WorldWideScience

Sample records for reactor experiment loop

  1. Reactor design, cold-model experiment and CFD modeling for chemical looping combustion

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Shaohua; Ma, Jinchen; Hu, Xintao; Zhao, Haibo; Wang, Baowen; Zheng, Chuguang [Huazhong Univ. of Science and Technology, Wuhan (China). State Key Lab. of Coal Combustion

    2013-07-01

    Chemical looping combustion (CLC) is an efficient, clean and cheap technology for CO{sub 2} capture, and an interconnected fluidized bed is more appropriate solution for CLC. This paper aims to design a reactor system for CLC, carry out cold-model experiment of the system, and model fuel reactor using commercial CFD software. As for the CLC system, the air reactor (AR) is designed as a fast fluidized bed while the fuel reactor (FR) is a bubbling bed; a cyclone is used for solid separation of the AR exit flow. The AR and FR are separated by two U-type loop seals to remain gas sealed. Considered the chemical kinetics of oxygen carrier, fluid dynamics, pressure balance and mass balance of the system simultaneously, some key design parameters of a CH{sub 4}-fueled and Fe{sub 2}O{sub 3}/Al{sub 2}O{sub 3}-based CLC reactor (thermal power of 50 kWth) are determined, including key geometric parameters (reactor cross-sectional area and reactor height) and operation parameters (bed material quantity, solid circulation rate, apparent gas velocity of each reactor). A cold-model bench having same geometric parameters with its prototype is built up to study the effects of various operation conditions (including gas velocity in the reactors and loop seals, and bed material height, etc.) on the solids circulation rate, gas leakage, and pressure balance. It is witnessed the cold-model system is able to meet special requirements for CLC system such as gas sealing between AR and FR, the circulation rate and particles residence time. Furthermore, the thermal FR reactor with oxygen carrier of Fe{sub 2}O{sub 3}/Al{sub 2}O{sub 3} and fuel of CH{sub 4} is simulated by commercial CFD solver FLUENT. It is found that for the design case the combustion efficiency of CH{sub 4} reaches 88.2%. A few part of methane is unburned due to fast, large bubbles rising through the reactor.

  2. Selective purge for hydrogenation reactor recycle loop

    Science.gov (United States)

    Baker, Richard W.; Lokhandwala, Kaaeid A.

    2001-01-01

    Processes and apparatus for providing improved contaminant removal and hydrogen recovery in hydrogenation reactors, particularly in refineries and petrochemical plants. The improved contaminant removal is achieved by selective purging, by passing gases in the hydrogenation reactor recycle loop or purge stream across membranes selective in favor of the contaminant over hydrogen.

  3. Physical aspects of liquid-impelled loop reactors

    NARCIS (Netherlands)

    Sonsbeek, van H.

    1992-01-01

    The liquid-impelled loop reactor (LLR) is a reactor that consists of two parts : the main tube and the circulation tube. Both parts are in open connection at the bottom and at the top. The reactor is filled with a liquid phase: the continuous phase. Another liquid phase is injected in the

  4. CaMn0.875Ti0.125O3 as oxygen carrier for chemical-looping combustion with oxygen uncoupling (CLOU)—Experiments in a continuously operating fluidized-bed reactor system

    KAUST Repository

    Rydén, Magnus

    2011-03-01

    Particles of the perovskite material CaMn0.875Ti0.125O3 has been examined as oxygen carrier for chemical-looping with oxygen uncoupling, and for chemical-looping combustion of natural gas, by 70h of experiments in a circulating fluidized-bed reactor system. For the oxygen uncoupling experiments, it was found that the particles released O2 in gas phase at temperatures above 720°C when the fuel reactor was fluidized with CO2. The effect increased with increased temperature, and with the O2 partial pressure in the air reactor. At 950°C, the O2 concentration in the outlet from the fuel reactor was in the order of 4.0vol%, if the particles were oxidized in air. For the chemical-looping combustion experiments the combustion efficiency with standard process parameters was in the order of 95% at 950°C, using 1000kg oxygen carrier per MW natural gas, of which about 30% was located in the fuel reactor. Reducing the fuel flow so that 1900kg oxygen carrier per MW natural gas was used improved the combustion efficiency to roughly 99.8%. The particles retained their physical properties, reactivity with CH4 and ability to release gas-phase O2 reasonably well throughout the testing period and there were no problems with the fluidization or formation of solid carbon in the reactor. X-ray diffraction showed that the particles underwent changes in their phase composition though. © 2010 Elsevier Ltd.

  5. Packed Bed Reactor Technology for Chemical-Looping Combustion

    NARCIS (Netherlands)

    Noorman, S.; van Sint Annaland, M.; Kuipers, J.A.M.

    2007-01-01

    Chemical-looping combustion (CLC) has emerged as an alternative for conventional power production processes to intrinsically integrate power production and CO2 capture. In this work a new reactor concept for CLC is proposed, based on dynamically operated packed bed reactors. With analytical

  6. Innovative hybrid pile oscillator technique in the Minerve reactor: open loop vs. closed loop

    Directory of Open Access Journals (Sweden)

    Geslot Benoit

    2018-01-01

    Full Text Available Pile oscillator techniques are powerful methods to measure small reactivity worth of isotopes of interest for nuclear data improvement. This kind of experiments has long been implemented in the Mineve experimental reactor, operated by CEA Cadarache. A hybrid technique, mixing reactivity worth estimation and measurement of small changes around test samples is presented here. It was made possible after the development of high sensitivity miniature fission chambers introduced next to the irradiation channel. A test campaign, called MAESTRO-SL, took place in 2015. Its objective was to assess the feasibility of the hybrid method and investigate the possibility to separate mixed neutron effects, such as fission/capture or scattering/capture. Experimental results are presented and discussed in this paper, which focus on comparing two measurements setups, one using a power control system (closed loop and another one where the power is free to drift (open loop. First, it is demonstrated that open loop is equivalent to closed loop. Uncertainty management and methods reproducibility are discussed. Second, results show that measuring the flux depression around oscillated samples provides valuable information regarding partial neutron cross sections. The technique is found to be very sensitive to the capture cross section at the expense of scattering, making it very useful to measure small capture effects of highly scattering samples.

  7. Innovative hybrid pile oscillator technique in the Minerve reactor: open loop vs. closed loop

    Science.gov (United States)

    Geslot, Benoit; Gruel, Adrien; Bréaud, Stéphane; Leconte, Pierre; Blaise, Patrick

    2018-01-01

    Pile oscillator techniques are powerful methods to measure small reactivity worth of isotopes of interest for nuclear data improvement. This kind of experiments has long been implemented in the Mineve experimental reactor, operated by CEA Cadarache. A hybrid technique, mixing reactivity worth estimation and measurement of small changes around test samples is presented here. It was made possible after the development of high sensitivity miniature fission chambers introduced next to the irradiation channel. A test campaign, called MAESTRO-SL, took place in 2015. Its objective was to assess the feasibility of the hybrid method and investigate the possibility to separate mixed neutron effects, such as fission/capture or scattering/capture. Experimental results are presented and discussed in this paper, which focus on comparing two measurements setups, one using a power control system (closed loop) and another one where the power is free to drift (open loop). First, it is demonstrated that open loop is equivalent to closed loop. Uncertainty management and methods reproducibility are discussed. Second, results show that measuring the flux depression around oscillated samples provides valuable information regarding partial neutron cross sections. The technique is found to be very sensitive to the capture cross section at the expense of scattering, making it very useful to measure small capture effects of highly scattering samples.

  8. Primary loop simulation of the SP-100 space nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Borges, Eduardo M.; Braz Filho, Francisco A.; Guimaraes, Lamartine N.F., E-mail: eduardo@ieav.cta.b, E-mail: fbraz@ieav.cta.b, E-mail: guimarae@ieav.cta.b [Instituto de Estudos Avancados (IEAv/DCTA) Sao Jose dos Campos, SP (Brazil)

    2011-07-01

    Between 1983 and 1992 the SP-100 space nuclear reactor development project for electric power generation in a range of 100 to 1000 kWh was conducted in the USA. Several configurations were studied to satisfy different mission objectives and power systems. In this reactor the heat is generated in a compact core and refrigerated by liquid lithium, the primary loops flow are controlled by thermoelectric electromagnetic pumps (EMTE), and thermoelectric converters produce direct current energy. To define the system operation point for an operating nominal power, it is necessary the simulation of the thermal-hydraulic components of the space nuclear reactor. In this paper the BEMTE-3 computer code is used to EMTE pump design performance evaluation to a thermalhydraulic primary loop configuration, and comparison of the system operation points of SP-100 reactor to two thermal powers, with satisfactory results. (author)

  9. The simulation of thermohydraulic phenomena in a pressurized water reactor primary loop

    Energy Technology Data Exchange (ETDEWEB)

    Popp, M

    1987-01-01

    Several important fluid flow and heat transfer phenomena essential to nuclear power reactor safety were investigated. Scaling and modeling laws for pressurized water reactors are reviewed and a new scaling approach focusing on the overall loop behavior is presented. Scaling criteria for one- and two-phase natural circulation are developed, as well as a simplified model describing the first phase of a small break loss of coolant accident. Reactor vessel vent valve effects are included in the analysis of steady one-phase natural circulation flow. Two new dimensionless numbers, which uniquely describe one-phase flow in natural circulation loops, were deduced and are discussed. A scaled model of the primary loop of a typical Babcock and Wilcox reactor was designed, built, and tested. The particular prototype modeled was the TMI unit 2 reactor. The electrically heated, stainless steel model operates at a maximum pressure of 300 psig and has a maximum heat input of 188 kW. The model is about 4 times smaller in height than the prototype reactor, with a nominal volume scale of 1:500. Experiments were conducted establishing subcooled natural circulation in the model loop. Both steady flow and power transients were investigated.

  10. DESIGN RULE FOR PUMPING DEVICES IN A LOOP POLYMERIZATION REACTOR

    NARCIS (Netherlands)

    MOZARYN, W; JANSSEN, LPBM; VANDIERENDONCK, LL

    The mutual influence on the pumping capacity of independently driven pumping-stirring devices in a polymerization loop reactor is presented. A simple method for the calculation of the minimum rotational speed for an axial pumping stirrer in relation to the pumping action of a helical ribbon is

  11. An investigation of tritium transfer in reactor loops

    Science.gov (United States)

    Ilyasova, O. H.; Mosunova, N. A.

    2017-09-01

    The work is devoted to the important task of the numerical simulation and analysis of the tritium behaviour in the reactor loops. The simulation was carried out by HYDRA-IBRAE/LM code, which is being developed in Nuclear safety institute of the Russian Academy of Sciences. The code is intended for modeling of the liquid metal flow (sodium, lead and lead-bismuth) on the base of non-homogeneous and non-equilibrium two-fluid model. In order to simulate tritium transfer in the code, the special module has been developed. Module includes the models describing the main phenomena of tritium behaviour in reactor loops: transfer, permeation, leakage, etc. Because of shortage of the experimental data, a lot of analytical tests and comparative calculations were considered. Some of them are presented in this work. The comparison of estimation results and experimental and analytical data demonstrate not only qualitative but also good quantitative agreement. It is possible to confirm that HYDRA-IBRAE/LM code allows modeling tritium transfer in reactor loops.

  12. U-loop reactor modelling for optimization. Part 2: Mass transfer

    Energy Technology Data Exchange (ETDEWEB)

    Roende Andersen, B.; Bagterp Joergensen, J.; Bay Joergensen, S.

    2005-09-01

    The present report is a continuation of Andersen et al. [2005] where the heat loss from the U-loop reactor pilot plant Gas estimated for operation without fermentation. The proposed model described the dependency on circulation velocity, air flow rate and reactor temperature. The model could not describe variations in heat exchanger flow rate. Since the heat exchanger flow rate is actuator for the reactor temperature control during cooling, this dependency is important. The most likely explanation of the mismatch between the model and the experimental values are inadequate calibration of the flow and temperature sensors. The energy balance for the reactor cannot be included in the present model before these sensors have been calibrated. This calibration still awaits. Initially separate mass transfer experiments were planned and a master project is devoted to mass transfer in autumn 2005. Until these experiments are performed, the mass transfer is attempted estimated from a fermentation experiment, but the experiment is not designed for estimating mass transfer. The fermentation experiment studied in this report was delayed till July 2005, consequently only a crude model has been set up. The most reasonable description of the U-loop reactor is obtained by partial differential equations due to the distributed nature of the reactor. However given the limited time that only allowed a crude model to be formulated with a number of well mixed tanks in series. Furthermore the significant pressure variations through the reactor have been ignored. This means that only a rough average value of the mass transfer coefficient can be estimated. This value can be compared at different operating conditions for this U-loop reactor but the value cannot be sensibly compared to other reactor types. Anyway, the crude model does indicate some of the process limitations. The experiment was planned to run under methanol limitation with excess of oxygen and nitrogen. The results from the

  13. 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.

  14. Chemical looping combustion in a rotating bed reactor--finding optimal process conditions for prototype reactor.

    Science.gov (United States)

    Håkonsen, Silje Fosse; Blom, Richard

    2011-11-15

    A lab-scale rotating bed reactor for chemical looping combustion has been designed, constructed, and tested using a CuO/Al(2)O(3) oxygen carrier and methane as fuel. Process parameters such as bed rotating frequency, gas flows, and reactor temperature have been varied to find optimal performance of the prototype reactor. Around 90% CH(4) conversion and >90% CO(2) capture efficiency based on converted methane have been obtained. Stable operation has been accomplished over several hours, and also--stable operation can be regained after intentionally running into unstable conditions. Relatively high gas velocities are used to avoid fully reduced oxygen carrier in part of the bed. Potential CO(2) purity obtained is in the range 30 to 65%--mostly due to air slippage from the air sector--which seems to be the major drawback of the prototype reactor design. Considering the prototype nature of the first version of the rotating reactor setup, it is believed that significant improvements can be made to further avoid gas mixing in future modified and up-scaled reactor versions.

  15. Chemical looping reactor system design double loop circulating fluidized bed (DLCFB)

    Energy Technology Data Exchange (ETDEWEB)

    Bischi, Aldo

    2012-05-15

    Chemical looping combustion (CLC) is continuously gaining more importance among the carbon capture and storage (CCS) technologies. It is an unmixed combustion process which takes place in two steps. An effective way to realize CLC is to use two interconnected fluidized beds and a metallic powder circulating among them, acting as oxygen carrier. The metallic powder oxidizes at high temperature in one of the two reactors, the air reactor (AR). It reacts in a highly exothermic reaction with the oxygen of the injected fluidising air. Afterwards the particles are sent to the other reactor where the fuel is injected, the fuel reactor (FR). There, they transport heat and oxygen necessary for the reaction with the injected fuel to take place. At high temperatures, the particle's oxygen reacts with the fuel producing Co2 and steam, and the particles are ready to start the loop again. The overall reaction, the sum of the enthalpy changes of the oxygen carrier oxidation and reduction reactions, is the same as for the conventional combustion. Two are the key features, which make CLC promising both for costs and capture efficiency. First, the high inherent irreversibility of the conventional combustion is avoided because the energy is utilized stepwise. Second, the Co2 is intrinsically separated within the process; so there is in principle no need either of extra carbon capture devices or of expensive air separation units to produce oxygen for oxy-combustion. A lot of effort is taking place worldwide on the development of new chemical looping oxygen carrier particles, reactor systems and processes. The current work is focused on the reactor system: a new design is presented, for the construction of an atmospheric 150kWth prototype working with gaseous fuel and possibly with inexpensive oxygen carriers derived from industrial by-products or natural minerals. It consists of two circulating fluidized beds capable to operate in fast fluidization regime; this will increase the

  16. CO{sub 2} absorption characteristics of a jet loop reactor with a two-fluid swirl nozzle in an alkaline solution

    Energy Technology Data Exchange (ETDEWEB)

    Cha, Go-Eun; Lim, Jun-Heok; Lee, Tae-Yoon; Lee, Jea-Keun [Pukyong National University, Busan (Korea, Republic of); Sung, Ho-Jin [Institute for Advanced Engineering, Yongin (Korea, Republic of)

    2014-04-15

    To investigate the performance of a jet loop reactor with the two-fluid swirl nozzle (TSN), CO{sub 2} absorption experiments in an alkaline solution were performed. The experimental results obtained in the reactor were compared with those in a jet loop reactor with the two-fluid conventional nozzle (TCN). The neutralization time of alkaline solution and the CO{sub 2} removal efficiency were used as the indices for a comparison of the reactor performance. Due to the swirling flow, the neutralization times of alkaline solutions by CO{sub 2} in the reactor with the TSN were shortened compared with those in the reactor with the TCN. Also, the instantaneous and/or overall CO{sub 2} removal efficiencies in the reactor with the TSN were higher than those in the reactor with the TCN at the same liquid circulation flow rate.

  17. Mono fermentation of grass silage by means of loop reactors.

    Science.gov (United States)

    Koch, Konrad; Wichern, Marc; Lübken, Manfred; Horn, Harald

    2009-12-01

    A loop reactor was operated for mono fermentation of grass silage without manure addition under mesophilic conditions (38 degrees C). An averaged specific biogas production of 0.50 m(N)(3) per kg volatile solids (VS) with a methane concentration of 52% at an organic loading rate of up to 3.5 kg(VS)/(m(3) d) was obtained. The retention time varied from 440 days at 1.0 kg(VS)/(m(3) d) to 50 days at 3.5 kg(VS)/(m(3) d). The degradation level was more than 60% based on VS and 75% based on COD. The first-order hydrolysis rate constant of the process was estimated to be 0.6 d(-1). Despite the relative high ammonium concentration of up to 4 g/l, the system worked stable for an operation period of 310 days. In particular the TS content in the fermenter was found to be a key parameter and should not exceed 12% in order to avoid instabilities.

  18. Studies on the closed-loop digital control of multi-modular reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bernard, J.A. (Massachusetts Inst. of Tech., Cambridge, MA (United States). Nuclear Reactor Lab.); Henry, A.F.; Lanning, D.D.; Meyer, J.E. (Massachusetts Inst. of Tech., Cambridge, MA (United States). Dept. of Nuclear Engineering)

    1992-11-01

    This report describes the theoretical development and the evaluation via both experiment and simulation of digital methods for the closed-loop control of power, temperature, and steam generator level in multi-modular reactors. The major conclusion of the research reported here is that the technology is currently available to automate many aspects of the operation of multi-modular plants. This will in turn minimize the number of required personnel and thus contain both operating and personnel costs, allow each module to be operated at a different power level thereby staggering the times at which refuelings would be needed, and maintain the competitiveness of US industry relative to foreign vendors who are developing and applying advanced control concepts. The technology described in this report is appropriate to the proposed multi-modular reactor designs and to present-generation pressurized water reactors. Its extension to boiling water reactors is possible provided that the commitment is made to create a real-time model of a BWR. The work reported here was performed by the Massachusetts Institute of Technology (MIT) under contract to the Oak Ridge National Laboratory (ORNL) and to the United States Department of Energy (Division of Industry and University Programs, Contract No. DE-FG07-90ER12930.)

  19. Studies on the closed-loop digital control of multi-modular reactors. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Bernard, J.A. [Massachusetts Inst. of Tech., Cambridge, MA (United States). Nuclear Reactor Lab.; Henry, A.F.; Lanning, D.D.; Meyer, J.E. [Massachusetts Inst. of Tech., Cambridge, MA (United States). Dept. of Nuclear Engineering

    1992-11-01

    This report describes the theoretical development and the evaluation via both experiment and simulation of digital methods for the closed-loop control of power, temperature, and steam generator level in multi-modular reactors. The major conclusion of the research reported here is that the technology is currently available to automate many aspects of the operation of multi-modular plants. This will in turn minimize the number of required personnel and thus contain both operating and personnel costs, allow each module to be operated at a different power level thereby staggering the times at which refuelings would be needed, and maintain the competitiveness of US industry relative to foreign vendors who are developing and applying advanced control concepts. The technology described in this report is appropriate to the proposed multi-modular reactor designs and to present-generation pressurized water reactors. Its extension to boiling water reactors is possible provided that the commitment is made to create a real-time model of a BWR. The work reported here was performed by the Massachusetts Institute of Technology (MIT) under contract to the Oak Ridge National Laboratory (ORNL) and to the United States Department of Energy (Division of Industry and University Programs, Contract No. DE-FG07-90ER12930.)

  20. Nuclear materials testing in the loops of the NRU research reactor using material test bundles

    Energy Technology Data Exchange (ETDEWEB)

    Leung, T.C.; Walters, L. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

    2014-07-01

    The NRU research reactor has been used to obtain data to understand and quantify the effects of irradiation on nuclear reactor components through their in-service lives and to develop improved designs and components. Apart from the Mark-4 and Mark-7 fast neutron rod material testing facilities in NRU, the high-pressure/high-temperature experimental loops provide an environment similar to the CANDU reactor core, where test materials are subjected to simulated power reactor conditions. Nuclear materials are tested in the loops using Material Test Bundles (MTB). This paper describes how the MTB is designed to operate in the NRU loops. It also describes the physics calculation of the 89-energy-group neutron spectrum in the MTB and its comparison with the spectrum in CANDU power reactors. The predictions of spectral effects on nuclear material behaviour, such as material damage and helium generation are summarized. (author)

  1. Irradiation Testing Vehicles for Fast Reactors from Open Test Assemblies to Closed Loops

    Energy Technology Data Exchange (ETDEWEB)

    Sienicki, James J. [Argonne National Lab. (ANL), Argonne, IL (United States); Grandy, Christopher [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-12-15

    A review of irradiation testing vehicle approaches and designs that have been incorporated into past Sodium-Cooled Fast Reactors (SFRs) or envisioned for incorporation has been carried out. The objective is to understand the essential features of the approaches and designs so that they can inform test vehicle designs for a future U.S. Fast Test Reactor. Fast test reactor designs examined include EBR-II, FFTF, JOYO, BOR-60, PHÉNIX, JHR, and MBIR. Previous designers exhibited great ingenuity in overcoming design and operational challenges especially when the original reactor plant’s mission changed to an irradiation testing mission as in the EBRII reactor plant. The various irradiation testing vehicles can be categorized as: Uninstrumented open assemblies that fit into core locations; Instrumented open test assemblies that fit into special core locations; Self-contained closed loops; and External closed loops. A special emphasis is devoted to closed loops as they are regarded as a very desirable feature of a future U.S. Fast Test Reactor. Closed loops are an important technology for irradiation of fuels and materials in separate controlled environments. The impact of closed loops on the design of fast reactors is also discussed in this report.

  2. A novel reactor configuration for packed bed chemical-looping combustion of syngas

    NARCIS (Netherlands)

    Hamers, H.P.; Gallucci, F.; Cobden, P.D.; Kimball, E.; Sint Annaland, M. van

    2013-01-01

    This study reports on the application of chemical looping combustion (CLC) in pressurized packed bed reactors using syngas as a fuel. High pressure operation of CLC in packed bed has a different set of challenges in terms of material properties, cycle and reactor design compared to fluidized bed

  3. Simulation, Control and Optimization of Single Cell Protein Production in a U-Loop Reactor

    DEFF Research Database (Denmark)

    Engoulevent, Franck Guillaume; Jørgensen, John Bagterp

    2012-01-01

    potential for alternative sources of protein production. Single cell protein (SCP) is protein produced by growth of micro organisms. Among these micro organisms, Methylococcus Capsulatus is particular interesting as it can grow on either methane or methanol and contains 70% protein. The U-Loop reactor...... is particular useful for production of SCP by M. Capsulatus as it has good gas-liquid mass transfer capabilities and also the capability to remove the signicant amount of heat developed by the reaction. In this paper we describe an implementation of a model to simulate SCP production in the U-Loop reactor. We...... report simulation results. In addition we design and compare dierent regulatory control systems for regulation of SCP production in the U-Loop reactor. The purpose of the regulatory control systems is to keep the process at a steady state and to reject disturbances. We design and implement such control...

  4. Rotary Bed Reactor for Chemical-Looping Combustion with Carbon Capture. Part 1: Reactor Design and Model Development

    KAUST Repository

    Zhao, Zhenlong

    2013-01-17

    Chemical-looping combustion (CLC) is a novel and promising technology for power generation with inherent CO2 capture. Currently, almost all of the research has been focused on developing CLC-based interconnected fluidized-bed reactors. In this two-part series, a new rotary reactor concept for gas-fueled CLC is proposed and analyzed. In part 1, the detailed configuration of the rotary reactor is described. In the reactor, a solid wheel rotates between the fuel and air streams at the reactor inlet and exit. Two purging sectors are used to avoid the mixing between the fuel stream and the air stream. The rotary wheel consists of a large number of channels with copper oxide coated on the inner surface of the channels. The support material is boron nitride, which has high specific heat and thermal conductivity. Gas flows through the reactor at elevated pressure, and it is heated to a high temperature by fuel combustion. Typical design parameters for a thermal capacity of 1 MW have been proposed, and a simplified model is developed to predict the performances of the reactor. The potential drawbacks of the rotary reactor are also discussed. © 2012 American Chemical Society.

  5. Ageing management experience at NUR Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Melllal, Sabrina; Rezig, Mohamed; Zamoun, Rachid; Ameur, Azeddin [Nuclear Research Center of Draria, Algiers (Algeria)

    2013-07-01

    NUR is a 1 MW, open pool reactor moderated and cooled by light water. It was commissioned in 1989. NUR is used for education and training in Nuclear Engineering and related topics for COMENA and National Scientific Community. It is also used to perform R and D works and services at national and regional levels. In this presentation, we describe the methodology and the main development activities related to the ageing management at NUR reactor. These activities include inspection actions and development actions to introduce modifications, to solve obsolescence issues in view to implement the required preventive and curative maintenance programs and to improve the performances of the installation. These actions involved mainly the Operation Assistance System of the Reactor (OAS), the secondary cooling loop, the cooling tower. A new OAS using a new technology and having more possibilities than the older one was introduced in the control system of the reactor. The OAS hardware structure, software structure and the main functions performed are presented. The second loop is entirely refurbished. Two new cooling towers are installed and connected to the main heat exchanger with new piping and valves. The architecture of this new installation is described and the performance assessed. Other actions which involve auxiliary systems like emergency electrical system, air pneumatic system and automatic fire extinguishing are presented.

  6. Defluoridation of drinking water by electrocoagulation/electroflotation in a stirred tank reactor with a comparative performance to an external-loop airlift reactor

    OpenAIRE

    Essadki, Abdel Hafid; Gourich, Bouchaib; Vial, Christophe; Delmas, Henri; Bennajah, Mounir

    2009-01-01

    Defluoridation using batch electrocoagulation/electroflotation (EC/EF) was carried out in two reactors for comparison purpose: a stirred tank reactor (STR) close to a conventional EC cell and an external-loop airlift reactor (ELAR) that was recently described as an innovative reactor for EC. The respective influences of current density, initial concentration and initial pH on the efficiency of defluoridation were investigated. The same trends were observed in both reactors, but the efficiency...

  7. Double Chooz and Reactor Theta13 Experiments

    CERN Document Server

    ,

    2016-01-01

    This is a contribution paper from the Double Chooz experiment to the special issue of NPB on neutrino oscillations. The physics and history of the reactor theta13 experiments, as well as Double Chooz experiment and its neutrino oscillation analyses are reviewed.

  8. Results of the Nucifer reactor neutrino experiment

    Energy Technology Data Exchange (ETDEWEB)

    Buck, Christian; Lindner, Manfred [MPIK Heidelberg (Germany)

    2016-07-01

    Nuclear reactors are a strong and pure source of electron antineutrinos. With neutrino experiments close to compact reactor cores new insights into neutrino properties and reactor physics can be obtained. The Nucifer experiment is one of the pioneers in this class of very short baseline projects. Its detector to reactor distance is only about 7 m. The data obtained in the last years allowed to estimate the plutonium concentration in the reactor core by the neutrino flux measurement. This is of interest for safeguard applications and non proliferation efforts. The antineutrinos in Nucifer are detected via the inverse beta decay on free protons. Those Hydrogen nuclei are provided by 850 liters of organic liquid scintillator. For higher detection efficiency and background reduction the liquid is loaded with Gadolinium. Despite all shielding efforts and veto systems the background induced by the reactor activity and cosmogenic particles is still the main challenge in the experiment. The principle of the Nucifer detector is similar to the needs of upcoming experiments searching for sterile neutrinos. Therefore, the Nucifer results are also valuable input for the understanding and optimization of those next generation projects. The observation of sterile neutrinos would imply new physics beyond the standard model.

  9. Economic Optimizing Control for Single-Cell Protein Production in a U-Loop Reactor

    DEFF Research Database (Denmark)

    Drejer, André; Ritschel, Tobias Kasper Skovborg; Jørgensen, Sten Bay

    2017-01-01

    The production of single-cell protein (SCP) in a U-loop reactor by a methanotroph is a cost efficient sustainable alternative to protein from fish meal obtained by over-fishing the oceans. SCP serves as animal feed. In this paper, we present a mathematical model that describes the dynamics of SCP...

  10. Feasibility of simultaneous nitrification and denitrification in a pilot-scale airlift-loop reactor

    NARCIS (Netherlands)

    Hao, X.; Nieuwstad, T.J.

    1993-01-01

    Airlift-loop reactors have become competitive in treating municipal wastewater. Many studies have verified that high COD conversion efficiencies can be reached. When simultaneous nitrification and denitrification could be realised to the same degree as is possible in low loaded conventional

  11. The N-Reactor fuel testing program in the KER loops

    Energy Technology Data Exchange (ETDEWEB)

    Evans, T.W.

    1963-04-01

    The KER loops are four pressurized water-cooled in-reactor testing loops in which fuel, coolant and corrosion tests are being conducted for the N Reactor program. Two of the loops, KER-3 and KER-4, are equipped with N-size Zr-2 process tubes and have sufficient pumping and heat dissipation capacity to permit testing of 14-element charges of N fuel. The other two loops, HER-1 and KER-2, are fitted with Zr-2 process tubes of 2.1 in. inner diameter (vs 2.7 in. for N). The small loops can be used to test charges of N inner tubes or other elements sized at about 1.8 in. outer diameter or less. An inert sleeve must be used with N inner tubes to assure adequate coolant flow over both surfaces of the elements. The KE reactor active zone is 28 in. long com pared to 35 in. for N so the length of test charges is somewhat shorter than actual N charges will be. The purpose of this report is to review the testing environment, to discuss the testing program and to present the schedule as presently seen.

  12. Detecting Dark Photons with Reactor Neutrino Experiments

    Science.gov (United States)

    Park, H. K.

    2017-08-01

    We propose to search for light U (1 ) dark photons, A', produced via kinetically mixing with ordinary photons via the Compton-like process, γ e-→A'e-, in a nuclear reactor and detected by their interactions with the material in the active volumes of reactor neutrino experiments. We derive 95% confidence-level upper limits on ɛ , the A'-γ mixing parameter, ɛ , for dark-photon masses below 1 MeV of ɛ mass dark photons.

  13. Steam drum level control studies of a natural circulation multi loop reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, Rajesh; Contractor, A.D.; Srivastava, Abhishek; Lele, H.G. [Bhabha Atomic Research Centre, Trombay, Mumbai (India). Reactor Safety Div.; Vaze, K.K. [Bhabha Atomic Research Centre, Trombay, Mumbai (India). Reactor Design and Development Group

    2013-12-15

    The proposed heavy water moderated and light water cooled pressure tube type boiling water reactor works on natural circulation at all power levels. It has parallel inter-connected loops with 452 boiling channels in the main heat transport system configuration. These multiple (four) interconnected loops influence the steam drum level control adversely through the common reactor inlet header. Alternate design studies made earlier for efficient control of SD levels have shown favorable results. This has lead to explore further the present scheme with the compartmentalization of CRIH into four compartments catering to four loops separately. The conventional 3-element level control has been found to be working satisfactorily. The interconnections between ECCS header and inlet header compartments have also increased the safety margin for various LOCA and design basis events. The paper deals with the SD level control aspects for this novel MHT configuration which has been analyzed for various PIEs (Postulated Initiating Events) and found to be satisfactory. (orig.)

