactor that have been changed since the last report

  3. Laboratory Measurements of Biomass Cook-stove Emissions Aged in an Oxidation Flow Reactor: Influence of Combustion and Aging Conditions on Aerosols

    Science.gov (United States)

    Grieshop, A. P.; Reece, S. M.; Sinha, A.; Wathore, R.

    2016-12-01

    Combustion in rudimentary and improved cook-stoves used by billions in developing countries can be a regionally dominant contributor to black carbon (BC), primary organic aerosols (POA) and precursors for secondary organic aerosol (SOA). Recent studies suggest that SOA formed during photo-oxidation of primary emissions from biomass burning may make important contribution to its atmospheric impacts. However, the extent to which stove type and operating conditions affect the amount, composition and characteristics of SOA formed from the aging of cookstoves emissions is still largely undetermined. Here we present results from experiments with a field portable oxidation flow reactor (F-OFR) designed to assess aging of cook-stove emissions in both laboratory and field settings. Laboratory tests results are used to compare the quantity and properties of fresh and aged emissions from a traditional open fire and twp alternative stove designs operated on the standard and alternate testing protocols. Diluted cookstove emissions were exposed to a range of oxidant concentrations in the F-OFR. Primary emissions were aged both on-line, to study the influence of combustion variability, and sampled from batched emissions in a smog chamber to examine different aging conditions. Data from real-time particle- and gas-phase instruments and integrated filter samples were collected up and down stream of the OFR. The properties of primary emissions vary strongly with stove type and combustion conditions (e.g. smoldering versus flaming). Experiments aging diluted biomass emissions from distinct phases of stove operation (smoldering and flaming) showed peak SOA production for both phases occurred between 3 and 6 equivalent days of aging with slightly greater production observed in flaming phase emissions. Changing combustion conditions had a stronger influence than aging on POA+SOA `emission factors'. Aerosol Chemical Speciation Monitor data show a substantial evolution of aerosol

  4. H Reactor

    Data.gov (United States)

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

  5. Specific schedule conditions for the formation of personnel of A or B category working in nuclear facilities. Option nuclear reactor-borne

    CERN Document Server

    Int. At. Energy Agency, Wien

    2002-01-01

    This document describes the specific dispositions relative to the nuclear reactor-borne domain, for the formation to the conventional and radiation risks prevention of personnel of A or B category working in nuclear facilities. The application domain, the applicable documents, the liability, the specificity of the nuclear reactor-borne and of the retraining, the Passerelle formation, are presented. (A.L.B.)

  6. Osiris reactor descriptive report

    International Nuclear Information System (INIS)

    1976-03-01

    OSIRIS is a swimming pool reactor of 70 MW thermal power. Its main purpose is the irradiation of reactor materials in high neutron flux. A description is given of the air conditioning, ventilation, and radioactive gas removal system. (R.L.)

  7. Thermal hydraulic conditions inducing incipient cracking in the 900 MWe unit 93 D reactor coolant pump shafts; Pompes primaires 93 D des tranches de 900 MW. Conditions thermo-hydrauliques d`amorcage des fissures d`arbres

    Energy Technology Data Exchange (ETDEWEB)

    Bore, C.

    1995-12-31

    From 1987, 900 MWe plant operating feedback revealed cracking in the lower part of the reactor coolant pump shafts, beneath the thermal ring. Metallurgical examinations established that this was due to a thermal fatigue phenomenon known as thermal crazing, occurring after a large number of cycles. Analysis of thermal hydraulic conditions initiating the cracks does not allow exact quantification of the thermal load inducing cracking. Only qualitative analyses are thus possible, the first of which, undertaken by the pump manufacturer, Jeumont Industrie, showed that the cracks could not be due to the major transients (stop-start, injection cut-off), which were too few in number. Another explanation was then put forward: the thermal ring, shrunk onto the shaft it is required to protect against thermal shocks, loosens to allow an alternating downflow of cold water from the shaft seals and an upflow of hot water from the primary system. However, approximate calculations showed that the flow involved would be too slight to initiate the cracking observed. A more stringent analysis undertaken with the 2D flow analysis code MELODIE subsequently refuted the possibility of alternating flows beneath the ring establishing that only a hot water upflow occurred due to a `viscosity pump` phenomenon. Crack initiation was finally considered to be due to flowrate variations beneath the ring, with the associated temperature fluctuations. This flowrate fluctuation could be due to an unidentified transient phenomenon or to a variation in pump operating conditions. This analysis of the hydraulic conditions initiating the cracks disregards shaft surface residual stresses. These are tensile stresses and show that loads less penalizing than those initially retained could cause incipient cracking. Thermal ring modifications to reduce these risks were proposed and implemented. In addition, final metallurgical treatment of the shafts was altered and implemented. (Abstract Truncated)

  8. Enhanced degradation of phenolic compounds in coal gasification wastewater by a novel integration of micro-electrolysis with biological reactor (MEBR) under the micro-oxygen condition.

    Science.gov (United States)

    Ma, Weiwei; Han, Yuxing; Xu, Chunyan; Han, Hongjun; Ma, Wencheng; Zhu, Hao; Li, Kun; Wang, Dexin

    2018-03-01

    The aim of this work was to study an integration of micro-electrolysis with biological reactor (MEBR) for strengthening removal of phenolic compounds in coal gasification wastewater (CGW). The results indicated MEBR achieved high efficiencies in removal of COD and phenolic compounds as well as improvement of biodegradability of CGW under the micro-oxygen condition. The integrated MEBR process was more favorable to improvement of the structural stability of activated sludge and biodiversity of specific functional microbial communities. Especially, Shewanella and Pseudomonas were enriched to accelerate the extracellular electron transfer, finally facilitating the degradation of phenolic compounds. Moreover, MEBR process effectively relieved passivation of Fe-C filler surface and prolonged lifespan of Fe-C filler. Accordingly, the synergetic effect between iron-carbon micro-electrolysis (ICME) and biological action played a significant role in performance of the integrated process. Therefore, the integrated MEBR was a promising practical process for enhancing CGW treatment. Copyright © 2017 Elsevier Ltd. All rights reserved.

  9. Effect of bacterial lipase on anaerobic co-digestion of slaughterhouse wastewater and grease in batch condition and continuous fixed-bed reactor.

    Science.gov (United States)

    Affes, Maha; Aloui, Fathi; Hadrich, Fatma; Loukil, Slim; Sayadi, Sami

    2017-10-10

    This study aimed to investigate the effects of bacterial lipase on biogas production of anaerobic co-digestion of slaughterhouse wastewater (SHWW) and hydrolyzed grease (HG). A neutrophilic Staphylococcus xylosus strain exhibiting lipolytic activity was used to perform microbial hydrolysis pretreatment of poultry slaughterhouse lipid rich waste. Optimum proportion of hydrolyzed grease was evaluated by determining biochemical methane potential. A high biogas production was observed in batch containing a mixture of slaughterhouse composed of 75% SHWW and 25% hydrolyzed grease leading to a biogas yield of 0.6 L/g COD introduced. Fixed bed reactor (FBR) results confirmed that the proportion of 25% of hydrolyzed grease gives the optimum condition for the digester performance. Biogas production was significantly high until an organic loading rate (OLR) of 2 g COD/L. d. This study indicates that the use of biological pre-treatment and FBR for the co-digestion of SHWW and hydrolyzed grease is feasible and effective.

  10. CAMDYN: a new model to describe the axial motion of molten fuel inside the pin of a fast breeder reactor during accident conditions

    International Nuclear Information System (INIS)

    Peter, G.

    1991-01-01

    The new in-pin fuel motion model CAMDYN (Cavity Material Dynamics) describes the axial motion of both partially and fully molten fuel inside the pin of a fast breeder reactor during accident conditions. The motion of the two types of molten fuel and the imbedded fission gas bubbles is treated both before and after cladding failure. The basic modelling approach consists of the treatment of two one-dimensional flows which are coupled by interaction terms. Each of these flows is treated compressively and with axially variable flow cross sections. The mass and energy equations of both fields are solved explicitly using upwind differencing on a fixed Eulerian grid. The two momentum equations are solved simultaneously, using the convective momentum fluxes of the previous timestep. Both partially and fully molten fuel can move axially into a central hole extending to the plenum in the case of certain hollow pellet designs. The fuel temperature calculation includes the determination of a radial temperature profile. A simple conduction freezing model is included. After cladding failure, ejection into the coolant channel is modeled

  11. Full-scale model development of the WWER-440 reactor fuel rod bundle for core temperature regime study under reflooding conditions

    International Nuclear Information System (INIS)

    Bezrukov, Yu.A.; Logvinov, S.A.; Levchuk, S.V.; Nakladnov, V.D.; Onshin, V.P.; Sokolov, A.S.

    1982-01-01

    Consideration is given to the issues of a full scale WWER-440 fuel rod bundle imitation. An imitator contains a molybdenum heating rod inclosed in stainless steel shell. The shell diameter is 9 mm, the heated length is 2500 mm, the total len.o.th is 2855 mm. 125 fuel rod imitators are set in the bundle mock-up. The experiments were run on a test facility imitating the WWER-440 reactor primary loop, providing the conditions of the loop breaking. The mock-up thermal hydraulics has been studied during the refloodino. stage. The mock-up was heated up to predetermined initial temperature at a low power level with saturated steam cooling. Then the steam input was stopped, the power level rarapidly rised up to a given value and the cooling water injected. Simultaneously with water injection all the measured parameters monitoring was started. Both at the top spraying and combined cooling temperature oscillations in the upper and middle parts of the mock-up were observed. At the bottom reflooding the mock-up cooling down took more time, thereat temperature inthe upper part first slowly rised during reflooding then decreased and then dropped abruptly at thefront coming up [ru

  12. Investigation of the three-dimensional thermoelastic deformation of the core structure of a fast breeder reactor under stationary working conditions

    International Nuclear Information System (INIS)

    Yong-Su, Hoang.

