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Sample records for braunschweig experimental reactor

  1. Irrigation of treated wastewater in Braunschweig, Germany

    DEFF Research Database (Denmark)

    Ternes, T.A.; Bonerz, M.; Herrmann, N.

    2007-01-01

    In this study the fate of pharmaceuticals and personal care products which are irrigated on arable land with treated municipal waste-water was investigated. In Braunschweig, Germany, wastewater has been irrigated continuously for more than 45 years. In the winter time only the effluent...... of the sewage treatment plant (STP) of Braunschweig is used for irrigation, while during summer digested sludge is mixed with the effluent. In the present case study six wells and four lysimeters located in one of the irrigated agricultural fields were monitored with regard to the occurrence of 52...... pharmaceuticals and two personal care products (PPCPs; e.g. betablockers, antibiotics, antiphlogistics, carbamazepine, musk fragrances, iodinated contrast media (ICM) and estrogens). No differences in PPCP pollution of the groundwater were found due to irrigation of STP effluents with and without addition...

  2. Experimental Breeder Reactor I Preservation Plan

    Energy Technology Data Exchange (ETDEWEB)

    Julie Braun

    2006-10-01

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

  3. Seclazone Reactor Modeling And Experimental Validation

    Energy Technology Data Exchange (ETDEWEB)

    Osinga, T. [ETH-Zuerich (Switzerland); Olalde, G. [CNRS Odeillo (France); Steinfeld, A. [PSI and ETHZ (Switzerland)

    2005-03-01

    A numerical model is formulated for the SOLZINC solar chemical reactor for the production of Zn by carbothermal reduction of ZnO. The model involves solving, by the finite-volume technique, a 1D unsteady state energy equation that couples heat transfer to the chemical kinetics for a shrinking packed bed exposed to thermal radiation. Validation is accomplished by comparison with experimentally measured temperature profiles and Zn production rates as a function of time, obtained for a 5-kW solar reactor tested at PSI's solar furnace. (author)

  4. Plasma heating systems planned for the Argonne experimental power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bertoncini, P.; Brooks, J.; Fasolo, J.; Mills, F.; Moretti, A.; Norem, J.

    1976-01-01

    A scoping study and conceptual design of a tokamak experimental power reactor (TEPR) have been completed. The design objectives of the TEPR are to operate for ten years at or near electrical power breakeven conditions with a duty factor of greater than or equal to 50 percent and to demonstrate the feasibility of tokamak fusion power reactor techniques. These objectives can be met by a design which has a major radius of 6.25 m and a plasma radius of 2.1 m. Parameters for this reactor are listed, and a diagram is given. This paper will describe TEPR plasma heating systems. Neutral beam heating and rf heating are described.

  5. Upgrading program of the experimental fast reactor Joyo

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-07-01

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

  6. Investigation of mixing chamber for experimental FGD reactor

    Directory of Open Access Journals (Sweden)

    Novosád Jan

    2017-01-01

    Full Text Available This article deals with numerical investigation of flow and mixing of air and sulphur dioxide SO2 in designated mixing chamber. The mixing chamber is a part of experimental laboratory reactor designed for simulating the flue gas desulfurization (FGD process. Aim of this work is the numerical investigation of effect of different mixing chamber geometries to mixture composition, especially to mass fraction of sulphur dioxide. Using of similar concentration of sulphur dioxide in the experimental reactor as in the real process is necessary to be able to make additional research. Conclusion describes the effect of different geometries of mixing chamber to mixing. The aim of this work is to develop perfectly works mixing chamber, which will be manufactured and then implemented into experimental FGD reactor. The results will be validated by experiment after the mixing chamber will be manufactured.

  7. Experimental Criticality Benchmarks for SNAP 10A/2 Reactor Cores

    Energy Technology Data Exchange (ETDEWEB)

    Krass, A.W.

    2005-12-19

    This report describes computational benchmark models for nuclear criticality derived from descriptions of the Systems for Nuclear Auxiliary Power (SNAP) Critical Assembly (SCA)-4B experimental criticality program conducted by Atomics International during the early 1960's. The selected experimental configurations consist of fueled SNAP 10A/2-type reactor cores subject to varied conditions of water immersion and reflection under experimental control to measure neutron multiplication. SNAP 10A/2-type reactor cores are compact volumes fueled and moderated with the hydride of highly enriched uranium-zirconium alloy. Specifications for the materials and geometry needed to describe a given experimental configuration for a model using MCNP5 are provided. The material and geometry specifications are adequate to permit user development of input for alternative nuclear safety codes, such as KENO. A total of 73 distinct experimental configurations are described.

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

    Energy Technology Data Exchange (ETDEWEB)

    Slugeň, Vladimír, E-mail: vladimir.slugen@stuba.sk; Pecko, Stanislav; Sojak, Stanislav

    2016-01-15

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

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

    Science.gov (United States)

    Slugeň, Vladimír; Pecko, Stanislav; Sojak, Stanislav

    2016-01-01

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

  10. Incipient Transient Detection in Reactor Systems: Experimental and Theoretical Investigation

    Energy Technology Data Exchange (ETDEWEB)

    Lefteri H. Tsoukalas; S.T. Revankar; X Wang; R. Sattuluri

    2005-09-27

    The main goal of this research was to develop a method for detecting reactor system transients at the earliest possible time through a comprehensive experimental, testing and benchmarking program. This approach holds strong promise for developing new diagnostic technologies that are non-intrusive, generic and highly portable across different systems. It will help in the design of new generation nuclear power reactors, which utilize passive safety systems with a reliable and non-intrusive multiphase flow diagnostic system to monitor the function of the passive safety systems. The main objective of this research was to develop an improved fuzzy logic based detection method based on a comprehensive experimental testing program to detect reactor transients at the earliest possible time, practically at their birth moment. A fuzzy logic and neural network based transient identification methodology and implemented in a computer code called PROTREN was considered in this research and was compared with SPRT (Sequentially Probability Ratio Testing) decision and Bayesian inference. The project involved experiment, theoretical modeling and a thermal-hydraulic code assessment. It involved graduate and undergraduate students participation providing them with exposure and training in advanced reactor concepts and safety systems. In this final report, main tasks performed during the project period are summarized and the selected results are presented. Detailed descriptions for the tasks and the results are presented in previous yearly reports (Revankar et al 2003 and Revankar et al 2004).

  11. 9 CFR 319.182 - Braunschweiger and liver sausage or liverwurst.

    Science.gov (United States)

    2010-01-01

    ... 9 Animals and Animal Products 2 2010-01-01 2010-01-01 false Braunschweiger and liver sausage or liverwurst. 319.182 Section 319.182 Animals and Animal Products FOOD SAFETY AND INSPECTION SERVICE... AND VOLUNTARY INSPECTION AND CERTIFICATION DEFINITIONS AND STANDARDS OF IDENTITY OR COMPOSITION Cooked...

  12. Experimental Investigation of Effect on Hydrate Formation in Spray Reactor

    Directory of Open Access Journals (Sweden)

    Jianzhong Zhao

    2015-01-01

    Full Text Available The effects of reaction condition on hydrate formation were conducted in spray reactor. The temperature, pressure, and gas volume of reaction on hydrate formation were measured in pure water and SDS solutions at different temperature and pressure with a high-pressure experimental rig for hydrate formation. The experimental data and result reveal that additives could improve the hydrate formation rate and gas storage capacity. Temperature and pressure can restrict the hydrate formation. Lower temperature and higher pressure can promote hydrate formation, but they can increase production cost. So these factors should be considered synthetically. The investigation will promote the advance of gas storage technology in hydrates.

  13. Sociodemographic determinants of spatial disparities in early childhood caries: An ecological analysis in Braunschweig, Germany.

    Science.gov (United States)

    Meyer, Frederic; Karch, André; Schlinkmann, Kristin Maria; Dreesman, Johannes; Horn, Johannes; Rübsamen, Nicole; Sudradjat, Henny; Schubert, Rainer; Mikolajczyk, Rafael

    2017-10-01

    To identify spatial disparities in dental caries experience (measured by dmft (decayed missing filled teeth) index) of children in the city of Braunschweig and to evaluate whether these disparities can be explained by sociodemographic characteristics. We examined the dental health of children aged 3-6 years visiting a daycare centre (DCC) in the metropolitan area of Braunschweig between 2009 and 2014 by combining data on dental health from the annual visits of the local health service with aggregated data on sociodemographic factors for Braunschweig's city districts. We assessed longitudinal patterns of change in average dmft index at district level from 2009 to 2014 using a finite mixture model. We analysed spatial autocorrelation of the district's average dmft indices by Moran's I to identify spatial clusters. With a spatial lag model, we evaluated whether sociodemographic risk factors (data from 2012) were associated with high dmft scores (data from 2014) and whether spatial disparities remained after adjusting for these sociodemographic characteristics. The average dmft index decreased slightly (β=-0.048; Psociodemographic characteristics. The findings support the need of targeted interventions and prevention measures in regions with less favourable sociodemographic characteristics. © 2017 John Wiley & Sons A/S. Published by John Wiley & Sons Ltd.

  14. Methanol synthesis in a countercurrent gas-solid-solid trickle flow reactor. An experimental study

    NARCIS (Netherlands)

    Kuczynski, M.; Oyevaar, M.H.; Pieters, R.T.; Westerterp, K.R.

    1987-01-01

    The synthesis of methanol from CO and H2 was executed in a gas-solid-solid trickle flow reactor. The reactor consisted of three tubular reactor sections with cooling sections in between. The catalyst was Cu on alumina, the adsorbent was a silica-alumina powder and the experimental range 498–523 K,

  15. [Hygiene in Urological Surgeries - Results of the Health Authority's Visit to all Urological Surgeries in Braunschweig].

    Science.gov (United States)

    Buhr-Riehm, B; Lenz, T

    2015-07-01

    Following a patient complaint, the Health Department carried out a hygiene inspection of a urological practice in Braunschweig in February 2013. The topic of the complaint was that a patient assumed having acquired a resistant pathogen in the practice. In the subsequent visit, significant hygiene defects were found, particularly with regard to the processing of medical devices. This led to a decision to commit all urological practices in Braunschweig to hygiene inspections as part of a priority project. In retrospect, the hygiene surveys were justified. Deficiencies included inadequate preparation of medical products, procedures in practice inconsistent with hygiene plans, poor knowledge of hygiene procedures among assistant staff and doctors, lack of expertise of assistant staff and lack of hygiene risk awareness by doctors. Positive experiences were: open communication in a good atmosphere with the Practice managers, willingness to change, good cooperation between the Health Authority and the Labor Inspectorate and Physicians' Association. The claimed deficits were corrected by spring 2014 by the practice operators. The consulting expertise of the health authorities was made use of continuously. © Georg Thieme Verlag KG Stuttgart · New York.

  16. Tokamak experimental power reactor conceptual design. Volume I

    Energy Technology Data Exchange (ETDEWEB)

    1976-08-01

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

  17. Integral Reactor Containment Condensation Model and Experimental Validation

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Qiao [Oregon State Univ., Corvallis, OR (United States); Corradini, Michael [Univ. of Wisconsin, Madison, WI (United States)

    2016-05-02

    This NEUP funded project, NEUP 12-3630, is for experimental, numerical and analytical studies on high-pressure steam condensation phenomena in a steel containment vessel connected to a water cooling tank, carried out at Oregon State University (OrSU) and the University of Wisconsin at Madison (UW-Madison). In the three years of investigation duration, following the original proposal, the planned tasks have been completed: (1) Performed a scaling study for the full pressure test facility applicable to the reference design for the condensation heat transfer process during design basis accidents (DBAs), modified the existing test facility to route the steady-state secondary steam flow into the high pressure containment for controllable condensation tests, and extended the operations at negative gage pressure conditions (OrSU). (2) Conducted a series of DBA and quasi-steady experiments using the full pressure test facility to provide a reliable high pressure condensation database (OrSU). (3) Analyzed experimental data and evaluated condensation model for the experimental conditions, and predicted the prototypic containment performance under accidental conditions (UW-Madison). A film flow model was developed for the scaling analysis, and the results suggest that the 1/3 scaled test facility covers large portion of laminar film flow, leading to a lower average heat transfer coefficient comparing to the prototypic value. Although it is conservative in reactor safety analysis, the significant reduction of heat transfer coefficient (50%) could under estimate the prototypic condensation heat transfer rate, resulting in inaccurate prediction of the decay heat removal capability. Further investigation is thus needed to quantify the scaling distortion for safety analysis code validation. Experimental investigations were performed in the existing MASLWR test facility at OrST with minor modifications. A total of 13 containment condensation tests were conducted for pressure

  18. Introduction of Nuclear Instrumentations and Radiation Measurements in Experimental Fast Reactor 「JOYO」

    OpenAIRE

    大戸 敏弘; 鈴木 惣十

    1992-01-01

    This report introduces the nuclear instrumentation system and major R&D (research and development) activities using radiation measurement techniques in Experimental Fast Reactor "JOYO". In the introduction of the nuclear instrumentation system, following items are described; (1)system function (2)roles as a reactor plant equipment (3)specifications and charactelistics of neutron detectors, (4)construction and layout of the system. For reactor dosimetry at various irradiation tests and surveil...

  19. Review of accident analyses of RB experimental reactor

    Directory of Open Access Journals (Sweden)

    Pešić Milan P.

    2003-01-01

    Full Text Available The RB reactor is a uranium fuel heavy water moderated critical assembly that has been put and kept in operation by the VTNCA Institute of Nuclear Sciences, Belgrade, Serbia and Montenegro, since April 1958. The first complete Safety Analysis Report of the RB reactor was prepared in 1961/62 yet, the first accident analysis had been made in late 1958 with the aim to examine a power transition and the total equivalent doses received by the staff during the reactivity accident that occurred on October 15, 1958. Since 1960, the RB reactor has been modified a few times. Beside the initial natural uranium metal fuel rods, new types of fuel (TVR-S types of Russian origin consisting of 2% enriched uranium metal and 80% enriched UO2 dispersed in aluminum matrix, have been available since 1962 and 1976 respectively. Modifications of the control and safety systems of the reactor were made occasionally. Special reactor cores were designed and constructed using all three types of fuel elements as well as the coupled fast-thermal ones. The Nuclear Safety Committee of the VINĆA Institute, an independent regulatory body, approved for usage all these modifications of the RB reactor on the basis of the Preliminary Safety Analysis Reports, which, beside proposed technical modifications and new regulation rules, included safety analyses of various possible accidents. A special attention was given (and a new safety methodology was proposed to thorough analyses of the design-based accidents related to the coupled fast-thermal cores that included central zones of the reactor filled by the fuel elements without any moderator. In this paper, an overview of some accidents, methodologies and computation tools used for the accident analyses of the RB reactor is given.

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

    NARCIS (Netherlands)

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

    2005-01-01

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

  1. Experimental and simulated dosimetry of the university of Utah TRIGA reactor

    Science.gov (United States)

    Marble, Benjamin James

    Simulated neutron and gamma transport enable the gamma dose to be estimated at the surface of the University of Utah TRIGA Reactor UUTR pool. These results are benchmarked against experimental results for model verification. This model is useful for future licensing and possible reactor power upgrades. MCNP5 was utilized for the UUTR simulation and comparison with thermoluminescent detectors TLDs.

  2. An evaluation of alternative reactor vessel cutting technologies for the experimental boiling water reactor at Argonne National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Boing, L.E.; Henley, D.R. (Argonne National Lab., IL (USA)); Manion, W.J.; Gordon, J.W. (Nuclear Energy Services, Inc., Danbury, CT (USA))

    1989-12-01

    Metal cutting techniques that can be used to segment the reactor pressure vessel of the Experimental Boiling Water Reactor (EBWR) at Argonne National Laboratory (ANL) have been evaluated by Nuclear Energy Services. Twelve cutting technologies are described in terms of their ability to perform the required task, their performance characteristics, environmental and radiological impacts, and cost and schedule considerations. Specific recommendations regarding which technology should ultimately be used by ANL are included. The selection of a cutting method was the responsibility of the decommissioning staff at ANL, who included a relative weighting of the parameters described in this document in their evaluation process. 73 refs., 26 figs., 69 tabs.

  3. COMPUTATIONAL AND EXPERIMENTAL MODELING OF SLURRY BUBBLE COLUMN REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Paul C.K. Lam; Isaac K. Gamwo; Dimitri Gidaspow

    2002-05-01

    The objective of this study was to develop a predictive experimentally verified computational fluid dynamics (CFD) model for gas-liquid-solid flow. A three dimensional transient computer code for the coupled Navier-Stokes equations for each phase was developed and is appended in this report. The principal input into the model is the viscosity of the particulate phase which was determined from a measurement of the random kinetic energy of the 800 micron glass beads and a Brookfield viscometer. The details are presented in the attached paper titled ''CFD Simulation of Flow and Turbulence in a Slurry Bubble Column''. This phase of the work is in press in a referred journal (AIChE Journal, 2002) and was presented at the Fourth International Conference on Multiphase Flow (ICMF 2001) in New Orleans, May 27-June 1, 2001 (Paper No. 909). The computed time averaged particle velocities and concentrations agree with Particle Image Velocimetry (PIV) measurements of velocities and concentrations, obtained using a combination of gamma-ray and X-ray densitometers, in a slurry bubble column, operated in the bubbly-coalesced fluidization regime with continuous flow of water. Both the experiment and the simulation show a down-flow of particles in the center of the column and up-flow near the walls and nearly uniform particle concentration. Normal and shear Reynolds stresses were constructed from the computed instantaneous particle velocities. The PIV measurement and the simulation produced instantaneous particle velocities. The PIV measurement and the simulation produced similar nearly flat horizontal profiles of turbulent kinetic energy of particles. To better understand turbulence we studied fluidization in a liquid-solid bed. This work was also presented at the Fourth International Conference on Multiphase Flow (ICMF 2001, Paper No. 910). To understand turbulence in risers, measurements were done in the IIT riser with 530 micron glass beads using a PIV

  4. Experimental study of radiation dose rate at different strategic points of the BAEC TRIGA Research Reactor.

    Science.gov (United States)

    Ajijul Hoq, M; Malek Soner, M A; Salam, M A; Haque, M M; Khanom, Salma; Fahad, S M

    2017-12-01

    The 3MW TRIGA Mark-II Research Reactor of Bangladesh Atomic Energy Commission (BAEC) has been under operation for about thirty years since its commissioning at 1986. In accordance with the demand of fundamental nuclear research works, the reactor has to operate at different power levels by utilizing a number of experimental facilities. Regarding the enquiry for safety of reactor operating personnel and radiation workers, it is necessary to know the radiation level at different strategic points of the reactor where they are often worked. In the present study, neutron, beta and gamma radiation dose rate at different strategic points of the reactor facility with reactor power level of 2.4MW was measured to estimate the rising level of radiation due to its operational activities. From the obtained results high radiation dose is observed at the measurement position of the piercing beam port which is caused by neutron leakage and accordingly, dose rate at the stated position with different reactor power levels was measured. This study also deals with the gamma dose rate measurements at a fixed position of the reactor pool top surface for different reactor power levels under both Natural Convection Cooling Mode (NCCM) and Forced Convection Cooling Mode (FCCM). Results show that, radiation dose rate is higher for NCCM in compared with FCCM and increasing with the increase of reactor power. Thus, concerning the radiological safety issues for working personnel and the general public, the radiation dose level monitoring and the experimental analysis performed within this paper is so much effective and the result of this work can be utilized for base line data and code verification of the nuclear reactor. Copyright © 2017 Elsevier Ltd. All rights reserved.

  5. The Jules Horowitz reactor (JHR), a European material testing reactor (MTR), with extended experimental capabilities

    Energy Technology Data Exchange (ETDEWEB)

    Ballagny, A.; Bergamaschi, Y.; Bouilloux, Y.; Bravo, X.; Guigon, B.; Rommens, M.; Tremodeux, P. [CEA Cadarache, Dir. de l' Energie Nucleaire DEN, 13 - Saint-Paul-lez-Durance (France)]|[CEA Saclay Dir. de l' Energie Nucleaire DEN, 91 - Gif sur Yvette (France)

    2003-07-01

    The Jules Horowitz Reactor (JHR) is the European MTR (Material Testing Reactor) designed to provide, after 2010, the necessary knowledge for keeping the existing power plants in operation and to design innovative reactors types with new objectives such as: minimizing the radioactive waste production, taking into account additional safety requirements, preventing risks of nuclear proliferation... To achieve such an ambitious objective. The JHR is designed with a high flexibility in order to satisfy the current demand from European industry, research and to be able to accommodate future requirements. The JHR will offer a wide range of performances and services in gathering, in a single site at Cadarache, all the necessary functionalities and facilities for an effective production of results: e.g. fuel fabrication laboratories, preparation of the instrumented devices, interpretation of the experiments, modelling. The JHR must rely on a top level scientific environment based on experts teams from CEA and EC and local universities. With a thermal flux of 7,4.10{sup 14} ncm{sup -2} s{sup -1} and a fast flux of 6,4.10{sup 14} ncm{sup -2}s{sup -1}, it is possible to carry out irradiation experiments on materials and fuels whatever the reactor type considered. It will also be possible to carry out locally, fast neutron irradiation to achieve damage effect up to 25 dpa/year. (dpa = displacement per atom.) The study of the fuels behavior under accidental conditions, from analytical experiments, on a limited amount of irradiated fuel, is a major objective of the project. These oriented safety tests are possible by taking into account specific requirements in the design of the facility such as the tightness level of the containment building, the addition of an alpha hot cell and a laboratory for on line fission products measurement. (authors)

  6. ESTIMATION OF ROUTINE DISCHARGE OF RADIONUCLIDES ON POWER REACTOR EXPERIMENTAL RDE

    OpenAIRE

    Udiyani, P.M; Sri Kuntjoro

    2017-01-01

    Experimental power reactor (RDE) which is planned to be constructed by BATAN is a kind of High Temperature Gas Cooled Reactor (HTGR) with 10 MWth power. HTGR is a helium gas-cooled reactor with TRISO-coated fuel that is able to confine fission products remained in the core. Although the fission products released into the environment are very small, in order to comply the regulations the study about environmental radiation on normal or routine operation condition need to be performed. Estimati...

  7. Experimental Investigation of Biogas Reforming in Gliding Arc Plasma Reactors

    Directory of Open Access Journals (Sweden)

    P. Thanompongchart

    2014-01-01

    Full Text Available Biogas is an important renewable energy source. Its utilization is restricted to vicinity of farm areas, unless pipeline networks or compression facilities are established. Alternatively, biogas may be upgraded into synthetic gas via reforming reaction. In this work, plasma assisted reforming of biogas was investigated. A laboratory gliding arc plasma setup was developed. Effects of CH4/CO2 ratio (1, 2.33, 9, feed flow rate (16.67–83.33 cm3/s, power input (100–600 W, number of reactor, and air addition (0–60% v/v on process performances in terms of yield, selectivity, conversion, and energy consumption were investigated. High power inputs and long reaction time from low flow rates, or use of two cascade reactors were found to promote dry reforming of biogas. High H2 and CO yields can be obtained at low energy consumption. Presence of air enabled partial oxidation reforming that produced higher CH4 conversion, compared to purely dry CO2 reforming process.

  8. Detection efficiency of the neutron detector BELEN-48 measured at the PTB Braunschweig

    Energy Technology Data Exchange (ETDEWEB)

    Marta, Michele [GSI Helmholtzzentrum fuer Schwerionenforschung GmbH, Darmstadt (Germany); II. Physikalisches Institut, Justus-Liebig Universitaet Giessen (Germany); Agramunt, Jorge; Tain, Jose Luis [IFIC-CSIC University of Valencia, Valencia (Spain); Caballero-Folch, Roger; Cortes, Guillem; Riego, Albert [INTE-DFEN, Universitat Politecnica de Catalunya, Barcelona (Spain); Dillmann, Iris [GSI Helmholtzzentrum fuer Schwerionenforschung GmbH, Darmstadt (Germany); II. Physikalisches Institut, Justus-Liebig Universitaet Giessen (Germany); TRIUMF, Vancouver (Canada); Erhard, Martin; Giesen, Ulrich; Nolte, Ralf; Roettger, Stefan [Physikalisch-Technische Bundesanstalt (PTB), Braunschweig (Germany); Fraile, Luis M. [Universidad Complutense de Madrid (Spain)

    2014-07-01

    The BEta-deLayEd Neutron detector BELEN-48 is a highly efficient detector of β-delayed neutrons, for nuclear structure, nuclear astrophysics and reactor studies. It consists of 48 {sup 3}He proportional counters arranged in a polyethylene matrix in a way that the detection efficiency remains constant for neutron energies from thermal up to a few MeV. In order to validate MCNPX simulations, the detection efficiency has been calibrated with well-known (p,n) and (α,n) reactions on {sup 7}Li, {sup 13}C and {sup 51}V producing neutrons with energies between 0.1 and 5 MeV. The experiment has been performed at the neutron metrology facility of PTB, which allowed the measurement of yields and angular distributions with a calibrated monitor. The new results indicate anisotropies, which are not reported in literature and have been taken into account to obtain the experimental efficiencies for BELEN.

  9. Experiments in the experimental fast reactor VENUS-F: The FREYA project; Experimentos en el reactor rapido experimental VENUS-F: El proyecto FREYA

    Energy Technology Data Exchange (ETDEWEB)

    Villamarin, D.; Becares, V.; Cano, D.; Gonzalez, E.

    2011-07-01

    Due to the high flexibility of operation of the reactor VENUS-E, FREYA project has two main objectives. The first is the end of the study monitoring techniques reactivity and serve as validation of simulation codes. The second objective is to provide experimental support for design and licensing MYRRHA / FASTEE and TRF in collaboration with CDTy LEADER projects of the 7th Framework Programme of the EU.

  10. Experimental and computational investigation of flow of pebbles in a pebble bed nuclear reactor

    Science.gov (United States)

    Khane, Vaibhav B.

    The Pebble Bed Reactor (PBR) is a 4th generation nuclear reactor which is conceptually similar to moving bed reactors used in the chemical and petrochemical industries. In a PBR core, nuclear fuel in the form of pebbles moves slowly under the influence of gravity. Due to the dynamic nature of the core, a thorough understanding about slow and dense granular flow of pebbles is required from both a reactor safety and performance evaluation point of view. In this dissertation, a new integrated experimental and computational study of granular flow in a PBR has been performed. Continuous pebble re-circulation experimental set-up, mimicking flow of pebbles in a PBR, is designed and developed. Experimental investigation of the flow of pebbles in a mimicked test reactor was carried out for the first time using non-invasive radioactive particle tracking (RPT) and residence time distribution (RTD) techniques to measure the pebble trajectory, velocity, overall/zonal residence times, flow patterns etc. The tracer trajectory length and overall/zonal residence time is found to increase with change in pebble's initial seeding position from the center towards the wall of the test reactor. Overall and zonal average velocities of pebbles are found to decrease from the center towards the wall. Discrete element method (DEM) based simulations of test reactor geometry were also carried out using commercial code EDEM(TM) and simulation results were validated using the obtained benchmark experimental data. In addition, EDEM(TM) based parametric sensitivity study of interaction properties was carried out which suggests that static friction characteristics play an important role from a packed/pebble beds structural characterization point of view. To make the RPT technique viable for practical applications and to enhance its accuracy, a novel and dynamic technique for RPT calibration was designed and developed. Preliminary feasibility results suggest that it can be implemented as a non

  11. Experimental investigation of a pilot-scale jet bubbling reactor for wet flue gas desulphurisation

    DEFF Research Database (Denmark)

    Zheng, Yuanjing; Kiil, Søren; Johnsson, Jan Erik

    2003-01-01

    In the present work, an experimental parameter study was conducted in a pilot-scale jet bubbling reactor for wet flue gas desulphurisation (FGD). The pilot plant is downscaled from a limestone-based, gypsum producing full-scale wet FGD plant. Important process parameters, such as slurry pH, inlet...... flue gas concentration of SO2, reactor temperature, and slurry concentration of Cl- have been varied. The degree of desulphurisation, residual limestone content of the gypsum, liquid phase concentrations, and solids content of the slurry were measured during the experimental series. The SO2 removal...

  12. Validation of High-Fidelity Reactor Physics Models for Support of the KJRR Experimental Campaign in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nigg, David W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nielsen, Joseph W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Norman, Daren R. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-07-01

    The Korea Atomic Energy Research Institute is currently in the process of qualifying a Low-Enriched Uranium fuel element design for the new Ki-Jang Research Reactor (KJRR). As part of this effort, a prototype KJRR fuel element was irradiated for several operating cycles in the Northeast Flux Trap of the Advanced Test Reactor (ATR) at the Idaho National Laboratory. The KJRR fuel element contained a very large quantity of fissile material (618g 235U) in comparison with historical ATR experiment standards (<1g 235U), and its presence in the ATR flux trap was expected to create a neutronic configuration that would be well outside of the approved validation envelope for the reactor physics analysis methods used to support ATR operations. Accordingly it was necessary, prior to high-power irradiation of the KJRR fuel element in the ATR, to conduct an extensive set of new low-power physics measurements with the KJRR fuel element installed in the ATR Critical Facility (ATRC), a companion facility to the ATR that is located in an immediately adjacent building, sharing the same fuel handling and storage canal. The new measurements had the objective of expanding the validation envelope for the computational reactor physics tools used to support ATR operations and safety analysis to include the planned KJRR irradiation in the ATR and similar experiments that are anticipated in the future. The computational and experimental results demonstrated that the neutronic behavior of the KJRR fuel element in the ATRC is well-understood, both in terms of its general effects on core excess reactivity and fission power distributions, its effects on the calibration of the core lobe power measurement system, as well as in terms of its own internal fission rate distribution and total fission power per unit ATRC core power. Taken as a whole, these results have significantly extended the ATR physics validation envelope, thereby enabling an entire new class of irradiation experiments.

  13. Experimental characterization of slurry bubble-column reactor hydrodynamics

    Energy Technology Data Exchange (ETDEWEB)

    Shollenberger, K.A.; Torczynski, J.R.; Jackson, N.B.; O`Hern, T.J.

    1997-09-01

    Sandia`s program to develop, implement, and apply diagnostics for hydrodynamic characterization of slurry bubble column reactors (SBCRs) at industrially relevant conditions is discussed. Gas liquid flow experiments are performed on an industrial scale. Gamma densitometry tomography (GDT) is applied to measure radial variations in gas holdup at one axial location. Differential pressure (DP) measurements are used to calculate volume averaged gas holdups along the axis of the vessel. The holdups obtained from DP show negligible axial variation for water but significant variations for oil, suggesting that the air water flow is fully developed (minimal flow variations in the axial direction) but that the air oil flow is still developing at the GDT measurement location. The GDT and DP gas holdup results are in good agreement for the air water flow but not for the air oil flow. Strong flow variations in the axial direction may be impacting the accuracy of one or both of these techniques. DP measurements are also acquired at high sampling frequencies (250 Hz) and are interpreted using statistical analyses to determine the physical mechanism producing each frequency component in the flow. This approach did not yield the information needed to determine the flow regime in these experiments. As a first step toward three phase material distribution measurements, electrical impedance tomography (EIT) and GDT are applied to a liquid solid flow to measure solids holdup. Good agreement is observed between both techniques and known values.

  14. Experimental Studies of NGNP Reactor Cavity Cooling System With Water

    Energy Technology Data Exchange (ETDEWEB)

    Corradini, Michael; Anderson, Mark; Hassan, Yassin; Tokuhiro, Akira

    2013-01-16

    This project will investigate the flow behavior that can occur in the reactor cavity cooling system (RCCS) with water coolant under the passive cooling-mode of operation. The team will conduct separate-effects tests and develop associated scaling analyses, and provide system-level phenomenological and computational models that describe key flow phenomena during RCCS operation, from forced to natural circulation, single-phase flow and two-phase flow and flashing. The project consists of the following tasks: Task 1. Conduct separate-effects, single-phase flow experiments and develop scaling analyses for comparison to system-level computational modeling for the RCCS standpipe design. A transition from forced to natural convection cooling occurs in the standpipe under accident conditions. These tests will measure global flow behavior and local flow velocities, as well as develop instrumentation for use in larger scale tests, thereby providing proper flow distribution among standpipes for decay heat removal. Task 2. Conduct separate-effects experiments for the RCCS standpipe design as two-phase flashing occurs and flow develops. As natural circulation cooling continues without an ultimate heat sink, water within the system will heat to temperatures approaching saturation , at which point two-phase flashing and flow will begin. The focus is to develop a phenomenological model from these tests that will describe the flashing and flow stability phenomena. In addition, one could determine the efficiency of phase separation in the RCCS storage tank as the two-phase flashing phenomena ensues and the storage tank vents the steam produced. Task 3. Develop a system-level computational model that will describe the overall RCCS behavior as it transitions from forced flow to natural circulation and eventual two-phase flow in the passive cooling-mode of operation. This modeling can then be used to test the phenomenological models developed as a function of scale.

  15. The CABRI fast neutron Hodoscope: Renovation, qualification program and first results following the experimental reactor restart

    Science.gov (United States)

    Chevalier, V.; Mirotta, S.; Guillot, J.; Biard, B.

    2018-01-01

    The CABRI experimental pulse reactor, located at the Cadarache nuclear research center, southern France, is devoted to the study of Reactivity Initiated Accidents (RIA). For the purpose of the CABRI International Program (CIP), managed and funded by IRSN, in the framework of an OECD/NEA agreement, a huge renovation of the facility has been conducted since 2003. The Cabri Water Loop was then installed to ensure prototypical Pressurized Water Reactor (PWR) conditions for testing irradiated fuel rods. The hodoscope installed in the CABRI reactor is a unique online fuel motion monitoring system, operated by IRSN and dedicated to the measurement of the fast neutrons emitted by the tested rod during the power pulse. It is one of the distinctive features of the CABRI reactor facility, which is operated by CEA. The system is able to determine the fuel motion, if any, with a time resolution of 1 ms and a spatial resolution of 3 mm. The hodoscope equipment has been upgraded as well during the CABRI facility renovation. This paper presents the main outcomes achieved with the hodoscope since October 2015, date of the first criticality of the CABRI reactor in its new Cabri Water Loop configuration. Results obtained during reactor commissioning phase functioning, either in steady-state mode (at low and high power, up to 23 MW) or in transient mode (start-up, possibly beyond 20 GW), are discussed.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-01-15

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

  17. Antenna design for fast ion collective Thomson scattering diagnostic for the international thermonuclear experimental reactor

    DEFF Research Database (Denmark)

    Leipold, Frank; Furtula, Vedran; Salewski, Mirko

    2009-01-01

    Fast ion physics will play an important role for the international thermonuclear experimental reactor (ITER), where confined alpha particles will affect and be affected by plasma dynamics and thereby have impacts on the overall confinement. A fast ion collective Thomson scattering (CTS) diagnostic...

  18. Experimental research in neutron physic and thermal-hydraulic at the CDTN Triga reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mesquita, Amir Z.; Souza, Rose Mary G.P.; Ferreira, Andrea V.; Pinto, Antonio J.; Costa, Antonio C.L.; Rezende, Hugo C., E-mail: amir@cdtn.b, E-mail: souzarm@cdtn.b, E-mail: avf@cdtn.b, E-mail: ajp@cdtn.b, E-mail: aclc@cdtn.b, E-mail: hcr@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The IPR-R1 TRIGA (Training, Research, Isotopes production, General Atomics) at Nuclear Technology Development Center (CDTN) is a pool type reactor cooled by natural circulation of light water and an open surface. TRIGA reactors, developed by General Atomics (GA), are the most widely used research reactor in the world and characterized by inherent safety. The IPR-R1 is the only Brazilian nuclear research reactor available and able to perform experiments in which interaction between neutronic and thermal-hydraulic areas occurs. The IPR-R1 has started up on November 11th, 1960. At that time the maximum thermal power was 30 kW. The present forced cooling system was built in the 70th and the power was upgraded to 100 kW. Recently the core configuration and instrumentation was upgraded again to 250 kW at steady state, and is awaiting the license of CNEN to operate definitely at this new power. This paper describes the experimental research project carried out in the IPR-R1 reactor that has as objective evaluate the behaviour of the reactor operational parameters, and mainly to investigate the influence of temperature on the neutronic variables. The research was supported by Research Support Foundation of the State of Minas Gerais (FAPEMIG) and Brazilian Council for Scientific and Technological Development (CNPq). The research project meets the recommendations of the IAEA, for safety, modernization and development of strategic plan for research reactors utilization. This work is in line with the strategic objectives of Brazil, which aims to design and construct the Brazilian Multipurpose research Reactor (RMB). (author)

  19. Experimental estimation of moderator temperature coefficient of reactivity of the IPEN/MB-01 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Rubens C. da; Bitelli, Ulysses D.; Mura, Luiz Ernesto C., E-mail: rubensrcs@usp.br, E-mail: ubitelli@ipen.br, E-mail: credidiomura@gmail.com [Universidade de Sao Paulo (PNV/POLI/USP), SP (Brazil). Arquitetura Naval e Departamento de Engenharia Oceanica; Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2017-07-01

    The aim of this article is to present the procedure for the experimental estimation of the Moderator Temperature Coefficient of Reactivity of the IPEN/MB-01 Research Reactor, a parameter that has an important role in the physics and the control operations of any reactor facility. At the experiment, the IPEN/MB-01 reactor went critical at the power of 1W (1% of its total power), and whose core configuration was 28 x 26 rectangular array of UO{sub 2} fuel rods, inside a light water (moderator) tank. In addition, there was a heavy water (D{sub 2}O) reflector installed in the West side of the core to obtain an adequate neutron reflection along the experiment. The moderator temperature was increased in steps of 4 °C, and the measurement of the mean moderator temperature was acquired using twelve calibrated thermocouples, placed around the reactor core. As a result, the mean value of -4.81 pcm/°C was obtained for such coefficient. The curves of ρ(T) (Reactivity x Temperature) and α{sup M}{sub T}(T)(Moderator Temperature Coefficient of Reactivity x Temperature) were developed using data from an experimental measurement of the integral reactivity curves through the Stable Period and Inverse Kinetics Methods, that was carried out at the reactor with the same core configuration. Such curves were compared and showed a very similar behavior between them. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Shuck, Author B.

    1953-04-20

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-04-01

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

  2. The D&D of the Experimental Boiling Water Reactor (EBWR)

    Energy Technology Data Exchange (ETDEWEB)

    Fellhauer, C.R.; Boling, L.E.; Yule, T.J.; Bhattacharyya, S.K.

    1996-03-01

    Argonne National Laboratory has completed the D&D of the Experimental Boiling Water Reactor. The Project consisted of decontaminating and for packaging as radioactive waste the reactor vessel and internals, contaminated piping systems, miscellaneous tanks, pumps, and associated equipment. The D&D work involved dismantling process equipment and associated plumbing, ductwork drain lines, etc., performing size reduction of reactor vessel internals in the fuel pool, packaging and manifesting all radioactive and mixed waste, and performing a thorough survey of the facility after the removal of activated and contaminated material. Non-radioactive waste was disposed of in the ANL-E landfill or recycled. In January 1996 the EBWR facility was formally decommissioned and transferred from EM-40 to EM-30. This paper will discuss the details of this ten year effort.

  3. On the possibility of experimentally confirming the hypothesis of reactor antineutrino passage into a sterile state

    Science.gov (United States)

    Serebrov, A. P.; Fomin, A. K.; Zinov'ev, V. G.; Loginov, Yu. E.; Onegin, M. S.; Gagarskiy, A. M.; Petrov, G. A.; Solovei, V. A.; Chernyi, A. V.; Zherebtsov, O. M.; Martem'yanov, V. P.; Tsinoev, V. G.; Tarasenkov, V. G.; Aleshin, V. I.; Petelin, A. L.; Pavlov, S. V.; Svyatkin, M. N.; Izhutov, A. L.; Sazontov, S. A.; Ryazanov, D. K.; Gromov, M. O.; Khramkov, N. S.; Rykalin, V. I.

    2013-07-01

    The "Neutrino-4" experiment for the 100-MW SM-3 reactor has been developed with the aim of testing the reactor antineutrino anomaly at Petersburg Nuclear Physics Institute. The advantages of this reactor for studying the antineutrino anomaly are (i) a low background level and (ii) small dimensions (35 × 42 × 42 cm) of the active zone. Operation of a position-sensitive antineutrino detector comprising five working sections and moving so as to cover a region of distances within 6-13 m from the active zone has been simulated by the Monte-Carlo method. The range of experimental sensitivity with respect to the oscillation parameters Δ m 2 and sin22θ is determined, which will make it possible to confirm the hypothesis of antineutrino oscillations into a sterile state.

  4. Material unaccounted for at the Southwest Experimental Fast Oxide Reactor: The SEFOR MUF

    Energy Technology Data Exchange (ETDEWEB)

    Higinbotham, W.A.

    1994-11-07

    The U.S. Atomic Energy Commission contracted with the General Electric Company to design, construct, and operate the Southwest Experimental Fast Oxide Reactor (SEFOR) to measure the Doppler effect for fast neutron breeder reactors. It contracted with Nuclear Fuel Services to fabricate the fuel rods for the reactor. When the reactor went critical in May, 1969, it appeared that some of the mixed uranium-plutonium oxide (MOX) fuel rods did not contain the specified quantity of plutonium. The SEFOR operators soon found several fuel rods which appeared to be low in plutonium. The safeguards group at Brookhaven was asked to look into the problem and, if possible, determine how much plutonium was missing from the unirradiated rods and from the larger number which had been slightly irradiated in the reactor. It was decided that the plutonium content of the unirradiated and irradiated rods could be measured relative to a reference rod using a high resolution gamma-ray detector and also by neutron measurements using an auto-correlation circuit recently developed at the Naval Research Laboratory (NRL). During the next two years, Brookhaven personnel and C.V. Strain of NRL made several trips to the SEFOR reactor. About 250 of the 775 rods were measured by two or more methods, using a sodium-iodide detector, a high-resolution germanium detector, a neutron detector, or the reactor (to measure reactivity). The research team concluded that 4.6 {+-} 0.46 kg of plutonium was missing out of the 433 kg that the rods should have contained. This report describes the SEFOR experiment and the procedures used to determine the material unaccounted for, or MUF.

  5. Experimental investigation of a new method for advanced fast reactor shutdown cooling

    Science.gov (United States)

    Pakholkov, V. V.; Kandaurov, A. A.; Potseluev, A. I.; Rogozhkin, S. A.; Sergeev, D. A.; Troitskaya, Yu. I.; Shepelev, S. F.

    2017-07-01

    We consider a new method for fast reactor shutdown cooling using a decay heat removal system (DHRS) with a check valve. In this method, a coolant from the decay heat exchanger (DHX) immersed into the reactor upper plenum is supplied to the high-pressure plenum and, then, inside the fuel subassemblies (SAs). A check valve installed at the DHX outlet opens by the force of gravity after primary pumps (PP-1) are shut down. Experimental studies of the new and alternative methods of shutdown cooling were performed at the TISEY test facility at OKBM. The velocity fields in the upper plenum of the reactor model were obtained using the optical particle image velocimetry developed at the Institute of Applied Physics (Russian Academy of Sciences). The study considers the process of development of natural circulation in the reactor and the DHRS models and the corresponding evolution of the temperature and velocity fields. A considerable influence of the valve position in the displacer of the primary pump on the natural circulation of water in the reactor through the DHX was discovered (in some modes, circulation reversal through the DHX was obtained). Alternative DHRS designs without a shell at the DHX outlet with open and closed check valve are also studied. For an open check valve, in spite of the absence of a shell, part of the flow is supplied through the DHX pipeline and then inside the SA simulators. When simulating power modes of the reactor operation, temperature stratification of the liquid was observed, which increased in the cooling mode via the DHRS. These data qualitatively agree with the results of tests at BN-600 and BN-800 reactors.

  6. COMPUTATIONAL AND EXPERIMENTAL MODELING OF SLURRY BUBBLE COLUMN REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Paul Lam; Dimitri Gidaspow

    2000-09-01

    The objective if this study was to develop a predictive experimentally verified computational fluid dynamics (CFD) model for gas-liquid-solid flow. A three dimensional transient computer code for the coupled Navier-Stokes equations for each phase was developed. The principal input into the model is the viscosity of the particulate phase which was determined from a measurement of the random kinetic energy of the 800 micron glass beads and a Brookfield viscometer. The computed time averaged particle velocities and concentrations agree with PIV measurements of velocities and concentrations, obtained using a combination of gamma-ray and X-ray densitometers, in a slurry bubble column, operated in the bubbly-coalesced fluidization regime with continuous flow of water. Both the experiment and the simulation show a down-flow of particles in the center of the column and up-flow near the walls and nearly uniform particle concentration. Normal and shear Reynolds stresses were constructed from the computed instantaneous particle velocities. The PIV measurement and the simulation produced instantaneous particle velocities. The PIV measurement and the simulation produced similar nearly flat horizontal profiles of turbulent kinetic energy of particles. This phase of the work was presented at the Chemical Reaction Engineering VIII: Computational Fluid Dynamics, August 6-11, 2000 in Quebec City, Canada. To understand turbulence in risers, measurements were done in the IIT riser with 530 micron glass beads using a PIV technique. The results together with simulations will be presented at the annual meeting of AIChE in November 2000.

  7. Experimental and Kinetic Modeling Study of Ethyl Levulinate Oxidation in a Jet-Stirred Reactor

    KAUST Repository

    Wang, Jui-Yang

    2017-06-01

    A jet-stirred reactor was designed and constructed in the Clean Combustion Research Center (CCRC) at King Abdullah University of Science and Technology (KAUST); was validated with n-heptane, iso-octane oxidation and cyclohexene pyrolysis. Different configurations of the setup have been tested to achieve good agreement with results from the literature. Test results of the reactor indicated that installation of a pumping system at the downstream side in the experimental apparatus was necessary to avoid the reoccurrence of reactions in the sampling probe. Experiments in ethyl levulinate oxidation were conducted in the reactor under several equivalence ratios, from 600 to 1000 K, 1 bar and 2 s residence time. Oxygenated species detected included methyl vinyl ketone, levulinic acid and ethyl acrylate. Ethylene, methane, carbon monoxide, hydrogen, oxygen and carbon dioxide were further quantified with a gas chromatography, coupled with a flame ionization detector and a thermal conductivity detector. The ethyl levulinate chemical kinetic model was first developed by Dr. Stephen Dooley, Trinity College Dublin, and simulated under the same conditions, using the Perfect-Stirred Reactor code in Chemkin software. In comparing the simulation results with experimental data, some discrepancies were noted; predictions of ethylene production were not well matched. The kinetic model was improved by updating several classes of reactions: unimolecular decomposition, H-abstraction, C-C and C-O beta-scissions of fuel radicals. The updated model was then compared again with experimental results and good agreement was achieved, proving that the concerted eliminated reaction is crucial for the kinetic mechanism formulation of ethyl levulinate. In addition, primary reaction pathways and sensitivity analysis were performed to describe the role of molecular structure in combustion (800 and 1000 K for ethyl levulinate oxidation in the jet-stirred reactor).

  8. Results of theoretical and experimental studies of hydrodynamics of circulation loops in circulating fluidized bed reactors and systems with interconnected reactors

    Science.gov (United States)

    Ryabov, G. A.; Folomeev, O. M.; Sankin, D. A.; Melnikov, D. A.

    2015-02-01

    Problems of the calculation of circulation loops in circulating fluidized bed reactors and systems with interconnected reactors (polygeneration systems for the production of electricity, heat, and useful products and chemical cycles of combustion and gasification of solid fuels)are considered. A method has been developed for the calculation of circulation loop of fuel particles with respect to boilers with circulating fluidized bed (CFB) and systems with interconnected reactors with fluidized bed (FB) and CFB. New dependences for the connection between the fluidizing agent flow (air, gas, and steam) and performance of reactors and for the whole system (solids flow rate, furnace and cyclone pressure drops, and bed level in the riser) are important elements of this method. Experimental studies of hydrodynamics of circulation loops on the aerodynamic unit have been conducted. Experimental values of pressure drop of the horizontal part of the L-valve, which satisfy the calculated dependence, have been obtained.

  9. Experimentation on the anaerobic filter reactor for biogas production using rural domestic wastewater

    Science.gov (United States)

    Leju Celestino Ladu, John; Lü, Xi-wu; Zhong, Zhaoping

    2017-08-01

    The biogas production from anaerobic filter (AF) reactor was experimented in Taihu Lake Environmental Engineering Research Center of Southeast University, Wuxi, China. Two rounds of experimental operations were conducted in a laboratory scale at different Hydraulic retention time (HRT) and wastewater temperature. The biogas production rate during the experimentation was in the range of 4.63 to 11.78 L/d. In the first experimentation, the average gas production rate was 10.08 L/d, and in the second experimentation, the average gas production rate was 4.97 L/d. The experimentation observed the favorable Hydraulic Retention Time and wastewater temperature in AF was three days and 30.95°C which produced the gas concentration of 11.78 L/d. The HRT and wastewater temperature affected the efficiency of the AF process on the organic matter removal and nutrients removal as well. It can be deduced from the obtained results that HRT and wastewater temperature directly affects the efficiency of the AF reactor in biogas production. In conclusion, anaerobic filter treatment of organic matter substrates from the rural domestic wastewater increases the efficiency of the AF reactor on biogas production and gives a number of benefits for the management of organic wastes as well as reduction in water pollution. Hence, the operation of the AF reactor in rural domestic wastewater treatment can play an important element for corporate economy of the biogas plant, socio-economic aspects and in the development of effective and feasible concepts for wastewater management, especially for people in rural low-income areas.

  10. Kinetics of vinyl acetate emulsion polymerization in a pulsed tubular reactor: comparison between experimental and simulation results

    OpenAIRE

    Sayer C.; Palma M.; Giudici R.

    2002-01-01

    A new reactor, the pulsed sieve plate column (PSPC), was developed to perform continuous emulsion polymerization reactions. This reactor combines the enhanced flexibility of tubular reactors with the mixing behavior provided by sieved plates and by the introduction of pulses that is important to prevent emulsion destabilization. The main objective of this work is to study the kinetics of vinyl acetate (VA) emulsion polymerization reactions performed in this PSPC. Therefore, both experimental ...

  11. High energy resolution characteristics on 14MeV neutron spectrometer for fusion experimental reactor

    Energy Technology Data Exchange (ETDEWEB)

    Iguchi, Tetsuo [Tokyo Univ., Tokai, Ibaraki (Japan). Nuclear Engineering Research Lab.; Takada, Eiji; Nakazawa, Masaharu

    1996-10-01

    A 14MeV neutron spectrometer suitable for an ITER-like fusion experimental reactor is now under development on the basis of a recoil proton counter telescope principle in oblique scattering geometry. To verify its high energy resolution characteristics, preliminary experiments are made for a prototypical detector system. The comparison results show reasonably good agreement and demonstrate the possibility of energy resolution of 2.5% in full width at half maximum for 14MeV neutron spectrometry. (author)

  12. An experimental study of hypervapotron structure in external reactor vessel cooling

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Yufeng; Zhang, Ming [State Nuclear Power Technology R& D Center (Beijing), Beijing (China); Hou, Fangxin [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing (China); Gao, Tianfang [State Nuclear Power Technology R& D Center (Beijing), Beijing (China); Chen, Peipei, E-mail: chenpeipei@snptc.com.cn [State Power Investment Group Corporation, Beijing (China)

    2016-07-15

    Highlights: • Experiments are performed to study the application of hypervapotron in ERVC design. • CHF experiments on two surfaces are conducted under different flow conditions. • Hypervapotron improves CHF performance by 40–60% compared with smooth surface. • Visualization shows fin structure removes vapor mushroom for better liquid supply. - Abstract: In vessel retention (IVR) is one of the key strategies for many advanced LWR designs to mitigate postulated severe accidents. The success of IVR substantially relies on external reactor vessel cooling (ERVC) by which the decay heat is removed from the melt core in the reactor vessel lower head. The main challenge of IVR is to provide an adequate safety margin of ERVC against critical heat flux (CHF) of subcooled flow boiling in the reactor lower head flow channel. Due to uncertainties in corium melt pool configuration, large CHF margin of ERVC is usually required by regulatory authorities to demonstrate reliability of severe accident mitigation methods. Various CHF enhancement designs have been proposed and studied in literature. In this paper, an experimental study of hypervapotron structure as a novel design to improve CHF performance of ERVC is conducted. Hypervapotron is chosen as one of the potential engineering options for International Thermonuclear Experimental Reactor (ITER) program as a divertor structure to remove highly intense heat from fusion chamber. This study is to conduct CHF experiments at typical PWR ERVC working conditions. The CHF experiments are performed in a 30 mm by 61 mm rectangular flow channel with a 200 mm long heated surface along the flow direction. Both smooth and hypervapotron surface are tested at various inclination angles of the test section to simulate various positions of the reactor lower head. The hypervapotron is found to have a 40–60% CHF improvement compared with the smooth surface. The high speed visualization indicates that hypervapotron is able to

  13. Advanced Reactors-Intermediate Heat Exchanger (IHX) Coupling: Theoretical Modeling and Experimental Validation

    Energy Technology Data Exchange (ETDEWEB)

    Utgikar, Vivek [Univ. of Idaho, Moscow, ID (United States); Sun, Xiaodong [The Ohio State Univ., Columbus, OH (United States); Christensen, Richard [The Ohio State Univ., Columbus, OH (United States); Sabharwall, Piyush [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-12-29

    The overall goal of the research project was to model the behavior of the advanced reactorintermediate heat exchange system and to develop advanced control techniques for off-normal conditions. The specific objectives defined for the project were: 1. To develop the steady-state thermal hydraulic design of the intermediate heat exchanger (IHX); 2. To develop mathematical models to describe the advanced nuclear reactor-IHX-chemical process/power generation coupling during normal and off-normal operations, and to simulate models using multiphysics software; 3. To develop control strategies using genetic algorithm or neural network techniques and couple these techniques with the multiphysics software; 4. To validate the models experimentally The project objectives were accomplished by defining and executing four different tasks corresponding to these specific objectives. The first task involved selection of IHX candidates and developing steady state designs for those. The second task involved modeling of the transient and offnormal operation of the reactor-IHX system. The subsequent task dealt with the development of control strategies and involved algorithm development and simulation. The last task involved experimental validation of the thermal hydraulic performances of the two prototype heat exchangers designed and fabricated for the project at steady state and transient conditions to simulate the coupling of the reactor- IHX-process plant system. The experimental work utilized the two test facilities at The Ohio State University (OSU) including one existing High-Temperature Helium Test Facility (HTHF) and the newly developed high-temperature molten salt facility.

  14. Development of a membrane-assisted fluidized bed reactor - 2 - Experimental demonstration and modeling for the partial oxidation of methanol

    NARCIS (Netherlands)

    Deshmukh, S.A.R.K.; Laverman, J.A.; van Sint Annaland, M.; Kuipers, J.A.M.

    2005-01-01

    A small laboratory-scale membrane-assisted fluidized bed reactor (MAFBR) was constructed in order to experimentally demonstrate the reactor concept for the partial oxidation of methanol to formaldehyde. Methanol conversion and product selectivities were measured at various overall fluidization

  15. Experimental and Numerical Evaluation of the By-Pass Flow in a Catalytic Plate Reactor for Hydrogen Production

    DEFF Research Database (Denmark)

    Sigurdsson, Haftor Örn; Kær, Søren Knudsen

    2011-01-01

    Numerical and experimental study is performed to evaluate the reactant by-pass flow in a catalytic plate reactor with a coated wire mesh catalyst for steam reforming of methane for hydrogen generation. By-pass of unconverted methane is evaluated under different wire mesh catalyst width to reactor...

  16. Conceptual studies of toroidal field magnets for the tokamak experimental power reactor. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Buncher, B.R.; Chi, J.W.H.; Fernandez, R.

    1976-10-26

    This report documents the principal results of a Conceptual Design Study for the Superconducting Toroidal Field System for a Tokamak Experimental Power Reactor. Two concepts are described for peak operating fields at the windings of 8 tesla, and 12 tesla, respectively. The design and manufacturing considerations are treated in sufficient detail that cost and schedule estimates could be developed. Major uncertainties in the design are identified and their potential impact discussed, along with recommendations for the necessary research and development programs to minimize these uncertainties. The minimum dimensions of a sub-size test coil for experimental qualification of the full size design are developed and a test program is recommended.

  17. Experimental and kinetic modeling study of 3-methylheptane in a jet-stirred reactor

    KAUST Repository

    Karsenty, Florent

    2012-08-16

    Improving the combustion of conventional and alternative fuels in practical applications requires the fundamental understanding of large hydrocarbon combustion chemistry. The focus of the present study is on a high-molecular-weight branched alkane, namely, 3-methylheptane, oxidized in a jet-stirred reactor. This fuel, along with 2-methylheptane, 2,5-dimethylhexane, and n-octane, are candidate surrogate components for conventional diesel fuels derived from petroleum, synthetic Fischer-Tropsch diesel and jet fuels derived from coal, natural gas, and/or biomass, and renewable diesel and jet fuels derived from the thermochemical treatment of bioderived fats and oils. This study presents new experimental results along with a low- and high-temperature chemical kinetic model for the oxidation of 3-methylheptane. The proposed model is validated against these new experimental data from a jet-stirred reactor operated at 10 atm, over the temperature range of 530-1220 K, and for equivalence ratios of 0.5, 1, and 2. Significant effort is placed on the understanding of the effects of methyl substitution on important combustion properties, such as fuel reactivity and species formation. It was found that 3-methylheptane reacts more slowly than 2-methylheptane at both low and high temperatures in the jet-stirred reactor. © 2012 American Chemical Society.

  18. Experimental assessment of accident scenarios for the high temperature reactor fuel system

    Energy Technology Data Exchange (ETDEWEB)

    Seeger, O.; Avincola, V.; Bottomley, P.D.W.; Rondinella, V.V. [European Commission Joint Research Centre - Institute for Transuranium Elements (JRC-ITU) (Germany)

    2012-11-01

    The High Temperature Reactor (HTR) is an advanced reactor concept with particular safety features. Fuel elements are constituted by a graphite matrix containing sub-mm-sized fuel particles with TRISO (tri-isotropic) coating designed to provide high fission product retention. Passive safety features of the HTR include a low power density in the core compared to other reactor designs; this ensures sufficient heat transport in a loss of coolant accident scenario. The temperature during such events would not exceed 1600 C, remaining well below the melting point of the fuel. An experimental assessment of the fuel behaviour under severe accident conditions is necessary to confirm the fission product retention of TRISO coated particles and to validate relevant computer codes. Though helium is used as coolant for the HTR system, additional corrosion effects come into play in case of an in-leakage affecting the primary circuit. The experimental scope of the present work focuses on two key aspects associated with the HTR fuel safety. Fission product retention at high temperatures (up to {proportional_to}1800 C) is analyzed with the so-called cold finger apparatus (KueFA: Kuehlfinger-Apparatur), while the performance of HTR fuel elements in case of air/steam ingress accidents is assessed with a high temperature corrosion apparatus (KORA: Korrosions-Apparatur). (orig.)