  14. A reduced fidelity model for the rotary chemical looping combustion reactor

    KAUST Repository

    Iloeje, Chukwunwike O.

    2017-01-11

    The rotary chemical looping combustion reactor has great potential for efficient integration with CO capture-enabled energy conversion systems. In earlier studies, we described a one-dimensional rotary reactor model, and used it to demonstrate the feasibility of continuous reactor operation. Though this detailed model provides a high resolution representation of the rotary reactor performance, it is too computationally expensive for studies that require multiple model evaluations. Specifically, it is not ideal for system-level studies where the reactor is a single component in an energy conversion system. In this study, we present a reduced fidelity model (RFM) of the rotary reactor that reduces computational cost and determines an optimal combination of variables that satisfy reactor design requirements. Simulation results for copper, nickel and iron-based oxygen carriers show a four-order of magnitude reduction in simulation time, and reasonable prediction accuracy. Deviations from the detailed reference model predictions range from 3% to 20%, depending on oxygen carrier type and operating conditions. This study also demonstrates how the reduced model can be modified to deal with both optimization and design oriented problems. A parametric study using the reduced model is then applied to analyze the sensitivity of the optimal reactor design to changes in selected operating and kinetic parameters. These studies show that temperature and activation energy have a greater impact on optimal geometry than parameters like pressure or feed fuel fraction for the selected oxygen carrier materials.

  15. Strategic need for a multi-purpose thermal hydraulic loop for support of advanced reactor technologies

    Energy Technology Data Exchange (ETDEWEB)

    O' Brien, James E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sabharwall, Piyush [Idaho National Lab. (INL), Idaho Falls, ID (United States); Yoon, Su -Jong [Idaho National Lab. (INL), Idaho Falls, ID (United States); Housley, Gregory K. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-09-01

    This report presents a conceptual design for a new high-temperature multi fluid, multi loop test facility for the INL to support thermal hydraulic, materials, and thermal energy storage research for nuclear and nuclear-hybrid applications. In its initial configuration, the facility will include 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 with this facility include the characterization and performance evaluation of candidate compact heat exchangers such as printed circuit heat exchangers (PCHEs) at prototypical operating conditions, flow and heat transfer issues related to core thermal hydraulics in advanced helium-cooled and salt-cooled reactors, and evaluation of corrosion behavior of new cladding materials and accident-tolerant fuels for LWRs at prototypical conditions. Based on its relevance to advanced reactor systems, the new facility has been named the Advanced Reactor Technology Integral System Test (ARTIST) facility. Research performed in this facility will advance the state of the art and technology readiness level of high temperature intermediate heat exchangers (IHXs) for nuclear applications while establishing the INL as a center of excellence for the development and certification of this technology. The thermal energy storage capability will support research and demonstration activities related to process heat delivery for a variety of hybrid energy systems and grid stabilization strategies. Experimental results obtained from this research will assist in development of reliable predictive models for thermal hydraulic design and safety codes over the range of expected advanced reactor operating conditions. Proposed/existing IHX heat transfer and friction correlations and criteria will be assessed with information on materials compatibility and instrumentation

  16. US graphite reactor D&D experience

    Energy Technology Data Exchange (ETDEWEB)

    Garrett, S.M.K.; Williams, N.C.

    1997-02-01

    This report describes the results of the U.S. Graphite Reactor Experience Task for the Decommissioning Strategy Plan for the Leningrad Nuclear Power Plant (NPP) Unit 1 Study. The work described in this report was performed by the Pacific Northwest National Laboratory (PNNL) for the Department of Energy (DOE).

  17. A novel reactor configuration for packed bed chemical-looping combustion of syngas

    Energy Technology Data Exchange (ETDEWEB)

    Hamers, H.P.; Gallucci, F.; Van Sint Annaland, M. [Multiphase Reactor Group, Chemical Process Intensification, Department of Chemical Engineering and Chemistry, Eindhoven University of Technology, P.O. Box 513, 5600 MB Eindhoven (Netherlands); Cobden, P.D. [Energy Research Centre of the Netherlands (ECN), P.O. Box 1, 1755 ZG Petten (Netherlands); Kimball, E. [TNO Gas Treatment, P.O. Box 6000, 2600 JA Delft (Netherlands)

    2013-08-15

    This study reports on the application of chemical looping combustion (CLC) in pressurized packed bed reactors using syngas as a fuel. High pressure operation of CLC in packed bed has a different set of challenges in terms of material properties, cycle and reactor design compared to fluidized bed operation. However, high pressure operation allows the use of inherently more efficient power cycles than low pressure fluidized bed solutions. This paper quantifies the challenges in high pressure operation and introduces a novel reactor concept with which those challenges can be addressed. Continuous cyclic operation of a packed bed CLC system is simulated in a 1D numerical reactor model. Importantly, it is demonstrated that the temperature profiles that can occur in a packed bed reactor as a result of the different process steps do not accumulate, and have a negligible effect on the overall performance of the system. Moreover, it has been shown that an even higher energy efficiency can be achieved by feeding the syngas from the opposite direction during the reduction step (i.e. countercurrent operation). Unfortunately, in this configuration mode, more severe temperature fluctuations occur in the reactor exhaust, which is disadvantageous for the operation of a downstream gas turbine. Finally, a novel reactor configuration is introduced in which the desired temperature rise for obtained hot pressured air suitable for a gas turbine is obtained by carrying out the process with two packed bed reactor in series (two-stage CLC). This is shown to be a good alternative to the single bed configuration, and has the added advantage of decreasing the demands on both the oxygen carrier and the reactor materials and design specification.

  18. CFD-PBE simulation of gas-phase hydrodynamics in a gas-liquid-solid combined loop reactor

    National Research Council Canada - National Science Library

    Qi Nana Zhang Kai Xu Gang Yang Yongping Zhang Hu

    2013-01-01

    The computational fluid dynamics (CFD)-population balance equations (PBE) coupled model is employed to investigate the hydrodynamics in a gas-slurry internal loop reactor with external slurry circulation...

  19. Neutrino Oscillation Experiments at Nuclear Reactors

    CERN Document Server

    Gratta, Giorgio

    2000-01-01

    In this paper I give an overview of the status of neutrino oscillation experiments performed using nuclear reactors as sources of neutrinos. I review the present generation of experiments (Chooz and Palo Verde) with baselines of about 1 km as well as the next generation that will search for oscillations with a baseline of about 100 km. While the present detectors provide essential input towards the understanding of the atmospheric neutrino anomaly, in the future, the KamLAND reactor experiment represents our best opportunity to study very small mass neutrino mixing in laboratory conditions. In addition KamLAND with its very large fiducial mass and low energy threshold, will also be sensitive to a broad range of different physics.

  20. 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.

  1. Comprehensive Prediction of Thermosyphon Characteristics in Reactor Passive Cooling System Simulation Loop FASSIP-01

    Directory of Open Access Journals (Sweden)

    H. Tjahjono

    2017-12-01

    Full Text Available Passive cooling mechanism for a nuclear reactor has been proven to be very important since the Fukushima Daiichi Reactor accident that was caused by active cooling system malfunction due to total loss of electrical power source. In the Center for Nuclear Reactor Technology and Safety of BATAN, the cooling mechanism was studied by using a natural circulation test loop named FASSIP-01 that applied thermosyphon mechanism of water inside pipes of 1” diameter. This study aimed to analytically predictthe thermal characteristics of the loop including its response time towards steady condition usingthe MATLAB calculation program. This prediction derived the influence of several parameters such as the heat transfer coefficient of the cooler side (h-cooler, the heater power, the elevation difference between the heater and cooler(DZ, and the effects of the insulation thickness of pipe (IT on the flowrate, temperature, and the heat power distribution across all components in the loop. The result showed that byavoiding boiling condition, for transferring the heater power of 1000 W and 2000 W,the needed h-cooler exceeds 200 and 400 W m-2°C-1, respectively. For a h-cooler of 200 W m-2°C-1, the circulation flow rate increased from 0.04 to 0.06 kg/s-1 for heater power increase from 1000 W to 2000 W. Those flow rates were decreased to 0.037 and 0.052 kgs-1 by increasing h-cooler to 1000 W m-2°C-1.The results were in agreement with other studies on rectangular loops in the literature.The time needed to reach 95 % towards steady state was predicted to be more than 13 hours. Reduction of this time to less than five hours was possible by reducing the heater tank volume from 100 L to 30 L or by modifying the starting heater input power.

  2. Novel Magnetically Fluidized Bed Reactor Development for the Looping Process: Coal to Hydrogen Production R&D

    Energy Technology Data Exchange (ETDEWEB)

    Mei, Renwei; Hahn, David; Klausner, James; Petrasch, Jorg; Mehdizadeh, Ayyoub; Allen, Kyle; Rahmatian, Nima; Stehle, Richard; Bobek, Mike; Al-Raqom, Fotouh; Greek, Ben; Li, Like; Chen, Chen; Singh, Abhishek; Takagi, Midori; Barde, Amey; Nili, Saman

    2013-09-30

    The coal to hydrogen project utilizes the iron/iron oxide looping process to produce high purity hydrogen. The input energy for the process is provided by syngas coming from gasification process of coal. The reaction pathways for this process have been studied and favorable conditions for energy efficient operation have been identified. The Magnetically Stabilized Porous Structure (MSPS) is invented. It is fabricated from iron and silica particles and its repeatable high performance has been demonstrated through many experiments under various conditions in thermogravimetric analyzer, a lab-scale reactor, and a large scale reactor. The chemical reaction kinetics for both oxidation and reduction steps has been investigated thoroughly inside MSPS as well as on the surface of very smooth iron rod. Hydrogen, CO, and syngas have been tested individually as the reducing agent in reduction step and their performance is compared. Syngas is found to be the most pragmatic reducing agent for the two-step water splitting process. The transport properties of MSPS including porosity, permeability, and effective thermal conductivity are determined based on high resolution 3D CT x-ray images obtained at Argonne National Laboratory and pore-level simulations using a lattice Boltzmann Equation (LBE)-based mesoscopic model developed during this investigation. The results of those measurements and simulations provide necessary inputs to the development of a reliable volume-averaging-based continuum model that is used to simulate the dynamics of the redox process in MSPS. Extensive efforts have been devoted to simulate the redox process in MSPS by developing a continuum model consist of various modules for conductive and radiative heat transfer, fluid flow, species transport, and reaction kinetics. Both the Lagrangian and Eulerian approaches for species transport of chemically reacting flow in porous media have been investigated and verified numerically. Both approaches lead to correct

  3. 78 FR 58575 - Review of Experiments for Research Reactors

    Science.gov (United States)

    2013-09-24

    ... COMMISSION Review of Experiments for Research Reactors AGENCY: Nuclear Regulatory Commission. ACTION... Guide (RG) 2.4, ``Review of Experiments for Research Reactors.'' The guide is being withdrawn because... Experiments for Research Reactors,'' (ADAMS Accession No. ML003740131) because its guidance no longer provides...

  4. Multiscale Modeling of a Packed Bed Chemical Looping Reforming (PBCLR Reactor

    Directory of Open Access Journals (Sweden)

    Arpit Singhal

    2017-12-01

    Full Text Available Packed bed reactors are broadly used in industry and are under consideration for novel reactor concepts such as packed bed chemical looping reforming (PBCLR. Mass and heat transfer limitations in and around the particles in packed bed reactors strongly affect the behavior of these units. This study employs a multiscale modeling methodology to simulate a PBCLR reactor. Specifically, small-scale particle-resolved direct numerical simulation is utilized to improve large-scale mass transfer models for use in an industrial scale 1D model. Existing intra-particle mass transfer models perform well for simple first order reactions, but several model enhancements were required to model the more complex steam methane reforming reaction system. Three specific aspects required enhanced modeling: the generation of additional gas volume by the reforming reactions, the lack of clear reaction orders in the equilibrium reactions, and the diffusion of multiple reactant species into the particle. Large-scale simulations of the PBCLR reactor with the enhanced 1D model showed that the highly reactive Ni-based catalyst/oxygen carrier employed allows for the use of large particle sizes and high gas flowrates, offering potential for process intensification.

  5. MULTI-LOOP CONTROL DESIGN IN MULTIVARIABLE (2X2 CONTINUOUS STIRRED TANK REACTOR

    Directory of Open Access Journals (Sweden)

    Abdul Wahid

    2015-06-01

    Full Text Available With this study, the design and tuning of multi-loop for multivariable (2x2 CSTR will be made in order to achieve optimum CSTR control performance. This study used Bequette model reactor and MATLAB software and is expected to be able to cope with disturbances in the reactor so that the reactor system is able to stabilize quickly despite the distractions. In this study, the design will be made using multi-loop approach, along with PI controller as the next step. Then, BLT and auto-tune tuning method will be used in PI controller and given disturbances to both of tuning method. The controller performances are then compared. Results of the study are then analyzed for discussions and conclusions. Results from this study have shown that in terms of disturbance rejection, BLT is better than auto-tune based on comparison between both of controller performances. For IAE for the case of temperature, BLT is 30% better than auto-tune, but it is almost the same for the case of concentration. For settling time for the case of concentration, BLT is 30% better than auto-tune, and for the case of temperature, BLT is 18% better than auto-tune. For rise time for the case of concentration and temperature, BLT is 30% better than auto-tune.

  6. Waste Heat Recovery from the Advanced Test Reactor Secondary Coolant Loop

    Energy Technology Data Exchange (ETDEWEB)

    Donna Post Guillen

    2012-11-01

    This study investigated the feasibility of using a waste heat recovery system (WHRS) to recover heat from the Advanced Test Reactor (ATR) secondary coolant system (SCS). This heat would be used to preheat air for space heating of the reactor building, thus reducing energy consumption, carbon footprint, and energy costs. Currently, the waste heat from the reactor is rejected to the atmosphere via a four-cell, induced-draft cooling tower. Potential energy and cost savings are 929 kW and $285K/yr. The WHRS would extract a tertiary coolant stream from the SCS loop and pump it to a new plate and frame heat exchanger, from which the heat would be transferred to a glycol loop for preheating outdoor air supplied to the heating and ventilation system. The use of glycol was proposed to avoid the freezing issues that plagued and ultimately caused the failure of a WHRS installed at the ATR in the 1980s. This study assessed the potential installation of a new WHRS for technical, logistical, and economic feasibility.

  7. 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

  8. A Semi-Batch Reactor Experiment for the Undergraduate Laboratory

    Science.gov (United States)

    Derevjanik, Mario; Badri, Solmaz; Barat, Robert

    2011-01-01

    This experiment and analysis offer an economic yet challenging semi-batch reactor experience. Household bleach is pumped at a controlled rate into a batch reactor containing pharmaceutical hydrogen peroxide solution. Batch temperature, product molecular oxygen, and the overall change in solution conductivity are metered. The reactor simulation…

  9. Design of a rotary reactor for chemical-looping combustion. Part 1: Fundamentals and design methodology

    KAUST Repository

    Zhao, Zhenlong

    2014-04-01

    Chemical-looping combustion (CLC) is a novel and promising option for several applications including carbon capture (CC), fuel reforming, H 2 generation, etc. Previous studies demonstrated the feasibility of performing CLC in a novel rotary design with micro-channel structures. In the reactor, a solid wheel rotates between the fuel and air streams at the reactor inlet, and depleted air and product streams at exit. The rotary wheel consists of a large number of micro-channels with oxygen carriers (OC) coated on the inner surface of the channel walls. In the CC application, the OC oxidizes the fuel while the channel is in the fuel zone to generate undiluted CO2, and is regenerated while the channel is in the air zone. In this two-part series, the effect of the reactor design parameters is evaluated and its performance with different OCs is compared. In Part 1, the design objectives and criteria are specified and the key parameters controlling the reactor performance are identified. The fundamental effects of the OC characteristics, the design parameters, and the operating conditions are studied. The design procedures are presented on the basis of the relative importance of each parameter, enabling a systematic methodology of selecting the design parameters and the operating conditions with different OCs. Part 2 presents the application of the methodology to the designs with the three commonly used OCs, i.e., nickel, copper, and iron, and compares the simulated performances of the designs. © 2013 Elsevier Ltd. All rights reserved.

  10. Determination of the Clean Air Delivery Rate (CADR) of Photocatalytic Oxidation (PCO) Purifiers for Indoor Air Pollutants Using a Closed-Loop Reactor. Part I: Theoretical Considerations.

    Science.gov (United States)

    Dumont, Éric; Héquet, Valérie

    2017-03-06

    This study demonstrated that a laboratory-scale recirculation closed-loop reactor can be an efficient technique for the determination of the Clean Air Delivery Rate (CADR) of PhotoCatalytic Oxidation (PCO) air purification devices. The recirculation closed-loop reactor was modeled by associating equations related to two ideal reactors: one is a perfectly mixed reservoir and the other is a plug flow system corresponding to the PCO device itself. Based on the assumption that the ratio between the residence time in the PCO device and the residence time in the reservoir τP/τR tends to 0, the model highlights that a lab closed-loop reactor can be a suitable technique for the determination of the efficiency of PCO devices. Moreover, if the single-pass removal efficiency is lower than 5% of the treated flow rate, the decrease in the pollutant concentration over time can be characterized by a first-order decay model in which the time constant is proportional to the CADR. The limits of the model are examined and reported in terms of operating conditions (experiment duration, ratio of residence times, and flow rate ranges).

  11. Determination of the Clean Air Delivery Rate (CADR of Photocatalytic Oxidation (PCO Purifiers for Indoor Air Pollutants Using a Closed-Loop Reactor. Part I: Theoretical Considerations

    Directory of Open Access Journals (Sweden)

    Éric Dumont

    2017-03-01

    Full Text Available This study demonstrated that a laboratory-scale recirculation closed-loop reactor can be an efficient technique for the determination of the Clean Air Delivery Rate (CADR of PhotoCatalytic Oxidation (PCO air purification devices. The recirculation closed-loop reactor was modeled by associating equations related to two ideal reactors: one is a perfectly mixed reservoir and the other is a plug flow system corresponding to the PCO device itself. Based on the assumption that the ratio between the residence time in the PCO device and the residence time in the reservoir τP/τR tends to 0, the model highlights that a lab closed-loop reactor can be a suitable technique for the determination of the efficiency of PCO devices. Moreover, if the single-pass removal efficiency is lower than 5% of the treated flow rate, the decrease in the pollutant concentration over time can be characterized by a first-order decay model in which the time constant is proportional to the CADR. The limits of the model are examined and reported in terms of operating conditions (experiment duration, ratio of residence times, and flow rate ranges.

  12. Simulation of the Fuel Reactor of a Coal-Fired Chemical Looping Combustor

    Science.gov (United States)

    Mahalatkar, Kartikeya; O'Brien, Thomas; Huckaby, E. David; Kuhlman, John

    2009-06-01

    Responsible carbon management (CM) will be required for the future utilization of coal for power generation. CO2 separation is the more costly component of CM, not sequestration. Most methods of capture require a costly process of gas separation to obtain a CO2-rich gas stream. However, recently a process termed Chemical Looping Combustion (CLC) has been proposed, in which an oxygen-carrier is used to provide the oxygen for combustion. This process quite naturally generates a separate exhaust gas stream containing mainly H2O and CO2 but requires two reaction vessels, an Air Reactor (AR) and a Fuel Reactor (FR). The carrier (M for metal, the usual carrier) is oxidized in the AR. This highly exothermic process provides heat for power generation. The oxidized carrier (MO) is separated from this hot, vitiated air stream and transported to the FR where it oxidizes the hydrocarbon fuel, yielding an exhaust gas stream of mainly H2O and CO2. This process is usually slightly endothermic so that the carrier must also transport the necessary heat of reaction. The reduced carrier (M) is then returned to the air reactor for regeneration, hence the term "looping." The net chemical reaction and energy release is identical to that of conventional combustion of the fuel. However, CO2 separation is easily achieved, the only operational penalty being the slight pressure losses required to circulate the carrier. CLC requires many unit operations involving gas-solid or granular flow. To utilize coal in the fuel reactor, in either a moving bed or bubbling fluidized bed, the granular flow is especially critical. The solid coal fuel must be heated by the recycled metal oxide, driving off moisture and volatile material. The remaining char must be gasified by H2O (or CO2), which is recycled from the product stream. The gaseous product of these reactions must then contact the MO before leaving the bed to obtain complete conversion to H2O and CO2. Further, the reduced M particles must be

  13. Hydrodynamics of an external-loop airlift reactor with inserted membrane

    Directory of Open Access Journals (Sweden)

    Predrag S. Kojić

    Full Text Available Abstract The objective of this study was to investigate the hydrodynamics of an external-loop airlift membrane reactor (ELAMR. The ELAMR was operated in two modes: without (mode A and with bubbles in the downcomer (mode B, depending on the liquid level in the gas separator. The influence of the gas distributor’s geometry and various alcohol solutions on the hydrodynamics of the ELAMR was studied. Results for the gas holdup and the downcomer liquid velocity are commented with respect to an external-loop airlift reactor of the same geometry but without the membrane in the downcomer (ELAR. Due to the presence of the membrane in the downcomer, acting as the local hydrodynamic resistance, the gas holdup in the riser of the ELAMR increases maximally by 16%, while the liquid velocity in the downcomer decreases up to 50%. The values of the gas holdup and liquid velocity predicted by the application of empirical power law correlations and a feed forward back propagation neural network (ANN are in very good agreement with experimental values.

  14. Experiment Needs and Facilities Study Appendix A Transient Reactor Test Facility (TREAT) Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    None

    1976-09-01

    The TREAT Upgrade effort is designed to provide significant new capabilities to satisfy experiment requirements associated with key LMFBR Safety Issues. The upgrade consists of reactor-core modifications to supply the physics performance needed for the new experiments, an Advanced TREAT loop with size and thermal-hydraulics capabilities needed for the experiments, associated interface equipment for loop operations and handling, and facility modifications necessary to accommodate operations with the Loop. The costs and schedules of the tasks to be accomplished under the TREAT Upgrade project are summarized. Cost, including contingency, is about 10 million dollars (1976 dollars). A schedule for execution of 36 months has been established to provide the new capabilities in order to provide timely support of the LMFBR national effort. A key requirement for the facility modifications is that the reactor availability will not be interrupted for more than 12 weeks during the upgrade. The Advanced TREAT loop is the prototype for the STF small-bundle package loop. Modified TREAT fuel elements contain segments of graphite-matrix fuel with graded uranium loadings similar to those of STF. In addition, the TREAT upgrade provides for use of STF-like stainless steel-UO{sub 2} TREAT fuel for tests of fully enriched fuel bundles. This report will introduce the Upgrade study by presenting a brief description of the scope, performance capability, safety considerations, cost schedule, and development requirements. This work is followed by a "Design Description". Because greatly upgraded loop performance is central to the upgrade, a description is given of Advanced TREAT loop requirements prior to description of the loop concept. Performance requirements of the upgraded reactor system are given. An extensive discussion of the reactor physics calculations performed for the Upgrade concept study is provided. Adequate physics performance is essential for performance of experiments with

  15. A simple model of reactor cores for reactor neutrino flux calculations for the KamLAND experiment

    Energy Technology Data Exchange (ETDEWEB)

    Nakajima, K. [Research Center for Neutrino Science, Tohoku University, Sendai 980-8578 (Japan)]. E-mail: kyo@awa.tohoku.ac.jp; Inoue, K. [Research Center for Neutrino Science, Tohoku University, Sendai 980-8578 (Japan); Owada, K. [Research Center for Neutrino Science, Tohoku University, Sendai 980-8578 (Japan); Suekane, F. [Research Center for Neutrino Science, Tohoku University, Sendai 980-8578 (Japan); Suzuki, A. [Research Center for Neutrino Science, Tohoku University, Sendai 980-8578 (Japan); Hirano, G. [TEPCO Systems Corporation, Tokyo 135-0034 (Japan); Kosaka, S. [TEPCO Systems Corporation, Tokyo 135-0034 (Japan); Ohta, T. [Tokyo Electric Power Company, Tokyo 100-8560 (Japan); Tanaka, H. [Tokyo Electric Power Company, Tokyo 100-8560 (Japan)

    2006-12-21

    KamLAND is a reactor neutrino oscillation experiment with a very long baseline. This experiment successfully measured oscillation phenomena of reactor antineutrinos coming mainly from 53 reactors in Japan. In order to extract the results, it is necessary to accurately calculate time-dependent antineutrino spectra from all the reactors. A simple model of reactor cores and code implementing it were developed for this purpose. This paper describes the model of the reactor cores used in the KamLAND reactor analysis.

  16. Advanced Reactor Fuels Irradiation Experiment Design Objectives

    Energy Technology Data Exchange (ETDEWEB)

    Chichester, Heather Jean MacLean [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hayes, Steven Lowe [Idaho National Lab. (INL), Idaho Falls, ID (United States); Dempsey, Douglas [Idaho National Lab. (INL), Idaho Falls, ID (United States); Harp, Jason Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    This report summarizes the objectives of the current irradiation testing activities being undertaken by the Advanced Fuels Campaign relative to supporting the development and demonstration of innovative design features for metallic fuels in order to realize reliable performance to ultra-high burnups. The AFC-3 and AFC-4 test series are nearing completion; the experiments in this test series that have been completed or are in progress are reviewed and the objectives and test matrices for the final experiments in these two series are defined. The objectives, testing strategy, and test parameters associated with a future AFC test series, AFC-5, are documented. Finally, the future intersections and/or synergies of the AFC irradiation testing program with those of the TREAT transient testing program, emerging needs of proposed Versatile Test Reactor concepts, and the Joint Fuel Cycle Study program’s Integrated Recycle Test are discussed.

  17. Simulation of Reactors for Antineutrino Experiments Using DRAGON

    OpenAIRE

    Winslow, L.

    2011-01-01

    From the discovery of the neutrino to the precision neutrino oscillation measurements in KamLAND, nuclear reactors have proven to be an important source of antineutrinos. As their power and our knowledge of neutrino physics has increased, more sensitive measurements have become possible. The next generation of reactor antineutrino experiments require more detailed simulations of the reactor core. Many of the reactor simulation codes are proprietary which makes detailed studies difficult. Here...

  18. 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

    Highlights: • Gas as a coolant in Gen-IV reactors, history and development. • Main physical parameters comparison of gas coolants: carbon dioxide, helium, hydrogen with water. • Forced convection in turbulent pipe flow. • Gas cooled fast reactor concept comparisons to very high temperature reactor concept. • High temperature helium loop: concept, development, mechanism, design and constraints. - Abstract: Rapidly increasing energy and electricity demands, global concerns over the climate changes and strong dependence on foreign fossil fuel supplies are powerfully influencing greater use of nuclear power. In order to establish the viability of next-generation reactor concepts to meet tomorrow's needs for clean and reliable energy production the fundamental research and development issues need to be addressed for the Generation-IV nuclear-energy systems. Generation-IV reactor concepts are being developed to use more advanced materials, coolants and higher burn-ups fuels, while keeping a nuclear reactor safe and reliable. One of the six Generation-IV concepts is a very high temperature reactor (VHTR). The VHTR concept uses a graphite-moderated core with a once-through uranium fuel cycle, using high temperature helium as the coolant. Because helium is naturally inert and single-phase, the helium-cooled reactor can operate at much higher temperatures, leading to higher efficiency. Current VHTR concepts will use fuels such as uranium dioxide, uranium carbide, or uranium oxycarbide. Since some of these fuels are new in nuclear industry and due to their unknown properties and behavior within VHTR conditions it is very important to address these issues by investigate their characteristics within conditions close to those in VHTRs. This research can be performed in a research reactor with in-pile helium loop designed and constructed in Research Center Rez Ltd. One of the topics analyzed in this article are also physical characteristic and benefits of gas

  19. Preliminary decay heat calculations for the fuel loaded irradiation loop device of the RMB multipurpose Brazilian reactor

    Energy Technology Data Exchange (ETDEWEB)

    Campolina, Daniel; Costa, Antonio Carlos L. da; Andrade, Edison P., E-mail: campolina@cdtn.br, E-mail: aclp@cdtn.br, E-mail: epa@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (SETRE/CDTN/CNEN-MG), Belo Horizonte, MG (Brazil). Servico de Tecnologia de Reatores

    2017-07-01

    The structuring project of the Brazilian Multipurpose Reactor (RMB) is responsible for meeting the capacity to develop and test materials and nuclear fuel for the Brazilian Nuclear Program. An irradiation test device (Loop) capable of performing fuel test for power reactor rods is being conceived for RMB reflector. In this work preliminary neutronic calculations have been carried out in order to determine parameters to the cooling system of the Loop basic design. The heat released as a result of radioactive decay of fuel samples was calculated using ORIGEN-ARP and it resulted less than 200 W after 1 hour of irradiation interruption. (author)

  20. 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.

  1. HYDRODYNAMICS AND MASS-TRANSFER CHARACTERISTICS OF A LOOP-VENTURI REACTOR WITH A DOWNFLOW LIQUID JET EJECTOR

    NARCIS (Netherlands)

    CRAMERS, PHMR; BEENACKERS, AACM; VANDIERENDONCK, LL

    1992-01-01

    The hydrodynamics and mass transfer characteristics of a loop-venturi reactor have been investigated using a downflow liquid jet ejector. The specific interfacial area of the ejector and the main holding vessel were determined separately. The cobalt catalyzed sulfite oxidation was used as a model

  2. Method for carrying out biotechnological processes by means of a multi-phase system in a loop reactor.

    NARCIS (Netherlands)

    Tramper, J.

    1987-01-01

    A method for carrying out biotechnological processes by means of multiphase system in a loop reactor, which system comprises an aqueous phase (11) and at least one organic solvent (9, 10) which is immiscible with water and which has a different density from water; one of the liquid components is

  3. Conceptual design of the integral test loop (I): Reactor coolant system and secondary system

    Energy Technology Data Exchange (ETDEWEB)

    Song, Chul Hwa; Lee, Seong Je; Kwon, Tae Soon; Moon, Sang Ki [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-10-01

    This report describes the conceptual design of the primary coolant system and the secondary system of the Integral Test Loop (ITL) which simulates overall thermal hydraulic phenomena of the primary system of a nuclear power plant during postulated accidents or transients. The design basis for the primary coolant system and secondary system is as follows ; Reference plant: Korean Standard Nuclear Plant (KSNP), Height ratio : 1/1, Volume ratio : 1/200, Power scale : Max. 15% of the scaled nominal power, Temperature, Pressure : Real plant conditions. The primary coolant system includes a reactor vessel, which contains a core simulator, a steam generator, a reactor coolant pump simulator, a pressurizer and piping, which consists of two hot legs, four cold legs and four intermediate legs. The secondary system consists of s steam discharge system, a feedwater supply system and a steam condensing system. This conceptual design report describes general configuration of the reference plant, and major function and operation of each system of the plant. Also described is the design philosophy of each component and system of the ITL, and specified are the design criteria and technical specifications of each component and system of the ITL in the report. 17 refs., 43 figs., 51 tabs. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byung Cheol; Chang, Ki Oak

    1997-05-01

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

  5. 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.