    1976-12-01

    In this study a method is described which has been developed in order to calculate three-dimensional deformation of the reactor core, taking into account thermal expansion. Two problem areas are of particular importance: 1) The spatial deflection of subassemblies in specified flexible supports and with specified clearances; 2) The investigation of the equilibrium configurations of the subassemblies in the planes of clamping (problem of clamping plane). - The elementary theory of beam deflection has been used to calculate the deformation of subassemblies. However, particular problems have been encountered as a result of flexibly designed support configurations having some spatial clearances. The problem has essentially been solved in two steps: a) Uniqueness analysis of the beam-support configuration; b) Calculation of the support loads and bending line for the unique beam-support configuration. - Basic difficulties currently prevent the problem of clamping plane being solved in a satisfactory manner. Therefore, a simplified clamping model was used for supports without spatial clearance and a parametric study was performed for supports having spatial clearance. The computation method developed is applied to the MARK I core of SNR 300. Core deformations are calculated under different support conditions for the subassemblies in the grid plate and in the upper clamping plane. (orig./HR) [de

  13. Behavior of U3Si2 Fuel and FeCrAl Cladding under Normal Operating and Accident Reactor Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Gamble, Kyle Allan Lawrence [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hales, Jason Dean [Idaho National Lab. (INL), Idaho Falls, ID (United States); Barani, Tommaso [Idaho National Lab. (INL), Idaho Falls, ID (United States); Pizzocri, Davide [Idaho National Lab. (INL), Idaho Falls, ID (United States); Pastore, Giovanni [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    As part of the Department of Energy's Nuclear Energy Advanced Modeling and Simulation program, an Accident Tolerant Fuel High Impact Problem was initiated at the beginning of fiscal year 2015 to investigate the behavior of \\usi~fuel and iron-chromium-aluminum (FeCrAl) claddings under normal operating and accident reactor conditions. The High Impact Problem was created in response to the United States Department of Energy's renewed interest in accident tolerant materials after the events that occurred at the Fukushima Daiichi Nuclear Power Plant in 2011. The High Impact Problem is a multinational laboratory and university collaborative research effort between Idaho National Laboratory, Los Alamos National Laboratory, Argonne National Laboratory, and the University of Tennessee, Knoxville. This report primarily focuses on the engineering scale research in fiscal year 2016 with brief summaries of the lower length scale developments in the areas of density functional theory, cluster dynamics, rate theory, and phase field being presented.

  14. Effects of transient and non-uniform distribution of heat flux on intensity of heat transfer and burnout conditions in the channels of nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Vitaly Osmachkin [Russian Research Center ' Kurchatov Institute' 1, Kurchatov sq, Moscow 123182 (Russian Federation)

    2005-07-01

    Full text of publication follows: The influence of power transient, changes of flow rate, inlet temperatures or pressure in cores of nuclear reactors on heat transfer and burnout conditions in channels depend on rate of such violations. Non-uniform distribution of the heat flux is also important factor for heat transfer and development of crisis phenomenon. Such effects may be significant for NPPs safety. But they have not yet generally accepted interpretation. Steady state approach is often recommended for use in calculations. In the paper a review of experimental observed so-called non-equilibrium effects is presented. The effects of space and time factors are displaying due delay in reformation turbulence intensity, velocity, temperatures or void fraction profiles, water film flow on the surface of heated channels. For estimation of such effect different methods are used. Modern computer codes based on two or three fluids approaches are considered as most effective. But simple and clear correlations may light up the mechanics of effects on heat transfer and improve general understanding of scale and significance of the transient events. In the paper the simplified methods for assessment the influence of lags in the development of distributions of parameters of flow, the relaxation of temporal or space violations are considered. They are compared with more sophisticated approaches. Velocities of disturbance fronts moving along the channels are discussed also. (author)

  15. A detailed TEM and SEM study of Ni-base alloys oxide scales formed in primary conditions of pressurized water reactor

    International Nuclear Information System (INIS)

    Sennour, Mohamed; Marchetti, Loic; Martin, Frantz; Perrin, Stephane; Molins, Regine; Pijolat, Michele

    2010-01-01

    The oxide film formed on nickel-based alloys in pressurized water reactors (PWR) primary coolant conditions (325 o C, aqueous media) is very thin, in the range of 1-100 nm thick, depending on the surface state and on the corrosion test duration. The nature and the structure of this scale have been investigated by Transmission Electron Microscopy (TEM) and Scanning Electron Microscopy (SEM). TEM observations revealed an oxide layer divided in two parts. The internal layer was mainly composed of a continuous spinel layer, identified as a mixed iron and nickel chromite (Ni (1-x) Fe x Cr 2 O 4 ). Moreover, nodules of Cr 2 O 3 , with a size about 5 nm, were present at the interface between this spinel and the alloy. No chromium depletion was observed in the alloy, at the alloy/oxide interface. The external layer is composed of large crystallites corresponding to a spinel structure rich in iron (Ni (1-z) Fe (2+z) O 4 ) resulting from precipitation phenomena. SEM and TEM observations showed a link between the nucleation and/or the growth of crystallites of nickel ferrite and the crystallographic orientation of the substrate. A link between the presence of surface defects and the nucleation of the crystallites was also underlined by SEM observations. Partially hydrated nickel hydroxide, was also observed by TEM in the external scale. Based on these results, some considerations about the mechanism of formation of this oxide layer are discussed.

  16. HOMOGENEOUS NUCLEAR POWER REACTOR

    Science.gov (United States)

    King, L.D.P.

    1959-09-01

    A homogeneous nuclear power reactor utilizing forced circulation of the liquid fuel is described. The reactor does not require fuel handling outside of the reactor vessel during any normal operation including complete shutdown to room temperature, the reactor being selfregulating under extreme operating conditions and controlled by the thermal expansion of the liquid fuel. The liquid fuel utilized is a uranium, phosphoric acid, and water solution which requires no gus exhaust system or independent gas recombining system, thereby eliminating the handling of radioiytic gas.

  17. Validation of nonlinear FEA models of a thin-walled elbow under extreme loading conditions for Sodium-cooled Fast Reactors

    International Nuclear Information System (INIS)

    Watakabe, Tomoyoshi; Wakai, Takashi; Jin, Chuanrong; Usui, Yoshiya; Sakai, Shinkichi; Ooshika, Junji; Tsukimori, Kazuyuki

    2015-01-01

    For the purpose of confirming failure modes and safety margin, some studies on the ultimate strength of thin-walled piping components for Sodium-cooled Fast Reactors (SFRs) under extreme loading conditions such as large earthquakes have been reported these several years. Nonlinear finite element analysis has been applied in these studies to simulate buckling and yielding with large deformation, whose accuracy is dependent on the element type, the mesh size, the elasto-plastic model and so on. It is important to check the validation of a finite element model for nonlinear analysis especially under extreme loading conditions. This paper presents static and dynamic analyses of a thin-walled elbow with large deformation under large seismic loading, and discusses the validation of the FEA models comparing with experimental results. The finite element analysis models in this study are generated by shell elements for a stainless steel pipe elbow of diameter-to-thickness ratio 59:1 similar to the main pipe of SFRs, which is used for shaking table tests. At first, a static analysis is carried out for an in-plane monotonic bending test, in order to confirm that the shell element is appropriate to the large deformation analysis and the material parameters are proper for the strain level in the experiments. And then, a dynamic in-plane bending test with the maximum acceleration of 11.7G is simulated by the nonlinear FEA with stiffness-proportional damping. The influence of mesh sizes on results is investigated, to determine proper mesh sizes and reduce the computational cost. Finally, comparing the results of the FEM analyses with those of experiments, it is concluded that the appropriately generated FEA models are effective and give accurate results for nonlinear analyses of the thin-walled elbow under large seismic loading. (author)

  18. COUNTER-INTERVENTION, INVITATION, BOTH, OR NEITHER ?

    African Journals Online (AJOL)

    eliasn

    2 See Testimony by David H. Shinn, Somalia: U.S. Govrenemnt Policy and Challenges, ... a state of relative peace—one that can be best described as cold peace. ... denied sending its troops to Somalia until it formally declared war on 24 ...... Lebanese government, the invitation of the USSR by the dubious government of.

  19. Siting of research reactors

    International Nuclear Information System (INIS)

    1987-01-01

    The purpose of this document is to develop criteria for siting and the site-related design basis for research reactors. The concepts presented in this document are intended as recommendations for new reactors and are not suggested for backfitting purposes for facilities already in existence. In siting research reactors serious consideration is given to minimizing the effects of the site on the reactor and the reactor on the site and the potential impact of the reactor on the environment. In this document guidance is first provided on the evaluation of the radiological impact of the installation under normal reactor operation and accident conditions. A classification of research reactors in groups is then proposed, together with a different approach for each group, to take into account the relevant safety problems associated with facilities of different characteristics. Guidance is also provided for both extreme natural events and for man-induced external events which could affect the safe operation of the reactor. Extreme natural events include earthquakes, flooding for river or coastal sites and extreme meteorological phenomena. The feasibility of emergency planning is finally considered for each group of reactors

  20. Biological Action Spectra (invited paper)

    International Nuclear Information System (INIS)

    Gruijl, F.R. de

    2000-01-01

    Ultraviolet (UV) radiation induces a wide variety of biological responses: ranging in humans from well-known short-term effects like sunburn to long-term effects like skin cancer. The wavelength dependencies ('action spectra') of the responses can differ significantly, depending on the UV-targeted molecules (their absorption spectra), their localisation (transmission to the target depth) and the photochemical reactions involved (e.g. quantum yields, competing reaction). An action spectrum (e.g. of sunburn) is usually determined in a wavelength by wavelength analysis of the response. This is not always possible (e.g. in case of skin cancer), and an action spectrum may then be extracted mathematically from differences in responses to broadband UV sources of various spectral compositions (yielding 'biological spectral weights'). However, relative spectral weights may shift with exposure levels and contributions from different wavelengths may not always add up. Under these circumstances conventional analyses will yield different action spectra for different experimental conditions. (author)

  1. Transport of volatile fission products in the fuel-to-sheath gap of defective fuel elements during normal and reactor accident conditions

    International Nuclear Information System (INIS)

    Lewis, B.J.; Bonin, H.W.