  19. Experimental measurement and CFD simulation on the hydrodynamics of an internal-loop airlift reactor

    Directory of Open Access Journals (Sweden)

    Liew Shi Yan

    2017-01-01

    Full Text Available This paper concerns with the experimental measurement and computational fluid dynamics simulation on local hydrodynamics of a gas-liquid internal-loop airlift reactor. The aim of this work is to study the sensitivity of the drag models and the significance of considering the lift force on the predictive accuracy of the simulation. The experimental analysis was carried out using laser Doppler anemometry at three different heights (i.e. Y = 0.20 m, 0.30 m and 0.38 m across the riser and downcomerat volumetric flow rate of 0.30 m3/h to provide validation for the simulation results. A transient three-dimensional gasliquid internal-loop airlift reactor was carried out using FLUENT 16.2 by implementing the two-fluid model approach. The Eulerian-Eulerian multiphase and standard κ-ε dispersed turbulence model wereemployed in this study. Results suggest that the spherical drag model performed poorly and that the drag model governed by Rayleigh-Taylor shows promising accuracy in the prediction of overall mean axial liquid velocity. On the other hand, the consideration of lift model shows slightly improvement in accuracy. These findings may serve as a guidance for future scale-up and design of airlift reactor studies

  20. Experimental facility for development of high-temperature reactor technology: instrumentation needs and challenges

    Directory of Open Access Journals (Sweden)

    Sabharwall Piyush

    2015-01-01

    Full Text Available A high-temperature, multi-fluid, multi-loop test facility is under development at the Idaho National Laboratory for support of thermal hydraulic materials, and system integration research for high-temperature reactors. The experimental facility includes a high-temperature helium loop, a liquid salt loop, and a hot water/steam loop. The three loops will be thermally coupled through an intermediate heat exchanger (IHX and a secondary heat exchanger (SHX. Research topics to be addressed include the characterization and performance evaluation of candidate compact heat exchangers such as printed circuit heat exchangers (PCHEs at prototypical operating conditions. Each loop will also include an interchangeable high-temperature test section that can be customized to address specific research issues associated with each working fluid. This paper also discusses needs and challenges associated with advanced instrumentation for the multi-loop facility, which could be further applied to advanced high-temperature reactors. Based on its relevance to advanced reactor systems, the new facility has been named the Advanced Reactor Technology Integral System Test (ARTIST facility. A preliminary design configuration of the ARTIST facility will be presented with the required design and operating characteristics of the various components. The initial configuration will include a high-temperature (750 °C, high-pressure (7 MPa helium loop thermally integrated with a molten fluoride salt (KF-ZrF4 flow loop operating at low pressure (0.2 MPa, at a temperature of ∼450 °C. The salt loop will be thermally integrated with the steam/water loop operating at PWR conditions. Experiment design challenges include identifying suitable materials and components that will withstand the required loop operating conditions. The instrumentation needs to be highly accurate (negligible drift in measuring operational data for extended periods of times, as data collected will be

  1. Experimental Evaluation of a Water Shield for a Surface Power Reactor

    Science.gov (United States)

    Pearson, J. Boise; Reid, Robert S.

    2006-01-01

    As part of the Vision for Space Exploration the end of the next decade will bring man back to the surface of the moon. One of the most critical issues for the establishment of human presence on the moon will be the availability of compact power sources. The establishment of man on the moon will require power from greater than 10's of kWt's in follow on years. Nuclear reactors are extremely we11 suited to meet the needs for power generation on the lunar or Martian surface. reactor system. Several competing concepts exist for lightweight, safe, robust shielding systems such as a water shield, lithium hydride (LiH), Boron Carbide, and others. Water offers several potential advantages, including reduced cost, reduced technical risk, and reduced mass. Water has not typically been considered for space reactor applications because of the need for gravity to remove the potential for radiation streaming paths. The water shield concept relies on predictions of passive circulation of the shield water by natural convection to adequately cool the shield. This prediction needs to be experimentally evaluated, especially for shields with complex geometries. MSFC has developed the experience and fac necessary to do this evaluation in the Early Flight Fission - Test Facility (EFF-TF).

  2. 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.

  3. Generation of an activation map for decommissioning planning of the Berlin Experimental Reactor-II

    Science.gov (United States)

    Lapins, Janis; Guilliard, Nicole; Bernnat, Wolfgang

    2017-09-01

    The BER-II is an experimental facility with 10 MW that was operated since 1974. Its planned operation will end in 2019. To support the decommissioning planning, a map with the overall distribution of relevant radionuclides has to be created according to the state of the art. In this paper, a procedure to create these 3-d maps using a combination of MCNP and deterministic methods is presented. With this approach, an activation analysis is performed for the whole reactor geometry including the most remote parts of the concrete shielding.

  4. ESTIMATION OF ROUTINE DISCHARGE OF RADIONUCLIDES ON POWER REACTOR EXPERIMENTAL RDE

    Directory of Open Access Journals (Sweden)

    P.M. Udiyani

    2017-02-01

    Full Text Available Experimental power reactor (RDE which is planned to be constructed by BATAN is a kind of High Temperature Gas Cooled Reactor (HTGR with 10 MWth power. HTGR is a helium gas-cooled reactor with TRISO-coated fuel that is able to confine fission products remained in the core. Although the fission products released into the environment are very small, in order to comply the regulations the study about environmental radiation on normal or routine operation condition need to be performed. Estimation of radiology in the environment involves the source term released into the environment under routine operation condition. The purpose of this study is to estimate the source term released into the environment based on postulation of normal or routine operations of RDE. The research approach starts with an assumption that there are defects and impurities in the TRISO fuel because of limitation during the fabrication. Mechanism of fission products release from the fuel to the environment was created based on the safety features design of RDE. Radionuclides inventories in the reactor were calculated using ORIGEN-2 whose library has been modified for HTGR type, and the assumptions of defects of the TRISO fuel and release fraction for each compartment of RDE safety system used a reference parameter. The results showed that the important source terms of RDE are group of noble gases (Kr and Xe, halogen (I, Sr, Cs, H-3, and Ag. Activities of RDE source terms for routine operations have no significant difference with the HTGR source terms with the same power. Keywords: routine discharge, radionuclide, source term, RDE, HTGR

  5. Approaches to experimental validation of high-temperature gas-cooled reactor components

    Energy Technology Data Exchange (ETDEWEB)

    Belov, S.E. [Joint Stock Company ' Afrikantov OKB Mechanical Engineering' , Burnakovsky Proezd, 15, Nizhny Novgorod 603074 (Russian Federation); Borovkov, M.N., E-mail: borovkov@okbm.nnov.ru [Joint Stock Company ' Afrikantov OKB Mechanical Engineering' , Burnakovsky Proezd, 15, Nizhny Novgorod 603074 (Russian Federation); Golovko, V.F.; Dmitrieva, I.V.; Drumov, I.V.; Znamensky, D.S.; Kodochigov, N.G. [Joint Stock Company ' Afrikantov OKB Mechanical Engineering' , Burnakovsky Proezd, 15, Nizhny Novgorod 603074 (Russian Federation); Baxi, C.B.; Shenoy, A.; Telengator, A. [General Atomics, 3550 General Atomics Court, CA (United States); Razvi, J., E-mail: Junaid.Razvi@ga.com [General Atomics, 3550 General Atomics Court, CA (United States)

    2012-10-15

    Highlights: Black-Right-Pointing-Pointer Computational and experimental investigations of thermal and hydrodynamic characteristics for the equipment. Black-Right-Pointing-Pointer Vibroacoustic investigations. Black-Right-Pointing-Pointer Studies of the electromagnetic suspension system on GT-MHR turbo machine rotor models. Black-Right-Pointing-Pointer Experimental investigations of the catcher bearings design. - Abstract: The special feature of high-temperature gas-cooled reactors (HTGRs) is stressed operating conditions for equipment due to high temperature of the primary circuit helium, up to 950 Degree-Sign C, as well as acoustic and hydrodynamic loads upon the gas path elements. Therefore, great significance is given to reproduction of real operation conditions in tests. Experimental investigation of full-size nuclear power plant (NPP) primary circuit components is not practically feasible because costly test facilities will have to be developed for the power of up to hundreds of megawatts. Under such conditions, the only possible process to validate designs under development is representative tests of smaller scale models and fragmentary models. At the same time, in order to take in to validated account the effect of various physical factors, it is necessary to ensure reproduction of both individual processes and integrated tests incorporating needed integrated investigations. Presented are approaches to experimental validation of thermohydraulic and vibroacoustic characteristics for main equipment components and primary circuit path elements under standard loading conditions, which take account of their operation in the HTGR. Within the framework of the of modular helium reactor project, including a turbo machine in the primary circuit, a new and difficult problem is creation of multiple-bearing flexible vertical rotor. Presented are approaches to analytical and experimental validation of the rotor electromagnetic bearings, catcher bearings, flexible rotor

  6. Modeling and Experimental Studies of Mercury Oxidation and Adsorption in a Fixed-Bed Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Buitrago, Paula A.; Morrill, Mike; Lighty, JoAnn S.; Silcox, Geoffrey D.

    2009-06-15

    This report presents experimental and modeling mercury oxidation and adsorption data. Fixed-bed and single-particle models of mercury adsorption were developed. The experimental data were obtained with two reactors: a 300-W, methane-fired, tubular, quartz-lined reactor for studying homogeneous oxidation reactions and a fixed-bed reactor, also of quartz, for studying heterogeneous reactions. The latter was attached to the exit of the former to provide realistic combustion gases. The fixed-bed reactor contained one gram of coconut-shell carbon and remained at a temperature of 150°C. All methane, air, SO2, and halogen species were introduced through the burner to produce a radical pool representative of real combustion systems. A Tekran 2537A Analyzer coupled with a wet conditioning system provided speciated mercury concentrations. At 150°C and in the absence of HCl or HBr, the mercury uptake was about 20%. The addition of 50 ppm HCl caused complete capture of all elemental and oxidized mercury species. In the absence of halogens, SO2 increased the mercury adsorption efficiency to up to 30 percent. The extent of adsorption decreased with increasing SO2 concentration when halogens were present. Increasing the HCl concentration to 100 ppm lessened the effect of SO2. The fixed-bed model incorporates Langmuir adsorption kinetics and was developed to predict adsorption of elemental mercury and the effect of multiple flue gas components. This model neglects intraparticle diffusional resistances and is only applicable to pulverized carbon sorbents. It roughly describes experimental data from the literature. The current version includes the ability to account for competitive adsorption between mercury, SO2, and NO2. The single particle model simulates in-flight sorbent capture of elemental mercury. This model was developed to include Langmuir and Freundlich isotherms, rate equations, sorbent feed rate, and

  7. Overview of the TIBER 2 (Thermal Ignition/Burn Experimental Reactor) design

    Science.gov (United States)

    Henning, C. D.; Logan, B. G.

    1987-10-01

    The TIBER 2 Tokamak Ignition/Burn Experimental Reactor design is the result of efforts by numerous people and institutions, including many fusion laboratories, universities, and industries. While subsystems will be covered extensively in other reports, this overview will attempt to place the work in perspective. Major features of the design are compact size, low cost, and steady-state operation. These are achieved through plasma shaping and innovative features such as radiation tolerant magnets and optimized shielding. While TIBER 2 can operate in a pulsed mode, steady-state is preferred for nuclear testing. Current drive is achieved by a combination of lower hybrid and neutral beams. In addition, 10 MW of ECR is added for disruption control and current drive profiling. The TIBER 2 design has been the US option in preparation for the International Thermonuclear Experimental Reactor (ITER). Other equivalent national designs are the NET in Europe, the FER in Japan and the OTR in the USSR. These designs will help set the basis for the new international design effort.

  8. Modeling and experimental validation of hydrodynamics in an ultrasonic batch reactor.

    Science.gov (United States)

    Ajmal, M; Rusli, S; Fieg, G

    2016-01-01

    Simulation of hydrodynamics in ultrasonic batch reactor containing immobilized enzymes as catalyst is done. A transducer with variable power and constant frequency (24 kHz) is taken as source of ultrasound (US). Simulation comprises two steps. In first step, acoustic pressure field is simulated and in second step effect of this field on particle trajectories is simulated. Simulation results are compared with experimentally determined particle trajectories using PIV Lab (particle image velocimetry). Effect of varying ultrasonic power, positioning and number of ultrasonic sources on particle trajectories is studied. It is observed that catalyst particles tend to orientate according to pattern of acoustic pressure field. An increase in ultrasonic power increases particle velocity and also brings more particles into motion. Simulation results are found to be in agreement with experimentally determined data. Copyright © 2015 Elsevier B.V. All rights reserved.

  9. Kinetics of vinyl acetate emulsion polymerization in a pulsed tubular reactor: comparison between experimental and simulation results

    Directory of Open Access Journals (Sweden)

    Sayer C.

    2002-01-01

    Full Text Available A new reactor, the pulsed sieve plate column (PSPC, was developed to perform continuous emulsion polymerization reactions. This reactor combines the enhanced flexibility of tubular reactors with the mixing behavior provided by sieved plates and by the introduction of pulses that is important to prevent emulsion destabilization. The main objective of this work is to study the kinetics of vinyl acetate (VA emulsion polymerization reactions performed in this PSPC. Therefore, both experimental studies and reaction simulations were performed. Results showed that it is possible to obtain high conversions with rather low residence times in the PSPC.

  10. CFD and experimental investigation of sloshing parameters for the safety assessment of HLM reactors

    Energy Technology Data Exchange (ETDEWEB)

    Myrillas, Konstantinos, E-mail: myrillas@vki.ac.be [von Karman Institute for Fluid Dynamics, Chaussée de Waterloo 72, B-1640 Rhode-St-Genèse (Belgium); Planquart, Philippe, E-mail: philippe.planquart@vki.ac.be [von Karman Institute for Fluid Dynamics, Chaussée de Waterloo 72, B-1640 Rhode-St-Genèse (Belgium); Simonini, Alessia, E-mail: Simonini@vki.ac.be [von Karman Institute for Fluid Dynamics, Chaussée de Waterloo 72, B-1640 Rhode-St-Genèse (Belgium); Buchlin, Jean-Marie, E-mail: buchlin@vki.ac.be [von Karman Institute for Fluid Dynamics, Chaussée de Waterloo 72, B-1640 Rhode-St-Genèse (Belgium); Schyns, Marc, E-mail: mschyns@SCKCEN.BE [SCK-CEN, Boeretang 200, B-2400 Mol (Belgium)

    2017-02-15

    Highlights: • Comparison of sloshing behavior in cylindrical tank using mercury and water. • Flow visualization of liquid sloshing in resonance case. • CFD simulations of sloshing with OpenFOAM, using the VOF method. • Qualitative and quantitative comparison of experimental and numerical results. • Evaluation of sloshing forces on the tank walls from numerical simulations. - Abstract: For the safety assessment of Heavy Liquid Metal nuclear reactors under seismic excitation, sloshing phenomena can be of great concern. The earthquake motions are transferred to the liquid coolant which oscillates inside the vessel, exerting additional forces on the walls and internal structures. The present study examines the case of MYRRHA, a multi-purpose experimental reactor with LBE as coolant, developed by SCK·CEN. The sloshing behavior of liquid metals is studied through a comparison between mercury and water in a cylindrical tank. Experimental investigation of sloshing is carried out using optical techniques with the shaking table facility SHAKESPEARE at the von Karman Institute. Emphasis is given on the resonance case, where maximum forces occur on the tank walls. The experimental cases are reproduced numerically with the CFD software OpenFOAM, using the VOF method to track the liquid interface. The non-linear nature of sloshing is observed through visualization, where swirling is shown in the resonance case. The complex behavior is well reproduced by the CFD simulations, providing good qualitative validation of the numerical tools. A quantitative comparison of the maximum liquid elevation inside the tank shows higher values for the liquid metal than for water. Some discrepancies are revealed in CFD results and the differences are quantified. From simulations it is verified that the forces scale with the density ratio, following similar evolution in time. Overall, water is demonstrated to be a valid option as a working liquid in order to evaluate the sloshing

  11. Experimental study and modeling of a high-temperature solar chemical reactor for hydrogen production from methane cracking

    Energy Technology Data Exchange (ETDEWEB)

    Abanades, Stephane; Flamant, Gilles [Processes, Materials, and Solar Energy Laboratory, CNRS (PROMES-CNRS, UPR 8521), 7 Rue du Four Solaire, 66120 Odeillo Font-Romeu (France)

    2007-07-15

    A high-temperature fluid-wall solar reactor was developed for the production of hydrogen from methane cracking. This laboratory-scale reactor features a graphite tubular cavity directly heated by concentrated solar energy, in which the reactive flowing gas dissociates to form hydrogen and carbon black. The solar reactor characterization was achieved with: (a) a thorough experimental study on the reactor performance versus operating conditions and (b) solar reactor modeling. The results showed that the conversion of CH{sub 4} and yield of H{sub 2} can exceed 97% and 90%, respectively, and these depend strongly on temperature and on fluid-wall heat transfer and reaction surface area. In addition to the experimental study, a 2D computational model coupling transport phenomena was developed to predict the mapping of reactor temperature and of species concentration, and the reaction extent at the outlet. The model was validated and kinetics of methane decomposition were identified from simulations and comparison to experimental results. (author)

  12. Formulation and experimental evaluation of closed-form control laws for the rapid maneuvering of reactor neutronic power

    Energy Technology Data Exchange (ETDEWEB)

    Bernard, J.A. (Massachusetts Inst. of Tech., Cambridge, MA (USA). Nuclear Reactor Lab.)

    1989-09-01

    This report describes both the theoretical development and the experimental evaluation of a novel, robust methodology for the time-optimal adjustment of a reactor's neutronic power under conditions of closed-loop digital control. Central to the approach are the MIT-SNL Period-Generated Minimum Time Control Laws' which determine the rate at which reactivity should be changed in order to cause a reactor's neutronic power to conform to a specified trajectory. Using these laws, reactor power can be safely raised by five to seven orders of magnitude in a few seconds. The MIT-SNL laws were developed to facilitate rapid increases of neutronic power on spacecraft reactors operating in an SDI environment. However, these laws are generic and have other applications including the rapid recovery of research and test reactors subsequent to an unanticipated shutdown, power increases following the achievement of criticality on commercial reactors, power adjustments on commercial reactors so as to minimize thermal stress, and automated startups. The work reported here was performed by the Massachusetts Institute of Technology under contract to the Sandia National Laboratories. Support was also provided by the US Department of Energy's Division of University and Industry Programs. The work described in this report is significant in that a novel solution to the problem of time-optimal control of neutronic power was identified, in that a rigorous description of a reactor's dynamics was derived in that the rate of change of reactivity was recognized as the proper control signal, and in that extensive experimental trials were conducted of these newly developed concepts on actual nuclear reactors. 43 refs., 118 figs., 11 tabs.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-08-01

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

  14. Development of gas cooled reactors and experimental setup of high temperature helium loop for in-pile operation

    Energy Technology Data Exchange (ETDEWEB)

    Miletić, Marija, E-mail: marija_miletic@live.com [Czech Technical University in Prague, Prague (Czech Republic); Fukač, Rostislav, E-mail: fuk@cvrez.cz [Research Centre Rez Ltd., Rez (Czech Republic); Pioro, Igor, E-mail: Igor.Pioro@uoit.ca [University of Ontario Institute of Technology, Oshawa (Canada); Dragunov, Alexey, E-mail: Alexey.Dragunov@uoit.ca [University of Ontario Institute of Technology, Oshawa (Canada)

    2014-09-15

    coolants. The purpose of the high temperature helium loop (HTHL) is to simulate technical and chemical conditions of VHTR's coolant. The loop is intended to serve an as experimental device for fatigue and creep tests of construction metallic materials for gas-cooled reactors and it should be also employed for research in field of gaseous coolant chemistry. The loop will serve also for tests of nuclear graphite, dosing and helium purification systems. Because the VHTR is a new reactor concept, major technical uncertainties remain relative to helium-cooled advanced reactor systems. This paper summarizes also the concept of the HTHL in the Research Centre Rez Ltd., its design, utilization and future plans for experimental setup.

  15. Conceptual design of superconducting magnet systems for the Argonne Tokamak Experimental Power Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wang, S.T.; Turner, L.R.; Mills, F.E.; DeMichele, D.W.; Smelser, P.; Kim, S.H.

    1976-01-01

    As an integral effort in the Argonne Tokamak Experimental Power Reactor Conceptual Design, the conceptual design of a 10-tesla, pure-tension superconducting toroidal-field (TF) coil system has been developed in sufficient detail to define a realistic design for the TF coil system that could be built based upon the current state of technology with minimum technological extrapolations. A conceptual design study on the superconducting ohmic-heating (OH) coils and the superconducting equilibrium-field (EF) coils were also completed. These conceptual designs are developed in sufficient detail with clear information on high current ac conductor design, cooling, venting provision, coil structural support and zero loss poloidal coil cryostat design. Also investigated is the EF penetration into the blanket and shield.

  16. A large scale fullerenes synthesis solar reactor modelling and first experimental results

    Energy Technology Data Exchange (ETDEWEB)

    Guillard, T.; Flamand, G.; Robert, J.F.; Rivoire, B.; Olalde, G.; Alvarez, L. [Centre National de la Recherche Scientifique (CNRS-IMP), 66 - Font-Romeu (France); Laplaze, D. [Universite de Montpellier, GDPC, 34 (France)

    1999-03-01

    After the promising results obtained with a 2 kW solar furnace for fullerenes and nano-tubes synthesis, a large scale production project using the 1 MW Odeillo solar furnace started in 1997. This paper presents the first experimental results obtained with a concept-validation vessel and the comparison with a numerical simulation of the target thermal behavior. It is shown that a 6 mm i.d. graphite rod heated by a 500 W/cm{sup 2} incident solar flux density (I{sub s}) reaches a front temperature of 2800 K, in agreement with the thermal model. On this basis, accurate prediction of maximum working temperature of the 1 MW reactor is proposed: 3400 K for I{sub s} = 900 W/cm{sup 2}. (authors)

  17. Review of the International Thermonuclear Experimental Reactor (ITER) detailed design report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-04-18

    Dr. Martha Krebs, Director, Office of Energy Research at the US Department of Energy (DOE), wrote to the Fusion Energy Sciences Advisory Committee (FESAC), in letters dated September 23 and November 6, 1996, requesting that FESAC review the International Thermonuclear Experimental Reactor (ITER) Detailed Design Report (DDR) and provide its view of the adequacy of the DDR as part of the basis for the United States decision to enter negotiations with the other interested Parties regarding the terms and conditions for an agreement for the construction, operations, exploitation and decommissioning of ITER. The letter from Dr. Krebs, referred to as the Charge Letter, provided context for the review and a set of questions of specific interest.

  18. Development plan for the External Hazards Experimental Group. Light Water Reactor Sustainability Program

    Energy Technology Data Exchange (ETDEWEB)

    Coleman, Justin Leigh [Idaho National Lab. (INL), Idaho Falls, ID (United States); Smith, Curtis Lee [Idaho National Lab. (INL), Idaho Falls, ID (United States); Burns, Douglas Edward [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kammerer, Annie [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-03-01

    This report describes the development plan for a new multi-partner External Hazards Experimental Group (EHEG) coordinated by Idaho National Laboratory (INL) within the Risk-Informed Safety Margin Characterization (RISMC) technical pathway of the Light Water Reactor Sustainability Program. Currently, there is limited data available for development and validation of the tools and methods being developed in the RISMC Toolkit. The EHEG is being developed to obtain high-quality, small- and large-scale experimental data validation of RISMC tools and methods in a timely and cost-effective way. The group of universities and national laboratories that will eventually form the EHEG (which is ultimately expected to include both the initial participants and other universities and national laboratories that have been identified) have the expertise and experimental capabilities needed to both obtain and compile existing data archives and perform additional seismic and flooding experiments. The data developed by EHEG will be stored in databases for use within RISMC. These databases will be used to validate the advanced external hazard tools and methods.

  19. Reactor operation

    CERN Document Server

    Shaw, J

    2013-01-01

    Reactor Operation covers the theoretical aspects and design information of nuclear reactors. This book is composed of nine chapters that also consider their control, calibration, and experimentation.The opening chapters present the general problems of reactor operation and the principles of reactor control and operation. The succeeding chapters deal with the instrumentation, start-up, pre-commissioning, and physical experiments of nuclear reactors. The remaining chapters are devoted to the control rod calibrations and temperature coefficient measurements in the reactor. These chapters also exp

  20. Computational and Experimental Investigations of the Coolant Flow in the Cassette Fissile Core of a KLT-40S Reactor

    Science.gov (United States)

    Dmitriev, S. M.; Varentsov, A. V.; Dobrov, A. A.; Doronkov, D. V.; Pronin, A. N.; Sorokin, V. D.; Khrobostov, A. E.

    2017-07-01

    Results of experimental investigations of the local hydrodynamic and mass-exchange characteristics of a coolant flowing through the cells in the characteristic zones of a fuel assembly of a KLT-40S reactor plant downstream of a plate-type spacer grid by the method of diffusion of a gas tracer in the coolant flow with measurement of its velocity by a five-channel pneumometric probe are presented. An analysis of the concentration distribution of the tracer in the coolant flow downstream of a plate-type spacer grid in the fuel assembly of the KLT-40S reactor plant and its velocity field made it possible to obtain a detailed pattern of this flow and to determine its main mechanisms and features. Results of measurement of the hydraulic-resistance coefficient of a plate-type spacer grid depending on the Reynolds number are presented. On the basis of the experimental data obtained, recommendations for improvement of the method of calculating the flow rate of a coolant in the cells of the fissile core of a KLT-40S reactor were developed. The results of investigations of the local hydrodynamic and mass-exchange characteristics of the coolant flow in the fuel assembly of the KLT-40S reactor plant were accepted for estimating the thermal and technical reliability of the fissile cores of KLT-40S reactors and were included in the database for verification of computational hydrodynamics programs (CFD codes).

  1. Analysis of the optimal fuel composition for the Indonesian experimental power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Liem, Peng Hong [Nippon Advanced Information Service (NAIS Co., Inc.), Ibaraki (Japan); Sembiring, Tagor Malem [National Nuclear Energy Agency of Indonesia, Banten (Indonesia). Center for Nuclear Reactor Technology and Safety; Arbie, Bakri; Subki, Iyos [PT MOTAB Technology, Jakarta Barat (Indonesia)

    2017-03-15

    The optimal fuel composition of the 10 MWth Experimental Power Reactor (RDE), to be built by the Indonesian National Nuclear Energy Agency (BATAN), is a very important design parameter since it will directly affect the fuel cost, new and spent fuel storage capacity, and other back-end environmental burden. The RDE is a very small sized pebble-bed high temperature gas-cooled reactor (HTGR) with low enriched uranium (LEU) UO{sub 2} TRISO fuel under multipass or once-through-then-out fueling scheme. A scoping study on fuel composition parameters, namely heavy metal (HM) loading per pebble and uranium enrichment is conducted. All burnup, criticality calculations and core equilibrium search are carried out by using BATAN-MPASS, a general in-core fuel management code for pebble bed HTGRs, featured with many automatic equilibrium searching options as well as thermal-hydraulic calculation capability. The RDE User Requirement Document issued by BATAN is used to derive the main core design parameters and constraints. The scoping study is conducted over uranium enrichment in the range of 10 to 20 w/o and HM loading in the range of 4 g to 10 g/pebble. Fissile loading per unit energy generated (kg/GWd) is taken as the objective function for the present scoping study. The analysis results show that the optimal HM loading is around 8 g/pebble. Under the constraint of 80 GWd/t fuel discharge burnup imposed by the technical specification, the uranium enrichment for the optimal HM loading is approximately 13 w/o.

  2. Thermo-kinetic instabilities in model reactors. Examples in experimental tests

    Science.gov (United States)

    Lavadera, Marco Lubrano; Sorrentino, Giancarlo; Sabia, Pino; de Joannon, Mara; Cavaliere, Antonio; Ragucci, Raffaele

    2017-11-01

    The use of advanced combustion technologies (such as MILD, LTC, etc.) is among the most promising methods to reduce emission of pollutants. For such technologies, working temperatures are enough low to boost the formation of several classes of pollutants, such as NOx and soot. To access this temperature range, a significant dilution as well as preheating of reactants is required. Such conditions are usually achieved by a strong recirculation of exhaust gases that simultaneously dilute and pre-heat the fresh reactants. These peculiar operative conditions also imply strong fuel flexibility, thus allowing the use of low calorific value (LCV) energy carriers with high efficiency. However, the intersection of low combustion temperatures and highly diluted mixtures with intense pre-heating alters the evolution of the combustion process with respect to traditional flames, leading to features such as the susceptibility to oscillations, which are undesirable during combustion. Therefore, an effective use of advanced combustion technologies requires a thorough analysis of the combustion kinetic characteristics in order to identify optimal operating conditions and control strategies with high efficiency and low pollutant emissions. The present work experimentally and numerically characterized the ignition and oxidation processes of methane and propane, highly diluted in nitrogen, at atmospheric pressure, in a Plug Flow Reactor and a Perfectly Stirred Reactor under a wide range of operating conditions involving temperatures, mixture compositions and dilution levels. The attention was focused particularly on the chemistry of oscillatory phenomena and multistage ignitions. The global behavior of these systems can be qualitatively and partially quantitatively modeled using the detailed kinetic models available in the literature. Results suggested that, for diluted conditions and lower adiabatic flame temperatures, the competition among several pathways, i.e. intermediate- and

  3. Basic experiments during loss of vacuum event (LOVE) in fusion experimental reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ogawa, Masuro; Kunugi, Tomoaki; Seki, Yasushi (JAERI, Ibaraki (Japan))

    1993-06-01

    If a loss of vacuum event (LOVE) occurs due to damage of the vacuum vessel of a nuclear fusion experimental reactor, some chemical reactions such as a graphic oxidation and a buoyancy-driven exchange flow take place after equalization of the gas pressure between the inside and outside of the vacuum vessel. The graphite oxidation would generate inflammable carbon monoxide and release tritium retained in the graphite. The exchange flow through the breaches may transport the carbon monoxide and tritium out of the vacuum vessel. To add confidence to the safety evaluations and analyses, it is important to grasp the basic phenomena such as the exchange flow and the graphite oxidation. Experiments of the exchange flow and the graphite oxidation were carried out to obtain the exchange flow rate and the rate constant for the carbon monoxide combustion, respectively. These experimental results were compared with existing correlations. The authors plan a scaled-model test and a full-scale model test for the LOVE.

  4. Experimental and numerical thermal-hydraulics investigation of a molten salt reactor concept core

    Energy Technology Data Exchange (ETDEWEB)

    Yamaji, Bogdan; Aszodi, Attila [Budapest Univ. of Technology and Economics (Hungary). Inst. of Nuclear Techniques

    2017-09-15

    In the paper measurement results of experimental modelling of a molten salt fast reactor concept will be presented and compared with three-dimensional computational fluid dynamics (CFD) simulation results. Purpose of this article is twofold, on one hand to introduce a geometry modification in order to avoid the disadvantages of the original geometry and discuss new measurement results. On the other hand to present an analysis in order to suggest a method of proper numerical modelling of the problem based on the comparison of calculation results and measurement data for the new, modified geometry. The investigated concept has a homogeneous cylindrical core without any internal structures. Previous measurements on the scaled and segmented plexiglas model of the concept core and simulation results have shown that this core geometry could be optimized for better thermal-hydraulics characteristics. In case of the original geometry strong undesired flow separation could develop, that could negatively affect the characteristics of the core from neutronics point of view as well. An internal flow distributor plate was designed and installed with the purpose of optimizing the flow field in the core by enhancing its uniformity. Particle image velocimetry (PIV) measurement results of the modified experimental model will be presented and compared to numerical simulation results with the purpose of CFD model validation.

  5. Experimental and numerical studies of microwave-plasma interaction in a MWPECVD reactor

    Directory of Open Access Journals (Sweden)

    A. Massaro

    2016-12-01

    Full Text Available This work deals with and proposes a simple and compact diagnostic method able to characterize the interaction between microwave and plasma without the necessity of using an external diagnostic tool. The interaction between 2.45 GHz microwave and plasma, in a typical ASTeX-type reactor, is investigated from experimental and numerical view points. The experiments are performed by considering plasmas of three different gas mixtures: H2, CH4-H2 and CH4-H2-N2. The two latter are used to deposit synthetic undoped and n-doped diamond films. The experimental setup equipped with a matching network enables the measurements of very low reflected power. The reflected powers show ripples due to the mismatching between wave and plasma impedance. Specifically, the three types of plasma exhibit reflected power values related to the variation of electron-neutral collision frequency among the species by changing the gas mixture. The different gas mixtures studied are also useful to test the sensitivity of the reflected power measurements to the change of plasma composition. By means of a numerical model, only the interaction of microwave and H2 plasma is examined allowing the estimation of plasma and matching network impedances and of reflected power that is found about eighteen times higher than that measured.

  6. Study of process of water disinfection it saw energy solar using an experimental reactor; Estudo do proceso de desinfeccao de agua via energia solar utilizando um reator experimental

    Energy Technology Data Exchange (ETDEWEB)

    Batista, C. H.; Prado, L. R.; Lima, A. S.; Egues, S. M. S.; Araujo, P. M. M.

    2008-07-01

    In this work, was conducted an experimental study of the efficiency of a solar reactor in the disinfection of drinking water using photolysis (UV) and heterogeneous photo catalysis (TiO{sub 2}/UV). The experiments were conducted in batch mode, evaluating the effects of reactor inclination and the presence of a solar concentrator. The results indicated that the employed system was capable to promote the complete disinfection in 150 min using only the photo thermic effect, and in 120 min with the addition of immobilized TiO{sub 2} and the solar concentrator. (Author)

  7. Estudio del comportamiento de reactores discontinuos y semicontinuos: modelización y comprobación experimental

    OpenAIRE

    Grau Vilalta, Ma. Dolors

    1999-01-01

    L'objectiu primordial d'aquest treball és la comparació entre el funcionament d'un reactor discontinu i un de semicontinu. Per això es porta a terme la modelització matemàtica d'ambdós, utilitzant programes propis emprant el llenguatge Fortran 77, a més del simulador ISIM i el software MATLAB. La validació dels models matemàtics s'efectua, en primer lloc, a partir de dades de la bibliografia. A partir d'aquí, es realitzen proves experimentals en una planta piloto amb un reactor encamisat de v...

  8. Experimental and Analytic Study on the Core Bypass Flow in a Very High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Richard Schultz

    2012-04-01

    Core bypass flow has been one of key issues in the very high temperature reactor (VHTR) design for securing core thermal margins and achieving target temperatures at the core exit. The bypass flow in a prismatic VHTR core occurs through the control element holes and the radial and axial gaps between the graphite blocks for manufacturing and refueling tolerances. These gaps vary with the core life cycles because of the irradiation swelling/shrinkage characteristic of the graphite blocks such as fuel and reflector blocks, which are main components of a core's structure. Thus, the core bypass flow occurs in a complicated multidimensional way. The accurate prediction of this bypass flow and counter-measures to minimize it are thus of major importance in assuring core thermal margins and securing higher core efficiency. Even with this importance, there has not been much effort in quantifying and accurately modeling the effect of the core bypass flow. The main objectives of this project were to generate experimental data for validating the software to be used to calculate the bypass flow in a prismatic VHTR core, validate thermofluid analysis tools and their model improvements, and identify and assess measures for reducing the bypass flow. To achieve these objectives, tasks were defined to (1) design and construct experiments to generate validation data for software analysis tools, (2) determine the experimental conditions and define the measurement requirements and techniques, (3) generate and analyze the experimental data, (4) validate and improve the thermofluid analysis tools, and (5) identify measures to control the bypass flow and assess its performance in the experiment.

  9. Review of nuclear data improvement needs for nuclear radiation measurement techniques used at the CEA experimental reactor facilities

    Directory of Open Access Journals (Sweden)

    Destouches Christophe

    2016-01-01

    Full Text Available The constant improvement of the neutron and gamma calculation codes used in experimental nuclear reactors goes hand in hand with that of the associated nuclear data libraries. The validation of these calculation schemes always requires the confrontation with integral experiments performed in experimental reactors to be completed. Nuclear data of interest, straight as cross sections, or elaborated ones such as reactivity, are always derived from a reaction rate measurement which is the only measurable parameter in a nuclear sensor. So, in order to derive physical parameters from the electric signal of the sensor, one needs specific nuclear data libraries. This paper presents successively the main features of the measurement techniques used in the CEA experimental reactor facilities for the on-line and offline neutron/gamma flux characterizations: reactor dosimetry, neutron flux measurements with miniature fission chambers and Self Power Neutron Detector (SPND and gamma flux measurements with chamber ionization and TLD. For each technique, the nuclear data necessary for their interpretation will be presented, the main identified needs for improvement identified and an analysis of their impact on the quality of the measurement. Finally, a synthesis of the study will be done.

  10. EXPERIMENTAL STUDY OF LOCAL HYDRODYNAMICS AND MASS EXCHANGE PROCESSES OF COOLANT IN FUEL ASSEMBLIES OF PRESSURIZED WATER REACTORS

    Directory of Open Access Journals (Sweden)

    S. M. Dmitriev

    2016-01-01

    Full Text Available The results of experimental studies of local hydrodynamics and mass exchange of coolant flow behind spacer and mixing grids of different structural versions that were developed for fuel assemblies of domestic and foreign nuclear reactors are presented in the article. In order to carry out the study the models of the following fuel assemblies have been fabricated: FA for VVER and VBER, FA-KVADRAT for PWR-reactor and FA for KLT-40C reactor. All the models have been fabricated with a full geometrical similarity with full-scale fuel assemblies. The study was carried out by simulating the flow of coolant in a core by air on an aerodynamic test rig. In order to measure local hydrodynamic characteristics of coolant flow five-channel Pitot probes were used that enable to measure the velocity vector in a point by its three components. The tracerpropane method was used for studying mass transfer processes. Flow hydrodynamics was studied by measuring cross-section velocities of coolant flow and coolant rates according to the model cells. The investigation of mass exchange processes consisted of a study of concentration distribution for tracer in experimental model, in determination of attenuation lengths of mass transfer processes behind mixing grids, in calculating of inter-cellar mass exchange coefficient. The database on coolant flow in fuel assemblies for different types of reactors had been accumulated that formed the basis of the engineering substantiation of reactor cores designs. The recommendations on choice of optimal versions of mixing grids have been taken into consideration by implementers of the JSC “OKBM Afrikantov” when creating commissioned fuel assemblies. The results of the study are used for verification of CFD-codes and CFD programs of detailed cell-by-cell calculation of reactor cores in order to decrease conservatism for substantiation of thermal-mechanical reliability.

  11. Experimental and statistical investigation of thermally induced failure in reactor fuel particles

    Energy Technology Data Exchange (ETDEWEB)

    Lunsford, J.L.; Imprescia, R.J.; Bowman, A.L.; Radosevich, C.E.

    1980-10-01

    An incomplete experimental study into the failure statistics of fuel particle for the high-temperature gas-cooled reactor (HTGR) is described. Fuel particles failure was induced by thermal ramping from room temperature to temperatures in the vicinity of 2273/sup 0/K to 2773/sup 0/K in 2 to 30 h and detected by the appearance of /sup 85/Kr in the helium carrier gas used to sweep the furnace. The concentration of krypton, a beta emitter, was detected by measuring the current that resulted when the helium sweep gas was passed through an ionization chamber. TRISO fuel particles gave a krypton concentration profile as a function of time that built up in several minutes and decayed in a fraction of an hour. This profile, which was temperature independent, was similar to the impulse response of the ionization chamber, suggesting that the TRISO particles failed instantaneously and completely. BISO fuel particles gave a krypton concentration profile as a function of time that built up in a fraction of an hour and decayed in a fraction of a day. This profile was strongly temperature dependent, suggesting that krypton release was diffusion controlled, i.e., that the krypton was diffusing through a sound coat, or that the BISO coating failed but that the krypton was unable to escape the kernel without diffusion, or that a combination of pre- and postfailure diffusion accompanied partial or complete failure.

  12. Calculation of fast neutron flux in reactor pressure tubes and experimental facilities

    Energy Technology Data Exchange (ETDEWEB)

    Barnett, P.C. [Canadian General Electric (Canada)

    1968-07-15

    The computer program EPITHET was used to calculate the fast neutron flux (>1 MeV) in several reactor pressure tubes and experimental facilities in order to compare the fast neutron flux in the different cases and to provide a self-consistent set of flux values which may be used to relate creep strain to fast neutron flux . The facilities considered are shown below together with the calculated fast neutron flux (>1 MeV). Fast flux 10{sup 13} n/cm{sup 2}s: NPD 1.14, Douglas Point 2.66, Pickering 2.89, Gentilly 2.35, SGHWR 3.65, NRU U-1 and U-2 3.25'' pressure tube - 19 element fuel 3.05, NRU U-1 and U-2 4.07'' pressure tube - 28 element fuel 3.18, NRU U-1 and U-2 4.07'' pressure tube - 18 element fuel 2.90, NRX X-5 0.88, PRTR Mk I fuel 2.81, PRTR HPD fuel 3.52, WR-1 2.73, Mk IV creep machine (NRX) 0.85, Mk VI creep machine (NRU) 2.04, Biaxial creep insert (NRU U-49) 2.61.

  13. Analysis of Experimental Data for High Burnup PWR Spent Fuel Isotopic Validation - Vandellos II Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ilas, Germina [ORNL; Gauld, Ian C [ORNL

    2011-01-01

    This report is one of the several recent NUREG/CR reports documenting benchmark-quality radiochemical assay data and the use of the data to validate computer code predictions of isotopic composition for spent nuclear fuel, to establish the uncertainty and bias associated with code predictions. The experimental data analyzed in the current report were acquired from a high-burnup fuel program coordinated by Spanish organizations. The measurements included extensive actinide and fission product data of importance to spent fuel safety applications, including burnup credit, decay heat, and radiation source terms. Six unique spent fuel samples from three uranium oxide fuel rods were analyzed. The fuel rods had a 4.5 wt % {sup 235}U initial enrichment and were irradiated in the Vandellos II pressurized water reactor operated in Spain. The burnups of the fuel samples range from 42 to 78 GWd/MTU. The measurements were used to validate the two-dimensional depletion sequence TRITON in the SCALE computer code system.

  14. Fission product iodine release and retention in nuclear reactor accidents— experimental programme at PSI

    Science.gov (United States)

    Bruchertseifer, H.; Cripps, R.; Guentay, S.; Jaeckel, B.

    2003-01-01

    Iodine radionuclides constitute one of the most important fission products of uranium and plutonium. If the volatile forms would be released into the environment during a severe accident, a potential health hazard would then ensue. Understanding its behaviour is an important prerequisite for planning appropriate mitigation measures. Improved and extensive knowledge of the main iodine species and their reactions important for the release and retention processes in the reactor containment is thus mandatory. The aim of PSI's radiolytical studies is to improve the current thermodynamic and kinetic databases and the models for iodine used in severe accident computer codes. Formation of sparingly soluble silver iodide (AgI) in a PWR containment sump can substantially reduce volatile iodine fraction in the containment atmosphere. However, the effectiveness is dependent on its radiation stability. The direct radiolytic decomposition of AgI and the effect of impurities on iodine volatilisation were experimentally determined at PSI using a remote-controlled and automated high activity 188W/Re generator (40 GBq/ml). Low molecular weight organic iodides are difficult to be retained in engineered safety systems. Investigation of radiolytic decomposition of methyl iodide in aqueous solutions, combined with an on-line analysis of iodine species is currently under investigation at PSI.

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

    Energy Technology Data Exchange (ETDEWEB)

    Stekly, Z.J.J.; Lucas, E.J. (eds.)

    1976-12-01

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

  16. Bulk-bronzied graphites for plasma-facing components in ITER (International Thermonuclear Experimental Reactor)

    Energy Technology Data Exchange (ETDEWEB)

    Hirooka, Y.; Conn, R.W.; Doerner, R.; Khandagle, M. (California Univ., Los Angeles, CA (USA). Inst. of Plasma and Fusion Research); Causey, R.; Wilson, K. (Sandia National Labs., Livermore, CA (USA)); Croessmann, D.; Whitley, J. (Sandia National Labs., Albuquerque, NM (USA)); Holland, D.; Smolik, G. (Idaho National Engineering Lab., Idaho Falls, ID (USA)); Matsuda, T.; Sogabe, T. (Toyo Tanso Co. Ltd., O

    1990-06-01

    Newly developed bulk-boronized graphites and boronized C-C composites with a total boron concentration ranging from 1 wt % to 30 wt % have been evaluated as plasma-facing component materials for the International Thermonuclear Experimental Reactor (ITER). Bulk-boronized graphites have been bombarded with high-flux deuterium plasmas at temperatures between 200 and 1600{degree}C. Plasma interaction induced erosion of bulk-boronized graphites is observed to be a factor of 2--3 smaller than that of pyrolytic graphite, in regimes of physical sputtering, chemical sputtering and radiation enhanced sublimation. Postbombardment thermal desorption spectroscopy indicates that bulk-boronized graphites enhance recombinative desorption of deuterium, which leads to a suppression of the formation of deuterocarbon due to chemical sputtering. The tritium inventory in graphite has been found to decrease by an order of magnitude due to 10 wt % bulk-boronization at temperatures above 1000{degree}C. The critical heat flux to induce cracking for bulk-boronized graphites has been found to be essentially the same as that for non-boronized graphites. Also, 10 wt % bulk-boronization of graphite hinders air oxidation nearly completely at 800{degree}C and reduces the steam oxidation rate by a factor of 2--3 at around 1100 and 1350{degree}C. 38 refs., 5 figs.

  17. Economic impacts on the United States of siting decisions for the international thermonuclear experimental reactor

    Energy Technology Data Exchange (ETDEWEB)

    Peerenboom, J.P.; Hanson, M.E.; Huddleston, J.R. [and others

    1996-08-01

    This report presents the results of a study that examines and compares the probable short-term economic impacts of the International Thermonuclear Experimental Reactor (ITER) on the United States (U.S.) if (1) ITER were to be sited in the U.S., or (2) ITER were to be sited in one of the other countries that, along with the U.S., is currently participating in the ITER program. Life-cycle costs associated with ITER construction, operation, and decommissioning are analyzed to assess their economic impact. A number of possible U.S. host and U.S. non-host technology and cost-sharing arrangements with the other ITER Parties are examined, although cost-sharing arrangements and the process by which the Parties will select a host country and an ITER site remain open issues. Both national and local/regional economic impacts, as measured by gross domestic product, regional output, employment, net exports, and income, are considered. These impacts represent a portion of the complex, interrelated set of economic considerations that characterize U.S. host and U.S. non-host participation in ITER. A number of other potentially important economic and noneconomic considerations are discussed qualitatively.

  18. An experimental study on coolability through the external reactor vessel cooling according to RPV insulation design

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Kyoung Ho; Koo, Kil Mo; Park, Rae Joon; Cho, Young Ro; Kim, Sang Baik

    2004-01-01

    LAVA-ERVC experiments have been performed to investigate the effect of insulation design features on the water accessibility and coolability in case of the external reactor vessel cooling. Alumina iron thermite melt was used as corium stimulant. And the hemispherical test vessel is linearly scaled-down of RPV lower plenum. 4 tests have been performed varying the melt composition and the configuration of the insulation system. Due to the limited steam venting capacity through the insulation, steam binding occurred inside the annulus in the LAVA- ERVC-1, 2 tests which were performed for simulating the KSNP insulation design. This steam binding brought about incident heat up of the vessel outer surface at the upper part in the LAVA-ERVC-1, 2 tests. On the contrary, in the LAVA-ERVC-3, 4 tests which were performed for simulating the APR1400 insulation design, the temperatures of the vessel outer surface maintained near saturation temperature. Sufficient water ingression and steam venting through the insulation lead to effective cooldown of the vessel characterized by nucleate boiling in the LAVA-ERVC-3, 4 tests. From the LAVA-ERVC experimental results, it could be preliminarily concluded that if pertinent modification of the insulation design focused on the improvement of water ingression and steam venting should be preceded the possibility of in-vessel corium retention through the external vessel cooling could be considerably increased.

  19. A statistical experimental design to remove sulfate by crystallization in a fluidized-bed reactor

    Directory of Open Access Journals (Sweden)

    Mark Daniel G. de Luna

    2017-05-01

    Full Text Available This study used crystallization in a fluidized-bed reactor as an alternative technology to the conventional chemical precipitation to remove sulfate. The Box-Behnken Design was used to study the effects and interactions of seed dosage of synthetic gypsum, initial sulfate concentration and molar ratio of calcium to sulfate on conversion and removal of sulfate. The optimum conditions of conversion and removal of sulfate were determined and used to treat the simulated acid mine drainage (AMD wastewater. The effect of inorganic ions CO32−, NH4+ and Al3+ on sulfate conversion was also investigated. Experimental results indicated that seed dosage, initial sulfate concentration and molar ratio of calcium to sulfate are all significant parameters in the sulfate removal by fluidized-bed crystallization. The optimal conditions of 4 g seed L−1, 119.7 mM of initial sulfate concentration and [Ca2+]/[SO42−] molar ratio of 1.48 resulted in sulfate conversion of 82% and sulfate removal of 67%. Conversion and removal of sulfate in the simulated AMD wastewater were 79 and 63%, respectively. When ammonium or aluminum was added to the synthetic sulfate wastewater, significant conversion of sulfate was achieved.

  20. Economic Impacts on the United States of Siting Decisions for the International Thermonuclear Experimental Reactor

    Science.gov (United States)

    Peerenboom, J. P.; Hanson, M. E.; Huddleston, J. R.; Wolsko, T. D.

    1997-12-01

    This paper presents the results of a study that examines and compares the probable short-term economic impacts of the International Thermonuclear Experimental Reactor (ITER) on the United States (U.S.) if (1) ITER were to be sited in the U.S., or (2) ITER were to be sited in one of the other countries that, along with the U.S., is currently participating in the ITER program. Life-cycle costs associated with ITER construction, operation, and decommissioning are analyzed to assess their economic impact. A number of possible U.S. host and U.S. non-host technology and cost-sharing arrangements with the other ITER Parties are examined, although cost-sharing arrangements and the process by which the Parties will select a host country and an ITER site remain open issues. Both national and local/regional economic impacts, as measured by gross domestic product, regional output, employment, net exports, and income, are considered. These impacts represent a portion of the complex, interrelated set of economic considerations that characterize U.S. host and U.S. non-host participation in ITER. A number of other potentially important economic and noneconomic considerations are discussed qualitatively.

  1. Conceptual studies of toroidal field magnets for the tokamak (fusion) experimental power reactor. Final report

    Energy Technology Data Exchange (ETDEWEB)

    1976-11-01

    This report presents the results of ''Conceptual Studies of Toroidal Field Magnets for the Tokamak Experimental Power Reactor'' performed for the Energy Research and Development Administration, Oak Ridge Operations. Two conceptual coil designs are developed. One design approach to produce a specified 8 Tesla maximum field uses a novel NbTi superconductor design cooled by pool-boiling liquid helium. For a highest practicable field design, a unique NbSn/sub 3/ conductor is used with forced-flow, single-phase liquid helium cooling to achieve a 12 Tesla peak field. Fabrication requirements are also developed for these approximately 7 meter horizontal bore by 11 meter vertical bore coils. Cryostat design approaches are analyzed and a hybrid cryostat approach selected. Structural analyses are performed for approaches to support in-plane and out-of-plane loads and a structural approach selected. In addition to the conceptual design studies, cost estimates and schedules are prepared for each of the design approaches, major uncertainties and recommendations for research and development identified, and test coil size for demonstration recommended.

  2. Transmission thermography for inspecting the busbar insulation layer in thermonuclear experimental reactor

    Science.gov (United States)

    Chen, Dapeng; Zhang, Guang; Zhang, Xiaolong; Zeng, Zhi

    2014-11-01

    In Thermonuclear Experimental Reactor, Superconducting Busbar is used for current transmission between magnet coils and current leads. The work temperature of the Busbar is about 4K because of liquid helium via inside. The large temperature grad from 300K to 4K could lead to the defects and damages occur on the insulation layer, which is made of glass fiber and polyimide and has a big different thermal expansion coefficient compared with the metal inner cylinder. This paper aims at developing an infrared transmission non-destructive evaluation (NDE) method for inspecting the insulation layer of Superconducting Busbar; theoretical model of transient heat conduction under a continuous inner heat source for cylindrical structure is described in the paper; a Busbar specimen which is designed with three delamination defects of different depths is heated inside by pouring hot water and monitored by an infrared detector located outside. Results demonstrate excellent detection performance for delamination defects in the insulation layer by using transmission thermography, all of the three defects of different depths can be visualized clearly in the thermal images, and the deeper defect has a better signal contrast, which is also shown in the temperature difference between defects and sound area vs. time curves. The results of light pulse thermography is also shown as a comparison, and it is found that the thermal images obtained by the transmission thermography has a much better signal contrast than that of the pulse thermography. In order to verify the experiments, finite element method is applied to simulate the heat conduction in the Busbar under the continuous inside heating, and it is found that the simulated temperature vs. time and simulated temperature difference vs. time curves are basically coincident with the experimental results. In addition, the possibility of in-service inspection for Busbar insulation layer in ITER item is discussed.

  3. A novel reverse flow reactor coupling endothermic and exothermic reactions: an experimental study

    NARCIS (Netherlands)

    van Sint Annaland, M.; Nijssen, R.C.

    2002-01-01

    A new reactor concept is studied for highly endothermic heterogeneously catalysed gas phase reactions at high temperatures with rapid but reversible catalyst deactivation. The reactor concept aims to achieve an indirect coupling of energy necessary for endothermic reactions and energy released by

  4. Experimental and numerical stability investigations on natural circulation boiling water reactors

    CERN Document Server

    Marcel, CP

    2007-01-01

    In the design of novel nuclear reactors active systems are replaced by passive ones in order to reduce the risk of failure. For that reason natural circulation is being considered as the primary cooling mechanism in next generation nuclear reactor designs

  5. Design considerations and experimental observations for the TAMU air-cooled reactor cavity cooling system for the VHTR

    Energy Technology Data Exchange (ETDEWEB)

    Sulaiman, S. A., E-mail: shamsulamri@tamu.edu; Dominguez-Ontiveros, E. E., E-mail: elvisdom@tamu.edu; Alhashimi, T., E-mail: jbudd123@tamu.edu; Budd, J. L., E-mail: dubaiboy@tamu.edu; Matos, M. D., E-mail: mailgoeshere@gmail.com; Hassan, Y. A., E-mail: yhasssan@tamu.edu [Department of Nuclear Engineering, Texas A and M University, College Station, TX, 77843-3133 (United States)

    2015-04-29

    The Reactor Cavity Cooling System (RCCS) is a promising passive decay heat removal system for the Very High Temperature Reactor (VHTR) to ensure reliability of the transfer of the core residual and decay heat to the environment under all off-normal circumstances. A small scale experimental test facility was constructed at Texas A and M University (TAMU) to study pertinent multifaceted thermal hydraulic phenomena in the air-cooled reactor cavity cooling system (RCCS) design based on the General Atomics (GA) concept for the Modular High Temperature Gas-Cooled Reactor (MHTGR). The TAMU Air-Cooled Experimental Test Facility is ⅛ scale from the proposed GA-MHTGR design. Groundwork for experimental investigations focusing into the complex turbulence mixing flow behavior inside the upper plenum is currently underway. The following paper illustrates some of the chief design considerations used in construction of the experimental test facility, complete with an outline of the planned instrumentation and data acquisition methods. Computational Fluid Dynamics (CFD) simulations were carried out to furnish some insights on the overall behavior of the air flow in the system. CFD simulations assisted the placement of the flow measurement sensors location. Preliminary experimental observations of experiments at 120oC inlet temperature suggested the presence of flow reversal for cases involving single active riser at both 5 m/s and 2.25 m/s, respectively and four active risers at 2.25 m/s. Flow reversal may lead to thermal stratification inside the upper plenum by means of steady state temperature measurements. A Particle Image Velocimetry (PIV) experiment was carried out to furnish some insight on flow patterns and directions.

  6. Design considerations and experimental observations for the TAMU air-cooled reactor cavity cooling system for the VHTR

    Science.gov (United States)

    Sulaiman, S. A.; Dominguez-Ontiveros, E. E.; Alhashimi, T.; Budd, J. L.; Matos, M. D.; Hassan, Y. A.

    2015-04-01

    The Reactor Cavity Cooling System (RCCS) is a promising passive decay heat removal system for the Very High Temperature Reactor (VHTR) to ensure reliability of the transfer of the core residual and decay heat to the environment under all off-normal circumstances. A small scale experimental test facility was constructed at Texas A&M University (TAMU) to study pertinent multifaceted thermal hydraulic phenomena in the air-cooled reactor cavity cooling system (RCCS) design based on the General Atomics (GA) concept for the Modular High Temperature Gas-Cooled Reactor (MHTGR). The TAMU Air-Cooled Experimental Test Facility is ⅛ scale from the proposed GA-MHTGR design. Groundwork for experimental investigations focusing into the complex turbulence mixing flow behavior inside the upper plenum is currently underway. The following paper illustrates some of the chief design considerations used in construction of the experimental test facility, complete with an outline of the planned instrumentation and data acquisition methods. Computational Fluid Dynamics (CFD) simulations were carried out to furnish some insights on the overall behavior of the air flow in the system. CFD simulations assisted the placement of the flow measurement sensors location. Preliminary experimental observations of experiments at 120oC inlet temperature suggested the presence of flow reversal for cases involving single active riser at both 5 m/s and 2.25 m/s, respectively and four active risers at 2.25 m/s. Flow reversal may lead to thermal stratification inside the upper plenum by means of steady state temperature measurements. A Particle Image Velocimetry (PIV) experiment was carried out to furnish some insight on flow patterns and directions.

  7. Hydrocarbon pyrolysis reactor experimentation and modeling for the production of solar absorbing carbon nanoparticles

    Science.gov (United States)

    Frederickson, Lee Thomas

    Much of combustion research focuses on reducing soot particulates in emissions. However, current research at San Diego State University (SDSU) Combustion and Solar Energy Laboratory (CSEL) is underway to develop a high temperature solar receiver which will utilize carbon nanoparticles as a solar absorption medium. To produce carbon nanoparticles for the small particle heat exchange receiver (SPHER), a lab-scale carbon particle generator (CPG) has been built and tested. The CPG is a heated ceramic tube reactor with a set point wall temperature of 1100-1300°C operating at 5-6 bar pressure. Natural gas and nitrogen are fed to the CPG where natural gas undergoes pyrolysis resulting in carbon particles. The gas-particle mixture is met downstream with dilution air and sent to the lab scale solar receiver. To predict soot yield and general trends in CPG performance, a model has been setup in Reaction Design CHEMKIN-PRO software. One of the primary goals of this research is to accurately measure particle properties. Mean particle diameter, size distribution, and index of refraction are calculated using Scanning Electron Microscopy (SEM) and a Diesel Particulate Scatterometer (DPS). Filter samples taken during experimentation are analyzed to obtain a particle size distribution with SEM images processed in ImageJ software. These results are compared with the DPS, which calculates the particle size distribution and the index of refraction from light scattering using Mie theory. For testing with the lab scale receiver, a particle diameter range of 200-500 nm is desired. Test conditions are varied to understand effects of operating parameters on particle size and the ability to obtain the size range. Analysis of particle loading is the other important metric for this research. Particle loading is measured downstream of the CPG outlet and dilution air mixing point. The air-particle mixture flows through an extinction tube where opacity of the mixture is measured with a 532 nm

  8. Parametric experimental tests of steam gasification of pine wood in a fluidized bed reactor

    Directory of Open Access Journals (Sweden)

    L. Vecchione

    2013-09-01

    Full Text Available Among Renewable Energy Sources (RES, biomass represent one of the most common and suitable solution in order to contribute to the global energy supply and to reduce greenhouse gases (GHG emissions. The disposal of some residual biomass, as pruning from pine trees, represent a problem for agricultural and agro-industrial sectors. But if the residual biomass are used for energy production can become a resource. The most suitable energy conversion technology for the above-mentioned biomass is gasification process because the high C/N ratio and the low moisture content, obtained from the analysis. In this work a small-pilot bubbling-bed gasification plant has been designed, constructed and used in order to obtain, from the pine trees pruning, a syngas with low tar and char contents and high hydrogen content. The activities showed here are part of the activities carried out in the European 7FP UNIfHY project. In particular the aim of this work is to develop experimental test on a bench scale steam blown fluidized bed biomass gasifier. These tests will be utilized in future works for the simulations of a pilot scale steam fluidized bed gasifier (100 kWth fed with different biomass feedstock. The results of the tests include produced gas and tar composition as well gas, tar and char yield. Tests on a bench scale reactor (8 cm I.D. were carried out varying steam to biomass ratio from 0.5, 0.7 and 1 to 830°C.