  6. Hydrodynamic analysis of a three-fluidized bed reactor cold flow model for chemical looping hydrogen generation. Pressure characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Xue, Zhipeng; Xiang, Wenguo; Chen, Shiyi; Wang, Dong [Southeast Univ., Nanjing (China). School of Energy and Environment

    2013-07-01

    Chemical looping hydrogen generation (CLHG) can produce pure hydrogen with inherent separation of CO{sub 2} from fossils fuel. The process involves a metal oxide, as an oxygen carrier, such as iron oxide. The CLHG system consists of three reactors: a fuel reactor (FR), a steam reactor (SR) and an air reactor (AR). In the FR, the fuel gases react with iron oxides (hematite Fe{sub 2}O{sub 3}, magnetite Fe{sub 3}O{sub 4}, wuestite FeO), generating reduced iron oxides (FeO or even Fe), and with full conversion of gaseous fuels, pure CO{sub 2} can be obtained after cooling the flue gas from the fuel reactor; in the SR, FeO and Fe reacts with steam to generate magnetite (Fe{sub 3}O{sub 4}) and H{sub 2}, the latter representing the final target product of the process; in the AR, the magnetite is oxidized back to hematite which is used in another cycle. A cold flow model of three-fluidized bed for CLHG corresponding to 50 KW hot units has been built. A major novelty of this facility is the compact fuel reactor, which integrates a bubble and a fast fluidized bed to avoid the incomplete conversion of the fuel gas caused by the thermodynamics equilibrium. In order to study the pressure characteristics and the solids concentration of the system, especially in the fuel reactor, the gas velocity of three reactors, gas flow of L-type value, total solids inventory (TSI) and the secondary air of fuel reactor were varied. Results show that the pressure and the solids concentration are strongly influenced by the fluidizing-gas velocity of three reactors. Moreover, the entrainment of the upper part of fuel reactor increases as the total solids inventory increases, and the operating range of the FR can be changed by introducing secondary air or increasing the total solids inventory.

  7. Closed Loop Experiment Manager (CLEM—An Open and Inexpensive Solution for Multichannel Electrophysiological Recordings and Closed Loop Experiments

    Directory of Open Access Journals (Sweden)

    Hananel Hazan

    2017-10-01

    Full Text Available There is growing need for multichannel electrophysiological systems that record from and interact with neuronal systems in near real-time. Such systems are needed, for example, for closed loop, multichannel electrophysiological/optogenetic experimentation in vivo and in a variety of other neuronal preparations, or for developing and testing neuro-prosthetic devices, to name a few. Furthermore, there is a need for such systems to be inexpensive, reliable, user friendly, easy to set-up, open and expandable, and possess long life cycles in face of rapidly changing computing environments. Finally, they should provide powerful, yet reasonably easy to implement facilities for developing closed-loop protocols for interacting with neuronal systems. Here, we survey commercial and open source systems that address these needs to varying degrees. We then present our own solution, which we refer to as Closed Loop Experiments Manager (CLEM. CLEM is an open source, soft real-time, Microsoft Windows desktop application that is based on a single generic personal computer (PC and an inexpensive, general-purpose data acquisition board. CLEM provides a fully functional, user-friendly graphical interface, possesses facilities for recording, presenting and logging electrophysiological data from up to 64 analog channels, and facilities for controlling external devices, such as stimulators, through digital and analog interfaces. Importantly, it includes facilities for running closed-loop protocols written in any programming language that can generate dynamic link libraries (DLLs. We describe the application, its architecture and facilities. We then demonstrate, using networks of cortical neurons growing on multielectrode arrays (MEA that despite its reliance on generic hardware, its performance is appropriate for flexible, closed-loop experimentation at the neuronal network level.

  8. Simulation of Reactors for Antineutrino Experiments Using DRAGON

    CERN Document Server

    Winslow, L

    2011-01-01

    From the discovery of the neutrino to the precision neutrino oscillation measurements in KamLAND, nuclear reactors have proven to be an important source of antineutrinos. As their power and our knowledge of neutrino physics has increased, more sensitive measurements have become possible. The next generation of reactor antineutrino experiments require more detailed simulations of the reactor core. Many of the reactor simulation codes are proprietary which makes detailed studies difficult. Here we present the results of the open source DRAGON code and compare it to other industry standards for reactor modeling. We use published data from the Takahama reactor to determine the quality of the simulations. The propagation of the uncertainty to the antineutrino flux is also discussed.

  9. PREDICTION OF GAS HOLD-UP IN A COMBINED LOOP AIR LIFT FLUIDIZED BED REACTOR USING NEWTONIAN AND NON-NEWTONIAN LIQUIDS

    Directory of Open Access Journals (Sweden)

    Sivakumar Venkatachalam

    2011-09-01

    Full Text Available Many experiments have been conducted to study the hydrodynamic characteristics of column reactors and loop reactors. In this present work, a novel combined loop airlift fluidized bed reactor was developed to study the effect of superficial gas and liquid velocities, particle diameter, fluid properties on gas holdup by using Newtonian and non-Newtonian liquids. Compressed air was used as gas phase. Water, 5% n-butanol, various concentrations of glycerol (60 and 80% were used as Newtonian liquids, and different concentrations of carboxy methyl cellulose aqueous solutions (0.25, 0.6 and 1.0% were used as non-Newtonian liquids. Different sizes of spheres, Bearl saddles and Raschig rings were used as solid phases. From the experimental results, it was found that the increase in superficial gas velocity increases the gas holdup, but it decreases with increase in superficial liquid velocity and viscosity of liquids. Based on the experimental results a correlation was developed to predict the gas hold-up for Newtonian and non-Newtonian liquids for a wide range of operating conditions at a homogeneous flow regime where the superficial gas velocity is approximately less than 5 cm/s

  10. Defluoridation of drinking water by electrocoagulation/electroflotation in a stirred tank reactor with a comparative performance to an external-loop airlift reactor

    Energy Technology Data Exchange (ETDEWEB)

    Essadki, A.H., E-mail: essadki@hotmail.com [Ecole Superieure de Technologie de Casablanca, BP 8012, Oasis, Casablanca (Morocco); Gourich, B. [Ecole Superieure de Technologie de Casablanca, BP 8012, Oasis, Casablanca (Morocco); Vial, Ch. [Laboratoire de Genie Chimique et Biochimique, LGCB-UBP/ENSCCF, 24 avenue des Landais, BP 206, 63174 Aubiere Cedex (France); Delmas, H. [Laboratoire de Genie Chimique, ENSIACET-INPT, 5 rue Paulin Talabot, 31106 Toulouse (France); Bennajah, M. [Ecole Superieure de Technologie de Casablanca, BP 8012, Oasis, Casablanca (Morocco); Laboratoire de Genie Chimique, ENSIACET-INPT, 5 rue Paulin Talabot, 31106 Toulouse (France)

    2009-09-15

    Defluoridation using batch electrocoagulation/electroflotation (EC/EF) was carried out in two reactors for comparison purpose: a stirred tank reactor (STR) close to a conventional EC cell and an external-loop airlift reactor (ELAR) that was recently described as an innovative reactor for EC. The respective influences of current density, initial concentration and initial pH on the efficiency of defluoridation were investigated. The same trends were observed in both reactors, but the efficiency was higher in the STR at the beginning of the electrolysis, whereas similar values were usually achieved after 15 min operation. The influence of the initial pH was explained using the analyses of sludge composition and residual soluble aluminum species in the effluents, and it was related to the prevailing mechanisms of defluoridation. Fluoride removal and sludge reduction were both favored by an initial pH around 4, but this value required an additional pre-treatment for pH adjustment. Finally, electric energy consumption was similar in both reactors when current density was lower than 12 mA/cm{sup 2}, but mixing and complete flotation of the pollutants were achieved without additional mechanical power in the ELAR, using only the overall liquid recirculation induced by H{sub 2} microbubbles generated by water electrolysis, which makes subsequent treatments easier to carry out.

  11. Defluoridation of drinking water by electrocoagulation/electroflotation in a stirred tank reactor with a comparative performance to an external-loop airlift reactor.

    Science.gov (United States)

    Essadki, A H; Gourich, B; Vial, Ch; Delmas, H; Bennajah, M

    2009-09-15

    Defluoridation using batch electrocoagulation/electroflotation (EC/EF) was carried out in two reactors for comparison purpose: a stirred tank reactor (STR) close to a conventional EC cell and an external-loop airlift reactor (ELAR) that was recently described as an innovative reactor for EC. The respective influences of current density, initial concentration and initial pH on the efficiency of defluoridation were investigated. The same trends were observed in both reactors, but the efficiency was higher in the STR at the beginning of the electrolysis, whereas similar values were usually achieved after 15min operation. The influence of the initial pH was explained using the analyses of sludge composition and residual soluble aluminum species in the effluents, and it was related to the prevailing mechanisms of defluoridation. Fluoride removal and sludge reduction were both favored by an initial pH around 4, but this value required an additional pre-treatment for pH adjustment. Finally, electric energy consumption was similar in both reactors when current density was lower than 12mA/cm(2), but mixing and complete flotation of the pollutants were achieved without additional mechanical power in the ELAR, using only the overall liquid recirculation induced by H(2) microbubbles generated by water electrolysis, which makes subsequent treatments easier to carry out.

  12. A two-compartment fluidized bed reactor for CO{sub 2} capture by chemical-looping combustion

    Energy Technology Data Exchange (ETDEWEB)

    Kronberger, B.; Loeffler, G.; Hofbauer, H. [Vienna University of Technology, Institute of Chemical Engineering, Getreidemarkt 9/1663, A-1060 Vienna (Austria); Johansson, E.; Mattisson, T.; Lyngfelt, A. [Chalmers University of Technology, Department of Energy Conversion, S-412 96 Gothenburg (Sweden)

    2004-12-01

    Chemical-looping combustion (CLC) is a combustion method for a gaseous fuel with inherent separation of the greenhouse gas carbon dioxide. A CLC system consists of two reactors, an air reactor and a fuel reactor, and an oxygen carrier circulating between the two reactors. The oxygen carrier transfers the oxygen from the air to the fuel. The flue gas from the fuel reactor consists of carbon dioxide and water, while the flue gas from the air reactor is nitrogen from the air. A two-compartment fluidized bed CLC system was designed and tested using a flow model in order to find critical design parameters. Gas velocities and slot design were varied, and the solids circulation rate and gas leakage between the reactors were measured. The solids circulation rate was found to be sufficient. The gas leakage was somewhat high but could be reduced by altering the slot design. Finally, a hot laboratory CLC system is presented with an advanced design for the slot and also with the possibility for inert gas addition into the downcomer for solids flow increase. (Abstract Copyright [2004], Wiley Periodicals, Inc.)

  13. An overview of the Daya Bay Reactor Neutrino Experiment

    CERN Document Server

    Cao, Jun

    2016-01-01

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

  14. An overview of the Daya Bay reactor neutrino experiment

    OpenAIRE

    Cao, Jun; Luk, Kam-Biu

    2016-01-01

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

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

    Science.gov (United States)

    Polzin, Kurt A.; Godfroy, Thomas J.

    2008-01-01

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

  16. 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.

  17. 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.

  18. The detector system of the Daya Bay reactor neutrino experiment

    OpenAIRE

    An, F. P.; Carr, R.; McKeown, R.D.; Tsang, R. H. M.; Wu, F.F.

    2016-01-01

    The Daya Bay experiment was the first to report simultaneous measurements of reactor antineutrinos at multiple baselines leading to the discovery of ν¯e oscillations over km-baselines. Subsequent data has provided the world׳s most precise measurement of sin^2 2θ_(13) and the effective mass splitting Δm^2_(ee). The experiment is located in Daya Bay, China where the cluster of six nuclear reactors is among the world׳s most prolific sources of electron antineutrinos. Multiple antineutrino detect...

  19. Coupling of high temperature nuclear reactor with chemical plant by means of steam loop with heat pump

    Directory of Open Access Journals (Sweden)

    Kopeć Mariusz

    2017-01-01

    Full Text Available High temperature nuclear reactors (HTR can be used as an excellent, emission-free source of technological heat for various industrial applications. Their outlet helium temperature (700°-900°C allows not only for heat supply to all processes below 600°C (referred to as “steam class”, but also enables development of clean nuclear-assisted hydrogen production or coal liquefaction technologies with required temperatures up to 900°C (referred to as “chemical class”. This paper presents the results of analyses done for various configurations of the steam transport loop coupled with the high-temperature heat pump designed for “chemical class” applications. The advantages and disadvantages as well as the key issues are discussed in comparison with alternative solutions, trying to answer the question whether the system with the steam loop and the hightemperature heat pump is viable and economically justified.

  20. Power-level regulation and simulation of nonlinear pressurized water reactor core with xenon oscillation using H-infinity loop shaping control

    Directory of Open Access Journals (Sweden)

    Li Gang

    2016-01-01

    Full Text Available This investigation is to solve the power-level control issue of a nonlinear pressurized water reactor core with xenon oscillations. A nonlinear pressurized water reactor core is modeled using the lumped parameter method, and a linear model of the core is then obtained through the small perturbation linearization way. The H∞loop shapingcontrolis utilized to design a robust controller of the linearized core model.The calculated H∞loop shaping controller is applied to the nonlinear core model. The nonlinear core model and the H∞ loop shaping controller build the nonlinear core power-level H∞loop shaping control system.Finally, the nonlinear core power-level H∞loop shaping control system is simulatedconsidering two typical load processes that are a step load maneuver and a ramp load maneuver, and simulation results show that the nonlinear control system is effective.

  1. Spatial and temporal variations of microbial community in a mixed plug-flow loop reactor fed with dairy manure

    OpenAIRE

    Li, Yueh-Fen; Chen, Po-Hsu; Yu, Zhongtang

    2014-01-01

    Mixed plug-flow loop reactor (MPFLR) has been widely adopted by the US dairy farms to convert cattle manure to biogas. However, the microbiome in MPFLR digesters remains unexplored. In this study, the microbiome in a MPFLR digester operated on a mega-dairy farm was examined thrice over a 2 month period. Within 23 days of retention time, 55–70% of total manure solid was digested. Except for a few minor volatile fatty acids (VFAs), total VFA concentration and pH remained similar along the cours...

  2. Investigating the spectral anomaly with different reactor antineutrino experiments

    Directory of Open Access Journals (Sweden)

    C. Buck

    2017-02-01

    Full Text Available The spectral shape of reactor antineutrinos measured in recent experiments shows anomalies in comparison to neutrino reference spectra. New precision measurements of the reactor neutrino spectra as well as more complete input in nuclear data bases are needed to resolve the observed discrepancies between models and experimental results. This article proposes the combination of experiments at reactors which are highly enriched in U235 with commercial reactors with typically lower enrichment to gain new insights into the origin of the anomalous neutrino spectrum. The presented method clarifies, if the spectral anomaly is either solely or not at all related to the predicted U235 spectrum. Considering the current improvements of the energy scale uncertainty of present-day experiments, a significance of three sigma and above can be reached. As an example, we discuss the option of a direct comparison of the measured shape in the currently running Double Chooz near detector and the upcoming Stereo experiment. A quantitative feasibility study emphasizes that a precise understanding of the energy scale systematics is a crucial prerequisite in recent and next generation experiments investigating the spectral anomaly.

  3. Investigating the spectral anomaly with different reactor antineutrino experiments

    Science.gov (United States)

    Buck, C.; Collin, A. P.; Haser, J.; Lindner, M.

    2017-02-01

    The spectral shape of reactor antineutrinos measured in recent experiments shows anomalies in comparison to neutrino reference spectra. New precision measurements of the reactor neutrino spectra as well as more complete input in nuclear data bases are needed to resolve the observed discrepancies between models and experimental results. This article proposes the combination of experiments at reactors which are highly enriched in 235U with commercial reactors with typically lower enrichment to gain new insights into the origin of the anomalous neutrino spectrum. The presented method clarifies, if the spectral anomaly is either solely or not at all related to the predicted 235U spectrum. Considering the current improvements of the energy scale uncertainty of present-day experiments, a significance of three sigma and above can be reached. As an example, we discuss the option of a direct comparison of the measured shape in the currently running Double Chooz near detector and the upcoming Stereo experiment. A quantitative feasibility study emphasizes that a precise understanding of the energy scale systematics is a crucial prerequisite in recent and next generation experiments investigating the spectral anomaly.

  4. Effect of scale-up on average shear rates for aerated non-Newtonian liquids in external loop airlift reactors.

    Science.gov (United States)

    Al-Masry

    1999-02-01

    Average shear rates have been estimated experimentally in a 700-dm3 external loop airlift reactor. Aqueous pseudoplastic carboxymethylcellulose and xanthan gum solutions were used to simulate non-Newtonian behavior of biological media. Average shear rates of non-Newtonian solutions were found by analogy with Newtonian glycerol solutions using downcomer liquid velocity as the measurable parameter. Due to the complexity of local shear rate measurement, an average shear rate was assumed to exist and is proportional to superficial gas velocity. The data from this work and those in the literature were used in producing a new correlation for estimating average shear rates as a function of superficial gas velocity, geometry, and dispersion height. Wall shear rates were found to be significant. The ratio of wall shear rates to bulk shear rates were varied from 5% to 40%. Furthermore, it has been found that shear rates generated in airlift loop reactors are lower than those generated in bubble columns. Copyright 1999 John Wiley & Sons, Inc.

  5. Latest Results from the Daya Bay Reactor Neutrino Experiment

    CERN Multimedia

    CERN. Geneva

    2014-01-01

    Among all the fundamental particles that have been experimentally observed, neutrinos remain one of the least understood. The Daya Bay Reactor Neutrino Experiment in China consists of eight identical detectors placed underground at different baselines from three groups of nuclear reactors, a configuration that is ideally suited for studying the properties of these elusive particles. This talk will present three sets of results that have just recently been released by the Daya Bay Collaboration: (i) a precision measurement of the oscillation parameters that drive the disappearance of electron antineutrinos at short baselines, (ii) a search for sterile neutrino mixing, and (iii) a high-statistics determination of the absolute flux and spectrum of reactor-produced electron antineutrinos. All of these results extend the limits of our knowledge in their respective areas and thus shed new light on neutrinos and the physics that surround them.

  6. 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.

  7. 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.

  8. Preliminary Feasibility, Design, and Hazard Analysis of a Boiling Water Test Loop Within the Idaho National Laboratory Advanced Test Reactor National Scientific User Facility

    Energy Technology Data Exchange (ETDEWEB)

    Douglas M. Gerstner

    2009-05-01

    The Advanced Test Reactor (ATR) is a pressurized light-water reactor with a design thermal power of 250 MW. The principal function of the ATR is to provide a high neutron flux for testing reactor fuels and other materials. The ATR and its support facilities are located at the Idaho National Laboratory (INL). A Boiling Water Test Loop (BWTL) is being designed for one of the irradiation test positions within the. The objective of the new loop will be to simulate boiling water reactor (BWR) conditions to support clad corrosion and related reactor material testing. Further it will accommodate power ramping tests of candidate high burn-up fuels and fuel pins/rods for the commercial BWR utilities. The BWTL will be much like the pressurized water loops already in service in 5 of the 9 “flux traps” (region of enhanced neutron flux) in the ATR. The loop coolant will be isolated from the primary coolant system so that the loop’s temperature, pressure, flow rate, and water chemistry can be independently controlled. This paper presents the proposed general design of the in-core and auxiliary BWTL systems; the preliminary results of the neutronics and thermal hydraulics analyses; and the preliminary hazard analysis for safe normal and transient BWTL and ATR operation.

  9. Design, Construction and Testing of an In-Pile Loop for PWR (Pressurized Water Reactor) Simulation.

    Science.gov (United States)

    1987-06-01

    concentrations. It was found that raising coolant pH would decrease visual crud concentrations, but steam generator plenum and primary coolant pipe wall...built, the Loop duplicates the core and Steam Generator fluid surface film differential temperatures, bulk fluid temperatures, and wall fluid shear...2.4.2 Loop Plenum Design 39 2.4.3 Neutron Absorber Plug 44 2.4.4 Plenum Section Instrumentation 45 2.5 Steam Generator /Heat Rejection Section 45 5 2.5.1

  10. Hydrodynamics of a hybrid circulating fluidized bed reactor with a partitioned loop seal system

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Dal-Hee; Moon, Jong-Ho; Jin, Gyoung Tae; Shun, Dowon [Korea Institute of Energy Research, Daejeon (Korea, Republic of); Yun, Minyoung; Park, Chan Seung; Norbeck, Joseph M. [University of California, Riverside (United States)

    2015-07-15

    A circulating fluidized bed (CFB) with a hybrid design has been developed and optimized for steam hydrogasification. The hybrid CFB is composed of a bubbling fluidized bed (BFB) type combustor and a fast fluidized bed (FB) type gasifier. Char is burnt in the combustor and the generated heat is supplied to the gasifier along with the bed materials. Two different types of fluidized beds are connected to each other with a newly developed partitioned loop seal to avoid direct contact between two separate gas streams flowing in each fluidized bed. Gas mixing tests were carried out with Air and Argon in a cold model hybrid CFB to test the loop seal efficiency. Increase in solid inventory in the loop seal can improve the gas separation efficiency. It can be realized at higher gas velocity in fast bed and with higher solid inventory in the loop seal system. In addition, bed hydrodynamics was investigated with varying gas flow conditions and particle sizes in order to obtain a full understanding of changes of solid holdup in the FB. The solid holdup in the FB increased with increasing gas velocity in the BFB. Conversely, increase in gas velocity in the FB contributed to reducing the solid holdup in the FB. It was observed that changing the particle size of bed material does not have a big impact on hydrodynamic parameters.

  11. Reactor choices for chemical looping combustion (CLC) dependencies on materials characteristics

    NARCIS (Netherlands)

    Kimball, E.; Lambert, A.; Fossdal, A.; Leenman, R.N.; Comte, E.; Bos, W.A.P. van den; Blom, R.

    2013-01-01

    The physio-chemical stability of the oxygen carrier material during chemical looping combustion (CLC) operation is crucial. In the present paper we discuss the challenges connected to operating a metal oxide base material in a cyclic manner between oxidizing and reducing atmospheres. Especially,

  12. Startup of reactors for anoxic ammonium oxidation: experiences from the first full-scale anammox reactor in Rotterdam.

    Science.gov (United States)

    van der Star, Wouter R L; Abma, Wiebe R; Blommers, Dennis; Mulder, Jan-Willem; Tokutomi, Takaaki; Strous, Marc; Picioreanu, Cristian; van Loosdrecht, Mark C M

    2007-10-01

    The first full-scale anammox reactor in the world was started in Rotterdam (NL). The reactor was scaled-up directly from laboratory-scale to full-scale and treats up to 750 kg-N/d. In the initial phase of the startup, anammox conversions could not be identified by traditional methods, but quantitative PCR proved to be a reliable indicator for growth of the anammox population, indicating an anammox doubling time of 10-12 days. The experience gained during this first startup in combination with the availability of seed sludge from this reactor, will lead to a faster startup of anammox reactors in the future. The anammox reactor type employed in Rotterdam was compared to other reactor types for the anammox process. Reactors with a high specific surface area like the granular sludge reactor employed in Rotterdam provide the highest volumetric loading rates. Mass transfer of nitrite into the biofilm is limiting the conversion of those reactor types that have a lower specific surface area. Now the first full-scale commercial anammox reactor is in operation, a consistent and descriptive nomenclature is suggested for reactors in which the anammox process is employed.

  13. Cephalosporin C production by a highly productive Cephalosporium acremonium strain in an airlift tower loop reactor with static mixers.

    Science.gov (United States)

    Zhou, W; Holzhauer-Rieger, K; Bayer, T; Schügerl, K

    1993-04-01

    The production of cephalosporin C (CPC) and its precursors penicillin N (PEN N), deacetoxycephalosporin C (DAOC) and deacetylcephalosporin C (DAC), with a highly productive strain of Cephalosporin acremonium, was investigated in an 80-1 airlift tower loop reactor with four static mixer modules (Type SMV, Sulzer) (ATLRM) on a complex medium containing 50 g l-1 peanut flour (PF). The most important key parameters such as glucose concentration and cell mass concentration were monitored during a fed-batch cultivation process. The concentrations of products CPC, PEN N, DAOC an DAC were determined on line by HPLC. The influences of four motionless mixers on the dissolved oxygen concentration (DOC), oxygen transfer rate, the cell growth and the CPC production, as well as the reactor performance, were evaluated. The results were compared with the performance of an airlift tower loop reactor (ATLR) without static mixers as well as with a stirred tank reactor (STR). A comparison of cultivations in the ATLRM and ATLR with 50 g l-1 PF indicates that the obtained maximal CPC concentration and the (CPC + DAC + DAOC) concentration were 7% and 22% higher in the ATLRM (4.96 and 7.46 g l-1) than in the ATLR (4.63 and 6.13 g l-1) respectively. The maximal CPC volumetric productivity in the ATLRM (55.1 mg l-1 h-1) was also considerably higher than that in the ATLR (48.5 mg l-1 h-1). The specific power input was reduced from 2.36 to 1.5 kW m-3, the specific productivity pertaining to the power input was improved from 1.96 to 3.31 g W-1. On the other hand, cultivation in the ATLRM had a lower maximum CPC concentration and volumetric productivity than those in STR (7.2 g l-1 and 71.2 mg l-1 h-1) with the same medium due to the lower shear stress levels and the lower specific power input (1.5 vs. 3.0 kW m-3); but the specific power imput-based yield coefficient was in the ATLRM (3.31 g W-1) higher than in the STR (2.40 g W-1). By increasing the amount of PF, it was possible to enhance the

  14. Contributions of Cu-rich clusters, dislocation loops and nanovoids to the irradiation-induced hardening of Cu-bearing low-Ni reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Bergner, F., E-mail: f.bergner@hzdr.de [Helmholtz-Zentrum Dresden-Rossendorf, Bautzner Landstr. 400, 01328 Dresden (Germany); Gillemot, F. [Centre for Energy Research of the Hungarian Academy of Sciences, 29-33 Konkoly-Thege street, 1121 Budapest XII (Hungary); Hernández-Mayoral, M.; Serrano, M. [Division of Materials, CIEMAT, Avenida Complutense 22, 28040 Madrid (Spain); Török, G. [Wigner Research Center for Physics of the Hungarian Academy of Sciences, 29-33 Konkoly-Thege street, 1121 Budapest XII (Hungary); Ulbricht, A.; Altstadt, E. [Helmholtz-Zentrum Dresden-Rossendorf, Bautzner Landstr. 400, 01328 Dresden (Germany)

    2015-06-15

    Highlights: • TEM and SANS were applied to estimate mean size and number density of loops, nanovoids and Cu-rich clusters. • A three-feature dispersed-barrier hardening model was applied to estimate the yield stress increase. • The values and errors of the dimensionless obstacle strength were estimated in a consistent way. • Nanovoids are stronger obstacles for dislocation glide than dislocation loops, loops are stronger than Cu-rich clusters. • For reactor-relevant conditions, Cu-rich clusters contribute most to hardening due to their high number density. - Abstract: Dislocation loops, nanovoids and Cu-rich clusters (CRPs) are known to represent obstacles for dislocation glide in neutron-irradiated reactor pressure vessel (RPV) steels, but a consistent experimental determination of the respective obstacle strengths is still missing. A set of Cu-bearing low-Ni RPV steels and model alloys was characterized by means of SANS and TEM in order to specify mean size and number density of loops, nanovoids and CRPs. The obstacle strengths of these families were estimated by solving an over-determined set of linear equations. We have found that nanovoids are stronger than loops and loops are stronger than CRPs. Nevertheless, CRPs contribute most to irradiation hardening because of their high number density. Nanovoids were only observed for neutron fluences beyond typical end-of-life conditions of RPVs. The estimates of the obstacle strength are critically compared with reported literature data.

  15. Overview of Experiments for Physics of Fast Reactors from the International Handbooks of Evaluated Criticality Safety Benchmark Experiments and Evaluated Reactor Physics Benchmark Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Bess, J. D.; Briggs, J. B.; Gulliford, J.; Ivanova, T.; Rozhikhin, E. V.; Semenov, M. Yu.; Tsibulya, A. M.; Koscheev, V. N.

    2017-07-01

    Overview of Experiments to Study the Physics of Fast Reactors Represented in the International Directories of Critical and Reactor Experiments John D. Bess Idaho National Laboratory Jim Gulliford, Tatiana Ivanova Nuclear Energy Agency of the Organisation for Economic Cooperation and Development E.V.Rozhikhin, M.Yu.Sem?nov, A.M.Tsibulya Institute of Physics and Power Engineering The study the physics of fast reactors traditionally used the experiments presented in the manual labor of the Working Group on Evaluation of sections CSEWG (ENDF-202) issued by the Brookhaven National Laboratory in 1974. This handbook presents simplified homogeneous model experiments with relevant experimental data, as amended. The Nuclear Energy Agency of the Organization for Economic Cooperation and Development coordinates the activities of two international projects on the collection, evaluation and documentation of experimental data - the International Project on the assessment of critical experiments (1994) and the International Project on the assessment of reactor experiments (since 2005). The result of the activities of these projects are replenished every year, an international directory of critical (ICSBEP Handbook) and reactor (IRPhEP Handbook) experiments. The handbooks present detailed models of experiments with minimal amendments. Such models are of particular interest in terms of the settlements modern programs. The directories contain a large number of experiments which are suitable for the study of physics of fast reactors. Many of these experiments were performed at specialized critical stands, such as BFS (Russia), ZPR and ZPPR (USA), the ZEBRA (UK) and the experimental reactor JOYO (Japan), FFTF (USA). Other experiments, such as compact metal assembly, is also of interest in terms of the physics of fast reactors, they have been carried out on the universal critical stands in Russian institutes (VNIITF and VNIIEF) and the US (LANL, LLNL, and others.). Also worth mentioning

  16. Alpha Particle Physics Experiments in the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Budny, R.V.; Darrow, D.S.; Medley, S.S.; Nazikian, R.; Zweben, S.J.; et al.

    1998-12-14

    Alpha particle physics experiments were done on the Tokamak Fusion Test Reactor (TFTR) during its deuterium-tritium (DT) run from 1993-1997. These experiments utilized several new alpha particle diagnostics and hundreds of DT discharges to characterize the alpha particle confinement and wave-particle interactions. In general, the results from the alpha particle diagnostics agreed with the classical single-particle confinement model in magnetohydrodynamic (MHD) quiescent discharges. Also, the observed alpha particle interactions with sawteeth, toroidal Alfvén eigenmodes (TAE), and ion cyclotron resonant frequency (ICRF) waves were roughly consistent with theoretical modeling. This paper reviews what was learned and identifies what remains to be understood.

  17. Recent Results from Daya Bay Reactor Neutrino Experiment

    CERN Document Server

    Hu, Bei-Zhen

    2015-01-01

    The Daya Bay reactor neutrino experiment announced the discovery of a non-zero value of \\sin^22\\theta_{13} with significance better than 5 \\sigma in 2012. The experiment is continuing to improve the precision of \\sin^22\\theta_{13} and explore other physics topics. In this talk, I will show the current oscillation and mass-squared difference results which are based on the combined analysis of the measured rates and energy spectra of antineutrino events, an independent measurement of \\theta_{13} using IBD events where delayed neutrons are captured on hydrogens, and a search for light sterile neutrinos.