    1995-01-01

    An analytical treatment has been used to model the vapour transport of radioactive fission products released into the fuel-to-sheath gap of defective nuclear fuel elements. The model accounts for both diffusive and bulk-convective transport. Convective transport becomes important as a result of a significant release of gaseous fission products into the gap during a high-temperature reactor accident. However, during normal reactor operation, diffusion is shown to be the dominant process of transport. The model is based on an analysis of several in-reactor tests with operating defective fuel elements, and high-temperature annealing experiments with irradiated fuel specimens. ((orig.))

  2. Thai research reactor

    International Nuclear Information System (INIS)

    Aramrattana, M.

    1987-01-01

    The Office of Atomic Energy for Peace (OAEP) was established in 1962, as a reactor center, by the virtue of the Atomic Energy for Peace Act, under operational policy and authority of the Thai Atomic Energy for Peace Commission (TAEPC); and under administration of Ministry of Science, Technology and Energy. It owns and operates the only Thai Research Reactor (TRR-1/M1). The TRR-1/M1 is a mixed reactor system constituting of the old MTR type swimming pool, irradiation facilities and cooling system; and TRIGA Mark III core and control instrumentation. The general performance of TRR-1/M1 is summarized in Table I. The safe operation of TRR-1/M1 is regulated by Reactor Safety Committee (RSC), established under TAEPC, and Health Physics Group of OAEP. The RCS has responsibility and duty to review of and make recommendations on Reactor Standing Orders, Reactor Operation Procedures, Reactor Core Loading and Requests for Reactor Experiments. In addition,there also exist of Emergency Procedures which is administered by OAEP. The Reactor Operation Procedures constitute of reactor operating procedures, system operating procedures and reactor maintenance procedures. At the level of reactor routine operating procedures, there is a set of Specifications on Safety and Operation Limits and Code of Practice from which reactor shift supervisor and operators must follow in order to assure the safe operation of TRR-1/M1. Table II is the summary of such specifications. The OAEP is now upgrading certain major components of the TRR-1/M1 such as the cooling system, the ventilation system and monitoring equipment to ensure their adequately safe and reliable performance under normal and emergency conditions. Furthermore, the International Atomic Energy Agency has been providing assistance in areas of operation and maintenance and safety analysis. (author)

  3. Reactor BR2. Introduction

    International Nuclear Information System (INIS)

    Gubel, P.

    2002-01-01

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

  4. Reactor BR2. Introduction

    Energy Technology Data Exchange (ETDEWEB)

    Gubel, P

    2002-04-01

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

  5. Biological Action Spectra (invited paper)

    Energy Technology Data Exchange (ETDEWEB)

    Gruijl, F.R. de

    2000-07-01

    Ultraviolet (UV) radiation induces a wide variety of biological responses: ranging in humans from well-known short-term effects like sunburn to long-term effects like skin cancer. The wavelength dependencies ('action spectra') of the responses can differ significantly, depending on the UV-targeted molecules (their absorption spectra), their localisation (transmission to the target depth) and the photochemical reactions involved (e.g. quantum yields, competing reaction). An action spectrum (e.g. of sunburn) is usually determined in a wavelength by wavelength analysis of the response. This is not always possible (e.g. in case of skin cancer), and an action spectrum may then be extracted mathematically from differences in responses to broadband UV sources of various spectral compositions (yielding 'biological spectral weights'). However, relative spectral weights may shift with exposure levels and contributions from different wavelengths may not always add up. Under these circumstances conventional analyses will yield different action spectra for different experimental conditions. (author)

  6. Exposure from Appliances (invited paper)

    Energy Technology Data Exchange (ETDEWEB)

    Cartwright, R.A

    1999-07-01

    The dosemeter studies on exposures to ELF electromagnetic irradiation from domestic equipment all suggest the whole-body doses are low, whilst some exposure to extremities could be high but of brief duration. Old style electric overblankets, however, are an exception and produce high fields and relatively high whole-body doses if switched on during the night. Relatively few epidemiological studies have addressed these issues. All of the studies have associated problems of interpretations. Two isolated studies throw up on association with the frequency of spontaneous abortion and electric blanket use whilst another links adult AML and electric shaver usage. Both results could be fortuitous. More consistency appears from three studies of childhood leukaemia. Here statistically significant associations between electric blanket use in pregnancies appear in two separate studies, as does hair dryer use in the case of children: other appliances use associations have been reported. These results are critically assessed. The United Kingdom Childhood Cancer Study (UKCCS) addresses some of these issues as part of a wider ranging assessment of EMF at home and at school. External sources are incorporated with the measurement of domestic ambient conditions in an attempt to make an overall assessment of total exposure. (author)

  7. Invitation to medical image processing

    International Nuclear Information System (INIS)

    Kitasaka, Takayuki; Suenaga, Yasuhito; Mori, Kensaku

    2010-01-01

    This medical essay explains the present state of CT image processing technology about its recognition, acquisition and visualization for computer-assisted diagnosis (CAD) and surgery (CAS), and future view. Medical image processing has a series of history of its original start from the discovery of X-ray to its application to diagnostic radiography, its combination with the computer for CT, multi-detector raw CT, leading to 3D/4D images for CAD and CAS. CAD is performed based on the recognition of normal anatomical structure of human body, detection of possible abnormal lesion and visualization of its numerical figure into image. Actual instances of CAD images are presented here for chest (lung cancer), abdomen (colorectal cancer) and future body atlas (models of organs and diseases for imaging), a recent national project: computer anatomy. CAS involves the surgical planning technology based on 3D images, navigation of the actual procedure and of endoscopy. As guidance to beginning technological image processing, described are the national and international community like related academic societies, regularly conducting congresses, textbooks and workshops, and topics in the field like computed anatomy of an individual patient for CAD and CAS, its data security and standardization. In future, protective medicine is in authors' view based on the imaging technology, e.g., daily life CAD of individuals ultimately, as exemplified in the present body thermometer and home sphygmometer, to monitor one's routine physical conditions. (T.T.)

  8. Hillslope Soils and Life (Invited)

    Science.gov (United States)

    Amundson, R.; Owen, J. J.; Heimsath, A. M.; Yoo, K.; Dietrich, W. E.

    2013-12-01

    That hillslope processes are impacted by biology has been long understood, but the complexities of the abiotic-biotic processes and their feedbacks are quantitatively emerging with the growing body of pertinent literature. The concept that plants modulate both the disaggregation and transport of soil particles on hillslopes was clearly articulated by G.K. Gilbert. Yet earlier, James Hutton (starting from very different intellectual boundary conditions) argued that soil, which results from the dynamic balance of rock destruction and removal, is a prerequisite for plants - a concept that underscores the need to more deeply examine the feedback of geomorphic processes on terrestrial ecosystems. We compiled the results of recent studies that have been conducted on gentle convex hillslopes across a broad range of rainfall. We found that vegetated landscapes appear to have strong controls on hillslope soil thickness, landscape denudation rates, and soil residence times. The restricted range in residence times - despite large differences in climate - appear in turn to sustain relatively high levels of both nitrogen (N) and phosphorus (P) fertility, suggesting ecological resilience and resistance to non-anthropogenic environmental perturbations. At the most arid end of Earth's climate vegetation disappears, but not all water. The loss of plants shifts soil erosion to abiotic processes, with a corresponding thinning or loss of the soil mantle. This reinforces the hypothesis that a planet without vegetation, but with a hydrologic cycle, would be largely devoid of soil-mantled hillslopes and would be driven toward hillslope morphologies that differ from the familiar convex-up forms of biotic landscapes. While our synthesis of the effects of vegetation on soil production and soil thickness provides a quantitative view of the suggestions of Gilbert, it also identifies that vegetation itself responds to the geomorphic processes, as believed by Hutton. There is a complex

  9. Calculation estimation of the possibilities of using the off-reactor control of WWER for the determination of change in the multiplication properties of the core in changing its configuration under the reactor shutdown conditions

    International Nuclear Information System (INIS)

    Dementiev, V.G.; Kamyshan, A.N.; Shishkov, L.K.

    2001-01-01

    In the refueling and in the process of first core loading the control of subcriticality for ensuring the reactor safety becomes important. In principle this control can be provided by the use of off-reactor neutron detectors. However a low neutron flux in the area of detectors location and the presence of spatial effects make it necessary to employ spatial methods of accumulation and processing the experimental information. This paper gives a description of plans of works in this direction and the preliminary estimation of the effect (Authors)

  10. Challenge, Confrontation, and Exhortation as Intentional Invitations by Professional Helpers.

    Science.gov (United States)

    Schmidt, John J.