  9. Fast reactor: an experimental study of thermohydraulic processes in different operating regimes

    Science.gov (United States)

    Opanasenko, A. N.; Sorokin, A. P.; Zaryugin, D. G.; Trufanov, A. A.

    2017-05-01

    Results of integrated water model studies of temperature fields and a flow pattern of a nonisothermal primary coolant in the elements of the fast neutron reactor (hereinafter, fast reactor) primary circuit with primary sodium in different regimes, such as forced circulation (FC), transition to the reactor cooldown and emergency cooldown with natural coolant convection, are presented. It is shown that, under the influence of lift forces on the nonisothermal coolant flow in the upper chamber at the periphery of its bottom region over the side shields, a stable cold coolant isothermal zone is formed, whose dimensions increase with increase of total water flowrate. An essential and stable coolant temperature stratification is detected in the peripheral area of the upper (hot) chamber over the side shields, in the pressure and cold side chambers, in the elevator baffle, in the cooling system of the reactor vessel, and in the outlet of intermediate and autonomous heat exchangers in different operating regimes. Large gradients and temperature fluctuations are registered at the interface of stratified and recycling formations. In all of the studied cooldown versions, the coolant outlet temperature at the core fuel assembly is decreased and the coolant temperature in the peripheral zone of the upper chamber is increased compared to the FC. High performance of a passive emergency cooldown system of a fast reactor (BN-1200) with submersible autonomous heat exchangers (AHE) is confirmed. Thus, in a normal operation regime, even in case of malfunction of three submersible AHEs, the temperature of the equipment inside the reactor remains within acceptable limits and decay heat removal from the reactor does not exceed safe operation limits. The obtained results can be used both for computer code verification and for approximate estimate of the reactor plant parameters on the similarity criteria basis.

  10. Experimental evidences of95 mTc production in a nuclear reactor.

    Science.gov (United States)

    Cohen, I M; Robles, A; Mendoza, P; Airas, R M; Montoya, E H

    2018-02-02

    95 m Tc has been identified as by-product in some solutions of 99 m Tc obtained by irradiation of molybdenum trioxide in a reactor neutron flux. The characterization was carried out using both measurements by gamma spectrometry and half-life determination. The possible ways that lead to the 95 m Tc production in a nuclear reactor are discussed. Copyright © 2018. Published by Elsevier Ltd.

  11. Experimental evaluation of gamma fluence-rate predictions from Argon-41 releases to the atmosphere over a nuclear research reactor site

    DEFF Research Database (Denmark)

    Rojas-Palma, C.; Aage, H.K.; Astrup, P.

    2004-01-01

    An experimental study of radionuclide dispersion in the atmosphere has been conducted at the BR1 research reactor in Mol, Belgium. Artificially generated aerosols ('white smoke') were mixed with the routine releases of Ar-41 in the reactor's 60-m tall venting stack. The detailed plume geometry...

  12. Analysis and Experimental Qualification of an Irradiation Capsule Design for Testing Pressurized Water Reactor Fuel Cladding in the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Kurt R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Daily, Charles R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Petrie, Christian M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-09-01

    The Advanced Fuels Campaign within the Fuel Cycle Research and Development program of the Department of Energy Office of Nuclear Energy is currently investigating a number of advanced nuclear fuel cladding concepts to improve the accident tolerance of light water reactors. Alumina-forming ferritic alloys (e.g., FeCrAl) are some of the leading candidates to replace traditional zirconium alloys due to their superior oxidation resistance, provided no prohibitive irradiation-induced embrittlement occurs. Oak Ridge National Laboratory has developed experimental designs to irradiate thin-walled cladding tubes with representative pressurized water reactor geometry in the High Flux Isotope Reactor (HFIR) under relevant temperatures. These designs allow for post-irradiation examination (PIE) of cladding that closely resembles expected commercially viable geometries and microstructures. The experiments were designed using relatively inexpensive rabbit capsules for the irradiation vehicle. The simplistic designs combined with the extremely high neutron flux in the HFIR allow for rapid testing of a large test matrix, thus reducing the time and cost needed to advanced cladding materials closer to commercialization. The designs are flexible in that they allow for testing FeCrAl alloys, stainless steels, Inconel alloys, and zirconium alloys (as a reference material) both with and without hydrides. This will allow a direct comparison of the irradiation performance of advanced cladding materials with traditional zirconium alloys. PIE will include studies of dimensional change, microstructure variation, mechanical performance, etc. This work describes the capsule design, neutronic and thermal analyses, and flow testing that were performed to support the qualification of this new irradiation vehicle.

  13. Experimental and Thermalhydraulic Code Assessment of the Transient Behavior of the Passive Condenser System in an Advanced Boiling Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S.T. Revankar; W. Zhou; Gavin Henderson

    2008-07-08

    The main goal of the project was to study analytically and experimentally the condensation heat transfer for the passive condenser system such as GE Economic Simplified Boiling Water Reactor (ESBWR). The effect of noncondensable gas in condenser tube and the reduction of secondary pool water level to the condensation heat transfer coefficient was the main focus in this research. The objectives of this research were to : 1) obtain experimental data on the local and tube averaged condensation heat transfer rates for the PCCS with non-condensable and with change in the secondary pool water, 2) assess the RELAP5 and TRACE computer code against the experimental data, and 3) develop mathematical model and ehat transfer correlation for the condensation phenomena for system code application. The project involves experimentation, theoretical model development and verification, and thermal- hydraulic codes assessment.

  14. Experimental and MCNP5 based evaluation of neutron and gamma flux in the irradiation ports of the University of Utah research reactor

    Directory of Open Access Journals (Sweden)

    Noble Brooklyn

    2012-01-01

    Full Text Available Neutron and gamma flux environment of various irradiation ports in the University of Utah training, research, isotope production, general atomics reactor were experimentally assessed and fully modeled using the MCNP5 code. The experimental measurements were based on the cadmium ratio in the irradiation ports of the reactor, flux profiling using nickel wire, and gamma dose measurements using thermo luminescence dosimeter. Full 3-D MCNP5 reactor model was developed to obtain the neutron flux distributions of the entire reactor core and to compare it with the measured flux focusing at the irradiation ports. Integration of all these analysis provided the updated comprehensive neutron-gamma flux maps of the existing irradiation facilities of the University of Utah TRIGA reactor.

  15. Evaluation of integral continuing experimental capability (CEC) concepts for light water reactor research: PWR scaling concepts

    Energy Technology Data Exchange (ETDEWEB)

    Condie, K G; Larson, T K; Davis, C B; McCreery, G E

    1987-02-01

    In this report reactor transients and thermal-hydraulic phenomena of importance (based on probabilistic risk assessment and the International Code Assessment Program) to reactor safety were examined and identified. Established scaling methodologies were used to develop potential concepts for integral thermal-hydraulic testing facilities. Advantages and disadvantages of each concept are evaluated. Analysis is conducted to examine the scaling of various phenomena in each of the selected concepts. Results generally suggest that a facility capable of operating at typical reactor operating conditions will scale most phenomena reasonably well. Although many phenomena in facilities using Freon or water at nontypical pressure will scale reasonably well, those phenomena that are heavily dependent on quality (heat transfer or critical flow for example) can be distorted. Furthermore, relation of data produced in facilities operating with nontypical fluids or at nontypical pressures to large plants will be a difficult and time consuming process.

  16. Verification and Validation of the PLTEMP/ANL Code for Thermal-Hydraulic Analysis of Experimental and Test Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kalimullah, M. [Argonne National Lab. (ANL), Argonne, IL (United States); Olson, Arne P. [Argonne National Lab. (ANL), Argonne, IL (United States); Feldman, E. E. [Argonne National Lab. (ANL), Argonne, IL (United States); Hanan, N. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-04-07

    The document compiles in a single volume several verification and validation works done for the PLTEMP/ANL code during the years of its development and improvement. Some works that are available in the open literature are simply referenced at the outset, and are not included in the document. PLTEMP has been used in conversion safety analysis reports of several US and foreign research reactors that have been licensed and converted. A list of such reactors is given. Each chapter of the document deals with the verification or validation of a specific model. The model verification is usually done by comparing the code with hand calculation, Microsoft spreadsheet calculation, or Mathematica calculation. The model validation is done by comparing the code with experimental data or a more validated code like the RELAP5 code.

  17. Screening of Gas-Cooled Reactor Thermal-Hydraulic and Safety Analysis Tools and Experimental Database

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Won Jae; Kim, Min Hwan; Lee, Seung Wook (and others)

    2007-08-15

    This report is a final report of I-NERI Project, 'Screening of Gas-cooled Reactor Thermal Hydraulic and Safety Analysis Tools and Experimental Database 'jointly carried out by KAERI, ANL and INL. In this study, we developed the basic technologies required to develop and validate the VHTR TH/safety analysis tools and evaluated the TH/safety database information. The research tasks consist of; 1) code qualification methodology (INL), 2) high-level PIRTs for major nucleus set of events (KAERI, ANL, INL), 3) initial scaling and scoping analysis (ANL, KAERI, INL), 4) filtering of TH/safety tools (KAERI, INL), 5) evaluation of TH/safety database information (KAERI, INL, ANL) and 6) key scoping analysis (KAERI). The code qualification methodology identifies the role of PIRTs in the R and D process and the bottom-up and top-down code validation methods. Since the design of VHTR is still evolving, we generated the high-level PIRTs referencing 600MWth block-type GT-MHR and 400MWth pebble-type PBMR. Nucleus set of events that represents the VHTR safety and operational transients consists of the enveloping scenarios of HPCC (high pressure conduction cooling: loss of primary flow), LPCC/Air-Ingress (low pressure conduction cooling: loss of coolant), LC (load changes: power maneuvering), ATWS (anticipated transients without scram: reactivity insertion), WS (water ingress: water-interfacing system break) and HU (hydrogen-side upset: loss of heat sink). The initial scaling analysis defines dimensionless parameters that need to be reflected in mixed convection modeling and the initial scoping analysis provided the reference system transients used in the PIRTs generation. For the PIRTs phenomena, we evaluated the modeling capability of the candidate TH/safety tools and derived a model improvement need. By surveying and evaluating the TH/safety database information, a tools V and V matrix has been developed. Through the key scoping analysis using available database, the

  18. Experimental investigation into fast pyrolysis of biomass using an entrained flow reactor

    Science.gov (United States)

    Bohn, M.; Benham, C.

    1981-02-01

    Pyrolysis experiments were performed with steam as a carrier gas and two different feedstocks - wheat straw and powdered material derived from municipal solid waste (ECO-II TM). Reactor wall temperature was varied from 7000 to 1400 C. Gas composition data from the ECO-II tests were comparable to previously reported data but ethylene yield appeared to vary with reactor wall temperature and residence time. The important conclusion from the wheat straw tests is that olefin yields are about one half that obtained from ECO-II. Evidence was found that high olefin yields from ECO-II are due to the presence of plastics in the feedstock.

  19. Evaluation of activation detectors for the SPHINX project at the LR-0 experimental reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lahodova, Zdena; Viererbl, Ladislav [Research Center Rez Ltd (Czech Republic); Novak, Evzen; Svadlenkova, Marie; Rypar, Vojtech [Nuclear Power and Safety Division, Nuclear Research Institute Rez plc (Czech Republic)

    2008-07-01

    This article summarizes the measurements of neutron fluence distributions carried out at the LR-0 research reactor (Czech Republic) in the frame of the SPHINX project. The influence of fluoride-salts or graphite filling in the SR-0 modules on neutron spectrum was studied using activation detectors. The activation detectors (Mn, Ni, In and Au) were evaluated to determine the changes in neutron field. The In and Au detectors were also irradiated with a cadmium cover. Five different configurations of reactor core (EROS) were realized. (authors)

  20. Experimental evaluation of methane dry reforming process on a membrane reactor to hydrogen production

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Fabiano S.A.; Benachour, Mohand; Abreu, Cesar A.M. [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Dept. of Chemical Engineering], Email: f.aruda@yahoo.com.br

    2010-07-01

    In a fixed bed membrane reactor evaluations of methane-carbon dioxide reforming over a Ni/{gamma}- Al{sub 2}O{sub 3} catalyst were performed at 773 K, 823 K and 873 K. A to convert natural gas into syngas a fixed-bed reactor associate with a selective membrane was employed, where the operating procedures allowed to shift the chemical equilibrium of the reaction in the direction of the products of the process. Operations under hydrogen permeation, at 873 K, promoted the increase of methane conversion, circa 83%, and doubled the yield of hydrogen production, when compared with operations where no hydrogen permeation occurred. (author)

  1. Hydrogen production by reforming of liquid hydrocarbons in a membrane reactor for portable power generation-Experimental studies

    Science.gov (United States)

    Damle, Ashok S.

    One of the most promising technologies for lightweight, compact, portable power generation is proton exchange membrane (PEM) fuel cells. PEM fuel cells, however, require a source of pure hydrogen. Steam reforming of hydrocarbons in an integrated membrane reactor has potential to provide pure hydrogen in a compact system. Continuous separation of product hydrogen from the reforming gas mixture is expected to increase the yield of hydrogen significantly as predicted by model simulations. In the laboratory-scale experimental studies reported here steam reforming of liquid hydrocarbon fuels, butane, methanol and Clearlite ® was conducted to produce pure hydrogen in a single step membrane reformer using commercially available Pd-Ag foil membranes and reforming/WGS catalysts. All of the experimental results demonstrated increase in hydrocarbon conversion due to hydrogen separation when compared with the hydrocarbon conversion without any hydrogen separation. Increase in hydrogen recovery was also shown to result in corresponding increase in hydrocarbon conversion in these studies demonstrating the basic concept. The experiments also provided insight into the effect of individual variables such as pressure, temperature, gas space velocity, and steam to carbon ratio. Steam reforming of butane was found to be limited by reaction kinetics for the experimental conditions used: catalysts used, average gas space velocity, and the reactor characteristics of surface area to volume ratio. Steam reforming of methanol in the presence of only WGS catalyst on the other hand indicated that the membrane reactor performance was limited by membrane permeation, especially at lower temperatures and lower feed pressures due to slower reconstitution of CO and H 2 into methane thus maintaining high hydrogen partial pressures in the reacting gas mixture. The limited amount of data collected with steam reforming of Clearlite ® indicated very good match between theoretical predictions and

  2. Students' Assessment of Interactive Distance Experimentation in Nuclear Reactor Physics Laboratory Education

    Science.gov (United States)

    Malkawi, Salaheddin; Al-Araidah, Omar

    2013-01-01

    Laboratory experiments develop students' skills in dealing with laboratory instruments and physical processes with the objective of reinforcing the understanding of the investigated subject. In nuclear engineering, where research reactors play a vital role in the practical education of students, the high cost and long construction time of research…

  3. Experimental and numerical stability investigations on natural circulation boiling water reactors

    NARCIS (Netherlands)

    Marcel, C.P.

    2007-01-01

    The stability of natural circulation boiling water reactors is investigated with a strong emphasis on experiments. Two different facilities are used for such a task: the GENESIS facility (to which a void reactivity feedback system is artificially added) and the CIRCUS facility. In addition,

  4. Experimental computer-controlled instrumentation system for the research reactor DR2

    DEFF Research Database (Denmark)

    Goodstein, L.P.

    1969-01-01

    An instrumentation system has been developed for one of the Danish Atomic Energy Commission's research reactors as part of an experiment on the advantages to be gained by the use of digital computers in a process plant application. Problem areas to be investigated include (a) reliability and safety...

  5. Experimental analysis of arsenic precipitation during microbial sulfate and iron reduction in model aquifer sediment reactors

    Science.gov (United States)

    Kirk, Matthew F.; Roden, Eric E.; Crossey, Laura J.; Brealey, Adrian J.; Spilde, Michael N.

    2010-05-01

    Microbial SO 42- reduction limits accumulation of aqueous As in reducing aquifers where the sulfide that is produced forms minerals that sequester As. We examined the potential for As partitioning into As- and Fe-sulfide minerals in anaerobic, semi-continuous flow bioreactors inoculated with 0.5% (g mL -1) fine-grained alluvial aquifer sediment. A fluid residence time of three weeks was maintained over a ca. 300-d incubation period by replacing one-third of the aqueous phase volume of the reactors with fresh medium every seven days. The medium had a composition comparable to natural As-contaminated groundwater with slightly basic pH (7.3) and 7.5 μM aqueous As(V) and also contained 0.8 mM acetate to stimulate microbial activity. Medium was delivered to a reactor system with and without 10 mmol L -1 synthetic goethite (α-FeOOH). In both reactors, influent As(V) was almost completely reduced to As(III). Pure As-sulfide minerals did not form in the Fe-limited reactor. Realgar (As 4S 4) and As 2S 3(am) were undersaturated throughout the experiment. Orpiment (As 2S 3) was saturated while sulfide content was low (˜50 to 150 μM), but precipitation was likely limited by slow kinetics. Reaction-path modeling suggests that, even if these minerals had formed, the dissolved As content of the reactor would have remained at hazardous levels. Mackinawite (Fe 1 + xS; x ⩽ 0.07) formed readily in the Fe-bearing reactor and held dissolved sulfide at levels below saturation for orpiment and realgar. The mackinawite sequestered little As (<0.1 wt.%), however, and aqueous As accumulated to levels above the influent concentration as microbial Fe(III) reduction consumed goethite and mobilized adsorbed As. A relatively small amount of pyrite (FeS 2) and greigite (Fe 3S 4) formed in the Fe-bearing reactor when we injected a polysulfide solution (Na 2S 4) to a final concentration of 0.5 mM after 216, 230, 279, and 286 days. The pyrite, and to a lesser extent the greigite, that formed

  6. COMPUTATIONAL AND EXPERIMENTAL MODELING OF THREE-PHASE SLURRY-BUBBLE COLUMN REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    Isaac K. Gamwo; Dimitri Gidaspow

    1999-09-01

    Considerable progress has been achieved in understanding three-phase reactors from the point of view of kinetic theory. In a paper in press for publication in Chemical Engineering Science (Wu and Gidaspow, 1999) we have obtained a complete numerical solution of bubble column reactors. In view of the complexity of the simulation a better understanding of the processes using simplified analytical solutions is required. Such analytical solutions are presented in the attached paper, Large Scale Oscillations or Gravity Waves in Risers and Bubbling Beds. This paper presents analytical solutions for bubbling frequencies and standing wave flow patterns. The flow patterns in operating slurry bubble column reactors are not optimum. They involve upflow in the center and downflow at the walls. It may be possible to control flow patterns by proper redistribution of heat exchangers in slurry bubble column reactors. We also believe that the catalyst size in operating slurry bubble column reactors is not optimum. To obtain an optimum size we are following up on the observation of George Cody of Exxon who reported a maximum granular temperature (random particle kinetic energy) for a particle size of 90 microns. The attached paper, Turbulence of Particles in a CFB and Slurry Bubble Columns Using Kinetic Theory, supports George Cody's observations. However, our explanation for the existence of the maximum in granular temperature differs from that proposed by George Cody. Further computer simulations and experiments involving measurements of granular temperature are needed to obtain a sound theoretical explanation for the possible existence of an optimum catalyst size.

  7. Theoretical and experimental study of the dark signal in CMOS image sensors affected by neutron radiation from a nuclear reactor

    Science.gov (United States)

    Xue, Yuanyuan; Wang, Zujun; He, Baoping; Yao, Zhibin; Liu, Minbo; Ma, Wuying; Sheng, Jiangkun; Dong, Guantao; Jin, Junshan

    2017-12-01

    The CMOS image sensors (CISs) are irradiated with neutron from a nuclear reactor. The dark signal in CISs affected by neutron radiation is studied theoretically and experimentally. The Primary knock-on atoms (PKA) energy spectra for 1 MeV incident neutrons are simulated by Geant4. And the theoretical models for the mean dark signal, dark signal non-uniformity (DSNU) and dark signal distribution versus neutron fluence are established. The results are found to be in good agreement with the experimental outputs. Finally, the dark signal in the CISs under the different neutron fluence conditions is estimated. This study provides the theoretical and experimental evidence for the displacement damage effects on the dark signal CISs.

  8. 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.

  9. Theoretical and Experimental Evaluation of the Temperature Distribution in a Dry Type Air Core Smoothing Reactor of HVDC Station

    Directory of Open Access Journals (Sweden)

    Yu Wang

    2017-05-01

    Full Text Available The outdoor ultra-high voltage (UHV dry-type air-core smoothing reactors (DASR of High Voltage Direct Current systems are equipped with a rain cover and an acoustic enclosure. To study the convective heat transfer between the DASR and the surrounding air, this paper presents a coupled model of the temperature and fluid field based on the structural features and cooling manner. The resistive losses of encapsulations calculated by finite element method (FEM were used as heat sources in the thermal analysis. The steady fluid and thermal field of the 3-D reactor model were solved by the finite volume method (FVM, and the temperature distribution characteristics of the reactor were obtained. Subsequently, the axial and radial temperature distributions of encapsulation were investigated separately. Finally, an optical fiber temperature measurement scheme was used for an UHV DASR under natural convection conditions. Comparative analysis showed that the simulation results are in good agreement with the experimental data, which verifies the rationality and accuracy of the numerical calculation. These results can serve as a reference for the optimal design and maintenance of UHV DASRs.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-06-28

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

  11. Experimental studies on catalytic hydrogen recombiners for light water reactors; Experimentelle Untersuchungen zu katalytischen Wasserstoffkombinatoren fuer Leichtwasserreaktoren

    Energy Technology Data Exchange (ETDEWEB)

    Drinovac, P.

    2006-06-19

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1990-11-01

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

  13. A statistical experimental design to remove sulfate by crystallization in a fluidized-bed reactor

    OpenAIRE

    Mark Daniel G. de Luna; Rance, Diana Pearl M.; Luzvisminda M. Bellotindos; Lu, Ming-Chun

    2016-01-01

    This study used crystallization in a fluidized-bed reactor as an alternative technology to the conventional chemical precipitation to remove sulfate. The Box-Behnken Design was used to study the effects and interactions of seed dosage of synthetic gypsum, initial sulfate concentration and molar ratio of calcium to sulfate on conversion and removal of sulfate. The optimum conditions of conversion and removal of sulfate were determined and used to treat the simulated acid mine drainage (AMD) wa...

  14. Stress state dependence of in-reactor creep and swelling. Part 2: Experimental results

    Science.gov (United States)

    Hall, M. M., Jr.; Flinn, J. E.

    2010-01-01

    Irradiation creep constitutive equations, which were developed in Part I, are used here to analyze in-reactor creep and swelling data obtained ca. 1977-1979 as part of the US breeder reactor program. The equations were developed according to the principles of incremental continuum plasticity for the purpose of analyzing data obtained from a novel irradiation experiment that was conducted, in part, using Type 304 stainless steel that had been previously irradiated to significant levels of void swelling. Analyses of these data support an earlier observation that all stress states, whether tensile, compressive, shear or mixed, can affect both void swelling and interactions between irradiation creep and swelling. The data were obtained using a set of five unique multiaxial creep-test specimens that were designed and used for the first time in this study. The data analyses demonstrate that the constitutive equations derived in Part I provide an excellent phenomenological representation of the interactive creep and swelling phenomena. These equations provide nuclear power reactor designers and analysts with a first-of-its-kind structural analysis tool for evaluating irradiation damage-dependent distortion of complex structural components having gradients in neutron damage rate, temperature and stress state.

  15. The SPES3 Experimental Facility Design for the IRIS Reactor Simulation

    Directory of Open Access Journals (Sweden)

    Mario Carelli

    2009-01-01

    Full Text Available IRIS is an advanced integral pressurized water reactor, developed by an international consortium led by Westinghouse. The licensing process requires the execution of integral and separate effect tests on a properly scaled reactor simulator for reactor concept, safety system verification, and code assessment. Within the framework of an Italian R&D program on Nuclear Fission, managed by ENEA and supported by the Ministry of Economic Development, the SPES3 facility is under design and will be built and operated at SIET laboratories. SPES3 simulates the primary, secondary, and containment systems of IRIS with 1 : 100 volume scale, full elevation, and prototypical thermal-hydraulic conditions. The simulation of the facility with the RELAP5 code and the execution of the tests will provide a reliable tool for data extrapolation and safety analyses of the final IRIS design. This paper summarises the main design steps of the SPES3 integral test facility, underlying choices and phases that lead to the final design.

  16. Experimental Study on Ash-Returned Reactor of CFB Atmospheric Air Gasifier

    Science.gov (United States)

    Shihong, Zhang; Luning, Tian; Xianrong, Zhou; Hanping, Chen; Haiping, Yang; Xianhua, Wang

    In an attempt to improve the gasification efficiency and decrease the carbon content in fly ash of atmospheric air CFB gasifiers, an innovatory equipment by name ash-returned reactor is put forward by SKLCC. Ash-returned reactor is an ash-returned apparatus on line of ash circulation, typically like "U" type valve in CFB boilers, with additional function of some extent combustion of residual carbon and increase the furnace inlet temperature of returning ash, and hence the coal conversion of gasifiers is enhanced. As to its configuration compared to conventional "U" type valve, ash-returned rector has two distinguished features of several times of height scale of fluidizing transportation region to meet the combustion reaction time need and appropriate heat transfer tube bundles arranged in the region to moderate the local temperature so as to avoid slagging. And hence, corresponding to the structure renovation, the material transportation and regulation performance of ash-returned reactor is primarily investigated through a series of experiments in a cold lab-scale facility in this paper. The heat transfer characteristic of the tube bundles is then researched and its influential factors are further discussed. These works lay a foundation on the following study of hot state experiments and industrial applications.

  17. Experimental Studies on Assemblies 1 and 2 of the Fast Reactor FR-0. Part 2

    Energy Technology Data Exchange (ETDEWEB)

    Hellstrand, E.; Andersson, T.L.; Brunfelter, B.; Kockum, J.; Londen, S.O.; Tiren, L.I.

    1965-12-15

    In a first part of this report, published as AE-195, an account was given of critical mass determinations and measurements of flux distribution and reaction ratios in the first assemblies of the fast zero power reactor FR0. This second part of the report deals with various investigations involving the measurement of reactivity. Control rod calibrations have been made using the positive period, the inverse multiplication, the rod drop and the pulsed source techniques, and show satisfactory agreement between the various methods. The reactivity worths of samples of different materials and different sizes have been measured at the core centre. Comparisons with perturbation calculations show that the regular and adjoint fluxes are well predicted in the central region of the core. The variation in the prompt neutron life-time with reactivity has been studied by means of the pulsed source and the Rossi-{alpha} techniques. Comparison with one region calculations reveals large discrepancies, indicating that this simple model is inadequate. Some investigations of streaming effects in an empty channel in the reactor and of interaction effects between channels have been made and are compared with theoretical estimates. Measurements of the reactivity worth of an air gap between the reactor halves and of the temperature coefficient are also described in the report. The work has been performed as a joint effort by AB Atomenergi and the Research Institute of National Defence.

  18. Modeling and Experimental Studies of Mercury Oxidation and Adsorption in a Fixed-Bed and Entrained-Flow Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Buitrago, Paula A. [Univ. of Utah, Salt Lake City, UT (United States); Morrill, Mike [Univ. of Utah, Salt Lake City, UT (United States); Lighty, JoAnn S. [Univ. of Utah, Salt Lake City, UT (United States); Silcox, Geoffrey D. [Univ. of Utah, Salt Lake City, UT (United States)

    2009-06-01

    This report presents experimental and modeling mercury oxidation and adsorption data. Fixed-bed and single-particle models of mercury adsorption were developed. The experimental data were obtained with two reactors: a 300-W, methane-fired, tubular, quartz-lined reactor for studying homogeneous oxidation reactions and a fixed-bed reactor, also of quartz, for studying heterogeneous reactions. The latter was attached to the exit of the former to provide realistic combustion gases. The fixed-bed reactor contained one gram of coconut-shell carbon and remained at a temperature of 150°C. All methane, air, SO2, and halogen species were introduced through the burner to produce a radical pool representative of real combustion systems. A Tekran 2537A Analyzer coupled with a wet conditioning system provided speciated mercury concentrations. At 150°C and in the absence of HCl or HBr, the mercury uptake was about 20%. The addition of 50 ppm HCl caused complete capture of all elemental and oxidized mercury species. In the absence of halogens, SO2 increased the mercury adsorption efficiency to up to 30 percent. The extent of adsorption decreased with increasing SO2 concentration when halogens were present. Increasing the HCl concentration to 100 ppm lessened the effect of SO2. The fixed-bed model incorporates Langmuir adsorption kinetics and was developed to predict adsorption of elemental mercury and the effect of multiple flue gas components. This model neglects intraparticle diffusional resistances and is only applicable to pulverized carbon sorbents. It roughly describes experimental data from the literature. The current version includes the ability to account for competitive adsorption between mercury, SO2, and NO2. The single particle model simulates in-flight sorbent capture of elemental mercury. This model was developed to include Langmuir and Freundlich isotherms, rate equations, sorbent feed rate, and

  19. Experimental studies into the thermal-hydraulic performance of the VK-300 reactor based on a draft tube model

    Directory of Open Access Journals (Sweden)

    N.P. Serdun

    2015-12-01

    Full Text Available The paper presents an experimental study into the thermal-hydraulic performance of the VK-300 reactor based on a model of a single draft tube at a pressure of 3.4MPa, various flow rates and the model inlet relative enthalpies of –0.05 to 0.2. The experimental procedures include generation of a steam-water mixture circulation with a preset flow rate and a relative enthalpy through the test section at a pressure of 3.3 to 3.4MPa, and measurement of thermal-hydraulic parameters within the circuit's representative upflow and downflow lengths of practical interest. There have been confirmed the designs used to support the reactor facility serviceability and the assumptions concerning the thermal-hydraulic performance of a natural circulation circuit used in the analysis thereof. It has been shown that, across the analyzed range of the relative enthalpy values, the draft tube has an annular-dispersed or an annular flow of the steam-water mixture, both providing for the significant separation of the steam-water mixture (Ksep=0.4 at the draft tube edges and in the mixing chamber. The perforation in the upper part of the draft tubes allows the separation coefficient to be increased at the first stage and creates more favorable conditions for the second-stage separation. The measured values of the void fraction in the mixing chamber and in the draft tube are in a satisfactory agreement with calculations based on Z.L. Miropolskiy's method and the RELAP code and may be used to verify the VK-300 thermal-hydraulic codes. It has been shown that steam may enter the ring slit that simulates the annular space and reach the reactor core inlet. Further investigations need to be conducted to study this effect for its guaranteed exclusion and for the development of emergency response procedures.

  20. Experimental Study of the Effect of Graphite Dispersion on the Heat Transfer Phenomena in a Reactor Cavity Cooling System

    Energy Technology Data Exchange (ETDEWEB)

    Vaghetto, Rodolfo; Capone, Luigi; Hassan, Yassin A

    2011-05-31

    An experimental activity was performed to observe and study the effects of graphite dispersion and deposition on thermal-hydraulic phenomena in a reactor cavity cooling system (RCCS). The small-scale RCCS experimental facility (16.5 x 16.5 x 30.4 cm) used for this activity represents half of the reactor cavity with an electrically heated vessel. Water flowing through five vertical pipes removes the heat produced in the vessel and releases it into the environment by mixing with cold water in a large tank. The particle image velocimetry technique was used to study the velocity field of the air inside the cavity. A set of 52 thermocouples was installed in the facility to monitor the temperature profiles of the vessel, pipe walls, and air. Ten grams of a fine graphite powder (average particle size 2 m) was injected into the cavity through a spraying nozzle placed at the bottom of the vessel. The temperatures and air velocity field were recorded and compared with the measurements obtained before the graphite dispersion, showing a decrease of the temperature surfaces that was related to an increase in their emissivity. The results contribute to the understanding of RCCS capability in an accident scenario.

  1. Uncertainty analysis and flow measurements in an experimental mock-up of a molten salt reactor concept

    Energy Technology Data Exchange (ETDEWEB)

    Yamaji, Bogdan; Aszodi, Attila [Budapest University of Technology and Economics (Hungary). Inst. of Nuclear Techniques

    2016-09-15

    In the paper measurement results from the experimental modelling of a molten salt reactor concept will be presented along with detailed uncertainty analysis of the experimental system. Non-intrusive flow measurements are carried out on the scaled and segmented mock-up of a homogeneous, single region molten salt fast reactor concept. Uncertainty assessment of the particle image velocimetry (PIV) measurement system applied with the scaled and segmented model is presented in detail. The analysis covers the error sources of the measurement system (laser, recording camera, etc.) and the specific conditions (de-warping of measurement planes) originating in the geometry of the investigated domain. Effect of sample size in the ensemble averaged PIV measurements is discussed as well. An additional two-loop-operation mode is also presented and the analysis of the measurement results confirm that without enhancement nominal and other operation conditions will lead to strong unfavourable separation in the core flow. It implies that use of internal flow distribution structures will be necessary for the optimisation of the core coolant flow. Preliminary CFD calculations are presented to help the design of a perforated plate located above the inlet region. The purpose of the perforated plate is to reduce recirculation near the cylindrical wall and enhance the uniformity of the core flow distribution.

  2. Experimental investigation and model validation of a CaO/Ca(OH)2 fluidized bed reactor for thermochemical energy storage applications

    OpenAIRE

    Álvarez Criado, Yolanda; Huille, Alfred; Rougé, Sylvie; Abanades García, Juan Carlos

    2016-01-01

    The CaO/Ca(OH)2 hydration/dehydration chemical loop has long been recognized as a potential candidate for application in energy storage systems for concentrated solar plants. However, the technology still remains at a conceptual level because little information has been published on the performance of the key reactors in the system. In this work, we experimentally investigate the hydration and dehydration reactors in a 5.5 kWth batch fluidized bed reactor, in conditions relevant to larger sys...

  3. Feasibility Study of a Novel Membrane Reactor for Syngas Production. Part 1: Experimental Study of O2 Permeation through Perovskite Membranes under Reducing and Non-Reducing Atmospheres

    NARCIS (Netherlands)

    Zhang Wenxing, Z.W.; Zhang, Wenxing; Smit, J.; van Sint Annaland, M.; Kuipers, J.A.M.

    2007-01-01

    In this contribution, the feasibility of a novel membrane reactor for energy efficient syngas production is investigated by means of an experimental and a simulation study. In Part 1, a detailed experimental study is performed on the O2 permeation through a perovskite membrane with composition

  4. Selective hydrogenation in trickle-bed reactor. Experimental and modelling including partial wetting.

    OpenAIRE

    Dietz, Adrian; Julcour-Lebigue, Carine; Wilhelm, Anne-Marie; Delmas, Henri

    2003-01-01

    International audience; A steady state model of a trickle bed reactor is developed for the consecutive hydrogenation of 1,5,9-cyclododecatriene on a Pd/Al2O3 catalyst. Various experiments have shown that the selectivity of this reaction towards the product of interest is much lower in co-current down-flow (trickle-bed) than in up-flow. This is due to uneven liquid distribution and to partial wetting of the catalyst surface at low liquid flow rates. The non-isothermal heterogeneous model propo...

  5. Experimental study of the temperature distribution in the TRIGA IPR-R1 Brazilian research reactor; Investigacao experimental da distribuicao de temperaturas no reator nuclear de pesquisa TRIGA IPR-R1

    Energy Technology Data Exchange (ETDEWEB)

    Mesquita, Amir Zacarias

    2005-07-01

    The TRIGA-IPR-R1 Research Nuclear Reactor has completed 44 years in operation in November 2004. Its initial nominal thermal power was 30 kW. In 1979 its power was increased to 100 kW by adding new fuel elements to the reactor. Recently some more fuel elements were added to the core increasing the power to 250 kW. The TRIGA-IPR-R1 is a pool type reactor with a natural circulation core cooling system. Although the large number of experiments had been carried out with this reactor, mainly on neutron activation analysis, there is not many data on its thermal-hydraulics processes, whether experimental or theoretical. So a number of experiments were carried out with the measurement of the temperature inside the fuel element, in the reactor core and along the reactor pool. During these experiments the reactor was set in many different power levels. These experiments are part of the CDTN/CNEN research program, and have the main objective of commissioning the TRIGA-IPR-R1 reactor for routine operation at 250 kW. This work presents the experimental and theoretical analyses to determine the temperature distribution in the reactor. A methodology for the calibration and monitoring the reactor thermal power was also developed. This methodology allowed adding others power measuring channels to the reactor by using thermal processes. The fuel thermal conductivity and the heat transfer coefficient from the cladding to the coolant were also experimentally valued. lt was also presented a correlation for the gap conductance between the fuel and the cladding. The experimental results were compared with theoretical calculations and with data obtained from technical literature. A data acquisition and processing system and a software were developed to help the investigation. This system allows on line monitoring and registration of the main reactor operational parameters. The experiments have given better comprehension of the reactor thermal-fluid dynamics and helped to develop numerical

  6. Experimental Development and Demonstration of Ultrasonic Measurement Diagnostics for Sodium Fast Reactor Thermal-hydraulics

    Energy Technology Data Exchange (ETDEWEB)

    Tokuhiro, Akira; Jones, Byron

    2013-09-13

    This research project will address some of the principal technology issues related to sodium-cooled fast reactors (SFR), primarily the development and demonstration of ultrasonic measurement diagnostics linked to effective thermal convective sensing under normatl and off-normal conditions. Sodium is well-suited as a heat transfer medium for the SFR. However, because it is chemically reactive and optically opaque, it presents engineering accessibility constraints relative to operations and maintenance (O&M) and in-service inspection (ISI) technologies that are currently used for light water reactors. Thus, there are limited sensing options for conducting thermohydraulic measurements under normal conditions and off-normal events (maintenance, unanticipated events). Acoustic methods, primarily ultrasonics, are a key measurement technology with applications in non-destructive testing, component imaging, thermometry, and velocimetry. THis project would have yielded a better quantitative and qualitative understanding of the thermohydraulic condition of solium under varied flow conditions. THe scope of work will evaluate and demonstrate ultrasonic technologies and define instrumentation options for the SFR.

  7. Experimental study of hydrodynamic and operation start of a baffled anaerobic reactor treating sewage

    Directory of Open Access Journals (Sweden)

    Ana Carolina Silveira Perico

    2009-12-01

    Full Text Available It is important to provide individual sanitation systems for sewage peri-urban communities or rural areas to minimize impacts on the environment and human health caused by sewage discharge in natura into water resources. In this context, the anaerobic digestion of effluent has been one of the main considered technologies due to easy implementation, material minimization and reduction in waste production. The objective of this work was to study a Baffled Anaerobic Reactor (BAR including its hydrodynamic characteristics, percentile of inoculum to be applied and reactor operation start. It was concluded that the flow is dispersed with 3.84% of dead spaces and that 20% of the cow manure provided best results; however, due to the high fiber content of the manure, its use is not recommended as inoculum. The BAR system, composed of four chambers, presented good performance for sewage treatment of a rural community in terms of organic substance removal (COD, turbidity and solids meeting effluent disposal standards of these parameters considering the Federal and Minas Gerais State legislation, in Brazil, even in a transient phase of operation, at temperatures below 20°C. However, the effluents from the BAR can’t be released into water bodies without other parameters such as nitrogen, phosphorus, fecal coliforms, and others are investigated to be conforming to those standards.

  8. Theoretical and experimental study of the photocatalytic activity of ZnO coated tubular reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ríos-Valdovinos, E.; Amézaga-Madrid, P.; Antúnez-Flores, W.; Pola-Albores, F.; Pizá-Ruiz, P.; Miki-Yoshida, M., E-mail: mario.miki@cimav.edu.mx

    2014-01-15

    Highlights: • High quality ZnO thin films were deposited on the internal surface of fused silica tubing. • Surface carrier concentration was calculated theoretically under external irradiation. • Influence of film thickness on photocatalytic activity was explained by this model. • An optimum thickness around 60–70 nm was determined to get highest activity. -- Abstract: ZnO thin films were deposited inside of fused silica tubing by aerosol assisted chemical vapor deposition technique. The films were transparent, uniform, highly adherent and non-light scattering. Photocatalytic activity of internally ZnO coated tubing was evaluated by discoloration of a methyl orange aqueous solution in a batch reactor. Tubing was externally irradiated with UV-A at room temperature. A one dimensional model was proposed to calculate the spatial distribution of the carrier density and the films’ surface charge carrier concentration. This model can explain the influence of the films thickness on the photocatalytic activity. Results showed that the photocatalytic activity largely depends on the film thickness. For external irradiation of the films the optimum thickness was around 60–70 nm, for which the photocatalytic activity was maximum. The photonic efficiency of internally ZnO coated tubular reactors was evaluated as a function of initial colorant concentration, irradiation time and intensity. Furthermore, due to the high activity of the ZnO films, the films were repeatedly exposed to UV-A irradiation cycles, followed by activity measurement.

  9. Development of Safety Analysis Codes and Experimental Validation for a Very High Temperature Gas-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chang, H. Oh, PhD; Cliff Davis; Richard Moore

    2004-11-01

    The very high temperature gas-cooled reactors (VHTGRs) are those concepts that have average coolant temperatures above 900 degrees C or operational fuel temperatures above 1250 degrees C. These concepts provide the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation and nuclear hydrogen generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperatures to support process heat applications, such as desalination and cogeneration, the VHTGR's higher temperatures are suitable for particular applications such as thermochemical hydrogen production. However, the high temperature operation can be detrimental to safety following a loss-of-coolant accident (LOCA) initiated by pipe breaks caused by seismic or other events. Following the loss of coolant through the break and coolant depressurization, air from the containment will enter the core by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structures and fuel. The oxidation will release heat and accelerate the heatup of the reactor core. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. The Idaho National Engineering and Environmental Laboratory (INEEL) has investigated this event for the past three years for the HTGR. However, the computer codes used, and in fact none of the world's computer codes, have been sufficiently developed and validated to reliably predict this event. New code development, improvement of the existing codes, and experimental validation are imperative to narrow the uncertaninty in the predictions of this type of accident. The objectives of this Korean/United States collaboration are to develop advanced computational methods for VHTGR safety analysis codes and to validate these computer codes.

  10. 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.

  11. Experimental assessment of accident scenarios for the high temperature reactor fuel system

    Energy Technology Data Exchange (ETDEWEB)

    Seeger, O.; Laurie, M.; Bottomley, P.D.W.; Rondinella, V.V. [European Commission, Joint Research Center, Karlsruhe (Germany). Inst. for Transuranium Elements; Avincola, V. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany). Inst. fuer Angewandte Materialien (IAM-AWP); Allelein, H.J. [RWTH Aachen Univ. (Germany). Lehrstuhl fuer Reaktorsicherheit und -technik

    2013-11-15

    The High Temperature Reactor (HTR) is characterized by an advanced design with passive safety features. Fuel elements are constituted by a graphite matrix containing sub-mm-sized fuel particles with TRi-ISOtropic (TRISO) coating, designed to provide high fission product retention. During a loss of coolant accident scenario in a HTR the maximum temperature is foreseen to be in the range of 1,600 to 1,650 C, remaining well below the melting point of the fuel. Two key aspects associated with the safety of HTR fuel are assessed in this paper: fission product retention at temperatures up to 1,800 C is analyzed with the Cold Finger Apparatus (KueFA) while the behaviour of HTR-relevant fuel materials in an oxidizing environment is studied with the Corrosion Apparatus KORA. The KueFA is used to observe the combined effects of Depressurization and LOss of Forced Circulation (DLOFC) accident scenarios on HTR fuel. Originally designed at the Forschungszentrum Juelich (FZJ), an adapted KueFA operates on irradiated fuel in hot cell at JRC-ITU. A fuel pebble is heated in helium atmosphere for several hundred hours, mimicking accident temperatures up to 1,800 C and realistic temperature transients. Nongaseous volatile fission products released from the fuel condense on a water cooled stainless steel plate dubbed 'Cold Finger'. Exchanging plates frequently during the experiment and analyzing plate deposits by means of High Purity Germanium (HPGe) gamma spectroscopy allows a reconstruction of the fission product release as a function of time and temperature. To achieve a good quantification of the release, a careful calibration of the setup is necessary and a collimator needs to be used in some cases. The analysis of condensation plates from recent KueFA tests shows that fission product release quantification is possible at high and low activity levels. Another relevant HTR accident scenario is air ingress into the reactor vessel as a consequence of a DLOFC incident. In

  12. Experimental investigation of the solar carbothermic reduction of ZnO using a two-cavity solar reactor

    Energy Technology Data Exchange (ETDEWEB)

    Frommherz, U.; Osinga, T.; Steinfeld, A.; Wieckert, C.

    2003-03-01

    Zinc production by solar carbothermic reduction of ZnO offers a CO{sub 2} emission reduction by a factor of 5 vis-a-vis the conventional fossil-fuel-based electrolytic or Imperial Smelting processes. Zinc can serve as a fuel in Zn-air fuel cells or can be further reacted with H{sub 2}O to form high-purity H{sub 2}. In either case, the product ZnO can be solar-recycled to Zn. We report on experimental results obtained with a 5 kW solar chemical reactor prototype that features two cavities in series, with the inner one functioning as the solar absorber and the outer one as the reaction chamber. Tests were conducted at PSI's Solar Furnace and ETH's High-Flux Solar Simulator to investigate the effect of process temperature (range 1350-1600 K) and reducing agent type (beech charcoal, activated charcoal, petcoke) on the reactor's performance and on the chemical conversion. In a typical 40-min solar experiment at 1500 K, 500 g of a 1:0.8 stoichiometric ZnO-C mixture were processed into Zn(g), CO, and CO{sub 2}. Thermal efficiencies of up to 20% were achieved. (author)

  13. Experimental Design for Evaluating Selected Nondestructive Measurement Technologies - Advanced Reactor Technology Milestone: M3AT-16PN2301043

    Energy Technology Data Exchange (ETDEWEB)

    Ramuhalli, Pradeep [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Hirt, Evelyn H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Pitman, Stan G. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Dib, Gerges [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Roy, Surajit [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Good, Morris S. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Walker, Cody M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-07-16

    The harsh environments in advanced reactors (AdvRx) increase the possibility of degradation of safety-critical passive components, and therefore pose a particular challenge for deployment and extended operation of these concepts. Nondestructive evaluation technologies are an essential element for obtaining information on passive component condition in AdvRx, with the development of sensor technologies for nondestructively inspecting AdvRx passive components identified as a key need. Given the challenges posed by AdvRx environments and the potential needs for reducing the burden posed by periodic in-service inspection of hard-to-access and hard-to-replace components, a viable solution may be provided by online condition monitoring of components. This report identifies the key challenges that will need to be overcome for sensor development in this context, and documents an experimental plan for sensor development, test, and evaluation. The focus of initial research and development is on sodium fast reactors, with the eventual goal of the research being developing the necessary sensor technology, quantifying sensor survivability and long-term measurement reliability for nondestructively inspecting critical components. Materials for sensor development that are likely to withstand the harsh environments are described, along with a status on the fabrication of reference specimens, and the planned approach for design and evaluation of the sensor and measurement technology.

  14. Combined numerical and experimental investigation into the coolant flow hydrodynamics and mass transfer behind the spacer grid in fuel assemblies of the floating power unit reactor

    Directory of Open Access Journals (Sweden)

    S.M. Dmitriev

    2015-09-01

    Full Text Available The results of experimental investigations into the local hydrodynamics and inter-cell mass transfer of the coolant flow in representative zones of the KLT-40C reactor FAs behind the plate-type spacer grid are presented. The investigations were conducted on an aerodynamic rig using the admixture diffusion method (the tracer-gas method. A study into the spatial dispersion of the absolute flow velocity projections and into the distribution of the tracer concentration allowed specify the coolant flow pattern behind the FA plate-type spacer grid of the KLT-40C reactor. The results of measuring the flow friction coefficient in the plate-type spacer grid, depending on the Reynolds number, are presented. Based on the obtained experimental data, recommendations have been provided for updating the procedures to calculate the coolant flow rates for the KLT-40C reactor core by-channel codes. The results of investigating the coolant flow local hydrodynamics and mass transfer in the KLT-40C reactor FAs have been adopted for practical use by Afrikantov OKBM for estimating the heat-engineering reliability of the KLT-40C reactor cores and have been data based for verification of CFD codes and detailed by-channel calculation of the KLT-40C reactor core.

  15. Monitoring of dry anaerobic fermentation in experimental facility with use of biofilm reactor

    Directory of Open Access Journals (Sweden)

    Milan Šinkora

    2011-01-01

    -called biofilm reactor. An external reactor with a cultivated bacterial biofilm on an immovable carrier with the percolate flowing through it has been constructed in laboratory conditions for this purpose. The choice of suitable percolate strategy (this means the frequency of sprinkling and the amount of percolate directly influences the process of anaerobic fermentation.

  16. Experimental Investigations of Physical and Chemical Properties for Microalgae HTL Bio-Crude Using a Large Batch Reactor

    Directory of Open Access Journals (Sweden)

    Farhad M. Hossain

    2017-04-01

    Full Text Available As a biofuel feedstock, microalgae has good scalability and potential to supply a significant proportion of world energy compared to most types of biofuel feedstock. Hydrothermal liquefaction (HTL is well-suited to wet biomass (such as microalgae as it greatly reduces the energy requirements associated with dewatering and drying. This article presents experimental analyses of chemical and physical properties of bio-crude oil produced via HTL using a high growth-rate microalga Scenedesmus sp. in a large batch reactor. The overarching goal was to investigate the suitability of microalgae HTL bio-crude produced in a large batch reactor for direct application in marine diesel engines. To this end we characterized the chemical and physical properties of the bio-crudes produced. HTL literature mostly reports work using very small batch reactors which are preferred by researchers, so there are few experimental and parametric measurements for bio-crude physical properties, such as viscosity and density. In the course of this study, a difference between traditionally calculated values and measured values was noted. In the parametric study, the bio-crude viscosity was significantly closer to regular diesel and biodiesel standards than transesterified (FAME microalgae biodiesel. Under optimised conditions, HTL bio-crude’s high density (0.97–1.04 kg·L−1 and its high viscosity (70.77–73.89 mm2·s−1 had enough similarity to marine heavy fuels. although the measured higher heating value, HHV, was lower (29.8 MJ·kg−1. The reaction temperature was explored in the range 280–350 °C and bio-crude oil yield and HHV reached their maxima at the highest temperature. Slurry concentration was explored between 15% and 30% at this temperature and the best HHV, O:C, and N:C were found to occur at 25%. Two solvents (dichloromethane and n-hexane were used to recover the bio-crude oil, affecting the yield and chemical composition of the bio-crude.

  17. Experimental investigation of pebble flow dynamics using radioactive particle tracking technique in a scaled-down Pebble Bed Modular Reactor (PBMR)

    Energy Technology Data Exchange (ETDEWEB)

    Khane, Vaibhav; Said, I.A.; Al-Dahhan, Muthanna H., E-mail: aldahhanm@mst.edu

    2016-06-15

    Highlights: • Pebble Flow fields at Pebble Bed Modular Reactor was investigated. • Radioactive Particle Tracking (RPT) technique has been used. • Plug flow type velocity profile is suggested at upper cylindrical region. - Abstract: The Pebble Bed Modular Reactor (PBMR) is a type of very-high-temperature reactor (VHTR) that is conceptually very similar to moving bed reactors used in the chemical and petrochemical industries. In a PBMR core, nuclear fuel is in the form of pebbles and moves slowly under the influence of gravity. In this work, an integrated experimental and computational study of granular flow in a scaled-down cold flow PBMR was performed. A continuous pebble re-circulation experimental set-up, mimicking the flow of pebbles in a PBMR was designed and developed. An experimental investigation of pebble flow dynamics in a scaled down test reactor was carried out using a non-invasive radioactive particle tracking (RPT) technique that used a cobalt-60 based tracer to mimic pebbles in terms of shape, size and density. A cross-correlation based position reconstruction algorithm and RPT calibration data were used to obtain results about Lagrangian trajectories, the velocity field, and residence time distributions. The RPT technique results a serve as a benchmark data for assessing contact force models used in the discrete element method (DEM) simulations.

  18. Numerical and experimental investigation of surface vortex formation in coolant reservoirs of reactor safety systems

    Energy Technology Data Exchange (ETDEWEB)

    Pandazis, Peter [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany); Babcsany, Boglarka [Budapest Univ. of Technology and Economics (Hungary). Inst. of Nuclear Techniques

    2016-11-15

    The reliable operation of the emergency coolant pumps and passive gravitational injection systems are an important safety issue during accident scenarios with coolant loss in pressurized water reactors. Because of the pressure drop and flow disturbances surface vortices develops at the pump intakes if the water level decreasing below a critical value. The induced swirling flow and gas entrainment lead to flow limitation and to pump failures and damages. The prediction of the critical submergence to avoid surface vortex building is difficult because it depends on many geometrical and fluid dynamical parameters. An alternative and new method has been developed for the investigation of surface vortices. The method based on the combination of CFD results with the analytical vortex model of Burgers and Rott. For further investigation the small scale experiments from the Institute of Nuclear Techniques of the Budapest University of Technology and Economics are used which were inspired from flow limitation problems during the draining of the bubble condenser trays at a VVER type nuclear power plants.

  19. The thermal decomposition of the benzyl radical in a heated micro-reactor. I. Experimental findings

    Science.gov (United States)

    Buckingham, Grant T.; Ormond, Thomas K.; Porterfield, Jessica P.; Hemberger, Patrick; Kostko, Oleg; Ahmed, Musahid; Robichaud, David J.; Nimlos, Mark R.; Daily, John W.; Ellison, G. Barney

    2015-01-01

    The pyrolysis of the benzyl radical has been studied in a set of heated micro-reactors. A combination of photoionization mass spectrometry (PIMS) and matrix isolation infrared (IR) spectroscopy has been used to identify the decomposition products. Both benzyl bromide and ethyl benzene have been used as precursors of the parent species, C6H5CH2, as well as a set of isotopically labeled radicals: C6H5CD2, C6D5CH2, and C6H513CH2. The combination of PIMS and IR spectroscopy has been used to identify the earliest pyrolysis products from benzyl radical as: C5H4=C=CH2, H atom, C5H4—C ≡ CH, C5H5, HCCCH2, and HC ≡ CH. Pyrolysis of the C6H5CD2, C6D5CH2, and C6H513CH2 benzyl radicals produces a set of methyl radicals, cyclopentadienyl radicals, and benzynes that are not predicted by a fulvenallene pathway. Explicit PIMS searches for the cycloheptatrienyl radical were unsuccessful, there is no evidence for the isomerization of benzyl and cycloheptatrienyl radicals: C6H5CH2⇋C7H7. These labeling studies suggest that there must be other thermal decomposition routes for the C6H5CH2 radical that differ from the fulvenallene pathway.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-06-01

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

  1. Experimental Breeder Reactor II (EBR-II) Fuel-Performance Test Facility (FPTF)

    Energy Technology Data Exchange (ETDEWEB)

    Pardini, J.A.; Brubaker, R.C.; Veith, D.J.; Giorgis, G.C.; Walker, D.E.; Seim, O.S.

    1982-01-01

    The Fuel-Performance Test Facility (FPTF) is the latest in a series of special EBR-II instrumented in-core test facilities. A flow control valve in the facility is programmed to vary the coolant flow, and thus the temperature, in an experimental-irradiation subassembly beneath it and coupled to it. In this way, thermal transients can be simulated in that subassembly without changing the temperatures in surrounding subassemblies. The FPTF also monitors sodium flow and temperature, and detects delayed neutrons in the sodium effluent from the experimental-irradiation subassembly beneath it. This facility also has an acoustical detector (high-temperature microphone) for detecting sodium boiling.

  2. The role of clusters in gas-solids reactors. An experimental study.

    NARCIS (Netherlands)

    Venderbosch, R.H.

    1998-01-01

    This PhD-work is meant to determine the contact efficiency experimentally for fluidization of fine particles over a wide range of superficial gas velocities (dp<200 mm and 0.1

  3. 14MeV neutron irradiation experiment on window materials for fusion experimental reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sato, Fuminobu; Oyama, Yukio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Iida, Toshiyuki

    1997-06-01

    Data on wavelength spectra of photons emitted from window material during neutron and gamma-ray irradiation has been required for design of next D-T burning fusion reactor such as ITER. Thus, a photon measurement system has been developed to analyze wavelength spectra of photons emitted from the optical window materials during 14MeV-neutron irradiation, and the system consisted of a sample holder, a radiation-resistant optical fiber, a photon counting analyzer and other electronic devices. The irradiation experiments for synthesized sapphire, high-purity silica glass and synthesized quartz were performed using a fusion neutron source FNS. As for all the sample, number of photon emission was proportional to the 14MeV-neutron flux in the range of 10{sup 6}-10{sup 11}n/cm{sup 2}/sec. The photon emission efficiency of F-center luminescence of the sapphire was 2200 {+-} 700photons/MeV, while the efficiency of F{sup +}-center luminescence was two order less than that of F-center. The wavelength spectra of the high-purity silica glass had a large peak around 450nm, which was concerned with decay of self-trapped excitons in oxygen vacancies. Its photon emission efficiency for 14MeV-neutrons has been found to be about 5 {+-} 3photons/MeV in visible range, while that for gamma-rays to be about 135 {+-} 50photons/MeV. The spectrum of photons emitted from the quartz had two large peaks around not only 450nm but also 650nm, and the photon emission efficiency in the wavelength range of 350-750nm was 14 {+-} 4photons/MeV. (author)

  4. The thermal decomposition of the benzyl radical in a heated micro-reactor. I. Experimental findings

    Energy Technology Data Exchange (ETDEWEB)

    Buckingham, Grant T.; Ormond, Thomas K. [Department of Chemistry and Biochemistry, University of Colorado, Boulder, Colorado 80309-0215 (United States); National Renewable Energy Laboratory, 15013 Denver West Parkway, Golden, Colorado 80401 (United States); Porterfield, Jessica P.; Ellison, G. Barney [Department of Chemistry and Biochemistry, University of Colorado, Boulder, Colorado 80309-0215 (United States); Hemberger, Patrick [Molecular Dynamics Group, Paul Scherrer Institut, CH-5232 Villigen-PSI (Switzerland); Kostko, Oleg; Ahmed, Musahid [Chemical Sciences Division, Lawrence Berkeley National Laboratories, Berkeley, California 94720 (United States); Robichaud, David J.; Nimlos, Mark R. [National Renewable Energy Laboratory, 15013 Denver West Parkway, Golden, Colorado 80401 (United States); Daily, John W. [Department of Mechanical Engineering, Center for Combustion and Environmental Research,University of Colorado, Boulder, Colorado 80309-0427 (United States)

    2015-01-28

    The pyrolysis of the benzyl radical has been studied in a set of heated micro-reactors. A combination of photoionization mass spectrometry (PIMS) and matrix isolation infrared (IR) spectroscopy has been used to identify the decomposition products. Both benzyl bromide and ethyl benzene have been used as precursors of the parent species, C{sub 6}H{sub 5}CH{sub 2}, as well as a set of isotopically labeled radicals: C{sub 6}H{sub 5}CD{sub 2}, C{sub 6}D{sub 5}CH{sub 2}, and C{sub 6}H{sub 5}{sup 13}CH{sub 2}. The combination of PIMS and IR spectroscopy has been used to identify the earliest pyrolysis products from benzyl radical as: C{sub 5}H{sub 4}=C=CH{sub 2}, H atom, C{sub 5}H{sub 4}—C ≡ CH, C{sub 5}H{sub 5}, HCCCH{sub 2}, and HC ≡ CH. Pyrolysis of the C{sub 6}H{sub 5}CD{sub 2}, C{sub 6}D{sub 5}CH{sub 2}, and C{sub 6}H{sub 5}{sup 13}CH{sub 2} benzyl radicals produces a set of methyl radicals, cyclopentadienyl radicals, and benzynes that are not predicted by a fulvenallene pathway. Explicit PIMS searches for the cycloheptatrienyl radical were unsuccessful, there is no evidence for the isomerization of benzyl and cycloheptatrienyl radicals: C{sub 6}H{sub 5}CH{sub 2}⇋C{sub 7}H{sub 7}. These labeling studies suggest that there must be other thermal decomposition routes for the C{sub 6}H{sub 5}CH{sub 2} radical that differ from the fulvenallene pathway.