  18. INFLUENCE OF THE GAS-DENSITY ON THE GAS ENTRAINMENT RATE AND GAS HOLD-UP IN LOOP-VENTURI REACTORS

    NARCIS (Netherlands)

    CRAMERS, PHMR; VANDIERENDONCK, LL; BEENACKERS, AACM

    1992-01-01

    The hydrodynamics of a loop-venturi reactor were investigated using a downflow liquid jet ejector. Both the gas entrainment rate of the ejector and the gas hold-up in the main holding vessel were shown to be influenced by the gas density. The amount of volumetrically entrained gas as well as the gas

  19. Death rate in a small air-lift loop reactor of vero cells grown on solid microcarriers and in macroporous microcarriers

    NARCIS (Netherlands)

    Martens, D E; Nollen, E A; Hardeveld, M; Velden-de Groot, C A; Gooijer, C D; Beuvery, E C; Tramper, J

    The death rate of Vero cells grown on Cytodex-3 microcarrierswas studied as a function of the gas flow rate in a smallair-lift loop reactor. The death rate may be described byfirst-order death-rate kinetics. The first-order death-rateconstant as calculated from the decrease in viable cells,

  20. Effect of Spray System on Fission Product Distribution in Containment During a Severe Accident in a Two-Loop Pressurized Water Reactor

    National Research Council Canada - National Science Library

    Dehjourian, Mehdi; Rahgoshay, Mohammad; Sayareh, Reza; Jahanfarnia, Gholamreza; Shirani, Amir Saied

    2016-01-01

    The containment response during the first 24 hours of a low-pressure severe accident scenario in a nuclear power plant with a two-loop Westinghouse-type pressurized water reactor was simulated with the CONTAIN 2.0 computer code...

  1. Operation experience of the Indonesian multipurpose research reactor RSG-GAS

    Energy Technology Data Exchange (ETDEWEB)

    Hastowo, Hudi; Tarigan, Alim [Multipurpose Reactor Center, National Nuclear Energy Agency of the Republic of Indonesia (PRSG-BATAN), Kawasan PUSPIPTEK Serpong, Tangerang (Indonesia)

    1999-08-01

    RSG-GAS is a multipurpose research reactor with nominal power of 30 MW, operated by BATAN since 1987. The reactor is an open pool type, cooled and moderated with light water, using the LEU-MTR fuel element in the form of U{sub 3}O{sub 8}-Al dispersion. Up to know, the reactor have been operated around 30,000 hours to serve the user. The reactor have been utilized to produce radioisotope, neutron beam experiments, irradiation of fuel element and its structural material, and reactor physics experiments. This report will explain in further detail concerning operational experience of this reactor, i.e. reactor operation data, reactor utilization, research program, technical problems and it solutions, plant modification and improvement, and development plan to enhance better reactor operation performance and its utilization. (author)

  2. The detector system of the Daya Bay reactor neutrino experiment

    Science.gov (United States)

    An, F. P.; Bai, J. Z.; Balantekin, A. B.; Band, H. R.; Beavis, D.; Beriguete, W.; Bishai, M.; Blyth, S.; Brown, R. L.; Butorov, I.; Cao, D.; Cao, G. F.; Cao, J.; Carr, R.; Cen, W. R.; Chan, W. T.; Chan, Y. L.; Chang, J. F.; Chang, L. C.; Chang, Y.; Chasman, C.; Chen, H. Y.; Chen, H. S.; Chen, M. J.; Chen, Q. Y.; Chen, S. J.; Chen, S. M.; Chen, X. C.; Chen, X. H.; Chen, X. S.; Chen, Y. X.; Chen, Y.; Cheng, J. H.; Cheng, J.; Cheng, Y. P.; Cherwinka, J. J.; Chidzik, S.; Chow, K.; Chu, M. C.; Cummings, J. P.; de Arcos, J.; Deng, Z. Y.; Ding, X. F.; Ding, Y. Y.; Diwan, M. V.; Dong, L.; Dove, J.; Draeger, E.; Du, X. F.; Dwyer, D. A.; Edwards, W. R.; Ely, S. R.; Fang, S. D.; Fu, J. Y.; Fu, Z. W.; Ge, L. Q.; Ghazikhanian, V.; Gill, R.; Goett, J.; Gonchar, M.; Gong, G. H.; Gong, H.; Gornushkin, Y. A.; Grassi, M.; Greenler, L. S.; Gu, W. Q.; Guan, M. Y.; Guo, R. P.; Guo, X. H.; Hackenburg, R. W.; Hahn, R. L.; Han, R.; Hans, S.; He, M.; He, Q.; He, W. S.; Heeger, K. M.; Heng, Y. K.; Higuera, A.; Hinrichs, P.; Ho, T. H.; Hoff, M.; Hor, Y. K.; Hsiung, Y. B.; Hu, B. Z.; Hu, L. M.; Hu, L. J.; Hu, T.; Hu, W.; Huang, E. C.; Huang, H. Z.; Huang, H. X.; Huang, P. W.; Huang, X.; Huang, X. T.; Huber, P.; Hussain, G.; Isvan, Z.; Jaffe, D. E.; Jaffke, P.; Jen, K. L.; Jetter, S.; Ji, X. P.; Ji, X. L.; Jiang, H. J.; Jiang, W. Q.; Jiao, J. B.; Johnson, R. A.; Joseph, J.; Kang, L.; Kettell, S. H.; Kohn, S.; Kramer, M.; Kwan, K. K.; Kwok, M. W.; Kwok, T.; Lai, C. Y.; Lai, W. C.; Lai, W. H.; Langford, T. J.; Lau, K.; Lebanowski, L.; Lee, J.; Lee, M. K. P.; Lei, R. T.; Leitner, R.; Leung, J. K. C.; Lewis, C. A.; Li, B.; Li, C.; Li, D. J.; Li, F.; Li, G. S.; Li, J.; Li, N. Y.; Li, Q. J.; Li, S. F.; Li, S. C.; Li, W. D.; Li, X. B.; Li, X. N.; Li, X. Q.; Li, Y.; Li, Y. F.; Li, Z. B.; Liang, H.; Liang, J.; Lin, C. J.; Lin, G. L.; Lin, P. Y.; Lin, S. X.; Lin, S. K.; Lin, Y. C.; Ling, J. J.; Link, J. M.; Littenberg, L.; Littlejohn, B. R.; Liu, B. J.; Liu, C.; Liu, D. W.; Liu, H.; Liu, J. L.; Liu, J. C.; Liu, S.; Liu, S. S.; Liu, X.; Liu, Y. B.; Lu, C.; Lu, H. Q.; Lu, J. S.; Luk, A.; Luk, K. B.; Luo, T.; Luo, X. L.; Ma, L. H.; Ma, Q. M.; Ma, X. Y.; Ma, X. B.; Ma, Y. Q.; Mayes, B.; McDonald, K. T.; McFarlane, M. C.; McKeown, R. D.; Meng, Y.; Mitchell, I.; Mohapatra, D.; Monari Kebwaro, J.; Morgan, J. E.; Nakajima, Y.; Napolitano, J.; Naumov, D.; Naumova, E.; Newsom, C.; Ngai, H. Y.; Ngai, W. K.; Nie, Y. B.; Ning, Z.; Ochoa-Ricoux, J. P.; Olshevskiy, A.; Pagac, A.; Pan, H.-R.; Patton, S.; Pearson, C.; Pec, V.; Peng, J. C.; Piilonen, L. E.; Pinsky, L.; Pun, C. S. J.; Qi, F. Z.; Qi, M.; Qian, X.; Raper, N.; Ren, B.; Ren, J.; Rosero, R.; Roskovec, B.; Ruan, X. C.; Sands, W. R.; Seilhan, B.; Shao, B. B.; Shih, K.; Song, W. Y.; Steiner, H.; Stoler, P.; Stuart, M.; Sun, G. X.; Sun, J. L.; Tagg, N.; Tam, Y. H.; Tanaka, H. K.; Tang, W.; Tang, X.; Taychenachev, D.; Themann, H.; Torun, Y.; Trentalange, S.; Tsai, O.; Tsang, K. V.; Tsang, R. H. M.; Tull, C. E.; Tung, Y. C.; Viaux, N.; Viren, B.; Virostek, S.; Vorobel, V.; Wang, C. H.; Wang, L. S.; Wang, L. Y.; Wang, L. Z.; Wang, M.; Wang, N. Y.; Wang, R. G.; Wang, T.; Wang, W.; Wang, W. W.; Wang, X. T.; Wang, X.; Wang, Y. F.; Wang, Z.; Wang, Z.; Wang, Z. M.; Webber, D. M.; Wei, H. Y.; Wei, Y. D.; Wen, L. J.; Wenman, D. L.; Whisnant, K.; White, C. G.; Whitehead, L.; Whitten, C. A.; Wilhelmi, J.; Wise, T.; Wong, H. C.; Wong, H. L. H.; Wong, J.; Wong, S. C. F.; Worcester, E.; Wu, F. F.; Wu, Q.; Xia, D. M.; Xia, J. K.; Xiang, S. T.; Xiao, Q.; Xing, Z. Z.; Xu, G.; Xu, J. Y.; Xu, J. L.; Xu, J.; Xu, W.; Xu, Y.; Xue, T.; Yan, J.; Yang, C. G.; Yang, L.; Yang, M. S.; Yang, M. T.; Ye, M.; Yeh, M.; Yeh, Y. S.; Yip, K.; Young, B. L.; Yu, G. Y.; Yu, Z. Y.; Zeng, S.; Zhan, L.; Zhang, C.; Zhang, F. H.; Zhang, H. H.; Zhang, J. W.; Zhang, K.; Zhang, Q. X.; Zhang, Q. M.; Zhang, S. H.; Zhang, X. T.; Zhang, Y. C.; Zhang, Y. H.; Zhang, Y. M.; Zhang, Y. X.; Zhang, Y. M.; Zhang, Z. J.; Zhang, Z. Y.; Zhang, Z. P.; Zhao, J.; Zhao, Q. W.; Zhao, Y. F.; Zhao, Y. B.; Zheng, L.; Zhong, W. L.; Zhou, L.; Zhou, N.; Zhou, Z. Y.; Zhuang, H. L.; Zimmerman, S.; Zou, J. H.

    2016-03-01

    The Daya Bay experiment was the first to report simultaneous measurements of reactor antineutrinos at multiple baselines leading to the discovery of νbare oscillations over km-baselines. Subsequent data has provided the world's most precise measurement of sin2 2θ13 and the effective mass splitting Δ mee2. The experiment is located in Daya Bay, China where the cluster of six nuclear reactors is among the world's most prolific sources of electron antineutrinos. Multiple antineutrino detectors are deployed in three underground water pools at different distances from the reactor cores to search for deviations in the antineutrino rate and energy spectrum due to neutrino mixing. Instrumented with photomultiplier tubes, the water pools serve as shielding against natural radioactivity from the surrounding rock and provide efficient muon tagging. Arrays of resistive plate chambers over the top of each pool provide additional muon detection. The antineutrino detectors were specifically designed for measurements of the antineutrino flux with minimal systematic uncertainty. Relative detector efficiencies between the near and far detectors are known to better than 0.2%. With the unblinding of the final two detectors' baselines and target masses, a complete description and comparison of the eight antineutrino detectors can now be presented. This paper describes the Daya Bay detector systems, consisting of eight antineutrino detectors in three instrumented water pools in three underground halls, and their operation through the first year of eight detector data-taking.

  3. Reactor electron antineutrino disappearance in the Double Chooz experiment

    CERN Document Server

    Abe, Y; Anjos, J C dos; Barriere, J C; Bergevin, M; Bernstein, A; Bezerra, T J C; Bezrukhov, L; Blucher, E; Bowden, N S; Buck, C; Busenitz, J; Cabrera, A; Caden, E; Camilleri, L; Carr, R; Cerrada, M; Chang, P -J; Chimenti, P; Classen, T; Collin, A P; Conover, E; Conrad, J M; Crespo-Anadón, J I; Crum, K; Cucoanes, A; D'Agostino, M V; Damon, E; Dawson, J V; Dazeley, S; Dietrich, D; Djurcic, Z; Dracos, M; Durand, V; Ebert, J; Efremenko, Y; Elnimr, M; Etenko, A; Fallot, M; Fechner, M; von Feilitzsch, F; Felde, J; Franco, D; Franke, A J; Franke, M; Furuta, H; Gama, R; Gil-Botella, I; Giot, L; Goger-Neff, M; Gonzalez, L F G; Goodman, M C; Goon, J TM; Greiner, D; Haag, N; Hagner, C; Hara, T; Hartmann, F X; Haser, J; Hatzikoutelis, A; Hayakawa, T; Hofmann, M; Horton-Smith, G A; Hourlier, A; Ishitsuka, M; Jochum, J; Jollet, C; Jones, C L; Kaether, F; Kalousis, L N; Kamyshkov, Y; Kaplan, D M; Kawasaki, T; Keefer, G; Kemp, E; de Kerret, H; Kibe, Y; Konno, T; Kryn, D; Kuze, M; Lachenmaier, T; Lane, C E; Langbrandtner, C; Lasserre, T; Letourneau, A; Lhuillier, D; Lima, H P; Lindner, M; López-Castanõ, J M; LoSecco, J M; Lubsandorzhiev, B K; Lucht, S; McKee, D; Maeda, J; Maesano, C N; Mariani, C; Maricic, J; Martino, J; Matsubara, T; Mention, G; Meregaglia, A; Miletic, T; Milincic, R; Miyata, H; Mueller, Th A; Nagasaka, Y; Nakajima, K; Novella, P; Obolensky, M; Oberauer, L; Onillon, A; Osborn, A; Ostrovskiy, I; Palomares, C; Pepe, I M; Perasso, S; Perrin, P; Pfahler, P; Porta, A; Potzel, W; Reichenbacher, J; Reinhold, B; Remoto, A; Rohling, M; Roncin, R; Roth, S; Sakamoto, Y; Santorelli, R; Sato, F; Schonert, S; Schoppmann, S; Schwetz, T; Shaevitz, M H; Shimojima, S; Shrestha, D; Sida, J-L; Sinev, V; Skorokhvatov, M; Smith, E; Spitz, J; Stahl, A; Stancu, I; Stokes, L F F; Strait, M; Stuken, A; Suekane, F; Sukhotin, S; Sumiyoshi, T; Sun, Y; Svoboda, R; Terao, K; Tonazzo, A; Toups, M; Thi, H H Trinh; Valdiviesso, G; Veyssiere, C; Wagner, S; Watanabe, H; White, B; Wiebusch, C; Winslow, L; Worcester, M; Wurm, M; Yermia, F; Zimmer, V

    2012-01-01

    The Double Chooz experiment has observed 8,249 candidate electron antineutrino events in 227.93 live days with 33.71 GW-ton-years (reactor power x detector mass x livetime) exposure using a 10.3 cubic meter fiducial volume detector located at 1050 m from the reactor cores of the Chooz nuclear power plant in France. The expectation in case of theta13 = 0 is 8,937 events. The deficit is interpreted as evidence of electron antineutrino disappearance. From a rate plus spectral shape analysis we find sin^2 2{\\theta}13 = 0.109 \\pm 0.030(stat) \\pm 0.025(syst). The data exclude the no-oscillation hypothesis at 99.9% CL (3.1{\\sigma}).

  4. Status Report on Scoping Reactor Physics and Sensitivity/Uncertainty Analysis of LR-0 Reactor Molten Salt Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Mueller, Donald E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Patton, Bruce W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division

    2016-08-31

    Experiments are being planned at Research Centre Rež (RC Rež) to use the FLiBe (2 7LiF-BeF2) salt from the Molten Salt Reactor Experiment (MSRE) to perform reactor physics measurements in the LR-0 low power nuclear reactor. These experiments are intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems utilizing FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL) is performing sensitivity/uncertainty (S/U) analysis of these planned experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objective of these analyses is to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a status update on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. The S/U analyses will be used to inform design of FLiBe-based experiments using the salt from MSRE.

  5. MHTGR: New production reactor summary of experience base

    Energy Technology Data Exchange (ETDEWEB)

    1988-03-01

    Worldwide interest in the Modular High-Temperature Gas-Cooled Reactor (MHTGR) stems from the capability of the system to retain the advanced fuel and thermal performance while providing unparalleled levels of safety. The small power level of the MHTGR and its passive systems give it a margin of safety not attained by other concepts being developed for power generation. This report covers the experience base for the key nuclear system, components, and processes related to the MHTGR-NPR. 9 refs., 39 figs., 9 tabs.

  6. Educational reactor-physics experiments with the critical assemble TCA

    Energy Technology Data Exchange (ETDEWEB)

    Tsutsui, Hiroaki; Okubo, Masaaki; Igashira, Masayuki [Tokyo Inst. of Tech. (Japan); Horiki, Oichiro; Suzaki, Takenori

    1997-10-01

    The Tank-Type Critical Assembly (TCA) of Japan Atomic Energy Research Institute is research equipment for light water reactor physics. In the present report, the lectures given to the graduate students of Tokyo Institute of Technology who participated in the educational experiment course held on 26-30 August at TCA are rearranged to provide useful information for those who will implement educational basic experiments with TCA in the future. This report describes the principles, procedures, and data analyses for (1) Critical approach and Exponential experiment, (2) Measurement of neutron flux distribution, (3) Measurement of power distribution, (4) Measurement of fuel rod worth distribution, and (5) Measurement of safety plate worth by the rod drop method. (author)

  7. Chemical Looping Combustion of Hematite Ore with Methane and Steam in a Fluidized Bed Reactor

    Directory of Open Access Journals (Sweden)

    Samuel Bayham

    2017-08-01

    Full Text Available Chemical looping combustion is considered an indirect method of oxidizing a carbonaceous fuel, utilizing a metal oxide oxygen carrier to provide oxygen to the fuel. The advantage is the significantly reduced energy penalty for separating out the CO2 for reuse or sequestration in a carbon-constrained world. One of the major issues with chemical looping combustion is the cost of the oxygen carrier. Hematite ore is a proposed oxygen carrier due to its high strength and resistance to mechanical attrition, but its reactivity is rather poor compared to tailored oxygen carriers. This problem is further exacerbated by methane cracking, the subsequent deposition of carbon and the inability to transfer oxygen at a sufficient rate from the core of the particle to the surface for fuel conversion to CO2. Oxygen needs to be readily available at the surface to prevent methane cracking. The purpose of this work was to demonstrate the use of steam to overcome this issue and improve the conversion of the natural gas to CO2, as well as to provide data for computational fluid dynamics (CFD validation. The steam will gasify the deposited carbon to promote the methane conversion. This work studies the performance of hematite ore with methane and steam mixtures in a 5 cm fluidized bed up to approximately 140 kPa. Results show an increased conversion of methane in the presence of steam (from 20–45% without steam to 60–95% up to a certain point, where performance decreases. Adding steam allows the methane conversion to carbon dioxide to be similar to the overall methane conversion; it also helped to prevent carbon accumulation from occurring on the particle. In general, the addition of steam to the feed gas increased the methane conversion. Furthermore, the addition of steam caused the steam methane reforming reaction to form more hydrogen and carbon monoxide at higher steam and methane concentrations, which was not completely converted at higher concentrations and

  8. Hydrogen production from urea wastewater using a combination of urea thermal hydrolyser-desorber loop and a hydrogen-permselective membrane reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rahimpour, M.R.; Mottaghi, H.R.; Barmaki, M.M. [Department of Chemical Engineering, School of Chemical and Petroleum Engineering, Shiraz University, Shiraz71345 (Iran)

    2010-06-15

    This work presents novel application of palladium-based membrane in a wastewater treatment loop of urea plant for hydrogen production. Urea wastewater treatment loop is based on combined thermal hydrolysis-desorption operations. The wastewater of urea plant includes ammonia and urea which in the current treatment loop; urea decomposes to ammonia and carbon dioxide. The catalytic hydrogen-permselective membrane reactor is proposed for hydrogen production from desorbed ammonia of urea wastewater which much of it discharges to air and causes environmental pollution. Therefore hydrogen is produced from decomposition of ammonia on nickel-alumina catalyst bed simultaneously and permeates from reaction side to shell side through thin layer of palladium-silver membrane. Also a sweep gas is used in the shell side for increasing driving force. In this way, 4588 tons/yr hydrogen is produced and environmental problem of urea plant is solved. The membrane reactor and urea wastewater treatment loop are modeled mathematically and the predicted data of the model are consistent with the experimental and plant data that show validity of the model. Also the effects of key parameters on the performance of catalytic hydrogen-permselective membrane reactor such as the temperature, pressure, thickness of Pd-Ag layer, configuration of flow and sweep gas flow ratio were examined. (author)

  9. Operation of the NETL Chemical Looping Reactor with Natural Gas and a Novel Copper-Iron Material

    Energy Technology Data Exchange (ETDEWEB)

    Straub, Douglas [National Energy Technology Lab. (NETL), Morgantown, WV (United States); Bayham, Samuel [National Energy Technology Lab. (NETL), Morgantown, WV (United States); Weber, Justin [National Energy Technology Lab. (NETL), Morgantown, WV (United States)

    2017-02-21

    The proposed Clean Power Plan requires CO2 emission reductions of 30% by 2030 and further reductions are targeted by 2050. The current strategies to achieve the 30% reduction targets do not include options for coal. However, the 2016 Annual Energy Outlook suggests that coal will continue to provide more electricity than renewable sources for many regions of the country in 2035. Therefore, cost effective options to reduce greenhouse gas emissions from fossil fuel power plants are vital in order to achieve greenhouse gas reduction targets beyond 2030. As part of the U.S. Department of Energy’s Advanced Combustion Program, the National Energy Technology Laboratory’s Research and Innovation Center (NETL R&IC) is investigating the feasibility of a novel combustion concept in which the GHG emissions can be significantly reduced. This concept involves burning fuel and air without mixing these two reactants. If this concept is technically feasible, then CO2 emissions can be significantly reduced at a much lower cost than more conventional approaches. This indirect combustion concept has been called Chemical Looping Combustion (CLC) because an intermediate material (i.e., a metal-oxide) is continuously cycled to oxidize the fuel. This CLC concept is the focus of this research and will be described in more detail in the following sections. The solid material that is used to transport oxygen is called an oxygen carrier material. The cost, durability, and performance of this material is a key issue for the CLC technology. Researchers at the NETL R&IC have developed an oxygen carrier material that consists of copper, iron, and alumina. This material has been tested extensively using lab scale instruments such as thermogravimetric analysis (TGA), scanning electron microscopy (SEM), mechanical attrition (ASTM D5757), and small fluidized bed reactor tests. This report will describe the results from a realistic, circulating, proof-of-concept test that was

  10. Operational experiences and coping with ageing effects of the IRT-Sofia research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Krezhov, K. [Inst. for Nuclear Research and Nuclear Energy, Sofia (Bulgaria)

    1995-12-31

    The present paper gives a review on the efforts to cope with reactor equipment ageing effects and describes the major experience gained in the maintenance work necessary to keep the reactor in good condition throughout the years. Also, a short description of the modernization project with preserved reactor power level of 2 MW is given. (orig.)

  11. 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.

  12. COUNTERCURRENT FLOW LIMITATION EXPERIMENTS AND MODELING FOR IMPROVED REACTOR SAFETY

    Energy Technology Data Exchange (ETDEWEB)

    Vierow, Karen

    2008-09-26

    This project is investigating countercurrent flow and “flooding” phenomena in light water reactor systems to improve reactor safety of current and future reactors. To better understand the occurrence of flooding in the surge line geometry of a PWR, two experimental programs were performed. In the first, a test facility with an acrylic test section provided visual data on flooding for air-water systems in large diameter tubes. This test section also allowed for development of techniques to form an annular liquid film along the inner surface of the “surge line” and other techniques which would be difficult to verify in an opaque test section. Based on experiences in the air-water testing and the improved understanding of flooding phenomena, two series of tests were conducted in a large-diameter, stainless steel test section. Air-water test results and steam-water test results were directly compared to note the effect of condensation. Results indicate that, as for smaller diameter tubes, the flooding phenomena is predominantly driven by the hydrodynamics. Tests with the test sections inclined were attempted but the annular film was easily disrupted. A theoretical model for steam venting from inclined tubes is proposed herein and validated against air-water data. Empirical correlations were proposed for air-water and steam-water data. Methods for developing analytical models of the air-water and steam-water systems are discussed, as is the applicability of the current data to the surge line conditions. This report documents the project results from July 1, 2005 through June 30, 2008.

  13. Spatial and temporal variations of microbial community in a mixed plug-flow loop reactor fed with dairy manure.

    Science.gov (United States)

    Li, Yueh-Fen; Chen, Po-Hsu; Yu, Zhongtang

    2014-07-01

    Mixed plug-flow loop reactor (MPFLR) has been widely adopted by the US dairy farms to convert cattle manure to biogas. However, the microbiome in MPFLR digesters remains unexplored. In this study, the microbiome in a MPFLR digester operated on a mega-dairy farm was examined thrice over a 2 month period. Within 23 days of retention time, 55-70% of total manure solid was digested. Except for a few minor volatile fatty acids (VFAs), total VFA concentration and pH remained similar along the course of the digester and over time. Metagenomic analysis showed that although with some temporal variations, the bacterial community was rather stable spatially in the digester. The methanogenic community was also stable both spatially and temporally in the digester. Among methanogens, genus Methanosaeta dominated in the digester. Quantitative polymerase chain reaction (qPCR) analysis and metagenomic analysis yielded different relative abundance of individual genera of methanogens, especially for Methanobacterium, which was predominant based on qPCR analysis but undetectable by metagenomics. Collectively, the results showed that only small microbial and chemical gradients existed within the digester, and the digestion process occurred similarly throughout the MPFLR digester. The findings of this study may help improve the operation and design of this type of manure digesters. © 2014 The Authors. Microbial Biotechnology published by John Wiley & Sons Ltd and Society for Applied Microbiology.

  14. High purity H2 by sorption-enhanced chemical looping reforming of waste cooking oil in a packed bed reactor.

    Science.gov (United States)

    Pimenidou, P; Rickett, G; Dupont, V; Twigg, M V

    2010-12-01

    High purity hydrogen (>95%) was produced at 600 degrees C and 1 atm by steam reforming of waste cooking oil at a molar steam to carbon ratio of 4 using chemical looping, a process that features redox cycles of a Ni catalyst with the in-situ carbonation/calcination of a CO(2) sorbent (dolomite) in a packed bed reactor under alternated feedstreams of fuel-steam and air. The fuel and steam conversion were higher with the sorbent present than without it. Initially, the dolomite carbonation was very efficient (100%), and 98% purity hydrogen was produced, but the carbonation decreased to around 56% with a purity of 95% respectively in the following cycles. Reduction of the nickel catalyst occurred alongside steam reforming, water gas shift and carbonation, with H(2) produced continuously under fuel-steam feeds. Catalyst and CO(2)-sorbent regeneration was observed, and long periods of autothermal operation within each cycle were demonstrated. 2010 Elsevier Ltd. All rights reserved.

  15. A Loop Current experiment: Field and remote measurements

    Science.gov (United States)

    Hamilton, Peter; Lugo-Fernández, Alexis; Sheinbaum, Julio

    2016-12-01

    An overview of a new comprehensive observational study of the Loop Current (LC) in the eastern Gulf of Mexico that encompassed full-depth and near-bottom moorings, pressure-equipped inverted echo sounders (PIES) and remote sensing is presented. The study array was designed to encompass the LC from the Campeche Bank to the west Florida escarpment. This overview centers about principal findings as they pertain to mesoscale dynamics. Two companion papers provide in-depth analyses. Three LC anticyclonic eddy separation events were observed with good 3D spatial coverage over the 2½ year extent of the field study; the three separations exhibited similar processes after the LC had extended into the eastern Gulf. Large scale (∼300 km wavelength, 40-60 day periods) southward propagating meanders developed on the eastern side of the LC over deep (∼3000 m) water that were the result of baroclinic instability between the upper layer meandering jet and lower layer cyclones and anticyclones. The lower layer was only highly energetic during relatively short (∼2-3 months) intervals just prior to or during eddy detachments because of baroclinic instability. The steepening of the meanders lead to a pinch-off of LC eddies. The deep lower-layer eddies, constrained by the closed topography of the southeastern Gulf, propagated westward across the detachment zone and appear to assist in achieving separation. Small scale (∼50-100 km, periods ∼10 days) frontal eddies, observed on the western side of the LC along the Campeche Bank slope, decay over the deep water of the northern part of an extended LC, and have little influence on lower layer eddies, the east side meanders and the eddy detachment processes.

  16. Progress report on neutron beam experiments at Dalat Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nguyen Canh Hai; Tran Tuan Anh [Nuclear Physics Department, Nuclear Research Institute, Dalat (Viet Nam)

    2000-10-01

    The conduct and the utilizations of neutron beams at Dalat Nuclear Research Reactor was reported. In 1998 and 1999 the filtered thermal neutron beam at the beam tube using substances Si, Ti, C, Pb was extracted. The investigations on physical characteristics of reactor; neutron spectra and fluxes at beam tube; safety conditions have been carried out by calculations and experiments. The physical characteristics for the purposes of prompt gamma neutron activation analysis (PGNAA) and nuclear data measurement were improved. The delayed neutron analysis method is used to detect the neutron emission fragments produced by neutron irradiation of uranium and thorium. This method is to determine the concentrations of uranium and thorium simultaneously and detect 10{sup -6} g of uranium and 8x10{sup -6} g thorium in 10 g sample with a precision of 8 per cent. Beside the delayed neutron analysis facility, a gamma spectrometer system with HP-Ge 90 cm{sup 3} semiconductor detector was installed at pneumatic transfer for cyclic activation analysis (CAA). The CAA method has given analytical results quickly and sensitively for the isotopes with half-lives in order of seconds to minutes. (author)

  17. Experimental and analytical studies of merging plasma loops on the Caltech solar loop experiment

    Science.gov (United States)

    Pitigoi-Aron, Gabriela

    and personal factors and perceptions with emphasis on mentors' influence; (5) Negative influence of salary difference with respect to private practitioners. The findings of this study were similar to the available studies on foreign-trained dentists and to most of the studies already done on domestically trained dentists. The major factors found were comparable with the up-to-date literature. The elevated research drive, the intellectual challenges, the work environment, the desire to teach, and the mentors' influence were among those which mirrored almost perfectly the other studies. Some fine differences were found for foreign-trained dentists, such as a lighter financial burden caused by smaller student debt and the irrelevance of military practice experience. The study provides a number of suggestions for enhancing the recruiting and retaining process for dental academia: (1) Support and enhance the research capacity of dental schools; (2) Create structures to develop mentors; (3) Invest to build prestige; (4) Find creative ways to offset lower salaries; (5) Foster a pleasant academic working environment; (6) Use international activities to recruit international dentists. The study revealed factors that have been influential in participants' decisions to choose an academic career, in general and at Pacific. It is hoped that this study will be a useful reference in the increasingly difficult endeavor of adding and retaining world-class dental educators.

  18. On possibility of realization NEUTRINO-4 experiment on search for oscillations of the reactor antineutrino into a sterile state

    OpenAIRE

    Serebrov, A. P.; Fomin, A. K.; Zinoviev, V. G.; Ivochkin, V. G.; Loginov, Yu. E.; Petrov, G. A.; Solovey, V. A.; Chernyi, A. V.; Zherebtsov, O. M.; Samoylov, R. M.; Martemyanov, V. P.; Tsinoev, V. G.; Tarasenkov, V. G.; Aleshin, V. I.; Petelin, A. L.

    2013-01-01

    One has investigated possibility of performing NEUTRINO-4 experiment on search for reactor neutrino oscillations into a sterile state at research reactors. The simulated experiment has been conducted at 16 MW reactor WWR-M in PNPI with the purpose of implementing a full scale experiment with the help of 100 MW reactor SM-3 in RIAR. Background conditions for making such an experiment have been examined at both reactors. The conclusion has been made on possible implementation of a full scale ex...