    1996-01-01

    Examines intentional invitations that challenge, confront, exhort, and persuade people to change their behaviors. Assumes that the sender controls the "intention" and that the receiver determines the degree of "inviting." Suggests that elements of the invitational model serve as a framework to create acceptable inducements in…

  11. Berger Peter L., Invitation à la sociologie

    Directory of Open Access Journals (Sweden)

    Nicolas Marquis

    2011-03-01

    Full Text Available L’invitation à la sociologie de Peter Berger est la réédition d’un ouvrage initiale­ment paru en 1963. Né en 1929, P. Berger est un des grands noms de la sociologie amé­ricaine actuelle, notamment connu chez nous pour un ouvrage coécrit avec Th. Luck­­mann et paru trois ans après la première édition de Invitation à la sociologie : The Social Construction of Reality, qui constitue une étape importante dans l’approche dite du constructivisme social en réactivant la tradition phénoménologique en...

  12. Nuclear power reactor safety

    International Nuclear Information System (INIS)

    Pon, G.A.

    1976-10-01

    This report is based on the Atomic Energy of Canada Limited submission to the Royal Commission on Electric Power Planning on the safety of CANDU reactors. It discusses normal operating conditions, postulated accident conditions, and safety systems. The release of radioactivity under normal and accident conditions is compared to the limits set by the Atomic Energy Control Regulations. (author)

  13. Invited presentations. College on soil physics 2003

    International Nuclear Information System (INIS)

    Gabriels, D.M.; Ghirardi, G.; Nielsen, D.R.; Pla Sentis, I.; Skidmore, E.L.

    2004-01-01

    The present book is a partial compilation of contributions from selected former participants of the College on Soil Physics invited to make presentations related to their achievements as a result of attending the College. It also serves as a testimony of the existing links between soil physicists throughout the world strengthened by the support and programs of the International Centre for Theoretical Physics originally envisioned by Abdus Salam to foster the growth of advanced studies and physics research in developing countries

  14. Invited presentations. College on soil physics 2003

    Energy Technology Data Exchange (ETDEWEB)

    Gabriels, D M [Univ. Ghent (Belgium); Ghirardi, G [Univ. Trieste (Italy); Nielsen, D R [Univ. California (United States); Pla Sentis, I [Univ. Lleida (Spain); Skidmore, E L [Kansas State Univ. (United States)

    2004-05-15

    The present book is a partial compilation of contributions from selected former participants of the College on Soil Physics invited to make presentations related to their achievements as a result of attending the College. It also serves as a testimony of the existing links between soil physicists throughout the world strengthened by the support and programs of the International Centre for Theoretical Physics originally envisioned by Abdus Salam to foster the growth of advanced studies and physics research in developing countries.

  15. Substitution of the old console with a fully computerized one. Major maintenance and enhancements of the ventilation and air conditioning system of the reactor building. Research activities at the TRIGA L.E.N.A. plant in Pavia, Italy

    International Nuclear Information System (INIS)

    Orvini, E.; Piazzoli, A.; Borio, A.

    2008-01-01

    The TRIGA Mark II reactor of the University of Pavia was operated in the last two years on a routine basis accomplishing different purposes: - Development of B.N.C.T. for diffused tumours of liver; - Neutron activation analysis in matrix of geochemistry, archaeology and environment interest; - Electron spin resonance (ESR) study of radical processes; - Study of trace elements impact in environmental matrix and human health; - High purity control on electronic samples-microchips; - Fast neutron radiation damage investigation; - Certification of standard reference materials and data quality assurance by neutron activation analysis; - Age evaluation by Ar-K method in geological matrix; - Trace elements determination for provenience studies in archaeology; - Basic experiments on fission of Am-242 layers for the space nuclear engine project; - In field search of explosive using prompt gamma emission induced by neutron capture on nitrates. In the period of time between July 1998 and June 2000 the reactor was operated at full power (i.e. 250 kW) for a total amount of 1375 hours. The total fuel element burn-up was 15.07 MWD. During this period of time two major upgrading activities were planned: the installation of a new Instrumentation and Control System (ICS) for the reactor and the installation of a new Air Conditioning and Filtering System (ACFS) for the reactor building. The new ICS is a microprocessor-based design, incorporating the use of one logarithmic wide range neutron flux monitoring channel (NLW-1000), two current mode neutron monitoring safety channels (NP-1000 and NPP-1000) and a linear multi-range neutron flux monitoring channel (Keithley 485). The ICS configuration was personalized by General Atomic according to the requirements and to the technical prescriptions of the Pavia reactor, especially for what concerns the SCRAM inputs and the detectors architecture. Besides, since one the two safety channels, the NPP-1000, has been developed as an advanced

  16. Radio-contaminant behaviour in the cover-gas space and the containment building of a sodium-cooled fast reactor in accident conditions

    International Nuclear Information System (INIS)

    Mathe, Emmanuel

    2014-01-01

    In the context of the Generation IV initiative, the consequences of a severe-accident (SA) in a sodium-cooled fast reactor must be studied. A SFR (Sodium cooled Fast Reactor) severe accident involves the disruption of the core by super-criticality involving the destruction of a certain number of fuel assemblies. Subsequently the interaction between hot fuel and liquid sodium can lead to a vapor explosion which could create a breach in the primary system. Some contaminated liquid sodium would thus be ejected into the containment building. In this situation, the evaluation of potential releases to the environment (the source term) must forecast the quantity and the chemical speciation of the radio-contaminants likely to be released from the containment building. One critical risk of a SA is the production of contaminated aerosols in the containment building by spray ejection of primary-system sodium. Being pyrophoric, the sodium droplets react with oxygen first oxidizing then burning, with significant heat of combustion. As well as evaluating the consequences of a pressure rise inside the containment, the evolution of the sodium must be assessed since not only is it activated and contaminated but, in oxide form, very toxic. Ultimately, the aerosols are the main radiological risk acting as the vector for radionuclide transport to the environment in the event of a problem with the confinement. These aerosols could evolve and interact with the FP (Fissile Products) and these interactions could modify the physical and chemical nature of the PF. We model a large part of the events that occur during a SA inside a SFR from the sodium spray fire to the reaction between sodium aerosols and PF (iodine). At first, we develop a numerical model (NATRAC) that simulates the sodium spray fire, calculates the temperature and the pressure inside the containment as well as the mass of aerosols produced during this kind of fire. The simulation has been validated with different

  17. A Suggestion for Restoring Public Support for Public Schools: Learn to Invite Patrons.

    Science.gov (United States)

    Roebuck, Flora N.; Russell, Gene; Wedel, Mack

    1998-01-01

    Proposes that the increasing disavowal of public schools results from experience with day-by-day disinviting behaviors. The suggested remedy is for educators to evaluate their on-the-job responses to their patrons. Wherever subminimal interpersonal conditions are identified, participants should learn to invite people to education. Presents a case…

  18. Effects of Personalization and Invitation Email Length on Web-Based Survey Response Rates

    Science.gov (United States)

    Trespalacios, Jesús H.; Perkins, Ross A.

    2016-01-01

    Individual strategies to increase response rate and survey completion have been extensively researched. Recently, efforts have been made to investigate a combination of interventions to yield better response rates for web-based surveys. This study examined the effects of four different survey invitation conditions on response rate. From a large…

  19. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2002-01-01

    SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised

  20. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2001-04-01

    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised.

  1. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2002-04-01

    SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised.

  2. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2001-01-01

    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised

  3. Reactor operation

    CERN Document Server

    Shaw, J

    2013-01-01

    Reactor Operation covers the theoretical aspects and design information of nuclear reactors. This book is composed of nine chapters that also consider their control, calibration, and experimentation.The opening chapters present the general problems of reactor operation and the principles of reactor control and operation. The succeeding chapters deal with the instrumentation, start-up, pre-commissioning, and physical experiments of nuclear reactors. The remaining chapters are devoted to the control rod calibrations and temperature coefficient measurements in the reactor. These chapters also exp

  4. Reactor safeguards

    CERN Document Server

    Russell, Charles R

    1962-01-01

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

  5. Removal of organic substances and oxidation of ammonium nitrogen by a down-flow hanging sponge (DHS) reactor under high salinity conditions.

    Science.gov (United States)

    Uemura, Shigeki; Suzuki, Saori; Abe, Kenichi; Kubota, Keiichi; Yamaguchi, Takashi; Ohashi, Akiyoshi; Takemura, Yasuyuki; Harada, Hideki

    2010-07-01

    A down-flow hanging sponge (DHS) reactor, constructed by connecting three identical treatment units in series, was fed with highly saline artificial coke-plant wastewater containing 1400 mg L(-1) of phenol in terms of chemical oxygen demand (COD) and 500 mg-NL(-1) of ammonium nitrogen. The COD was removed by the 1st unit, achieving 92% removal at an average COD loading rate of 3.0 kg-COD m(-3)d(-1) for all units, with oxidation of ammonium nitrogen occurring primarily in the two downstream units. Microbial assays of the different units of the reactor revealed greater numbers of nitrifying bacteria in the 2nd and 3rd units than in the 1st unit, corresponding with the observed ammonium oxidation pattern of the reactor. These findings suggest that a succession of microflora was successfully established along the DHS. Copyright (c) 2010 Elsevier Ltd. All rights reserved.

  6. Quantification of the inert chemical oxygen demand of raw wastewater and evaluation of soluble microbial product production in demo-scale upflow anaerobic sludge blanket reactors under different operational conditions.