  5. Three-phase packed bed reactor with an evaporating solvent—I. Experimental: the hydrogenation of 2,4,6-trinitrotoluene in methanol

    NARCIS (Netherlands)

    van Gelder, K.B.; Damhof, J.K.; Kroijenga, P.J.; Westerterp, K.R.

    1990-01-01

    In this paper we present experimental data on the three-phase hydrogenation of 2,4,6-trinitrotoluene (TNT) to triaminotoluene. The experiments are performed in a cocurrent upflow packed bed reactor. Methanol is used as an evaporating solvent. The influence of the main operating parameters, the

  6. A scaled experimental study of control blade insertion dynamics in Pebble-Bed Fluoride-Salt-Cooled High-Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Buster, Grant C., E-mail: grant.buster@gmail.com; Laufer, Michael R.; Peterson, Per F.

    2016-07-15

    Highlights: • A granular dynamics scaling methodology is discussed. • Control blade insertion in a representative pebble-bed core is experimentally studied. • Control blade insertion forces and pebble displacements are experimentally measured. • X-ray tomography techniques are used to observe pebble displacement distributions. - Abstract: Direct control element insertion into a pebble-bed reactor core is proposed as a viable control system in molten-salt-cooled pebble-bed reactors. Unlike helium-cooled pebble-bed reactors, this reactor type uses spherical fuel elements with near-neutral buoyancy in the molten-salt coolant, thus reducing contact forces on the fuel elements. This study uses the X-ray Pebble Bed Recirculation Experiment facility to measure the force required to insert a control element directly into a scaled pebble-bed. The required control element insertion force, and therefore the contact force on fuel elements, is measured to be well below recommended limits. Additionally, X-ray tomography is used to observe how the direct insertion of a control element physically displaces spherical fuel elements. The tomography results further support the viability of direct control element insertion into molten-salt-cooled pebble-bed reactor cores.

  7. An Experimental Investigation of Sewage Sludge Gasification in a Fluidized Bed Reactor

    Science.gov (United States)

    Calvo, L. F.; García, A. I.; Otero, M.

    2013-01-01

    The gasification of sewage sludge was carried out in a simple atmospheric fluidized bed gasifier. Flow and fuel feed rate were adjusted for experimentally obtaining an air mass : fuel mass ratio (A/F) of 0.2 gasification. This allowed improving the process heat transfer and, therefore, gasification efficiency. The heating value of the produced gas was 8.4 MJ/Nm, attaining a hot gas efficiency of 70% and a cold gas efficiency of 57%. PMID:24453863

  8. An analytical and experimental investigation of natural circulation transients in a model pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Massoud, M

    1987-01-01

    Natural Circulation phenomena in a simulated PWR was investigated experimentally and analytically. The experimental investigation included determination of system characteristics as well as system response to the imposed transient under symmetric and asymmetric operations. System characteristics were used to obtain correlation for heat transfer coefficient in heat exchangers, system flow resistance, and system buoyancy heat. Asymmetric transients were imposed to study flow oscillation and possible instability. The analytical investigation encompassed development of mathematical model for single-phase, steady-state and transient natural circulation as well as modification of existing model for two-phase flow analysis of phenomena such as small break LOCA, high pressure coolant injection and pump coast down. The developed mathematical model for single-phase analysis was computer coded to simulate the imposed transients. The computer program, entitled ''Symmetric and Asymmetric Analysis of Single-Phase Flow (SAS),'' were employed to simulate the imposed transients. It closely emulated the system behavior throughout the transient and subsequent steady-state. Modifications for two-phase flow analysis included addition of models for once-through steam generator and electric heater rods. Both programs are faster than real time. Off-line, they can be used for prediction and training applications while on-line they serve for simulation and signal validation. The programs can also be used to determine the sensitivity of natural circulation behavior to variation of inputs such as secondary distribution and power transients.

  9. Radiation Damage in Nuclear Fuel for Advanced Burner Reactors: Modeling and Experimental Validation

    Energy Technology Data Exchange (ETDEWEB)

    Jensen, Niels Gronbech; Asta, Mark; Ozolins, Nigel Browning' Vidvuds; de Walle, Axel van; Wolverton, Christopher

    2011-12-29

    The consortium has completed its existence and we are here highlighting work and accomplishments. As outlined in the proposal, the objective of the work was to advance the theoretical understanding of advanced nuclear fuel materials (oxides) toward a comprehensive modeling strategy that incorporates the different relevant scales involved in radiation damage in oxide fuels. Approaching this we set out to investigate and develop a set of directions: 1) Fission fragment and ion trajectory studies through advanced molecular dynamics methods that allow for statistical multi-scale simulations. This work also includes an investigation of appropriate interatomic force fields useful for the energetic multi-scale phenomena of high energy collisions; 2) Studies of defect and gas bubble formation through electronic structure and Monte Carlo simulations; and 3) an experimental component for the characterization of materials such that comparisons can be obtained between theory and experiment.

  10. Analysis of radiological accident emissions of a lead-cooled experimental reactor. LEADER Project; Analisis radiologico de las emisiones en caso de accidente de un reactor experimental refrigerado por plomo. Proyecto LEADER

    Energy Technology Data Exchange (ETDEWEB)

    Gomez Salcedo, F.; Cortes Martin, A.

    2013-07-01

    The LEADER project develops a conceptual level industrial size reactor cooled lead and a demonstration plant of this technology. The project objectives are to define the characteristics and design to installation scale reactor using available technologies and short-term components and assess safety aspects conducting a preliminary analysis of the impact of the facility.

  11. Contribution to modeling of the reflooding of a severely damaged reactor core using PRELUDE experimental results

    Energy Technology Data Exchange (ETDEWEB)

    Bachrata, A.; Fichot, F.; Repetto, G. [Institut de Radioprotection et de Surete Nucleaire IRSN, Cadarache (France); Quintard, M. [Universite de Toulouse, INPT, UPS, IMFT Institut de Mecanique des Fluides de Toulouse, Allee Camille Soula, F-31400 Toulouse (France); CNRS, IMFT, F-31400 Toulouse (France); Fleurot, J. [Institut de Radioprotection et de Surete Nucleaire IRSN, Cadarache (France)

    2012-07-01

    In case of accident at a nuclear power plant, water sources may not be available for a long period of time and the core heats up due to the residual power. The reflooding (injection of water into core) may be applied if the availability of safety injection is recovered during accident. If the injection becomes available only in the late phase of accident, water will enter a core configuration that will differ significantly from original rod-bundle geometry. Any attempt to inject water after significant core degradation can lead to further fragmentation of core material. The fragmentation of fuel rods may result in the formation of a 'debris bed'. The typical particle size in a debris bed might reach few millimeters (characteristic length-scale: 1 to 5 mm), i.e., a high permeability porous medium. The French 'Institut de Radioprotection et de Surete Nucleaire' is developing experimental programs (PEARL and PRELUDE) and simulation tools (ICARE-CATHARE and ASTEC) to study and optimize the severe accident management strategy and to assess the probabilities to stop the progress of in-vessel core degradation. It is shown that the quench front exhibits either a ID behaviour or a 2D one, depending on injection rate or bed characteristics. The PRELUDE experiment covers a rather large range of variation of parameters, for which the developed model appears to be quite predictive. (authors)

  12. High power 1 MeV neutral beam system and its application plan for the international tokamak experimental reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hemsworth, R.S. [ITER Joint Central Team, Naka, Ibaraki (Japan)

    1997-03-01

    This paper describes the Neutral Beam Injection system which is presently being designed for the International Tokamak Experimental Reactor, ITER, in Europe Japan and Russia, with co-ordination by the Joint Central Team of ITER at Naka, Japan. The proposed system consists of three negative ion based neutral injectors, delivering a total of 50 MW of 1 MeV D{sup 0} to the ITER plasma for a pulse length of >1000 s. Each injectors uses a single caesiated volume arc discharge negative ion source, and a multi-grid, multi-aperture accelerator, to produce about 40 A of 1 MeV D{sup -}. This will be neutralized by collisions with D{sub 2} in a sub-divided gas neutralizer, which has a conversion efficiency of about 60%. The charged fraction of the beam emerging from the neutralizer is dumped in an electrostatic residual ion dump. A water cooled calorimeter can be moved into the beam path to intercept the neutral beam, allowing commissioning of the injector independent of ITER. ITER is scheduled to produce its first plasma at the beginning of 2008, and the planning of the R and D, construction and installation foresees the neutral injection system being available from the start of ITER operations. (author)

  13. Modeling of secondary emission processes in the negative ion based electrostatic accelerator of the International Thermonuclear Experimental Reactor

    Directory of Open Access Journals (Sweden)

    G. Fubiani

    2008-01-01

    Full Text Available The negative ion electrostatic accelerator for the neutral beam injector of the International Thermonuclear Experimental Reactor (ITER is designed to deliver a negative deuterium current of 40 A at 1 MeV. Inside the accelerator there are several types of interactions that may create secondary particles. The dominating process originates from the single and double stripping of the accelerated negative ion by collision with the residual molecular deuterium gas (≃29% losses. The resulting secondary particles (positive ions, neutrals, and electrons are accelerated and deflected by the electric and magnetic fields inside the accelerator and may induce more secondaries after a likely impact with the accelerator grids. This chain of reactions is responsible for a non-negligible heat load on the grids and must be understood in detail. In this paper, we will provide a comprehensive summary of the physics involved in the process of secondary emission in a typical ITER-like negative ion electrostatic accelerator together with a precise description of the numerical method and approximations involved. As an example, the multiaperture-multigrid accelerator concept will be discussed.

  14. Analysis of quench-vent pressures for present design of ITER (International Thermonuclear Experimental Reactor) TF (toroidal field) coils

    Energy Technology Data Exchange (ETDEWEB)

    Slack, D.S.

    1989-09-20

    The International Thermonuclear Experimental Reactor (ITER) is a new tokamak design project with joint participation from Japan, the European Community, the Union of the Soviet Union, and the United States. This paper examines the effects of a quench within the toroidal field (TF) coils based on current ITER design. It is a preliminary, rough analysis. Its intent is to assist ITER designers while more accurate computer codes are being developed and to provide a check against these more rigorous solutions. Rigorous solutions to the quench problem are very complex involving three- dimensional heat transfer, extreme changes in heat capacities and copper resistivity, and varying flow dynamics within the conductors. This analysis addresses all these factors in an approximate way. The result is much less accurate than a rigorous analysis. Results here could be in error as much as 30 to 40 percent. However, it is believed that this paper can still be very useful to the coil designer. Coil pressures and temperatures vs time into a quench are presented. Rate of helium vent, energy deposition in the coil, and depletion of magnetic stored energy are also presented. Peak pressures are high (about 43 MPa). This is due to the very long vent path length (446 m), small hydraulic diameters, and high current densities associated with ITER's cable-in-conduit design. The effects of these pressures as well as the ability of the coil to be self protecting during a quench are discussed. 3 refs., 1 fig., 3 tabs.

  15. Recent progress on the hydrogen storage properties of ZrCo-based alloys applied in International Thermonuclear Experimental Reactor (ITER

    Directory of Open Access Journals (Sweden)

    Feng Wang

    2017-02-01

    Full Text Available Under the development of International Thermonuclear Experimental Reactor (ITER system aimed at realizing the controllable fusion reaction to solve the energy crisis fundamentally, there is an urgent need to find an appropriate material for tritium handling. ZrCo alloy is considered to be a promising candidate for the storage and delivery of hydrogen isotopes due to the favorable characteristics such as low plateau pressure for absorption, high dissociation pressure at moderate temperature and better ability of trapping 3He. However, the hydrogen induced disproportionation and the slower recovery/deliverty rate of ZrCo-based alloys have limited their further application in ITER system. This review summarizes the efforts towards enhancing the hydrogen storage properties of ZrCo-based alloys including element substitution, surface modification, disproportionation mechanism investigation and the isotope effect study. Element substitution and surface modification play positive role to improve the anti-disproportionation ability and kinetic property of the alloys. However, the ZrCo-based alloys require to be further modified by more attempts such as new composition, novelty modification method or catalyst addition in order to better satisfy the application demands for tritium handling. Moreover, new insight for further understanding the inner disproportionation mechanisms of this material is needed by combining the advance characterization and theoretical analysis, which is in favor of addressing the disproportionation problem of the ZrCo-based alloys essentially.

  16. Commissioning of the STAR test section for experimental simulation of loss of coolant accident using the EC-208 instrumented fuel assembly of the IEA-R1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Maprelian, Eduardo; Torres, Walmir M.; Prado, Adelk C.; Umbehaun, Pedro E.; Franca, Renato L.; Santos, Samuel C.; Macedo, Luiz A.; Sabundjian, Gaiane, E-mail: emaprel@ipen.br, E-mail: wmtorres@ipen.br, E-mail: acprado@ipen.br, E-mail: umbehaun@ipen.br, E-mail: rlfranca@ipen.br, E-mail: samuelcs@ipen.br, E-mail: lamacedo@ipen.br, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SO (Brazil)

    2015-07-01

    The three basic safety functions of Research Reactors (RR) are the safe shutdown of the reactor, the proper cooling of the decay heat of the fuel elements and the confinement of radioactive materials. Compared to Nuclear Power Reactors, RR power release is small, yet its three safety functions must be met to ensure the integrity of the reactor. During a loss of coolant accident (LOCA) in pool type RR, partial or complete loss of pool water may occur, with consequent partial or complete uncovering of the fuel assemblies. In such an accident, the decay heat removal safety function must not be compromised. The Test Section for Experimental Simulation of Loss of Coolant Accident (STAR) is in commissioning phase. This test section will provide experimental data on partial and total uncovering of the EC-208 instrumented fuel assembly (IFA) irradiated in the IEA-R1. Experimental results will be useful in validation of computer codes for RR safety analysis, particularly on heat removal efficiency aspects (safety function) in accident conditions. STAR comprises a base on which is installed the IFA, the cylindrical stainless steel hull, the compressed air system for the test section emptying and refilling, and the instrumentation for temperature and level measurements. The commissioning tests or pre-operational check, consist of several preliminary tests to verify experimental procedures, the difficulties during assembling of STAR in the pool, the difficulties in control the emptying and refilling velocities, as well as, the repeatability capacity, tests of equipment, valves and systems and tests of instrumentation and data acquisition system. Safety, accuracy and easiness of operation will be checked. (author)

  17. Nitritation performance and biofilm development of co- and counter-diffusion biofilm reactors: Modeling and experimental comparison

    DEFF Research Database (Denmark)

    Wang, Rongchang; Terada, Akihiko; Lackner, Susanne

    2009-01-01

    A comparative study was conducted on the start-up performance and biofilm development in two different biofilm reactors with aim of obtaining partial nitritation. The reactors were both operated under oxygen limited conditions, but differed in geometry. While substrates (O-2, NH3) co...... results showed that the counter-diffusion biofilms developed faster and attained a larger maximum biofilm thickness than the co-diffusion biofilms. Under oxygen limited condition (DO

  18. Thermal-Hydraulic Analysis of an Experimental Reactor Cavity Cooling System with Air. Part I: Experiments; Part II: Separate Effects Tests and Modeling

    Energy Technology Data Exchange (ETDEWEB)

    Corradin, Michael [Univ. of Wisconsin, Madison, WI (United States). Dept. of Engineering Physics; Anderson, M. [Univ. of Wisconsin, Madison, WI (United States). Dept. of Engineering Physics; Muci, M. [Univ. of Wisconsin, Madison, WI (United States). Dept. of Engineering Physics; Hassan, Yassin [Texas A & M Univ., College Station, TX (United States); Dominguez, A. [Texas A & M Univ., College Station, TX (United States); Tokuhiro, Akira [Univ. of Idaho, Moscow, ID (United States); Hamman, K. [Univ. of Idaho, Moscow, ID (United States)

    2014-10-15

    This experimental study investigates the thermal hydraulic behavior and the heat removal performance for a scaled Reactor Cavity Cooling System (RCCS) with air. A quarter-scale RCCS facility was designed and built based on a full-scale General Atomics (GA) RCCS design concept for the Modular High Temperature Gas Reactor (MHTGR). The GA RCCS is a passive cooling system that draws in air to use as the cooling fluid to remove heat radiated from the reactor pressure vessel to the air-cooled riser tubes and discharged the heated air into the atmosphere. Scaling laws were used to preserve key aspects and to maintain similarity. The scaled air RCCS facility at UW-Madison is a quarter-scale reduced length experiment housing six riser ducts that represent a 9.5° sector slice of the full-scale GA air RCCS concept. Radiant heaters were used to simulate the heat radiation from the reactor pressure vessel. The maximum power that can be achieved with the radiant heaters is 40 kW with a peak heat flux of 25 kW per meter squared. The quarter-scale RCCS was run under different heat loading cases and operated successfully. Instabilities were observed in some experiments in which one of the two exhaust ducts experienced a flow reversal for a period of time. The data and analysis presented show that the RCCS has promising potential to be a decay heat removal system during an accident scenario.

  19. Experimental evaluation of two different types of reactors for CO2 removal from gaseous stream by bottom ash accelerated carbonation.

    Science.gov (United States)

    Lombardi, L; Carnevale, E A; Pecorini, I

    2016-12-01

    Low methane content landfill gas may be enriched by removing carbon dioxide. An innovative process, based on carbon dioxide capture and storage by means of accelerated carbonation of bottom ash is proposed and studied for the above purpose. Within this research framework we devoted a preliminary research activity to investigate the possibility of improving the way the contact between bottom ash and landfill gas takes place: this is the scope of the work reported in this paper. Two different types of reactors - fixed bed and rotating drum - were designed and constructed for this purpose. The process was investigated at laboratory scale. As the aim of this phase was the comparison of the performances of the two different reactors, we used a pure stream of CO2 to preliminarily evaluate the reactor behaviors in the most favorable condition for the process (i.e. maximum CO2 partial pressure at ambient condition). With respect to the simple fixed bed reactor concept, some modifications were proposed, consisting of separating the ash bed in three layers. With the three layer configuration we would like to reduce the possibility for the gas to follow preferential paths through the ash bed. However, the results showed that the process performances are not significantly influenced by the multiple layer arrangement. As an alternative to the fixed bed reactor, the rotating drum concept was selected in order to provide continuous mixing of the solids. Two operating parameters were considered and varied during the tests: the filling ratio and the rotating speed. Better performances were observed for lower filling ratio while the rotating speed showed minor importance. Finally the performances of the two reactors were compared. The rotating drum reactor is able to provide improved carbon dioxide removal with respect to the fixed bed one, especially when the rotating reactor is operated at low filling ratio values and slow rotating speed values. Comparing the carbon dioxide

  20. Dr Robert Aymar, Director of the International Thermonuclear Experimental Reactor (ITER), was nominated to succeed Professor Luciano Maiani as CERN's Director General, to take office on 1 January 2004.

    CERN Document Server

    2002-01-01

    Dr Robert Aymar, Director of the International Thermonuclear Experimental Reactor (ITER), was nominated to succeed Professor Luciano Maiani as CERN's Director General, to take office on 1 January 2004.

  1. Experimental study of local coolant hydrodynamics in TVS-Kvadrat PWR reactor fuel assembly using mixing spacer grids with different types of deflectors

    Directory of Open Access Journals (Sweden)

    S.M. Dmitriev

    2015-12-01

    Full Text Available Results of experimental studies of local hydrodynamic characteristics of coolant flow in fuel assemblies of RWR reactors using different types of mixing spacer grids are presented. Specific features and regularities of coolant flow in fuel pin bundles of TVS-KVADRAT fuel assemblies with different types of mixing spacer grids were revealed in the course of experiments. Analysis of space distribution of projections of absolute flow velocity allowed detailed description of coolant flow beyond the spacer grid with installation of three different types of deflectors. Optimal design of deflector for spacer grid of the TVS-KVADRAT fuel assembly in the standard cell in the area of guiding channels was identified. Results of studies of local hydrodynamics of coolant flow in the TVS-KVADRAT fuel assembly are accepted for subsequent practical application by the JSC Afrikantov Experimental Design Bureau for Mechanical Engineering (OKBM in the evaluations of thermal engineering reliability of PWR reactor cores and were included in the database for verification of computational fluid dynamic codes (CFD-codes and implementation of detailed cell array calculations of PWR reactor cores.

  2. Autothermal reforming of methane with integrated CO2 capture in a novel fluidized bed membrane reactor. Part 1: experimental demonstration

    NARCIS (Netherlands)

    Gallucci, F.; van Sint Annaland, M.; Kuipers, J.A.M.

    2008-01-01

    Two fluidized bed membrane reactor concepts for hydrogen production via autothermal reforming of methane with integrated CO2 capture are proposed. Ultra-pure hydrogen is obtained via hydrogen perm-selective Pd-based membranes, while the required reaction energy is supplied by oxidizing part of the

  3. Experimental study of a cocurrent upflow packed bed bubble column reactor: pressure drop, holdup and interfacial area

    NARCIS (Netherlands)

    Molga, E.J.; Westerterp, K.R.

    1997-01-01

    Gas¿liquid interfacial areas have been determined by means of chemically enhanced absorption of CO2 into DEA in a packed bed bubble column reactor with an inner diameter of 156 mm. The influence of the gas velocity and particle diameter on the interfacial areas, pressure drops and liquid holdups has

  4. Fluidised bed membrane reactor for ultrapure hydrogen production via methane steam reforming: Experimental demonstration and model validation

    NARCIS (Netherlands)

    Patil, C.S.; van Sint Annaland, M.; Kuipers, J.A.M.

    2007-01-01

    Hydrogen is emerging as a future alternative for mobile and stationary energy carriers in addition to its use in chemical and petrochemical applications. A novel multifunctional reactor concept has been developed for the production of ultrapure hydrogen View the MathML source from light hydrocarbons

  5. Development of safety analysis codes and experimental validation for a very high temperature gas-cooled reactor Final report

    Energy Technology Data Exchange (ETDEWEB)

    Chang Oh

    2006-03-01

    The very high-temperature gas-cooled reactor (VHTR) is envisioned as a single- or dual-purpose reactor for electricity and hydrogen generation. The concept has average coolant temperatures above 9000C and operational fuel temperatures above 12500C. The concept provides the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperature to support process heat applications, such as coal gasification, desalination or cogenerative processes, the VHTR’s higher temperatures allow broader applications, including thermochemical hydrogen production. However, the very high temperatures of this reactor concept can be detrimental to safety if a loss-of-coolant accident (LOCA) occurs. Following the loss of coolant through the break and coolant depressurization, air will enter the core through the break by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heatup of the reactor core and the release of toxic gasses (CO and CO2) and fission products. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. Prior to the start of this Korean/United States collaboration, no computer codes were available that had been sufficiently developed and validated to reliably simulate a LOCA in the VHTR. Therefore, we have worked for the past three years on developing and validating advanced computational methods for simulating LOCAs in a VHTR. Research Objectives As described above, a pipe break may lead to significant fuel damage and fission product release in the VHTR. The objectives of this Korean/United States collaboration were to develop and validate advanced computational methods for VHTR safety analysis. The methods that have been developed are now

  6. Experimental determination of spectral indices by scanning of fuel rod in the IPEN/MB-01 reactor; Determinacao experimental de indices espectrais por varredura gama de vareta combustivel no reator IPEN/MB-01

    Energy Technology Data Exchange (ETDEWEB)

    Fanaro, Leda Cristina Cabelo Bernardes

    2009-07-01

    In this work, the spectral indexes 28{rho}{sup *} and 25{delta}{sup *}, and gamma efficiency factor in the IPEN/MB-01 reactor were determined experimentally employing a rod scanning technique. In the case of 28{rho}{sup *}, this method has the advantage of eliminating most of the correction factors derived from the calculations. Only the fission yield factor and the relative fission rate in the {sup 235}U remain in the determination of the 25{delta}{sup *}. The experiments were performed with different thicknesses of cadmium sleeves: 0.55 mm, 1.10 mm and 2.20 mm. The final experimental uncertainty achieved in the experiment, less that 1 %, and the excellent geometrical and material data characterization of the IPEN/MB-01 reactor allow to use the results as benchmark for validate calculation methods and related nuclear data libraries. The comparison between calculated values and experimental values was performed by employing the MCNP-5 code and the nuclear data libraries: ENDF/B-VI.8, ENDF/B-VII.0, JENDL-3.3 and JEFF-3.1. The results demonstrate that the difference among libraries is very small. Also, the comparison between calculated values and experimental values shows that there has been considerable progress in the recent nuclear data libraries. The best result is obtained with ENDF/B-VII.0 nuclear data library, and the highest discrepancy was obtained with JEFF-3.1 and JENDL-3.3 nuclear data libraries. (author)

  7. Energy efficient electrocoagulation using a new flow column reactor to remove nitrate from drinking water - Experimental, statistical, and economic approach.

    Science.gov (United States)

    Hashim, Khalid S; Shaw, Andy; Al Khaddar, Rafid; Pedrola, Montserrat Ortoneda; Phipps, David

    2017-07-01

    In this investigation, a new bench-scale electrocoagulation reactor (FCER) has been applied for drinking water denitrification. FCER utilises the concepts of flow column to mix and aerate the water. The water being treated flows through the perforated aluminium disks electrodes, thereby efficiently mixing and aerating the water. As a result, FCER reduces the need for external stirring and aerating devices, which until now have been widely used in the electrocoagulation reactors. Therefore, FCER could be a promising cost-effective alternative to the traditional lab-scale EC reactors. A comprehensive study has been commenced to investigate the performance of the new reactor. This includes the application of FCER to remove nitrate from drinking water. Estimation of the produced amount of H2 gas and the yieldable energy from it, an estimation of its preliminary operating cost, and a SEM (scanning electron microscope) investigation of the influence of the EC process on the morphology of the surface of electrodes. Additionally, an empirical model was developed to reproduce the nitrate removal performance of the FCER. The results obtained indicated that the FCER reduced the nitrate concentration from 100 to 15 mg/L (World Health Organization limitations for infants) after 55 min of electrolysing at initial pH of 7, GBE of 5 mm, CD of 2 mA/cm2, and at operating cost of 0.455 US $/m3. Additionally, it was found that FCER emits H2 gas enough to generate a power of 1.36 kW/m3. Statistically, the relationship between the operating parameters and nitrate removal could be modelled with R2 of 0.848. The obtained SEM images showed a large number dents on anode's surface due to the production of aluminium hydroxides. Crown Copyright © 2017. Published by Elsevier Ltd. All rights reserved.

  8. Remote Reactor Monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Bernstein, Adam [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dazeley, Steve [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dobie, Doug [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Marleau, Peter [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Brennan, Jim [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gerling, Mark [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sumner, Matthew [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sweany, Melinda [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2014-10-21

    The overall goal of the WATCHMAN project is to experimentally demonstrate the potential of water Cerenkov antineutrino detectors as a tool for remote monitoring of nuclear reactors. In particular, the project seeks to field a large prototype gadolinium-doped, water-based antineutrino detector to demonstrate sensitivity to a power reactor at ~10 kilometer standoff using a kiloton scale detector. The technology under development, when fully realized at large scale, could provide remote near-real-time information about reactor existence and operational status for small operating nuclear reactors out to distances of many hundreds of kilometers.

  9. Determination of the Clean Air Delivery Rate (CADR) of Photocatalytic Oxidation (PCO) Purifiers for Indoor Air Pollutants Using a Closed-Loop Reactor. Part II: Experimental Results.

    Science.gov (United States)

    Héquet, Valérie; Batault, Frédéric; Raillard, Cécile; Thévenet, Frédéric; Le Coq, Laurence; Dumont, Éric

    2017-03-06

    The performances of a laboratory PhotoCatalytic Oxidation (PCO) device were determined using a recirculation closed-loop pilot reactor. The closed-loop system was modeled by associating equations related to two ideal reactors: a perfectly mixed reservoir with a volume of VR = 0.42 m³ and a plug flow system corresponding to the PCO device with a volume of VP = 5.6 × 10-3 m³. The PCO device was composed of a pleated photocatalytic filter (1100 cm²) and two 18-W UVA fluorescent tubes. The Clean Air Delivery Rate (CADR) of the apparatus was measured under different operating conditions. The influence of three operating parameters was investigated: (i) light irradiance I from 0.10 to 2.0 mW·cm-2; (ii) air velocity v from 0.2 to 1.9 m·s-1; and (iii) initial toluene concentration C₀ (200, 600, 1000 and 4700 ppbv). The results showed that the conditions needed to apply a first-order decay model to the experimental data (described in Part I) were fulfilled. The CADR values, ranging from 0.35 to 3.95 m³·h-1, were mainly dependent on the light irradiance intensity. A square root influence of the light irradiance was observed. Although the CADR of the PCO device inserted in the closed-loop reactor did not theoretically depend on the flow rate (see Part I), the experimental results did not enable the confirmation of this prediction. The initial concentration was also a parameter influencing the CADR, as well as the toluene degradation rate. The maximum degradation rate rmax ranged from 342 to 4894 ppbv/h. Finally, this study evidenced that a recirculation closed-loop pilot could be used to develop a reliable standard test method to assess the effectiveness of PCO devices.

  10. Determination of the Clean Air Delivery Rate (CADR of Photocatalytic Oxidation (PCO Purifiers for Indoor Air Pollutants Using a Closed-Loop Reactor. Part II: Experimental Results

    Directory of Open Access Journals (Sweden)

    Valérie Héquet

    2017-03-01

    Full Text Available The performances of a laboratory PhotoCatalytic Oxidation (PCO device were determined using a recirculation closed-loop pilot reactor. The closed-loop system was modeled by associating equations related to two ideal reactors: a perfectly mixed reservoir with a volume of VR = 0.42 m3 and a plug flow system corresponding to the PCO device with a volume of VP = 5.6 × 10−3 m3. The PCO device was composed of a pleated photocatalytic filter (1100 cm2 and two 18-W UVA fluorescent tubes. The Clean Air Delivery Rate (CADR of the apparatus was measured under different operating conditions. The influence of three operating parameters was investigated: (i light irradiance I from 0.10 to 2.0 mW·cm−2; (ii air velocity v from 0.2 to 1.9 m·s−1; and (iii initial toluene concentration C0 (200, 600, 1000 and 4700 ppbv. The results showed that the conditions needed to apply a first-order decay model to the experimental data (described in Part I were fulfilled. The CADR values, ranging from 0.35 to 3.95 m3·h−1, were mainly dependent on the light irradiance intensity. A square root influence of the light irradiance was observed. Although the CADR of the PCO device inserted in the closed-loop reactor did not theoretically depend on the flow rate (see Part I, the experimental results did not enable the confirmation of this prediction. The initial concentration was also a parameter influencing the CADR, as well as the toluene degradation rate. The maximum degradation rate rmax ranged from 342 to 4894 ppbv/h. Finally, this study evidenced that a recirculation closed-loop pilot could be used to develop a reliable standard test method to assess the effectiveness of PCO devices.

  11. Experimental research on molten salt thermofluid technology using a high-temperature molten salt loop applied for a fusion reactor Flibe blanket

    Energy Technology Data Exchange (ETDEWEB)

    Toda, Saburo; Chiba, Shinya E-mail: schiba@karma.qse.tohoku.ac.jp; Yuki, Kazuhisa; Omae, Masahiro; Sagara, Akio

    2002-12-01

    Experimental research on molten salt thermofluid technology using a high-temperature molten salt loop (MSL) is described in this paper. The MSL was designed to be able to use Flibe as a coolant, however, a simulant, heat transfer salt (HTS) has to be used alternatively since Flibe is difficult to operate under avoiding a biohazard of Be. Experiment on heat-transfer enhancement, that is required for applying to cool the high heat flux components of fusion reactors, is ongoing. Preliminary experimental results showed that an internal structure of a mixing chamber in the MSL was important to obtain accurate bulk temperatures under severe thermal conditions. For operating the loop, careful handling are needed to proceed how to melt the salt and to circulate it in starting the operation of the MSL. It is concluded that several improvements proposed from the present experiences should be applied for the future Flibe operation.

  12. 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...

  13. Development of Flow-Through Polymeric Membrane Reactor for Liquid Phase Reactions: Experimental Investigation and Mathematical Modeling

    Directory of Open Access Journals (Sweden)

    Endalkachew Chanie Mengistie

    2017-01-01

    Full Text Available Incorporating metal nanoparticles into polymer membranes can endow the membranes with additional functions. This work explores the development of catalytic polymer membrane through synthesis of palladium nanoparticles based on the approaches of intermatrix synthesis (IMS inside surface functionalized polyethersulfone (PES membrane and its application to liquid phase reactions. Flat sheet PES membranes have been successfully modified via UV-induced graft polymerization of acrylic acid monomer. Palladium nanoparticles have been synthesized by chemical reduction of palladium precursor loaded on surface modified membranes, an approach to the design of membranes modified with nanomaterials. The catalytic performances of the nanoparticle incorporated membranes have been evaluated by the liquid phase reduction of p-nitrophenol using NaBH4 as a reductant in flow-through membrane reactor configuration. The nanocomposite membranes containing palladium nanoparticles were catalytically efficient in achieving a nearly 100% conversion and the conversion was found to be dependent on the flux, amount of catalyst, and initial concentration of nitrophenol. The proposed mathematical model equation represents satisfactorily the reaction and transport phenomena in flow-through catalytic membrane reactor.

  14. Development of Safety Analysis Codes and Experimental Validation for a Very High Temperature Gas-Cooled Reactor - FY-05 Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    Chang Oh

    2005-09-01

    The very high temperature gas-cooled reactors (VHTGRs) are those concepts that have average coolant temperatures above 9000C or operational fuel temperatures above 12500C. These concepts provide the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation and nuclear hydrogen generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperatures to support process heat applications, such as desalination and cogeneration, the VHTGR’s higher temperatures are suitable for particular applications such as thermochemical hydrogen production. However, the high temperature operation can be detrimental to safety following a loss-of-coolant accident (LOCA) initiated by pipe breaks caused by seismic or other events. Following the loss of coolant through the break and coolant depressurization, air from the containment will enter the core by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structures and fuel. The oxidation will release heat and accelerate the heatup of the reactor core. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. The Idaho National Laboratory (INL) has investigated this event for the past three years for the HTGR. However, the computer codes used, and in fact none of the world’s computer codes, have been sufficiently developed and validated to reliably predict this event. New code development, improvement of the existing codes, and experimental validation are imperative to narrow the uncertainty in the predictions of this type of accident. The objectives of this Korean/United States collaboration are to develop advanced computational methods for VHTGR safety analysis codes and to validate these computer codes.

  15. Experimental and modeling study of the effect of CO and H2 on the urea DeNO(x) process in a 150kW laboratory reactor.

    Science.gov (United States)

    Javed, M Tayyeb; Nimmo, W; Gibbs, B M

    2008-01-01

    An experimental and modeling investigation has been performed to study the effect of process additives, H2 and CO on NO(x) removal from flue gases by a selective non-catalytic reduction process using urea as a reducing agent. Experiments were performed with a flow reactor in which flue gas was generated by the combustion of propane in air at 3% excess oxygen and the desired levels of initial NO(x) (500ppm) were achieved by doping the flame with ammonia. Experiments were performed throughout the temperature range of interest, i.e. from 850 to 1200 degrees C for investigation of the effects of the process additives on the performance of aqueous urea DeNO(x). Subsequently, computational kinetic modeling with SENKIN code was performed to analyze the performance of urea providing a direct comparison of modeling prediction with experimental measurements. With CO addition, a downwards shift of 215 degrees C in the peak reduction temperature from 1125 to 910 degrees C was observed during the experimentation while the kinetic modeling suggests it to be 150 degrees C, i.e. from 1020 to 870 degrees C. The addition of H2 impairs the peak NO(x) reduction but suggests a low temperature application of the process. A downward shift of 250 degrees C in the peak reduction temperature, from 1020 to 770 degrees C, was observed during kinetic modeling studies. The kinetic modeling shows a good qualitative agreement with the experimental observations and reveals additional information about the process.

  16. Sequential Aeration of Membrane-Aerated Biofilm Reactors for High-Rate Autotrophic Nitrogen Removal: Experimental Demonstration

    DEFF Research Database (Denmark)

    Pellicer i Nàcher, Carles; Sun, Sheng-Peng; Lackner, Susanne

    2010-01-01

    One-stage autotrophic nitrogen (N) removal, requiring the simultaneous activity of aerobic and anaerobic ammonium oxidizing bacteria (AOB and AnAOB), can be obtained in spatially redox-stratified biofilms. However, previous experience with Membrane-Aerated Biofilm Reactors (MABRs) has revealed......S rRNA gene confirmed that sequential aeration, even at elevated average O2 loads, stimulated the abundance of AnAOB and AOB and prevented the increase in NOB. Nitrous oxide (N2O) emissions were 100-fold lower compared to other anaerobic ammonium oxidation (Anammox)-nitritation systems. Hence...... a difficulty in reducing the abundance and activity of nitrite oxidizing bacteria (NOB), which drastically lowers process efficiency. Here we show how sequential aeration is an effective strategy to attain autotrophic N removal in MABRs: Two separate MABRs, which displayed limited or no N removal under...

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

    Directory of Open Access Journals (Sweden)

    Massimo Raboni

    2013-12-01

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

  18. Defluoridation of drinking water using a new flow column-electrocoagulation reactor (FCER) - Experimental, statistical, and economic approach.

    Science.gov (United States)

    Hashim, Khalid S; Shaw, Andy; Al Khaddar, Rafid; Ortoneda Pedrola, Montserrat; Phipps, David

    2017-07-15

    A new batch, flow column electrocoagulation reactor (FCER) that utilises a perforated plate flow column as a mixer has been used to remove fluoride from drinking water. A comprehensive study has been carried out to assess its performance. The efficiency of fluoride removal (R%) as a function of key operational parameters such as initial pH, detention time (t), current density (CD), inter-electrode distance (ID) and initial concentration (C 0 ) has been examined and an empirical model has been developed. A scanning electron microscopy (SEM) investigation of the influence of the EC process on morphology of the surface of the aluminium electrodes, showed the erosion caused by aluminium loss. A preliminary estimation of the reactor's operating cost is suggested, allowing for the energy from recycling of hydrogen gas hydrogen gas produced amount. The results obtained showed that 98% of fluoride was removed within 25 min of electrolysis at pH of 6, ID of 5 mm, and CD of 2 mA/cm 2 . The general relationship between fluoride removal and operating parameters could be described by a linear model with R 2 of 0.823. The contribution of the operating parameters to the suggested model followed the order: t > CD > C 0  > ID > pH. The SEM images obtained showed that, after the EC process, the surface of the anodes, became non-uniform with a large number of irregularities due to the generation of aluminium hydroxides. It is suggested that these do not materially affect the performance. A provisional estimate of the operating cost was 0.379 US $/m 3 . Additionally, it has been found that 0.6 kW/m 3 is potentially recoverable from the H 2 gas. Crown Copyright © 2017. Published by Elsevier Ltd. All rights reserved.

  19. Development of BWR regional stability experimental facility SIRIUS-F, which simulates thermal-hydraulics-neutronics coupling in reactor core, and stability evaluation of ABWR

    Energy Technology Data Exchange (ETDEWEB)

    Masahiro Furuya; Fumio Inada [Central Research Institute of Electric Power Industry (CRIEPI) 2-11-1 Iwado-kita, Komae, Tokyo 201-8511 (Japan); Takanori Fukahori [Global Nuclear Fuel Japan (GNF-J) 2-3-1 Uchikawa, Yokosuka, Kanagawa 239-0836 (Japan); Shinya Mizokami [Tokyo Electric Power Company (TEPCO) 1-1-3 Uchisaiwai-cho, Chiyoda, Tokyo 100-0011 (Japan)

    2005-07-01

    Full text of publication follows: The SIRIUS-F facility was designed and constructed for highly accurate simulation of channel, core-wide and regional instabilities of an ABWR. A real-time simulation is performed for the modal-point kinetics of reactor neutronics and fuel-rod conduction on the basis of a measured void fraction in a reactor core section of the facility. A noise analysis method was performed to calculate decay ratios from dominant poles of transfer function on the basis of the AR method by applying time series of a core inlet flow rate. By utilizing this method, one can estimate stability at any specific operating point online without assuming excess conservative conditions. Channel and regional stability experiments were conducted for a wide range of operating conditions including maximum power points along the minimum pump speed line and the natural circulation line of the ABWR. The decay ratios and the resonance frequencies are in good agreement with those from the design analysis code, ODYSY. The SIRIUS-F experimental results demonstrated stability characteristics such as a stabilizing effect of the power, and reviled a sufficiently large stability margin even under hypothetical conditions of power enlargement. (authors)

  20. Experimental study and nuclear model calculations on the 192Os(p,n)192Ir reaction: Comparison of reactor and cyclotron production of the therapeutic radionuclide 192Ir.

    Science.gov (United States)

    Hilgers, K; Sudár, S; Qaim, S M

    2005-07-01

    In a search for an alternative route of production of the important therapeutic radionuclide (192)Ir (T(1/2)=78.83 d), the excitation function of the reaction (192)Os(p,n)(192)Ir was investigated from its threshold up to 20 MeV. Thin samples of enriched (192)Os were obtained by electrodeposition on Ni, and the conventional stacked-foil technique was used for cross section measurements. The experimental data were compared with the results of theoretical calculations using the codes EMPIRE-II and ALICE-IPPE. Good agreement was found with EMPIRE-II, but slightly less with the ALICE-IPPE calculations. The theoretical thick target yield of (192)Ir over the energy range E(p)=16-->8 MeV amounts to only 0.16MBq/muA.h. A comparison of the reactor and cyclotron production methods is given. In terms of yield and radionuclidic purity of (192)Ir the reactor method appears to be superior; the only advantage of the cyclotron method could be the higher specific activity of the product.

  1. Experimental study and nuclear model calculations on the $^{192}Os (p, n)^{192}$Ir reaction Comparison of reactor and cyclotron production of the therapeutic radionuclide $^{192}$Ir

    CERN Document Server

    Hilgers, K; Sudar, S; 10.1016/j.apradiso.2004.12.010

    2005-01-01

    In a search for an alternative route of production of the important therapeutic radionuclide /sup 192/Ir (T/sub 1/2/=78.83 d), the excitation function of the reaction /sup 192/Os(p, n)/sup 192/Ir was investigated from its threshold up to 20MeV. Thin samples of enriched /sup 192/Os were obtained by electrodeposition on Ni, and the conventional stacked-foil technique was used for cross section measurements. The experimental data were compared with the results of theoretical calculations using the codes EMPIRE-II and ALICE-IPPE. Good agreement was found with EMPIRE-II, but slightly less with the ALICE-IPPE calculations. The theoretical thick target yield of /sup 192/Ir over the energy range E/sub p/=16 to 8MeV amounts to only 0.16MBq/ mu A.h. A comparison of the reactor and cyclotron production methods is given. In terms of yield and radionuclidic purity of /sup 192/Ir the reactor method appears to be superior; the only advantage of the cyclotron method could be the higher specific activity of the product.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-07-28

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

  3. Should Degree Programs in Biomedical and Health Informatics be Dedicated or Integrated? : Reflections and Recommendations after more than 40 Years of Medical Informatics Education at TU Braunschweig, including 10 Years of B.Sc. and 15 Years of M.Sc. Integrated Degree Curricula.

    Science.gov (United States)

    Haux, Reinhold; Marschollek, Michael; Wolf, Klaus-Hendrik; Zeisberg, Ute

    2017-07-01

    Education in biomedical and health informatics (BMHI) has been established in many countries throughout the world. For degree programs in BMHI we can distinguish between those that are completely stand-alone or dedicated to the discipline vs. those that are integrated within another program. After running integrated degree medical informatics programs at TU Braunschweig for 10 years at the B.Sc. and for 15 years at the M.Sc level, we (1) report about this educational approach, (2) analyze recommendations on, implementations of, and experiences with degree educational programs in BMHI worldwide, (3) summarize our lessons learned with the integrated approach at TU Braunschweig, and (4) suggest an answer to the question, whether degree programs in biomedical and health informatics should be dedicated or integrated. According to our experience at TU Braunschweig and based on our analysis of publications, there is a clear dominance of dedicated degree programs in BMHI. The specialization in medical informatics within a computer science program, as offered at TU Braunschweig, may be a good way of implementing an integrated, informatics-based approach to medical informatics, in particular if a dual degree option can be chosen. The option of curricula leading to double degrees, i.e. in this case to two separate degrees in computer science and in medical informatics might, however, be a better solution.

  4. Experimental and numerical investigation of gas/liquid phase boundaries representing the reference level for hydrostatic level measurements in boiling water reactors; Experimentelle und numerische Untersuchung von Gas/Liquid-Phasengrenzflaechen als Referenzwert fuer die hydrostatische Fuellstandsmessung in Siedewasserreaktoren

    Energy Technology Data Exchange (ETDEWEB)

    Schulz, Stephan

    2013-12-17

    The experimental and numerical investigation of gas/liquid phase boundaries representing the reference level for hydrostatic level measurements in boiling water reactors is considered as relevant for reactor safety research. The experiments allow a quantification of the transition processes in hydrostatic level measurement devices that were up to now only assessed by phenomenological descriptions. Experimental studies covered the topology and stability of water/vapor phase boundaries and the numerical description using CFD codes, including modeling of the surface topology and modeling of the heat and mass transport.

  5. Reactor Safety Research Programs

    Energy Technology Data Exchange (ETDEWEB)

    Edler, S. K.

    1981-07-01

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

  6. Experimental and numerical investigation of coolant mixing in a model of reactor pressure vessel down-comer and in cold leg inlets

    Directory of Open Access Journals (Sweden)

    Hutli Ezddin

    2017-01-01

    Full Text Available Thermal fatigue and pressurized thermal shock phenomena are the main problems for the reactor pressure vessel and the T-junctions both of them depend on the mixing of the coolant. The mixing process, flow and temperature distribution has been investigated experimentally using particle image velocimetry, laser induced fluorescence, and simulated by CFD tools. The obtained results showed that the ratio of flow rate between the main pipe and the branch pipe has a big influence on the mixing process. The particle image velocimetry/planar laser-induced fluorescence measurements technologies proved to be suitable for the investigation of turbulent mixing in the complicated flow system: both velocity and temperature distribution are important parameters in the determination of thermal fatigue and pressurized thermal shock. Results of the applied these techniques showed that both of them can be used as a good provider for data base and to validate CFD results.

  7. Experimental Investigation of the Root Cause Mechanism and Effectiveness of Mitigating Actions for Axial Offset Anomaly in Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Said Abdel-Khalik

    2005-07-02

    Axial offset anomaly (AOA) in pressurized water reactors refers to the presence of a significantly larger measured negative axial offset deviation than predicted by core design calculations. The neutron flux depression in the upper half of high-power rods experiencing significant subcooled boiling is believed to be caused by the concentration of boron species within the crud layer formed on the cladding surface. Recent investigations of the root-cause mechanism for AOA [1,2] suggest that boron build-up on the fuel is caused by precipitation of lithium metaborate (LiBO2) within the crud in regions of subcooled boiling. Indirect evidence in support of this hypothesis was inferred from operating experience at Callaway, where lithium return and hide-out were, respectively, observed following power reductions and power increases when AOA was present. However, direct evidence of lithium metaborate precipitation within the crud has, heretofore, not been shown because of its retrograde solubility. To this end, this investigation has been undertaken in order to directly verify or refute the proposed root-cause mechanism of AOA, and examine the effectiveness of possible mitigating actions to limit its impact in high power PWR cores.

  8. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2001-04-01

    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised.

  9. 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

  10. Development and verification test of integral reactor major components - Development of MCP impeller design, performance prediction code and experimental verification

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Myung Kyoon; Oh, Woo Hyoung; Song, Jae Wook [Korea Advanced Institute of Science and Technology, Taejon (Korea)

    1999-03-01

    The present study is aimed at developing a computational code for design and performance prediction of an axial-flow pump. The proposed performance prediction method is tested against a model axial-flow pump streamline curvature method. The preliminary design is made by using the ideal velocity triangles at inlet and exit and the three dimensional blade shape is calculated by employing the free vortex design method. Then the detailed blading design is carried out by using experimental database of double circular arc cambered hydrofoils. To computationally determine the design incidence, deviation, blade camber, solidity and stagger angle, a number of correlation equations are developed form the experimental database and a theorical formula for the lift coefficient is adopted. A total of 8 equations are solved iteratively using an under-relaxation factor. An experimental measurement is conducted under a non-cavitating condition to obtain the off-design performance curve and also a cavitation test is carried out by reducing the suction pressure. The experimental results are very satisfactorily compared with the predictions by the streamline curvature method. 28 refs., 26 figs., 11 tabs. (Author)

  11. Fluid flow separation in a reactor pressure vessel during an ECC injection. Single phase flow and two phase flow (air-water) experimental results

    Energy Technology Data Exchange (ETDEWEB)

    Thierry Bichet; Alain Martin [EDF - Research and Development Division - Fluid Mechanics and Heat Transfert 6, quai Watier - B.P. 49 - 78401 Chatou CEDEX 01 (France); Frederic Beaud [EDF/ Industry - Basic Design Department., 12-14, Avenue Dutrievoz 69628 Villeurbanne CEDEX (France)

    2005-07-01

    Full text of publication follows: Within the framework of the nuclear power plant lifetime issue, the assessment of the French 900 MWe (3-loops) series reactor pressure vessel (RPV) integrity has been performed. A simplified analysis has shown that the most severe loading conditions are given by the small break loss of coolant accidents due to the pressurized injection of cold water (9 deg. C) into the cold leg and down comer of the RPV. During these transient scenarios, single or two-phase (uncovered cold leg) flows have been shown in the cold leg, depending on the crack size and RPV model (900 MWe or 1300 MWe). An experimental study has been carried out, on the one hand, to consolidate the numerical results obtained with CFD home code (Code-Saturne) which mainly showed the stratified flow in the cold leg and the fluid flow separation and its oscillations in the down comer during a single phase scenario. These physical phenomena are important for the thermal RPV loading assessment. On the other hand, the absence of experimental two-phase data necessitated to carry out an experimental study around the mixing area behavior (free surface, stratified flow) during an ECC injection with an uncovered cold leg. The new EDF R and D mock up, called HYBISCUS, is a facility which is made out of Plexiglas (atmosphere pressure) and represents a half scale CP0 geometry with one cold leg and part of the down comer. The mock up modularity allows us to insert representative ECC nozzles and a thermal shield. In reference to the reactor scenarios, the experimental operating conditions are derived from the conservation of the density effects (Froude number). For that, a heated salted water flow is used to represent the ECC injection whereas water represents the cold leg fluid. This mock up has been defined in order to represent single phase flow (cold leg and down comer full of water) or two-phase flow (uncovered cold leg) ECC scenarios. This paper reports experimental results

  12. Reactor flux calculations

    Energy Technology Data Exchange (ETDEWEB)

    Lhuillier, D. [Commissariat à l' Énergie Atomique et aux Énergies Alternatives, Centre de Saclay, IRFU/SPhN, 91191 Gif-sur-Yvette (France)

    2013-02-15

    The status of the prediction of reactor anti-neutrino spectra is presented. The most accurate method is still the conversion of total β spectra of fissionning isotopes as measured at research reactors. Recent re-evaluations of the conversion process led to an increased predicted flux by few percent and were at the origin of the so-called reactor anomaly. The up to date predictions are presented with their main sources of error. Perspectives are given on the complementary ab-initio predictions and upcoming experimental cross-checks of the predicted spectrum shape.

  13. Experimental studies and computational benchmark on heavy liquid metal natural circulation in a full height-scale test loop for small modular reactors

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Yong-Hoon, E-mail: chaotics@snu.ac.kr [Department of Energy Systems Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 08826 (Korea, Republic of); Cho, Jaehyun [Korea Atomic Energy Research Institute, 111 Daedeok-daero, 989 Beon-gil, Yuseong-gu, Daejeon 34057 (Korea, Republic of); Lee, Jueun; Ju, Heejae; Sohn, Sungjune; Kim, Yeji; Noh, Hyunyub; Hwang, Il Soon [Department of Energy Systems Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 08826 (Korea, Republic of)

    2017-05-15

    Highlights: • Experimental studies on natural circulation for lead-bismuth eutectic were conducted. • Adiabatic wall boundaries conditions were established by compensating heat loss. • Computational benchmark with a system thermal-hydraulics code was performed. • Numerical simulation and experiment showed good agreement in mass flow rate. • An empirical relation was formulated for mass flow rate with experimental data. - Abstract: In order to test the enhanced safety of small lead-cooled fast reactors, lead-bismuth eutectic (LBE) natural circulation characteristics have been studied. We present results of experiments with LBE non-isothermal natural circulation in a full-height scale test loop, HELIOS (heavy eutectic liquid metal loop for integral test of operability and safety of PEACER), and the validation of a system thermal-hydraulics code. The experimental studies on LBE were conducted under steady state as a function of core power conditions from 9.8 kW to 33.6 kW. Local surface heaters on the main loop were activated and finely tuned by trial-and-error approach to make adiabatic wall boundary conditions. A thermal-hydraulic system code MARS-LBE was validated by using the well-defined benchmark data. It was found that the predictions were mostly in good agreement with the experimental data in terms of mass flow rate and temperature difference that were both within 7%, respectively. With experiment results, an empirical relation predicting mass flow rate at a non-isothermal, adiabatic condition in HELIOS was derived.

  14. Local Neutron Flux Distribution Measurements by Wire-Dosimetry in the AMMON Experimental Program in the EOLE Reactor

    Directory of Open Access Journals (Sweden)

    Gruel A.

    2016-01-01

    Full Text Available Dosimetry measurements were carried out during the AMMON experimental program, in the EOLE facility. Al-0.1 wt% Au wires were positioned along curved fuel plates of JHR-type assemblies to investigate the azimuthal and axial gold capture rate profiles, directly linked to the thermal and epithermal flux. After irradiation, wires were cut into small segments (a few mm, and the gold capture rate of each part was measured by gamma spectrometry on the MADERE platform. This paper presents results in the “hafnium” configuration, and more specifically the azimuthal flux profile characterization. The final uncertainty on each measured wire lies below 1% (at 2 standard deviations. Experimental profiles are in a good agreement against Monte Carlo calculations, and the 4% capture rate increase at the plate edge is well observed. The flux dissymmetry due to assembly position in the core is also measured, and shows a 10% discrepancy between the two edges of the plate.

  15. Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    1968-09-01

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

  16. The Fast-spectrum Transmutation Experimental Facility FASTEF: Main design achievements (part 2: Reactor building design and plant layout) within the FP7-CDT collaborative project of the European Commission

    Energy Technology Data Exchange (ETDEWEB)

    De Bruyn, D.; Engelen, J. [Belgian Nuclear Research Centre SCK CEN, Boeretang 200, 2400 Mol (Belgium); Ortega, A.; Aguado, M. P. [Empresarios Agrupados A.I.E., Magallanes 3, 28015 Madrid (Spain)

    2012-07-01

    MYRRHA (Multi-purpose hybrid Research Reactor for High-tech Applications) is the flexible experimental accelerator-driven system (ADS) in development at SCK-CEN in replacement of its material testing reactor BR2. SCK-CEN in association with 17 European partners from industry, research centres and academia, responded to the FP7 (Seventh Framework Programme) call from the European Commission to establish a Central Design Team (CDT) for the design of a Fast Spectrum Transmutation Experimental Facility (FASTEF) able to demonstrate efficient transmutation and associated technology through a system working in subcritical and/or critical mode. The project has started on April 01, 2009 for a period of three years. In this paper, we present the latest concept of the reactor building and the plant layout. The FASTEF facility has evolved quite a lot since the intermediate reporting done at the ICAPP'10 and ICAPP'11 conferences 1,2. Many iterations have been performed to take into account the safety requirements. The present configuration enables an easy operation and maintenance of the facility, including the possibility to change large components of the reactor. In a companion paper 3, we present the latest configuration of the reactor core and primary system. (authors)

  17. Progress of thermal hydraulic evaluation methods and experimental studies on a sodium-cooled fast reactor and its safety in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kamide, Hideki, E-mail: kamide.hideki@jaea.go.jp; Ohshima, Hiroyuki, E-mail: ohshima.hiroyuki@jaea.go.jp; Sakai, Takaaki, E-mail: sakai.takaaki@jaea.go.jp; Tanaka, Masaaki, E-mail: tanaka.masaaki@jaea.go.jp

    2017-02-15

    Highlights: • Thermal hydraulic issues for safety design criteria of sodium cooled fast reactors. • Measurement of velocity data in a subchannel surrounded by wire wrapped fuel-pins. • Statistical evaluation of core hot spot temperature during natural circulation. • Simulation of dynamics of molten fuel pool in a core disruptive accident. • V&V procedure of a multi-dimensional thermal hydraulic code on thermal striping. - Abstract: In the framework of the Generation-IV International Forum, the safety design criteria (SDC) incorporating safety-related R&D results on innovative technologies and lessons learned from Fukushima Dai-ichi nuclear power plants accident has been established to provide the set of general criteria for the safety designs of structures, systems and components of Generation-IV Sodium-cooled Fast Reactors (Gen-IV SFRs). A number of thermal-hydraulic evaluations are necessary to meet the concept of the criteria in the design studies of Gen-IV SFRs. This paper focuses on four kinds of thermal-hydraulic issues associated with the SDC, i.e., fuel subassembly thermal-hydraulics, natural circulation decay heat removal, core disruptive accidents, and thermal striping. Progress of evaluation methods on these issues is shown with activities on verification and validation (V&V) and experimental studies towards commercialization of SFR in Japan. These evaluation methods are planned to be eventually integrated into a comprehensive numerical simulation system that can be applied to all possible phenomena in SFR systems and that can be expected to become an effective tool for the development of human resource and the handing our knowledge and technologies down.

  18. Analysis and interpretation of residence time distribution experimental curves in FM01-LC reactor using axial dispersion and plug dispersion exchange models with closed-closed boundary conditions

    Energy Technology Data Exchange (ETDEWEB)

    Rivera, Fernando F. [Departamento de Quimica, Universidad Autonoma Metropolitana-Iztapalapa, San Rafael Atlixco 186, C.P. 09340, Mexico, D.F. (Mexico); Cruz-Diaz, Martin R., E-mail: mcruz@tese.edu.m [Division de Quimica y Bioquimica, Tecnologico de Estudios Superiores de Ecatepec, Av. Tecnologico S/N Esq. Av. Hank Gonzalez, Valle de Anahuac, C.P. 55120, Ecatepec, Edo. de Mex (Mexico); Rivero, Eligio P. [Departamento de Ingenieria y Tecnologia, Universidad Nacional Autonoma de Mexico, Facultad de Estudios Superiores Cuautitlan, Av. Primero de Mayo, Cuautitlan Izcalli, C.P. 54740, Edo. de Mex (Mexico); Gonzalez, Ignacio [Departamento de Quimica, Universidad Autonoma Metropolitana-Iztapalapa, San Rafael Atlixco 186, C.P. 09340, Mexico, D.F. (Mexico)

    2010-12-15

    The liquid phase mixing flow pattern at low (20 < Re < 120) and intermediate liquid flow rate (120 < Re < 400) was studied by means of residence time distribution (RTD) experimental curve in an up-flow Filter Press electrochemical reactor (FM01-LC) bench scale. For this purpose, a plastic turbulence promoter was used with stainless-steel and platinised titanium structural meshes as electrodes in channel configuration. To visualize and determine the mixing flow pattern in the liquid phase, the stimulus-response technique was employed using dextran blue (D{sub M} = 1.058 x 10{sup -11} m{sup 2} s{sup -1}, 25 {sup o}C, in water) as model tracer. A theoretical analysis and approximation RTD experimental curves with axial dispersion model (ADM) and plug dispersion exchange model (PDE), with 'closed-closed vessel' boundary conditions were used in order to establish a better approximation of the axial dispersion, stagnant zones, channelling and by-pass (preference flow) effects present at low and intermediate Re. RTD curves show that the liquid flow pattern in the FM01-LC deviates considerably from axial dispersion model at low Re, where the FM01-LC exhibits large channelling, stagnant zones, and dead zone. The PDE model represents fairly this deviation from ideal flow (less dead zone).