  19. Resolving the Reactor Neutrino Anomaly with the KATRIN Neutrino Experiment

    CERN Document Server

    Barrett, J A Formaggio J

    2011-01-01

    The KArlsruhe TRItium Neutrino experiment (KATRIN) combines an ultra-luminous molecular tritium source with an integrating high-resolution spectrometer to gain sensitivity to the absolute mass scale of neutrinos. The projected sensitivity of the experiment on the electron neutrino mass is 200 meV at 90% C.L. With such unprecedented resolution, the experiment is also sensitive to physics beyond the Standard Model, particularly to the existence of additional sterile neutrinos at the eV mass scale. A recent analysis of available reactor data appears to favor the existence of such such a sterile neutrino with a mass splitting of $|\\Delta m_{\\rm sterile}|^2 \\ge 1.5$ eV$^2$ and mixing strength of $\\sin^2{2\\theta_{\\rm sterile}} = 0.17\\pm 0.08$ at 95% C.L. Upcoming tritium beta decay experiments should be able to rule out or confirm the presence of the new phenomenon for a substantial fraction of the allowed parameter space.

  20. Status of the JUNO reactor anti-neutrino experiment

    Science.gov (United States)

    Lu, Haoqi; JUNO Collaboration

    2017-06-01

    The Jiangmen Underground Neutrino Observatory (JUNO) is a reactor antineutrino experiment with the aim to determine the neutrino mass hierarchy. The detector will be filled with 20 kilotons of liquid scintillator and instrumented with 18000 20-inch PMTs to achieve an unprecedented energy resolution of 3%@1 MeV. A 35.4 m diameter acrylic sphere will be built as a liquid scintillator vessel.The detector will be constructed in a 700-m-deep-underground laboratory to reduce cosmogenic muon flux. An external veto cosisting of a water Cherenkov detector and a top tracker will be used for cosmogenic muon detection and background reduction. The mass hierarchy sensitivity is expected to reach 3-4σ after 6 years of data taking. Civil construction and detector R&D are underway. Data taking is expected to start in 2020.

  1. Reduced-scale water test of natural circulation for decay heat removal in loop-type sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Murakami, T., E-mail: murakami@criepi.denken.or.jp [Central Research Institute of Electric Power Industry, 1646 Abiko, Chiba (Japan); Eguchi, Y., E-mail: eguchi@criepi.denken.or.jp [Central Research Institute of Electric Power Industry, 1646 Abiko, Chiba (Japan); Oyama, K., E-mail: kazuhiro_oyama@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 2-34-17 Jinguumae, Shibuya, Tokyo (Japan); Watanabe, O., E-mail: osamu4_watanabe@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 2-34-17 Jinguumae, Shibuya, Tokyo (Japan)

    2015-07-15

    Highlights: • The natural circulation characteristics of a loop-type SFR are examined by a water test. • The performance of decay heat removal system is evaluated using a similarity law. • The effects of flow deviation in the parallel piping of a primary loop are clarified. • The reproducibility of the natural circulation test is confirmed. - Abstract: Water tests of a loop-type sodium-cooled fast reactor have been conducted to physically evaluate the natural circulation characteristics. The water test apparatus was manufactured as a 1/10-scale mock-up of the Japan Sodium-Cooled Fast Reactor, which adopts a decay heat removal system (DHRS) utilizing natural circulation. Tests simulating a variety of events and operation conditions clarified the thermal hydraulic characteristics and core-cooling performance of the natural circulation in the primary loop. Operation conditions such as the duration of the pump flow coast-down and the activation time of the DHRS affect the natural circulation characteristics. A long pump flow coast-down cools the upper plenum of the reactor vessel (RV). This causes the loss of the buoyant force in the RV. The test result indicates that a long pump flow coast-down tends to result in a rapid increase in the core temperature because of the loss of the buoyant force. The delayed activation of the DHRS causes a decrease in the natural circulation flow rate and a temperature rise in the RV. Flow rate deviation and a reverse flow appear in the parallel cold-leg piping in some events, which cause thermal stratification in the cold-leg piping. The DHRS prevents the core temperature from fatally rise even for the most severe design-basis event, in which sodium leakage in a secondary loop of the DHRS and the opening failure of a single damper of the air cooler occur simultaneously. In the water test for the case of siphon break in the primary loop, which is one of the design extension conditions, a circulation flow consisting of ascendant

  2. Rotary Bed Reactor for Chemical-Looping Combustion with Carbon Capture. Part 2: Base Case and Sensitivity Analysis

    KAUST Repository

    Zhao, Zhenlong

    2013-01-17

    Part 1 (10.1021/ef3014103) of this series describes a new rotary reactor for gas-fueled chemical-looping combustion (CLC), in which, a solid wheel with microchannels rotates between the reducing and oxidizing streams. The oxygen carrier (OC) coated on the surfaces of the channels periodically adsorbs oxygen from air and releases it to oxidize the fuel. A one-dimensional model is also developed in part 1 (10.1021/ef3014103). This paper presents the simulation results based on the base-case design parameters. The results indicate that both the fuel conversion efficiency and the carbon separation efficiency are close to unity. Because of the relatively low reduction rate of copper oxide, fuel conversion occurs gradually from the inlet to the exit. A total of 99.9% of the fuel is converted within 75% of the channel, leading to 25% redundant length near the exit, to ensure robustness. In the air sector, the OC is rapidly regenerated while consuming a large amount of oxygen from air. Velocity fluctuations are observed during the transition between sectors because of the complete reactions of OCs. The gas temperature increases monotonically from 823 to 1315 K, which is mainly determined by the solid temperature, whose variations with time are limited within 20 K. The overall energy in the solid phase is balanced between the reaction heat release, conduction, and convective cooling. In the sensitivity analysis, important input parameters are identified and varied around their base-case values. The resulting changes in the model-predicted performance revealed that the most important parameters are the reduction kinetics, the operating pressure, and the feed stream temperatures. © 2012 American Chemical Society.

  3. Study on the Neutrino Oscillation with a Next Generation Medium-Baseline Reactor Experiment

    Directory of Open Access Journals (Sweden)

    Chang Dong Shin

    2014-01-01

    Full Text Available For over fifty years, reactor experiments have played an important role in neutrino physics, in both discoveries and precision measurements. One of the methods to verify the existence of neutrino is the observation of neutrino oscillation phenomena. Electron antineutrinos emitted from a reactor provide the measurement of the small mixing angle θ13, providing rich programs of neutrino properties, detector development, nuclear monitoring, and application. Using reactor neutrinos, future reactor neutrino experiments, more precise measurements of θ12,  Δm122, and mass hierarchy will be explored. The precise measurement of θ13 would be crucial for measuring the CP violation parameters at accelerators. Therefore, reactor neutrino physics will assist in the complete understanding of the fundamental nature and implications of neutrino masses and mixing. In this paper, we investigated several characteristics of RENO-50, which is a future medium-baseline reactor neutrino oscillation experiment, by using the GloBES simulation package.

  4. 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.

  5. 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.

  6. Study on the Neutrino Oscillation with a Next Generation Medium-Baseline Reactor Experiment

    OpenAIRE

    Chang Dong Shin; Kyung Kwang Joo

    2014-01-01

    For over fifty years, reactor experiments have played an important role in neutrino physics, in both discoveries and precision measurements. One of the methods to verify the existence of neutrino is the observation of neutrino oscillation phenomena. Electron antineutrinos emitted from a reactor provide the measurement of the small mixing angle θ13 , providing rich programs of neutrino properties, detector development, nuclear monitoring, and application. Using reactor neutrinos, future reacto...

  7. 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)

  8. Experiences of closed-loop insulin delivery among pregnant women with Type 1 diabetes.

    Science.gov (United States)

    Farrington, C; Stewart, Z A; Barnard, K; Hovorka, R; Murphy, H R

    2017-10-01

    To explore the experiences of pregnant women with Type 1 diabetes, and the relationships between perceptions of glucose control, attitudes to technology and glycaemic responses with regard to closed-loop insulin delivery. We recruited 16 pregnant women with Type 1 diabetes [mean ± sd age 34.1 ± 4.6 years, duration of diabetes 23.6 ± 7.2 years, baseline HbA 1c 51±5 mmol/mol (6.8 ± 0.6%)] to a randomized crossover trial of sensor-augmented pump therapy vs automated closed-loop therapy. Questionnaires (Diabetes Technology Questionnaire, Hypoglycaemia Fear Survey) were completed before and after each intervention, with qualitative interviews at baseline and follow-up. Women described the benefits and burdens of closed-loop systems during pregnancy. Feelings of improved glucose control, excitement and empowerment were counterbalanced by concerns about device visibility, obsessive data checking and diminished attentiveness to hyper- and hypoglycaemia symptoms. Responding to questionnaires, eight participants felt less worry about overnight hypoglycaemia and that diabetes 'did not run their lives'; however, five reported that closed-loop increased time thinking about diabetes, and three felt it made sleep and preventing hyperglycaemia more problematic. Women slightly overestimated their glycaemic response to closed-loop therapy. Most became more positive in their technology attitudes throughout pregnancy. Women with more positive technology attitudes had higher degrees of overestimation, and poorer levels of glycaemic control. Women displayed complex psychosocial responses to closed-loop therapy in pregnancy. Perceptions of glycaemic response may diverge from biomedical data. © 2017 Diabetes UK.

  9. FOEHN: The critical experiment for the Franco-German High Flux Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Scharmer, K.; Eckert, H. G. [Kernforschungszentrum Karlsruhe GmbH (Germany)

    1991-01-01

    A critical experiment for the Franco-German High Flux Reactor was carried out in the French reactor EOLE (CEN Cadarache). The purpose of the experiment was to check the calculation methods in a realistic geometry and to measure effects that can only be calculated imprecisely (e.g. beam hole effects). The structure of the experiment and the measurement and calculation methods are described. A detailed comparison between theoretical and experimental results was performed. 30 refs., 105 figs.

  10. Recent performance experience with US light water reactor self-actuating safety and relief valves

    Energy Technology Data Exchange (ETDEWEB)

    Hammer, C.G.

    1996-12-01

    Over the past several years, there have been a number of operating reactor events involving performance of primary and secondary safety and relief valves in U.S. Light Water Reactors. There are several different types of safety and relief valves installed for overpressure protection of various safety systems throughout a typical nuclear power plant. The following discussion is limited to those valves in the reactor coolant systems (RCS) and main steam systems of pressurized water reactors (PWR) and in the RCS of boiling water reactors (BWR), all of which are self-actuating having a setpoint controlled by a spring-loaded disk acting against system fluid pressure. The following discussion relates some of the significant recent experience involving operating reactor events or various testing data. Some of the more unusual and interesting operating events or test data involving some of these designs are included, in addition to some involving a number of similar events and those which have generic applicability.

  11. Effect of Spray System on Fission Product Distribution in Containment During a Severe Accident in a Two-Loop Pressurized Water Reactor

    Directory of Open Access Journals (Sweden)

    Mehdi Dehjourian

    2016-08-01

    Full Text Available The containment response during the first 24 hours of a low-pressure severe accident scenario in a nuclear power plant with a two-loop Westinghouse-type pressurized water reactor was simulated with the CONTAIN 2.0 computer code. The accident considered in this study is a large-break loss-of-coolant accident, which is not successfully mitigated by the action of safety systems. The analysis includes pressure and temperature responses, as well as investigation into the influence of spray on the retention of fission products and the prevention of hydrogen combustion in the containment.

  12. System-Level Heat Transfer Analysis, Thermal- Mechanical Cyclic Stress Analysis, and Environmental Fatigue Modeling of a Two-Loop Pressurized Water Reactor. A Preliminary Study

    Energy Technology Data Exchange (ETDEWEB)

    Mohanty, Subhasish [Argonne National Lab. (ANL), Argonne, IL (United States); Soppet, William [Argonne National Lab. (ANL), Argonne, IL (United States); Majumdar, Saurin [Argonne National Lab. (ANL), Argonne, IL (United States); Natesan, Ken [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-01-03

    This report provides an update on an assessment of environmentally assisted fatigue for light water reactor components under extended service conditions. This report is a deliverable in April 2015 under the work package for environmentally assisted fatigue under DOE's Light Water Reactor Sustainability program. In this report, updates are discussed related to a system level preliminary finite element model of a two-loop pressurized water reactor (PWR). Based on this model, system-level heat transfer analysis and subsequent thermal-mechanical stress analysis were performed for typical design-basis thermal-mechanical fatigue cycles. The in-air fatigue lives of components, such as the hot and cold legs, were estimated on the basis of stress analysis results, ASME in-air fatigue life estimation criteria, and fatigue design curves. Furthermore, environmental correction factors and associated PWR environment fatigue lives for the hot and cold legs were estimated by using estimated stress and strain histories and the approach described in NUREG-6909. The discussed models and results are very preliminary. Further advancement of the discussed model is required for more accurate life prediction of reactor components. This report only presents the work related to finite element modelling activities. However, in between multiple tensile and fatigue tests were conducted. The related experimental results will be presented in the year-end report.

  13. Validation of the TRACE code for the system dynamic simulations of the molten salt reactor experiment and the preliminary study on the dual fluid molten salt reactor

    Energy Technology Data Exchange (ETDEWEB)

    He, Xun

    2016-06-14

    Molten Salt Reactor (MSR), which was confirmed as one of the six Generation IV reactor types by the GIF (Generation IV International Forum in 2008), recently draws a lot of attention all around the world. Due to the application of liquid fuels the MSR can be regarded as the most special one among those six GEN-IV reactor types in a sense. A unique advantage of using liquid nuclear fuel lies in that the core melting accident can be thoroughly eliminated. Besides, a molten salt reactor can have several fuel options, for instance, the fuel can be based on {sup 235}U, {sup 232}Th-{sup 233}U, {sup 238}U-{sup 239}Pu cycle or even the spent nuclear fuel (SNF), so the reactor can be operated as a breeder or as an actinides burner both with fast, thermal or epi-thermal neutron spectrum and hence, it has excellent features of the fuel sustainability and for the non-proliferation. Furthermore, the lower operating pressure not only means a lower risk of the explosion as well as the radioactive leakage but also implies that the reactor vessel and its components can be lightweight, thus lowering the cost of equipments. So far there is no commercial MSR being operated. However, the MSR concept and its technical validation dates back to the 1960s to 1970s, when the scientists and engineers from ORNL (Oak Ridge National Laboratory) in the United States managed to build and run the world's first civilian molten salt reactor called MSRE (Molten Salt Reactor Experiment). The MSRE was an experimental liquid-fueled reactor with 10 MW thermal output using {sup 4}LiF-BeF{sub 2}-ZrF{sub 4}-UF{sub 4} as the fuel also as the coolant itself. The MSRE is usually taken as a very important reference case for many current researches to validate their codes and simulations. Without exception it works also as a benchmark for this thesis. The current thesis actually consists of two main parts. The first part is about the validation of the current code for the old MSRE concept, while the second

  14. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover

    2009-09-01

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

  16. An alternative method of determining the neutrino mass ordering in reactor neutrino experiments

    Directory of Open Access Journals (Sweden)

    S.M. Bilenky

    2017-09-01

    Full Text Available We discuss a novel alternative method of determining the neutrino mass ordering in medium baseline experiments with reactor antineutrinos. Results on the potential sensitivity of the new method are also presented.

  17. An alternative method of determining the neutrino mass ordering in reactor neutrino experiments

    Science.gov (United States)

    Bilenky, S. M.; Capozzi, F.; Petcov, S. T.

    2017-09-01

    We discuss a novel alternative method of determining the neutrino mass ordering in medium baseline experiments with reactor antineutrinos. Results on the potential sensitivity of the new method are also presented.

  18. Light water reactor fuel response during reactivity initiated accident experiments

    Energy Technology Data Exchange (ETDEWEB)

    MacDonald, P. E.; McCardell, R. K.; Martinson, Z. R.; Seiffert, S. L.

    1979-01-01

    Experimental results from six recent Power Burst Facility (PBF) reactivity initiated accident (RIA) tests are compared with data from previous Special Power Excursion Reactor Test (SPERT), and Japanese Nuclear Safety Research Reactor (NSRR) tests. The RIA fuel behavior experimental program recently started in the PBF is being conducted with coolant conditions typical of hot-startup conditions in a commercial boiling water reactor. The SPERT and NSRR test programs investigated the behavior of single or small clusters of light water reactor (LWR) type fuel rods under approximate room temperature and atmospheric pressure conditions in capsules containing stagnant water. As observed in the SPERT and NSRR tests, energy deposition, and consequent enthalpy increase in the PBF test fuel, appears to be the single most important variable. However, the consequences of failure at boiling water hot-startup system conditions appear to be more severe than previously observed in either the stagnant capsule SPERT or NSRR tests. Metallographic examination of both previously unirradiated and irradiated PBF fuel rod cross sections revealed extensive variation in cladding wall thicknesses (involving considerable plastic flow) and fuel shattering along grain boundaries in both restructured and unrestructured fuel regions. Oxidation of the cladding resulted in fracture at the location of cladding thinning and disintegration of the rods during quench. In addition,swelling of the gaseous and potentially volatile fission products in previously irradiated fuel resulted in volume increases of up to 180% and blockage of the coolant channels within the flow shrouds surrounding the fuel rods.

  19. Rate-Only analysis with reactor-off data in the Double Chooz experiment

    CERN Document Server

    Novella, P

    2013-01-01

    Among ongoing reactor-based experiments, Double Chooz is unique in obtaining data when the reactor cores are brought down for maintenance. These reactor-off data allow for a clean measurement of the backgrounds of the experiment, thus being of uppermost importance for the theta13 oscillation analysis. While the oscillation results published by the collaboration in 2011 and 2012 rely on background models derived from reactor-on data, in this talk we present an independent study based on the handle provided by 7.53 days of reactor-off data. A global fit to both theta13 and the total background is performed by analyzing the observed neutrino rate as a function of the non-oscillated expected rate for different reactor power conditions. The result presented in this talk is fully consistent with the one already published by Double Chooz. As they both yield almost the same precision, this work stands as a prove of the reliability of the background estimates and the oscillation analysis of the experiment.

  20. Belgian experience in applying the {open_quotes}leak-before-break{close_quotes} concept to the primary loop piping

    Energy Technology Data Exchange (ETDEWEB)

    Gerard, R.; Malekian, C.; Meessen, O. [Tractebel Energy Engineering, Brussels (Belgium)

    1997-04-01

    The Leak Before Break (LBB) concept allows to eliminate from the design basis the double-ended guillotine break of the primary loop piping, provided it can be demonstrated by a fracture mechanics analysis that a through-wall flaw, of a size giving rise to a leakage still well detectable by the plant leak detection systems, remains stable even under accident conditions (including the Safe Shutdown Earthquake (SSE)). This concept was successfully applied to the primary loop piping of several Belgian Pressurized Water Reactor (PWR) units, operated by the Utility Electrabel. One of the main benefits is to permit justification of supports in the primary loop and justification of the integrity of the reactor pressure vessel and internals in case of a Loss Of Coolant Accident (LOCA) in stretch-out conditions. For two of the Belgian PWR units, the LBB approach also made it possible to reduce the number of large hydraulic snubbers installed on the primary coolant pumps. Last but not least, the LBB concept also facilitates the steam generator replacement operations, by eliminating the need for some pipe whip restraints located close to the steam generator. In addition to the U.S. regulatory requirements, the Belgian safety authorities impose additional requirements which are described in details in a separate paper. An novel aspect of the studies performed in Belgium is the way in which residual loads in the primary loop are taken into account. Such loads may result from displacements imposed to close the primary loop in a steam generator replacement operation, especially when it is performed using the {open_quote}two cuts{close_quotes} technique. The influence of such residual loads on the LBB margins is discussed in details and typical results are presented.

  1. Experiences with austenitic steels in boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Wachter, O. [Preussische Elektrizitaets-AG (Preussenelektra), Hannover (Germany); Bruemmer, G. [Hamburgische Electricitaets-Werke AG., Hamburg (Germany)

    1997-05-01

    Stabilized austenitic steels are susceptible to intergranular stress corrosion cracking (IGSCC) under boiling water reactor (BWR) conditions. This important finding for the German nuclear power station industry arises from the detection of cracks during the last 3 years in reactor hot water pipes made from titanium-stabilized steel AISI 321 in six BWRs and in reactor core components made from the niobium-stabilized steel AISI 347 in one BWR. All the observed cracks had a common feature: they had their origin in the chromium carbide precipitates at the grain boundaries and in the associated chromium-depleted region near the grain boundary. These microstructural features in the heat-affected zones of the hot water pipe weldments were caused by the heat input during deposition of the root bead. The TiC partially dissolved in the region near the fusion line and the released carbon reacted to form chromium-rich M{sub 23}C{sub 6}. Regarding the cracks found in the core shroud and the core grid plates, it was shown that a sensitizing heat treatment of rings taken from the same heat of steel could give rise to a microstructure susceptible to IGSCC in the region of a weldment. High carbon contents coupled with low stabilization ratios led to sensitization. Residual stresses developed during welding provided the significant contributions to the tensile stress necessary for IGSCC. With regard to the service medium, the influence of the electrochemical corrosion potential (ECP) was recognized as a dominant factor, together with the conductivity. The corrosion potential was mainly determined by the radiolytic formation of H{sub 2}O{sub 2}; with increasing distance from the core, the H{sub 2}O{sub 2} content decreased owing to catalytic decomposition. For the pipes the problem of IGSCC could be resolved by the use of optimized steel (lower carbon content with maximum allowable stabilization ratio).

  2. Indication for the disappearance of reactor electron antineutrinos in the Double Chooz experiment

    CERN Document Server

    Abe, Y; Akiri, T; Anjos, J C dos; Ardellier, F; Barbosa, A F; Baxter, A; Bernstein, A; Bezerra, T J C; Bezrukhov, L; Blucher, E; Bongrand, M; Bowden, N S; Buck, C; Busenitz, J; Cabrera, A; Caden, E; Camilleri, L; Carr, R; Cerrada, M; Chang, P -J; Chimenti, P; Classen, T; Collin, A; Conover, E; Conrad, J M; Cormon, S; Crespo-Anadón, J I; Cribier, M; Crum, K; Cucoanes, A; D'Agostino, M V; Damon, E; Dawson, J V; Dazeley, S; Dierckxsens, M; Dietrich, D; Djurcic, Z; Dracos, M; Durand, V; Efremenko, Y; Endo, Y; Etenko, A; Falk, E; Fallot, M; Fechner, M; von Feilitzsch, F; Felde, J; Fernandes, S M; Franco, D; Franke, A; Franke, M; Furuta, H; Gama, R; Gil-Botella, I; Giot, L; Göger-Neff, M; Gonzalez, L F G; Goodman, M C; Goon, J TM; Greiner, D; Guillon, B; Haag, N; Hagner, C; Hara, T; Hartmann, F X; Hartnell, J; Haruna, T; Haser, J; Hatzikoutelis, A; Hayakawa, T; Hofmann, M; Horton-Smith, G; Ishitsuka, M; Jochum, J; Jollet, C; Jones, C L; Kaether, F; Kalousis, L; Kamyshkov, Y; Kaplan, D; Kawasaki, T; Keefer, G; Kemp, E; de Kerret, H; Kibe, Y; Konno, T; Kryn, D; Kuze, M; Lachenmaier, T; Lane, C E; Langbrandtner, C; Lasserre, T; Letourneau, A; Lhuillier, D; Lima, H P; Lindner, M; Liu, Y; López-Castanõ, J M; LoSecco, J M; Lubsandorzhiev, B K; Lucht, S; McKee, D; Maeda, J; Maesano, C N; Mariani, C; Maricic, J; Martino, J; Matsubara, T; Mention, G; Meregaglia, A; Miletic, T; Milincic, R; Milzstajn, A; Miyata, H; Motta, D; Mueller, Th A; Nagasaka, Y; Nakajima, K; Novella, P; Obolensky, M; Oberauer, L; Onillon, A; Osborn, A; Ostrovskiy, I; Palomares, C; Peeters, S J M; Pepe, I M; Perrin, P; Pfahler, P; Porta, A; Potzel, W; Queval, R; Reichenbacher, J; Reinhold, B; Remoto, A; Reyna, D; Röhling, M; Roth, S; Rubin, H A; Sakamoto, Y; Santorelli, R; Sato, F; Schönert, S; Schoppmann, S; Schwan, U; Schwetz, T; Shaevitz, M; Shrestha, D; Sida, J-L; Sinev, V; Skorokhvatov, M; Smith, E; Stahl, A; Stancu, I; Strait, M; Stüken, A; Suekane, F; Sukhotin, S; Sumiyoshi, T; Sun, Y; Sun, Z; Svoboda, R; Tabata, H; Tamura, N; Terao, K; Tonazzo, A; Toups, M; Thi, H H Trinh; Veyssiere, C; Vignaud, D; Wagner, S; Watanabe, H; White, B; Wiebusch, C; Winslow, L; Worcester, M; Wurm, M; Yanovitch, E; Yermia, F; Zbiri, K; Zimmer, V

    2011-01-01

    The Double Chooz Experiment presents an indication of reactor electron antineutrino disappearance consistent with neutrino oscillations. A ratio of 0.944 $\\pm$ 0.016 (stat) $\\pm$ 0.040 (syst) observed to predicted events was obtained in 101 days of running at the Chooz Nuclear Power Plant in France, with two 4.25 GW$_{th}$ reactors. The results were obtained from a single 10 m$^3$ fiducial volume detector located 1050 m from the two reactor cores. The reactor antineutrino flux prediction used the Bugey4 measurement as an anchor point. The deficit can be interpreted as an indication of a non-zero value of the still unmeasured neutrino mixing parameter \\sang. Analyzing both the rate of the prompt positrons and their energy spectrum we find \\sang = 0.086 $\\pm$ 0.041 (stat) $\\pm$ 0.030 (syst), or, at 90% CL, 0.015 $<$ \\sang $\\ <$ 0.16.

  3. Study for Reactor Monitoring using Anti-neutrino Detection in the Neos experiment

    Energy Technology Data Exchange (ETDEWEB)

    Han, Bo Young; Sun, Gwang Min [KAERI, Daejeon (Korea, Republic of); Jeon, Eun Ju [ISB, Daejeon (Korea, Republic of); and others

    2016-05-15

    In this study we describe a feasibility study of reactor monitoring using antineutrino detection in the Neutrino Experiment for Oscillation at Short baseline (NEOS) at Hanbit power plant. Recently, in the perspective of nonproliferation issues and misuse of nuclear energy as a fast-growing nuclear energy industry, the application of anti-neutrino measurement has been proposed and the feasibility studies has been carried out as a novel technology for monitoring the burning process of nuclear power reactor. The NEOS detector with 1000 L Gd-doped liquid scintillator was installed in tendon gallery at Hanbit power station unit 5 and has been collecting close to 2000 IBD events per day with the signal to noise ratio of ∼ 20. As a preliminary result, we demonstrate the possibility of monitoring nuclear power reactor with the IBD counting rate during reactor power ON, ramping up, and OFF.

  4. A High Precision Reactor Neutrino Detector for the Double Chooz Experiment

    CERN Document Server

    Suekane, Fumihiko

    2009-01-01

    Double Chooz is a reactor neutrino experiment which investigates the last neutrino mixing angle; theta-13. It is necessary to measure reactor neutrino disappearance with precision 1% or better to detect finite value of theta-13. This requirement is the most strict compared to other reactor neutrino experiments performed so far. The Double Chooz experiment makes use of a number of techniques to reduce the possible errors to achieve the sensitivity. The detector is now under construction and it is expected to take first neutrino data in 2009 and to measure sin^22theta-13 with a sensitivity of 0.03 (90%C.L.) In this proceedings, the technical concepts of Double Chooz detector are explained stressing on how it copes with the systematic errors.

  5. Application of a Virtual Reactivity Feedback Control Loop in Non-Nuclear Testing of a Fast Spectrum Reactor

    Science.gov (United States)

    Bragg-Sitton, Shannon M.; Forsbacka, Matthew

    2004-01-01

    For a compact, fast-spectrum reactor, reactivity feedback is dominated by core deformation at elevated temperature. Given the use of accurate deformation measurement techniques, it is possible to simulate nuclear feedback in non-nuclear electrically heated reactor tests. Implementation of simulated reactivity feedback in response to measured deflection is being tested at the NASA Marshall Space Flight Center Early Flight Fission Test Facility (EFF-TF). During tests of the SAFE-100 reactor prototype, core deflection was monitored using a high resolution camera. "virtual" reactivity feedback was accomplished by applying the results of Monte Carlo calculations (MCNPX) to core deflection measurements; the computational analysis was used to establish the reactivity worth of van'ous core deformations. The power delivered to the SAFE-100 prototype was then dusted accordingly via kinetics calculations, The work presented in this paper will demonstrate virtual reactivity feedback as core power was increased from 1 kilowatt(sub t), to 10 kilowatts(sub t), held approximately constant at 10 kilowatts (sub t), and then allowed to decrease based on the negative thermal reactivity coefficient.

  6. Reactor physics experiments related to transmutation in the KUCA

    Energy Technology Data Exchange (ETDEWEB)

    Shiroya, Seiji [Kyoto Univ., Kumatori, Osaka (Japan). Research Reactor Inst.

    1997-11-01

    At the Kyoto University Critical Assembly (KUCA), {sup 237}Np/{sup 235}U fission rate ratios are being measured using the back-to-back type double fission chamber to examine the nuclear data and the computational method for the transmutation of minor actinides (MA) in light water reactors (LWRs). The neutron spectra of cores are systematically being varied by changing the moderator-to-fuel volume ratio (V{sub m}/V{sub f}). The measured data are being compared with the calculated results by SRAC with three different nuclear data files. It has been indicated that the calculated results with JENDL-3.2 agreed better with the measured ones than those with JENDL-3.1 and ENDF/B-VI, although the calculated results underestimated the measured ones by around 10%. (author)

  7. Application of a Loop-Type Laboratory Biofilm Reactor to the Evaluation of Biofilm for Some Metallic Materials and Polymers such as Urinary Stents and Catheters

    Directory of Open Access Journals (Sweden)

    Hideyuki Kanematsu

    2016-10-01

    Full Text Available A laboratory biofilm reactor (LBR was modified to a new loop-type closed system in order to evaluate novel stents and catheter materials using 3D optical microscopy and Raman spectroscopy. Two metallic specimens, pure nickel and cupronickel (80% Cu-20% Ni, along with two polymers, silicone and polyurethane, were chosen as examples to ratify the system. Each set of specimens was assigned to the LBR using either tap water or an NB (Nutrient broth based on peptone from animal foods and beef extract mainly—cultured solution with E-coli formed over 48–72 h. The specimens were then analyzed using Raman Spectroscopy. 3D optical microscopy was employed to corroborate the Raman Spectroscopy results for only the metallic specimens since the inherent roughness of the polymer specimens made such measurements difficult. The findings suggest that the closed loop-type LBR together with Raman spectroscopy analysis is a useful method for evaluating biomaterials as a potential urinary system.