    Science.gov (United States)

    Aquino, Sergio F; Gloria, Roberto M; Silva, Silvana Q; Chernicharo, Carlos A L

    2009-06-01

    This paper investigates the production of soluble microbial products (SMPs) in demonstration-scale upflow anaerobic sludge blanket reactors operated under different conditions and fed with raw wastewater. The results showed that 9.2 +/- 1.3% of the influent soluble chemical oxygen demand (COD) could be considered inert to anaerobic treatment and that the amount of COD produced by biomass varied from 30 to 70 mg x L(-1), accounting for 45 to 63% of the soluble effluent COD. The accumulation of SMP appeared to be dependent on the hydraulic retention time (HRT) applied to the reactors, with a larger accumulation of SMP observed at the lowest HRT (5 hours); this may have been due to stress conditions caused by high upflow velocity (1.1 m x h(-1)). In terms of residual COD characterization, ultrafiltration results showed that higher amounts of high molecular weight compounds were found when HRT was the lowest (5 hours), and that the molecular weight distribution depended on the operational condition of the reactors. Biodegradability tests showed that the low and high molecular weight SMPs were only partially degraded anaerobically (10 to 60%) and that the high molecular weight SMPs were difficult to degrade aerobically.

  7. Pretreatment of vinasse from the sugar refinery industry under non-sterile conditions by Trametes versicolor in a fluidized bed bioreactor and its effect when coupled to an UASB reactor.

    Science.gov (United States)

    España-Gamboa, Elda; Vicent, Teresa; Font, Xavier; Dominguez-Maldonado, Jorge; Canto-Canché, Blondy; Alzate-Gaviria, Liliana

    2017-01-01

    During hydrous ethanol production from the sugar refinery industry in Mexico, vinasse is generated. Phenolic compounds and melanoidins contribute to its color and make degradation of the vinasse a difficult task. Although anaerobic digestion (AD) is feasible for vinasse treatment, the presence of recalcitrant compounds can be toxic or inhibitory for anaerobic microorganism. Therefore, this study presents new data on the coupled of the FBR (Fluidized Bed Bioreactor) to the UASB (Upflow Anaerobic Sludge Blanket) reactor under non-sterile conditions by T. versicolor . Nevertheless, for an industrial application, it is necessary to evaluate the performance in this kind of proposal system. Therefore, this study used a FBR for the removal of phenolic compounds (67%) and COD (38%) at non-sterile conditions. Continuous operation of the FBR was successfully for 26 days according to the literature. When the FBR was coupled to the UASB reactor, we obtained a better quality of effluent, furthermore methane content and yield were 74% and 0.18 m 3 CH 4 / kg COD removal respectively. This study demonstrated the possibility of using for an industrial application the coupled of the FBR to the UASB reactor under non-sterile conditions. Continuous operation of the FBR was carried out successfully for 26 days, which is the highest value found in the literature.

  8. Indian Chemical Engineering Congress 1995: 48th annual session of Indian Institute of Chemical Engineers: abstracts and invited lectures

    International Nuclear Information System (INIS)

    1995-01-01

    The 48th Annual Session of Indian Institute of Chemical Engineers was held in Kalpakkam during December 27-30, 1995. The book contains the proceeding of the conference, both abstracts and invited lectures. The topics covered included various aspects pertaining to chemical engineering and technology along with the chemical and engineering processes relevant to nuclear fuel cycle like uranium ore processing, fuel fabrication, reactor operation, fuel reprocessing and radioactive waste management. Papers relevant to INIS are indexed separately

  9. Nuclear reactors

    International Nuclear Information System (INIS)

    Middleton, J.E.

    1977-01-01

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

  10. Reactor decommissioning

    International Nuclear Information System (INIS)

    Lawton, H.

    1984-01-01

    A pioneering project on the decommissioning of the Windscale Advanced Gas-cooled Reactor, by the UKAEA, is described. Reactor data; policy; waste management; remote handling equipment; development; and recording and timescales, are all briefly discussed. (U.K.)

  11. RA Reactor

    International Nuclear Information System (INIS)

    1978-02-01

    In addition to basic characteristics of the RA reactor, organizational scheme and financial incentives, this document covers describes the state of the reactor components after 18 years of operation, problems concerned with obtaining the licence for operation with 80% fuel, problems of spent fuel storage in the storage pool of the reactor building and the need for renewal of reactor equipment, first of all instrumentation [sr

  12. Multiregion reactors

    International Nuclear Information System (INIS)

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

    1979-08-01

    The study of reflected reactors can be done employing the multigroup diffusion method. The neutron conservation equations, inside the intervals, can be written by fluxes and group constants. A reflected reactor (one and two groups) for a slab geometry is studied, aplying the continuity of flux and current in the interface. At the end, the appropriated solutions for a infinite cylindrical reactor and for a spherical reactor are presented. (Author) [pt

  13. Experimental demonstration of the reverse flow catalytic membrane reactor concept for energy efficient syngas production. Part 1: Influence of operating conditions

    NARCIS (Netherlands)

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

    2007-01-01

    In this contribution the technical feasibility of the reverse flow catalytic membrane reactor (RFCMR) concept with porous membranes for energy efficient syngas production is investigated. In earlier work an experimental proof of principle was already provided [Smit, J., Bekink, G.J., van Sint

  14. Nuclear reactor

    International Nuclear Information System (INIS)

    Hattori, Sadao; Sato, Morihiko.

    1994-01-01

    Liquid metals such as liquid metal sodium are filled in a reactor container as primary coolants. A plurality of reactor core containers are disposed in a row in the circumferential direction along with the inner circumferential wall of the reactor container. One or a plurality of intermediate coolers are disposed at the inside of an annular row of the reactor core containers. A reactor core constituted with fuel rods and control rods (module reactor core) is contained at the inside of each of the reactor core containers. Each of the intermediate coolers comprises a cylindrical intermediate cooling vessels. The intermediate cooling vessel comprises an intermediate heat exchanger for heat exchange of primary coolants and secondary coolants and recycling pumps for compulsorily recycling primary coolants at the inside thereof. Since a plurality of reactor core containers are thus assembled, a great reactor power can be attained. Further, the module reactor core contained in one reactor core vessel may be small sized, to facilitate the control for the reactor core operation. (I.N.)

  15. Nuclear reactor safety system

    International Nuclear Information System (INIS)

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

    1983-01-01

    The invention provides a safety system for a nuclear reactor which uses a parallel combination of computer type look-up tables each of which receives data on a particular parameter (from transducers located in the reactor system) and each of which produces the functional counterpart of that particular parameter. The various functional counterparts are then added together to form a control signal for shutting down the reactor. The functional counterparts are developed by analysis of experimental thermal and hydraulic data, which are used to form expressions that define safe conditions

  16. Reactor power control system

    International Nuclear Information System (INIS)

    Tomisawa, Teruaki.

    1981-01-01

    Purpose: To restore reactor-power condition in a minimum time after a termination of turbine bypass by reducing the throttling of the reactor power at the time of load-failure as low as possible. Constitution: The transient change of the internal pressure of condenser is continuously monitored. When a turbine is bypassed, a speed-control-command signal for a coolant recirculating pump is generated according as the internal pressure of the condenser. When the signal relating to the internal pressure of the condenser indicates insufficient power, a reactor-control-rod-drive signal is generated. (J.P.N.)

  17. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Butterfield, C.E.; Waite, E.

    1982-01-01

    A nuclear reactor fuel element comprising a column of vibration compacted fuel which is retained in consolidated condition by a thimble shaped plug. The plug is wedged into gripping engagement with the wall of the sheath by a wedge. The wedge material has a lower coefficient of expansion than the sheath material so that at reactor operating temperature the retainer can relax sufficient to accommodate thermal expansion of the column of fuel. (author)

  18. OECD Halden reactor project

    International Nuclear Information System (INIS)

    1978-01-01

    This report summarizes the activities of the OECD Halden Reactor Project for the year 1976. The main items reported on are: a) the process supervision and control which have focused on core monitoring and control, and operator-process communication; b) the fuel performance and safety behavior which have provided data and analytical descriptions of the thermal, mechanical and chemical behavior of fuel under various operating conditions; c) the reactor operations and d) the administration and finance

  19. L'invisible invité

    OpenAIRE

    Frémy, Anne

    2012-01-01

    Tiré du site Internet de Onestar Press: ""They’re a kind of cabin or hut, very basic and awkwardly assembled, wedged into the space of the balcony between the guardrail and the wall of the building.The vertical sides are made of plywood, canvas or tarp, and are topped by a vegetal roofing that is more or less dry, more or less green, threaded with palm wattle. These huts are opaque and seem to communicate with apartments…" Excerpt from L’Invisible invité by Anne Frémy, english by Bruce Bender...

  20. 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'

    International Nuclear Information System (INIS)

    Ura, Tamaki; Takamasa, Tomoji; Nishimura, Hajime

    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)

  1. 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)

  2. Nuclear power reactors

    International Nuclear Information System (INIS)

    1982-11-01

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

  3. Research reactors

    International Nuclear Information System (INIS)

    Merchie, Francois

    2015-10-01

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

  4. Reactor physics and reactor computations

    International Nuclear Information System (INIS)

    Ronen, Y.; Elias, E.

    1994-01-01

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

  5. Reactor scram device for FBR type reactor

    International Nuclear Information System (INIS)

    Kumasaka, Katsuyuki; Arashida, Genji; Itooka, Satoshi.

    1991-01-01

    In a control rod attaching structure in a reactor scram device of an FBR type reactor, an anti-rising mechanism proposed so far against external upward force upon occurrence of earthquakes relies on the engagement of a mechanical structure but temperature condition is not taken into consideration. Then, in the present invention, a material having curie temperature characteristics and which exhibits ferromagnetism only under low temperature condition and a magnet device are disposed to one of a movable control rod and a portion secured to the reactor. Alternatively, a bimetal member or a shape memory alloy which actuates to fix to the mating member only under low temperature condition is secured. The fixing device is adapted to operate so as to secure the control rods when the low temperature state is caused depending on the temperature condition. With such a constitution, when the control rods are separated from a driving device, they are prevented from rising even if they undergo external upward force due to earthquakes and so on, which can improve the reactor safety. (N.H.)