  19. Experimental study on fluid mixing in a fuel subassembly of a fast reactor. Temperature field around heated pin with cross flow

    Energy Technology Data Exchange (ETDEWEB)

    Miyakoshi, Hiroyuki; Kamide, Hideki; Tanaka, Masaaki; Yamamoto, Kazuhiro [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center

    2002-03-01

    High burnup of the core is one of means to reduce the cost of a fast reactor and fuel cycle system. However, it is not enough to investigate thermohydraulics in the core, in which fuel and wrapper tube are deformed due to irradiation under high burnup condition. In this study, sodium experiment was performed to investigate fluid mixing in a wire-wrapped 37-pin subassembly model, which had local blockage and cross flow around the blockage. Such cross flow is one of elements of thermohydraulics in a deformed subassembly. The experimental results is useful to develop numerical simulation method for the deformed subassembly. Seven pins, each had different relative position to the blockage, were heated individually in the experiments. Temperature field in the subassembly was measured. Influences of the flow rate and heater power were also examined. A horizontal cross flow occurred in upstream region toward the blockage. It was observed that the temperature field was influenced by this cross flow. The measured temperature field showed that there was a bypass flow around the blockage, which flowed toward the center of subassembly. The cross flow due to the bypass flow reached the 3rd row of pins from the blockage. The swirl flow, resulted from the spacer wire, also influenced the temperature field. The obtained experimental data will be used to develop and verify a numerical simulation method for a deformed fuel subassembly. (author)

  20. REVIEW OF EXPERIMENTAL CAPABILITIES AND HYDRODYNAMIC DATA FOR VALIDATION OF CFD-BASED PREDICTIONS FOR SLURRY BUBBLE COLUMN REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Donna Post Guillen; Daniel S. Wendt; Steven P. Antal; Michael Z. Podowski

    2007-11-01

    The purpose of this paper is to document the review of several open-literature sources of both experimental capabilities and published hydrodynamic data to aid in the validation of a Computational Fluid Dynamics (CFD) based model of a slurry bubble column (SBC). The review included searching the Web of Science, ISI Proceedings, and Inspec databases, internet searches as well as other open literature sources. The goal of this study was to identify available experimental facilities and relevant data. Integral (i.e., pertaining to the SBC system), as well as fundamental (i.e., separate effects are considered), data are included in the scope of this effort. The fundamental data is needed to validate the individual mechanistic models or closure laws used in a Computational Multiphase Fluid Dynamics (CMFD) simulation of a SBC. The fundamental data is generally focused on simple geometries (i.e., flow between parallel plates or cylindrical pipes) or custom-designed tests to focus on selected interfacial phenomena. Integral data covers the operation of a SBC as a system with coupled effects. This work highlights selected experimental capabilities and data for the purpose of SBC model validation, and is not meant to be an exhaustive summary.

  1. REVIEW OF EXPERIMENTAL CAPABILITIES AND HYDRODYNAMIC DATA FOR VALIDATION OF CFD BASED PREDICTIONS FOR SLURRY BUBBLE COLUMN REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Donna Post Guillen; Daniel S. Wendt

    2007-11-01

    The purpose of this paper is to document the review of several open-literature sources of both experimental capabilities and published hydrodynamic data to aid in the validation of a Computational Fluid Dynamics (CFD) based model of a slurry bubble column (SBC). The review included searching the Web of Science, ISI Proceedings, and Inspec databases, internet searches as well as other open literature sources. The goal of this study was to identify available experimental facilities and relevant data. Integral (i.e., pertaining to the SBC system), as well as fundamental (i.e., separate effects are considered), data are included in the scope of this effort. The fundamental data is needed to validate the individual mechanistic models or closure laws used in a Computational Multiphase Fluid Dynamics (CMFD) simulation of a SBC. The fundamental data is generally focused on simple geometries (i.e., flow between parallel plates or cylindrical pipes) or custom-designed tests to focus on selected interfacial phenomena. Integral data covers the operation of a SBC as a system with coupled effects. This work highlights selected experimental capabilities and data for the purpose of SBC model validation, and is not meant to be an exhaustive summary.

  2. General meeting. Technical reunion: the numerical and experimental simulation applied to the Reactor Physics; Assemblee generale. Reunion technique: la simulation numerique et experimentale appliquee a la physique des reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-10-01

    The SFEN (French Society on Nuclear Energy), organized the 18 october 2001 at Paris, a technical day on the numerical and experimental simulation, applied to the reactor Physics. Nine aspects were discussed, giving a state of the art in the domain:the french nuclear park; the future technology; the controlled thermonuclear fusion; the new organizations and their implications on the research and development programs; Framatome-ANP markets and industrial code packages; reactor core simulation at high temperature; software architecture; SALOME; DESCARTES. (A.L.B.)

  3. CONVECTION REACTOR

    Science.gov (United States)

    Hammond, R.P.; King, L.D.P.

    1960-03-22

    An homogeneous nuclear power reactor utilizing convection circulation of the liquid fuel is proposed. The reactor has an internal heat exchanger looated in the same pressure vessel as the critical assembly, thereby eliminating necessity for handling the hot liquid fuel outside the reactor pressure vessel during normal operation. The liquid fuel used in this reactor eliminates the necessity for extensive radiolytic gas rocombination apparatus, and the reactor is resiliently pressurized and, without any movable mechanical apparatus, automatically regulates itself to the condition of criticality during moderate variations in temperature snd pressure and shuts itself down as the pressure exceeds a predetermined safe operating value.

  4. Experimental Measurements of Heat Transfer through a Lunar Regolith Simulant in a Vibro-Fluidized Reactor Oven

    Science.gov (United States)

    Nayagam, Vedha; Berger, Gordon M.; Sacksteder, Kurt R.; Paz, Aaron

    2012-01-01

    Extraction of mission consumable resources such as water and oxygen from the planetary environment provides valuable reduction in launch-mass and potentially extends the mission duration. Processing of lunar regolith for resource extraction necessarily involves heating and chemical reaction of solid material with processing gases. Vibrofluidization is known to produce effective mixing and control of flow within granular media. In this study we present experimental results for vibrofluidized heat transfer in lunar regolith simulants (JSC-1 and JSC-1A) heated up to 900 C. The results show that the simulant bed height has a significant influence on the vibration induced flow field and heat transfer rates. A taller bed height leads to a two-cell circulation pattern whereas a single-cell circulation was observed for a shorter height. Lessons learned from these test results should provide insight into efficient design of future robotic missions involving In-Situ Resource Utilization.

  5. Dosimetry and radiobiology at the new RA-3 reactor boron neutron capture therapy (BNCT) facility: Application to the treatment of experimental oral cancer

    Energy Technology Data Exchange (ETDEWEB)

    Pozzi, E. [Research and Production Reactors, National Atomic Energy Commission, Ezeiza Atomic Center (Argentina); Department of Radiobiology, National Atomic Energy Commission, Constituyentes Atomic Center (Argentina)], E-mail: epozzi@cnea.gov.ar; Nigg, D.W. [Idaho National Laboratory, Idaho Falls (United States); Miller, M.; Thorp, S.I. [Instrumentation and Control Department, National Atomic Energy Commission, Ezeiza Atomic Center (Argentina); Heber, E.M. [Department of Radiobiology, National Atomic Energy Commission, Constituyentes Atomic Center (Argentina); Zarza, L.; Estryk, G. [Research and Production Reactors, National Atomic Energy Commission, Ezeiza Atomic Center (Argentina); Monti Hughes, A.; Molinari, A.J.; Garabalino, M. [Department of Radiobiology, National Atomic Energy Commission, Constituyentes Atomic Center (Argentina); Itoiz, M.E. [Department of Radiobiology, National Atomic Energy Commission, Constituyentes Atomic Center (Argentina); Department of Oral Pathology, Faculty of Dentistry, University of Buenos Aires (Argentina); Aromando, R.F. [Department of Oral Pathology, Faculty of Dentistry, University of Buenos Aires (Argentina); Quintana, J. [Research and Production Reactors, National Atomic Energy Commission, Ezeiza Atomic Center (Argentina); Trivillin, V.A.; Schwint, A.E. [Department of Radiobiology, National Atomic Energy Commission, Constituyentes Atomic Center (Argentina)

    2009-07-15

    The National Atomic Energy Commission of Argentina (CNEA) constructed a novel thermal neutron source for use in boron neutron capture therapy (BNCT) applications at the RA-3 research reactor facility located in Buenos Aires. The aim of the present study was to perform a dosimetric characterization of the facility and undertake radiobiological studies of BNCT in an experimental model of oral cancer in the hamster cheek pouch. The free-field thermal flux was 7.1x10{sup 9} n cm{sup -2} s{sup -1} and the fast neutron flux was 2.5x10{sup 6} n cm{sup -2} s{sup -1}, indicating a very well-thermalized neutron field with negligible fast neutron dose. For radiobiological studies it was necessary to shield the body of the hamster from the neutron flux while exposing the everted cheek pouch bearing the tumors. To that end we developed a lithium (enriched to 95% in {sup 6}Li) carbonate enclosure. Groups of tumor-bearing hamsters were submitted to BPA-BNCT, GB-10-BNCT, (GB-10+BPA)-BNCT or beam only treatments. Normal (non-cancerized) hamsters were treated similarly to evaluate normal tissue radiotoxicity. The total physical dose delivered to tumor with the BNCT treatments ranged from 6 to 8.5 Gy. Tumor control at 30 days ranged from 73% to 85%, with no normal tissue radiotoxicity. Significant but reversible mucositis in precancerous tissue surrounding tumors was associated to BPA-BNCT. The therapeutic success of different BNCT protocols in treating experimental oral cancer at this novel facility was unequivocally demonstrated.

  6. Combined coagulation-flocculation and sequencing batch reactor with phosphorus adjustment for the treatment of high-strength landfill leachate: experimental kinetics and chemical oxygen demand fractionation.

    Science.gov (United States)

    El-Fadel, M; Matar, F; Hashisho, J

    2013-05-01

    The treatability of high-strength landfill leachate is challenging and relatively limited. This study examines the feasibility of treating high-strength landfill leachate (chemical oxygen demand [COD]: 7,760-11,770 mg/L, biochemical oxygen demand [BOD5]: 2,760-3,569 mg/L, total nitrogen [TN] = 980-1,160 mg/L) using a sequencing batch reactor (SBR) preceded by a coagulation-flocculation process with phosphorus nutritional balance under various mixing and aeration patterns. Simulations were also conducted to define kinetic parameters and COD fractionation. Removal efficiencies reached 89% for BOD5, 60% for COD, and 72% for TN, similar to and better than reported studies, albeit with a relatively lower hydraulic retention time (HRT) and solid retention time (SRT). The coupled experimental and simulation results contribute in filling a gap toward managing high-strength landfill leachate and providing guidelines for corresponding SBR applications. The treatability of high-strength landfill leachate, which is challenging and relatively limited, was demonstrated using a combined coagulation-flocculation with SBR technology and nutrient balance adjustment. The most suitable coagulant, kinetic design parameters, and COD fractionation were defined using coupled experimental and simulation results contributing in filling a gap toward managing high-strength leachate by providing guidelines for corresponding SBR applications and anticipating potential constraints related to the non-biodegradable COD fraction. In this context, while the combined coagulation-flocculation and SBR process improved removal efficiencies, posttreatment may be required for high-strength leachate, depending on discharge standards and ultimate usage of the treated leachate.

  7. Dosimetry and radiobiology at the new RA-3 reactor boron neutron capture therapy (BNCT) facility: application to the treatment of experimental oral cancer.

    Science.gov (United States)

    Pozzi, E; Nigg, D W; Miller, M; Thorp, S I; Heber, E M; Zarza, L; Estryk, G; Monti Hughes, A; Molinari, A J; Garabalino, M; Itoiz, M E; Aromando, R F; Quintana, J; Trivillin, V A; Schwint, A E

    2009-07-01

    The National Atomic Energy Commission of Argentina (CNEA) constructed a novel thermal neutron source for use in boron neutron capture therapy (BNCT) applications at the RA-3 research reactor facility located in Buenos Aires. The aim of the present study was to perform a dosimetric characterization of the facility and undertake radiobiological studies of BNCT in an experimental model of oral cancer in the hamster cheek pouch. The free-field thermal flux was 7.1 x 10(9) n cm(-2)s(-1) and the fast neutron flux was 2.5 x 10(6) n cm(-2)s(-1), indicating a very well-thermalized neutron field with negligible fast neutron dose. For radiobiological studies it was necessary to shield the body of the hamster from the neutron flux while exposing the everted cheek pouch bearing the tumors. To that end we developed a lithium (enriched to 95% in (6)Li) carbonate enclosure. Groups of tumor-bearing hamsters were submitted to BPA-BNCT, GB-10-BNCT, (GB-10+BPA)-BNCT or beam only treatments. Normal (non-cancerized) hamsters were treated similarly to evaluate normal tissue radiotoxicity. The total physical dose delivered to tumor with the BNCT treatments ranged from 6 to 8.5 Gy. Tumor control at 30 days ranged from 73% to 85%, with no normal tissue radiotoxicity. Significant but reversible mucositis in precancerous tissue surrounding tumors was associated to BPA-BNCT. The therapeutic success of different BNCT protocols in treating experimental oral cancer at this novel facility was unequivocally demonstrated.

  8. The improvement of control rod in experimental fast reactor JOYO. The development of a sodium bonded type control rod

    Energy Technology Data Exchange (ETDEWEB)

    Soga, T.; Miyakawa, S.; Mitsugi, T. [Japan Nuclear Cycle Development Inst., Oarai Engineering Center, Irradiation Center, Irradiation and Administration Section, Oarai, Ibaraki (Japan)

    1999-06-01

    that the cladding strength sufficient to withstand stress accounting for decreased thickness by the ACCI zone. (5) The wet wash and storage method was selected for disposing of the spent sodium bonded control rods, based upon experimental results at the JOYO facilities. The effects from storing sodium bonded control rods in wet storage were evaluated. The results indicated that these effect would not pose a safety problem. (author)

  9. REACTOR COOLING

    Science.gov (United States)

    Quackenbush, C.F.

    1959-09-29

    A nuclear reactor with provisions for selectively cooling the fuel elements is described. The reactor has a plurality of tubes extending throughout. Cylindrical fuel elements are disposed within the tubes and the coolant flows through the tubes and around the fuel elements. The fuel elements within the central portion of the reactor are provided with roughened surfaces of material. The fuel elements in the end portions of the tubes within the reactor are provlded with low conduction jackets and the fuel elements in the region between the central portion and the end portions are provided with smooth surfaces of high heat conduction material.

  10. Multifunctional reactors

    NARCIS (Netherlands)

    Westerterp, K.R.

    1992-01-01

    Multifunctional reactors are single pieces of equipment in which, besides the reaction, other functions are carried out simultaneously. The other functions can be a heat, mass or momentum transfer operation and even another reaction. Multifunctional reactors are not new, but they have received much

  11. Development and trial manufacturing of 1/2-scale partial mock-up of blanket box structure for fusion experimental reactor

    Science.gov (United States)

    Hashimoto, Toshiyuki; Takatsu, Hideyuki; Sato, Satoshi

    1994-07-01

    Conceptual design of breeding blanket has been discussed during the CDA (Conceptual Design Activities) of ITER (International Thermonuclear Experimental Reactor). Structural concept of breeding blanket is based on box structure integrated with first wall and shield, which consists of three coolant manifolds for first wall, breeding and shield regions. The first wall must have cooling channels to remove surface heat flux and nuclear heating. The box structure includes plates to form the manifolds and stiffening ribs to withstand enormous electromagnetic load, coolant pressure and blanket internal (purge gas) pressure. A 1/2-scale partial model of the blanket box structure for the outboard side module near midplane is manufactured to estimate the fabrication technology, i.e. diffusion bonding by HIP (Hot Isostatic Pressing) and EBW (Electron Beam Welding) procedure. Fabrication accuracy is a key issue to manufacture first wall panel because bending deformation during HIP may not be small for a large size structure. Data on bending deformation during HIP was obtained by preliminary manufacturing of HIP elements. For the shield structure, it is necessary to reduce the welding strain and residual stress of the weldment to establish the fabrication procedure. Optimal shape of the parts forming the manifolds, welding locations and welding sequence have been investigated. In addition, preliminary EBW tests have been performed in order to select the EBW conditions, and fundamental data on built-up shield have been obtained. Especially, welding deformation by joining the first wall panel to the shield has been measured, and total deformation to build-up shield by EBW has been found to be smaller than 2 mm. Consequently, the feasibility of fabrication technologies has been successfully demonstrated for a 1m-scaled box structure including the first wall with cooling channels by means of HIP, EBW and TIG (Tungsten Inert Gas arc)-welding.

  12. Physics design of a 100 keV acceleration grid system for the diagnostic neutral beam for international tokamak experimental reactor.

    Science.gov (United States)

    Singh, M J; De Esch, H P L

    2010-01-01

    This paper describes the physics design of a 100 keV, 60 A H(-) accelerator for the diagnostic neutral beam (DNB) for international tokamak experimental reactor (ITER). The accelerator is a three grid system comprising of 1280 apertures, grouped in 16 groups with 80 apertures per beam group. Several computer codes have been used to optimize the design which follows the same philosophy as the ITER Design Description Document (DDD) 5.3 and the 1 MeV heating and current drive beam line [R. Hemsworth, H. Decamps, J. Graceffa, B. Schunke, M. Tanaka, M. Dremel, A. Tanga, H. P. L. De Esch, F. Geli, J. Milnes, T. Inoue, D. Marcuzzi, P. Sonato, and P. Zaccaria, Nucl. Fusion 49, 045006 (2009)]. The aperture shapes, intergrid distances, and the extractor voltage have been optimized to minimize the beamlet divergence. To suppress the acceleration of coextracted electrons, permanent magnets have been incorporated in the extraction grid, downstream of the cooling water channels. The electron power loads on the extractor and the grounded grids have been calculated assuming 1 coextracted electron per ion. The beamlet divergence is calculated to be 4 mrad. At present the design for the filter field of the RF based ion sources for ITER is not fixed, therefore a few configurations of the same have been considered. Their effect on the transmission of the electrons and beams through the accelerator has been studied. The OPERA-3D code has been used to estimate the aperture offset steering constant of the grounded grid and the extraction grid, the space charge interaction between the beamlets and the kerb design required to compensate for this interaction. All beamlets in the DNB must be focused to a single point in the duct, 20.665 m from the grounded grid, and the required geometrical aimings and aperture offsets have been calculated.

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

    Energy Technology Data Exchange (ETDEWEB)

    1978-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Koch, M.; Kazimi, M.S.

    1991-04-01

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

  15. NUCLEAR REACTOR

    Science.gov (United States)

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

    1958-01-21

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

  16. NUCLEAR REACTOR

    Science.gov (United States)

    Grebe, J.J.

    1959-07-14

    High temperature reactors which are uniquely adapted to serve as the heat source for nuclear pcwered rockets are described. The reactor is comprised essentially of an outer tubular heat resistant casing which provides the main coolant passageway to and away from the reactor core within the casing and in which the working fluid is preferably hydrogen or helium gas which is permitted to vaporize from a liquid storage tank. The reactor core has a generally spherical shape formed entirely of an active material comprised of fissile material and a moderator material which serves as a diluent. The active material is fabricated as a gas permeable porous material and is interlaced in a random manner with very small inter-connecting bores or capillary tubes through which the coolant gas may flow. The entire reactor is divided into successive sections along the direction of the temperature gradient or coolant flow, each section utilizing materials of construction which are most advantageous from a nuclear standpoint and which at the same time can withstand the operating temperature of that particular zone. This design results in a nuclear reactor characterized simultaneously by a minimum critiral size and mass and by the ability to heat a working fluid to an extremely high temperature.

  17. Experimental data and numerical predictions of a single-phase flow in a batch square stirred tank reactor with a rotating cylinder agitator

    Science.gov (United States)

    Escamilla-Ruíz, I. A.; Sierra-Espinosa, F. Z.; García, J. C.; Valera-Medina, A.; Carrillo, F.

    2017-09-01

    Single-phase flows in stirred tank reactors have useful characteristics for a wide number of industrial applications. Usually, reactors are cylindrical vessels and complex impeller designs, which are often highly energy consuming and produce complicated flow patterns. Therefore, a novel configuration consisting of a square stirred tank reactor is proposed in this study with potential advantages over conventional reactors. In the present work hydrodynamics and turbulence have been studied for a single-phase flow in steady state operating in batch condition. The flow was induced by drag from a rotating cylinder with two diameters. The effects of drag from the stirrer as well as geometrical parameters of the system on the hydrodynamic behavior were investigated using Computational Fluids Dynamics (CFD) and non-intrusive Laser Doppler Anemometry, (LDA). Data obtained from LDA measurements were used for the validation of the CFD simulations, and to detecting the macro-instabilities inside the tank, based on the time series analysis for three rotational speeds N = 180, 1000 and 2000 rpm. The numerical results revealed the formation of flow patterns and macro-vortex structures in the upper part of the tank as consequence of the Reynolds number and the stream discharge emanated from the cylindrical stirrer. Moreover, increasing the cylinder diameter has an impact on the number of recirculation loops as well as the energy consumption of the entire system showing better performance in the presence of turbulent flows.

  18. Reactor Neutrinos

    CERN Document Server

    Lasserre, T; Lasserre, Thierry; Sobel, Henry W.

    2005-01-01

    We review the status and the results of reactor neutrino experiments, that toe the cutting edge of neutrino research. Short baseline experiments have provided the measurement of the reactor neutrino spectrum, and are still searching for important phenomena such as the neutrino magnetic moment. They could open the door to the measurement of coherent neutrino scattering in a near future. Middle and long baseline oscillation experiments at Chooz and KamLAND have played a relevant role in neutrino oscillation physics in the last years. It is now widely accepted that a new middle baseline disappearance reactor neutrino experiment with multiple detectors could provide a clean measurement of the last undetermined neutrino mixing angle theta13. We conclude by opening on possible use of neutrinos for Society: NonProliferation of Nuclear materials and Geophysics.

  19. NEUTRONIC REACTORS

    Science.gov (United States)

    Wigner, E.P.; Young, G.J.

    1958-10-14

    A method is presented for loading and unloading rod type fuel elements of a neutronic reactor of the heterogeneous, solld moderator, liquid cooled type. In the embodiment illustrated, the fuel rods are disposed in vertical coolant channels in the reactor core. The fuel rods are loaded and unloaded through the upper openings of the channels which are immersed in the coolant liquid, such as water. Unloading is accomplished by means of a coffer dam assembly having an outer sleeve which is placed in sealing relation around the upper opening. A radiation shield sleeve is disposed in and reciprocable through the coffer dam sleeve. A fuel rod engaging member operates through the axial bore in the radiation shield sleeve to withdraw the fuel rod from its position in the reactor coolant channel into the shield, the shield snd rod then being removed. Loading is accomplished in the reverse procedure.

  20. Safety of VVER-440 reactors

    CERN Document Server

    Slugen, Vladimir

    2011-01-01

    Safety of VVER-440 Reactors endeavours to promote an increase in the safety of VVER-440 nuclear reactors via the improvement of fission products limitation systems and the implementation of special non-destructive spectroscopic methods for materials testing. All theoretical and experimental studies performed the by author over the last 25 years have been undertaken with the aim of improving VVER-440 defence in depth, which is one of the most important principle for ensuring safety in nuclear power plants. Safety of VVER-440 Reactors is focused on the barrier system through which the safety pri

  1. Supercritical-pressure light water cooled reactors

    CERN Document Server

    Oka, Yoshiaki

    2014-01-01

    This book focuses on the latest reactor concepts, single pass core and experimental findings in thermal hydraulics, materials, corrosion, and water chemistry. It highlights research on supercritical-pressure light water cooled reactors (SCWRs), one of the Generation IV reactors that are studied around the world. This book includes cladding material development and experimental findings on heat transfer, corrosion and water chemistry. The work presented here will help readers to understand the fundamental elements of reactor design and analysis methods, thermal hydraulics, materials and water

  2. Neutronics calculation of an heterogeneous compact and thermal core by means of deterministic and stochastic transport theory. Application to the experimental reactor of the University of Strasbourg; Modelisation neutronique d`un coeur thermique compact et heterogene en theorie du transport deterministe et probabiliste. Application au reacteur experimental de l`Universite de Strasbourg

    Energy Technology Data Exchange (ETDEWEB)

    Jammes, Ch

    1997-11-28

    The aim of this work is to create, validate theoretically and experimentally a calculation route for a thermal irradiation reactor. This is the research reactor of the University of Strasbourg, which presents all of characteristics of this reactor-type: compact and heterogeneous core, slab-type fuel with a high 235-uranium enrichment. This calculation route is based on the first use of the following two modern transport methods: the TDT method and the Monte Carlo method. The former, programmed within the APOLLO2 code, is a two dimensional collision probabilities method. The later, used by the TRIPOLI4 code, is a stochastic method. Both can be applied to complex geometries. After a few theoretical reminders about transport codes, a set of integral experiments is described which have been realized within the research reactor of the University of Strasbourg. One of them has been performed for this study. At the beginning of the theoretical part, significant errors are apparent due to the use of calculation route based on homogenization, condensation and the diffusion approximation. An extensive comparison between the discrete ordinates method and the TDT method carries out that the use of the TDT method is relevant for the studied reactor. The treatment of axial leakage with this method is the only disadvantage. Therefore, the use of the code TRIPOLI4 is recommended for a more accurate study of leakage within a reflector. By means of the experimental data, the ability of our calculation route is confirmed for essential neutronics questions such as the critical mass determination, the power distribution and the fuel management. (author)

  3. Experimental studies into the fluid dynamic performance of the coolant flow in the mixed core of the Temelin NPP VVER-1000 reactor

    Directory of Open Access Journals (Sweden)

    S.M. Dmitriev

    2015-11-01

    Full Text Available The paper presents the results of studies into the interassembly coolant interaction in the Temelin nuclear power plant (NPP VVER-1000 reactor core. An aerodynamic test bench was used to study the coolant flow processes in a TVSA-type fuel assembly bundle. To obtain more detailed information on the coolant flow dynamics, a VVER-1000 reactor core fragment was selected as the test model, which comprised two segments of a TVSA-12 PLUS fuel assembly and one segment of a TVSA-T assembly with stiffening angles and an interassembly gap. The studies into the coolant fluid dynamics consisted in measuring the velocity vector both in representative TVSA regions and inside the interassembly gap using a five-channel pneumometric probe. An analysis into the spatial distribution of the absolute flow velocity projections made it possible to detail the TVSA spacer, mixing and combined spacer grid flow pattern, identify the regions with the maximum transverse coolant flow, and determine the depth of the coolant flow disturbance propagation and redistribution in adjacent TVSA assemblies. The results of the studies into the interassembly coolant interaction among the adjacent TVSA assemblies are used at OKBM Afrikantov to update the VVER-1000 core thermal-hydraulic analysis procedures and have been added to the database for verification of computational fluid dynamics (CFD codes and for detailed cellwise analyses of the VVER-100 reactor cores.

  4. EBR-II Reactor Physics Benchmark Evaluation Report

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-12-28

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

  5. Neutronic reactor

    Science.gov (United States)

    Wende, Charles W. J.; Babcock, Dale F.; Menegus, Robert L.

    1983-01-01

    A nuclear reactor includes an active portion with fissionable fuel and neutron moderating material surrounded by neutron reflecting material. A control element in the active portion includes a group of movable rods constructed of neutron-absorbing material. Each rod is movable with respect to the other rods to vary the absorption of neutrons and effect control over neutron flux.

  6. Neutronic reactor

    Energy Technology Data Exchange (ETDEWEB)

    Babcock, D.F.; Menegus, R.L.; Wende, C.W.

    1983-01-04

    A nuclear reactor includes an active portion with fissionable fuel and neutron moderating material surrounded by neutron reflecting material. A control element in the active portion includes a group of movable rods constructed of neutron-absorbing material. Each rod is movable with respect to the other rods to vary the absorption of neutrons and effect control over neutron flux.

  7. Stabilized Spheromak Fusion Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Fowler, T

    2007-04-03

    The U.S. fusion energy program is focused on research with the potential for studying plasmas at thermonuclear temperatures, currently epitomized by the tokamak-based International Thermonuclear Experimental Reactor (ITER) but also continuing exploratory work on other plasma confinement concepts. Among the latter is the spheromak pursued on the SSPX facility at LLNL. Experiments in SSPX using electrostatic current drive by coaxial guns have now demonstrated stable spheromaks with good heat confinement, if the plasma is maintained near a Taylor state, but the anticipated high current amplification by gun injection has not yet been achieved. In future experiments and reactors, creating and maintaining a stable spheromak configuration at high magnetic field strength may require auxiliary current drive using neutral beams or RF power. Here we show that neutral beam current drive soon to be explored on SSPX could yield a compact spheromak reactor with current drive efficiency comparable to that of steady state tokamaks. Thus, while more will be learned about electrostatic current drive in coming months, results already achieved in SSPX could point to a productive parallel development path pursuing auxiliary current drive, consistent with plans to install neutral beams on SSPX in the near future. Among possible outcomes, spheromak research could also yield pulsed fusion reactors at lower capital cost than any fusion concept yet proposed.

  8. Turbulent precipitation of uranium oxalate in a vortex reactor - experimental study and modelling; Precipitation turbulente d'oxalate d'uranium en reacteur vortex - etude experimentale et modelisation

    Energy Technology Data Exchange (ETDEWEB)

    Sommer de Gelicourt, Y

    2004-03-15

    Industrial oxalic precipitation processed in an un-baffled magnetically stirred tank, the Vortex Reactor, has been studied with uranium simulating plutonium. Modelling precipitation requires a mixing model for the continuous liquid phase and the solution of population balance for the dispersed solid phase. Being chemical reaction influenced by the degree of mixing at molecular scale, that commercial CFD code does not resolve, a sub-grid scale model has been introduced: the finite mode probability density functions, and coupled with a model for the liquid energy spectrum. Evolution of the dispersed phase has been resolved by the quadrature method of moments, first used here with experimental nucleation and growth kinetics, and an aggregation kernel based on local shear rate. The promising abilities of this local approach, without any fitting constant, are strengthened by the similarity between experimental results and simulations. (author)

  9. Dynamic model of biological nitrogen removal in SBR reactors: Experimental study and model calibration; Modello di simulazione dinamica della rimozione biologica dell`azoto con reattori a flusso discontinuo (SBR): indagine sperimentale e taratura del modello

    Energy Technology Data Exchange (ETDEWEB)

    Andreottola, G. [Trento Univ. (Italy). Dipt. di Ingegneria Civile e Ambientale; Bortone, G.; Spadoni, S. [ENEA, Bologna (Italy). Area Energia Ambiente e Salute

    1995-03-01

    In the present study are presented the results of the experimental activity carried out in order to calibrate and verify the dynamical model developed by the Authors for nitrogen removal with SBR processes. Four track studies (two for the calibration phase and two for the verification one) have been carried out on the SBR lab-scale reactor. The experimental results have shown temporary increase of nitrite in the nitrification phase. Therefore, the SBR dynamic model has been modified, in order to better explain the dynamic behavior of the nitrogen removal process: nitrousation and nitratation processes have been explicitly described and a inhibition kinetic of nitratation by ammonia has been introduced. The new model, calibrated and verified, has successfully explained the dynamic behavior of the process. Further work will be carried out in order to better understand the inhibition dynamic of nitratation.

  10. Experimental investigation of iodine removal and containment depressurization in containment spray system test facility of 700 MWe Indian pressurized heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jain, Manish [Department of Mechanical Engineering, I.I.T., Bombay, Powai, Mumbai (India); Kandar, T.K.; Vhora, S.F.; Mohan, Nalini [Directorate of Technology Development, Nuclear Power Corporation of India Limited, Mumbai (India); Iyer, K.N. [Department of Mechanical Engineering, I.I.T., Bombay, Powai, Mumbai (India); Prabhu, S.V., E-mail: svprabhu@iitb.ac.in [Department of Mechanical Engineering, I.I.T., Bombay, Powai, Mumbai (India)

    2017-05-15

    Highlights: • Depressurization rate in a scaled down vessel filled with air and steam is studied. • Iodine removal rate in a scaled down vessel filled with steam/air is investigated. • Effect of SMD and vessel pressure on depressurization rate is studied. • Depressurization rate decreases with the increase in the droplet size (590 μm – 1 mm) • Decrease in pressure and iodine concentration with time follow exponential trend. - Abstract: As an additional safety measure in the new 700 MWe Indian pressurized heavy water reactors, the first of a kind system called containment Spray System is introduced. The system is designed to cater/mitigate the conditions after design basis accidents i.e., loss of coolant accident and main steam line break. As a contribution to the safety analysis of condition following loss-of-coolant accidents, experiments are carried out to establish the performance of the system. The loss of coolant is simulated by injecting saturated steam and iodine vapors into the containment vessel in which air is enclosed at atmospheric and room temperature, and then the steam-air mixture is cooled by sprays of water. The effect of water spray on the containment vessel pressure and the iodine scrubbing in a scaled down facility is investigated for the containment spray system of Indian pressurized heavy water reactors. The experiments are carried out in the scaled down vessel of the diameter of 2.0 m and height of 3.5 m respectively. Experiments are conducted with water at room temperature as the spray medium. Two different initial vessel pressure i.e. 0.7 bar and 1.0 bar are chosen for the studies as they are nearing the loss of coolant accident & main steam line break pressures in Indian pressurized heavy water reactors. These pressures are chosen based on the containment resultant pressures after a design basis accident. The transient temperature and pressure distribution of the steam in the vessel are measured during the depressurization

  11. Advanced Safeguards Approaches for New Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-12-15

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

  12. Jules Horowitz Reactor, basic design

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-07-01

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

  13. A new MTR fuel for a new MTR reactor: UMo for the Jules Horowitz reactor

    Energy Technology Data Exchange (ETDEWEB)

    Guigon, B. [CEA Cadarache, Dir. de l' Energie Nucleaire DEN, Reacteur Jules Horowitz, 13 - Saint-Paul-lez-Durance (France); Vacelet, H. [Compagnie pour l' Etude et la Realisation de Combustibles Atomiques, CERCA, Etablissement de Romans, 26 (France); Dornbusch, D. [Technicatome, Service d' Architecture Generale, 13 - Aix-en-Provence (France)

    2003-07-01

    Within some years, the Jules Horowitz Reactor will be the only working experimental reactor (material and fuel testing reactor) in France. It will have to provide facilities for a wide range of needs: from activation analysis to power reactor fuel qualification. In this paper will be presented the main characteristics of the Jules Horowitz Reactor: its total power, neutron flux, fuel element... Safety criteria will be explained. Finally merits and disadvantages of UMo compared to the standard U{sub 3}Si{sub 2} fuel will be discussed. (authors)

  14. Experimental results of acetone hydrogenation on a heat exchanger type reactor for solar chemical heat pump; Solar chemical heat pump ni okeru acetone suisoka hanno netsu kaishu jikken

    Energy Technology Data Exchange (ETDEWEB)

    Takashima, T.; Doi, T.; Tanaka, T.; Ando, Y. [Electrotechnical Laboratory, Tsukuba (Japan); Miyahara, R.; Kamoshida, J. [Shibaura Institute of Technology, Tokyo (Japan)

    1996-10-27

    With the purpose of converting solar heat energy to industrial heat energy, an experiment of acetone hydrogenation was carried out using a heat exchanger type reactor that recovers heat generated by acetone hydrogenation, an exothermic reaction, and supplies it to an outside load. In the experiment, a pellet-like activated carbon-supported ruthenium catalyst was used for the acetone hydrogenation with hydrogen and acetone supplied to the catalyst layer at a space velocity of 400-1,200 or so. In the external pipe of the double-pipe type reactor, a heating medium oil was circulated in parallel with the flow of the reactant, with the heat of reaction recovered that was generated from the acetone hydrogenation. In this experiment, an 1wt%Ru/C catalyst and a 5wt%Ru/C catalyst were used so as to examine the effects of variation in the space velocity. As a result, from the viewpoint of recovering the heat of reaction, it was found desirable to increase the reaction speed by raising catalytic density and also to supply the reactant downstream inside the reaction pipe by increasing the space velocity. 1 ref., 6 figs., 1 tab.

  15. Experimental and density functional theoretical study of the effects of Fenton’s reaction on the degradation of Bisphenol A in a high voltage plasma reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dai, Fei, E-mail: daif@clarkson.edu [Department of Civil and Environmental Engineering, 8 Clarkson Avenue, Clarkson University, Potsdam, 13699 NY (United States); Fan, Xiangru, E-mail: fanx@clarkson.edu [Department of Chemical and Biomolecular Engineering, 8 Clarkson Avenue, Clarkson University, Potsdam, 13699 NY (United States); Stratton, Gunnar R., E-mail: strattgr@clarkson.edu [Department of Chemical and Biomolecular Engineering, 8 Clarkson Avenue, Clarkson University, Potsdam, 13699 NY (United States); Bellona, Christopher L., E-mail: cbellona@clarkson.edu [Department of Civil and Environmental Engineering, 8 Clarkson Avenue, Clarkson University, Potsdam, 13699 NY (United States); Department of Civil and Environmental Engineering, 1500 Illinois St., Colorado School of Mines, Golden, 80401 CO (United States); Holsen, Thomas M., E-mail: tholsen@clarkson.edu [Department of Civil and Environmental Engineering, 8 Clarkson Avenue, Clarkson University, Potsdam, 13699 NY (United States); Crimmins, Bernard S., E-mail: bcrimmin@clarkson.edu [Department of Civil and Environmental Engineering, 8 Clarkson Avenue, Clarkson University, Potsdam, 13699 NY (United States); Xia, Xiaoyan, E-mail: xiax@clarkson.edu [Department of Civil and Environmental Engineering, 8 Clarkson Avenue, Clarkson University, Potsdam, 13699 NY (United States); Mededovic Thagard, Selma, E-mail: smededov@clarkson.edu [Department of Chemical and Biomolecular Engineering, 8 Clarkson Avenue, Clarkson University, Potsdam, 13699 NY (United States)

    2016-05-05

    Highlights: • Combining the Fenton reaction with the plasma treatment reduces Bisphenol A concentration below the detection limit within 30 min. • Carbon steel electrode in the plasma reactor can be used as a source of iron ions. • OH radical attack on Bisphenol A is the primary pathway for byproduct formation. - Abstract: A novel electrical discharge plasma reactor configuration with and without iron ions was evaluated for the degradation of 0.02 mM Bisphenol A (BPA). The pseudo-first-order reaction rate constant calculated for the plasma treatment of BPA with a stainless steel electrode in the presence of dissolved ferrous ion (Fe{sup 2+}) salts (termed plasma/Fenton treatment) was higher than in the plasma treatment in the absence of iron salts. At the optimal ferrous ion concentration, longer plasma treatment times resulted in higher BPA degradation rates, likely due to increased hydroxyl (OH) radical concentration formed through the decomposition of H{sub 2}O{sub 2}. Replacing the stainless steel with a carbon steel grounded electrode resulted in the release of iron ions from the carbon steel thereby increasing the rate of BPA removal and eliminating the need for iron salts. After the plasma/Fenton treatment, >97% of the residual iron salts were removed by coagulation/flocculation/sedimentation. Byproduct identification coupled with density functional theory (DFT) calculations confirmed that OH radical attack on BPA’s hydroxyl group is the primary pathway for byproduct formation.

  16. Experimental investigations on the coolability of prototypical particle beds with respect to reactor safety; Experimentelle Untersuchungen der Kuehlbarkeit prototypischer Schuettungskonfigurationen unter dem Aspekt der Reaktorsicherheit

    Energy Technology Data Exchange (ETDEWEB)

    Leininger, Simon

    2017-02-22

    In case of a severe accident in a light water reactor, continuous unavailability of cooling water to the reactor core may result in overheating of the fuel elements and finally the loss of core integrity. Under such conditions, a structure of heat-releasing particles of different size and shape may be formed by fragmentation of molten core material in several stages of the accident. The long-term coolability of such beds is of prime im-portance to avoid any damage to the reactor pressure vessel or even a release of fission products to the environment. In the frame of this work, specific experiments were con-ducted under prototypical conditions employing the existing DEBRIS test facility in order to gain further knowledge about the thermohydraulic behavior of such beds. In steady state boiling experiments, the pressure gradients in particle beds were meas-ured both for one- and multi-dimensional cooling water flow conditions and compared with one another in order to assess the flow behavior inside the bed. For these different flow conditions as well as for stratified bed configurations, the maximum removable heat flux densities were determined in the dryout experiments. E. g., it was found that an axial stratification of the permeability can significantly reduce the bed's coolability. For the first time, the quenching behavior of dry, superheated beds was investigated at elevated system pressure up to 0.5 MPa. In these experiments, the effect of system pressure on the coolability was quantified by means of the quenching time (time period to cool down the bed to saturation temperature). The investigated particle beds mainly consisted of non-spherical particles with well-defined geometry (cylinders and screws). It was shown that the effect of the particles geometry on the flow in a particle bed can be best estimated by using an equivalent particle diameter calculated for monodisperse particle beds from the product of the Sauter diameter and a shape factor and for

  17. Nuclear Reactors. Revised.

    Science.gov (United States)

    Hogerton, John F.

    This publication is one of a series of information booklets for the general public published by the United States Atomic Energy Commission. Among the topics discussed are: How Reactors Work; Reactor Design; Research, Teaching, and Materials Testing; Reactors (Research, Teaching and Materials); Production Reactors; Reactors for Electric Power…

  18. Multiphase Reactors: Models and Experimental Verification Les réacteurs polyphasés : modèles et vérification expérimentale

    Directory of Open Access Journals (Sweden)

    Dudukovic M. P.

    2006-11-01

    Full Text Available Identification and quantification of flow regimes, phase holdup distributions, flow patterns and backmixing is essential for proper design and scale-up of multiphase reactors. Existing models often suffer from inadequate experimental confirmation. Here, we describe recent progress made in our laboratory in characterizing liquid circulation and backmixing in bubble columns and in quantifying, via a phenomenological model, the behavior of trickle-beds in the low interaction regime. The need for imaging techniques such as fan-beam tomography and PET is illustrated. Liquid mixing in bubble columns caused by global liquid convection and by turbulent eddies is usually described by the axial dispersion model (ADM which, at least in the churn turbulent regime, lacks a theoretical basis. In the ADM the two mixing mechanisms are lumped into a single effective dispersion coefficient. An alternative backmixing description assumes multiple liquid circulation cells, with height equal to column diameter, to occur in the column. We have successfully demonstrated that instantaneous and time averaged liquid velocity profiles in the column can be determined by monitoring the motion of a neutrally bouyant tracer particle via a Computer Aided Radioactive Particle Tracking (CARPT technique. It can now be shown that a single strong liquid circulation cell exists in columns of a variety of diameters and at different operating conditions but that smaller circulation cells can form in the distributor region. Model predicted one dimensional axial time averaged liquid velocity profile agrees well with the data and applies to the middle section of the column. The CARPT technique allows the computation of Lagrangian autocorrelation coefficients, the rms fluctuating velocities, Lagrangian integral time scales and the turbulent dispersion coefficients. Thus CARPT provides the necessary information i. e. velocity profiles and turbulent dispersion coefficients, for

  19. Photocatalytic reactor

    Science.gov (United States)

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

    1999-01-01

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

  20. Hybrid adsorptive membrane reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-03-01

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

  1. D and DR Reactors

    Data.gov (United States)

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

  2. Combined Reactor and Microelectrode Measurements in Laboratory Grown Biofilms

    DEFF Research Database (Denmark)

    Larsen, Tove; Harremoës, Poul

    1994-01-01

    A combined biofilm reactor-/microelectrode experimental set-up has been constructed, allowing for simultaneous reactor mass balances and measurements of concentration profiles within the biofilm. The system consists of an annular biofilm reactor equipped with an oxygen microelectrode. Experiments...

  3. Reactor transient

    Energy Technology Data Exchange (ETDEWEB)

    Menegus, R.L.

    1956-05-31

    The authors are planning a calculation to be done on the Univac at the Louviers Building to estimate the effect of xenon transients, a high reactor power. This memorandum outlines the reasons why they prefer to do the work at Louviers rather than at another location, such as N.Y.U. They are to calculate the response of the reactor to a sudden change in position of the half rods. Qualitatively, the response will be a change in the rooftop ratio of the neutron flux. The rooftop ratio may oscillate with high damping, or, instead, it may oscillate for many cycles. It has not been possible for them to determine this response by hand calculation because of the complexity of the problem, and yet it is important for them to be certain that high power operation will not lead us to inherently unstable operation. Therefore they have resorted to machine computation. The system of differential equations that describes the response has seven dependent variables; therefore there are seven equations, each coupled with one or more of the others. The authors have discussed the problem with R.R. Haefner at the plant, and it is his opinion that the IBM 650 cannot adequately handle the system of seven equations because the characteristic time constants vary over a range of about 10{sup 8}. The Univac located at the Louviers Building is said to be satisfactory for this computation.

  4. Verification of Remote Inspection Techniques for Reactor Internal Structures of Liquid Metal Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Young Sang; Lee, Jae Han

    2007-02-15

    The reactor internal structures and components of a liquid metal reactor (LMR) are submerged in hot sodium of reactor vessel. The division 3 of ASME code section XI specifies the visual inspection as major in-service inspection (ISI) methods of reactor internal structures and components. Reactor internals of LMR can not be visually examined due to opaque liquid sodium. The under-sodium viewing techniques using an ultrasonic wave should be applied for the visual inspection of reactor internals. Recently, an ultrasonic waveguide sensor with a strip plate has been developed for an application to the under-sodium inspection. In this study, visualization technique, ranging technique and monitoring technique have been suggested for the remote inspection of reactor internals by using the waveguide sensor. The feasibility of these remote inspection techniques using ultrasonic waveguide sensor has been evaluated by an experimental verification.

  5. Experiment for search for sterile neutrino at SM-3 reactor

    Science.gov (United States)

    Serebrov, A. P.; Ivochkin, V. G.; Samoylov, R. M.; Fomin, A. K.; Zinoviev, V. G.; Neustroev, P. V.; Golovtsov, V. L.; Gruzinsky, N. V.; Solovey, V. A.; Cherniy, A. V.; Zherebtsov, O. M.; Martemyanov, V. P.; Zinoev, V. G.; Tarasenkov, V. G.; Aleshin, V. I.; Petelin, A. L.; Pavlov, S. V.; Izhutov, A. L.; Sazontov, S. A.; Ryazanov, D. K.; Gromov, M. O.; Afanasiev, V. V.; Matrosov, L. N.; Matrosova, M. Yu.

    2016-11-01

    In connection with the question of possible existence of sterile neutrino the laboratory on the basis of SM-3 reactor was created to search for oscillations of reactor antineutrino. A prototype of a neutrino detector with scintillator volume of 400 l can be moved at the distance of 6-11 m from the reactor core. The measurements of background conditions have been made. It is shown that the main experimental problem is associated with cosmic radiation background. Test measurements of dependence of a reactor antineutrino flux on the distance from a reactor core have been made. The prospects of search for oscillations of reactor antineutrino at short distances are discussed.

  6. Reference worldwide model for antineutrinos from reactors

    Science.gov (United States)

    Baldoncini, Marica; Callegari, Ivan; Fiorentini, Giovanni; Mantovani, Fabio; Ricci, Barbara; Strati, Virginia; Xhixha, Gerti

    2015-03-01

    Antineutrinos produced at nuclear reactors constitute a severe source of background for the detection of geoneutrinos, which bring to the Earth's surface information about natural radioactivity in the whole planet. In this framework, we provide a reference worldwide model for antineutrinos from reactors, in view of reactors operational records yearly published by the International Atomic Energy Agency. We evaluate the expected signal from commercial reactors for ongoing (KamLAND and Borexino), planned (SNO +), and proposed (Juno, RENO-50, LENA, and Hanohano) experimental sites. Uncertainties related to reactor antineutrino production, propagation, and detection processes are estimated using a Monte Carlo-based approach, which provides an overall site-dependent uncertainty on the signal in the geoneutrino energy window on the order of 3%. We also implement the off-equilibrium correction to the reference reactor spectra associated with the long-lived isotopes, and we estimate a 2.4% increase of the unoscillated event rate in the geoneutrino energy window due to the storage of spent nuclear fuels in the cooling pools. We predict that the research reactors contribute to less than 0.2% to the commercial reactor signal in the investigated 14 sites. We perform a multitemporal analysis of the expected reactor signal over a time lapse of ten years using reactor operational records collected in a comprehensive database published at www.fe.infn.it/antineutrino.

  7. ANAEROBIC DIGESTION AND THE DENITRIFICATION IN UASB REACTOR

    OpenAIRE

    José Tavares de Sousa; Keliana Dantas Santos; Israel Nunes Henrique; Danielle Patrício Brasil; Eclésio Cavalcante Santos

    2008-01-01

    The environmental conditions in Brazil have been contributing to the development of anaerobic systems in the treatment of wastewaters, especially UASB - Upflow Anaerobic Sludge Blanket reactors. The classic biological process for removal of nutrients uses three reactors - Bardenpho System, therefore, this work intends an alternative system, where the anaerobic digestion and the denitrification happen in the same reactor reducing the number of reactors for two. The experimental system was con...

  8. Bioremediation of chlorpyrifos contaminated soil by two phase bioslurry reactor: Processes evaluation and optimization by Taguchi's design of experimental (DOE) methodology.

    Science.gov (United States)

    Pant, Apourv; Rai, J P N

    2018-04-15

    Two phase bioreactor was constructed, designed and developed to evaluate the chlorpyrifos remediation. Six biotic and abiotic factors (substrate-loading rate, slurry phase pH, slurry phase dissolved oxygen (DO), soil water ratio, temperature and soil micro flora load) were evaluated by design of experimental (DOE) methodology employing Taguchi's orthogonal array (OA). The selected six factors were considered at two levels L-8 array (2^7, 15 experiments) in the experimental design. The optimum operating conditions obtained from the methodology showed enhanced chlorpyrifos degradation from 283.86µg/g to 955.364µg/g by overall 70.34% of enhancement. In the present study, with the help of few well defined experimental parameters a mathematical model was constructed to understand the complex bioremediation process and optimize the approximate parameters upto great accuracy. Copyright © 2017 Elsevier Inc. All rights reserved.

  9. Nuclear reactor neutron shielding

    Science.gov (United States)

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

    2017-09-12

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

  10. Design of the experimental apparatus to obtain a thermal neutron beam, intermediate-energy neutrons (2-144 keV) and high-energy photons (6 MeV) by means of the TRIGA reactor at the ENEA Casaccia center

    CERN Document Server

    Laitano, R F

    1987-01-01

    Design of the experimental apparatus to obtain a thermal neutron beam, intermediate-energy neutrons (2-144 keV) and high-energy photons (6 MeV) by means of the TRIGA reactor at the ENEA Casaccia center

  11. Modelization of physical phenomena in research reactors with the help of new developments in transport methods, and methodology validation with experimental data; Modelisation des phenomenes physiques dans les reacteurs de recherche a l'aide de developpements realises dans les methodes de transport et qualification

    Energy Technology Data Exchange (ETDEWEB)

    Rauck, St

    2000-10-01

    The aim of this work is to develop a scheme for experimental reactors, based on transport equations. This type of reactors is characterized by a small core, a complex, very heterogeneous geometry and a large leakage. The possible insertion of neutron beams in the reflector and the presence of absorbers in the core increase the difficulty of the 3D-geometrical description and the physical modeling of the component parameters of the reactor. The Orphee reactor has been chosen for our study. Physical models (homogenization, collapsing cross section in few groups, albedo multigroup condition) have been developed in the APOLLO2 and CRONOS2 codes to calculate flux and power maps in a 3D-geometry, with different burnup and through transport equations. Comparisons with experimental measurements have shown the interest of taking into account anisotropy, steep flux gradients by using Sn methods, and on the other hand using a 12-group cross section library. The modeling of neutron beams has been done outside the core modeling through Monte Carlo calculations and with the total geometry, including a large thickness of heavy water. Thanks to this calculations, one can evaluate the neutron beams anti-reactivity and determinate the core cycle. We assure these methods more accurate than usual transport-diffusion calculations will be used for the conception of new research reactors. (author)

  12. First experimental results of particle re-suspension in a low pressure wind tunnel applied to the issue of dust in fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rondeau, Anthony, E-mail: anthony.rondeau@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN-RES/SCA, Gif-sur-Yvette 91192 (France); Merrison, Jonathan; Iversen, Jens Jacob [Department of Physics and Astronomy, Institute for Storage Ring Facilities, 8000 Aarhus C (Denmark); Peillon, Samuel; Sabroux, Jean-Christophe; Lemaitre, Pascal; Gensdarmes, François [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN-RES/SCA, Gif-sur-Yvette 91192 (France); Chassefière, Eric [Laboratoire Géosciences Paris Sud (GEOPS), UMR 8148, Université Paris Sud, 91403 Orsay Cedex (France)

    2015-10-15

    Highlights: • The first experimental data of dust re-suspension performed by controlled airflow (in terms of airflow velocity and fluid density). • The effect of the surrounding pressure in the re-suspension mechanism. • The friction (shear velocity) reduction at low pressure in the transient regime (Knudsen number close to one). • The importance of the adhesion forces between particles in dust mobilization by clustering. • The fact that the existing re-suspension models of the relevant literature do not take into account this clustering phenomenon. - Abstract: During the normal operating condition of the future ITER tokamak, a massive production of dust in the toroidal vacuum vessel is expected. This dust, originating from the erosion of tungsten and beryllium internal walls of the torus by the plasma, would be mobilized to some extent during a loss of vacuum accident (LOVA). For safety reasons, it is essential to quantify the re-suspended dust fraction during such an event. Here, we provide preliminary experimental data of dust re-suspension obtained in the wind tunnel of the European Space Agency (ESA) at low pressures (300, 130 and 10 mbar). The experimentations were performed with multilayer deposits. We used two powders with a median diameter at 15.5 μm and 21.8 μm. A negative influence of the low pressure in the re-suspension mechanism is observed. For example, given a re-suspension fraction of 10%, increasing friction shear velocities are derived for decreasing absolute pressures: 300 mbar/0.66 m s{sup −1}; 130 mbar/1.08 m s{sup −1}; and 10 mbar/1.84 m s{sup −1}. In addition, we highlight the friction reduction for Kundsen numbers greater than 0.1 by an analysis of the airflow forces.

  13. Photoneutron effects on pulse reactor kinetics for the Annular Core Research Reactor (ACRR).

    Energy Technology Data Exchange (ETDEWEB)

    Parma, Edward J., Jr.

    2009-06-01

    The Annular Core Research Reactor (ACRR) is a swimming-pool type pulsed reactor that maintains an epithermal neutron flux and a nine-inch diameter central dry cavity. One of its uses is neutron and gamma-ray irradiation damage studies on electronic components under transient reactor power conditions. In analyzing the experimental results, careful attention must be paid to the kinetics associated with the reactor to ensure that the transient behavior of the electronic device is understood. Since the ACRR fuel maintains a substantial amount of beryllium, copious quantities of photoneutrons are produced that can significantly alter the expected behavior of the reactor power, especially following a reactor pulse. In order to understand these photoneutron effects on the reactor kinetics, the KIFLE transient reactor-analysis code was modified to include the photoneutron groups associated with the beryllium. The time-dependent behavior of the reactor power was analyzed for small and large pulses, assuming several initial conditions including following several pulses during the day, and following a long steady-state power run. The results indicate that, for these types of initial conditions, the photoneutron contribution to the reactor pulse energy can have a few to tens of percent effect.

  14. Experience gathered in dismounting pressure vessel internals of the Kahl experimental reactor (VAK Kahl); Erfahrungen bei der Zerlegung der Druckgefaesseinbauten im VAK Kahl

    Energy Technology Data Exchange (ETDEWEB)

    Kalwa, H. [Versuchsatomkraftwerk Kahl GmbH, Karlstein am Main (Germany); Eickelpasch, N. [Versuchsatomkraftwerk Kahl GmbH, Karlstein am Main (Germany); Reiter, W. [Versuchsatomkraftwerk Kahl GmbH, Karlstein am Main (Germany)

    1995-11-01

    Mechanical cutting processes are preferred for dismantling of activated pressure vessel internals in the VAK Kahl. Up to now, sawing, nibbling, grinding and electroeroding have been tried. For a realistic under-water trial run on a scale of 1:1, an inactive experimental facility was erected in the empty turbine building of the VAK. (orig./DG) [Deutsch] Bei der Zerlegung der aktivierten RDB-Einbauten im Versuchsatomkraftwerk Kahl werden bevorzugt mechanische Trennverfahren angewendet. Erprobt und eingesetzt wurden bisher das Saegen, das Knabbern, das Schleifen und das Elektroerodieren. Fuer eine realitaetsnahe Erprobung unter Wasser im Massstab 1:1 wurde in VAK ein inaktiver Versuchsstand im bereits leergeraeumten Maschinenhaus erstellt. (orig./DG)

  15. In-situ Condition Monitoring of Components in Small Modular Reactors Using Process and Electrical Signature Analysis. Final report, volume 1. Development of experimental flow control loop, data analysis and plant monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Upadhyaya, Belle [Univ. of Tennessee, Knoxville, TN (United States); Hines, J. Wesley [Univ. of Tennessee, Knoxville, TN (United States); Damiano, Brian [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Mehta, Chaitanya [Univ. of Tennessee, Knoxville, TN (United States); Collins, Price [Univ. of Tennessee, Knoxville, TN (United States); Lish, Matthew [Univ. of Tennessee, Knoxville, TN (United States); Cady, Brian [Univ. of Tennessee, Knoxville, TN (United States); Lollar, Victor [Univ. of Tennessee, Knoxville, TN (United States); de Wet, Dane [Univ. of Tennessee, Knoxville, TN (United States); Bayram, Duygu [Univ. of Tennessee, Knoxville, TN (United States)

    2015-12-15

    The research and development under this project was focused on the following three major objectives: Objective 1: Identification of critical in-vessel SMR components for remote monitoring and development of their low-order dynamic models, along with a simulation model of an integral pressurized water reactor (iPWR). Objective 2: Development of an experimental flow control loop with motor-driven valves and pumps, incorporating data acquisition and on-line monitoring interface. Objective 3: Development of stationary and transient signal processing methods for electrical signatures, machinery vibration, and for characterizing process variables for equipment monitoring. This objective includes the development of a data analysis toolbox. The following is a summary of the technical accomplishments under this project: - A detailed literature review of various SMR types and electrical signature analysis of motor-driven systems was completed. A bibliography of literature is provided at the end of this report. Assistance was provided by ORNL in identifying some key references. - A review of literature on pump-motor modeling and digital signal processing methods was performed. - An existing flow control loop was upgraded with new instrumentation, data acquisition hardware and software. The upgrading of the experimental loop included the installation of a new submersible pump driven by a three-phase induction motor. All the sensors were calibrated before full-scale experimental runs were performed. - MATLAB-Simulink model of a three-phase induction motor and pump system was completed. The model was used to simulate normal operation and fault conditions in the motor-pump system, and to identify changes in the electrical signatures. - A simulation model of an integral PWR (iPWR) was updated and the MATLAB-Simulink model was validated for known transients. The pump-motor model was interfaced with the iPWR model for testing the impact of primary flow perturbations (upsets) on

  16. Continuous esterification to produce biodiesel by SPES/PES/NWF composite catalytic membrane in flow-through membrane reactor: experimental and kinetic studies.