  8. Sabatier Reactor System Integration with Microwave Plasma Methane Pyrolysis Post-Processor for Closed-Loop Hydrogen Recovery

    Science.gov (United States)

    Abney, Morgan B.; Miller, Lee A.; Williams, Tom

    2010-01-01

    The Carbon Dioxide Reduction Assembly (CRA) designed and developed for the International Space Station (ISS) represents the state-of-the-art in carbon dioxide reduction (CDRe) technology. The CRA produces water and methane by reducing carbon dioxide with hydrogen via the Sabatier reaction. The water is recycled to the Oxygen Generation Assembly (OGA) and the methane is vented overboard resulting in a net loss of hydrogen. The proximity to earth and the relative ease of logistics resupply from earth allow for a semi-closed system on ISS. However, long-term manned space flight beyond low earth orbit (LEO) dictates a more thoroughly closed-loop system involving significantly higher recovery of hydrogen, and subsequent recovery of oxygen, to minimize costs associated with logistics resupply beyond LEO. The open-loop ISS system for CDRe can be made closed-loop for follow-on missions by further processing methane to recover hydrogen. For this purpose, a process technology has been developed that employs a microwave-generated plasma to reduce methane to hydrogen and acetylene resulting in 75% theoretical recovery of hydrogen. In 2009, a 1-man equivalent Plasma Pyrolysis Assembly (PPA) was delivered to the National Aeronautics and Space Administration (NASA) for technical evaluation. The PPA has been integrated with a Sabatier Development Unit (SDU). The integrated process configuration incorporates a sorbent bed to eliminate residual carbon dioxide and water vapor in the Sabatier methane product stream before it enters the PPA. This paper provides detailed information on the stand-alone and integrated performance of both the PPA and SDU. Additionally, the integrated test stand design and anticipated future work are discussed.

  9. Experience of on-site disposal of production uranium-graphite nuclear reactor.

    Science.gov (United States)

    Pavliuk, Alexander O; Kotlyarevskiy, Sergey G; Bespala, Evgeny V; Zakharova, Elena V; Ermolaev, Vyacheslav M; Volkova, Anna G

    2018-04-01

    The paper reported the experience gained in the course of decommissioning EI-2 Production Uranium-Graphite Nuclear Reactor. EI-2 was a production Uranium-Graphite Nuclear Reactor located on the Production and Demonstration Center for Uranium-Graphite Reactors JSC (PDC UGR JSC) site of Seversk City, Tomsk Region, Russia. EI-2 commenced its operation in 1958, and was shut down on December 28, 1990, having operated for the period of 33 years all together. The extra pure grade graphite for the moderator, water for the coolant, and uranium metal for the fuel were used in the reactor. During the operation nitrogen gas was passed through the graphite stack of the reactor. In the process of decommissioning the PDC UGR JSC site the cavities in the reactor space were filled with clay-based materials. A specific composite barrier material based on clays and minerals of Siberian Region was developed for the purpose. Numerical modeling demonstrated the developed clay composite would make efficient geological barriers preventing release of radionuclides into the environment. Copyright © 2018 Elsevier Ltd. All rights reserved.

  10. Juicer Provides a One-Click System for Analyzing Loop-Resolution Hi-C Experiments.

    Science.gov (United States)

    Durand, Neva C; Shamim, Muhammad S; Machol, Ido; Rao, Suhas S P; Huntley, Miriam H; Lander, Eric S; Aiden, Erez Lieberman

    2016-07-01

    Hi-C experiments explore the 3D structure of the genome, generating terabases of data to create high-resolution contact maps. Here, we introduce Juicer, an open-source tool for analyzing terabase-scale Hi-C datasets. Juicer allows users without a computational background to transform raw sequence data into normalized contact maps with one click. Juicer produces a hic file containing compressed contact matrices at many resolutions, facilitating visualization and analysis at multiple scales. Structural features, such as loops and domains, are automatically annotated. Juicer is available as open source software at http://aidenlab.org/juicer/. Copyright © 2016 The Authors. Published by Elsevier Inc. All rights reserved.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1982-01-01

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

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

    NARCIS (Netherlands)

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

    1992-01-01

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

  13. Kinetic study of COS with tertiary alkanolamine solutions 1. Experiments in an intensely stirred batch reactor

    NARCIS (Netherlands)

    Littel, R.J.; Littel, R.J.; Versteeg, Geert; van Swaaij, Willibrordus Petrus Maria

    1992-01-01

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

  14. Operation of the NETL Chemical Looping Reactor with Natural Gas and a Novel Copper-Iron Material

    Energy Technology Data Exchange (ETDEWEB)

    Bayham, Sanuel [National Energy Technology Lab. (NETL), Morgantown, WV (United States); Straub, Doug [National Energy Technology Lab. (NETL), Morgantown, WV (United States); Weber, Justin [National Energy Technology Lab. (NETL), Morgantown, WV (United States)

    2017-02-01

    As part of the U.S. Department of Energy’s Advanced Combustion Program, the National Energy Technology Laboratory’s Research and Innovation Center (NETL R&IC) is investigating the feasibility of a novel combustion concept in which the GHG emissions can be significantly reduced. This concept involves burning fuel and air without mixing these two reactants. If this concept is technically feasible, then CO2 emissions can be significantly reduced at a much lower cost than more conventional approaches. This indirect combustion concept has been called Chemical Looping Combustion (CLC) because an intermediate material (i.e., a metaloxide) is continuously cycled to oxidize the fuel. This CLC concept is the focus of this research and will be described in more detail in the following sections.

  15. Resolving octant degeneracy at LBL experiment by combining Daya Bay reactor setup

    Science.gov (United States)

    Bora, Kalpana; Dutta, Debajyoti

    2014-03-01

    Long baseline Experiment(LBL) have promised to be a very powerful experimental setup to study various issues related to Neutrinos. Some ongoing and planned LBL and medium baseline experiments are- T2K, MINOS, NOvA, LBNE, LBNO etc. But, the long baseline experiments are crippled due to presence of some parameter degeneracies, like the Octant -degeneracy. In this work, we first show the presence of Octant degeneracy in LBL experiments and then combine it with Daya Bay Reactor experiment at different values of CP violation phase. We show that the Octant degeneracy in LBNE can be resolved completely with this proposal.

  16. Design Studies for a Multiple Application Thermal Reactor for Irradiation Experiments (MATRIX)

    Energy Technology Data Exchange (ETDEWEB)

    Pope, Michael A.; Gougar, Hans D.; Ryskamp, J. M.

    2015-03-01

    The Advanced Test Reactor (ATR) is a high power density test reactor specializing in fuel and materials irradiation. For more than 45 years, the ATR has provided irradiations of materials and fuels testing along with radioisotope production. Should unforeseen circumstances lead to the decommissioning of ATR, the U.S. Government would be left without a large-scale materials irradiation capability to meet the needs of its nuclear energy and naval reactor missions. In anticipation of this possibility, work was performed under the Laboratory Directed Research and Development (LDRD) program to investigate test reactor concepts that could satisfy the current missions of the ATR along with an expanded set of secondary missions. A survey was conducted in order to catalogue the anticipated needs of potential customers. Then, concepts were evaluated to fill the role for this reactor, dubbed the Multi-Application Thermal Reactor Irradiation eXperiments (MATRIX). The baseline MATRIX design is expected to be capable of longer cycle lengths than ATR given a particular batch scheme. The volume of test space in In-Pile-Tubes (IPTs) is larger in MATRIX than in ATR with comparable magnitude of neutron flux. Furthermore, MATRIX has more locations of greater volume having high fast neutron flux than ATR. From the analyses performed in this work, it appears that the lead MATRIX design can be designed to meet the anticipated needs of the ATR replacement reactor. However, this design is quite immature, and therefore any requirements currently met must be re-evaluated as the design is developed further.

  17. Status of the NGNP Fuel Experiment AGR-2 Irradiated in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Blaine Grover

    2012-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2), which utilized the same experiment design as well as control and monitoring systems as AGR-1, started irradiation in June 2010 and is currently scheduled to be completed in April 2013. The design of this experiment and support systems will be briefly discussed, followed by the progress and status of the experiment to date.

  18. A new MC-based method to evaluate the fission fraction uncertainty at reactor neutrino experiment

    CERN Document Server

    Ma, X B; Chen, Y X

    2016-01-01

    Uncertainties of fission fraction is an important uncertainty source for the antineutrino flux prediction in a reactor antineutrino experiment. A new MC-based method of evaluating the covariance coefficients between isotopes was proposed. It was found that the covariance coefficients will varying with reactor burnup and which may change from positive to negative because of fissioning balance effect, for example, the covariance coefficient between $^{235}$U and $^{239}$Pu changes from 0.15 to -0.13. Using the equation between fission fraction and atomic density, the consistent of uncertainty of fission fraction and the covariance matrix were obtained. The antineutrino flux uncertainty is 0.55\\% which does not vary with reactor burnup, and the new value is about 8.3\\% smaller.

  19. FLOWSHEET EVALUATION FOR THE DISSOLVING AND NEUTRALIZATION OF SODIUM REACTOR EXPERIMENT USED NUCLEAR FUEL

    Energy Technology Data Exchange (ETDEWEB)

    Daniel, W. E.; Hansen, E. K.; Shehee, T. C.

    2012-10-30

    This report includes the literature review, hydrogen off-gas calculations, and hydrogen generation tests to determine that H-Canyon can safely dissolve the Sodium Reactor Experiment (SRE; thorium fuel), Ford Nuclear Reactor (FNR; aluminum alloy fuel), and Denmark Reactor (DR-3; silicide fuel, aluminum alloy fuel, and aluminum oxide fuel) assemblies in the L-Bundles with respect to the hydrogen levels in the projected peak off-gas rates. This is provided that the number of L-Bundles charged to the dissolver is controlled. Examination of SRE dissolution for potential issues has aided in predicting the optimal batching scenario. The calculations detailed in this report demonstrate that the FNR, SRE, and DR-3 used nuclear fuel (UNF) are bounded by MURR UNF and may be charged using the controls outlined for MURR dissolution in a prior report.

  20. Supercritical Water Gasification of Biomass in a Ceramic Reactor: Long-Time Batch Experiments

    Directory of Open Access Journals (Sweden)

    Daniele Castello

    2017-10-01

    Full Text Available Supercritical water gasification (SCWG is an emerging technology for the valorization of (wet biomass into a valuable fuel gas composed of hydrogen and/or methane. The harsh temperature and pressure conditions involved in SCWG (T > 375 °C, p > 22 MPa are definitely a challenge for the manufacturing of the reactors. Metal surfaces are indeed subject to corrosion under hydrothermal conditions, and expensive special alloys are needed to overcome such drawbacks. A ceramic reactor could be a potential solution to this issue. Finding a suitable material is, however, complex because the catalytic effect of the material can influence the gas yield and composition. In this work, a research reactor featuring an internal alumina inlay was utilized to conduct long-time (16 h batch tests with real biomasses and model compounds. The same experiments were also conducted in batch reactors made of stainless steel and Inconel 625. The results show that the three devices have similar performance patterns in terms of gas production, although in the ceramic reactor higher yields of C2+ hydrocarbons were obtained. The SEM observation of the reacted alumina surface revealed a good resistance of such material to supercritical conditions, even though some intergranular corrosion was observed.

  1. Long-lived activation products in TRIGA Mark II research reactor concrete shield: calculation and experiment

    Science.gov (United States)

    Žagar, Tomaž; Božič, Matjaž; Ravnik, Matjaž

    2004-12-01

    In this paper, a process of long-lived activity determination in research reactor concrete shielding is presented. The described process is a combination of experiment and calculations. Samples of original heavy reactor concrete containing mineral barite were irradiated inside the reactor shielding to measure its long-lived induced radioactivity. The most active long-lived (γ emitting) radioactive nuclides in the concrete were found to be 133Ba, 60Co and 152Eu. Neutron flux, activation rates and concrete activity were calculated for actual shield geometry for different irradiation and cooling times using TORT and ORIGEN codes. Experimental results of flux and activity measurements showed good agreement with the results of calculations. Volume of activated concrete waste after reactor decommissioning was estimated for particular case of Jožef Stefan Institute TRIGA reactor. It was observed that the clearance levels of some important long-lived isotopes typical for barite concrete (e.g. 133Ba, 41Ca) are not included in the IAEA and EU basic safety standards.

  2. Background study for the KamLAND reactor neutrino experiment

    Energy Technology Data Exchange (ETDEWEB)

    Ichimura, K; Minekawa, Y [RCNS, Tohoku University, Sendai 980-8578 (Japan)], E-mail: ichimura@awa.tohoku.ac.jp, E-mail: yukie@awa.tohoku.ac.jp

    2008-07-15

    One of the goals of the KamLAND experiment is a search for anti-neutrino oscillation via inverse {beta} decay with the characteristic delayed-coincidence method in the liquid scintillator. For a more precise measurement than previous KamLAND result [1], we have improved the background estimations of ({alpha}, n) and fast neutrons. We present the estimated number of backgrounds in our data set from Mar. 2002 to May 2007.

  3. Operation Experiences of the Small Scale Nitrogen Loop with a Water-Cooling Printed Circuit Heat Exchanger

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chan Soo; Seo, Dong Un; Yoo, Tae Ho; Hong, Sung Deok; Kim, Yong Wan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    A small scale nitrogen loop is in-operation for the integrity and feasibility test of the hybrid-concept sulfur trioxide decomposer. The small-scale gas loop takes the place of the hot gas and the process gas loop. The hot gas loop simulates the intermediate loop of a nuclear hydrogen production system in that it is designed to withstand the maximum temperature of 1273K, the maximum pressure of 6.0 MPa, and to operate at a mass flow rate of 2.0 kg/min with 4.0 MPa. Nitrogen is used as the working fluid for simple high pressure gas experiments. The fluid temperature is controlled by adjusting the power of the heaters using direct voltage controllers. The accumulator maintains the primary system at the constant pressure. The inverter of the circulator and the bypass flow valve control the primary mass flow rate. In this paper, the operating experience is presented to estimate the performance of the primary system. A water-cooling printed circuit heat exchanger was used to cool the hot gas into the room temperature

  4. A Moral Experience Feedback Loop: Modeling a System of Moral Self-Cultivation in Everyday Life

    Science.gov (United States)

    Sherblom, Stephen A.

    2015-01-01

    This "systems thinking" model illustrates a common feedback loop by which people engage the moral world and continually reshape their moral sensibility. The model highlights seven processes that collectively form this feedback loop: beginning with (1) one's current moral sensibility which shapes processes of (2) perception, (3)…

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

    Energy Technology Data Exchange (ETDEWEB)

    Wood, RT

    2004-09-27

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

  6. The iso-response method: Measuring neuronal stimulus integration with closed-loop experiments

    Directory of Open Access Journals (Sweden)

    Tim eGollisch

    2012-12-01

    Full Text Available Throughout the nervous system, neurons integrate high-dimensional input streams and transform them into an output of their own. This integration of incoming signals involves filtering processes and complex nonlinear operations. The shapes of these filters and nonlinearities determine the computational features of single neurons and their functional roles within larger networks. A detailed characterization of signal integration is thus a central ingredient to understanding information processing in neural circuits. Conventional methods for measuring single-neuron response properties, such as reverse correlation, however, are often limited by the implicit assumption that stimulus integration occurs in a linear fashion. Here, we review a conceptual and experimental alternative that is based on exploring the space of those sensory stimuli that result in the same neural output. As demonstrated by recent results in the auditory and visual system, such iso-response stimuli can be used to identify the nonlinearities relevant for stimulus integration, disentangle subsequent neural processing steps, and determine their characteristics with unprecedented precision. Automated closed-loop experiments are crucial for this advance, allowing rapid search strategies for identifying iso-response stimuli during experiments. Prime targets for the method are feed-forward neural signaling chains in sensory systems, but the method has also been successfully applied to feedback systems. Depending on the specific question, iso-response may refer to a predefined firing rate, single-spike probability, first-spike latency, or other output measures. Examples from different studies show that substantial progress in understanding neural dynamics and coding can be achieved once rapid online data analysis and stimulus generation, adaptive sampling, and computational modelling are tightly integrated into experiments.

  7. The iso-response method: measuring neuronal stimulus integration with closed-loop experiments

    Science.gov (United States)

    Gollisch, Tim; Herz, Andreas V. M.

    2012-01-01

    Throughout the nervous system, neurons integrate high-dimensional input streams and transform them into an output of their own. This integration of incoming signals involves filtering processes and complex non-linear operations. The shapes of these filters and non-linearities determine the computational features of single neurons and their functional roles within larger networks. A detailed characterization of signal integration is thus a central ingredient to understanding information processing in neural circuits. Conventional methods for measuring single-neuron response properties, such as reverse correlation, however, are often limited by the implicit assumption that stimulus integration occurs in a linear fashion. Here, we review a conceptual and experimental alternative that is based on exploring the space of those sensory stimuli that result in the same neural output. As demonstrated by recent results in the auditory and visual system, such iso-response stimuli can be used to identify the non-linearities relevant for stimulus integration, disentangle consecutive neural processing steps, and determine their characteristics with unprecedented precision. Automated closed-loop experiments are crucial for this advance, allowing rapid search strategies for identifying iso-response stimuli during experiments. Prime targets for the method are feed-forward neural signaling chains in sensory systems, but the method has also been successfully applied to feedback systems. Depending on the specific question, “iso-response” may refer to a predefined firing rate, single-spike probability, first-spike latency, or other output measures. Examples from different studies show that substantial progress in understanding neural dynamics and coding can be achieved once rapid online data analysis and stimulus generation, adaptive sampling, and computational modeling are tightly integrated into experiments. PMID:23267315

  8. Summary of Thermocouple Performance During Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor and Out-of-Pile Thermocouple Testing in Support of Such Experiments

    Energy Technology Data Exchange (ETDEWEB)

    A. J. Palmer; DC Haggard; J. W. Herter; M. Scervini; W. D. Swank; D. L. Knudson; R. S. Cherry

    2011-07-01

    High temperature gas reactor experiments create unique challenges for thermocouple based temperature measurements. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time dependent change in composition and, as a consequence, a time dependent drift of the thermocouple signal. This drift is particularly severe for high temperature platinum-rhodium thermocouples (Types S, R, and B); and tungsten-rhenium thermocouples (Types C and W). For lower temperature applications, previous experiences with type K thermocouples in nuclear reactors have shown that they are affected by neutron irradiation only to a limited extent. Similarly type N thermocouples are expected to be only slightly affected by neutron fluxes. Currently the use of these Nickel based thermocouples is limited when the temperature exceeds 1000°C due to drift related to phenomena other than nuclear irradiation. High rates of open-circuit failure are also typical. Over the past ten years, three long-term Advanced Gas Reactor (AGR) experiments have been conducted with measured temperatures ranging from 700oC – 1200oC. A variety of standard Type N and specialty thermocouple designs have been used in these experiments with mixed results. A brief summary of thermocouple performance in these experiments is provided. Most recently, out of pile testing has been conducted on a variety of Type N thermocouple designs at the following (nominal) temperatures and durations: 1150oC and 1200oC for 2000 hours at each temperature, followed by 200 hours at 1250oC, and 200 hours at 1300oC. The standard Type N design utilizes high purity crushed MgO insulation and an Inconel 600 sheath. Several variations on the standard Type N design were tested, including Haynes 214 alloy sheath, spinel (MgAl2O4) insulation instead of MgO, a customized sheath developed at the University of Cambridge, and finally a loose assembly thermocouple with hard fired alumina

  9. Summary of thermocouple performance during advanced gas reactor fuel irradiation experiments in the advanced test reactor and out-of-pile thermocouple testing in support of such experiments

    Energy Technology Data Exchange (ETDEWEB)

    Palmer, A. J.; Haggard, DC; Herter, J. W.; Swank, W. D.; Knudson, D. L.; Cherry, R. S. [Idaho National Laboratory, P.O. Box 1625, MS 4112, Idaho Falls, ID, (United States); Scervini, M. [University of Cambridge, Department of Material Science and Metallurgy, 27 Charles Babbage Road, CB3 0FS, Cambridge, (United Kingdom)

    2015-07-01

    High temperature gas reactor experiments create unique challenges for thermocouple-based temperature measurements. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time-dependent change in composition and, as a consequence, a time-dependent drift of the thermocouple signal. This drift is particularly severe for high temperature platinum-rhodium thermocouples (Types S, R, and B) and tungsten-rhenium thermocouples (Type C). For lower temperature applications, previous experiences with Type K thermocouples in nuclear reactors have shown that they are affected by neutron irradiation only to a limited extent. Similarly, Type N thermocouples are expected to be only slightly affected by neutron fluence. Currently, the use of these nickel-based thermocouples is limited when the temperature exceeds 1000 deg. C due to drift related to phenomena other than nuclear irradiation. High rates of open-circuit failure are also typical. Over the past 10 years, three long-term Advanced Gas Reactor experiments have been conducted with measured temperatures ranging from 700 deg. C - 1200 deg. C. A variety of standard Type N and specialty thermocouple designs have been used in these experiments with mixed results. A brief summary of thermocouple performance in these experiments is provided. Most recently, out-of-pile testing has been conducted on a variety of Type N thermocouple designs at the following (nominal) temperatures and durations: 1150 deg. C and 1200 deg. C for 2,000 hours at each temperature, followed by 200 hours at 1250 deg. C and 200 hours at 1300 deg. C. The standard Type N design utilizes high purity, crushed MgO insulation and an Inconel 600 sheath. Several variations on the standard Type N design were tested, including a Haynes 214 alloy sheath, spinel (MgAl{sub 2}O{sub 4}) insulation instead of MgO, a customized sheath developed at the University of Cambridge, and finally a loose assembly

  10. Phenylnaphthalene Derivatives as Heat Transfer Fluids for Concentrating Solar Power: Loop Experiments and Final Report

    Energy Technology Data Exchange (ETDEWEB)

    McFarlane, Joanna [ORNL; Bell, Jason R [ORNL; Felde, David K [ORNL; Joseph III, Robert Anthony [ORNL; Qualls, A L [ORNL; Weaver, Samuel P [ORNL

    2013-02-01

    ORNL and subcontractor Cool Energy completed an investigation of higher-temperature, organic thermal fluids for solar thermal applications. Although static thermal tests showed promising results for 1-phenylnaphthalene, loop testing at temperatures to 450 C showed that the material isomerized at a slow rate. In a loop with a temperature high enough to drive the isomerization, the higher melting point byproducts tended to condense onto cooler surfaces. So, as experienced in loop operation, eventually the internal channels of cooler components such as the waste heat rejection exchanger may become coated or clogged and loop performance will decrease. Thus, pure 1-phenylnaphthalene does not appear to be a fluid that would have a sufficiently long lifetime (years to decades) to be used in a loop at the increased temperatures of interest. Hence a decision was made not to test the ORNL fluid in the loop at Cool Energy Inc. Instead, Cool Energy tested and modeled power conversion from a moderate-temperature solar loop using coupled Stirling engines. Cool Energy analyzed data collected on third and fourth generation SolarHeart Stirling engines operating on a rooftop solar field with a lower temperature (Marlotherm) heat transfer fluid. The operating efficiencies of the Stirling engines were determined at multiple, typical solar conditions, based on data from actual cycle operation. Results highlighted the advantages of inherent thermal energy storage in the power conversion system.

  11. Survey of Worldwide Light Water Reactor Experience with Mixed Uranium-Plutonium Oxide Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Cowell, B.S.; Fisher, S.E.

    1999-02-01

    The US and the Former Soviet Union (FSU) have recently declared quantities of weapons materials, including weapons-grade (WG) plutonium, excess to strategic requirements. One of the leading candidates for the disposition of excess WG plutonium is irradiation in light water reactors (LWRs) as mixed uranium-plutonium oxide (MOX) fuel. A description of the MOX fuel fabrication techniques in worldwide use is presented. A comprehensive examination of the domestic MOX experience in US reactors obtained during the 1960s, 1970s, and early 1980s is also presented. This experience is described by manufacturer and is also categorized by the reactor facility that irradiated the MOX fuel. A limited summary of the international experience with MOX fuels is also presented. A review of MOX fuel and its performance is conducted in view of the special considerations associated with the disposition of WG plutonium. Based on the available information, it appears that adoption of foreign commercial MOX technology from one of the successful MOX fuel vendors will minimize the technical risks to the overall mission. The conclusion is made that the existing MOX fuel experience base suggests that disposition of excess weapons plutonium through irradiation in LWRs is a technically attractive option.

  12. 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

  13. Feasibility of conducting a dynamic helium charging experiment for vanadium alloys in the advanced test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H.; Gomes, I.; Strain, R.V.; Smith, D.L. [Argonne National Lab., IL (United States); Matsui, H. [Tohoku Univ. (Japan)

    1996-10-01

    The feasibility of conducting a dynamic helium charging experiment (DHCE) for vanadium alloys in the water-cooled Advanced Test Reactor (ATR) is being investigated as part of the U.S./Monbusho collaboration. Preliminary findings suggest that such an experiment is feasible, with certain constraints. Creating a suitable irradiation position in the ATR, designing an effective thermal neutron filter, incorporating thermocouples for limited specimen temperature monitoring, and handling of tritium during various phases of the assembly and reactor operation all appear to be feasible. An issue that would require special attention, however, is tritium permeation loss through the capsule wall at the higher design temperatures (>{approx}600{degrees}C). If permeation is excessive, the reduced amount of tritium entering the test specimens would limit the helium generation rates in them. At the lower design temperatures (<{approx}425{degrees}C), sodium, instead of lithium, may have to be used as the bond material to overcome the tritium solubility limitation.

  14. 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.

  15. Probing light sterile neutrino signatures at reactor and Spallation Neutron Source neutrino experiments

    Science.gov (United States)

    Kosmas, T. S.; Papoulias, D. K.; Tórtola, M.; Valle, J. W. F.

    2017-09-01

    We investigate the impact of a fourth sterile neutrino at reactor and Spallation Neutron Source neutrino detectors. Specifically, we explore the discovery potential of the TEXONO and COHERENT experiments to subleading sterile neutrino effects through the measurement of the coherent elastic neutrino-nucleus scattering event rate. Our dedicated χ2-sensitivity analysis employs realistic nuclear structure calculations adequate for high purity sub-keV threshold Germanium detectors.

  16. Study of in-reactor creep of vanadium alloy in the HFIR RB-12J experiment

    Energy Technology Data Exchange (ETDEWEB)

    Strain, R.V.; Konicek, C.F.; Tsai, H. [Argonne National Lab., IL (United States)

    1996-10-01

    Biaxial creep specimens will be included in the HFIR RB-12J experiment to study in-reactor creep of the V-4Cr-4Ti alloy at {approx}500{degrees}C and 5 dpa. These specimens were fabricated with the 500-kg, heat (832665) material and pressurized to attain 0, 50, 100, 150, and 200 MPa mid-wall hoop stresses during the irradiation.

  17. Indication of reactor ν(e) disappearance in the Double Chooz experiment.

    Science.gov (United States)

    Abe, Y; Aberle, C; Akiri, T; dos Anjos, J C; Ardellier, F; Barbosa, A F; Baxter, A; Bergevin, M; Bernstein, A; Bezerra, T J C; Bezrukhov, L; Blucher, E; Bongrand, M; Bowden, N S; Buck, C; Busenitz, J; Cabrera, A; Caden, E; Camilleri, L; Carr, R; Cerrada, M; Chang, P-J; Chimenti, P; Classen, T; Collin, A P; Conover, E; Conrad, J M; Cormon, S; Crespo-Anadón, J I; Cribier, M; Crum, K; Cucoanes, A; D'Agostino, M V; Damon, E; Dawson, J V; Dazeley, S; Dierckxsens, M; Dietrich, D; Djurcic, Z; Dracos, M; Durand, V; Efremenko, Y; Elnimr, M; Endo, Y; Etenko, A; Falk, E; Fallot, M; Fechner, M; von Feilitzsch, F; Felde, J; Fernandes, S M; Franco, D; Franke, A J; Franke, M; Furuta, H; Gama, R; Gil-Botella, I; Giot, L; Göger-Neff, M; Gonzalez, L F G; Goodman, M C; Goon, J T M; Greiner, D; Guillon, B; Haag, N; Hagner, C; Hara, T; Hartmann, F X; Hartnell, J; Haruna, T; Haser, J; Hatzikoutelis, A; Hayakawa, T; Hofmann, M; Horton-Smith, G A; Ishitsuka, M; Jochum, J; Jollet, C; Jones, C L; Kaether, F; Kalousis, L; Kamyshkov, Y; Kaplan, D M; Kawasaki, T; Keefer, G; Kemp, E; de Kerret, H; Kibe, Y; Konno, T; Kryn, D; Kuze, M; Lachenmaier, T; Lane, C E; Langbrandtner, C; Lasserre, T; Letourneau, A; Lhuillier, D; Lima, H P; Lindner, M; Liu, Y; López-Castanõ, J M; LoSecco, J M; Lubsandorzhiev, B K; Lucht, S; McKee, D; Maeda, J; Maesano, C N; Mariani, C; Maricic, J; Martino, J; Matsubara, T; Mention, G; Meregaglia, A; Miletic, T; Milincic, R; Milzstajn, A; Miyata, H; Motta, D; Mueller, Th A; Nagasaka, Y; Nakajima, K; Novella, P; Obolensky, M; Oberauer, L; Onillon, A; Osborn, A; Ostrovskiy, I; Palomares, C; Peeters, S J M; Pepe, I M; Perasso, S; Perrin, P; Pfahler, P; Porta, A; Potzel, W; Queval, R; Reichenbacher, J; Reinhold, B; Remoto, A; Reyna, D; Röhling, M; Roth, S; Rubin, H A; Sakamoto, Y; Santorelli, R; Sato, F; Schönert, S; Schoppmann, S; Schwan, U; Schwetz, T; Shaevitz, M H; Shrestha, D; Sida, J-L; Sinev, V; Skorokhvatov, M; Smith, E; Spitz, J; Stahl, A; Stancu, I; Strait, M; Stüken, A; Suekane, F; Sukhotin, S; Sumiyoshi, T; Sun, Y; Sun, Z; Svoboda, R; Tabata, H; Tamura, N; Terao, K; Tonazzo, A; Toups, M; Trinh Thi, H H; Veyssiere, C; Wagner, S; Watanabe, H; White, B; Wiebusch, C; Winslow, L; Worcester, M; Wurm, M; Yanovitch, E; Yermia, F; Zbiri, K; Zimmer, V

    2012-03-30

    The Double Chooz experiment presents an indication of reactor electron antineutrino disappearance consistent with neutrino oscillations. An observed-to-predicted ratio of events of 0.944±0.016(stat)±0.040(syst) was obtained in 101 days of running at the Chooz nuclear power plant in France, with two 4.25 GW(th) reactors. The results were obtained from a single 10 m(3) fiducial volume detector located 1050 m from the two reactor cores. The reactor antineutrino flux prediction used the Bugey4 flux measurement after correction for differences in core composition. The deficit can be interpreted as an indication of a nonzero value of the still unmeasured neutrino mixing parameter sin(2)2θ(13). Analyzing both the rate of the prompt positrons and their energy spectrum, we find sin(2)2θ(13)=0.086±0.041(stat)±0.030(syst), or, at 90% C.L., 0.017

  18. Design of a rotary reactor for chemical-looping combustion. Part 2: Comparison of copper-, nickel-, and iron-based oxygen carriers

    KAUST Repository

    Zhao, Zhenlong

    2014-04-01

    Chemical-looping combustion (CLC) is a novel and promising option for several applications including carbon capture (CC), fuel reforming, H 2 generation, etc. Previous studies demonstrated the feasibility of performing CLC in a novel rotary design with micro-channel structures. Part 1 of this series studied the fundamentals of the reactor design and proposed a comprehensive design procedure, enabling a systematic methodology of designing and evaluating the rotary CLC reactor with different OCs and operating conditions. This paper presents the application of the methodology to the designs with three commonly used OCs, i.e., copper, nickel, and iron. The physical properties and the reactivities of the three OCs are compared at operating conditions suitable for the rotary CLC. Nickel has the highest reduction rate, but relatively slow oxidation reactivity while the iron reduction rate is most sensitive to the fuel concentration. The design parameters and the operating conditions for the three OCs are selected, following the strategies proposed in Part 1, and the performances are evaluated using a one-dimensional plug-flow model developed previously. The simulations show that for all OCs, complete fuel conversion and high carbon separation efficiency can be achieved at periodic stationary state with reasonable operational stabilities. The nickel-based design includes the smallest dimensions because of its fast reduction rate. The operation of nickel case is mainly limited to the slow oxidation rate, and hence a relatively large share of air sector is used. The iron-based design has the largest size, due to its slow reduction reactivity near the exit or in the fuel purge sector where the fuel concentration is low. The gas flow temperature increases monotonically for all the cases, and is mainly determined by the solid temperature. In the periodic state, the local temperature variation is within 40 K and the thermal distortion is limited. The design of the rotary CLC is

  19. GROWTH OF THE INTERNATIONAL CRITICALITY SAFETY AND REACTOR PHYSICS EXPERIMENT EVALUATION PROJECTS

    Energy Technology Data Exchange (ETDEWEB)

    J. Blair Briggs; John D. Bess; Jim Gulliford

    2011-09-01

    Since the International Conference on Nuclear Criticality Safety (ICNC) 2007, the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPhEP) have continued to expand their efforts and broaden their scope. Eighteen countries participated on the ICSBEP in 2007. Now, there are 20, with recent contributions from Sweden and Argentina. The IRPhEP has also expanded from eight contributing countries in 2007 to 16 in 2011. Since ICNC 2007, the contents of the 'International Handbook of Evaluated Criticality Safety Benchmark Experiments1' have increased from 442 evaluations (38000 pages), containing benchmark specifications for 3955 critical or subcritical configurations to 516 evaluations (nearly 55000 pages), containing benchmark specifications for 4405 critical or subcritical configurations in the 2010 Edition of the ICSBEP Handbook. The contents of the Handbook have also increased from 21 to 24 criticality-alarm-placement/shielding configurations with multiple dose points for each, and from 20 to 200 configurations categorized as fundamental physics measurements relevant to criticality safety applications. Approximately 25 new evaluations and 150 additional configurations are expected to be added to the 2011 edition of the Handbook. Since ICNC 2007, the contents of the 'International Handbook of Evaluated Reactor Physics Benchmark Experiments2' have increased from 16 different experimental series that were performed at 12 different reactor facilities to 53 experimental series that were performed at 30 different reactor facilities in the 2011 edition of the Handbook. Considerable effort has also been made to improve the functionality of the searchable database, DICE (Database for the International Criticality Benchmark Evaluation Project) and verify the accuracy of the data contained therein. DICE will be discussed in separate papers at ICNC 2011. The status of the

  20. Optimization of Fast Critical Experiments to Reduce Nuclear Data Uncertainties in Support of a Fast Burner Reactor Design Concept

    Science.gov (United States)

    Stover, Tracy E., Jr.