  6. Test report: Preliminary tests for the High Flux Reactor: Experimental determination of flow redistribution conditions at pressures between 4 and 5 kg/cm2 abs in a rectangular channel 2 mm thick and 60 cm long

    International Nuclear Information System (INIS)

    Schleisiek, K.; Dumaine, J.C.

    1989-01-01

    In the context of safety research for the OSIRIS reactor, tests have been performed on the Super BOB cell with a view to determining experimentally the internal characteristics (or ''S'' curves) of a channel with a rectangular heating cross-section 2 x 38 mm and 600 mm long. During these tests the maximum pressure at the channel exit was brought to 3 kg/cm 2 abs. The pressurization level in the High Flux Reactor will be higher. That is why tests have been carried out at maximum pressure of 5 kg/cm 2 abs allowable on the ''super BOB'' loop without modifying it. The first objective of this test series was to determine the ''S'' curves and the exchange coefficients experimentally. This document discusses the test conditions and test results

  7. Chernobyl reactor transient simulation study

    International Nuclear Information System (INIS)

    Gaber, F.A.; El Messiry, A.M.

    1988-01-01

    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

  8. Heterogeneous gas core reactor

    International Nuclear Information System (INIS)

    Han, K.I.

    1977-01-01

    Preliminary investigations of a heterogeneous gas core reactor (HGCR) concept suggest that this potential power reactor offers distinct advantages over other existing or conceptual reactor power plants. One of the most favorable features of the HGCR is the flexibility of the power producing system which allows it to be efficiently designed to conform to a desired optimum condition without major conceptual changes. The arrangement of bundles of moderator/coolant channels in a fissionable gas or mixture of gases makes a truly heterogeneous nuclear reactor core. It is this full heterogeneity for a gas-fueled reactor core which accounts for the novelty of the heterogeneous gas core reactor concept and leads to noted significant advantages over previous gas core systems with respect to neutron and fuel economy, power density, and heat transfer characteristics. The purpose of this work is to provide an insight into the design, operating characteristics, and safety of a heterogeneous gas core reactor system. The studies consist mainly of neutronic, energetic and kinetic analyses of the power producing and conversion systems as a preliminary assessment of the heterogeneous gas core reactor concept and basic design. The results of the conducted research indicate a high potential for the heterogeneous gas core reactor system as an electrical power generating unit (either large or small), with an overall efficiency as high as 40 to 45%. The HGCR system is found to be stable and safe, under the conditions imposed upon the analyses conducted in this work, due to the inherent safety of ann expanding gaseous fuel and the intrinsic feedback effects of the gas and water coolant

  9. Student Theological Research as an Invitation

    Directory of Open Access Journals (Sweden)

    William Badke

    2011-12-01

    Full Text Available Seminary students, despite having graduated from bachelors programs, struggle to make sense of the goals, processes, skills, and resources of research as graduate students. Beyond brief introductions to research, a scattered number of seminaries have developed either a separate theological information literacy course or have taken a through-the-curriculum approach to enhancing the information abilities of students. The former, however, separates information literacy from the curriculum, while the latter is difficult to implement and maintain. Living in a world of information glut, seminary professors are finding that traditional information dissemination models of education are becoming less viable. What is more, such models tend to teach students about a discipline rather than inviting them into it. These problems present a unique opportunity to place the teaching of information literacy at the foundation of theological education. With such an approach, students may be invited into the disciplines of their professors and enabled to practice these disciplines, thus becoming equipped to turn knowledge into praxis.

  10. Research reactors

    International Nuclear Information System (INIS)

    Kowarski, L.

    1955-01-01

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

  11. Alternating anoxic feast/aerobic famine condition for improving granular sludge formation in sequencing batch airlift reactor at reduced aeration rate.

    Science.gov (United States)

    Wan, Junfeng; Bessière, Yolaine; Spérandio, Mathieu

    2009-12-01

    In this study the influence of a pre-anoxic feast period on granular sludge formation in a sequencing batch airlift reactor is evaluated. Whereas a purely aerobic SBR was operated as a reference (reactor R2), another reactor (R1) was run with a reduced aeration rate and an alternating anoxic-aerobic cycle reinforced by nitrate feeding. The presence of pre-anoxic phase clearly improved the densification of aggregates and allowed granular sludge formation at reduced air flow rate (superficial air velocity (SAV)=0.63cms(-1)). A low sludge volume index (SVI(30)=45mLg(-1)) and a high MLSS concentration (9-10gL(-1)) were obtained in the anoxic/aerobic system compared to more conventional results for the aerobic reactor. A granular sludge was observed in the anoxic/aerobic system whilst only flocs were observed in the aerobic reference even when operated at a high aeration rate (SAV=2.83cms(-1)). Nitrification was maintained efficiently in the anoxic/aerobic system even when organic loading rate (OLR) was increased up to 2.8kgCODm(-3)d(-1). In the contrary nitrification was unstable in the aerobic system and dropped at high OLR due to competition between autotrophic and heterotrophic growth. The presence of a pre-anoxic period positively affected granulation process via different mechanisms: enhancing heterotrophic growth/storage deeper in the internal anoxic layer of granule, reducing the competition between autotrophic and heterotrophic growth. These processes help to develop dense granular sludge at a moderate aeration rate. This tends to confirm that oxygen transfer is the most limiting factor for granulation at reduced aeration. Hence the use of an alternative electron acceptor (nitrate or nitrite) should be encouraged during feast period for reducing energy demand of the granular sludge process.

  12. In-pile testing of ITER first wall mock-ups at relevant thermal loading conditions in the LVR-15 nuclear research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kysela, Jan [Research Centre Rez, Hlavni 130, 250 68 Husinec-Rez (Czech Republic); Entler, Slavomir, E-mail: slavomir.entler@cvrez.cz [Research Centre Rez, Hlavni 130, 250 68 Husinec-Rez (Czech Republic); Vsolak, Rudolf; Klabik, Tomas [Research Centre Rez, Hlavni 130, 250 68 Husinec-Rez (Czech Republic); Zlamal, Ondrej [CEZ, Duhova 2/1444, 140 53 Praha 4 (Czech Republic); Bellin, Boris; Zacchia, Francesco [Fusion for Energy, Josep Pla, 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain)

    2015-10-15

    Highlights: • Irradiated thermal fatigue testing of the ITER primary first wall mock-ups. • Cyclic heat flux of 0.5 MW/m{sup 2} in the neutron field of the nuclear reactor core. • 17,040 thermal cycles. • Radiation damage in the range of 0.41–1.17 dpa depending on the material. - Abstract: The TW3 in-pile rig enabled the thermal fatigue testing of ITER primary first wall mock-ups in the core of the nuclear reactor. This experiment investigated the neutron irradiation influence on the design performance under high heat flux testing. A thermal flux of 0.5 MW/m{sup 2} in the neutron field of the core of the LVR-15 nuclear reactor was applied. Within the scope of the tests with simultaneous neutron irradiation, the TW3 rig reached a record of 17,040 thermal cycles with the radiation damage in the range of 0.41–1.17 dpa depending on the material. Even after a high number of thermal cycles, while being irradiated by neutrons, no damage of the tested mock-ups was visually observed. Further testing and analysis will follow in the Forschungszentrum Juelich.

  13. Control of a reactor for conditioning of biogenic waste materials. Final report; Steuerung eines Reaktors zur Aufbereitung von Abfaellen mit biogenen Bestandteilen. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Bartha, B.; Brummack, J.; Kloeden, W.

    2002-10-01

    Mechanical-biological waste treatment implies biological drying and mechanical separation; it produces a pollutant-free fuel fraction with good calorific value and inorganic fractions available for recycling. Apart from some prototypes, mostly static aerobic reactors are used. For moist residues, dynamic reactors may be interesting as they permit coupling of thermal and mechanical processes, i.e. parallel biological drying and mechanical separation. A discontinuous rotary kiln reactor for biological drying of residues was developed in this project. (orig.) [German] Die mechanisch-biologische Aufbereitung von Restabfaellen (MBA) hat sich zu einem Systembaustein innerhalb der Restabfallentsorgung entwickelt. Das Hauptanwendungsziel besteht darin, durch die Kombination von biologischer Trocknung mit mechanischen Stofftrennverfahren eine heizwertreiche und schadstoffentfrachtete Fraktion mit Brennstoffeigenschaften, sowie verwertbare anorganische Fraktionen aus dem Restabfall zu gewinnen. Die Ausgangssituation zeigte, dass, abgesehen von einigen prototypischen Anlagen, in der MBA-Technologie hauptsaechlich statische, aerob arbeitende Reaktoren (Rottetunnel und Rotteboxen) fuer den biologischen Schritt eingesetzt werden. Fuer feuchte Restabfaelle erscheint es interessant, dynamische Reaktoren einzusetzen, die die Kopplung thermischer und mechanischer Prozesse erlauben, die also die biologische Trocknung und den mechanischen Aufschluss parallel zulassen. Im Rahmen des Projektes wurde die Steuerung eines diskontinuierlich betriebenen Drehrohrreaktors zur biologischen Trocknung von Restabfaellen entwickelt. (orig.)

  14. Nuclear reactor control column

    International Nuclear Information System (INIS)

    Bachovchin, D.M.