    Science.gov (United States)

    Shi, Wenying; He, Benqiao; Cao, Yuping; Li, Jianxin; Yan, Feng; Cui, Zhenyu; Zou, Zhiqun; Guo, Shiwei; Qian, Xiaomin

    2013-02-01

    A novel composite catalytic membrane (CCM) was prepared from sulfonated polyethersulfone (SPES) and polyethersulfone (PES) blend supported by non-woven fabrics, as a heterogeneous catalyst to produce biodiesel from continuous esterification of oleic acid with methanol in a flow-through mode. A kinetic model of esterification was established based on a plug-flow assumption. The effects of the CCM structure (thickness, area, porosity, etc.), reaction temperature and the external and internal mass transfer resistances on esterification were investigated. The results showed that the CCM structure had a significant effect on the acid conversion. The external mass transfer resistance could be neglected when the flow rate was over 1.2 ml min(-1). The internal mass transfer resistance impacted on the conversion when membrane thickness was over 1.779 mm. An oleic acid conversion kept over 98.0% for 500 h of continuous running. The conversions obtained from the model are in good agreement with the experimental data. Copyright © 2012 Elsevier Ltd. All rights reserved.

  17. Prolongation of the BOR-60 reactor operation

    Directory of Open Access Journals (Sweden)

    Alexey l. Izhutov

    2015-04-01

    Full Text Available The fast neutron reactor BOR-60 is one of the key experimental facilities worldwide to perform large-scale tests of fuel, absorbing, and structural materials for advanced reactors. The BOR-60 reactor was put into operation in December 1969, and by the end of 2014 it had been operating on power for ∼265,000 hours. BOR-60 still demonstrates potential capabilities to extend the lifetime of sodium-cooled fast reactors. The BOR-60 lifetime should have expired at the end of 2014. Over the past few years, a great scope of work has been performed to justify the possibility of extending its lifetime. The work included inspection of the equipment conditions, calculations and experimental research on operating parameters and the conditions of nonremovable components, investigation of the structural material samples after their long-term operation under irradiation, etc. Based on the results of the work performed, the residual lifetime was evaluated and the reactor operator made a decision to extend the lifetime period of the BOR-60 reactor. After considering both a set of documents about the reactor conditions and the positive decision of independent experts, the Regulatory Authority of the Russian Federation extended the BOR-60 operating license up to 2020.

  18. Experimental study of biological nitrogen and phosphorus removal in anaerobic/anoxic SBR reactor with separated biofilm nitrification; Indagine sperimentale sulla rimozione biologica di azoto e fosforo in un reattore SBR anaerobico/anossico con nitrificazione separata su biofiltro

    Energy Technology Data Exchange (ETDEWEB)

    Bortone, G.; Malaspina, F.; Stante, L.; Tilche, A. [ENEA, Bologna (Italy). Area Energia Ambiente e Salute; Andreottola, G. [Trento Univ. (Italy). Dipt. di Ingegneria Civile-Ambientale

    1995-04-01

    An Anaerobic/Anoxic Sequencing Batch Reactor (A/A SBR) with separated batch biofilm nitrification was tested for nutrient removal against a five step anaerobic anoxic/Oxic SBR (A/O SBR). Piggery waste water, particularly challenging for its low COD/N ratio, was used as feed. A/A SBR, especially in the configuration tested in the second experimental period, showed very good N (97%) and P (90%) removal capacities with excellent sludge settling properties (SVI 52 ml/gSS). On the other hand, organic carbon removal efficiency with nitrate was a little bit lower than with oxygen (90% in A/A SBR against 92% in A/O SBR). Batch biofilm nitrification was very effective, with very high nitrification rates (1.4 mgN gVSS{sup -1} h{sup -1}; 0.1 khN m{sup 3}d{sup -1}). Presence of poly P bacteria in the A/A SBR sludge was assessed through microscopic observation and from the high cellular poly-phosphate content.

  19. Neutronics qualification of the Jules Horowitz reactor fuel by interpretation of the VALMONT experimental program - Transposition of the uncertainties on the reactivity of JHR with JEF2.2 and JEFF3.1.1

    Energy Technology Data Exchange (ETDEWEB)

    Leray, O.; Hudelot, J. P. [DEN Nuclear Projects Laboratory, CEA Cadarache, F-13108 St Paul-Lez-Durance (France); Antony, M. [DEN Experimental Physics Laboratory, CEA Cadarache, F-13108 St Paul-Lez-Durance (France); Doederlein, C. [DEN Nuclear Projects Laboratory, CEA Cadarache, F-13108 St Paul-Lez-Durance (France); Santamarina, A. [DEN Reactor and Cycle Physics Service, CEA Cadarache, F-13108 St Paul-Lez-Durance (France); Bernard, D. [DEN Physical Studies Laboratory, CEA Cadarache, F-13108 St Paul-Lez-Durance (France); Vaglio-Gaudard, C. [DEN Nuclear Projects Laboratory, CEA Cadarache, F-13108 St Paul-Lez-Durance (France)

    2011-07-01

    The new European material testing Jules Horowitz Reactor (JHR), currently under construction in Cadarache center (CEA France), will use LEU (20% enrichment in {sup 235}U) fuels (U{sub 3}Si{sub 2} for the start up and UMoAl in the future) which are quite different from the industrial oxide fuel, for which an extensive neutronics qualification database has been established. The HORUS3D/N neutronics calculation scheme, used for the design and safety studies of the JHR, is being developed within the framework of a rigorous verification-validation-qualification methodology. In this framework, the experimental VALMONT (Validation of Aluminium Molybdenum uranium fuel for Neutronics) program has been performed in the MINERVE facility of CEA Cadarache (France), in order to qualify the capability of HORUS3D/N to accurately calculate the reactivity of the JHR reactor. The MINERVE facility using the oscillation technique provides accurate measurements of reactivity effect of samples. The VALMONT program includes oscillations of samples of UAl{sub x}/Al and UMo/Al with enrichments ranging from 0.2% to 20% and Uranium densities from 2.2 to 8 g/cm{sup 3}. The geometry of the samples and the pitch of the experimental lattice ensure maximum representativeness with the neutron spectrum expected for JHR. By comparing the effect of the sample with the one of a known fuel specimen, the reactivity effect can be measured in absolute terms and be compared to computational results. Special attention was paid to the rigorous determination and reduction of the experimental uncertainties. The calculational analysis of the VALMONT results was performed with the French deterministic code APOLLO2. A comparison of the impact of the different calculation methods, data libraries and energy meshes that were tested is presented. The interpretation of the VALMONT experimental program allowed the qualification of JHR fuel UMoAl8 (with an enrichment of 19.75% {sup 235}U) by the Minerve

  20. Experimental and CFD Studies of Coolant Flow Mixing within Scaled Models of the Upper and Lower Plenums of NGNP Gas-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hassan, Yassin [Texas A & M Univ., College Station, TX (United States); Anand, Nk [Texas A & M Univ., College Station, TX (United States)

    2016-03-30

    A 1/16th scaled VHTR experimental model was constructed and the preliminary test was performed in this study. To produce benchmark data for CFD validation in the future, the facility was first run at partial operation with five pipes being heated. PIV was performed to extract the vector velocity field for three adjacent naturally convective jets at statistically steady state. A small recirculation zone was found between the pipes, and the jets entered the merging zone at 3 cm from the pipe outlet but diverged as the flow approached the top of the test geometry. Turbulence analysis shows the turbulence intensity peaked at 41-45% as the jets mixed. A sensitivity analysis confirmed that 1000 frames were sufficient to measure statistically steady state. The results were then validated by extracting the flow rate from the PIV jet velocity profile, and comparing it with an analytic flow rate and ultrasonic flowmeter; all flow rates lie within the uncertainty of the other two methods for Tests 1 and 2. This test facility can be used for further analysis of naturally convective mixing, and eventually produce benchmark data for CFD validation for the VHTR during a PCC or DCC accident scenario. Next, a PTV study of 3000 images (1500 image pairs) were used to quantify the velocity field in the upper plenum. A sensitivity analysis confirmed that 1500 frames were sufficient to precisely estimate the flow. Subsequently, three (3, 9, and 15 cm) Y-lines from the pipe output were extracted to consider the output differences between 50 to 1500 frames. The average velocity field and standard deviation error that accrued in the three different tests were calculated to assess repeatability. The error was varied, from 1 to 14%, depending on Y-elevation. The error decreased as the flow moved farther from the output pipe. In addition, turbulent intensity was calculated and found to be high near the output. Reynolds stresses and turbulent intensity were used to validate the data by

  1. Studying the properties and behaviour of high-level radioactive wastes from the BOR-60 reactor U-Pu and U spent fuel experimental gas-fluoride reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Kirillovich, A.P.; Lavrinovich, Yu.G.; Vorobej, M.P.; Pimonov, Yu.I.

    1982-07-01

    The results of investigations of physical-chemical and radiation properties of fluoride radioactive wastes produced during experimental reprocessing of spent oxide uranium-plutonium fuel as well as high-level radioactive waste behaviour in the process of their six-year controlled storage are presented. Radioactive gas release from solid wastes, gaseous phase composition, radionuclide leaching rate are determined. Investigations are performed at a special bench. Energy release of the spent fuel and high-level radioactive wastes is determined by means of heat-conducting type calorimeter. Gas mixture composition in containers with wastes is determined by mass-spectrometric method at equilibrium temperature of high-level radioactive product self-heating and at external container heating up to 700 deg C. Thermal-physical characteristics of solid fluoride wastes are found by the differential thermography method under quasistationary heating. The results obtained show that about a half (44.8-60.9%) of fission product radioactivity is concentrated in fluorination wastes, specific heat release of which constitutes 50-52 W/kg, while ..beta..-activity exceeds 550 TBq/kg. Main contribution into ..beta..-activity is made by Ce, /sup 144/Pr, Ru, /sup 106/Rh, Zr, /sup 95/Nb, /sup 137/Cs. With waste storage time increase their thermal stability increases. It is concluded that the investigation results can be used for calculating the conditions of safe storage of high-level radiactive solid fluoride wastes and optimization of the technological process of spent fuel gas-fluoride reprossing.

  2. Hybrid plasmachemical reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lelevkin, V. M., E-mail: lelevkin44@mail.ru; Smirnova, Yu. G.; Tokarev, A. V. [Kyrgyz-Russian Slavic University (Kyrgyzstan)

    2015-04-15

    A hybrid plasmachemical reactor on the basis of a dielectric barrier discharge in a transformer is developed. The characteristics of the reactor as functions of the dielectric barrier discharge parameters are determined.

  3. Attrition reactor system

    Science.gov (United States)

    Scott, Charles D.; Davison, Brian H.

    1993-01-01

    A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur.

  4. Guidebook to nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Nero, A.V. Jr.

    1976-05-01

    A general introduction to reactor physics and theory is followed by descriptions of commercial nuclear reactor types. Future directions for nuclear power are also discussed. The technical level of the material is suitable for laymen.

  5. Equipment for neutron measurements at VR-1 Sparrow training reactor.

    Science.gov (United States)

    Kolros, Antonin; Huml, Ondrej; Kríz, Martin; Kos, Josef

    2010-01-01

    The VR-1 sparrow reactor is an experimental nuclear facility for training, student education and teaching purposes. The sparrow reactor is an educational platform for the basic experiments at the reactor physic and dosimetry. The aim of this article is to describe the new experimental equipment EMK310 features and possibilities for neutron detection by different gas filled detectors at VR-1 reactor. Among the EMK310 equipment typical attributes belong precise set-up, simple control, resistance to electromagnetic interference, high throughput (counting rate), versatility and remote controllability. The methods for non-linearity correction of pulse neutron detection system and reactimeter application are presented. Copyright 2009. Published by Elsevier Ltd.

  6. NUCLEAR REACTOR CONTROL SYSTEM

    Science.gov (United States)

    Epler, E.P.; Hanauer, S.H.; Oakes, L.C.

    1959-11-01

    A control system is described for a nuclear reactor using enriched uranium fuel of the type of the swimming pool and other heterogeneous nuclear reactors. Circuits are included for automatically removing and inserting the control rods during the course of normal operation. Appropriate safety circuits close down the nuclear reactor in the event of emergency.

  7. Nuclear reactor overflow line

    Science.gov (United States)

    Severson, Wayne J.

    1976-01-01

    The overflow line for the reactor vessel of a liquid-metal-cooled nuclear reactor includes means for establishing and maintaining a continuous bleed flow of coolant amounting to 5 to 10% of the total coolant flow through the overflow line to prevent thermal shock to the overflow line when the reactor is restarted following a trip. Preferably a tube is disposed concentrically just inside the overflow line extending from a point just inside the reactor vessel to an overflow tank and a suction line is provided opening into the body of liquid metal in the reactor vessel and into the annulus between the overflow line and the inner tube.

  8. Light water reactor safety

    CERN Document Server

    Pershagen, B

    2013-01-01

    This book describes the principles and practices of reactor safety as applied to the design, regulation and operation of light water reactors, combining a historical approach with an up-to-date account of the safety, technology and operating experience of both pressurized water reactors and boiling water reactors. The introductory chapters set out the basic facts upon which the safety of light water reactors depend. The central section is devoted to the methods and results of safety analysis. The accidents at Three Mile Island and Chernobyl are reviewed and their implications for light wate

  9. Nuclear reactor physics

    CERN Document Server

    Stacey, Weston M

    2010-01-01

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

  10. Nuclear reactors built, being built, or planned, 1991

    Energy Technology Data Exchange (ETDEWEB)

    Simpson, B.

    1992-07-01

    This document contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1991. The book is divided into three major sections: Section 1 consists of a reactor locator map and reactor tables; Section 2 includes nuclear reactors that are operating, being built, or planned; and Section 3 includes reactors that have been shut down permanently or dismantled. Sections 2 and 3 contain the following classification of reactors: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor is an American company -- working either independently or in cooperation with a foreign company (Part 4, in each section). Critical assembly refers to an assembly of fuel and assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5).

  11. Spinning fluids reactor

    Science.gov (United States)

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

    2012-11-20

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

  12. Massive computation methodology for reactor operation (MACRO)

    Energy Technology Data Exchange (ETDEWEB)

    Gustavsson, Cecilia; Pomp, Stephan; Sjoestrand, Henrik; Wallin, Gustav; Oesterlund, Michael [Division of applied nuclear physics, Department of physics and astronomy, Uppsala University, Laegerhyddsvaegen 1, 751 20 Uppsala (Sweden); Koning, Arjan; Rochman, Dimitri [Nuclear Research and consultancy Group (NRG) Westerduinweg 3, Petten (Netherlands); Bejmer, Klaes-Hakan [Vattenfall Nuclear Fuel AB, Jaemtlandsgatan 99, Vaellingby (Sweden); Henriksson, Hans [Vattenfall Research and Development AB, Jaemtlandsgatan 99, Vaellingby (Sweden)

    2010-07-01

    Today, nuclear data libraries do not handle uncertainties from nuclear data in a consistent manner and the reactor codes do not request uncertainties in nuclear data input. Thus, the output from these codes have unknown uncertainties. The plan is to use a method proposed by Koning and Rochman to investigate the propagation of nuclear data uncertainties into reactor physics codes and macroscopic parameters. A project (acronym MACRO) has started at Uppsala University in collaboration with A. Koning and with financial support from Vattenfall AB and the Swedish Research Council within the GENIUS (Generation IV research in universities of Sweden) project. In the proposed method the uncertainties in nuclear model parameters will be derived from theoretical considerations and comparisons of nuclear model results with experimental cross-section data. Given the probability distribution in the model parameters a large set of random, complete ENDF-formatted nuclear data libraries will be created using the TALYS code. The generated nuclear data libraries will then be used in neutron transport codes to obtain macroscopic reactor parameters. For this, models of reactor systems with proper geometry and elements will be used. This will be done for all data libraries and the variation of the final results will be regarded as a systematic uncertainty in the investigated reactor parameter. The understanding of these systematic uncertainties is especially important for the design and intercomparison of new reactor concepts, i.e., Generation IV, and optimization applications for current generation reactors is envisaged. (authors)

  13. Reactor Monitoring with Antineutrinos - A Progress Report

    Science.gov (United States)

    Bernstein, Adam

    2012-08-01

    The Reactor Safeguards regime is the name given to a set of protocols and technologies used to monitor the consumption and production of fissile materials in nuclear reactors. The Safeguards regime is administered by the International Atomic Energy Agency (IAEA), and is an essential component of the global Treaty on Nuclear Nonproliferation, recently renewed by its 189 remaining signators. (The 190th, North Korea, withdrew from the Treaty in 2003). Beginning in Russia in the 1980s, a number of researchers worldwide have experimentally demonstrated the potential of cubic meter scale antineutrino detectors for non-intrusive real-time monitoring of fissile inventories and power output of reactors. The detectors built so far have operated tens of meters from a reactor core, outside of the containment dome, largely unattended and with remote data acquisition for an entire 1.5 year reactor cycle, and have achieved levels of sensitivity to fissile content of potential interest for the IAEA safeguards regime. In this article, I will describe the unique advantages of antineutrino detectors for cooperative monitoring, consider the prospects and benefits of increasing the range of detectability for small reactors, and provide a partial survey of ongoing global research aimed at improving near-field and far field monitoring and discovery of nuclear reactors.

  14. Conduction heat transfer in a cylindrical dielectric barrier discharge reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sadat, H. [Laboratoire d' Etudes Thermiques, Universite de Poitiers, 40 Avenue du Recteur Pineau, 86022 Poitiers (France)], E-mail: hamou.sadat@univ-poitiers.fr; Dubus, N. [Laboratoire d' Etudes Thermiques, Universite de Poitiers, 40 Avenue du Recteur Pineau, 86022 Poitiers (France); Pinard, L.; Tatibouet, J.M.; Barrault, J. [Laboratoire en catalyse et chimie organique, Universite de Poitiers, 40 Avenue du Recteur Pineau, 86022 Poitiers (France)

    2009-04-15

    The thermal behaviour of a dielectric barrier discharge reactor is studied. The experimental tests are performed on a laboratory reactor with two working fluids: helium and air. A simple heat conduction model for calculating the heat loss is developed. By using temperature measurements in the internal and external electrodes, a thermal resistance of the reactor is defined. Finally, the percentage of the input power that is dissipated to the environment is given.

  15. Consumption of the electric power inside silent discharge reactors

    Energy Technology Data Exchange (ETDEWEB)

    Yehia, Ashraf, E-mail: yehia30161@yahoo.com [Department of Physics, Faculty of Science, Assiut University, Assiut 71516, Arab Republic of Egypt and Department of Physics, College of Science and Humanitarian Studies at Alkharj, Salman bin Abdulaziz University, P.O. Box 83, Alkharj 11942 (Saudi Arabia)

    2015-01-15

    An experimental study was made in this paper to investigate the relation between the places of the dielectric barriers, which cover the surfaces of the electrodes in the coaxial cylindrical reactors, and the rate of change of the electric power that is consumed in forming silent discharges. Therefore, silent discharges have been formed inside three coaxial cylindrical reactors. The dielectric barriers in these reactors were pasted on both the internal surface of the outer electrode in the first reactor and the external surface of the inner electrode in the second reactor as well as the surfaces of the two electrodes in the third reactor. The reactor under study has been fed by atmospheric air that flowed inside it with a constant rate at normal temperature and pressure, in parallel with the application of a sinusoidal ac voltage between the electrodes of the reactor. The electric power consumed in forming the silent discharges inside the three reactors was measured as a function of the ac peak voltage. The validity of the experimental results was investigated by applying Manley's equation on the same discharge conditions. The results have shown that the rate of consumption of the electric power relative to the ac peak voltage per unit width of the discharge gap improves by a ratio of either 26.8% or 80% or 128% depending on the places of the dielectric barriers that cover the surfaces of the electrodes inside the three reactors.

  16. Training experience at Experimental Breeder Reactor II

    Energy Technology Data Exchange (ETDEWEB)

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

    1978-01-01

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

  17. Chemical reactor modeling multiphase reactive flows

    CERN Document Server

    Jakobsen, Hugo A

    2014-01-01

    Chemical Reactor Modeling closes the gap between Chemical Reaction Engineering and Fluid Mechanics.  The second edition consists of two volumes: Volume 1: Fundamentals. Volume 2: Chemical Engineering Applications In volume 1 most of the fundamental theory is presented. A few numerical model simulation application examples are given to elucidate the link between theory and applications. In volume 2 the chemical reactor equipment to be modeled are described. Several engineering models are introduced and discussed. A survey of the frequently used numerical methods, algorithms and schemes is provided. A few practical engineering applications of the modeling tools are presented and discussed. The working principles of several experimental techniques employed in order to get data for model validation are outlined. The monograph is based on lectures regularly taught in the fourth and fifth years graduate courses in transport phenomena and chemical reactor modeling, and in a post graduate course in modern reactor m...

  18. Gas pollutant cleaning by a membrane reactor

    Directory of Open Access Journals (Sweden)

    Kaldis Sotiris

    2006-01-01

    Full Text Available An alternative technology for the removal of gas pollutants at the integrated gasification combined cycle process for power generation is the use of a catalytic membrane reactor. In the present study, ammonia decomposition in a catalytic reactor, with a simultaneous removal of hydrogen through a ceramic membrane, was investigated. A Ni/Al2O3 catalyst was prepared by the dry and wet impregnation method and characterized by the inductively coupled plasma method, scanning electron microscopy, X-ray diffraction, and N2 adsorption before and after activation. Commercially available a-Al2O3 membranes were also characterized and the permeabilities and permselectivities of H2, N2, and CO2 were measured by the variable volume method. In parallel with the experimental analysis, the necessary mathematical models were developed to describe the operation of the catalytic membrane reactor and to compare its performance with the conventional reactor. .

  19. Neutron fluxes in test reactors

    Energy Technology Data Exchange (ETDEWEB)

    Youinou, Gilles Jean-Michel [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-01-01

    Communicate the fact that high-power water-cooled test reactors such as the Advanced Test Reactor (ATR), the High Flux Isotope Reactor (HFIR) or the Jules Horowitz Reactor (JHR) cannot provide fast flux levels as high as sodium-cooled fast test reactors. The memo first presents some basics physics considerations about neutron fluxes in test reactors and then uses ATR, HFIR and JHR as an illustration of the performance of modern high-power water-cooled test reactors.

  20. Advanced Computational Thermal Fluid Physics (CTFP) and Its Assessment for Light Water Reactors and Supercritical Reactors

    Energy Technology Data Exchange (ETDEWEB)

    D.M. McEligot; K. G. Condie; G. E. McCreery; H. M. McIlroy; R. J. Pink; L.E. Hochreiter; J.D. Jackson; R.H. Pletcher; B.L. Smith; P. Vukoslavcevic; J.M. Wallace; J.Y. Yoo; J.S. Lee; S.T. Ro; S.O. Park

    2005-10-01

    Background: The ultimate goal of the study is the improvement of predictive methods for safety analyses and design of Generation IV reactor systems such as supercritical water reactors (SCWR) for higher efficiency, improved performance and operation, design simplification, enhanced safety and reduced waste and cost. The objective of this Korean / US / laboratory / university collaboration of coupled fundamental computational and experimental studies is to develop the supporting knowledge needed for improved predictive techniques for use in the technology development of Generation IV reactor concepts and their passive safety systems. The present study emphasizes SCWR concepts in the Generation IV program.

  1. HORIZONTAL BOILING REACTOR SYSTEM

    Science.gov (United States)

    Treshow, M.

    1958-11-18

    Reactors of the boiling water type are described wherein water serves both as the moderator and coolant. The reactor system consists essentially of a horizontal pressure vessel divided into two compartments by a weir, a thermal neutronic reactor core having vertical coolant passages and designed to use water as a moderator-coolant posltioned in one compartment, means for removing live steam from the other compartment and means for conveying feed-water and water from the steam compartment to the reactor compartment. The system further includes auxiliary apparatus to utilize the steam for driving a turbine and returning the condensate to the feed-water inlet of the reactor. The entire system is designed so that the reactor is self-regulating and has self-limiting power and self-limiting pressure features.

  2. Fast Spectrum Reactors

    CERN Document Server

    Todd, Donald; Tsvetkov, Pavel

    2012-01-01

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

  3. POTENTIAL BENCHMARKS FOR ACTINIDE PRODUCTION IN HANFORD REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    PUIGH RJ; TOFFER H

    2011-10-19

    A significant experimental program was conducted in the early Hanford reactors to understand the reactor production of actinides. These experiments were conducted with sufficient rigor, in some cases, to provide useful information that can be utilized today in development of benchmark experiments that may be used for the validation of present computer codes for the production of these actinides in low enriched uranium fuel.

  4. THERMAL NEUTRONIC REACTOR

    Science.gov (United States)

    Spinrad, B.I.

    1960-01-12

    A novel thermal reactor was designed in which a first reflector formed from a high atomic weight, nonmoderating material is disposed immediately adjacent to the reactor core. A second reflector composed of a moderating material is disposed outwardly of the first reflector. The advantage of this novel reflector arrangement is that the first reflector provides a high slow neutron flux in the second reflector, where irradiation experiments may be conducted with a small effect on reactor reactivity.

  5. Development of reactor graphite

    Science.gov (United States)

    Haag, G.; Mindermann, D.; Wilhelmi, G.; Persicke, H.; Ulsamer, W.

    1990-04-01

    The German graphite development programme for High Temperature Reactors has been based on the assumption that reactor graphite for core components with lifetime fluences of up to 4 × 10 22 neutrons per cm 2 (EDN) at 400°C can be manufactured from regular pitch coke. The use of secondary coke and vibrational moulding techniques have allowed production of materials with very small anisotropy, high strength, and high purity which are the most important properties of reactor graphite. A variety of graphite grades has been tested in fast neutron irradiation experiments. The results show that suitable graphites for modern High Temperature Reactors with spherical fuel elements are available.

  6. Oscillatory flow chemical reactors

    National Research Council Canada - National Science Library

    Slavnić Danijela S; Bugarski Branko M; Nikačević Nikola M

    2014-01-01

    .... However, the reactions of interests for the mentioned industry sectors are often slow, thus continuous tubular reactors would be impractically long for flow regimes which provide sufficient heat...

  7. Membrane reactors at Degussa.

    Science.gov (United States)

    Wöltinger, Jens; Karau, Andreas; Leuchtenberger, Wolfgang; Drauz, Karlheinz

    2005-01-01

    The review covers the development of membrane reactor technologies at Degussa for the synthesis of fine chemicals. The operation of fed-batch or continuous biocatalytic processes in the enzyme membrane reactor (EMR) is well established at Degussa. Degussa has experience of running EMRs from laboratory gram scale up to a production scale of several hundreds of tons per year. The transfer of the enzyme membrane reactor from biocatalysis to chemical catalysis in the chemzyme membrane reactor (CMR) is discussed. Various homogeneous catalysts have been investigated in the CMR, and the scope and limitation of this new technique is discussed.

  8. Pressurizing new reactors

    Energy Technology Data Exchange (ETDEWEB)

    Neill, J.S.

    1956-01-30

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

  9. 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.

  10. Antineutrino and gamma emission from the OSIRIS research reactor

    Science.gov (United States)

    Giot, Lydie; Fallot, Muriel

    2017-09-01

    For the first time, the summation method has been coupled with a complete reactor model, in order to predict the antineutrino emission of a research reactor. This work, discussed in the first part of this paper, allows us to predict the low energy part of the antineutrino spectrum, evidencing the important contribution of actinides to the antineutrino emission. Experimental conditions at short distance from research reactors are challenging, because the reactor itself produces huge gamma background that induce accidental and correlated backgrounds in an antineutrino target. The understanding of this background is of utmost importance and triggered the second part of the work presented here.

  11. Prospects for Tokamak Fusion Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sheffield, J.; Galambos, J.

    1995-04-01

    This paper first reviews briefly the status and plans for research in magnetic fusion energy and discusses the prospects for the tokamak magnetic configuration to be the basis for a fusion power plant. Good progress has been made in achieving fusion reactor-level, deuterium-tritium (D-T) plasmas with the production of significant fusion power in the Joint European Torus (up to 2 MW) and the Tokamak Fusion Test Reactor (up to 10 MW) tokamaks. Advances on the technologies of heating, fueling, diagnostics, and materials supported these achievements. The successes have led to the initiation of the design phases of two tokamaks, the International Thermonuclear Experimental Reactor (ITER) and the US Toroidal Physics Experiment (TPX). ITER will demonstrate the controlled ignition and extended bum of D-T plasmas with steady state as an ultimate goal. ITER will further demonstrate technologies essential to a power plant in an integrated system and perform integrated testing of the high heat flux and nuclear components required to use fusion energy for practical purposes. TPX will complement ITER by testing advanced modes of steady-state plasma operation that, coupled with the developments in ITER, will lead to an optimized demonstration power plant.

  12. Dynamic reactor modeling with applications to SPR and ZEDNA

    Energy Technology Data Exchange (ETDEWEB)

    Suo-Anttila, Ahti Jorma [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2011-12-01

    A dynamic reactor model has been developed for pulse-type reactor applications. The model predicts reactor power, axial and radial fuel expansion, prompt and delayed neutron population, and prompt and delayed gamma population. All model predictions are made as a function of time. The model includes the reactivity effect of fuel expansion on a dynamic timescale as a feedback mechanism for reactor power. All inputs to the model are calculated from first principles, either directly by solving systems of equations, or indirectly from Monte Carlo N-Particle Transport Code (MCNP) derived results. The model does not include any empirical parameters that can be adjusted to match experimental data. Comparisons of model predictions to actual Sandia Pulse Reactor SPR-III pulses show very good agreement for a full range of pulse magnitudes. The model is also applied to Z-pinch externally driven neutron assembly (ZEDNA) type reactor designs to model both normal and off-normal ZEDNA operations.

  13. Technical specifications, Hanford production reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gilbert, W.D. [comp.

    1962-06-25

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

  14. Reactor Materials Research

    Energy Technology Data Exchange (ETDEWEB)

    Van Walle, E

    2001-04-01

    The activities of the Reactor Materials Research Department of the Belgian Nuclear Research Centre SCK-CEN in 2000 are summarised. The programmes within the department are focussed on studies concerning (1) fusion, in particular mechanical testing; (2) Irradiation Assisted Stress Corrosion Cracking (IASCC); (3) nuclear fuel; and (4) Reactor Pressure Vessel Steel (RPVS)

  15. Space Nuclear Reactor Engineering

    Energy Technology Data Exchange (ETDEWEB)

    Poston, David Irvin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-03-06

    We needed to find a space reactor concept that could be attractive to NASA for flight and proven with a rapid turnaround, low-cost nuclear test. Heat-pipe-cooled reactors coupled to Stirling engines long identified as the easiest path to near-term, low-cost concept.

  16. REFLECTOR FOR NEUTRONIC REACTORS

    Science.gov (United States)

    Fraas, A.P.

    1963-08-01

    A reflector for nuclear reactors that comprises an assembly of closely packed graphite rods disposed with their major axes substantially perpendicular to the interface between the reactor core and the reflector is described. Each graphite rod is round in transverse cross section at (at least) its interface end and is provided, at that end, with a coaxial, inwardly tapering hole. (AEC)

  17. A porous medium approach for the fluid structure interaction modelling of a water pressurized nuclear reactor core fuel assemblies: simulation and experimentation; Une approche milieu poreux pour la modeisation de l'interaction fluide-structure des assemblages combustibles dans un coeur de reacteur a eau pressurisee: simulation et experimentation

    Energy Technology Data Exchange (ETDEWEB)

    Ricciardi, G.

    2008-10-15

    The designing of a pressurized water reactor core subjected to seismic loading, is a major concern of the nuclear industry. We propose, in this PhD report, to establish the global behaviour equations of the core, in term of a porous medium. Local equations of fluid and structure are space averaged on a control volume, thus we define an equivalent fluid and an equivalent structure, of which unknowns are defined on the whole space. The non-linear fuel assemblies behaviour is modelled by a visco-elastic constitutive law. The fluid-structure coupling is accounted for by a body force, the expression of that force is based on empirical formula of fluid forces acting on a tube subject to an axial flow. The resulting equations are solved using a finite element method. A validation of the model, on three experimental device, is proposed. The first one presents two fuel assemblies subjected to axial flow. One of the two fuel assemblies is deviated from its position of equilibrium and released, while the other is at rest. The second one presents a six assemblies row, immersed in water, placed on a shaking table that can simulate seismic loading. Finally, the last one presents nine fuel assemblies network, arranged in a three by three, subject to an axial flow. The displacement of the central fuel assembly is imposed. The simulations are in agreement with the experiments, the model reproduces the influence of the flow of fluid on the dynamics and coupling of the fuel assemblies. (author)

  18. Reactor vessel lower head integrity

    Energy Technology Data Exchange (ETDEWEB)

    Rubin, A.M.

    1997-02-01

    On March 28, 1979, the Three Mile Island Unit 2 (TMI-2) nuclear power plant underwent a prolonged small break loss-of-coolant accident that resulted in severe damage to the reactor core. Post-accident examinations of the TMI-2 reactor core and lower plenum found that approximately 19,000 kg (19 metric tons) of molten material had relocated onto the lower head of the reactor vessel. Results of the OECD TMI-2 Vessel Investigation Project concluded that a localized hot spot of approximately 1 meter diameter had existed on the lower head. The maximum temperature on the inner surface of the reactor pressure vessel (RPV) in this region reached 1100{degrees}C and remained at that temperature for approximately 30 minutes before cooling occurred. Even under the combined loads of high temperature and high primary system pressure, the TMI-2 RPV did not fail. (i.e. The pressure varied from about 8.5 to 15 MPa during the four-hour period following the relocation of melt to the lower plenum.) Analyses of RPV failure under these conditions, using state-of-the-art computer codes, predicted that the RPV should have failed via local or global creep rupture. However, the vessel did not fail; and it has been hypothesized that rapid cooling of the debris and the vessel wall by water that was present in the lower plenum played an important role in maintaining RPV integrity during the accident. Although the exact mechanism(s) of how such cooling occurs is not known, it has been speculated that cooling in a small gap between the RPV wall and the crust, and/or in cracks within the debris itself, could result in sufficient cooling to maintain RPV integrity. Experimental data are needed to provide the basis to better understand these phenomena and improve models of RPV failure in severe accident codes.

  19. Mirror reactor surface study

    Energy Technology Data Exchange (ETDEWEB)

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

    1976-09-01

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

  20. Nuclear reactor design

    CERN Document Server

    2014-01-01

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

  1. Status of French reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-08-01

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

  2. A performance evaluation of a microchannel reactor for the production of hydrogen from formic acid for electrochemical energy applications

    CSIR Research Space (South Africa)

    Ndlovu, IM

    2017-12-01

    Full Text Available An experimental evaluation of a microchannel reactor was completed to assess the reactor performance for the catalytic decomposition of vaporised formic acid (FA) for H2 production. Initially, X-ray powder diffraction (XRD), elemental mapping using...

  3. Observation of abundant heat production from a reactor device and of isotopic changes in the fuel

    OpenAIRE

    Levi, Giuseppe; Evelyn, Foschi; Bo, Hoistad; Roland, Pettesson; Lars, Tegnér; Hanno, Essén

    2014-01-01

    New results are presented from an extended experimental investigation of anomalous heat production in a special type of reactor tube operating at high temperatures. The reactor, named E-Cat, is charged with a small amount of hydrogen-loaded nickel powder plus some additives, mainly Lithium. The reaction is primarily initiated by heat from resistor coils around the reactor tube. Measurements of the radiated power from the reactor were performed with high-resolution ...

  4. Review of Transient Testing of Fast Reactor Fuels in the Transient REActor Test Facility (TREAT)

    Energy Technology Data Exchange (ETDEWEB)

    Jensen, C.; Wachs, D.; Carmack, J.; Woolstenhulme, N.

    2017-01-01

    The restart of the Transient REActor Test (TREAT) facility provides a unique opportunity to engage the fast reactor fuels community to reinitiate in-pile experimental safety studies. Historically, the TREAT facility played a critical role in characterizing the behavior of both metal and oxide fast reactor fuels under off-normal conditions, irradiating hundreds of fuel pins to support fast reactor fuel development programs. The resulting test data has provided validation for a multitude of fuel performance and severe accident analysis computer codes. This paper will provide a review of the historical database of TREAT experiments including experiment design, instrumentation, test objectives, and salient findings. Additionally, the paper will provide an introduction to the current and future experiment plans of the U.S. transient testing program at TREAT.

  5. Nuclear reactor control column

    Science.gov (United States)

    Bachovchin, Dennis 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 cross-sectional 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.

  6. Reactor power monitoring device

    Energy Technology Data Exchange (ETDEWEB)

    Oda, Naotaka; Igawa, Shinji; Kitazono, Hideaki

    1998-02-13

    The present invention provides a reactor power monitoring device capable of ensuring circumstance resistance, high reliability and high speed transmission even if an APRM is disposed in a reactor building (R/B). Namely, signal processing sections (APRM) for transmitting data to a central control chamber are distributed in the reactor building at an area at the lowest temperature among areas where the temperature control in an emergency state is regulated, and a transmission processing section (APRM-I/F) for transmitting data to the other systems is disposed to the central control chamber. An LPRM signal transmission processing section is constituted such that LPRM signals can be transmitted at a high speed by DMA. Set values relevant to reactor tripping (neutron flux high, thermal output high and sudden reduction of a reactor core flow rate) are stored in the APRM-I/F, and reactor tripping calculation is conducted in the APRM-I/F. With such procedure, a reactor power monitoring device having enhanced control function can be attained. (N.H.)

  7. Dissecting Reactor Antineutrino Flux Calculations

    Science.gov (United States)

    Sonzogni, A. A.; McCutchan, E. A.; Hayes, A. C.

    2017-09-01

    Current predictions for the antineutrino yield and spectra from a nuclear reactor rely on the experimental electron spectra from 235U, 239Pu, 241Pu and a numerical method to convert these aggregate electron spectra into their corresponding antineutrino ones. In the present work we investigate quantitatively some of the basic assumptions and approximations used in the conversion method, studying first the compatibility between two recent approaches for calculating electron and antineutrino spectra. We then explore different possibilities for the disagreement between the measured Daya Bay and the Huber-Mueller antineutrino spectra, including the 238U contribution as well as the effective charge and the allowed shape assumption used in the conversion method. We observe that including a shape correction of about +6 % MeV-1 in conversion calculations can better describe the Daya Bay spectrum. Because of a lack of experimental data, this correction cannot be ruled out, concluding that in order to confirm the existence of the reactor neutrino anomaly, or even quantify it, precisely measured electron spectra for about 50 relevant fission products are needed. With the advent of new rare ion facilities, the measurement of shape factors for these nuclides, for many of which precise beta intensity data from TAGS experiments already exist, would be highly desirable.

  8. Nuclear reactor reflector

    Science.gov (United States)

    Hopkins, R.J.; Land, J.T.; Misvel, M.C.

    1994-06-07

    A nuclear reactor reflector is disclosed that comprises a stack of reflector blocks with vertical water flow passages to cool the reflector. The interface between blocks is opposite support points for reactor fuel rods. Water flows between the reflector and the reactor barrel from passages in a bottom block. The top block contains a flange to limit this flow and the flange has a slot to receive an alignment pin that is welded to the barrel. The pin is held in the slot by two removable shims. Alignment bars extend the length of the stack in slots machined in each block when the stack is assembled. 12 figs.

  9. Fast Breeder Reactor studies

    Energy Technology Data Exchange (ETDEWEB)

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

    1980-07-01

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

  10. NEUTRONIC REACTOR CONTROL

    Science.gov (United States)

    Metcalf, H.E.

    1958-10-14

    Methods of controlling reactors are presented. Specifically, a plurality of neutron absorber members are adjustably disposed in the reactor core at different distances from the center thereof. The absorber members extend into the core from opposite faces thereof and are operated by motive means coupled in a manner to simultaneously withdraw at least one of the absorber members while inserting one of the other absorber members. This feature effects fine control of the neutron reproduction ratio by varying the total volume of the reactor effective in developing the neutronic reaction.

  11. Nuclear reactor reflector

    Science.gov (United States)

    Hopkins, Ronald J.; Land, John T.; Misvel, Michael C.

    1994-01-01

    A nuclear reactor reflector is disclosed that comprises a stack of reflector blocks with vertical water flow passages to cool the reflector. The interface between blocks is opposite support points for reactor fuel rods. Water flows between the reflector and the reactor barrel from passages in a bottom block. The top block contains a flange to limit this flow and the flange has a slot to receive an alignment pin that is welded to the barrel. The pin is held in the slot by two removable shims. Alignment bars extend the length of the stack in slots machined in each block when the stack is assembled.

  12. Microfluidic electrochemical reactors

    Science.gov (United States)

    Nuzzo, Ralph G [Champaign, IL; Mitrovski, Svetlana M [Urbana, IL

    2011-03-22

    A microfluidic electrochemical reactor includes an electrode and one or more microfluidic channels on the electrode, where the microfluidic channels are covered with a membrane containing a gas permeable polymer. The distance between the electrode and the membrane is less than 500 micrometers. The microfluidic electrochemical reactor can provide for increased reaction rates in electrochemical reactions using a gaseous reactant, as compared to conventional electrochemical cells. Microfluidic electrochemical reactors can be incorporated into devices for applications such as fuel cells, electrochemical analysis, microfluidic actuation, pH gradient formation.

  13. Neutron dosimetry at the reactor facility VENUS

    Energy Technology Data Exchange (ETDEWEB)

    Deboodt, P.; Vermeersch, F.; Vanhavere, F.; Minsart, G. [SCK-CEN, Belgian Nuclear Research Centre, Mol (Belgium)

    1997-09-01

    The reactor VENUS is a zero-power research reactor mainly devoted to studies on light water fuels. The need for undertaking a neutron spectrometric and dosimetric study became apparent when locally high neutron dose rates were measured. The spectrometric study is based on two approaches. The first is an experimental one in which the neutron spectrum was measured at three positions around the facility. The second is a theoretical one in which a numerical modelling of the neutron transport at the reactor site was performed in order to determine neutron spectra and fluence rates at different positions around the site. The measured and calculated spectra are interpreted in terms of the responses of different individual and environmental dosemeters. These responses are confronted with the in situ measurements. The impact of the ICRP 60 recommendations on the determined dose rates is also studied. (author).

  14. ANAEROBIC DIGESTION AND THE DENITRIFICATION IN UASB REACTOR

    Directory of Open Access Journals (Sweden)

    José Tavares de Sousa

    2008-01-01

    Full Text Available The environmental conditions in Brazil have been contributing to the development of anaerobic systems in the treatment of wastewaters, especially UASB - Upflow Anaerobic Sludge Blanket reactors. The classic biological process for removal of nutrients uses three reactors - Bardenpho System, therefore, this work intends an alternative system, where the anaerobic digestion and the denitrification happen in the same reactor reducing the number of reactors for two. The experimental system was constituted by two units: first one was a nitrification reactor with 35 L volume and 15 d of sludge age. This system was fed with raw sanitary waste. Second unit was an UASB, with 7.8 L and 6 h of hydraulic detention time, fed with ¾ of effluent nitrification reactor and ¼ of raw sanitary waste. This work had as objective to evaluate the performance of the UASB reactor. In terms of removal efficiency, of bath COD and nitrogen, it was verified that the anaerobic digestion process was not affected. The removal efficiency of organic material expressed in COD was 71%, performance already expected for a reactor of this type. It was also observed that the denitrification process happened; the removal nitrate efficiency was 90%. Therefore, the denitrification process in reactor UASB is viable.

  15. ANAEROBIC DIGESTION AND THE DENITRIFICATION IN UASB REACTOR

    Directory of Open Access Journals (Sweden)

    José Tavares de Sousa

    2008-12-01

    Full Text Available The environmental conditions in Brazil have been contributing to the development of anaerobic systems in the treatment of wastewaters, especially UASB – Upflow Anaerobic Sludge Blanket reactors. The classic biological process for removal of nutrients uses three reactors - Bardenpho System, therefore, this work intends an alternative system, where the anaerobic digestion and the denitrification happen in the same reactor reducing the number of reactors for two. The experimental system was constituted by two units: first one was a nitrification reactor with 35 L volume and 15 d of sludge age. This system was fed with raw sanitary waste. Second unit was an UASB, with 7.8 L and 6 h of hydraulic detention time, fed with ¾ of effluent nitrification reactor and ¼ of raw sanitary waste. This work had as objective to evaluate the performance of the UASB reactor. In terms of removal efficiency, of bath COD and nitrogen, it was verified that the anaerobic digestion process was not affected. The removal efficiency of organic material expressed in COD was 71%, performance already expected for a reactor of this type. It was also observed that the denitrification process happened; the removal nitrate efficiency was 90%. Therefore, the denitrification process in reactor UASB is viable.

  16. New reactor type proposed

    CERN Multimedia

    2003-01-01

    "Russian scientists at the Research Institute of Nuclear Power Engineering in Moscow are hoping to develop a new reactor that will use lead and bismuth as fuel instead of uranium and plutonium" (1/2 page).

  17. Department of Reactor Technology

    DEFF Research Database (Denmark)

    Risø National Laboratory, Roskilde

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

  18. Reactor Safety Research Programs Quarterly Report April- June 1981

    Energy Technology Data Exchange (ETDEWEB)

    Edler, S. K.

    1981-09-01

    This document summarizes the work performed by Pacific Northwest laboratory (PNL} from April1 through June 30, 1981, for the Division of Reactor Safety Research within the U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and postaccident coolability tests for the ESSOR reactor Super Sara Test Program, lspra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory {INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  19. Natural convection reactor

    Energy Technology Data Exchange (ETDEWEB)

    Babcock, D.F.; Bernath, L.; Menegus, R.L.; Ring, H.F.

    1956-05-01

    A previous report described the conceptual design of a plutonium producing reactor that may be characterized as follows: Power output (2000 MW); cooling - (natural convection of light water through the reactor, up through a draft tube to an evaporative cooling pond, then back to the reactor, and fuel (400 to 500 tons of uranium enriched to 1.2% U-235). Because this reactor would be cooled by the natural convection of light water, it is believed that the construction costs would be significantly less than for a Savannah or Hanford type reactor. Such expensive items as water treatment and water pumping facilities would be eliminated entirely. The inventory of 500 tons of slightly enriched uranium, however, is an unattractive feature. It represents not only a large dollar investment but also makes the reactor less attractive for construction during periods of national emergency because of the almost certain scarcity of even slightly enriched uranium at that time. The Atomic Energy Commission asked that the design be reviewed with the objective of reducing the inventory of uranium, The results of this review are given in this report.

  20. Development of essential system technologies for advanced reactor - Development of natural circulation analysis code for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Goon Cherl; Park, Ik Gyu; Kim, Jae Hak; Lee, Sang Min; Kim, Tae Wan [Seoul National University, Seoul (Korea)

    1999-04-01

    The objective of this study is to understand the natural circulation characteristics of integral type reactors and to develope the natural circulation analysis code for integral type reactors. This study is focused on the asymmetric 3-dimensional flow during natural circulation such as 1/4 steam generator section isolation and the inclination of the reactor systems. Natural circulation experiments were done using small-scale facilities of integral reactor SMART (System-Integrated Modular Advanced ReacTor). CFX4 code was used to investigate the flow patterns and thermal mixing phenomena in upper pressure header and downcomer. Differences between normal operation of all steam generators and the 1/4 section isolation conditions were observed and the results were used as the data 1/4 section isolation conditions were observed and the results were used as the data for RETRAN-03/INT code validation. RETRAN-03 code was modified for the development of natural circulation analysis code for integral type reactors, which was development of natural circulation analysis code for integral type reactors, which was named as RETRAN-03/INT. 3-dimensional analysis models for asymmetric flow in integral type reactors were developed using vector momentum equations in RETRAN-03. Analysis results using RETRAN-03/INT were compared with experimental and CFX4 analysis results and showed good agreements. The natural circulation characteristics obtained in this study will provide the important and fundamental design features for the future small and medium integral reactors. (author). 29 refs., 75 figs., 18 tabs.

  1. TCODE: a computer code for analysis of tritium and vacuum systems for tokamak fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Clemmer, R.G.

    1978-08-01

    TCODE can be used for either near-term experimental reactors or for commercial reactors. The code provides options for items that may be included in a commercial reactor such as a divertor, neutral beam heating, and a breeding blanket. The code was used to calculate tritium and vacuum system parameters for the near term reactors ITR, TNS-UP and EPR as well as for some commercial reactor designs, the UWMAK series. A selected sample of the tritium and vacuum parameters for these reactor designs is shown. Also shown are parameters for a hypothetical reactor UWMAK-III M having similar characteristics to UWMAK-III but with a higher fractional burnup (5.0% cf. 0.83%). The impact of the reactor design scenario upon major tritium and vacuum systems is discussed.

  2. Design Considerations for Economically Competitive Sodium Cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hongbin Zhang; Haihua Zhao

    2009-05-01

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

  3. Reactor Safety Planning for Prometheus Project, for Naval Reactors Information

    Energy Technology Data Exchange (ETDEWEB)

    P. Delmolino

    2005-05-06

    The purpose of this letter is to submit to Naval Reactors the initial plan for the Prometheus project Reactor Safety work. The Prometheus project is currently developing plans for cold physics experiments and reactor prototype tests. These tests and facilities may require safety analysis and siting support. In addition to the ground facilities, the flight reactor units will require unique analyses to evaluate the risk to the public from normal operations and credible accident conditions. This letter outlines major safety documents that will be submitted with estimated deliverable dates. Included in this planning is the reactor servicing documentation and shipping analysis that will be submitted to Naval Reactors.

  4. SPRAY CALCINATION REACTOR

    Science.gov (United States)

    Johnson, B.M.

    1963-08-20

    A spray calcination reactor for calcining reprocessin- g waste solutions is described. Coaxial within the outer shell of the reactor is a shorter inner shell having heated walls and with open regions above and below. When the solution is sprayed into the irner shell droplets are entrained by a current of gas that moves downwardly within the inner shell and upwardly between it and the outer shell, and while thus being circulated the droplets are calcined to solids, whlch drop to the bottom without being deposited on the walls. (AEC) H03 H0233412 The average molecular weights of four diallyl phthalate polymer samples extruded from the experimental rheometer were redetermined using the vapor phase osmometer. An amine curing agent is required for obtaining suitable silver- filled epoxy-bonded conductive adhesives. When the curing agent was modified with a 47% polyurethane resin, its effectiveness was hampered. Neither silver nor nickel filler impart a high electrical conductivity to Adiprenebased adhesives. Silver filler was found to perform well in Dow-Corning A-4000 adhesive. Two cascaded hot-wire columns are being used to remove heavy gaseous impurities from methane. This purified gas is being enriched in the concentric tube unit to approximately 20% carbon-13. Studies to count low-level krypton-85 in xenon are continuing. The parameters of the counting technique are being determined. The bismuth isotopes produced in bismuth irradiated for polonium production are being determined. Preliminary data indicate the presence of bismuth207 and bismuth-210m. The light bismuth isotopes are probably produced by (n,xn) reactions bismuth-209. The separation of uranium-234 from plutonium-238 solutions was demonstrated. The bulk of the plutonium is removed by anion exchange, and the remainder is extracted from the uranium by solvent extraction techniques. About 99% of the plutonium can be removed in each thenoyltrifluoroacetone extraction. The viscosity, liquid density, and

  5. Scaleable, High Efficiency Microchannel Sabatier Reactor Project

    Data.gov (United States)

    National Aeronautics and Space Administration — A Microchannel Sabatier Reactor System (MSRS) consisting of cross connected arrays of isothermal or graded temperature reactors is proposed. The reactor array...

  6. Monte Carlo simulations of the SLOWPOKE-2 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tan, A.; Buijs, A., E-mail: tana22@mcmaster.ca, E-mail: buijsa@mcmaster.ca [McMaster University, Hamilton, ON (Canada)

    2015-07-01

    The goal of this project is to study the transient behaviour of the SLOWPOKE-2 reactor using Monte-Carlo simulations. By validating the Monte-Carlo methods in G4-STORK with experimental measurements we hope to extend our understanding of reactor transients as well as further develop our methods to model the transients of the next generation reactor designs. A SLOWPOKE-2 reactor such as the one at RMC is modelled using simulation tools from GEANT4 and data taken from open literature. Simulations in G4-STORK find a neutron flux of order 10{sup 12} n cm{sup -2} s{sup -1} and a control rod worth of (4.9 2.0) mk compared to the experimentally measured worth of 5.45 mk. (author)

  7. Sodium fast reactor safety and licensing research plan. Volume II.

    Energy Technology Data Exchange (ETDEWEB)

    Ludewig, H. (Brokhaven National Laboratory, Upton, NY); Powers, D. A.; Hewson, John C.; LaChance, Jeffrey L.; Wright, A. (Argonne National Laboratory, Argonne, IL); Phillips, J.; Zeyen, R. (Institute for Energy Petten, Saint-Paul-lez-Durance, France); Clement, B. (IRSN/DPAM.SEMIC Bt 702, Saint-Paul-lez-Durance, France); Garner, Frank (Radiation Effects Consulting, Richland, WA); Walters, Leon (Advanced Reactor Concepts, Los Alamos, NM); Wright, Steve; Ott, Larry J. (Oak Ridge National Laboratory, Oak Ridge, TN); Suo-Anttila, Ahti Jorma; Denning, Richard (Ohio State University, Columbus, OH); Ohshima, Hiroyuki (Japan Atomic Energy Agency, Ibaraki, Japan); Ohno, S. (Japan Atomic Energy Agency, Ibaraki, Japan); Miyhara, S. (Japan Atomic Energy Agency, Ibaraki, Japan); Yacout, Abdellatif (Argonne National Laboratory, Argonne, IL); Farmer, M. (Argonne National Laboratory, Argonne, IL); Wade, D. (Argonne National Laboratory, Argonne, IL); Grandy, C. (Argonne National Laboratory, Argonne, IL); Schmidt, R.; Cahalen, J. (Argonne National Laboratory, Argonne, IL); Olivier, Tara Jean; Budnitz, R. (Lawrence Berkeley National Laboratory, Berkeley, CA); Tobita, Yoshiharu (Japan Atomic Energy Agency, Ibaraki, Japan); Serre, Frederic (Centre d' %C3%94etudes nucl%C3%94eaires de Cadarache, Cea, France); Natesan, Ken (Argonne National Laboratory, Argonne, IL); Carbajo, Juan J. (Oak Ridge National Laboratory, Oak Ridge, TN); Jeong, Hae-Yong (Korea Atomic Energy Research Institute, Daejeon, Korea); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Corradini, Michael (University of Wisconsin-Madison, Madison, WI); Thomas, Justin (Argonne National Laboratory, Argonne, IL); Wei, Tom (Argonne National Laboratory, Argonne, IL); Sofu, Tanju (Argonne National Laboratory, Argonne, IL); Flanagan, George F. (Oak Ridge National Laboratory, Oak Ridge, TN); Bari, R. (Brokhaven National Laboratory, Upton, NY); Porter D. (Idaho National Laboratory, Idaho Falls, ID); Lambert, J. (Argonne National Laboratory, Argonne, IL); Hayes, S. (Idaho National Laboratory, Idaho Falls, ID); Sackett, J. (Idaho National Laboratory, Idaho Falls, ID); Denman, Matthew R.

    2012-05-01

    Expert panels comprised of subject matter experts identified at the U.S. National Laboratories (SNL, ANL, INL, ORNL, LBL, and BNL), universities (University of Wisconsin and Ohio State University), international agencies (IRSN, CEA, JAEA, KAERI, and JRC-IE) and private consultation companies (Radiation Effects Consulting) were assembled to perform a gap analysis for sodium fast reactor licensing. Expert-opinion elicitation was performed to qualitatively assess the current state of sodium fast reactor technologies. Five independent gap analyses were performed resulting in the following topical reports: (1) Accident Initiators and Sequences (i.e., Initiators/Sequences Technology Gap Analysis), (2) Sodium Technology Phenomena (i.e., Advanced Burner Reactor Sodium Technology Gap Analysis), (3) Fuels and Materials (i.e., Sodium Fast Reactor Fuels and Materials: Research Needs), (4) Source Term Characterization (i.e., Advanced Sodium Fast Reactor Accident Source Terms: Research Needs), and (5) Computer Codes and Models (i.e., Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety). Volume II of the Sodium Research Plan consolidates the five gap analysis reports produced by each expert panel, wherein the importance of the identified phenomena and necessities of further experimental research and code development were addressed. The findings from these five reports comprised the basis for the analysis in Sodium Fast Reactor Research Plan Volume I.

  8. Integral Design Methodology of Photocatalytic Reactors for Air Pollution Remediation.

    Science.gov (United States)

    Passalía, Claudio; Alfano, Orlando M; Brandi, Rodolfo J

    2017-06-07

    An integral reactor design methodology was developed to address the optimal design of photocatalytic wall reactors to be used in air pollution control. For a target pollutant to be eliminated from an air stream, the proposed methodology is initiated with a mechanistic derived reaction rate. The determination of intrinsic kinetic parameters is associated with the use of a simple geometry laboratory scale reactor, operation under kinetic control and a uniform incident radiation flux, which allows computing the local superficial rate of photon absorption. Thus, a simple model can describe the mass balance and a solution may be obtained. The kinetic parameters may be estimated by the combination of the mathematical model and the experimental results. The validated intrinsic kinetics obtained may be directly used in the scaling-up of any reactor configuration and size. The bench scale reactor may require the use of complex computational software to obtain the fields of velocity, radiation absorption and species concentration. The complete methodology was successfully applied to the elimination of airborne formaldehyde. The kinetic parameters were determined in a flat plate reactor, whilst a bench scale corrugated wall reactor was used to illustrate the scaling-up methodology. In addition, an optimal folding angle of the corrugated reactor was found using computational fluid dynamics tools.

  9. Integral Design Methodology of Photocatalytic Reactors for Air Pollution Remediation

    Directory of Open Access Journals (Sweden)

    Claudio Passalía

    2017-06-01

    Full Text Available An integral reactor design methodology was developed to address the optimal design of photocatalytic wall reactors to be used in air pollution control. For a target pollutant to be eliminated from an air stream, the proposed methodology is initiated with a mechanistic derived reaction rate. The determination of intrinsic kinetic parameters is associated with the use of a simple geometry laboratory scale reactor, operation under kinetic control and a uniform incident radiation flux, which allows computing the local superficial rate of photon absorption. Thus, a simple model can describe the mass balance and a solution may be obtained. The kinetic parameters may be estimated by the combination of the mathematical model and the experimental results. The validated intrinsic kinetics obtained may be directly used in the scaling-up of any reactor configuration and size. The bench scale reactor may require the use of complex computational software to obtain the fields of velocity, radiation absorption and species concentration. The complete methodology was successfully applied to the elimination of airborne formaldehyde. The kinetic parameters were determined in a flat plate reactor, whilst a bench scale corrugated wall reactor was used to illustrate the scaling-up methodology. In addition, an optimal folding angle of the corrugated reactor was found using computational fluid dynamics tools.