    An optimization technique has been developed to select optimized experimental design specifications to produce data specifically designed to be assimilated to optimize a given reactor concept. Data from the optimized experiment is assimilated to generate posteriori uncertainties on the reactor concept's core attributes from which the design responses are computed. The reactor concept is then optimized with the new data to realize cost savings by reducing margin. The optimization problem iterates until an optimal experiment is found to maximize the savings. A new generation of innovative nuclear reactor designs, in particular fast neutron spectrum recycle reactors, are being considered for the application of closing the nuclear fuel cycle in the future. Safe and economical design of these reactors will require uncertainty reduction in basic nuclear data which are input to the reactor design. These data uncertainty propagate to design responses which in turn require the reactor designer to incorporate additional safety margin into the design, which often increases the cost of the reactor. Therefore basic nuclear data needs to be improved and this is accomplished through experimentation. Considering the high cost of nuclear experiments, it is desired to have an optimized experiment which will provide the data needed for uncertainty reduction such that a reactor design concept can meet its target accuracies or to allow savings to be realized by reducing the margin required due to uncertainty propagated from basic nuclear data. However, this optimization is coupled to the reactor design itself because with improved data the reactor concept can be re-optimized itself. It is thus desired to find the experiment that gives the best optimized reactor design. Methods are first established to model both the reactor concept and the experiment and to efficiently propagate the basic nuclear data uncertainty through these models to outputs. The representativity of the experiment

  1. Hydrogen/Oxygen Reactions at High Pressures and Intermediate Temperatures: Flow Reactor Experiments and Kinetic Modeling

    DEFF Research Database (Denmark)

    Hashemi, Hamid; Christensen, Jakob Munkholt; Glarborg, Peter

    of the mixture was varied from oxidizing to reducing conditions. Moreover, a series of experiments in an oxygen atmosphere instead of a nitrogen atmosphere has been done. A reaction mechanism based on a recent work by Burke et al. has been developed. In addition to modeling of the present experiments......, ignition occurs at the temperature of 775–800 K. In general, the present model provides a good agreement with the measurements in the flow reactor and with recent data on laminar burning velocity and ignition delay time....

  2. On-site underground background measurements for the KASKA reactor-neutrino experiment

    Science.gov (United States)

    Furuta, H.; Sakuma, K.; Aoki, M.; Fukuda, Y.; Funaki, Y.; Hara, T.; Haruna, T.; Ishihara, N.; Katsumata, M.; Kawasaki, T.; Kuze, M.; Maeda, J.; Matsubara, T.; Matsumoto, T.; Miyata, H.; Nagasaka, Y.; Nakagawa, T.; Nakajima, N.; Nitta, K.; Sakai, K.; Sakamoto, Y.; Suekane, F.; Sumiyoshi, T.; Tabata, H.; Tamura, N.; Tsuchiya, Y.

    2006-12-01

    On-site underground background measurements were performed for the planned reactor-neutrino oscillation experiment KASKA at Kashiwazaki-Kariwa nuclear power station in Niigata, Japan. A small-diameter boring hole was excavated down to 70 m underground level, and a detector unit for γ-ray and cosmic-muon measurements was placed at various depths to take data. The data were analyzed to obtain abundance of natural radioactive elements in the surrounding soil and rates of cosmic muons that penetrate the overburden. The results will be reflected in the design of the KASKA experiment.

  3. Mass hierarchy sensitivity of medium baseline reactor neutrino experiments with multiple detectors

    Directory of Open Access Journals (Sweden)

    Hong-Xin Wang

    2017-05-01

    Full Text Available We report the neutrino mass hierarchy (MH determination of medium baseline reactor neutrino experiments with multiple detectors, where the sensitivity of measuring the MH can be significantly improved by adding a near detector. Then the impact of the baseline and target mass of the near detector on the combined MH sensitivity has been studied thoroughly. The optimal selections of the baseline and target mass of the near detector are ∼12.5 km and ∼4 kton respectively for a far detector with the target mass of 20 kton and the baseline of 52.5 km. As typical examples of future medium baseline reactor neutrino experiments, the optimal location and target mass of the near detector are selected for the specific configurations of JUNO and RENO-50. Finally, we discuss distinct effects of the reactor antineutrino energy spectrum uncertainty for setups of a single detector and double detectors, which indicate that the spectrum uncertainty can be well constrained in the presence of the near detector.

  4. Overview of the 2014 Edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook)

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess; J. Blair Briggs; Jim Gulliford; Ian Hill

    2014-10-01

    The International Reactor Physics Experiment Evaluation Project (IRPhEP) is a widely recognized world class program. The work of the IRPhEP is documented in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook). Integral data from the IRPhEP Handbook is used by reactor safety and design, nuclear data, criticality safety, and analytical methods development specialists, worldwide, to perform necessary validations of their calculational techniques. The IRPhEP Handbook is among the most frequently quoted reference in the nuclear industry and is expected to be a valuable resource for future decades.

  5. Consolidated fuel reprocessing program: Criticality experiments with fast test reactor fuel pins in an organic moderator

    Energy Technology Data Exchange (ETDEWEB)

    Bierman, S.R.

    1986-12-01

    The results obtained in a series of criticality experiments performed as part of a joint program on criticality data development between the United States Department of Energy and the Power Reactor and Nuclear Fuel Development Corporation of Japan are presented in this report along with a complete description of the experiments. The experiments involved lattices of Fast Test Reactor (FTR) fuel pins in an organic moderator mixture similar to that used in the solvent extraction stage of fuel reprocessing. The experiments are designed to provide data for direct comparison with previously performed experimental measurements with water moderated lattices of FTR fuel pins. The same lattice arrangements and FTR fuel pin types are used in these organic moderated experimental assemblies as were used in the water moderated experiments. The organic moderator is a mixture of 38 wt % tributylphosphate in a normal paraffin hydrocarbon mixture of C{sub 11}H{sub 24} to C{sub 15}H{sub 32} molecules. Critical sizes of 1054.8, 599.2, 301.8, 199.5 and 165.3 fuel pins were obtained respectively for organic moderated lattices having 0.761 cm, 0.968 cm, 1.242 cm, 1.537 cm and 1.935 cm square lattice pitches as compared to 1046.9, 571.9, 293.9, 199.7 and 165.1 fuel pins for the same lattices water moderated.

  6. Exploring the Use of Design of Experiments in Industrial Processes Operating Under Closed-Loop Control

    DEFF Research Database (Denmark)

    Capaci, Francesca; Kulahci, Murat; Vanhatalo, Erik

    2017-01-01

    Industrial manufacturing processes often operate under closed-loop control, where automation aims to keep important process variables at their set-points. In process industries such as pulp, paper, chemical and steel plants, it is often hard to find production processes operating in open loop...... such processes. The Tennessee Eastman challenge process simulator is used as a test-bed to highlight two experimental scenarios. The first scenario explores the impact of experimental factors that may be considered as disturbances in the closed-loop system. The second scenario exemplifies a screening design...... using the set-points of controllers as experimental factors. We provide examples of how to analyze the two scenarios. ...

  7. Digital Phase Locked Loop Induction Motor Speed Controller: Design and Experiments

    Directory of Open Access Journals (Sweden)

    Mouna BEN HAMED

    2012-08-01

    Full Text Available Phase locked loop (PLL is a technique which has contributed significantly toward the technology advancement in communication and motor servo control systems. Inventions in PLL schemes combining with novel integrated circuit have made PLL devices important system components. The development of better modular PLL integrated circuit is continuing. As a result, it is expected that it will contribute to improvement in performance and reliability for communication and servo control systems. In this paper, the study of the speed control of induction motor (IM drives using digital phase locked loop (DPLL is discussed. A novel scalar law which compensates the slip frequency loop calculation is proposed. The overall investigated system is tested using a 1Kw IM. Different speed trajectories are considered covering the realistic operating range. The PLL IM drives controller is implemented all around the most popular integrated circuits 4046 PLL. Experimental results are presented to show the performance of the investigated control system.

  8. Hard real-time closed-loop electrophysiology with the Real-Time eXperiment Interface (RTXI.

    Directory of Open Access Journals (Sweden)

    Yogi A Patel

    2017-05-01

    Full Text Available The ability to experimentally perturb biological systems has traditionally been limited to static pre-programmed or operator-controlled protocols. In contrast, real-time control allows dynamic probing of biological systems with perturbations that are computed on-the-fly during experimentation. Real-time control applications for biological research are available; however, these systems are costly and often restrict the flexibility and customization of experimental protocols. The Real-Time eXperiment Interface (RTXI is an open source software platform for achieving hard real-time data acquisition and closed-loop control in biological experiments while retaining the flexibility needed for experimental settings. RTXI has enabled users to implement complex custom closed-loop protocols in single cell, cell network, animal, and human electrophysiology studies. RTXI is also used as a free and open source, customizable electrophysiology platform in open-loop studies requiring online data acquisition, processing, and visualization. RTXI is easy to install, can be used with an extensive range of external experimentation and data acquisition hardware, and includes standard modules for implementing common electrophysiology protocols.

  9. First Test of Lorentz Violation with a Reactor-based Antineutrino Experiment

    CERN Document Server

    Abe, Y; Anjos, J C dos; Bergevin, M; Bernstein, A; Bezerra, T J C; Bezrukhov, L; Blucher, E; Bowden, N S; Buck, C; Busenitz, J; Cabrera, A; Caden, E; Camilleri, L; Carr, R; Cerrada, M; Chang, P -J; Chimenti, P; Classen, T; Collin, A P; Conover, E; Conrad, J M; Crespo-Anadón, J I; Crum, K; Cucoanes, A; D'Agostino, M V; Damon, E; Dawson, J V; Dazeley, S; Dietrich, D; Djurcic, Z; Dracos, M; Durand, V; Ebert, J; Efremenko, Y; Elnimr, M; Erickson, A; Fallot, M; Fechner, M; von Feilitzsch, F; Felde, J; Fischer, V; Franco, D; Franke, A J; Franke, M; Furuta, H; Gama, R; Gil-Botella, I; Giot, L; Göger-Neff, M; Gonzalez, L F G; Goodman, M C; Goon, J TM; Greiner, D; Haag, N; Habib, S; Hagner, C; Hara, T; Hartmann, F X; Haser, J; Hatzikoutelis, A; Hayakawa, T; Hofmann, M; Horton-Smith, G A; Ishitsuka, M; Jochum, J; Jollet, C; Jones, C L; Kaether, F; Kalousis, L N; Kamyshkov, Y; Kaplan, D M; Katori, T; Kawasaki, T; Keefer, G; Kemp, E; de Kerret, H; Konno, T; Kryn, D; Kuze, M; Lachenmaier, T; Lane, C E; Lasserre, T; Letourneau, A; Lhuillier, D; Lima, H P; Lindner, M; López-Castanõ, J M; LoSecco, J M; Lubsandorzhiev, B K; Lucht, S; McKee, D; Maeda, J; Maesano, C N; Mariani, C; Maricic, J; Martino, J; Matsubara, T; Mention, G; Meregaglia, A; Meyer, M; Miletic, T; Milincic, R; Miyata, H; Mueller, Th A; Nagasaka, Y; Nakajima, K; Novella, P; Obolensky, M; Oberauer, L; Onillon, A; Osborn, A; Ostrovskiy, I; Palomares, C; Pepe, I M; Perasso, S; Perrin, P; Pfahler, P; Porta, A; Potzel, W; Pronost, G; Reichenbacher, J; Reinhold, B; Remoto, A; Röhling, M; Roncin, R; Roth, S; Rybolt, B; Sakamoto, Y; Santorelli, R; Sato, F; Schönert, S; Schoppmann, S; Schwetz, T; Shaevitz, M H; Shrestha, D; Sida, J -L; Sinev, V; Skorokhvatov, M; Smith, E; Spitz, J; Stahl, A; Stancu, I; Stokes, L F F; Strait, M; Stüken, A; Suekane, F; Sukhotin, S; Sumiyoshi, T; Sun, Y; Terao, K; Tonazzo, A; Toups, M; Thi, H H Trinh; Valdiviesso, G; Veyssiere, C; Wagner, S; Watanabe, H; White, B; Wiebusch, C; Winslow, L; Worcester, M; Wurm, M; Yanovitch, E; Yermia, F; Zimmer, V

    2012-01-01

    We present a search for Lorentz violation with 8249 candidate electron antineutrino events taken by the Double Chooz experiment in 227.9 live days of running. This analysis, featuring a search for a sidereal time dependence of the events, is the first test of Lorentz invariance using a reactor-based antineutrino source. No sidereal variation is present in the data and the disappearance results are consistent with sidereal time independent oscillations. Under the Standard-Model Extension (SME), we set the first limits on fourteen Lorentz violating coefficients associated with transitions between electron and tau flavor, and set two competitive limits associated with transitions between electron and muon flavor.

  10. Neutrino-4 experiment on the search for a sterile neutrino at the 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.; Chernyi, A. V.; Zherebtsov, O. M.; Martemyanov, V. P.; Tsinoev, 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.; Matrosov, M. Yu.

    2015-10-01

    In view of the possibility of the existence of a sterile neutrino, test measurements of the dependence of the reactor antineutrino flux on the distance from the reactor core has been performed on SM-2 reactor with the Neutrino-2 detector model in the range of 6-11 m. Prospects of the search for reactor antineutrinos at short distances have been discussed.

  11. Neutrino-4 experiment on the search for a sterile neutrino at the SM-3 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Serebrov, A. P., E-mail: serebrov@pnpi.spb.ru; Ivochkin, V. G.; Samoylov, R. M.; Fomin, A. K.; Zinoviev, V. G.; Neustroev, P. V.; Golovtsov, V. L.; Gruzinsky, N. V.; Solovey, V. A.; Chernyi, A. V.; Zherebtsov, O. M. [National Research Centre “Kurchatov Institute,”, Konstantinov Petersburg Nuclear Physics Institute (Russian Federation); Martemyanov, V. P.; Tsinoev, V. G.; Tarasenkov, V. G.; Aleshin, V. I. [National Research Centre “Kurchatov Institute,” (Russian Federation); Petelin, A. L.; Pavlov, S. V.; Izhutov, A. L.; Sazontov, S. A.; Ryazanov, D. K. [State Scientific Centre Research Institute of Atomic Reactors (Russian Federation); and others

    2015-10-15

    In view of the possibility of the existence of a sterile neutrino, test measurements of the dependence of the reactor antineutrino flux on the distance from the reactor core has been performed on SM-2 reactor with the Neutrino-2 detector model in the range of 6–11 m. Prospects of the search for reactor antineutrinos at short distances have been discussed.

  12. Nuclear heating experiments for the validation of fusion reactor shielding performance

    Energy Technology Data Exchange (ETDEWEB)

    Batistoni, P. [Associazione EURATOM-ENEA sulla Fusione, Frascati (Italy); Angelone, M. [Associazione EURATOM-ENEA sulla Fusione, Frascati (Italy); Pillon, M. [Associazione EURATOM-ENEA sulla Fusione, Frascati (Italy); Rado, V. [Associazione EURATOM-ENEA sulla Fusione, Frascati (Italy)

    1997-05-01

    Nuclear heating experiments were performed at the 14 MeV Frascati neutron generator, with the purpose of validating the shielding performance of the International Thermonuclear Experiment Reactor (ITER) shielding system (first wall, shielding blanket and vacuum vessel). The experiments consisted of the irradiation, with 14 MeV neutrons, of large assemblies that simulated, for size and material composition, the shielding system and the toroidal field magnet. The nuclear heating was measured with different thermo-luminescent detectors, as a function of the penetration depth inside the assemblies. The experimental results obtained were compared with the same quantities calculated with the same nuclear database and code used for nuclear heating calculations in the ITER design. (orig.)

  13. Evaluation of Concepts for Mulitiple Application Thermal Reactor for Irradiation eXperiments (MATRIX)

    Energy Technology Data Exchange (ETDEWEB)

    Pope, Michael A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ryskamp, John M. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2013-09-01

    The Advanced Test Reactor (ATR) is a high power density test reactor specializing in fuel and materials irradiation. For more than 45 years, the ATR has provided irradiations of materials and fuels testing along with radioisotope production. Originally operated primarily in support of the Offcie of Naval Reactors (NR), the mission has gradually expanded to cater to other customers, such as the DOE Office of Nuclear Energy (NE), private industry, and universities. Unforeseen circumstances may lead to the decommissioning of ATR, thus leaving the U.S. Government without a large-scale materials irradiation capability to meet the needs of its nuclear energy and naval reactor missions. In anticipation of this possibility, work was performed under the Laboratory Directed Research and Development (LDRD) program to investigate test reactor concepts that could satisfy the current missions of the ATR along with an expanded set of secondary missions. This work can be viewed as an update to a project from the 1990’s called the Broad Application Test Reactor (BATR). In FY 2012, a survey of anticipated customer needs was performed, followed by analysis of the original BATR concepts with fuel changed to low-enriched uranium. Departing from these original BATR designs, four concepts were identified for further analysis in FY2013. The project informally adopted the acronym MATRIX (Multiple-Application Thermal Reactor for Irradiation eXperiments). This report discusses analysis of the four MATRIX concepts along with a number of variations on these main concepts. Designs were evaluated based on their satisfaction of anticipated customer requirements and the “Cylindrical” variant was selected for further analysis of options. This downselection should be considered preliminary and the backup alternatives should include the other three main designs. The baseline Cylindrical MATRIX design is expected to be capable of higher burnup than the ATR (or longer cycle length given a

  14. In-Space technology experiments program. A high efficiency thermal interface (using condensation heat transfer) between a 2-phase fluid loop and heatpipe radiator: Experiment definition phase

    Science.gov (United States)

    Pohner, John A.; Dempsey, Brian P.; Herold, Leroy M.

    1990-01-01

    Space Station elements and advanced military spacecraft will require rejection of tens of kilowatts of waste heat. Large space radiators and two-phase heat transport loops will be required. To minimize radiator size and weight, it is critical to minimize the temperature drop between the heat source and sink. Under an Air Force contract, a unique, high-performance heat exchanger is developed for coupling the radiator to the transport loop. Since fluid flow through the heat exchanger is driven by capillary forces which are easily dominated by gravity forces in ground testing, it is necessary to perform microgravity thermal testing to verify the design. This contract consists of an experiment definition phase leading to a preliminary design and cost estimate for a shuttle-based flight experiment of this heat exchanger design. This program will utilize modified hardware from a ground test program for the heat exchanger.

  15. Comparison of three small-break loss-of-coolant accident tests with different break locations using the system-integrated modular advanced reactor-integral test loop facility to estimate the safety of the smart design

    Directory of Open Access Journals (Sweden)

    Hwang Bae

    2017-08-01

    Full Text Available Three small-break loss-of-coolant accident (SBLOCA tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor, i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal–hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

  16. Comparison of three small-break loss-of-coolant accident tests with different break locations using the system-integrated modular advanced reactor-integral test loop facility to estimate the safety of the smart design

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Hwang; Ryu, Sung Uk; Yi, Sung Jae; Park, Hyun Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Dong Eok [Dept. of Precision Mechanical Engineering, Kyungpook National University, Sangju (Korea, Republic of)

    2017-08-15

    Three small-break loss-of-coolant accident (SBLOCA) tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor), i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal–hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

  17. Full-scale 3-D finite element modeling of a two-loop pressurized water reactor for heat transfer, thermal–mechanical cyclic stress analysis, and environmental fatigue life estimation

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-12-15

    Highlights: • Full-scale 3-D finite element model. • Pressurized water reactor. • Heat transfer analysis. • Thermal–mechanical stress analysis. • Environmental fatigue life estimation. - Abstract: This paper discusses a system-level finite element model of a two-loop pressurized water reactor (PWR). Based on this model, system-level heat transfer analysis and subsequent sequentially coupled thermal–mechanical stress analysis were performed for typical thermal–mechanical fatigue cycles. The in-air fatigue lives of example components, such as the hot and cold legs, were estimated on the basis of stress analysis results, ASME in-air fatigue life estimation criteria, and fatigue design curves. Furthermore, environmental correction factors and associated PWR environment fatigue lives for the hot and cold legs were estimated by using estimated stress and strain histories and the approach described in US-NRC report: NUREG-6909.

  18. Modeling and Depletion Simulations for a High Flux Isotope Reactor Cycle with a Representative Experiment Loading

    Energy Technology Data Exchange (ETDEWEB)

    Chandler, David [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Betzler, Ben [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Hirtz, Gregory John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Ilas, Germina [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Sunny, Eva [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division

    2016-09-01

    The purpose of this report is to document a high-fidelity VESTA/MCNP High Flux Isotope Reactor (HFIR) core model that features a new, representative experiment loading. This model, which represents the current, high-enriched uranium fuel core, will serve as a reference for low-enriched uranium conversion studies, safety-basis calculations, and other research activities. A new experiment loading model was developed to better represent current, typical experiment loadings, in comparison to the experiment loading included in the model for Cycle 400 (operated in 2004). The new experiment loading model for the flux trap target region includes full length 252Cf production targets, 75Se production capsules, 63Ni production capsules, a 188W production capsule, and various materials irradiation targets. Fully loaded 238Pu production targets are modeled in eleven vertical experiment facilities located in the beryllium reflector. Other changes compared to the Cycle 400 model are the high-fidelity modeling of the fuel element side plates and the material composition of the control elements. Results obtained from the depletion simulations with the new model are presented, with a focus on time-dependent isotopic composition of irradiated fuel and single cycle isotope production metrics.

  19. Advancements in Development of Chemical-Looping Combustion: A Review

    Directory of Open Access Journals (Sweden)

    He Fang

    2009-01-01

    Full Text Available Chemical-looping combustion (CLC is a novel combustion technology with inherent separation of greenhouse CO2. Extensive research has been performed on CLC in the last decade with respect to oxygen carrier development, reaction kinetics, reactor design, system efficiencies, and prototype testing. Transition metal oxides, such as Ni, Fe, Cu, and Mn oxides, were reported as reactive species in the oxygen carrier particles. Ni-based oxygen carriers exhibited the best reactivity and stability during multiredox cycles. The performance of the oxygen carriers can be improved by changing preparation method or by making mixedoxides. The CLC has been demonstrated successfully in continuously operated prototype reactors based on interconnected fluidized-bed system in the size range of 0.3–50 kW. High fuel conversion rates and almost 100%  CO2 capture efficiencies were obtained. The CLC system with two interconnected fluidized-bed reactors was considered the most suitable reactor design. Development of oxygen carriers with excellent reactivity and stability is still one of the challenges for CLC in the near future. Experiences of building and operating the large-scale CLC systems are needed before this technology is used commercially. Chemical-looping reforming (CLR and chemical-looping hydrogen (CLH are novel chemical-looping techniques to produce synthesis gas and hydrogen deserving more attention and research.

  20. IEA-R1 Nuclear Research Reactor: 58 Years of Operating Experience and Utilization for Research, Teaching and Radioisotopes Production

    Energy Technology Data Exchange (ETDEWEB)

    Cardenas, Jose Patricio Nahuel; Filho, Tufic Madi; Saxena, Rajendra; Filho, Walter Ricci [Nuclear and Energy Research Institute, IPEN-CNEN/SP, Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP, Av. Prof. Lineu Prestes 2242 Cid Universitaria CEP: 05508-000- Sao Paulo-SP (Brazil)

    2015-07-01

    IEA-R1 research reactor at the Instituto de Pesquisas Energeticas e Nucleares (Nuclear and Energy Research Institute) IPEN, Sao Paulo, Brazil is the largest power research reactor in Brazil, with a maximum power rating of 5 MWth. It is being used for basic and applied research in the nuclear and neutron related sciences, for the production of radioisotopes for medical and industrial applications, and for providing services of neutron activation analysis, real time neutron radiography, and neutron transmutation doping of silicon. IEA-R1 is a swimming pool reactor, with light water as the coolant and moderator, and graphite and beryllium as reflectors. The reactor was commissioned on September 16, 1957 and achieved its first criticality. It is currently operating at 4.5 MWth with a 60-hour cycle per week. In the early sixties, IPEN produced {sup 131}I, {sup 32}P, {sup 198}Au, {sup 24}Na, {sup 35}S, {sup 51}Cr and labeled compounds for medical use. During the past several years, a concerted effort has been made in order to upgrade the reactor power to 5 MWth through refurbishment and modernization programs. One of the reasons for this decision was to produce {sup 99}Mo at IPEN. The reactor cycle will be gradually increased to 120 hours per week continuous operation. It is anticipated that these programs will assure the safe and sustainable operation of the IEA-R1 reactor for several more years, to produce important primary radioisotopes {sup 99}Mo, {sup 125}I, {sup 131}I, {sup 153}Sm and {sup 192}Ir. Currently, all aspects of dealing with fuel element fabrication, fuel transportation, isotope processing, and spent fuel storage are handled by IPEN at the site. The reactor modernization program is slated for completion by 2015. This paper describes 58 years of operating experience and utilization of the IEA-R1 research reactor for research, teaching and radioisotopes production. (authors)

  1. Integral neutronics experiments in analytical mockups for blanket of a hybrid reactor

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Rong, E-mail: liurongzy@163.com; Zhu, Tonghua; Lu, Xinxin; Wang, Xinhua; Yan, Xiaosong; Feng, Song; Yang, Yiwei; Wang, Mei; Jiang, Li

    2014-12-15

    Highlights: • For checking property of the hybrid blanket by integral experiments, three mockups are established. • In spherical mockup with depleted uranium and cubic mockup with natural uranium, the plutonium production rates and uranium fission rates are measured. • In spherical mockup with depleted uranium and LiPb, tritium production rates are measured. • The measured results are compared to the calculated ones with MCNP-4B code and ENDF/B-VI library data. - Abstract: The paper describes recent progress in integral neutronics experiments in the analytical mockups for the blanket in a fusion-fission hybrid energy reactor. A conceptual blanket of the hybrid reactor is mainly loaded with natural uranium and lithium material. In the fission fuel region, uranium material and light water are arranged alternately. The mockups of the conceptual blanket are designed and used for checking neutron property of the blanket by integral experiments. Based on materials available, the spherical fission mockup for fission research and plutonium production consists of three layers of depleted uranium shells and several layers of polyethylene and graphite shells. The spherical lithium mockup for tritium production consists of depleted uranium and LiPb alloy shells. The cubic mockup consists of natural uranium and polyethylene and its structure is basically consistent with one of the fuel region. In the mockups with the D-T neutron source, the plutonium production rates, uranium fission rates and tritium production rates are measured, separately. The measured results are compared to the calculated ones with MCNP-4B code and ENDF/B-VI library data.

  2. A comparison of coal char reactivity determined from thermogravimetric and laminar flow reactor experiments

    Energy Technology Data Exchange (ETDEWEB)

    Zolin, A.; Jensen, A.; Pedersen, L.S.; Dam-Johansen, K.; Toerslev, P. [Technical University of Denmark, Lyngby (Denmark). Dept. of Chemical Engineering

    1998-03-01

    The reactivity of nine different coals ranking from subbituminous to low-volatile bituminous has been studied by thermogravimetric analysis (TGA). At a standard set of conditions a qualitative fuel reactivity classification (ranking) with respect to one of the coals, Cerrejon, is presented. Particle reaction rates per unit external surface area and a normalized reactivity index based on raw experimental data were used as reactivity parameters to compare the fuels. The TGA chars were prepared at 900{degree}C with 15 min holding time and then combusted in a 20 mol % O{sub 2} environment at several temperatures in the range 450-650{degree}C. TGA reaction rate data were adequately interpreted by a random pore model. However, at 650{degree}C it is believed that particle ignition gave rise to a char reaction rate behavior that the model was incapable of describing properly. Except for two Southern Hemisphere coals, the reactivity ranking obtained with the TGA apparatus at a combustion temperature of 550{degree}C agrees well with a corresponding classification based on experiments carried out in another study with a laminar flow reactor (LFR) at {approximately} 1400{degree}C. The maximum difference in reaction rates based on external surface area between the coal chars in the low-temperature TGA experiments was 1 order of magnitude higher than in the high-temperature LFT experiments, due to the increasing effect of pore diffusion and thermal annealing of the coal chars in the LFR tests. The similarity in the reactivity ranking obtained for the Northern Hemisphere coals from both reactor systems indicates that a ranking can be performed by thermogravimetric analysis. This provides a simple means for determining a fuel reactivity ranking that could be applied to full scale suspension fired plants. 28 refs., 8 figs., 4 tabs.