    1982-01-01

    The nuclear reactor control column comprises a column disposed within the nuclear reactor core having a variable cross-section hollow channel and containing balls whose vertical location is determined by the flow of the reactor coolant through the column. The control column is divided into three basic sections wherein each of the sections has a different cross-sectional area. The uppermost section of the control column has the greatest crosssectional area, the intermediate section of the control column has the smallest cross-sectional area, and the lowermost section of the control column has the intermediate cross-sectional area. In this manner, the area of the uppermost section can be established such that when the reactor coolant is flowing under normal conditions therethrough, the absorber balls will be lifted and suspended in a fluidized bed manner in the upper section. However, when the reactor coolant flow falls below a predetermined value, the absorber balls will fall through the intermediate section and into the lowermost section, thereby reducing the reactivity of the reactor core and shutting down the reactor

  15. OECD Halden reactor project

    International Nuclear Information System (INIS)

    1979-01-01

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

  16. Seismic behaviour of gas cooled reactor components

    International Nuclear Information System (INIS)

    1990-08-01

    On invitation of the French Government the Specialists' Meeting on the Seismic Behaviour of Gas-Cooled Reactor Components was held at Gif-sur-Yvette, 14-16 November 1989. This was the second Specialists' Meeting on the general subject of gas-cooled reactor seismic design. There were 27 participants from France, the Federal Republic of Germany, Israel, Japan, Spain, Switzerland, the United Kingdom, the Soviet Union, the United States, the CEC and IAEA took the opportunity to present and discuss a total of 16 papers reflecting the state of the art of gained experiences in the field of their seismic qualification approach, seismic analysis methods and of the capabilities of various facilities used to qualify components and verify analytical methods. Since the first meeting, the sophistication and expanded capabilities of both the seismic analytical methods and the test facilities are apparent. The two main methods for seismic analysis, the impedance method and the finite element method, have been computer-programmed in several countries with the capability of each of the codes dependent on the computer capability. The correlations between calculation and tests are dependent on input assumptions such as boundary conditions, soil parameters and various interactions between the soil, the buildings and the contained equipment. The ability to adjust these parameters and match experimental results with calculations was displayed in several of the papers. The expanded capability of some of the new test facilities was graphically displayed by the description of the SAMSON vibration test facility at Juelich, FRG, capable of dynamically testing specimens weighing up to 25 tonnes, and the TAMARIS facility at the CEA laboratories in Gif-sur-Yvette where the largest table is capable of testing specimens weighing up to 100 tonnes. The proceedings of this meeting contain all 16 presented papers. A separate abstract was prepared for each of these papers. Refs, figs and tabs

  17. Reactor pressure boundary materials

    International Nuclear Information System (INIS)

    Hong, Jun Hwa; Chi, S. H.; Lee, B. S.

    2002-04-01

    With a long-term operation of nuclear power plants, the component materials are degraded under severe reactor conditions such as neutron irradiation, high temperature, high pressure and corrosive environment. It is necessary to establish the reliable and practical technologies for improving and developing the component materials and for evaluating the mechanical properties. Especially, it is very important to investigate the technologies for reactor pressure boundary materials such as reactor vessel and pipings in accordance with their critical roles. Therefore, this study was focused on developing and advancing the microstructural/micro-mechanical evaluation technologies, and on evaluating the neutron irradiation characteristics and radiation effects analysis technology of the reactor pressure boundary materials, and also on establishing a basis of nuclear material property database

  18. Reactor container

    International Nuclear Information System (INIS)

    Naruse, Yoshihiro.

    1990-01-01

    The thickness of steel shell plates in a reactor container embedded in sand cussions is monitored to recognize the corrosion of the steel shell plates. That is, the reactor pressure vessel is contained in a reactor container shell and the sand cussions are disposed on the lower outside of the reactor container shell to elastically support the shell. A pit is disposed at a position opposing to the sand cussions for measuring the thickness of the reactor container shell plates. The pit is usually closed by a closing member. In the reactor container thus constituted, the closing member can be removed upon periodical inspection to measure the thickness of the shell plates. Accordingly, the corrosion of the steel shell plates can be recognized by the change of the plate thickness. (I.S.)

  19. Hybrid reactors

    International Nuclear Information System (INIS)

    Moir, R.W.

    1980-01-01

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

  20. Nuclear reactor

    International Nuclear Information System (INIS)

    Batheja, P.; Huber, R.; Rau, P.

    1985-01-01

    Particularly for nuclear reactors of small output, the reactor pressure vessel contains at least two heat exchangers, which have coolant flowing through them in a circuit through the reactor core. The circuit of at least one heat exchanger is controlled by a slide valve, so that even for low drive forces, particularly in natural circulation, the required even loading of the heat exchanger is possible. (orig./HP) [de

  1. Plateau's problem an invitation to varifold geometry

    CERN Document Server

    Frederick J Almgren, Jr

    2001-01-01

    There have been many wonderful developments in the theory of minimal surfaces and geometric measure theory in the past 25 to 30 years. Many of the researchers who have produced these excellent results were inspired by this little book--or by Fred Almgren himself. The book is indeed a delightful invitation to the world of variational geometry. A central topic is Plateau's Problem, which is concerned with surfaces that model the behavior of soap films. When trying to resolve the problem, however, one soon finds that smooth surfaces are insufficient: Varifolds are needed. With varifolds, one can obtain geometrically meaningful solutions without having to know in advance all their possible singularities. This new tool makes possible much exciting new analysis and many new results. Plateau's problem and varifolds live in the world of geometric measure theory, where differential geometry and measure theory combine to solve problems which have variational aspects. The author's hope in writing this book was to encour...

  2. INVITATION to CERN pensioners and personnel

    CERN Multimedia

    DG Unit

    2009-01-01

    The Swiss Foundation for Research into Ageing, AETAS, will be holding its research award ceremony at CERN on Wednesday 10 June 2009. You are cordially invited to attend the ceremony and, in particular, a lecture given by Professor Osman RATIB Professor of Medicine, Head of the Department of Medical Imaging and Medical Informatics and Head of the Division of Nuclear Medicine at the University Hospital of Geneva "Progress in medical imaging" WEDNESDAY 10 JUNE 2009 AT 5:00 P.M. CERN MAIN AUDITORIUM (Bldg.500) The lecture will be followed by a discussion moderated by Professor Jean-Pierre Michel, President of the AETAS Foundation. Refreshments will be offered by the Foundation

  3. INVITATION to CERN pensioners and personnel

    CERN Multimedia

    GS Department

    2009-01-01

    The Swiss Foundation for Research into Ageing, AETAS, will be holding its research award ceremony at CERN on Wednesday 10 June 2009. You are cordially invited to attend the ceremony and, in particular, a lecture given by Professor Osman RATIB Professor of Medicine, Head of the Department of Medical Imaging and Medical Informatics and Head of the Division of Nuclear Medicine at the University Hospital of Geneva "Progress in medical imaging" WEDNESDAY 10 JUNE 2009 AT 5.00 P.M. CERN MAIN AUDITORIUM, (Bldg.500) The lecture will be followed by a discussion moderated by Professor Jean-Pierre Michel, President of the AETAS Foundation. Drinks will be offered by the Foundation

  4. Nuclear reactor core assembly

    International Nuclear Information System (INIS)

    Baxi, C.B.

    1978-01-01

    The object of the present invention is to provide a fast reactor core assembly design for use with a fluid coolant such as liquid sodium or carbon monoxide incorporating a method of increasing the percentage of coolant flow though the blanket elements relative to the total coolant flow through the blanket and fuel elements during shutdown conditions without using moving parts. It is claimed that deterioration due to reactor radiation or temperature conditions is avoided and ready modification or replacement is possible. (U.K.)

  5. Comparison between temperature distributions of an annular fuel rod of circular cross-section and of a hemoglobin shaped cross-section rod for PWR reactors in steady state conditions

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Maria Vitória A. de; Alvim, Antônio Carlos Marques, E-mail: moliveira@con.ufrj.br, E-mail: alvim@nuclear.ufrj.br [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

    2017-07-01

    The objective of this work is to make a comparison between the temperature distributions of an annular fuel rod of circular cross-section and a hemoglobin shaped cross-section for PWR reactors in steady state conditions. The motivation for this article is due to the fact that the symmetric form of the red globules particles allows the O{sub 2} gases to penetrate the center of the cell homogeneously and quickly. The diffusion equation of gases in any environment is very similar to the heat diffusion equation: Diffusion - Fick's Law; Heat Flow - Fourier; where, the temperature (T) replaces the concentration (c). In previous works the comparison between the shape of solid fuel rods with circular section, and a with hemoglobin-shaped cross-section has proved that this new format optimizes the heat transfer, decreasing the thermal resistance between the center of the UO{sub 2} pellets and the clad. With this, a significant increase in the specific power of the reactor was made possible (more precisely a 23% increase). Currently, the advantages of annular fuel rods are being studied and recent works have shown that 12 x 12 arrays of annular fuel rods perform better, increasing the specific power of the reactor by at least 20% in relation to solid fuel rods, without affecting the safety of the reactor. Our proposal is analyzing the temperature distribution in annular fuel rods with cross sections with red blood cell shape and compare with the theoretical results of the annular fuel rods of circular cross section, initially in steady state. (author)

  6. Comparison between temperature distributions of an annular fuel rod of circular cross-section and of a hemoglobin shaped cross-section rod for PWR reactors in steady state conditions

    International Nuclear Information System (INIS)

    Oliveira, Maria Vitória A. de; Alvim, Antônio Carlos Marques

    2017-01-01

    The objective of this work is to make a comparison between the temperature distributions of an annular fuel rod of circular cross-section and a hemoglobin shaped cross-section for PWR reactors in steady state conditions. The motivation for this article is due to the fact that the symmetric form of the red globules particles allows the O 2 gases to penetrate the center of the cell homogeneously and quickly. The diffusion equation of gases in any environment is very similar to the heat diffusion equation: Diffusion - Fick's Law; Heat Flow - Fourier; where, the temperature (T) replaces the concentration (c). In previous works the comparison between the shape of solid fuel rods with circular section, and a with hemoglobin-shaped cross-section has proved that this new format optimizes the heat transfer, decreasing the thermal resistance between the center of the UO 2 pellets and the clad. With this, a significant increase in the specific power of the reactor was made possible (more precisely a 23% increase). Currently, the advantages of annular fuel rods are being studied and recent works have shown that 12 x 12 arrays of annular fuel rods perform better, increasing the specific power of the reactor by at least 20% in relation to solid fuel rods, without affecting the safety of the reactor. Our proposal is analyzing the temperature distribution in annular fuel rods with cross sections with red blood cell shape and compare with the theoretical results of the annular fuel rods of circular cross section, initially in steady state. (author)