  10. Nuclear reactors built, being built, or planned, 1994

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-01

    This document contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1994. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE; from the US Nuclear Regulatory Commission (NRC); from the US reactor manufacturers who are the principal nuclear contractors for foreign reactor locations; from US and foreign embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: a commercial reactor locator map and tables of the characteristic and statistical data that follow; a table of abbreviations; tables of data for reactors operating, being built, or planned; and tables of data for reactors that have been shut down permanently or dismantled. The reactors are subdivided into the following parts: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor is a US company -- working either independently or in cooperation with a foreign company (Part 4). Critical assembly refers to an assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5).

  11. Nuclear reactors built, being built, or planned: 1995

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-08-01

    This report contains unclassified information about facilities built, being built, or planned in the US for domestic use or export as of December 31, 1995. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE; from the US Nuclear Regulatory Commission (NRC); from the US reactor manufacturers who are the principal nuclear contractors for foreign reactor locations; from US and foreign embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: (1) a commercial reactor locator map and tables of the characteristic and statistical data that follow; a table of abbreviations; (2) tables of data for reactors operating, being built, or planned; and (3) tables of data for reactors that have been shut down permanently or dismantled. The reactors are subdivided into the following parts: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor is a US company--working either independently or in cooperation with a foreign company (Part 4). Critical assembly refers to an assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5).

  12. Oxidative coupling of methane using inorganic membrane reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ma, Y.H.; Moser, W.R.; Dixon, A.G. [Worcester Polytechnic Institute, MA (United States)] [and others

    1995-12-31

    The goal of this research is to improve the oxidative coupling of methane in a catalytic inorganic membrane reactor. A specific target is to achieve conversion of methane to C{sub 2} hydrocarbons at very high selectivity and relatively higher yields than in fixed bed reactors by controlling the oxygen supply through the membrane. A membrane reactor has the advantage of precisely controlling the rate of delivery of oxygen to the catalyst. This facility permits balancing the rate of oxidation and reduction of the catalyst. In addition, membrane reactors minimize the concentration of gas phase oxygen thus reducing non selective gas phase reactions, which are believed to be a main route for formation of CO{sub x} products. Such gas phase reactions are a cause for decreased selectivity in oxidative coupling of methane in conventional flow reactors. Membrane reactors could also produce higher product yields by providing better distribution of the reactant gases over the catalyst than the conventional plug flow reactors. Modeling work which aimed at predicting the observed experimental trends in porous membrane reactors was also undertaken in this research program.

  13. Fast reactor technology R and D activities in China

    Energy Technology Data Exchange (ETDEWEB)

    Mi, Xu [China Institute of Atomic Energy, Beijing (China)

    2007-06-15

    The basic research on fast reactor technology was started in the mid-1960's in China. The emphasis was put on fast reactor neutronics, thermohydraulics, sodium technology, materials, fuels, safety, sodium devices and instrumentation. In 1987, the research turned to applied basic research with the conceptual design of a 60 MW experimental fast reactor as a target. The project of the China Experimental Fast Reactor (CEFR) with a thermal power 65 MW was launched in 1993. The R and D of fast reactor technology then carried out to serve a design demonstration connected with the different phases of the conceptual, preliminary and detailed design of the CEFR. Recently, three directions of fast reactor technology R and D activities have been considered, and some research programs have been developed. They are: (1) R and D related to the CEFR, i.e. experiments to be conducted on the CEFR for its safe operation, (2) R and D related to the projects of a prototype and the demonstration of fast reactors, and (3) advanced SFR technology within the framework of the international cooperation of INPRO and GIF.

  14. MERCHANT MARINE SHIP REACTOR

    Science.gov (United States)

    Mumm, J.F.; North, D.C. Jr.; Rock, H.R.; Geston, D.K.

    1961-05-01

    A nuclear reactor is described for use in a merchant marine ship. The reactor is of pressurized light water cooled and moderated design in which three passes of the water through the core in successive regions of low, intermediate, and high heat generation and downflow in a fuel region are made. The foregoing design makes a compact reactor construction with extended core life. The core has an egg-crate lattice containing the fuel elements confined between a lower flow baffle and upper grid plate, with the latter serving also as part of a turn- around manifold from which the entire coolant is distributed into the outer fuel elements for the second pass through the core. The inner fuel elements are cooled in the third pass.

  15. Nuclear Rocket Engine Reactor

    CERN Document Server

    Lanin, Anatoly

    2013-01-01

    The development of a nuclear rocket engine reactor (NRER ) is presented in this book. The working capacity of an active zone NRER under mechanical and thermal load, intensive neutron fluxes, high energy generation (up to 30 MBT/l) in a working medium (hydrogen) at temperatures up to 3100 K is displayed. Design principles and bearing capacity of reactors area discussed on the basis of simulation experiments and test data of a prototype reactor. Property data of dense constructional, porous thermal insulating and fuel materials like carbide and uranium carbide compounds in the temperatures interval 300 - 3000 K are presented. Technological aspects of strength and thermal strength resistance of materials are considered. The design procedure of possible emergency processes in the NRER is developed and risks for their origination are evaluated. Prospects of the NRER development for pilotless space devices and piloted interplanetary ships are viewed.

  16. Nuclear reactor safety device

    Science.gov (United States)

    Hutter, Ernest

    1986-01-01

    A safety device is disclosed for use in a nuclear reactor for axially repositioning a control rod with respect to the reactor core in the event of an upward thermal excursion. Such safety device comprises a laminated helical ribbon configured as a tube-like helical coil having contiguous helical turns with slidably abutting edges. The helical coil is disclosed as a portion of a drive member connected axially to the control rod. The laminated ribbon is formed of outer and inner laminae. The material of the outer lamina has a greater thermal coefficient of expansion than the material of the inner lamina. In the event of an upward thermal excursion, the laminated helical coil curls inwardly to a smaller diameter. Such inward curling causes the total length of the helical coil to increase by a substantial increment, so that the control rod is axially repositioned by a corresponding amount to reduce the power output of the reactor.

  17. Heat dissipating nuclear reactor

    Science.gov (United States)

    Hunsbedt, A.; Lazarus, J.D.

    1985-11-21

    Disclosed is a nuclear reactor containment adapted to retain and cool core debris in the unlikely event of a core meltdown and subsequent breach in the reactor vessel. The reactor vessel is seated in a cavity which has a thick metal sidewall that is integral with a thick metal basemat at the bottom of the cavity. The basemat extends beyond the perimeter of the cavity sidewall. Underneath the basemat is a porous bed with water pipes and steam pipes running into it. Water is introduced into the bed and converted into steam which is vented to the atmosphere. A plurality of metal pilings in the form of H-beams extend from the metal base plate downwardly and outwardly into the earth.

  18. Reactor for exothermic reactions

    Science.gov (United States)

    Smith, Jr., Lawrence A.; Hearn, Dennis; Jones, Jr., Edward M.

    1993-01-01

    A liquid phase process for oligomerization of C.sub.4 and C.sub.5 isoolefins or the etherification thereof with C.sub.1 to C.sub.6 alcohols wherein the reactants are contacted in a reactor with a fixed bed acid cation exchange resin catalyst at an LHSV of 5 to 20, pressure of 0 to 400 psig and temperature of 120.degree. to 300.degree. F. Wherein the improvement is the operation of the reactor at a pressure to maintain the reaction mixture at its boiling point whereby at least a portion but less than all of the reaction mixture is vaporized. By operating at the boiling point and allowing a portion of the reaction mixture to vaporize, the exothermic heat of reaction is dissipated by the formation of more boil up and the temperature in the reactor is controlled.

  19. Substantiation of physical concepts of fast reactors in Russia: experience and prospects

    Energy Technology Data Exchange (ETDEWEB)

    Alekseev, P.N. [Russian Research Center ' Kurchatov Institute' (RRC KI), 1, Kurchatov Sq., Moscow, 123182 (Russian Federation); Vasiliev, B.A. [Experimental Design Bureau of Machine Building (OKBM) 15, Burnakovskiy Pr., N. Novgorod, 603074 (Russian Federation); Kormilitsyn, M.V. [State Scientific Center of Russian Federation - Research Institute of Atomic Reactors (NIIAR) Dimitrovgrad-10, Ulianovsk Reg., 433510 (Russian Federation); Lopatkin, A.V. [N.A. Dollezhal Research and Development Institute of Power Engineering (NIKIET) 2/8, M. Krasnoselskaya Str., Moscow, 107140 (Russian Federation); Seleznev, E.F. [All-Russian Research Institute for Nuclear Power Plant Operation (VNIIAES) 25, Ferganskaya, Moscow, 109507 (Russian Federation); Khomyakov, Yu.S.; Tsybulia, A.M. [State Scientific Center of the Russian Federation - A. I. Leypunsky Institute for Physics and Power Engineering (SSC RF- IPPE) 1, Bondarenko Sq., Obninsk, Kaluga Reg., 249033 (Russian Federation); Tocheny, L.V. [International Science and Technology Center (ISTC) 32-34 Krasnoproletarskaya Ulitsa, Moscow, 127473 (Russian Federation)

    2008-07-01

    The fast reactor concept in Russia has accumulated unique experience, since its advent in the 1950's and up to the present, from the creation of the first experimental installation BR-1, experimental reactors BR-5 and BOR-60, the pilot industrial reactors BN-350 in Kazakhstan and up to the BN-600 at Beloyarsk Atomic Power Station. Investigations on the first experimental installations BR-1 and BR-5/-10 proved the propriety of the idea that it is possible to create nuclear reactors that can produce more nuclear fuel than they consume, i.e. the idea of breeding. The architecture of such reactors was also designed, producing a current leader among fast reactors with sodium coolant and oxide uranium-plutonium fuel. Operational experience of BOR-60, BN-350 and, particularly, BN-600 confirmed the engineering and technical feasibility of the concept of fast reactors, the possibility for its realization both for power production and for certain other purposes as well, such as desalinisation of sea water (BN-350) and for radionuclide production (BN-350, BN-600), and it enabled the development and verification of different models, computer methods and codes. The paper presents a review of experience in the creation of plants with fast reactors, scientific research on these installations, principal results, the current status of experimental data analysis, and prospective directions in the development of fast reactors and the corresponding experimental basis in Russia. (authors)

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

    Science.gov (United States)

    1985-02-01

    implies, this reactor uses gas as the primary coolant . The coolant has a higher exit temperature when leaving the core than the PWR water 6 AFWL-TN-84...nuclear reactors, coolants must be used to ensure material components are not subject to failure due to the temperature exceeding melting points...Reactors that use deu- terium (heavy water) as a coolant can use natural uranium as a fuel. The * Canadian reactor, CANDU , utilizes this concept

  1. Oscillatory flow chemical reactors

    Directory of Open Access Journals (Sweden)

    Slavnić Danijela S.

    2014-01-01

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

  2. Jet stability in the lithium fall reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kang, S.W.

    1978-05-04

    A preliminary analysis has been made of the various hydrodynamic aspects involved in the stability of a liquid-lithium jet in a laser-fusion reactor, which comprises a part of LLL's laser fusion power-generation concept. Various physical factors that may affect the jet breakup are delineated, and some approximate calculations are performed to determine their relative influences. Areas of uncertainty are pointed out, along with plans for experimental verification or further theoretical analysis.

  3. Identification of Chemical Reactor Plant’s Mathematical Model

    Directory of Open Access Journals (Sweden)

    Pyakillya Boris

    2015-01-01

    Full Text Available This work presents a solution of the identification problem of chemical reactor plant’s mathematical model. The main goal is to obtain a mathematical description of a chemical reactor plant from experimental data, which based on plant’s time response measurements. This data consists sequence of measurements for water jacket temperature and information about control input signal, which is used to govern plant’s behavior.

  4. Reactor Dosimetry State of the Art 2008

    Science.gov (United States)

    Voorbraak, Wim; Debarberis, Luigi; D'Hondt, Pierre; Wagemans, Jan

    2009-08-01

    Oral session 1: Retrospective dosimetry. Retrospective dosimetry of VVER 440 reactor pressure vessel at the 3rd unit of Dukovany NPP / M. Marek ... [et al.]. Retrospective dosimetry study at the RPV of NPP Greifswald unit 1 / J. Konheiser ... [et al.]. Test of prototype detector for retrospective neutron dosimetry of reactor internals and vessel / K. Hayashi ... [et al.]. Neutron doses to the concrete vessel and tendons of a magnox reactor using retrospective dosimetry / D. A. Allen ... [et al.]. A retrospective dosimetry feasibility study for Atucha I / J. Wagemans ... [et al.]. Retrospective reactor dosimetry with zirconium alloy samples in a PWR / L. R. Greenwood and J. P. Foster -- Oral session 2: Experimental techniques. Characterizing the Time-dependent components of reactor n/y environments / P. J. Griffin, S. M. Luker and A. J. Suo-Anttila. Measurements of the recoil-ion response of silicon carbide detectors to fast neutrons / F. H. Ruddy, J. G. Seidel and F. Franceschini. Measurement of the neutron spectrum of the HB-4 cold source at the high flux isotope reactor at Oak Ridge National Laboratory / J. L. Robertson and E. B. Iverson. Feasibility of cavity ring-down laser spectroscopy for dose rate monitoring on nuclear reactor / H. Tomita ... [et al.]. Measuring transistor damage factors in a non-stable defect environment / D. B. King ... [et al.]. Neutron-detection based monitoring of void effects in boiling water reactors / J. Loberg ... [et al.] -- Poster session 1: Power reactor surveillance, retrospective dosimetry, benchmarks and inter-comparisons, adjustment methods, experimental techniques, transport calculations. Improved diagnostics for analysis of a reactor pulse radiation environment / S. M. Luker ... [et al.]. Simulation of the response of silicon carbide fast neutron detectors / F. Franceschini, F. H. Ruddy and B. Petrović. NSV A-3: a computer code for least-squares adjustment of neutron spectra and measured dosimeter responses / J. G

  5. Reactor Materials Research

    Energy Technology Data Exchange (ETDEWEB)

    Van Walle, E

    2002-04-01

    The activities of SCK-CEN's Reactor Materials Research Department for 2001 are summarised. The objectives of the department are: (1) to evaluate the integrity and behaviour of structural materials used in nuclear power industry; (2) to conduct research to unravel and understand the parameters that determine the material behaviour under or after irradiation; (3) to contribute to the interpretation, the modelling of the material behaviour and to develop and assess strategies for optimum life management of nuclear power plant components. The programmes within the department are focussed on studies concerning (1) Irradiation Assisted Stress Corrosion Cracking (IASCC); (2) nuclear fuel; and (3) Reactor Pressure Vessel Steel.

  6. Nuclear reactor apparatus

    Science.gov (United States)

    Wade, Elman E.

    1978-01-01

    A lifting, rotating and sealing apparatus for nuclear reactors utilizing rotating plugs above the nuclear reactor core. This apparatus permits rotation of the plugs to provide under the plug refueling of a nuclear core. It also provides a means by which positive top core holddown can be utilized. Both of these operations are accomplished by means of the apparatus lifting the top core holddown structure off the nuclear core while stationary, and maintaining this structure in its elevated position during plug rotation. During both of these operations, the interface between the rotating member and its supporting member is sealingly maintained.

  7. Perspectives on reactor safety

    Energy Technology Data Exchange (ETDEWEB)

    Haskin, F.E. [New Mexico Univ., Albuquerque, NM (United States). Dept. of Chemical and Nuclear Engineering; Camp, A.L. [Sandia National Labs., Albuquerque, NM (United States)

    1994-03-01

    The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor, safety concepts. The course consists of five modules: (1) historical perspective; (2) accident sequences; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course.

  8. Safe reactor power

    Energy Technology Data Exchange (ETDEWEB)

    Menegus, R.L.; Ring, H.F.; Bernath, L.

    1956-05-15

    The upper limit on reactor operating power is established not only by safety considerations during steady-state operation but also by the requirement that during an accident no permanent damage be inflicted upon the reactor or the fuel charge. Two general categories of accidents are recognized; they are the ``nuclear runaway`` and the ``loss of coolant flow`` incidents. In this memorandum an incident of the latter type is analyzed. It is assumed that the safety rods function normally, and a method is defined for establishing the highest operating power that may be permitted if the postulated accident is to do no damage.

  9. High Flux Isotope Reactor (HFIR)

    Data.gov (United States)

    Federal Laboratory Consortium — The HFIR at Oak Ridge National Laboratory is a light-water cooled and moderated reactor that is the United States’ highest flux reactor-based neutron source. HFIR...

  10. Reactor operation environmental information document

    Energy Technology Data Exchange (ETDEWEB)

    Haselow, J.S.; Price, V.; Stephenson, D.E.; Bledsoe, H.W.; Looney, B.B.

    1989-12-01

    The Savannah River Site (SRS) produces nuclear materials, primarily plutonium and tritium, to meet the requirements of the Department of Defense. These products have been formed in nuclear reactors that were built during 1950--1955 at the SRS. K, L, and P reactors are three of five reactors that have been used in the past to produce the nuclear materials. All three of these reactors discontinued operation in 1988. Currently, intense efforts are being extended to prepare these three reactors for restart in a manner that protects human health and the environment. To document that restarting the reactors will have minimal impacts to human health and the environment, a three-volume Reactor Operations Environmental Impact Document has been prepared. The document focuses on the impacts of restarting the K, L, and P reactors on both the SRS and surrounding areas. This volume discusses the geology, seismology, and subsurface hydrology. 195 refs., 101 figs., 16 tabs.

  11. Thermal Hydraulic Tests for Reactor Core Safety

    Energy Technology Data Exchange (ETDEWEB)

    Moon, S. K.; Baek, W. P.; Chun, S. Y. (and others)

    2007-06-15

    The main objectives of the present project are to resolve the current issues of reactor core thermal hydraulics, to develop an advanced measurement and analytical techniques, and to perform reactor core safety verification tests. 6x6 reflood experiments, various heat transfer experiments using Freon, and experiments on the spacer grids effects on the post-dryout are carried out using spacer grids developed in Korea in order to resolve the current issues of the reactor core thermal hydraulics. In order to develop a reflood heat transfer model, the detailed reflood phenomena are visualized and measured using round tube and 2x2 rod bundle. A detailed turbulent mixing phenomenon for subchannels is measured using advanced measurement techniques such as LDV and PIV. MARS and MATRA codes developed in Korea are assessed, verified and improved using the obtained experimental data. Finally, a systematic quality assurance program and experimental data generation system has been constructed in order to increase the reliability of the experimental data.

  12. Reactor Safety Research Programs Quarterly Report October - December 1981

    Energy Technology Data Exchange (ETDEWEB)

    Edler, S. K.

    1982-03-01

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

  13. Reactor Safety Research Programs Quarterly Report July - September 1981

    Energy Technology Data Exchange (ETDEWEB)

    Edler, S. K.

    1982-01-01

    This document summarizes the work performed by Pacific Northwest laboratory (PNL) from July 1 through September 30, 1981, for the Division of Accident Evaluation, U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR} steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and postaccident coolability tests for the ESSOR reactor Super Sara Test Program, lspra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL), Idaho Falls, Idaho. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  14. Study of the uniform corrosion of an aluminium alloy used for the fuel cladding of the Jules Horowitz experimental reactor; Etude de la corrosion uniforme d'un alliage d'aluminium utilise comme gainage du combustible nucleaire du reacteur experimental Jules Horowitz

    Energy Technology Data Exchange (ETDEWEB)

    Wintergerst, M. [CEA Saclay, Dept. des Materiaux pour le Nucleaire (DEN/DANS/DMN/SEMI), 91 - Gif-sur-Yvette (France)

    2008-07-01

    For the Jules Horowitz new material testing reactor, an aluminium base alloy, AlFeNi, will be used for the cladding of the fuel plates. Taking into account the thermal properties of the alloy and of its oxide, the corrosion of the fuel cans presents many problems. The aim of this thesis is to provide a growing kinetic of the oxide layer at the surface of the AlFeNi fuel can in order to predict the life time of fuel element. Thus the mechanism of degradation of the cladding will be describe in order to integrate the different parameters of the operating reactor. (A.L.B.)

  15. Reactor operation safety information document

    Energy Technology Data Exchange (ETDEWEB)

    1990-01-01

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

  16. Fast reactor programme in India

    Indian Academy of Sciences (India)

    2015-09-04

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

  17. Engineering reactors for catalytic reactions

    Indian Academy of Sciences (India)

    Extensive studies have been conducted to establish sound basis for design and engineering of reactors for practising such catalytic reactions and for realizing improvements in reactor performance. In this article, application of recent (and not so recent) developments in engineering reactors for catalytic reactions is ...

  18. REACTOR FUEL ELEMENTS TESTING CONTAINER

    Science.gov (United States)

    Whitham, G.K.; Smith, R.R.

    1963-01-15

    This patent shows a method for detecting leaks in jacketed fuel elements. The element is placed in a sealed tank within a nuclear reactor, and, while the reactor operates, the element is sparged with gas. The gas is then led outside the reactor and monitored for radioactive Xe or Kr. (AEC)

  19. Fast reactor programme in India

    Indian Academy of Sciences (India)

    2015-09-04

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

  20. Nuclear Reactors and Technology

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-01-01

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

  1. Nuclear reactor building

    Science.gov (United States)

    Gou, P.F.; Townsend, H.E.; Barbanti, G.

    1994-04-05

    A reactor building for enclosing a nuclear reactor includes a containment vessel having a wetwell disposed therein. The wetwell includes inner and outer walls, a floor, and a roof defining a wetwell pool and a suppression chamber disposed there above. The wetwell and containment vessel define a drywell surrounding the reactor. A plurality of vents are disposed in the wetwell pool in flow communication with the drywell for channeling into the wetwell pool steam released in the drywell from the reactor during a LOCA for example, for condensing the steam. A shell is disposed inside the wetwell and extends into the wetwell pool to define a dry gap devoid of wetwell water and disposed in flow communication with the suppression chamber. In a preferred embodiment, the wetwell roof is in the form of a slab disposed on spaced apart support beams which define there between an auxiliary chamber. The dry gap, and additionally the auxiliary chamber, provide increased volume to the suppression chamber for improving pressure margin. 4 figures.

  2. JACKETED REACTOR FUEL ELEMENT

    Science.gov (United States)

    Smith, K.F.; Van Thyne, R.J.

    1958-12-01

    A fuel element is described for fast reactors comprised of a core of uranium metal containing material and a jacket around the core, the jacket consisting of from 2.5 to 15 percent of titanium, from 1 to 5 percent of niobium, and from 80 to 96.5 percent of vanadium.

  3. Nuclear rocket engine reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lanin, Anatoly

    2013-07-01

    Covers a new technology of nuclear reactors and the related materials aspects. Integrates physics, materials science and engineering Serves as a basic book for nuclear engineers and nuclear physicists. The development of a nuclear rocket engine reactor (NRER) is presented in this book. The working capacity of an active zone NRER under mechanical and thermal load, intensive neutron fluxes, high energy generation (up to 30 MBT/l) in a working medium (hydrogen) at temperatures up to 3100 K is displayed. Design principles and bearing capacity of reactors area discussed on the basis of simulation experiments and test data of a prototype reactor. Property data of dense constructional, porous thermal insulating and fuel materials like carbide and uranium carbide compounds in the temperatures interval 300 - 3000 K are presented. Technological aspects of strength and thermal strength resistance of materials are considered. The design procedure of possible emergency processes in the NRER is developed and risks for their origination are evaluated. Prospects of the NRER development for pilotless space devices and piloted interplanetary ships are viewed.

  4. WATER BOILER REACTOR

    Science.gov (United States)

    King, L.D.P.

    1960-11-22

    As its name implies, this reactor utilizes an aqueous solution of a fissionable element salt, and is also conventional in that it contains a heat exchanger cooling coil immersed in the fuel. Its novelty lies in the utilization of a cylindrical reactor vessel to provide a critical region having a large and constant interface with a supernatant vapor region, and the use of a hollow sleeve coolant member suspended from the cover assembly in coaxial relation with the reactor vessel. Cool water is circulated inside this hollow coolant member, and a gap between its outer wall and the reactor vessel is used to carry off radiolytic gases for recombination in an external catalyst chamber. The central passage of the coolant member defines a reflux condenser passage into which the externally recombined gases are returned and condensed. The large and constant interface between fuel solution and vapor region prevents the formation of large bubbles and minimizes the amount of fuel salt carried off by water vapor, thus making possible higher flux densities, specific powers and power densities.

  5. Fossil fuel furnace reactor

    Science.gov (United States)

    Parkinson, William J.

    1987-01-01

    A fossil fuel furnace reactor is provided for simulating a continuous processing plant with a batch reactor. An internal reaction vessel contains a batch of shale oil, with the vessel having a relatively thin wall thickness for a heat transfer rate effective to simulate a process temperature history in the selected continuous processing plant. A heater jacket is disposed about the reactor vessel and defines a number of independent controllable temperature zones axially spaced along the reaction vessel. Each temperature zone can be energized to simulate a time-temperature history of process material through the continuous plant. A pressure vessel contains both the heater jacket and the reaction vessel at an operating pressure functionally selected to simulate the continuous processing plant. The process yield from the oil shale may be used as feedback information to software simulating operation of the continuous plant to provide operating parameters, i.e., temperature profiles, ambient atmosphere, operating pressure, material feed rates, etc., for simulation in the batch reactor.

  6. NEUTRONIC REACTOR SHIELD

    Science.gov (United States)

    Fermi, E.; Zinn, W.H.

    1957-09-24

    The reactor radiation shield material is comprised of alternate layers of iron-containing material and compressed cellulosic material, such as masonite. The shielding material may be prefabricated in the form of blocks, which can be stacked together in ary desired fashion to form an effective shield.

  7. Neutronic Reactor Structure

    Science.gov (United States)

    Vernon, H. C.; Weinberg, A. M.

    1961-05-30

    The neutronic reactor is comprised of a core consisting of natural uranium and heavy water with a K-factor greater than unity. The core is surrounded by a reflector consisting of natural uranium and ordinary water with a Kfactor less than unity. (AEC)

  8. NEUTRONIC REACTOR CONTROL ELEMENT

    Science.gov (United States)

    Beaver, R.J.; Leitten, C.F. Jr.

    1962-04-17

    A boron-10 containing reactor control element wherein the boron-10 is dispersed in a matrix material is describeri. The concentration of boron-10 in the matrix varies transversely across the element from a minimum at the surface to a maximum at the center of the element, prior to exposure to neutrons. (AEC)

  9. Thermal Reactor Safety

    Energy Technology Data Exchange (ETDEWEB)

    1980-06-01

    Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods.

  10. Research on the reactor physics and reactor safety of VVER reactors. AER Symposium 2016

    Energy Technology Data Exchange (ETDEWEB)

    Kliem, S.

    2017-09-15

    The selected paperscan be attributed to the following main subjects: Reactor start-up tests and use of corresponding data for code validation, code development and application, approaches for safety analyses, closure of nuclear fuel cycle, prospective reactor concepts.

  11. Sequencing Bacth Reactors; Reactores biologicos secuenciados

    Energy Technology Data Exchange (ETDEWEB)

    Nolasco, D.; Manoharan, M.

    1999-06-01

    The application of sequencing batch reactors (SBR) for wastewater treatment is becoming increasingly popular. However, published information on process performance and construction costs for SBRs is scarce. For this reason. Environment Canada, the Ontario Ministry of the Environment (MOE), and the Water Environment Association of Ontario (WEAO) decided to sponsor a program to evaluate the performance of 75 municipal SBRs in Canada and the United States. Effluent quality, construction costs, and design and operating problems were investigated. Areas for optimization found as a result of this investigation were classified an prioritized based on their impact on operational costs, treatment capacity, effluent quality, and frequency of occurrence. A list of recommendations for process optimization was prepared. A construction cost comparison between activated sludge systems of continuous flow and SBRs was prepared. (Author) 12 refs.

  12. Fischer-Tropsch Slurry Reactor modeling

    Energy Technology Data Exchange (ETDEWEB)

    Soong, Y.; Gamwo, I.K.; Harke, F.W. [Pittsburgh Energy Technology Center, PA (United States)] [and others

    1995-12-31

    This paper reports experimental and theoretical results on hydrodynamic studies. The experiments were conducted in a hot-pressurized Slurry-Bubble Column Reactor (SBCR). It includes experimental results of Drakeol-10 oil/nitrogen/glass beads hydrodynamic study and the development of an ultrasonic technique for measuring solids concentration. A model to describe the flow behavior in reactors was developed. The hydrodynamic properties in a 10.16 cm diameter bubble column with a perforated-plate gas distributor were studied at pressures ranging from 0.1 to 1.36 MPa, and at temperatures from 20 to 200{degrees}C, using a dual hot-wire probe with nitrogen, glass beads, and Drakeol-10 oil as the gas, solid, and liquid phase, respectively. It was found that the addition of 20 oil wt% glass beads in the system has a slight effect on the average gas holdup and bubble size. A well-posed three-dimensional model for bed dynamics was developed from an ill-posed model. The new model has computed solid holdup distributions consistent with experimental observations with no artificial {open_quotes}fountain{close_quotes} as predicted by the earlier model. The model can be applied to a variety of multiphase flows of practical interest. An ultrasonic technique is being developed to measure solids concentration in a three-phase slurry reactor. Preliminary measurements have been made on slurries consisting of molten paraffin wax, glass beads, and nitrogen bubbles at 180 {degrees}C and 0.1 MPa. The data show that both the sound speed and attenuation are well-defined functions of both the solid and gas concentrations in the slurries. The results suggest possibilities to directly measure solids concentration during the operation of an autoclave reactor containing molten wax.

  13. Sandia Pulsed Reactor Facility (SPRF) calculator-assisted pulse analysis and display system

    Energy Technology Data Exchange (ETDEWEB)

    Estes, B.F.; Berry, D.T.

    1980-02-01

    Two solid-metal fast burst type reactors (SPR II and SPR III) are operated at the Sandia Pulsed Reactor Facility. Since startup of the reactors, oscilloscope traces have been used to record (by camera) the pulse (power) shape while log N systems have measured initial reactor period. Virtually no other pulse information is available. A decision was made to build a system that could collect the basic input data available from the reactor - fission chambers, photodiodes, and thermocouples - condition the signals and output the various parameters such as power, energy, temperature, period and lifetime on hard copy that would provide a record for operations personnel as well as the experimenter. Because the reactors operate in short time frames - pulse operation - it is convenient to utilize the classical Nordheim-Fuchs approximation of the diffusion equation to describe reactor behavior. This report describes the work performed to date in developing the calculator system and analytical models for computing the desired parameters.

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

    Energy Technology Data Exchange (ETDEWEB)

    Saha, P., E-mail: pradip.saha@ge.com [GE Hitachi Nuclear Energy, Wilmington, NC (United States); Aksan, N. [GRNSPG Group, University of Pisa (Italy); Andersen, J. [GE Hitachi Nuclear Energy, Wilmington, NC (United States); Yan, J. [Westinghouse Electric Co., Columbia, SC (United States); Simoneau, J.P. [AREVA, Lyon (France); Leung, L. [Atomic Energy of Canada Ltd., Chalk River, Ontario (Canada); Bertrand, F. [CEA, DEN, DER, F-13108 Saint-Paul-Lez-Durance (France); Aoto, K.; Kamide, H. [Japan Atomic Energy Agency, Chiyoda-ku, Tokyo (Japan)

    2013-11-15

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

  15. CER. Research reactors in France

    Energy Technology Data Exchange (ETDEWEB)

    Estrade, Jerome [CEA, DEN, DER, Saint-Paul-lez-Durance (France). Jules Horowitz Reactor (JHR)

    2012-10-15

    Networking and the establishment of coalitions between research reactors are important to guarantee a high technical quality of the facility, to assure well educated and trained personnel, to harmonize the codes of standards and the know-ledge of the personnel as well as to enhance research reactor utilization. In addition to the European co-operation, country-specific working groups have been established for many years, such as the French research reactor Club d'Exploitants des Reacteurs (CER). It is the association of French research reactors representing all types of research reactors from zero power up to high flux reactors. CER was founded in 1990 and today a number of 14 research reactors meet twice a year for an exchange of experience. (orig.)

  16. Methanogenesis in Thermophilic Biogas Reactors

    DEFF Research Database (Denmark)

    Ahring, Birgitte Kiær

    1995-01-01

    Methanogenesis in thermophilic biogas reactors fed with different wastes is examined. The specific methanogenic activity with acetate or hydrogen as substrate reflected the organic loading of the specific reactor examined. Increasing the loading of thermophilic reactors stabilized the process...... as indicated by a lower concentration of volatile fatty acids in the effluent from the reactors. The specific methanogenic activity in a thermophilic pilot-plant biogas reactor fed with a mixture of cow and pig manure reflected the stability of the reactor. The numbers of methanogens counted by the most...... against Methanothrix soehngenii or Methanothrix CALS-I in any of the thermophilic biogas reactors examined. Studies using 2-14C-labeled acetate showed that at high concentrations (more than approx. 1 mM) acetate was metabolized via the aceticlastic pathway, transforming the methyl-group of acetate...

  17. Nuclear reactor construction with bottom supported reactor vessel

    Science.gov (United States)

    Sharbaugh, John E.

    1987-01-01

    An improved liquid metal nuclear reactor construction has a reactor core and a generally cylindrical reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core within the pool. The reactor vessel has an open top end, a closed flat bottom end wall and a continuous cylindrical closed side wall interconnecting the top end and bottom end wall. The reactor also has a generally cylindrical concrete containment structure surrounding the reactor vessel and being formed by a cylindrical side wall spaced outwardly from the reactor vessel side wall and a flat base mat spaced below the reactor vessel bottom end wall. A central support pedestal is anchored to the containment structure base mat and extends upwardly therefrom to the reactor vessel and upwardly therefrom to the reactor core so as to support the bottom end wall of the reactor vessel and the lower end of the reactor core in spaced apart relationship above the containment structure base mat. Also, an annular reinforced support structure is disposed in the reactor vessel on the bottom end wall thereof and extends about the lower end of the core so as to support the periphery thereof. In addition, an annular support ring having a plurality of inward radially extending linear members is disposed between the containment structure base mat and the bottom end of the reactor vessel wall and is connected to and supports the reactor vessel at its bottom end on the containment structure base mat so as to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event. The reactor construction also includes a bed of insulating material in sand-like granular form, preferably being high density magnesium oxide particles, disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall on the containment

  18. Reactor core stability monitoring method for BWR type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kanemoto, Shigeru; Ebata, Shigeo.

    1992-09-01

    In an operation for a BWR type reactor, reactor power is usually increased or decreased by controlling both of control rods and reactor core flow rate. Under a certain condition, the reactor core is made unstable by the coupling of nuclear and thermohydrodynamic characteristics in the reactor. Therefore, the reactor power and the reactor core flow rate are changed within a range predetermined by a design calculation. However, if reactor core stability can be always measured and monitored, it is useful for safe operation, as well as an existent operation range can be extended to enable more effective operation. That is, autoregressive a coefficient is determined successively on real time based on fluctuation components of neutron flux signals. Based on the result, an amplification ratio, as a typical measure of the reactor core stability, is determined on a real time. A time constant of the successive calculation for the autoregressive coefficient can be made variable by the amplification ratio. Then, the amplification ratio is estimated at a constant accuracy. With such procedures, the reactor core stability can be monitored successively in an ON-line manner at a high accuracy, thereby enabling to improve the operation performance. (I.S.).

  19. Passive modular gas safety system for a reactor

    Energy Technology Data Exchange (ETDEWEB)

    Abalin, S.S.; Isaev, I.F.; Kulakov, A.A.; Sivokon, V.P.; Udovenko, A.N.; Ionaitis, R.R.

    1994-01-01

    Reactor safety systems have developed gradually. Today in particular, auxiliary systems are being developed which are based on nontraditional operational concepts, by using gaseous neutron absorbers. The Scientific-Research and Design Institute of Power Technology (NIKIET) and the Institute of Nuclear Reactors, Kurchatov Institute Reactor Science Center (RNTs), have done preliminary development and experimental verification of separate elements of this system, in which helium is used as the absorber. This article presents a rapid passive safety system based on gaseous absorber, which is made as autonomous modules as the final stage of reactor safety. Its effectiveness is discussed by using an RBMK reactor as an example. As opposed to traditional active, systems, it does not require a functioning power supply and information signals from outside the reactors system, which makes it stable against unsanctioned actions by personnel, the influence of other systems, and also outside actions (sabotage and natural calamities which could destroy the the nuclear power plant structure). Because the gas safety system can operate instantaneously (0.1-0.3 sec), in principle, it can shut down the reactor even with fast-neutron runaway, where traditional safety systems are ineffective.

  20. Advanced burner test reactor preconceptual design report.

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-12-16

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

  1. Creation of a neutrino laboratory for search for sterile neutrino at SM-3 reactor

    OpenAIRE

    Serebrov, A. P.; Ivochkin, V. G.; Samoylov, R. M.; Fomin, A. K.; Zinoviev, V. G.; Neustroev, P. V.; Golovtsov, V. L.; Gruzinsky, N. V.; Solovey, V. A.; Cherniy, A. V.; Zherebtsov, O. M.; Martemyanov, V. P.; Zinoev, V. G.; Tarasenkov, V. G.; Aleshin, V. I.

    2015-01-01

    In connection with the question of possible existence of sterile neutrino the laboratory on the basis of SM-3 reactor was created to search for oscillations of reactor antineutrino. A prototype of a neutrino detector with scintillator volume of 400 l can be moved at the distance of 6-11 m from the reactor core. The measurements of background conditions have been made. It is shown that the main experimental problem is associated with cosmic radiation background. Test measurements of dependence...

  2. Concept of the power-reactor-pumped laser for technology applications

    Science.gov (United States)

    Gulevich, Andrey V.; Dyachenko, Peter P.; Kononov, Victor N.; Kukharchuk, Oleg F.; Zrodnikov, Anatoly V.

    1998-09-01

    Conception of a high-power pulsed reactor-pumped laser system (RPLS) based on new physical principles (direct nuclear-to- optical energy conversion) for the technology and space application is discussed. The development of an energy model of RPLS consisting of the ignition two-core fast-burst reactor reactor module and a thermal subcritical laser module filled with an Ar-Xe laser active medium is reported. Some of the experimental results are also presented.

  3. Reactor Safety Research Programs Quarterly Report October - December 1980

    Energy Technology Data Exchange (ETDEWEB)

    Edler, S K

    1981-04-01

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from October 1 through December 31, 1980, for the Division of Reactor Safety Research within the U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining structural graphite strength, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NOE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation and postaccident coolability tests for the ESSOR Test Reactor Program, Ispra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  4. Reactor Safety Research Programs Quarterly Report July- September 1980

    Energy Technology Data Exchange (ETDEWEB)

    Edler, S. K.

    1980-12-01

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from April 1 through June 30, 1980, for the Division of Reactor Safety Research within the Nuclear Regulatory Commission {NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining structural graphite strength, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation and postaccident coolability tests for the ESSOR Test Reactor Program, Ispra, Italy; blowdown and reflood tests in the test facility at Cadarache, France; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  5. Reactor Safety Research Programs Quarterly Report April -June 1980

    Energy Technology Data Exchange (ETDEWEB)

    Edler, S. K.

    1980-11-01

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from April 1 through June 30, 1980, for the Division of Reactor Safety Research within the Nuclear Regulatory Commission {NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining structural graphite strength, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation and postaccident coolability tests for the ESSOR Test Reactor Program, Ispra, Italy; blowdown and reflood tests in the test facility at Cadarache, France; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  6. Compact fusion reactors

    CERN Multimedia

    CERN. Geneva

    2015-01-01

    Fusion research is currently to a large extent focused on tokamak (ITER) and inertial confinement (NIF) research. In addition to these large international or national efforts there are private companies performing fusion research using much smaller devices than ITER or NIF. The attempt to achieve fusion energy production through relatively small and compact devices compared to tokamaks decreases the costs and building time of the reactors and this has allowed some private companies to enter the field, like EMC2, General Fusion, Helion Energy, Lawrenceville Plasma Physics and Lockheed Martin. Some of these companies are trying to demonstrate net energy production within the next few years. If they are successful their next step is to attempt to commercialize their technology. In this presentation an overview of compact fusion reactor concepts is given.

  7. HOMOGENEOUS NUCLEAR REACTOR

    Science.gov (United States)

    Hammond, R.P.; Busey, H.M.

    1959-02-17

    Nuclear reactors of the homogeneous liquid fuel type are discussed. The reactor is comprised of an elongated closed vessel, vertically oriented, having a critical region at the bottom, a lower chimney structure extending from the critical region vertically upwardly and surrounded by heat exchanger coils, to a baffle region above which is located an upper chimney structure containing a catalyst functioning to recombine radiolyticallydissociated moderator gages. In operation the liquid fuel circulates solely by convection from the critical region upwardly through the lower chimney and then downwardly through the heat exchanger to return to the critical region. The gases formed by radiolytic- dissociation of the moderator are carried upwardly with the circulating liquid fuel and past the baffle into the region of the upper chimney where they are recombined by the catalyst and condensed, thence returning through the heat exchanger to the critical region.

  8. AIR COOLED NEUTRONIC REACTOR

    Science.gov (United States)

    Fermi, E.; Szilard, L.

    1958-05-27

    A nuclear reactor of the air-cooled, graphite moderated type is described. The active core consists of a cubicle mass of graphite, approximately 25 feet in each dimension, having horizontal channels of square cross section extending between two of the opposite faces, a plurality of cylindrical uranium slugs disposed in end to end abutting relationship within said channels providing a space in the channels through which air may be circulated, and a cadmium control rod extending within a channel provided in the moderator. Suitable shielding is provlded around the core, as are also provided a fuel element loading and discharge means, and a means to circulate air through the coolant channels through the fuel charels to cool the reactor.

  9. NEUTRONIC REACTOR CORE INSTRUMENT

    Science.gov (United States)

    Mims, L.S.

    1961-08-22

    A multi-purpose instrument for measuring neutron flux, coolant flow rate, and coolant temperature in a nuclear reactor is described. The device consists essentially of a hollow thimble containing a heat conducting element protruding from the inner wall, the element containing on its innermost end an amount of fissionsble materinl to function as a heat source when subjected to neutron flux irradiation. Thermocouple type temperature sensing means are placed on the heat conducting element adjacent the fissionable material and at a point spaced therefrom, and at a point on the thimble which is in contact with the coolant fluid. The temperature differentials measured between the thermocouples are determinative of the neutron flux, coolant flow, and temperature being measured. The device may be utilized as a probe or may be incorporated in a reactor core. (AE C)

  10. A COOLED NEUTRONIC REACTOR

    Science.gov (United States)

    Wigner, E.P.; Creutz, E.C.

    1960-03-15

    A nuclear reactor comprising a pair of graphite blocks separated by an air gap is described. Each of the blocks contains a plurality of channels extending from the gap through the block with a plurality of fuel elements being located in the channels. Means are provided for introducing air into the gap between the graphite blocks and for exhausting the air from the ends of the channels opposite the gap.

  11. Gaseous fuel reactor research

    Science.gov (United States)

    Thom, K.; Schneider, R. T.

    1977-01-01

    The paper reviews studies dealing with the concept of a gaseous fuel reactor and describes the structure and plans of the current NASA research program of experiments on uranium hexafluoride systems and uranium plasma systems. Results of research into the basic properties of uranium plasmas and fissioning gases are reported. The nuclear pumped laser is described, and the main results of experiments with these devices are summarized.

  12. The reactor core TRIGA Mark-III with fuels type 30/20; El nucleo del reactor TRIGA Mark-III con combustible tipo 30/20

    Energy Technology Data Exchange (ETDEWEB)

    Aguilar H, F., E-mail: fortunato.aguilar@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2012-10-15

    This work describes the calculation series carried out with the program MCNP5 in order to define the configuration of the reactor core with fuels 30/20 (fuels with 30% of uranium content in the Or-Zr-H mixture and a nominal enrichment of 20%). To select the configuration of the reactor core more appropriate to the necessities and future uses of the reactor, the following criterions were taken into account: a) the excess in the reactor reactivity, b) the switch out margin and c) to have new irradiation facilities inside the reactor core. Taking into account these criterions is proceeded to know the characteristics of the components that form the reactor core (dimensions, geometry, materials, densities and positions), was elaborated a base model of the reactor core, for the MCNP5 code, with a configuration composed by 85 fuel elements, 4 control bars and the corresponding structural elements. The high reactivity excess obtained with this model, gave the rule to realize other models of the reactor core in which the reactivity excess and the switch out margin were approximate to the values established in the technical specifications of the reactor operation. Several models were realized until finding the satisfactory model; this is composite for 74 fuels, 4 control bars and 6 additional experimental positions inside the reactor core. (Author)

  13. International working group on gas-cooled reactors. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    1981-01-15

    The purpose of the meeting was to provide a forum for exchange of information on safety and licensing aspects for gas-cooled reactors in order to provide comprehensive review of the present status and of directions for future applications and development. Contributions were made concerning the operating experience of the Fort St. Vrain (FSV) HTGR Power Plant in the United States of America, the experimental power station Arbeitsgemeinschaft Versuchsreaktor (AVR) in the Federal Republic of Germany, and the CO/sub 2/-cooled reactors in the United Kingdom such as Hunterson B and Hinkley Point B. The experience gained at each of these reactors has proved the high safety potential of Gas-cooled Reactor Power Plants.

  14. Use of tower reactors for continuous ethanol production

    Directory of Open Access Journals (Sweden)

    M.C. Viegas

    2002-04-01

    Full Text Available The purpose of this work was to develop a continuous fermentation system operating with a tower reactor using some flocculent yeast strains isolated from an industrial process. The strain was an used in the trial of the proposed system, composed of two serial glass tower reactor. The effects of the following variables were studied on the yield and productivity of the system: total reducing sugar (TRS, concentration in feeding, recycle flow in the second reactor, residence time and diameter/height ratio of the reactors. It was observed that the TRS concentration in feeding and residence time is the variables that interfere most with the productivity of the system. Yield was not affected by any of the variables within the range of values studied. All trials were performed according to a factorial experimental design (making up a total of 19 trials and the results were evaluated by response surface.

  15. MODELING OF TUBULAR ELECTROCHEMICAL REACTOR FOR DYE REMOVAL

    Directory of Open Access Journals (Sweden)

    V. VIJAYAKUMAR

    2017-06-01

    Full Text Available The aim of the present investigation is to model a tubular electrochemical reactor for the treatment of synthetic dye wastewater. The tubular reactor was modeled and solved by finite difference method. For the model solution, the column was divided into 11 nodes in the axial direction and the variation in the radial direction has been neglected. An initial dye concentration of 200 mg L-1was taken in the reservoir. The reactor was operated in a batch with recirculation operation. Based on preliminary experiments all parameters have been optimized. The model simulation is compared with the experimental value and it is observed that the model fairly matches well with the experiment. The modeling of tubular electrochemical reactors for dye waste water treatment could be useful in the design and scale up of electrochemical process.

  16. BOILER-SUPERHEATED REACTOR

    Science.gov (United States)

    Heckman, T.P.

    1961-05-01

    A nuclear power reactor of the type in which a liquid moderator-coolant is transformed by nuclear heating into a vapor that may be used to drive a turbo- generator is described. The core of this reactor comprises a plurality of freely suspended tubular fuel elements, called fuel element trains, within which nonboiling pressurized liquid moderator-coolant is preheated and sprayed through orifices in the walls of the trains against the outer walls thereof to be converted into vapor. Passage of the vapor ovcr other unwetted portions of the outside of the fuel elements causes the steam to be superheated. The moderatorcoolant within the fuel elements remains in the liqUid state, and that between the fuel elements remains substantiaily in the vapor state. A unique liquid neutron-absorber control system is used. Advantages expected from the reactor design include reduced fuel element failure, increased stability of operation, direct response to power demand, and circulation of a minimum amount of liquid moderatorcoolant. (A.G.W.)

  17. NUCLEAR REACTOR CORE DESIGN

    Science.gov (United States)

    Mahlmeister, J.E.; Peck, W.S.; Haberer, W.V.; Williams, A.C.

    1960-03-22

    An improved core design for a sodium-cooled, graphitemoderated nuclear reactor is described. The improved reactor core comprises a number of blocks of moderator material, each block being in the shape of a regular prism. A number of channels, extending the length of each block, are disposed around the periphery. When several blocks are placed in contact to form the reactor core, the channels in adjacent blocks correspond with each other to form closed conduits extending the length of the core. Fuel element clusters are disposed in these closed conduits, and liquid coolant is forced through the annulus between the fuel cluster and the inner surface of the conduit. In a preferred embodiment of the invention, the moderator blocks are in the form of hexagonal prisms with longitudinal channels cut into the corners of the hexagon. The main advantage of an "edge-loaded" moderator block is that fewer thermal neutrons are absorbed by the moderator cladding, as compared with a conventional centrally loaded moderator block.

  18. Nuclear reactor sealing system

    Science.gov (United States)

    McEdwards, James A.

    1983-01-01

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

  19. Development of an improved Sabatier reactor

    Energy Technology Data Exchange (ETDEWEB)

    Birbara, P.J.; Sribnik, F.

    1979-01-01

    This paper presents the results of recent experimental and analytical studies of a Sabatier reactor where carbon dioxide and hydrogen in the presence of a catalyst react to form water, methane, and heat. The work undertaken in this program was aimed at simplification of design and control concepts of Sabatier subsystems. To this end, effort was expended to the development of UASC-151G, a highly active, physically durable catalyst composed of ruthenium on alumina. UASC-151G is five times as active as that supplied for the SSP program. The use of this improved catalyst has very significant effects on the Sabatier reaction subsystem design including: (1) lower temperature starting capability, (2) simiplification of active control and instrumentation requirements, (3) simplified reactor design, (4) improved reliability, and (5) high conversion efficiencies using only small amounts of catalyst. Reasonable agreement between test and computer simulation has been obtained for temperature and lean component conversion efficiencies for both steady-state and cyclic operation.

  20. Nuclear research reactors in Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Cota, Anna Paula Leite; Mesquita, Amir Zacarias, E-mail: aplc@cdtn.b, E-mail: amir@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The rising concerns about global warming and energy security have spurred a revival of interest in nuclear energy, giving birth to a 'nuclear power renaissance' in several countries in the world. Particularly in Brazil, in the recent years, the nuclear power renaissance can be seen in the actions that comprise its nuclear program, summarily the increase of the investments in nuclear research institutes and the government target to design and build the Brazilian Multipurpose research Reactor (BMR). In the last 50 years, Brazilian research reactors have been used for training, for producing radioisotopes to meet demands in industry and nuclear medicine, for miscellaneous irradiation services and for academic research. Moreover, the research reactors are used as laboratories to develop technologies in power reactors, which are evaluated today at around 450 worldwide. In this application, those reactors become more viable in relation to power reactors by the lowest cost, by the operation at low temperatures and, furthermore, by lower demand for nuclear fuel. In Brazil, four research reactors were installed: the IEA-R1 and the MB-01 reactors, both at the Instituto de Pesquisas Energeticas Nucleares (IPEN, Sao Paulo); the Argonauta, at the Instituto de Engenharia Nuclear (IEN, Rio de Janeiro) and the IPR-R1 TRIGA reactor, at the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN, Belo Horizonte). The present paper intends to enumerate the characteristics of these reactors, their utilization and current academic research. Therefore, through this paper, we intend to collaborate on the BMR project. (author)

  1. Modeling a Packed Bed Reactor Utilizing the Sabatier Process

    Science.gov (United States)

    Shah, Malay G.; Meier, Anne J.; Hintze, Paul E.

    2017-01-01

    A numerical model is being developed using Python which characterizes the conversion and temperature profiles of a packed bed reactor (PBR) that utilizes the Sabatier process; the reaction produces methane and water from carbon dioxide and hydrogen. While the specific kinetics of the Sabatier reaction on the RuAl2O3 catalyst pellets are unknown, an empirical reaction rate equation1 is used for the overall reaction. As this reaction is highly exothermic, proper thermal control is of the utmost importance to ensure maximum conversion and to avoid reactor runaway. It is therefore necessary to determine what wall temperature profile will ensure safe and efficient operation of the reactor. This wall temperature will be maintained by active thermal controls on the outer surface of the reactor. Two cylindrical PBRs are currently being tested experimentally and will be used for validation of the Python model. They are similar in design except one of them is larger and incorporates a preheat loop by feeding the reactant gas through a pipe along the center of the catalyst bed. The further complexity of adding a preheat pipe to the model to mimic the larger reactor is yet to be implemented and validated; preliminary validation is done using the smaller PBR with no reactant preheating. When mapping experimental values of the wall temperature from the smaller PBR into the Python model, a good approximation of the total conversion and temperature profile has been achieved. A separate CFD model incorporates more complex three-dimensional effects by including the solid catalyst pellets within the domain. The goal is to improve the Python model to the point where the results of other reactor geometry can be reasonably predicted relatively quickly when compared to the much more computationally expensive CFD approach. Once a reactor size is narrowed down using the Python approach, CFD will be used to generate a more thorough prediction of the reactors performance.

  2. Kinetic Parameter Measurements in the MINERVE Reactor

    Science.gov (United States)

    Perret, Grégory; Geslot, Benoit; Gruel, Adrien; Blaise, Patrick; Di-Salvo, Jacques; De Izarra, Grégoire; Jammes, Christian; Hursin, Mathieu; Pautz, Andréas

    2017-01-01

    In the framework of an international collaboration, teams of the PSI and CEA research institutes measure the critical decay constant (α0 = β/A), delayed neutron fraction (β) and generation time (A) of the Minerve reactor using the Feynman-α, Power Spectral Density and Rossi-α neutron noise measurement techniques. These measurements contribute to the experimental database of kinetic parameters used to improve nuclear data files and validate modern methods in Monte Carlo codes. Minerve is a zero-power pool reactor composed of a central experimental test lattice surrounded by a large aluminum buffer and four high-enriched driver regions. Measurements are performed in three slightly subcritical configurations (-2 cents to -30 cents) using two high-efficiency 235U fission chambers in the driver regions. Measurement of α0 and β obtained by the two institutes and with the different techniques are consistent for the configurations envisaged. Slight increases of the β values are observed with the subcriticality level. Best estimate values are obtained with the Cross-Power Spectral Density technique at -2 cents, and are worth: β = 716.9±9.0 pcm, α0 = 79.0±0.6 s-1 and A = 90.7±1.4 μs. The kinetic parameters are predicted with MCNP5-v1.6 and TRIPOLI4.9 and the JEFF-3.1/3.1.1 and ENDF/B-VII.1 nuclear data libraries. The predictions for β and α0 overestimate the experimental results by 3-5% and 10-12%, respectively; that for A underestimate the experimental result by 6-7%. The discrepancies are suspected to come from the driven system nature of Minerve and the location of the detectors in the driver regions, which prevent accounting for the full reactor.

  3. Decay heat measurement on fusion reactor materials and validation of calculation code system

    Energy Technology Data Exchange (ETDEWEB)

    Maekawa, Fujio; Ikeda, Yujiro; Wada, Masayuki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-03-01

    Decay heat rates for 32 fusion reactor relevant materials irradiated with 14-MeV neutrons were measured for the cooling time period between 1 minute and 400 days. With using the experimental data base, validity of decay heat calculation systems for fusion reactors were investigated. (author)

  4. The catalytic combustion of natural gas in a membrane reactor with separate feed of reactants

    NARCIS (Netherlands)

    Neomagus, H.W.J.P.; Saracco, G.; Wessel, H.F.W.; Wessel, H.F.W.; Versteeg, Geert

    2000-01-01

    This paper provides an experimental and modelling analysis of the performance of a membrane reactor with separate feed of reactants for the combustion of methane. In this reactor concept methane and air streams are fed at opposite sides of a Pt/γ-Al2O3-activated porous membrane which hosts their

  5. Fast pyrolysis in a novel wire-mesh reactor: decomposition of pine wood and model compounds

    NARCIS (Netherlands)

    Hoekstra, E.; van Swaaij, Willibrordus Petrus Maria; Kersten, Sascha R.A.; Hogendoorn, Kees

    2012-01-01

    In fast pyrolysis, biomass decomposition processes are followed by vapor phase reactions. Experimental results were obtained in a unique wire-mesh reactor using pine wood, KCl impregnated pine wood and several model compounds (cellulose, xylan, lignin, levoglucosan, glucose). The wire-mesh reactor

  6. Fluoroptic measurements of the local heat transfer coefficient inside the rotating cone reactor

    NARCIS (Netherlands)

    Wagenaar, B.M.; Kuipers, J.A.M.; Versteeg, Geert; van Swaaij, Willibrordus Petrus Maria

    1994-01-01

    The rotating cone reactor is a novel reactor type for rapid thermal processing of solids. This paper focuses on the experimental determination of the gas-to-particle heat transfer coefficient. This quantity has been measured for several particle diameters (average size of 150, 280 and 425 ¿m) and

  7. Identification of Thermal Feedback Parameters for a TRIGA Mark-1 Reactor.

    Science.gov (United States)

    An identification procedure is given for determining the thermal feedback parameters for the TRIGA Mark-1 nuclear reactor at The University of Texas...feedback loops, is developed and then simplified. A brief description is given of the TRIGA reactor and the rest of the experimental equipment. The

  8. Spiral-shaped disinfection reactors

    KAUST Repository

    Ghaffour, Noreddine

    2015-08-20

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

  9. Acceptability of reactors in space

    Energy Technology Data Exchange (ETDEWEB)

    Buden, D.

    1981-04-01

    Reactors are the key to our future expansion into space. However, there has been some confusion in the public as to whether they are a safe and acceptable technology for use in space. The answer to these questions is explored. The US position is that when reactors are the preferred technical choice, that they can be used safely. In fact, it dies not appear that reactors add measurably to the risk associated with the Space Transportation System.

  10. Neutrino Oscillation Studies with Reactors

    CERN Document Server

    Vogel, Petr; Zhang, Chao

    2015-01-01

    Nuclear reactors are one of the most intense, pure, controllable, cost-effective, and well-understood sources of neutrinos. Reactors have played a major role in the study of neutrino oscillations, a phenomenon that indicates that neutrinos have mass and that neutrino flavors are quantum mechanical mixtures. Over the past several decades reactors were used in the discovery of neutrinos, were crucial in solving the solar neutrino puzzle, and allowed the determination of the smallest mixing angle $\\theta_{13}$. In the near future, reactors will help to determine the neutrino mass hierarchy and to solve the puzzling issue of sterile neutrinos.

  11. Modeling of Carbochlorination of Zircon in Fluidized Bed Reactor

    Science.gov (United States)

    Jazini, M. H.; Ghoreishi, S. M.; Dadkhah, A. A.

    2010-02-01

    The chlorination of zircon is an integral part of the overall process for the production of zirconium. A two-phase hydrodynamic-type model was used to simulate the fluidized bed zircon carbochlorination reactor. In the plug-plug (P-P) model, the flow of gas in both dense and bubble phases was considered as a plug flow compared with the plug-mixed (P-M) model in which the flow in dense phase was assumed to be mixed. The zircon conversion obtained by model was compared with experimental measurements for model validation. The results of the P-M model indicated a stronger correlation with experimental data. Using the validated model, the effects of the zircon inlet size distribution, reactor temperature, inlet gas concentration, chlorine conversion velocity, and converted zircon were investigated. The results demonstrated that higher reactor temperature, smaller zircon size, and higher inlet gas velocity and concentration enhanced the chlorination rate.

  12. Background studies for the MINER Coherent Neutrino Scattering reactor experiment

    Energy Technology Data Exchange (ETDEWEB)

    Agnolet, G.; Baker, W. [Department of Physics and Astronomy, and the Mitchell Institute for Fundamental Physics and Astronomy, Texas A& M University, College Station, TX 77843 (United States); Barker, D. [School of Physics & Astronomy, University of Minnesota, Minneapolis, MN 55455 (United States); Beck, R. [Department of Physics and Astronomy, and the Mitchell Institute for Fundamental Physics and Astronomy, Texas A& M University, College Station, TX 77843 (United States); Carroll, T.J.; Cesar, J. [Department of Physics, University of Texas at Austin, Austin, TX 78712 (United States); Cushman, P. [School of Physics & Astronomy, University of Minnesota, Minneapolis, MN 55455 (United States); Dent, J.B. [Department of Physics, University of Louisiana at Lafayette, Lafayette, LA 70504 (United States); De Rijck, S. [Department of Physics, University of Texas at Austin, Austin, TX 78712 (United States); Dutta, B. [Department of Physics and Astronomy, and the Mitchell Institute for Fundamental Physics and Astronomy, Texas A& M University, College Station, TX 77843 (United States); Flanagan, W. [Department of Physics, University of Texas at Austin, Austin, TX 78712 (United States); Fritts, M. [School of Physics & Astronomy, University of Minnesota, Minneapolis, MN 55455 (United States); Gao, Y. [Department of Physics and Astronomy, and the Mitchell Institute for Fundamental Physics and Astronomy, Texas A& M University, College Station, TX 77843 (United States); Department of Physics & Astronomy, Wayne State University, Detroit 48201 (United States); Harris, H.R.; Hays, C.C. [Department of Physics and Astronomy, and the Mitchell Institute for Fundamental Physics and Astronomy, Texas A& M University, College Station, TX 77843 (United States); Iyer, V. [School of Physical Sciences, National Institute of Science Education and Research, Jatni - 752050 (India); and others

    2017-05-01

    The proposed Mitchell Institute Neutrino Experiment at Reactor (MINER) experiment at the Nuclear Science Center at Texas A&M University will search for coherent elastic neutrino-nucleus scattering within close proximity (about 2 m) of a 1 MW TRIGA nuclear reactor core using low threshold, cryogenic germanium and silicon detectors. Given the Standard Model cross section of the scattering process and the proposed experimental proximity to the reactor, as many as 5–20 events/kg/day are expected. We discuss the status of preliminary measurements to characterize the main backgrounds for the proposed experiment. Both in situ measurements at the experimental site and simulations using the MCNP and GEANT4 codes are described. A strategy for monitoring backgrounds during data taking is briefly discussed.

  13. Accelerator based fusion reactor

    Science.gov (United States)

    Liu, Keh-Fei; Chao, Alexander Wu

    2017-08-01

    A feasibility study of fusion reactors based on accelerators is carried out. We consider a novel scheme where a beam from the accelerator hits the target plasma on the resonance of the fusion reaction and establish characteristic criteria for a workable reactor. We consider the reactions d+t\\to n+α,d+{{}3}{{H}\\text{e}}\\to p+α , and p+{{}11}B\\to 3α in this study. The critical temperature of the plasma is determined from overcoming the stopping power of the beam with the fusion energy gain. The needed plasma lifetime is determined from the width of the resonance, the beam velocity and the plasma density. We estimate the critical beam flux by balancing the energy of fusion production against the plasma thermo-energy and the loss due to stopping power for the case of an inert plasma. The product of critical flux and plasma lifetime is independent of plasma density and has a weak dependence on temperature. Even though the critical temperatures for these reactions are lower than those for the thermonuclear reactors, the critical flux is in the range of {{10}22}-{{10}24}~\\text{c}{{\\text{m}}-2}~{{\\text{s}}-1} for the plasma density {ρt}={{10}15}~\\text{c}{{\\text{m}}-3} in the case of an inert plasma. Several approaches to control the growth of the two-stream instability are discussed. We have also considered several scenarios for practical implementation which will require further studies. Finally, we consider the case where the injected beam at the resonance energy maintains the plasma temperature and prolongs its lifetime to reach a steady state. The equations for power balance and particle number conservation are given for this case.

  14. Burnup measurements on spent fuel elements of the RP-10 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Vela Mora, Mariano; Gallardo Padilla, Alberto; Palomino, Jose Luis Castro, E-mail: mvela@ipen.gob.p [Instituto Peruano de Energia Nuclear (IPEN/Peru), Lima (Peru). Grupo de Calculo, Analisis y Seguridad de Reactores; Terremoto, Luis Antonio Albiac, E-mail: laaterre@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    This work describes the measurement, using nondestructive gamma-ray spectroscopy, of the average burnup attained by Material Testing Reactor (MTR) fuel elements irradiated in the RP-10 research reactor. Measurements were performed at the reactor storage pool area using {sup 137}Cs as the only burnup monitor, even for spent fuel elements with cooling times much shorter than two years. The experimental apparatus was previously calibrated in efficiency to obtain absolute average burnup values, which were compared against corresponding ones furnished by reactor physics calculations. The mean deviation between both values amounts to 6%. (author)

  15. Nuclear Reactor Safety; (USA)

    Energy Technology Data Exchange (ETDEWEB)

    Cason, D.L.; Hicks, S.C. (eds.)

    1991-01-01

    This publication announces on an monthly basis the current worldwide information available on all safety-related aspects of reactors, including: accident analysis, safety systems, radiation protection, decommissioning and dismantling, and security measures. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database (EDB) during the past month. Also included are other US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency's Technology Data Exchange, the International Atomic Energy Agency's International Nuclear Information System, or government-to-government agreements.

  16. Iris reactor development

    Energy Technology Data Exchange (ETDEWEB)

    Paramonov, D.V.; Carelli, M.D. [Westinghouse Electric Corp., Baltimore, MD (United States); Miller, K. [BNFL, Inc., (United Kingdom); Lombardi, C.V.; Ricotti, M.E. [Polytechnic of Milan, Polimi (Italy); Todreas, N.E. [Masachussets Institute of Technology, MIT (United States); Greenspan, E. [University of California at Berkeley, UCB (United States); Yamamoto, K. [JAPC Japan Atomic Power Co., Tokyo (Japan); Nagano, A. [Mitsubishi Heavy Industries Ltd., Tokyo (Japan); Ninokata, H. [Tokyo Institut of Technology, TIT (Japan); Robertson, J. [Westinghouse and bechtel (United States); Oriolo, F. [Pisa Univ. (Italy)

    2001-07-01

    The development progress of the IRIS (International Reactor Innovative and Secure) nuclear power system is presented. IRIS is currently being developed by an international consortium of industry, laboratory, university and utility establishments, led by Westinghouse. It is aimed at achieving the four major objectives of the Generation IV nuclear systems, i.e., proliferation resistance, enhanced safety, economic competitiveness and reduced waste. The project first year activities, which are summarized here, were focused on core neutronics, in-vessel configuration, steam generator and containment design, safety approach and economic performance. Details of these studies are provided in parallel papers in these proceedings. (author)

  17. NEUTRONIC REACTOR CONSTRUCTION

    Science.gov (United States)

    Vernon, H.C.; Goett, J.J.

    1958-09-01

    A cover device is described for the fuel element receiving tube of a neutronic reactor of the heterogeneous, water cooled type wherein said tubes are arranged in a moderator with their longitudinal axes vertical. The cover is provided with means to support a rod-type fuel element from the bottom thereof and means to lock the cover in place, the latter being adapted for remote operation. This cover device is easily removable and seals the opening in the upper end of the fuel tube against leakage of coolant.

  18. Reactor coolant pump flywheel

    Science.gov (United States)

    Finegan, John Raymond; Kreke, Francis Joseph; Casamassa, John Joseph

    2013-11-26

    A flywheel for a pump, and in particular a flywheel having a number of high density segments for use in a nuclear reactor coolant pump. The flywheel includes an inner member and an outer member. A number of high density segments are provided between the inner and outer members. The high density segments may be formed from a tungsten based alloy. A preselected gap is provided between each of the number of high density segments. The gap accommodates thermal expansion of each of the number of segments and resists the hoop stress effect/keystoning of the segments.

  19. SODIUM DEUTERIUM REACTOR

    Science.gov (United States)

    Oppenheimer, E.D.; Weisberg, R.A.

    1963-02-26

    This patent relates to a barrier system for a sodium heavy water reactor capable of insuring absolute separation of the metal and water. Relatively cold D/sub 2/O moderator and reflector is contained in a calandria into which is immersed the fuel containing tubes. The fuel elements are cooled by the sodium which flows within the tubes and surrounds the fuel elements. The fuel containing tubes are surrounded by concentric barrier tubes forming annular spaces through which pass inert gases at substantially atmospheric pressure. Header rooms above and below the calandria are provided for supplying and withdrawing the sodium and inert gases in the calandria region. (AEC)

  20. Fast Reactor Fuel Type and Reactor Safety Performance

    Energy Technology Data Exchange (ETDEWEB)

    R. Wigeland; J. Cahalan

    2009-09-01

    Fast Reactor Fuel Type and Reactor Safety Performance R. Wigeland , Idaho National Laboratory J. Cahalan, Argonne National Laboratory The sodium-cooled fast neutron reactor is currently being evaluated for the efficient transmutation of the highly-hazardous, long-lived, transuranic elements that are present in spent nuclear fuel. One of the fundamental choices that will be made is the selection of the fuel type for the fast reactor, whether oxide, metal, carbide, nitride, etc. It is likely that a decision on the fuel type will need to be made before many of the related technologies and facilities can be selected, from fuel fabrication to spent fuel reprocessing. A decision on fuel type should consider all impacts on the fast reactor system, including safety. Past work has demonstrated that the choice of fuel type may have a significant impact on the severity of consequences arising from accidents, especially for severe accidents of low probability. In this paper, the response of sodium-cooled fast reactors is discussed for both oxide and metal fuel types, highlighting the similarities and differences in reactor response and accident consequences. Any fast reactor facility must be designed to be able to successfully prevent, mitigate, or accommodate all consequences of potential events, including accidents. This is typically accomplished by using multiple barriers to the release of radiation, including the cladding on the fuel, the intact primary cooling system, and most visibly the reactor containment building. More recently, this has also included the use of ‘inherent safety’ concepts to reduce or eliminate the potential for serious damage in some cases. Past experience with oxide and metal fuel has demonstrated that both fuel types are suitable for use as fuel in a sodium-cooled fast reactor. However, safety analyses for these two fuel types have also shown that there can be substantial differences in accident consequences due to the neutronic and

  1. Reactor Physics Analysis Models for a CANDU Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok

    2007-10-15

    Canada deuterium uranium (CANDU) reactor physics analysis is typically performed in three steps. At first, macroscopic cross-sections of the reference lattice is produced by modeling the reference fuel channel. Secondly macroscopic cross-sections of reactivity devices in the reactor are generated. The macroscopic cross-sections of a reactivity device are calculated as incremental cross-sections by subtracting macroscopic cross-sections of a three-dimensional lattice without reactivity device from those of a three-dimensional lattice with a reactivity device. Using the macroscopic cross-sections of the reference lattice and incremental cross-sections of the reactivity devices, reactor physics calculations are performed. This report summarizes input data of typical CANDU reactor physics codes, which can be utilized for the future CANDU reactor physics analysis.

  2. Important problems of future thermonuclear reactors*

    Directory of Open Access Journals (Sweden)

    Sadowski Marek J.

    2015-06-01

    Full Text Available This paper concerns important and difficult problems connected with a design and construction of thermonuclear reactors, which have to use nuclear fusion reactions of heavy isotopes of hydrogen, i.e., deuterium (D and tritium (T. There are described conditions in which such reactions can occur, and different methods of a high-temperature plasma generation, i.e., high-current electrical discharges, intense microwave pulses, and injection of energetic neutral atoms (NBI. There are also presented experimental facilities which can contain hot plasma for an appropriate period, and particularly so-called tokamaks. The second part presents the technical problems which must be solved in order to build a thermonuclear reactor, that might be used for energetic purposes. There are considered problems connected with a choice of constructional materials for a vacuum chamber, its internal parts, external windings generating a magnetic field, and necessary shields. The next part considers the handling of radioactive tritium; the using of alpha particles (4He for additional heating of plasma; recuperation of hydrogen isotopes absorbed in the tokamak internal parts, and a removal of a helium excess. There is presented a scheme of a future thermonuclear power plant and critical comments on a road map which should enable the construction of an industrial thermonuclear reactor (DEMO.

  3. High yields of hydrogen production from methanol steam reforming with a cross-U type reactor.

    Science.gov (United States)

    Zhang, Shubin; Zhang, Yufeng; Chen, Junyu; Zhang, Xuelin; Liu, Xiaowei

    2017-01-01

    This paper presents a numerical and experimental study on the performance of a methanol steam reformer integrated with a hydrogen/air combustion reactor for hydrogen production. A CFD-based 3D model with mass and momentum transport and temperature characteristics is established. The simulation results show that better performance is achieved in the cross-U type reactor compared to either a tubular reactor or a parallel-U type reactor because of more effective heat transfer characteristics. Furthermore, Cu-based micro reformers of both cross-U and parallel-U type reactors are designed, fabricated and tested for experimental validation. Under the same condition for reforming and combustion, the results demonstrate that higher methanol conversion is achievable in cross-U type reactor. However, it is also found in cross-U type reactor that methanol reforming selectivity is the lowest due to the decreased water gas shift reaction under high temperature, thereby carbon monoxide concentration is increased. Furthermore, the reformed gas generated from the reactors is fed into a high temperature proton exchange membrane fuel cell (PEMFC). In the test of discharging for 4 h, the fuel cell fed by cross-U type reactor exhibits the most stable performance.

  4. High yields of hydrogen production from methanol steam reforming with a cross-U type reactor.

    Directory of Open Access Journals (Sweden)

    Shubin Zhang

    Full Text Available This paper presents a numerical and experimental study on the performance of a methanol steam reformer integrated with a hydrogen/air combustion reactor for hydrogen production. A CFD-based 3D model with mass and momentum transport and temperature characteristics is established. The simulation results show that better performance is achieved in the cross-U type reactor compared to either a tubular reactor or a parallel-U type reactor because of more effective heat transfer characteristics. Furthermore, Cu-based micro reformers of both cross-U and parallel-U type reactors are designed, fabricated and tested for experimental validation. Under the same condition for reforming and combustion, the results demonstrate that higher methanol conversion is achievable in cross-U type reactor. However, it is also found in cross-U type reactor that methanol reforming selectivity is the lowest due to the decreased water gas shift reaction under high temperature, thereby carbon monoxide concentration is increased. Furthermore, the reformed gas generated from the reactors is fed into a high temperature proton exchange membrane fuel cell (PEMFC. In the test of discharging for 4 h, the fuel cell fed by cross-U type reactor exhibits the most stable performance.

  5. High yields of hydrogen production from methanol steam reforming with a cross-U type reactor

    Science.gov (United States)

    Zhang, Shubin; Chen, Junyu; Zhang, Xuelin; Liu, Xiaowei

    2017-01-01

    This paper presents a numerical and experimental study on the performance of a methanol steam reformer integrated with a hydrogen/air combustion reactor for hydrogen production. A CFD-based 3D model with mass and momentum transport and temperature characteristics is established. The simulation results show that better performance is achieved in the cross-U type reactor compared to either a tubular reactor or a parallel-U type reactor because of more effective heat transfer characteristics. Furthermore, Cu-based micro reformers of both cross-U and parallel-U type reactors are designed, fabricated and tested for experimental validation. Under the same condition for reforming and combustion, the results demonstrate that higher methanol conversion is achievable in cross-U type reactor. However, it is also found in cross-U type reactor that methanol reforming selectivity is the lowest due to the decreased water gas shift reaction under high temperature, thereby carbon monoxide concentration is increased. Furthermore, the reformed gas generated from the reactors is fed into a high temperature proton exchange membrane fuel cell (PEMFC). In the test of discharging for 4 h, the fuel cell fed by cross-U type reactor exhibits the most stable performance. PMID:29121067

  6. Solid oxide electrochemical reactor science.

    Energy Technology Data Exchange (ETDEWEB)

    Sullivan, Neal P. (Colorado School of Mines, Golden, CO); Stechel, Ellen Beth; Moyer, Connor J. (Colorado School of Mines, Golden, CO); Ambrosini, Andrea; Key, Robert J. (Colorado School of Mines, Golden, CO)

    2010-09-01

    Solid-oxide electrochemical cells are an exciting new technology. Development of solid-oxide cells (SOCs) has advanced considerable in recent years and continues to progress rapidly. This thesis studies several aspects of SOCs and contributes useful information to their continued development. This LDRD involved a collaboration between Sandia and the Colorado School of Mines (CSM) ins solid-oxide electrochemical reactors targeted at solid oxide electrolyzer cells (SOEC), which are the reverse of solid-oxide fuel cells (SOFC). SOECs complement Sandia's efforts in thermochemical production of alternative fuels. An SOEC technology would co-electrolyze carbon dioxide (CO{sub 2}) with steam at temperatures around 800 C to form synthesis gas (H{sub 2} and CO), which forms the building blocks for a petrochemical substitutes that can be used to power vehicles or in distributed energy platforms. The effort described here concentrates on research concerning catalytic chemistry, charge-transfer chemistry, and optimal cell-architecture. technical scope included computational modeling, materials development, and experimental evaluation. The project engaged the Colorado Fuel Cell Center at CSM through the support of a graduate student (Connor Moyer) at CSM and his advisors (Profs. Robert Kee and Neal Sullivan) in collaboration with Sandia.

  7. Status of reactor-shielding research in the US

    Energy Technology Data Exchange (ETDEWEB)

    Maienshein, F.C.

    1980-01-01

    While reactor programs change, shielding analysis methods are improved slowly. Version-V of ENDF/B provides improved data and Version-VI will be cost effective in advanced fission reactors are to be developed in the US. Benchmarks for data and methods validation are collected and distributed in the US in two series, one primarily for FBR-related experiments and one for LWR calculational methods. For LWR design, cavity streaming is now handled adequately, if with varying degrees of elegance. Investigations of improved detector response for LWRs rely upon transport methods. The great potential importance of pressure-vessel damage is dreflected in widespread studies to aid in the prediction of neutron fluences in vessels. For LMFBRS, the FFTF should give attenuation results on an operating reactor. For larger power reactors, the advantages of alternate shield materials appear compelling. A few other shielding studies appear to require experimental confirmation if LMFBRs are to be economically competitive. A coherent shielding program for the GCFR is nearing completion. For the fusion-reactor program, methods verification is under way, practical calculations are well advanced for test devices such as the TFTR and FMIT, and consideration is now given to shielding problems of large reactors, as in the ETF study.

  8. Metabolic modeling of synthesis gas fermentation in bubble column reactors.

    Science.gov (United States)

    Chen, Jin; Gomez, Jose A; Höffner, Kai; Barton, Paul I; Henson, Michael A

    2015-01-01

    A promising route to renewable liquid fuels and chemicals is the fermentation of synthesis gas (syngas) streams to synthesize desired products such as ethanol and 2,3-butanediol. While commercial development of syngas fermentation technology is underway, an unmet need is the development of integrated metabolic and transport models for industrially relevant syngas bubble column reactors. We developed and evaluated a spatiotemporal metabolic model for bubble column reactors with the syngas fermenting bacterium Clostridium ljungdahlii as the microbial catalyst. Our modeling approach involved combining a genome-scale reconstruction of C. ljungdahlii metabolism with multiphase transport equations that govern convective and dispersive processes within the spatially varying column. The reactor model was spatially discretized to yield a large set of ordinary differential equations (ODEs) in time with embedded linear programs (LPs) and solved using the MATLAB based code DFBAlab. Simulations were performed to analyze the effects of important process and cellular parameters on key measures of reactor performance including ethanol titer, ethanol-to-acetate ratio, and CO and H2 conversions. Our computational study demonstrated that mathematical modeling provides a complementary tool to experimentation for understanding, predicting, and optimizing syngas fermentation reactors. These model predictions could guide future cellular and process engineering efforts aimed at alleviating bottlenecks to biochemical production in syngas bubble column reactors.

  9. Decommissioning of commercial reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yui, Kohei [Japan Atomic Power Co., Tokyo (Japan)

    1997-02-01

    In the case of nuclear reactors, the diversion is often difficult as they are highly purposive, the disassembling is not easy as they are robust, and attention is required to handle the equipment containing radioactive substances. Decommissioning is defined as all the measures taken from the state that facilities become unused to the state of becoming green field. In Japan, already 40 years have elapsed since the effort for nuclear power was begun, and in this paper, the present state and future subjects of the decommissioning of nuclear power stations are summarized at the opportunity that the stop of commercial operation of Tokai Nuclear Power Station was decided recently. In the Tokai Nuclear Power Station, 166 MWe graphite-moderated, carbon dioxide-cooled reactor called improved Calder Hall type is installed, which started the operation in 1966. The circumstances of the decision to stop its operation are explained. The basic policy of the decommissioning of commercial nuclear power stations has been already published by the Advisory Committee for Energy. The state of the decommissioning in various foreign countries is reported. In Japan, the state of green field was realized in 1996 in the decommissioning of the JPDR in Japan Atomic Energy Research institute, and the decommissioning of the atomic powered ship ``Mutsu`` was completed. (K.I.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Author, Not Given

    1959-09-01

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

  11. Influence of denitrification reactor retention time distribution (RTD) on dissolved oxygen control and nitrogen removal efficiency.

    Science.gov (United States)

    Raboni, Massimo; Gavasci, Renato; Viotti, Paolo

    2015-01-01

    Low concentrations of dissolved oxygen (DO) are usually found in biological anoxic pre-denitrification reactors, causing a reduction in nitrogen removal efficiency. Therefore, the reduction of DO in such reactors is fundamental for achieving good nutrient removal. The article shows the results of an experimental study carried out to evaluate the effect of the anoxic reactor hydrodynamic model on both residual DO concentration and nitrogen removal efficiency. In particular, two hydrodynamic models were considered: the single completely mixed reactor and a series of four reactors that resemble plug-flow behaviour. The latter prove to be more effective in oxygen consumption, allowing a lower residual DO concentration than the former. The series of reactors also achieves better specific denitrification rates and higher denitrification efficiency. Moreover, the denitrification food to microrganism (F:M) ratio (F:MDEN) demonstrates a relevant synergic action in both controlling residual DO and improving the denitrification performance.

  12. Monitoring and Control Research Using a University Reactor and SBWR Test-Loop

    Energy Technology Data Exchange (ETDEWEB)

    Robert M. Edwards

    2003-09-28

    The existing hybrid simulation capability of the Penn State Breazeale nuclear reactor was expanded to conduct research for monitoring, operations and control. Hybrid simulation in this context refers to the use of the physical time response of the research reactor as an input signal to a real-time simulation of power-reactor thermal-hydraulics which in-turn provides a feedback signal to the reactor through positioning of an experimental changeable reactivity device. An ECRD is an aluminum tube containing an absorber material that is positioned in the central themble of the reactor kinetics were used to expand the hybrid reactor simulation (HRS) capability to include out-of-phase stability characteristics observed in operating BWRs.

  13. Design and operation of a filter reactor for continuous production of a selected pharmaceutical intermediate

    DEFF Research Database (Denmark)

    Christensen, Kim Müller; Pedersen, Michael Jønch; Dam-Johansen, Kim

    2012-01-01

    A novel filter reactor system for continuous production of selected pharmaceutical intermediates is presented and experimentally verified. The filter reactor system consists of a mixed flow reactor equipped with a bottom filter, to retain solid reactant particles, followed by a conventional plug...... flow reactor, where residual reactant is converted by titration. A chemical case study, production of the pharmaceutical intermediate allylcarbinol by a reaction between allylmagnesium chloride and 2-chloro-thioxanthone, in the presence of a side reaction is considered. The synthesis is conducted......-batch operation, are reduced impurity formation and the use of much lower reactor volumes (factor of 1000 based on the laboratory reactor) and less solvent consumption (from 5.8 to 2.3L/kg reactant). Added challenges include handling of continuous solid powder feeding, stable pumping of reactive slurries...

  14. Participation in the U.S. Department of Energy Reactor Sharing Program. Progress report

    Energy Technology Data Exchange (ETDEWEB)

    Mulder, R.U.; Benneche, P.E.; Hosticka, B.

    1997-03-01

    The objective of the DOE supported Reactor Sharing Program is to increase the availability of university nuclear reactor facilities to non-reactor-owning educational institutions. The educational and research programs of these user institutions is enhanced by the use of the nuclear facilities. Several methods have been used by the UVA Reactor Facility to achieve this objective. First, many college and secondary school groups toured the Reactor Facility and viewed the UVAR reactor and associated experimental facilities. Second, advanced undergraduate and graduate classes from area colleges and universities visited the facility to perform experiments in nuclear engineering and physics which would not be possible at the user institution. Third, irradiation and analysis services at the Facility have been made available for research by faculty and students from user institutions. Fourth, some institutions have received activated material from UVA for use at their institutions. These areas are discussed here.

  15. Startup of pilot-scale aerobic granular sludge reactor by stored granules.

    Science.gov (United States)

    Liu, Q S; Liu, Y; Tay, S T L; Show, K Y; Ivanov, V; Benjamin, M; Tay, J H

    2005-12-01

    The startup of a pilot-scale aerobic granular sludge reactor was investigated by seeding with 4-month stored aerobic granules. Two liters of granules were inoculated into the reactor (5.9% of reactor volume), which gave a biomass concentration of 1.03 g l(-1). Experimental results showed that seeding granules could be successfully maintained in the reactor. The microbial activity of seeding granules could be fully recovered to that of fresh granules after 2 days of operation, and new granules started to grow after day 5. Newly developed aerobic granules at stable period had similar size and morphology as seeding granules, and a biomass concentration of 6.0 g l(-1) was achieved in the reactor. The experiment demonstrated for the first time that stored aerobic granules could be used for easy and quick startup of aerobic granular sludge reactor.

  16. Radionuclide release from research reactor spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Curtius, H., E-mail: h.curtius@fz-juelich.de [Forschungszentrum Juelich, Institut fuer Energieforschung, IEF-6 Sicherheitsforschung und Reaktortechnik, Geb. 05.3, D-52425 Juelich (Germany); Kaiser, G.; Mueller, E.; Bosbach, D. [Forschungszentrum Juelich, Institut fuer Energieforschung, IEF-6 Sicherheitsforschung und Reaktortechnik, Geb. 05.3, D-52425 Juelich (Germany)

    2011-09-01

    Numerous investigations with respect to LWR fuel under non oxidizing repository relevant conditions were performed. The results obtained indicate slow corrosion rates for the UO{sub 2} fuel matrix. Special fuel-types (mostly dispersed fuels, high enriched in {sup 235}U, cladded with aluminium) are used in German research reactors, whereas in German nuclear power plants, UO{sub 2}-fuel (LWR fuel, enrichment in {sup 235}U up to 5%, zircaloy as cladding) is used. Irradiated research reactor fuels contribute less than 1% to the total waste volume. In Germany, the state is responsible for fuel operation and for fuel back-end options. The institute for energy research (IEF-6) at the Research Center Juelich performs investigation with irradiated research reactor spent fuels under repository relevant conditions. In the study, the corrosion of research reactor spent fuel has been investigated in MgCl{sub 2}-rich salt brine and the radionuclide release fractions have been determined. Leaching experiments in brine with two different research reactor fuel-types were performed in a hot cell facility in order to determine the corrosion behaviour and the radionuclide release fractions. The corrosion of two dispersed research reactor fuel-types (UAl{sub x}-Al and U{sub 3}Si{sub 2}-Al) was studied in 400 mL MgCl{sub 2}-rich salt brine in the presence of Fe{sup 2+} under static and initially anoxic conditions. Within these experimental parameters, both fuel types corroded in the experimental time period of 3.5 years completely, and secondary alteration phases were formed. After complete corrosion of the used research reactor fuel samples, the inventories of Cs and Sr were quantitatively detected in solution. Solution concentrations of Am and Eu were lower than the solubility of Am(OH){sub 3}(s) and Eu(OH){sub 3}(s) solid phases respectively, and may be controlled by sorption processes. Pu concentrations may be controlled by Pu(IV) polymer species, but the presence of Pu(V) and Pu

  17. The performance of structured packings in trickle-bed reactors.

    NARCIS (Netherlands)

    Frank, M.J.W.; Frank, M.J.W.; Kuipers, J.A.M.; Versteeg, Geert; van Swaaij, Willibrordus Petrus Maria

    1999-01-01

    An experimental study was carried out to investigate whether the use of structured packings might improve the mass transfer characteristics and the catalyst effectiveness of a trickle-bed reactor. Therefore, the performances of a structured packing, consisting of KATAPAK elements, and a dumped

  18. The Performance of Structured Packings in Trickle-Bed Reactors

    NARCIS (Netherlands)

    Frank, M.J.W.; Kuipers, J.A.M.; Versteeg, G.F.; Swaaij, W.P.M. van

    1999-01-01

    An experimental study was carried out to investigate whether the use of structured packings might improve the mass transfer characteristics and the catalyst effectiveness of a trickle-bed reactor. Therefore, the performances of a structured packing, consisting of KATAPAK elements, and a dumped

  19. Study of fast reactor safety test facilities. Preliminary report

    Energy Technology Data Exchange (ETDEWEB)

    Bell, G.I.; Boudreau, J.E.; McLaughlin, T.; Palmer, R.G.; Starkovich, V.; Stein, W.E.; Stevenson, M.G.; Yarnell, Y.L.

    1975-05-01

    Included are sections dealing with the following topics: (1) perspective and philosophy of fast reactor safety analysis; (2) status of accident analysis and experimental needs; (3) experiment and facility definitions; (4) existing in-pile facilities; (5) new facility options; and (6) data acquisition methods. (DG)

  20. Solvent refined coal reactor quench system

    Science.gov (United States)

    Thorogood, R.M.

    1983-11-08

    There is described an improved SRC reactor quench system using a condensed product which is recycled to the reactor and provides cooling by evaporation. In the process, the second and subsequent reactors of a series of reactors are cooled by the addition of a light oil fraction which provides cooling by evaporation in the reactor. The vaporized quench liquid is recondensed from the reactor outlet vapor stream. 1 fig.

  1. Brookhaven leak reactor to close

    CERN Multimedia

    MacIlwain, C

    1999-01-01

    The DOE has announced that the High Flux Beam Reactor at Brookhaven is to close for good. Though the news was not unexpected researchers were angry the decision had been taken before the review to assess the impact of reopening the reactor had been concluded (1 page).

  2. Rotary reactor and use thereof

    NARCIS (Netherlands)

    Bakker Wridzer, J.W.; Kapteijn, F.; Moulijn, J.A.

    1998-01-01

    The invention relates to a rotary reactor consisting of a number of tubular reaction compartments (A), each provided with a first end and a second end, a ceramic first reactor end plate (B) in which said first ends are received, and a second end plate (B) in which said second ends are received,

  3. Phosphorus removal from UASB reactor effluent by reactive media filtration.

    Science.gov (United States)

    Rodríguez-Gómez, Raúl; Renman, Gunno

    2017-08-01

    The phosphorus (P) and BOD7 removal performance of an upflow packed bed reactor (PBR) filled with two reactive filter media was studied over 50 weeks. The lower one-fifth of the reactor was filled with calcium-silicate-hydrate (Sorbulite®) and the upper four-fifth with calcium-silicate (Polonite®). A laboratory-scale upflow anaerobic sludge bed reactor (UASB) delivered wastewater to the PBR. A model was developed to describe the gradient in P concentration change in the reactor, based on reaction kinetics. The reaction terms were assumed to follow the Langmuir isotherm, based on the results obtained in a batch test. First, a comparison was made between experimental and simulated results. The capability of the model to forecast P removal capacity was then tested for three hypothetical cases: (i) reactor filled with Sorbulite and Polonite, (ii) reactor filled with only Sorbulite, and (iii) reactor filled with only Polonite. Finally, a sensitivity analysis was performed for the main parameters in the model. The average removal of P and BOD7 from the UASB effluent was 98% and 90%, respectively. The starting pH of the dual-medium effluent was 12.2 and decreased gradually over time to 11.1. The simulation both overestimated and underestimated mean measured P removal but was within the range of maximum and minimum measured values. The hypothetical cases revealed that most P was removed by Polonite due to calcium phosphate precipitation. The removal capacity of the two filter materials and their layer height in the reactor were the most sensitive parameters in the simulation.

  4. Radiation Shielding for Fusion Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Santoro, R.T.

    1999-10-01

    Radiation shielding requirements for fusion reactors present different problems than those for fission reactors and accelerators. Fusion devices, particularly tokamak reactors, are complicated by geometry constraints that complicate disposition of fully effective shielding. This paper reviews some of these shielding issues and suggested solutions for optimizing the machine and biological shielding. Radiation transport calculations are essential for predicting and confirming the nuclear performance of the reactor and, as such, must be an essential part of the reactor design process. Development and optimization of reactor components from the first wall and primary shielding to the penetrations and containment shielding must be carried out in a sensible progression. Initial results from one-dimensional transport calculations are used for scoping studies and are followed by detailed two- and three-dimensional analyses to effectively characterize the overall radiation environment. These detail model calculations are essential for accounting for the radiation leakage through ports and other penetrations in the bulk shield. Careful analysis of component activation and radiation damage is cardinal for defining remote handling requirements, in-situ replacement of components, and personnel access at specific locations inside the reactor containment vessel. Radiation shielding requirements for fusion reactors present different problems than those for fission reactors and accelerators. Fusion devices, particularly tokamak reactors, are complicated by geometry constraints that complicate disposition of fully effective shielding. This paper reviews some of these shielding issues and suggested solutions for optimizing the machine and biological shielding. Radiation transport calculations are essential for predicting and confirming the nuclear performance of the reactor and, as such, must be an essential part of the reactor design process. Development and optimization of reactor

  5. Thermal-Hydraulic Experiments and Modelling for Advanced Nuclear Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Song, C. H.; Chung, M. K.; Park, C. K. and others

    2005-04-15

    The objectives of the project are to study thermal hydraulic characteristics of reactor primary system for the verification of the reactor safety and to evaluate new safety concepts of new safety design features. To meet the research goal, several thermal hydraulic experiments were performed and related thermal hydraulic models were developed with the experimental data which were produced through the thermal hydraulic experiments. Followings are main research topics; - Multi-dimensional Phenomena in a Reactor Vessel Downcomer - Condensation Load and Thermal Mixing in the IRWST - Development of Thermal-Hydraulic Models for Two-Phase Flow - Development of Measurement Techniques for Two-Phase Flow - Supercritical Reactor T/H Characteristics Analysis From the above experimental and analytical studies, new safety design features of the advanced power reactors were verified and lots of the safety issues were also resolved.

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

    Energy Technology Data Exchange (ETDEWEB)

    Acena Moreno, V.

    1981-07-01

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

  7. Production test-080, physics testing at D reactor deactivation

    Energy Technology Data Exchange (ETDEWEB)

    Bailey, G.F.

    1967-06-15

    The purpose of this test is to provide a set of experimental data to test a compute code frequently used in nuclear safety analyses and to explore certain experimental techniques which may prove extremely valuable in the future. In addition, some basic physics parameters which will be measured may be used in an assessment of the feasibility of using a deactivated Hanford reactor for space-dependent transient tests.

  8. Evaluation of model parameters for simulating TiO(2) coated UV reactors.

    Science.gov (United States)

    Duran, J E; Taghipour, F; Mohseni, M

    2011-01-01

    A CFD-based model for simulating TiO(2) coated photocatalytic reactors used in drinking water treatment applications was preliminarily evaluated. The model includes aspects of hydrodynamics, mass transfer, UV-radiation field, and surface chemical reactions. Appropriate models for each of the associated physicochemical phenomena were experimentally or analytically examined. Once defined and evaluated, the individual models were integrated into a CFD-based model for simulating photocatalytic reactor performance, which was experimentally evaluated.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-07-01

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

  10. Neutrino Experiments at Reactors

    Science.gov (United States)

    Reines, F.; Gurr, H. S.; Jenkins, T. L.; Munsee, J. H.

    1968-09-09

    A description is given of the electron-antineutrino program using a large fission reactor. A search has been made for a neutral weak interaction via the reaction (electron antineutrino + d .> p + n + electron antineutrino), the reaction (electron antineutrino + d .> n + n + e{sup +}) has now been detected, and an effort is underway to observe the elastic scattering reaction (electron antineutrino + e{sup -} .> electron antineutrino + e{sup -}) as well as to measure more precisely the reaction (electron antineutrino + p .> n + e{sup+}). The upper limit on the elastic scattering reaction which we have obtained with our large composite NaI, plastic, liquid scintillation detector is now about 50 times the predicted value.

  11. Experimental Neutrino Physics: Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Lane, Charles E.; Maricic, Jelena

    2012-09-05

    Experimental studies of neutrino properties, with particular emphasis on neutrino oscillation, mass and mixing parameters. This research was pursued by means of underground detectors for reactor anti-neutrinos, measuring the flux and energy spectra of the neutrinos. More recent investigations have been aimed and developing detector technologies for a long-baseline neutrino experiment (LBNE) using a neutrino beam from Fermilab.

  12. Licensed reactor nuclear safety criteria applicable to DOE reactors

    Energy Technology Data Exchange (ETDEWEB)

    1993-11-01

    This document is a compilation and source list of nuclear safety criteria that the Nuclear Regulatory Commission (NRC) applies to licensed reactors; it can be used by DOE and DOE contractors to identify NRC criteria to be evaluated for application to the DOE reactors under their cognizance. The criteria listed are those that are applied to the areas of nuclear safety addressed in the safety analysis report of a licensed reactor. They are derived from federal regulations, USNRC regulatory guides, Standard Review Plan (SRP) branch technical positions and appendices, and industry codes and standards.

  13. Advanced fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tomita, Yukihiro [National Inst. for Fusion Science, Toki, Gifu (Japan)

    2003-04-01

    The main subjects on fusion research are now on D-T fueled fusion, mainly due to its high fusion reaction rate. However, many issues are still remained on the wall loading by the 14 MeV neutrons. In the case of D-D fueled fusion, the neutron wall loading is still remained, though the technology related to tritium breeding is not needed. The p-{sup 6}Li and p-{sup 11}B fueled fusions are not estimated to be the next generation candidate until the innovated plasma confinement technologies come in useful to achieve the high performance plasma parameters. The fusion reactor of D-{sup 3}He fuels has merits on the smaller neutron wall loading and tritium handling. However, there are difficulties on achieving the high temperature plasma more than 100 keV. Furthermore the high beta plasma is needed to decrease synchrotron radiation loss. In addition, the efficiency of the direct energy conversion from protons coming out from fusion reaction is one of the key parameters in keeping overall power balance. Therefore, open magnetic filed lines should surround the plasma column. In this paper, we outlined the design of the commercial base reactor (ARTEMIS) of 1 GW electric output power configured by D-{sup 3}He fueled FRC (Field Reversed Configuration). The ARTEMIS needs 64 kg of {sup 3}He per a year. On the other hand, 1 million tons of {sup 3}He is estimated to be in the moon. The {sup 3}He of about 10{sup 23} kg are to exist in gaseous planets such as Jupiter and Saturn. (Y. Tanaka)

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

    Directory of Open Access Journals (Sweden)

    V.I. Belozerov

    2016-12-01

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

  15. Study and optimization of an annular photocatalytic slurry reactor.

    Science.gov (United States)

    Camera-Roda, Giovanni; Santarelli, Francesco; Panico, Mauro

    2009-05-01

    The experimental results obtained for the photocatalytic degradation of a model organic dye in an annular slurry reactor are analyzed with the aid of a mathematical model. The model is used also to study the effects on the performances of many operative conditions: flow rate, photocatalyst concentration, power of the lamp, size of the photocatalytic particles, dimensions of the reactor. The investigation demonstrates that the rate of the process is often limited by the radiant energy transfer and that some simple rules can be followed in order to optimize different yields and the observed rate of reaction.

  16. UASB reactor hydrodynamics: residence time distribution and proposed modelling tools.

    Science.gov (United States)

    López, I; Borzacconi, L

    2010-05-01

    The hydrodynamic behaviour of UASB (Up Flow Anaerobic Sludge Blanket) reactors based on residence time distribution curves allows the implementation of global models, including the kinetic aspects of biological reactions. The most relevant hydrodynamic models proposed in the literature are discussed and compared with the extended tanks in series (ETIS) model. Although derived from the tanks in series model, the ETIS model's parameter is not an integer. The ETIS model can be easily solved in the Laplace domain and applied to a two-stage anaerobic digestion linear model. Experimental data from a 250 m3 UASB reactor treating malting wastewater are used to calibrate and validate the proposed model.

  17. 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.

  18. Maltodextrin hydrolysis in a fluidized-bed immobilized enzyme reactor

    Energy Technology Data Exchange (ETDEWEB)

    Vallat, I.; Monsan, P.; Riba, J.P.

    1986-02-01

    The present work deals with maltodextrin hydrolysis by glucoamylase immobilized onto corn stover in a fluidized bed reactor. An industrial enzyme preparation was convalently grafted onto corn stover, yielding an activity of up to 372 U/g and 1700 U/g for support particle sizes of 0.8 and 0.2 mm, respectively. A detailed kinetic study, using a differntial reactor, allowed the characterization of the influence of mass transfer resistance on the reaction catalyzed by immobilized glucoamylase. A simple and general mathematical model was then developed to describe the experimental conversion data and found to be vaild.

  19. Computational analysis of irradiation facilities at the JSI TRIGA reactor.

    Science.gov (United States)

    Snoj, Luka; Zerovnik, Gašper; Trkov, Andrej

    2012-03-01

    Characterization and optimization of irradiation facilities in a research reactor is important for optimal performance. Nowadays this is commonly done with advanced Monte Carlo neutron transport computer codes such as MCNP. However, the computational model in such calculations should be verified and validated with experiments. In the paper we describe the irradiation facilities at the JSI TRIGA reactor and demonstrate their computational characterization to support experimental campaigns by providing information on the characteristics of the irradiation facilities. Copyright © 2011 Elsevier Ltd. All rights reserved.

  20. Heating device for nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shiratori, Yoshitake; Ijima, Takashi; Katano, Yoshiaki; Saito, Masaki

    1996-05-31

    The present invention provides a control system of a heating device which elevates the temperature of a reactor from a normal temperature to an operation temperature by using a nuclear heating. Namely, the device of the present invention comprises (1) means for detecting reactor temperature, (2) means for detecting reactor power, (3) means for memorizing the corresponding relation of each value of the means (1) and means (2) as standard data when temperature is elevated at a predetermined temperature elevation rate, (4) means for calculating the power corresponding to the current temperature based on the standard data upon elevation of the reactor temperature, and (5) means for controlling the progress or retraction of the power control material of the reactor core based on the power calculated by the means (4). With such a constitution, since the current reactor power elevation rate corresponding to the coolants is controlled based on the standard data upon actual start-up of the reactor, the control for the temperature of coolants can be facilitated. (I.S.)

  1. Neutrino Physics with Accelerator Driven Subcritical Reactors

    Science.gov (United States)

    Ciuffoli, Emilio

    2017-09-01

    Accelerator Driven Subcritical System (ADS) reactors are being developed around the world, to produce energy and, at the same time, to provide an efficient way to dispose of and to recycle nuclear waste. Used nuclear fuel, by itself, cannot sustain a chain reaction; however in ADS reactors the additional neutrons which are required will be supplied by a high-intensity accelerator. This accelerator will produce, as a by-product, a large quantity of {\\bar{ν }}μ via muon Decay At Rest (µDAR). Using liquid scintillators, it will be possible to to measure the CP-violating phase δCP and to look for experimental signs of the presence of sterile neutrinos in the appearance channel, testing the LSND and MiniBooNE anomalies. Even in the first stage of the project, when the beam energy will be lower, it will be possible to produce {\\bar{ν }}e via Isotope Decay At Rest (IsoDAR), which can be used to provide competitive bounds on sterile neutrinos in the disappearance channel. I will consider several experimental setups in which the antineutrinos are created using accelerators that will be constructed as part of the China-ADS program.

  2. Advanced sodium fast reactor accident source terms :

    Energy Technology Data Exchange (ETDEWEB)

    Powers, Dana Auburn; Clement, Bernard; Denning, Richard; Ohno, Shuji; Zeyen, Roland

    2010-09-01

    An expert opinion elicitation has been used to evaluate phenomena that could affect releases of radionuclides during accidents at sodium-cooled fast reactors. The intent was to identify research needed to develop a mechanistic model of radionuclide release for licensing and risk assessment purposes. Experts from the USA, France, the European Union, and Japan identified phenomena that could affect the release of radionuclides under hypothesized accident conditions. They qualitatively evaluated the importance of these phenomena and the need for additional experimental research. The experts identified seven phenomena that are of high importance and have a high need for additional experimental research: High temperature release of radionuclides from fuel during an energetic event Energetic interactions between molten reactor fuel and sodium coolant and associated transfer of radionuclides from the fuel to the coolant Entrainment of fuel and sodium bond material during the depressurization of a fuel rod with breached cladding Rates of radionuclide leaching from fuel by liquid sodium Surface enrichment of sodium pools by dissolved and suspended radionuclides Thermal decomposition of sodium iodide in the containment atmosphere Reactions of iodine species in the containment to form volatile organic iodides. Other issues of high importance were identified that might merit further research as development of the mechanistic model of radionuclide release progressed.

  3. Random processes in nuclear reactors

    CERN Document Server

    Williams, M M R

    1974-01-01

    Random Processes in Nuclear Reactors describes the problems that a nuclear engineer may meet which involve random fluctuations and sets out in detail how they may be interpreted in terms of various models of the reactor system. Chapters set out to discuss topics on the origins of random processes and sources; the general technique to zero-power problems and bring out the basic effect of fission, and fluctuations in the lifetime of neutrons, on the measured response; the interpretation of power reactor noise; and associated problems connected with mechanical, hydraulic and thermal noise sources

  4. Nuclear reactor downcomer flow deflector

    Science.gov (United States)

    Gilmore, Charles B [Greensburg, PA; Altman, David A [Pittsburgh, PA; Singleton, Norman R [Murrysville, PA

    2011-02-15

    A nuclear reactor having a coolant flow deflector secured to a reactor core barrel in line with a coolant inlet nozzle. The flow deflector redirects incoming coolant down an annulus between the core barrel and the reactor vessel. The deflector has a main body with a front side facing the fluid inlet nozzle and a rear side facing the core barrel. The rear side of the main body has at least one protrusion secured to the core barrel so that a gap exists between the rear side of the main body adjacent the protrusion and the core barrel. Preferably, the protrusion is a relief that circumscribes the rear side of the main body.

  5. Reactor neutrons in nuclear astrophysics

    Science.gov (United States)

    Reifarth, René; Glorius, Jan; Göbel, Kathrin; Heftrich, Tanja; Jentschel, Michael; Jurado, Beatriz; Käppeler, Franz; Köster, Ulli; Langer, Christoph; Litvinov, Yuri A.; Weigand, Mario

    2017-09-01

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

  6. Fuel Fabrication and Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Karpius, Peter Joseph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-02-02

    The uranium from the enrichment plant is still in the form of UF6. UF6 is not suitable for use in a reactor due to its highly corrosive chemistry as well as its phase diagram. UF6 is converted into UO2 fuel pellets, which are in turn placed in fuel rods and assemblies. Reactor designs are variable in moderators, coolants, fuel, performance etc.The dream of energy ‘too-cheap to meter’ is no more, and now the nuclear power industry is pushing ahead with advanced reactor designs.

  7. Final report on in-reactor creep-fatigue deformation behaviour of a CuCrZr alloy: COFAT 1

    DEFF Research Database (Denmark)

    Singh, Bachu Narain; Tähtinen, S.; Moilanen, P.

    CrZr(HT1) alloy exposed concurrently to flux of neutrons and creep-fatigue cyclic loading directly in a fission reactor. Special experimental facilities were designed and fabricated for this purpose. A number of in-reactor creep-fatigue experiments were successfully carried out in the BR-2 reactor at Mol...... (Belgium). In the present report we first describe the experimental facilities and the details of the in-reactor creep-fatigue experiments carried out at 363 and 343K at a strain amplitude of 0.5% with holdtimes of 10 and 100s, respectively. For comparison purposes, similar creep-fatigue tests were...... performed outside of the reactor. (i.e. in the absence of neutron irradiation). During in-reactor tests, the mechanical response was continuously registered throughout the whole test. The results are first presented in the form of hysteresis loops confirming that the nature of deformation during these tests...

  8. Study of heat and synchrotron radiation transport in fusion tokamak plasmas. Application to the modelling of steady state and fast burn termination scenarios for the international experimental fusion reactor ITER

    Energy Technology Data Exchange (ETDEWEB)

    Villar Colome, J. [Association Euratom-CEA, Centre d`Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee]|[Universitat Polytechnica de Catalunya (Spain)

    1997-12-01

    The aim of this thesis is to give a global scope of the problem of energy transport within a thermonuclear plasma in the context of its power balance and the implications when modelling ITER operating scenarios. This is made in two phases. First, by furnishing new elements to the existing models of heat and synchrotron radiation transport in a thermonuclear plasma. Second, by applying the improved models to plasma engineering studies of ITER operating scenarios. The scenarios modelled are the steady state operating point and the transient that appears to have the biggest technological implications: the fast burn termination. The conduction-convection losses are modelled through the energy confinement time. This parameter is empirically obtained from the existing experimental data, since the underlying mechanisms are not well understood. In chapter 2 an expression for the energy confinement time is semi-analytically deduced from the Rebut-Lallia-Watkins local transport model. The current estimates of the synchrotron radiation losses are made with expressions of the dimensionless transparency factor deduced from a 0-dimensional cylindrical model proposed by Trubnikov in 1979. In chapter 3 realistic hypothesis for the cases of cylindrical and toroidal geometry are included in the model to deduce compact explicit expressions for the fast numerical computation of the synchrotron radiation losses. Numerical applications are provided for the cylindrical case. The results are checked against the existing models. In chapter 4, the nominal operating point of ITER and its thermal stability is studied by means of a 0-dimensional burn model of the thermonuclear plasma in ignition. This model is deduced by the elements furnished by the plasma particle and power balance. Possible heat overloading on the plasma facing components may provoke severe structural damage, implying potential safety problems related to tritium inventory and metal activation. In chapter 5, the assessment

  9. Nuclear reactor PBMR and cogeneration; Reactor nuclear PBMR y cogeneracion

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez S, J. R.; Alonso V, G., E-mail: ramon.ramirez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2013-10-15

    In recent years the nuclear reactor designs for the electricity generation have increased their costs, so that at the moment costs are managed of around the 5000 US D for installed kw, reason for which a big nuclear plant requires of investments of the order of billions of dollars, the designed reactors as modular of low power seek to lighten the initial investment of a big reactor dividing the power in parts and dividing in modules the components to lower the production costs, this way it can begin to build a module and finished this to build other, differing the long term investment, getting less risk therefore in the investment. On the other hand the reactors of low power can be very useful in regions where is difficult to have access to the electric net being able to take advantage of the thermal energy of the reactor to feed other processes like the water desalination or the vapor generation for the processes industry like the petrochemical, or even more the possible hydrogen production to be used as fuel. In this work the possibility to generate vapor of high quality for the petrochemical industry is described using a spheres bed reactor of high temperature. (Author)

  10. FASTER test reactor preconceptual design report summary

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-02-29

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

  11. FASTER Test Reactor Preconceptual Design Report

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-03-31

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

  12. ADAPTIVE CONTROL SYSTEM OF INDUSTRIAL REACTORS

    Directory of Open Access Journals (Sweden)

    Vyacheslav K. Mayevski

    2014-01-01

    Full Text Available This paper describes a mathematical model of an industrial chemical reactor for production of synthetic rubber. During reactor operation the model parameters vary considerably. To create a control algorithm performed transformation of mathematical model of the reactor in order to obtain a dependency that can be used to determine the model parameters are changing during reactor operation.

  13. Methodological developments and qualification of calculation schemes for the modelling of photonic heating in the experimental devices of the future Jules Horowitz material testing reactor (RJH); Developpements methodologiques et qualification de schemas de calcul pour la modelisation des echauffements photoniques dans les dispositifs experimentaux du futur reacteur d'irradiation technologiques Jules Horowitz (RJH)

    Energy Technology Data Exchange (ETDEWEB)

    Blanchet, D

    2006-07-01

    The objective of this work is to develop the modelling of the nuclear heating of the experimental devices of the future Jules Horowitz material testing reactor (RJH). The strong specific nuclear power produced (460 kW/l), induces so intense photonic fluxes which cause heating and large temperature gradients that it is necessary to control it by an adequate design. However, calculations of heating are penalized by the very large uncertainties estimated at a value of about 30% (2*{sigma}) coming from the gaps and uncertainties of the data of gamma emission present in the libraries of basic nuclear data. The experimental program ADAPh aims at reducing these uncertainties. Measurements by thermoluminescent detectors (TLD) and ionisation chambers are carried out in the critical assemblies EOLE (Mox) and Minerve (UO{sub 2}). The rigorous interpretation of these measurements requires specific developments based on Monte-Carlo simulations of coupled neutron-gamma and gamma-electron transport. The developments carried out are made different in particular by the modelling of cavities phenomena and delayed gamma emissions by the decay of fission products. The comparisons calculation-measurement made it possible to identify a systematic bias confirming a tendency of calculations to underestimate measurements. A Bayesian method of adjustment was developed in order to re-estimate the principal components of the gamma heating and to transpose the results obtained to the devices of the RJH, under conditions clearly and definitely representative. This work made possible to reduce significantly the uncertainties on the determination of the gamma heating from 30 to 15 per cent. (author)

  14. Odor Removal Characteristics of a Laminated Film-Electrode Packed-Bed Nonthermal Plasma Reactor

    Directory of Open Access Journals (Sweden)

    Hideya Kametaka

    2011-05-01

    Full Text Available Odor control has gained importance for ensuring a comfortable living environment. In this paper, the authors report the experimental results of a study on the detailed characteristics of a laminated film-electrode and a laminated film-electrode packed-bed nonthermal plasma reactor, which are types of dielectric barrier discharge (DBD reactor used for odor control. These plasma reactors can be potentially used for the decomposition of volatile organic compounds (VOCs and reduction of NOx. The reactor is driven by a low-cost 60-Hz neon transformer. Removal efficiencies under various experimental conditions are studied. The complete decomposition of the main odor component, namely, NH3, is achieved in a dry environment. The retention times are investigated for the complete removal of NH3 in the case of the film-electrode plasma reactor and the film-electrode packed-bed plasma reactor. The removal efficiency of the former reactor is lower than that of the latter reactor. Mixing another odor component such as CH3CHO in the gas stream has no significant effect on NH3 removal efficiency.

  15. Design of the HPLWR reactor core; Auslegung des HPLWR Reaktorkerns

    Energy Technology Data Exchange (ETDEWEB)

    Schulenberg, T. [Karlsruher Institut fuer Technologie, Karlsruhe (Germany). IKET; Maraczy, C. [KFKI Atomenergia Kutatointezet (AEKI), Budapest (Hungary); Heinecke, J. [AREVA NP GmbH, Erlangen (Germany); Bernnat, W. [Stuttgart Univ. (Germany). IKE

    2010-05-15

    The high performance light water reactor (HPLWR) is a LWR working with supercritical water as coolant medium and moderator. The operational pressure is 25 MPa and the fresh steam temperatures are above 500 C. In order to restrict the peak temperature in the reactor core to less than 630 C (upper limit of the corrosion resistance of stainless steel fuel cans) a three-step heating of the reactor core was proposed. The authors discuss the results of thermal hydraulic and neutronic calculations performed during the last three years. The coolant mixing is the key process of the concept. The design of the fuel and water cans is using double-walled constructions with ceramic insulations to avoid inadmissible heating o the moderator medium. Stress and deformation analysis of the core structures were performed. The calculated results still need experimental validation.

  16. Standard Guide for Benchmark Testing of Light Water Reactor Calculations

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This guide covers general approaches for benchmarking neutron transport calculations in light water reactor systems. A companion guide (Guide E2005) covers use of benchmark fields for testing neutron transport calculations and cross sections in well controlled environments. This guide covers experimental benchmarking of neutron fluence calculations (or calculations of other exposure parameters such as dpa) in more complex geometries relevant to reactor surveillance. Particular sections of the guide discuss: the use of well-characterized benchmark neutron fields to provide an indication of the accuracy of the calculational methods and nuclear data when applied to typical cases; and the use of plant specific measurements to indicate bias in individual plant calculations. Use of these two benchmark techniques will serve to limit plant-specific calculational uncertainty, and, when combined with analytical uncertainty estimates for the calculations, will provide uncertainty estimates for reactor fluences with ...

  17. Advances in process intensification through multifunctional reactor engineering.

    Energy Technology Data Exchange (ETDEWEB)

    Cooper, Marcia A.; Miller, James Edward; O' Hern, Timothy John; Gill, Walter; Evans, Lindsey R.

    2011-02-01

    A multifunctional reactor is a chemical engineering device that exploits enhanced heat and mass transfer to promote production of a desired chemical, combining more than one unit operation in a single system. The main component of the reactor system under study here is a vertical column containing packing material through which liquid(s) and gas flow cocurrently downward. Under certain conditions, a range of hydrodynamic regimes can be achieved within the column that can either enhance or inhibit a desired chemical reaction. To study such reactors in a controlled laboratory environment, two experimental facilities were constructed at Sandia National Laboratories. One experiment, referred to as the Two-Phase Experiment, operates with two phases (air and water). The second experiment, referred to as the Three-Phase Experiment, operates with three phases (immiscible organic liquid and aqueous liquid, and nitrogen). This report describes the motivation, design, construction, operational hazards, and operation of the both of these experiments. Data and conclusions are included.

  18. Development of essential system technologies for advanced reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Y. Y.; Hwang, Y. D.; Cho, B. H. and others

    1999-03-01

    Basic design of SMART adopts the new advanced technologies which were not applied in the existing 1000MWe PWR. However, the R and D experience on these advanced essential technologies is lacking in domestic nuclear industry. Recently, a research on these advanced technologies has been performed as a part of the mid-and-long term nuclear R and D program, but the research was limited only for the small scale fundamental study. The research on these essential technologies such as helically coiled tube steam generator, self pressurizer, core cooling by natural circulation required for the development of integral reactor SMART have not been conducted in full scale. This project, therefore, was performed for the development of analysis models and methodologies, system analysis and thermal hydraulic experiments on the essential technologies to be applied to the 300MWe capacity of integral reactor SMART and the advanced passive reactor expected to be developed in near future with the emphasis on experimental investigation. (author)

  19. Current Reactor Physics Benchmark Activities at the Idaho National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess; Margaret A. Marshall; Mackenzie L. Gorham; Joseph Christensen; James C. Turnbull; Kim Clark

    2011-11-01

    The International Reactor Physics Experiment Evaluation Project (IRPhEP) [1] and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) [2] were established to preserve integral reactor physics and criticality experiment data for present and future research. These valuable assets provide the basis for recording, developing, and validating our integral nuclear data, and experimental and computational methods. These projects are managed through the Idaho National Laboratory (INL) and the Organisation for Economic Co-operation and Development Nuclear Energy Agency (OECD-NEA). Staff and students at the Department of Energy - Idaho (DOE-ID) and INL are engaged in the development of benchmarks to support ongoing research activities. These benchmarks include reactors or assemblies that support Next Generation Nuclear Plant (NGNP) research, space nuclear Fission Surface Power System (FSPS) design validation, and currently operational facilities in Southeastern Idaho.

  20. Micro-Reactor Physics of MOX-Fueled Core

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, T.

    2001-06-17

    Recently, fuel assemblies of light water reactors have become complicated because of the extension of fuel burnup and the use of high-enriched Gd and mixed-oxide (MOX) fuel, etc. In conventional assembly calculations, the detailed flux distribution, spectrum distribution, and space dependence of self-shielding within a fuel pellet are not directly taken into account. The experimental and theoretical study of investigating these microscopic properties is named micro-reactor physics. The purpose of this work is to show the importance of micro-reactor physics in the analysis of MOX fuel assemblies. Several authors have done related studies; however, their studies are limited to fuel pin cells, and they are never mentioned with regard to burnup effect, which is important for actual core design.