  7. Heterogeneous reactors

    International Nuclear Information System (INIS)

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

    1979-08-01

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

  8. Guidelines for calculating radiation doses to the public from a release of airborne radioactive material under hypothetical accident conditions in nuclear reactors

    International Nuclear Information System (INIS)

    1989-09-01

    This Standard provides guidelines and a methodology for calculating effective doses and thyroid doses to people (either individually or collectively) in the path of airborne radioactive material released from a nuclear facility following a hypothetical accident. The specific radionuclides considered in the Standard are those associated with substances having the greatest potential for becoming airborne in reactor accidents (eg, tritium (HTO), noble gases and their daughters (Kr-Rb, Xe-Cs), and radioiodines (I)); and certain radioactive particulates (eg, Cs, Ru, Sr, Te) that may become airborne under exceptional circumstances

  9. Young Women's Experiences of Resisting Invitations to Use Illicit Drugs

    Science.gov (United States)

    Koehn, Corinne V.; O'Neill, Linda K.

    2011-01-01

    Ten young women were interviewed regarding their experiences of resisting invitations to use illicit drugs. Hermeneutic phenomenology was used to gather and analyze information. One key theme was the motivations that inspired women to refuse drug offers. Young women resisted drug invitations because of their desires to be authentic, protect their…

  10. 76 FR 63818 - Special Conditions: Boeing Model 767-400ER Series Airplanes; Seats With Inflatable Lapbelts

    Science.gov (United States)

    2011-10-14

    ... inflatable lapbelt must provide a consistent approach to energy absorption throughout that range of occupants... special conditions effective upon issuance. Comments Invited We invite interested people to take part in...

  11.   Personal invitations for population-based breast cancer screening

    DEFF Research Database (Denmark)

    Saalasti-Koskinen, Ulla; Mäkelä, Marjukka; Saarenmaa, Irma

    2010-01-01

    participation free of charge and the benefits of detecting breast cancer early. Harm associated with screening was seldom mentioned; no unit mentioned the possibility of false-negative results or overtreatment. CONCLUSION: The screening units provided very variable information, which often was biased toward......RATIONALE AND OBJECTIVES: Women who are invited for breast cancer screening should get enough information about the benefits and harms of screening to make an informed decision on participation. Personal invitations are an important source of information, because all invited women receive them....... The objective of this study was to evaluate the information breast cancer screening units send to women invited for screening in Finland. MATERIALS AND METHODS: A questionnaire was sent to all breast cancer screening units in Finland in 2005 and 2008, and the information (eg, invitations, results letters...

  12. Execution of programme of post-service study of the condition of nuclear icebreaker Lenin reactor 1 pressure vessel metal and perspectives of application of results to increase service life of nuclear icebreakers reactor vessels

    International Nuclear Information System (INIS)

    Platonov, P.Ya.; Shtrombakh, Ya.I.; Amaev, A.D.; Krasikov, E.A.; Korolev, Yu.N.; Zabusov, O.O.; Glushakov, G.M.

    2001-01-01

    With the aim of determining the irradiation-induced embrittlement of a base metal and a weld metal in a pressure vessel of the nuclear icebreaker Lenin after 18 years operation the specimens cut out of a vessel wall are used to study the chemical composition and to carry out impact tests. From the test results the temperature dependences of fracture energy are built which define the irradiation embrittlement of a low alloy steel. It is noted that the annealing at 475 deg C for 100 h results in complete restoration of impact strength. Based on the results obtained the following conclusions are formulated: a reactor vessel base metal has high resistance to brittle fracture and high radiation resistance; a weld metal possesses rather high radiation resistance but unsatisfactory ductile-brittle transition temperature (∼ 63 deg C); for cladded vessels there is a potential reserve in the form of enhanced radiation resistance of an undercladding layer; in the final stage of operation the coolant temperature is recommended to be kept at the highest possible level [ru

  13. Reactor safety

    International Nuclear Information System (INIS)

    Butz, H.P.; Heuser, F.W.; May, H.

    1985-01-01

    The paper comprises an introduction into nuclear physics bases, the safety concept generally speaking, safety devices of pwr type reactors, accident analysis, external influences, probabilistic safety assessment and risk studies. It further describes operational experience, licensing procedures under the Atomic Energy Law, research in reactor safety and the nuclear fuel cycle. (DG) [de

  14. Nuclear reactor

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  15. Reactor container

    International Nuclear Information System (INIS)

    Kato, Masami; Nishio, Masahide.

    1987-01-01

    Purpose: To prevent the rupture of the dry well even when the melted reactor core drops into a reactor pedestal cavity. Constitution: In a reactor container in which a dry well disposed above the reactor pedestal cavity for containing the reactor pressure vessel and a torus type suppression chamber for containing pressure suppression water are connected with each other, the pedestal cavity and the suppression chamber are disposed such that the flow level of the pedestal cavity is lower than the level of the pressure suppression water. Further, a pressure suppression water introduction pipeway for introducing the pressure suppression water into the reactor pedestal cavity is disposed by way of an ON-OFF valve. In case if the melted reactor core should fall into the pedestal cavity, the ON-OFF valve for the pressure suppression water introduction pipeway is opened to introduce the pressure suppression water in the suppression chamber into the pedestal cavity to cool the melted reactor core. (Ikeda, J.)

  16. RA Reactor

    International Nuclear Information System (INIS)

    1989-01-01

    This chapter includes the following: General description of the RA reactor, organization of work, responsibilities of leadership and operators team, regulations concerning operation and behaviour in the reactor building, regulations for performing experiments, regulations and instructions for inserting samples into experimental channels [sr

  17. Reactor physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, H.

    1998-01-01

    Progress in research on reactor physics in 1997 at the Belgian Nuclear Research Centre SCK/CEN is described. Activities in the following four domains are discussed: core physics, ex-core neutron transport, experiments in Materials Testing Reactors, international benchmarks

  18. Activity of the RA Reactor Physics group in 1980 - Definition of the Operation conditions for future safe and economical RA reactor operation with 80% enriched fuel; Prilog Ia - Rad sluzbe za fiziku reaktora RA u 1980. godini - Definisanje pogonskih uslova za dalji siguran i ekonomican rad reaktora RA sa 80% obogacenim gorivom

    Energy Technology Data Exchange (ETDEWEB)

    Martinc, R [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1980-12-15

    During 1980. the RA reactor was not in operation. That is why this period was devoted to definition of operating conditions for further reactor operation with 80% enriched fuel. The fuel elements which were in the core at the moment of shutdown in March 1979will not be used again (388 80% enriched fuel elements, and 511 2% enriched fuel elements). The reactor will be operated only with 80% enriched fuel, staring with initiat core configuration with 440 elements on the borders gradually changing to equi;librium core with 720 fuel elements. The analyses were concerned with safety issues of future operation. [Serbo-Croat] Reaktor RA tokom 1980 godine nije radio. Zbog toga je ovaj period iskoriscen za intenzivan rad na definisanju pogonskih uslova za dalji rad reaktora sa 80% obogacenim gorivom. Gorivo sa kojim je reaktor prestao da radi marta 1979. godine (388 gorivnih elemenata 80% obogacenoh i 511 elementa 2% obogaceno) nece biti vraceno u jezgro, vec ce reaktor raditi samo sa 80% obogacenim gorivom, pocev od konfiguracije sa 440 elemenata na periferiji do ravnotezne konfiguracije sa 720 elemenata. Sve nalize su se bavile sigurnosnim aspektima rada reaktora sa visokoobogacenim gorivom u jezgru.

  19. Reactor core

    International Nuclear Information System (INIS)

    Azekura, Kazuo; Kurihara, Kunitoshi.

    1992-01-01

    In a BWR type reactor, a great number of pipes (spectral shift pipes) are disposed in the reactor core. Moderators having a small moderating cross section (heavy water) are circulated in the spectral shift pipes to suppress the excess reactivity while increasing the conversion ratio at an initial stage of the operation cycle. After the intermediate stage of the operation cycle in which the reactor core reactivity is lowered, reactivity is increased by circulating moderators having a great moderating cross section (light water) to extend the taken up burnup degree. Further, neutron absorbers such as boron are mixed to the moderator in the spectral shift pipe to control the concentration thereof. With such a constitution, control rods and driving mechanisms are no more necessary, to simplify the structure of the reactor core. This can increase the fuel conversion ratio and control great excess reactivity. Accordingly, a nuclear reactor core of high conversion and high burnup degree can be attained. (I.N.)

  20. Reactor container

    International Nuclear Information System (INIS)

    Fukazawa, Masanori.

    1991-01-01

    A system for controlling combustible gases, it has been constituted at present such that the combustible gases are controlled by exhausting them to the wet well of a reactor container. In this system, however, there has been a problem, in a reactor container having plenums in addition to the wet well and the dry well, that the combustible gases in such plenums can not be controlled. In view of the above, in the present invention, suction ports or exhaust ports of the combustible gas control system are disposed to the wet well, the dry well and the plenums to control the combustible gases in the reactor container. Since this can control the combustible gases in the entire reactor container, the integrity of the reactor container can be ensured. (T.M.)