  3. Reaction Rate Benchmark Experiments with Miniature Fission Chambers at the Slovenian TRIGA Mark II Reactor

    Science.gov (United States)

    Štancar, Žiga; Kaiba, Tanja; Snoj, Luka; Barbot, Loïc; Destouches, Christophe; Fourmentel, Damien; Villard, Jean-François AD(; )

    2018-01-01

    A series of fission rate profile measurements with miniature fission chambers, developed by the Commisariat á l'énergie atomique et auxénergies alternatives, were performed at the Jožef Stefan Institute's TRIGA research reactor. Two types of fission chambers with different fissionable coating (235U and 238U) were used to perform axial fission rate profile measurements at various radial positions and several control rod configurations. The experimental campaign was supported by an extensive set of computations, based on a validated Monte Carlo computational model of the TRIGA reactor. The computing effort included neutron transport calculations to support the planning and design of the experiments as well as calculations to aid the evaluation of experimental and computational uncertainties and major biases. The evaluation of uncertainties was performed by employing various types of sensitivity analyses such as experimental parameter perturbation and core reaction rate gradient calculations. It has been found that the experimental uncertainty of the measurements is sufficiently low, i.e. the total relative fission rate uncertainty being approximately 5 %, in order for the experiments to serve as benchmark experiments for validation of fission rate profiles. The effect of the neutron flux redistribution due to the control rod movement was studied by performing measurements and calculations of fission rates and fission chamber responses in different axial and radial positions at different control rod configurations. It was confirmed that the control rod movement affects the position of the maximum in the axial fission rate distribution, as well as the height of the local maxima. The optimal detector position, in which the redistributions would have minimum effect on its signal, was determined.

  4. Proposal on experience learning of a nuclear reactor for children in future. A basic concept on a nuclear reactor facility for demonstration and education

    Energy Technology Data Exchange (ETDEWEB)

    Murata, Takashi [Kyoto Univ., Graduate School of Energy Science, Kyoto (Japan); Yoshiki, Nobuya; Kinehara, Yoshiki; Nakagawa, Haruo

    2001-12-01

    The Science Council of Japan indicates in a proposal on R and D on nuclear energy forward the 21st Century that it is important to expand the educational object on nuclear energy from colleges and gradual schools to elementary, middle high schools. And, the Committee of Japan Nuclear Energy Industries also proposed that as an effort forward security of reliability and popularization of knowledge, completeness of learning chance on energy and nuclear energy in education such as usage of general learning time, concept on establishment of educational reactor for demonstration and experience, is essential. Here was described on a concept on establishment of nuclear reactor for demonstration and experience at objectives of common national peoples, which was based on results of searches and investigations carried out by authors and aimed to supply to a field to grow up a literary adequately and widely capable of judging various information on the peoples by focusing to effectiveness of empirical learning as a method of promoting corrective understanding of common citizens on high class technical system and by establishment of the reactor aiming at general education on nuclear energy at a place easily accessible by common citizens, such as large city. (G.K.)

  5. Electrical Capacitance Volume Tomography for the Packed Bed Reactor ISS Flight Experiment

    Science.gov (United States)

    Marashdeh, Qussai; Motil, Brian; Wang, Aining; Liang-Shih, Fan

    2013-01-01

    Fixed packed bed reactors are compact, require minimum power and maintenance to operate, and are highly reliable. These features make this technology a highly desirable unit operation for long duration life support systems in space. NASA is developing an ISS experiment to address this technology with particular focus on water reclamation and air revitalization. Earlier research and development efforts funded by NASA have resulted in two hydrodynamic models which require validation with appropriate instrumentation in an extended microgravity environment. To validate these models, the instantaneous distribution of the gas and liquid phases must be measured.Electrical Capacitance Volume Tomography (ECVT) is a non-invasive imaging technology recently developed for multi-phase flow applications. It is based on distributing flexible capacitance plates on the peripheral of a flow column and collecting real-time measurements of inter-electrode capacitances. Capacitance measurements here are directly related to dielectric constant distribution, a physical property that is also related to material distribution in the imaging domain. Reconstruction algorithms are employed to map volume images of dielectric distribution in the imaging domain, which is in turn related to phase distribution. ECVT is suitable for imaging interacting materials of different dielectric constants, typical in multi-phase flow systems. ECVT is being used extensively for measuring flow variables in various gas-liquid and gas-solid flow systems. Recent application of ECVT include flows in risers and exit regions of circulating fluidized beds, gas-liquid and gas-solid bubble columns, trickle beds, and slurry bubble columns. ECVT is also used to validate flow models and CFD simulations. The technology is uniquely qualified for imaging phase concentrations in packed bed reactors for the ISS flight experiments as it exhibits favorable features of compact size, low profile sensors, high imaging speed, and

  6. Comparative study on neutron data in integral experiments of MYRRHA mockup critical cores in the VENUS-F reactor

    Science.gov (United States)

    Krása, Antonín; Kochetkov, Anatoly; Baeten, Peter; Vittiglio, Guido; Wagemans, Jan; Bécares, Vicente

    2017-09-01

    VENUS-F is a fast, zero-power reactor with 30% wt. metallic uranium fuel and solid lead as coolant simulator. It serves as a mockup of the MYRRHA reactor core. This paper describes integral experiments performed in two critical VENUS-F core configurations (with and without graphite reflector). Discrepancies between experiments and Monte Carlo calculations (MCNP5) of keff, fission rate spatial distribution and reactivity effects (lead void and fuel Doppler) depending on a nuclear data library used (JENDL-4.0, ENDF-B-VII.1, JEFF-3.1.2, 3.2, 3.3T2) are presented.

  7. Comparative study on neutron data in integral experiments of MYRRHA mockup critical cores in the VENUS-F reactor

    Directory of Open Access Journals (Sweden)

    Krása Antonín

    2017-01-01

    Full Text Available VENUS-F is a fast, zero-power reactor with 30% wt. metallic uranium fuel and solid lead as coolant simulator. It serves as a mockup of the MYRRHA reactor core. This paper describes integral experiments performed in two critical VENUS-F core configurations (with and without graphite reflector. Discrepancies between experiments and Monte Carlo calculations (MCNP5 of keff, fission rate spatial distribution and reactivity effects (lead void and fuel Doppler depending on a nuclear data library used (JENDL-4.0, ENDF-B-VII.1, JEFF-3.1.2, 3.2, 3.3T2 are presented.

  8. Initial Results from the CHOOZ Long Baseline Reactor Neutrino Oscillation Experiment

    CERN Document Server

    Apollonio, M

    1998-01-01

    Initial results are presented from CHOOZ, a long-baseline reactor-neutrino vacuum-oscillation experiment. Electron antineutrinos were detected by a liquid scintillation calorimeter located at a distance of about 1 km. The detector was constructed in a tunnel protected from cosmic rays by a 300 MWE rock overburden. This massive shielding strongly reduced potentially troublesome backgrounds due to cosmic-ray muons, leading to a background rate of about one event per day, more than an order of magnitude smaller than the observed neutrino signal. From the statistical agreement between detected and expected neutrino event rates, we find (at 90% confidence level) no evidence for neutrino oscillations in the electron antineutrino disappearance mode for the parameter region given approximately by deltam**2 > 0.9 10**(-3) eV**2 for maximum mixing and (sin(2 theta)**2) > 0.18 for large deltam**2.

  9. Analyzing the thermionic reactor critical experiments. [thermal spectrum of uranium 235 core

    Science.gov (United States)

    Niederauer, G. F.

    1973-01-01

    The Thermionic Reactor Critical Experiments (TRCE) consisted of fast spectrum highly enriched U-235 cores reflected by different thicknesses of beryllium or beryllium oxide with a transition zone of stainless steel between the core and reflector. The mixed fast-thermal spectrum at the core reflector interface region poses a difficult neutron transport calculation. Calculations of TRCE using ENDF/B fast spectrum data and GATHER library thermal spectrum data agreed within about 1 percent for the multiplication factor and within 6 to 8 percent for the power peaks. Use of GAM library fast spectrum data yielded larger deviations. The results were obtained from DOT R Theta calculations with leakage cross sections, by region and by group, extracted from DOT RZ calculations. Delineation of the power peaks required extraordinarily fine mesh size at the core reflector interface.

  10. Closing the loop on improvement: Packaging experience in the Software Engineering Laboratory

    Science.gov (United States)

    Waligora, Sharon R.; Landis, Linda C.; Doland, Jerry T.

    1994-01-01

    As part of its award-winning software process improvement program, the Software Engineering Laboratory (SEL) has developed an effective method for packaging organizational best practices based on real project experience into useful handbooks and training courses. This paper shares the SEL's experience over the past 12 years creating and updating software process handbooks and training courses. It provides cost models and guidelines for successful experience packaging derived from SEL experience.

  11. Fission Product Transport and Source Terms in HTRs: Experience from AVR Pebble Bed Reactor

    Directory of Open Access Journals (Sweden)

    Rainer Moormann

    2008-01-01

    Full Text Available Fission products deposited in the coolant circuit outside of the active core play a dominant role in source term estimations for advanced small pebble bed HTRs, particularly in design basis accidents (DBA. The deposited fission products may be released in depressurization accidents because present pebble bed HTR concepts abstain from a gas tight containment. Contamination of the circuit also hinders maintenance work. Experiments, performed from 1972 to 88 on the AVR, an experimental pebble bed HTR, allow for a deeper insight into fission product transport behavior. The activity deposition per coolant pass was lower than expected and was influenced by fission product chemistry and by presence of carbonaceous dust. The latter lead also to inconsistencies between Cs plate out experiments in laboratory and in AVR. The deposition behavior of Ag was in line with present models. Dust as activity carrier is of safety relevance because of its mobility and of its sorption capability for fission products. All metal surfaces in pebble bed reactors were covered by a carbonaceous dust layer. Dust in AVR was produced by abrasion in amounts of about 5 kg/y. Additional dust sources in AVR were ours oil ingress and peeling of fuel element surfaces due to an air ingress. Dust has a size of about 1  m, consists mainly of graphite, is partly remobilized by flow perturbations, and deposits with time constants of 1 to 2 hours. In future reactors, an efficient filtering via a gas tight containment is required because accidents with fast depressurizations induce dust mobilization. Enhanced core temperatures in normal operation as in AVR and broken fuel pebbles have to be considered, as inflammable dust concentrations in the gas phase.

  12. Simulation of in-reactor experiments with the ELOCA.Mk5 code. AECL research No. AECL-11133

    Energy Technology Data Exchange (ETDEWEB)

    Klein, M.E.; Arimescu, V.I.; Carlucci, L.N.

    1994-12-31

    ELOCA.Mk5 is a FORTRAN-77 computer code developed to model the thermo-mechanical response and associated fission-product release behavior of CANDU fuel elements during high-temperature transients such as large-break loss of coolant accidents (LOCA). This paper reports the results of model runs conducted to simulate two in-reactor LOCA experiments, using ELOCA.Mk5 in the Mk4S mode. Mk4S is a thermo-mechanical mode capable of performing a multi-segment analysis of a CANDU fuel element, accounting for axial variations in sheath temperatures, metallurgical regions, and reactor neutron flux. The first LOCA experiment consisted of four elements subjected to a coolant depressurization in the Power Burst Facility at Idaho Falls. The second consisted of a single fresh element, with an artificially set internal gas pressure, subjected to a coolant depressurization in the NRX reactor.

  13. Status of the NGNP graphite creep experiments AGC-1 and AGC-2 irradiated in the advanced test reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover

    2014-05-01

    The United States Department of Energy's Next Generation Nuclear Plant (NGNP) Program will be irradiating six nuclear graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the next generation nuclear plant (NGNP) very high temperature gas reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain six peripheral stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six peripheral stacks will have three different compressive loads applied to the top half of three diametrically opposite pairs of specimen stacks, while a seventh stack will not have a compressive load. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be sampling the sweep gas effluent to determine if any oxidation or off-gassing of the specimens occurs during irradiation of the experiment.

  14. Natural Circulation Characteristics of a Symmetric Loop under Inclined Conditions

    Directory of Open Access Journals (Sweden)

    Xingtuan Yang

    2014-01-01

    Full Text Available Natural circulation is an important process for primary loops of some marine integrated reactors. The reactor works under inclined conditions when severe accidents happen to the ship. In this paper, to investigate the characteristics of natural circulation, experiments were conducted in a symmetric loop under the inclined angle of 0~45°. A CFD model was also set up to predict the behaviors of the loop beyond the experimental scope. Total circulation flow rate decreases with the increase of inclined angle. Meanwhile one circulation is depressed while the other is enhanced, and accordingly the disparity between the branch circulations arises and increases with the increase of inclined angle. Circulation only takes place in one branch circuit at large inclined angle. Also based on the CFD model, the influences of flow resistance distribution and loop configuration on natural circulation are predicted. The numerical results show that to design the loop with the configuration of big altitude difference and small width, it is favorable to reduce the influence of inclination; however too small loop width will cause severe reduction of circulation ability at large angle inclination.

  15. Optimal design of radioactive particle tracking experiments for flow mapping in opaque multiphase reactors.

    Science.gov (United States)

    Roy, Shantanu; Larachi, Faical; Al-Dahhan, M H; Duduković, M P

    2002-03-01

    In the past decade, radioactive particle tracking techniques have emerged in the field of chemical engineering and have become increasingly popular for non-invasive flow mapping of the hydrodynamics in multiphase reactors. Based on gamma-ray sensitization of an array of scintillation detectors, the Computer Automated Radioactive Particle Tracking (CARPT) technique measures flow fields by monitoring the actual motion path of a single discrete radioactive flow follower which has the physical properties of the phase whose motion is being followed. A limitation to the accuracy of CARPT lies in the error associated with the reconstruction of the tracer particle position which affects the space-resolution capability of the technique. It is of interest, therefore, to minimize this error by choosing wisely the best hardware and an optimal configuration of CARPT detectors' array. Such choices are currently based on experience, without firm scientific basis. In this paper, through theoretical modeling and simulation, we describe how the accuracy of a radioactive particle tracking setup may be assessed a priori. Through an example of a proposed implementation of CARPT on a gas-solids riser, we demonstrate how this knowledge can be used for choosing the hardware required for the experiment. Finally, we show how the optimal arrangement of detectors can be effected for maximum accuracy for a given amount of monetary investment for the experiment.

  16. The Paucity Problem: Where Have All the Space Reactor Experiments Gone?

    Energy Technology Data Exchange (ETDEWEB)

    Bess, John D.; Marshall, Margaret A.

    2016-10-01

    The Handbooks of the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPhEP) together contain a plethora of documented and evaluated experiments essential in the validation of nuclear data, neutronics codes, and modeling of various nuclear systems. Unfortunately, only a minute selection of handbook data (twelve evaluations) are of actual experimental facilities and mockups designed specifically for space nuclear research. There is a paucity problem, such that the multitude of space nuclear experimental activities performed in the past several decades have yet to be recovered and made available in such detail that the international community could benefit from these valuable historical research efforts. Those experiments represent extensive investments in infrastructure, expertise, and cost, as well as constitute significantly valuable resources of data supporting past, present, and future research activities. The ICSBEP and IRPhEP were established to identify and verify comprehensive sets of benchmark data; evaluate the data, including quantification of biases and uncertainties; compile the data and calculations in a standardized format; and formally document the effort into a single source of verified benchmark data. See full abstract in attached document.

  17. Flip, feedback and fly: Using LOOP to enhance the professional experience of initial teacher education

    National Research Council Canada - National Science Library

    Phillipson, Shane N; Phillipson, Sivanes; Cooper, David G

    2015-01-01

    .... Problems can arise, however, when the evaluation of their professional experience against the Standards shifts from the providers of teacher education programmes to school-based supervising teachers...

  18. Picosecond lasing in ytterbium fibre laser with nonlinear optical loop mirror: experiment and numerical simulation

    Science.gov (United States)

    Borodkin, A. A.; Khudyakov, D. V.; Vartapetov, S. K.

    2015-02-01

    The operation regimes of a pulsed all-normal-dispersion polarisation-maintaining fibre laser with a nonlinear optical loop mirror are studied. The use of polarisation-maintaining fibres ensures polarisation and temperature stability of output radiation. The lasing and instability thresholds of the pulsed laser are determined experimentally. A spectral filter placed in the cavity makes it possible to change the centre wavelength of laser radiation within the range 1.02 - 1.05 μm with a spectral full width at half maximum of 2 nm. The average output power is 7 mW, which corresponds to a pulse energy of 0.8 nJ. The autocorrelation function width of the output pulse is 50 ps. The minimum pulse duration achieved after compression by an external pair of diffraction gratings is 1.8 ps. The dynamics of the temporal and spectral parameters of laser pulses is studied using mathematical simulation based on numerical solution of the nonlinear Schrödinger equation. The simulation results coincide with experimental data with a high accuracy.

  19. Picosecond lasing in ytterbium fibre laser with nonlinear optical loop mirror: experiment and numerical simulation

    Energy Technology Data Exchange (ETDEWEB)

    Borodkin, A A; Khudyakov, D V; Vartapetov, S K [Physics Instrumentation Center, A.M. Prokhorov General Physics Institute, Russian Academy of Sciences, Troitsk, Moscow Region (Russian Federation)

    2015-02-28

    The operation regimes of a pulsed all-normal-dispersion polarisation-maintaining fibre laser with a nonlinear optical loop mirror are studied. The use of polarisation-maintaining fibres ensures polarisation and temperature stability of output radiation. The lasing and instability thresholds of the pulsed laser are determined experimentally. A spectral filter placed in the cavity makes it possible to change the centre wavelength of laser radiation within the range 1.02 – 1.05 μm with a spectral full width at half maximum of 2 nm. The average output power is 7 mW, which corresponds to a pulse energy of 0.8 nJ. The autocorrelation function width of the output pulse is 50 ps. The minimum pulse duration achieved after compression by an external pair of diffraction gratings is 1.8 ps. The dynamics of the temporal and spectral parameters of laser pulses is studied using mathematical simulation based on numerical solution of the nonlinear Schrödinger equation. The simulation results coincide with experimental data with a high accuracy. (lasers)

  20. Large loop excision of the transformation zone for treating cervical intraepithelial neoplasia: a 12-year experience.

    Science.gov (United States)

    Paraskevaidis, E; Koliopoulos, G; Malamou-Mitsi, V; Zikopoulos, K; Paschopoulos, M; Pappa, L; Agnantis, N J; Loli, D E

    2001-01-01

    Although the existing evidence suggests that there is no obviously superior conservative method for treating cervical intraepithelial neoplasia (CIN), one of the most widely used is the large loop excision of the transformation zone (LLETZ). A total of 897 women who were treated with LLETZ at our colposcopy clinic from 1989 to 2000 were retrospectively studied. Forty women did not have significant cervical pathology (4.5% over-treatment rate). Clear margins of excision were obtained in 748 (88.5%) of the 845 cases of CIN or microinvasive cancers. Treatment failure rates were 4.7% for clear margins and 26.8% for involved or uncertain. LLETZ is a fast and reliable method of treating CIN and microinvasive carcinoma. Generalized cauterization of the resulting crater should be avoided and satellite HPV lesions ablated. Involved margins have a higher treatmentfailure rate, therefore a larger excision is recommended as cervical craters regenerate. Treatment in pregnant women can be delayed until postpartum provided they have adequate surveillance during pregnancy.

  1. Experiences in the emptying of waste silos containing solid nuclear waste from graphite- moderated reactors

    Energy Technology Data Exchange (ETDEWEB)

    Wall, S.; Schwarz, T. [RWE NUKEM Limited, B7 Windscale, Seascale, Cumbria CA20 1PF (United Kingdom)

    2003-07-01

    Before reactor sites can be handed over for ultimate decommissioning, at some sites silos containing waste from operations need to be emptied. The form and physical condition of the waste demands sophisticated retrieval technologies taking into account the onsite situation in terms of infrastructure and silo geometry. Furthermore, in the case of graphite moderated reactors, this waste usually includes several tonnes of graphite waste requiring special HVAC and dust handling measures. RWE NUKEM Group has already performed several contracts dealing with such emptying tasks. Of particular interest for the upcoming decommissioning projects in France might be the activities at Vandellos, Spain and Trawsfynnyd, UK. Retrieval System for Vandellos NPP is discussed. Following an international competitive tender exercise, RWE NUKEM won the contract to provide a turn-key retrieval system. This involved the design, manufacture and installation of a system built around the modules of a 200 kg capacity version of the ARTISAN manipulator system. The ARTISAN 200 manipulator, with remote slave arm detach facility, was deployed on a telescopic mast inserted into the silos through the roof penetrations. The manipulator deployed a range of tools to gather the waste and load it into a transfer basket, deployed through an adjacent penetration. After commissioning, the system cleared the vaults in less than the scheduled period with no failures. At the Trawsfynnyd Magnox plants two types of intermediate level waste (ILW) accumulated on site; namely Miscellaneous Activated Components (MAC) and Fuel Element Debris (FED). MAC is predominantly components that have been activated by the reactor core and then discharged. FED mainly consists of fuel cladding produced when fuel elements were prepared for dispatch to the reprocessing facility. RWE NUKEM Ltd. was awarded a contract to design, supply, commission and operate equipment to retrieve, pack and immobilize the two waste streams. Major

  2. Uncertainty quantification of calculated temperatures for advanced gas reactor fuel irradiation experiments

    Energy Technology Data Exchange (ETDEWEB)

    Pham, Binh Thi-Cam [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hawkes, Grant Lynn [Idaho National Lab. (INL), Idaho Falls, ID (United States); Einerson, Jeffrey James [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-08-01

    This paper presents the quantification of uncertainty of the calculated temperature data for the Advanced Gas Reactor (AGR) fuel irradiation experiments conducted in the Advanced Test Reactor at Idaho National Laboratory in support of the Advanced Reactor Technology Research and Development program. Recognizing uncertainties inherent in physics and thermal simulations of the AGR tests, the results of the numerical simulations are used in combination with statistical analysis methods to improve qualification of measured data. The temperature simulation data for AGR tests are also used for validation of the fission product transport and fuel performance simulation models. These crucial roles of the calculated fuel temperatures in ensuring achievement of the AGR experimental program objectives require accurate determination of the model temperature uncertainties. To quantify the uncertainty of AGR calculated temperatures, this study identifies and analyzes ABAQUS model parameters of potential importance to the AGR predicted fuel temperatures. The selection of input parameters for uncertainty quantification of the AGR calculated temperatures is based on the ranking of their influences on variation of temperature predictions. Thus, selected input parameters include those with high sensitivity and those with large uncertainty. Propagation of model parameter uncertainty and sensitivity is then used to quantify the overall uncertainty of AGR calculated temperatures. Expert judgment is used as the basis to specify the uncertainty range for selected input parameters. The input uncertainties are dynamic accounting for the effect of unplanned events and changes in thermal properties of capsule components over extended exposure to high temperature and fast neutron irradiation. The sensitivity analysis performed in this work went beyond the traditional local sensitivity. Using experimental design, analysis of pairwise interactions of model parameters was performed to establish

  3. Decay heat experiment and validation of calculation code systems for fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Maekawa, Fujio; Ikeda, Yujiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Wada, Masayuki

    1999-10-01

    Although accurate estimation of decay heat value is essential for safety analyses of fusion reactors against loss of coolant accidents and so on, no experimental work has been devoted to validating the estimation. Hence, a decay heat measurement experiment was performed as a task (T-339) of ITER/EDA. A new detector, the Whole Energy Absorption Spectrometer (WEAS), was developed for accurate and efficient measurements of decay heat. Decay heat produced in the thirty-two sample materials which were irradiated by 14-MeV neutrons at FNS/JAERI were measured with WEAS for a wide cooling time period from 1 min to 400 days. The data presently obtained were the first experimental decay heat data in the field of fusion. Validity of decay heat calculation codes of ACT4 and CINAC-V4, activation cross section libraries of FENDL/A-2.0 and JENDL Activation File, and decay data was investigated through analyses of the experiment. As a result, several points that should be modified were found in the codes and data. After solving the problems, it was demonstrated that decay heat valued calculated for most of samples were in good agreement with the experimental data. Especially for stainless steel 316 and copper, which were important materials for ITER, decay heat could be predicted with accuracy of {+-}10%. (author)

  4. Set of benchmark experiments on slit shielding compositions of thermonuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Andreev, M.I.; Afanasiev, V.V.; Belevitin, A.G.; Karaulov, A.V.; Romodanov, V.L. E-mail: rom@lng.mephi.msk.su; Sakharov, V.K.; Tikhomirov, G.V.; Vasiliev, A.P.; Kandiev, Ya. Z.; Lyutov, V.D.; Sokolov, Yu. A.; Terekhin, V.A.; Shmakov, V.M.; Androsenko, P.A.; Semenov, V.P.; Trykov, L.A.; Lopatkin, A.V.; Muratov, V.G

    2001-09-01

    The paper is based on the results of the ISTC project no. 180 that has recently been completed. The aim of the project was the development of methodical, hardware and design basis to carry out computational and experimental research on non-uniform shieldings of thermonuclear reactors. As a result a set of benchmark experiments were created. On their basis verification of the domestic and foreign computational codes with the nuclear data estimated was realized. For these purposes the iron hollow slits shielding compositions irradiated with 14.8 MeV energy neutrons were studied. The experimental installations allowed research of the shielding compositions with the following characteristics: a solid structure, a structure with one slit of a central symmetry, and the structures with asymmetric slits and with two slits. The thickness of shielding compositions in this research was 500 mm. The results of experiments were compared to the results of calculations by means of the MCNP-4a and PRIZMA computing codes with use of the FENDL-1.1, FENDL-2, JENDL-3.2 and BAS-78 libraries of nuclear data. The results of comparison made it possible to obtain the recommendations for use of these nuclear data.

  5. Preparations for deuterium--tritium experiments on the Tokamak Fusion Test Reactor*

    Energy Technology Data Exchange (ETDEWEB)

    Hawryluk, R.J.; Adler, H.; Alling, P.; Ancher, C.; Anderson, H.; Anderson, J.L.; Anderson, J.W.; Arunasalam, V.; Ascione, G.; Aschroft, D.; Barnes, C.W.; Barnes, G.; Batchelor, D.B.; Bateman, G.; Batha, S.; Baylor, L.A.; Beer, M.; Bell, M.G.; Biglow, T.S.; Bitter, M.; Blanchard, W.; Bonoli, P.; Bretz, N.L.; Brunkhorst, C.; Budny, R.; Burgess, T.; Bush, H.; Bush, C.E.; Camp, R.; Caorlin, M.; Carnevale, H.; Chang, Z.; Chen, L.; Cheng, C.Z.; Chrzanowski, J.; Collazo, I.; Collins, J.; Coward, G.; Cowley, S.; Cropper, M.; Darrow, D.S.; Daugert, R.; DeLooper, J.; Duong, H.; Dudek, L.; Durst, R.; Efthimion, P.C.; Ernst, D.; Faunce, J.; Fonck, R.J.; Fredd, E.; Fredrickson, E.; Fromm, N.; Fu, G.Y.; Furth, H.P.; Garzotto, V.; Gentile, C.; Gettelfinger, G.; Gilbert, J.; Gioia, J.; Goldfinger, R.C.; Golian, T.; Gorelenkov, N.; Gouge, M.J.; Grek, B.; Grisham, L.R.; Hammett, G.; Hanson, G.R.; Heidbrink, W.; Hermann, H.W.; Hill, K.W.; Hirshman, S.; Hoffman, D.J.; Hosea, J.; Hulse, R.A.; Hsuan, H.; Ja

    1994-05-01

    The final hardware modifications for tritium operation have been completed for the Tokamak Fusion Test Reactor (TFTR) [Fusion Technol. [bold 21], 1324 (1992)]. These activities include preparation of the tritium gas handling system, installation of additional neutron shielding, conversion of the toroidal field coil cooling system from water to a Fluorinert[sup TM] system, modification of the vacuum system to handle tritium, preparation, and testing of the neutral beam system for tritium operation and a final deuterium--deuterium (D--D) run to simulate expected deuterium--tritium (D--T) operation. Testing of the tritium system with low concentration tritium has successfully begun. Simulation of trace and high power D--T experiments using D--D have been performed. The physics objectives of D--T operation are production of [approx]10 MW of fusion power, evaluation of confinement, and heating in deuterium--tritium plasmas, evaluation of [alpha]-particle heating of electrons, and collective effects driven by alpha particles and testing of diagnostics for confined [alpha] particles. Experimental results and theoretical modeling in support of the D--T experiments are reviewed.

  6. Experiments and Modelling Techniques for Heat and Mass Transfer in Light Water Reactors

    Directory of Open Access Journals (Sweden)

    W. Ambrosini

    2009-01-01

    Full Text Available The paper summarizes the lesson learned from theoretical and experimental activities performed at the University of Pisa, Pisa, Italy, in past decades in order to develop a general methodology of analysis of heat and mass transfer phenomena of interest for nuclear reactor applications. An overview of previously published results is proposed, highlighting the rationale at the basis of the performed work and its relevant conclusions. Experimental data from different sources provided information for model development and assessment. They include condensation experiments performed at SIET (Piacenza, Italy on the PANTHERS prototypical PCCS module, falling film evaporation tests for simulating AP600-like outer shell spraying conditions, performed at the University of Pisa, experimental data concerning condensation on finned tubes, collected by CISE (Piacenza, Italy in the frame of the INCON EU Project, and experimental tests performed in the CONAN experimental facility installed at the University of Pisa. The experience gained in these activities is critically reviewed and discussed to highlight the relevant obtained conclusions and the perspectives for future work.

  7. Toward a Mechanistic Source Term in Advanced Reactors: A Review of Past U.S. SFR Incidents, Experiments, and Analyses

    Energy Technology Data Exchange (ETDEWEB)

    Bucknor, Matthew; Brunett, Acacia J.; Grabaskas, David

    2016-04-17

    In 2015, as part of a Regulatory Technology Development Plan (RTDP) effort for sodium-cooled fast reactors (SFRs), Argonne National Laboratory investigated the current state of knowledge of source term development for a metal-fueled, pool-type SFR. This paper provides a summary of past domestic metal-fueled SFR incidents and experiments and highlights information relevant to source term estimations that were gathered as part of the RTDP effort. The incidents described in this paper include fuel pin failures at the Sodium Reactor Experiment (SRE) facility in July of 1959, the Fermi I meltdown that occurred in October of 1966, and the repeated melting of a fuel element within an experimental capsule at the Experimental Breeder Reactor II (EBR-II) from November 1967 to May 1968. The experiments described in this paper include the Run-Beyond-Cladding-Breach tests that were performed at EBR-II in 1985 and a series of severe transient overpower tests conducted at the Transient Reactor Test Facility (TREAT) in the mid-1980s.

  8. Chemical Looping Combustion of Solid Fuels in a Laboratory Fluidized-bed Reactor Combustion de charges solides avec la boucle chimique dans un lit fluidisé de laboratoire

    Directory of Open Access Journals (Sweden)

    Leion H.

    2011-02-01

    Full Text Available When using solid fuel in a chemical looping system, the char fraction of the fuel needs to be gasified before syngas react with the oxygen carrier. This can be done inside the fuel reactor with fuel and oxygen carriers well mixed, and, since this gasification is comparably slow, this will be the time limiting step of such a system. An option is to use an oxygen carrier that is able to release gas-phase oxygen which can react with the fuel by normal combustion giving a significantly faster overall fuel conversion. This last option is generally referred to as Chemical Looping combustion with Oxygen Un-coupling (CLOU. In this work, an overview is given of parameters that affect the fuel conversion in laboratory CLC and CLOU experiments. The main factor determining the fuel conversion, in both CLC and CLOU, is the fuel itself. High-volatile fuels are generally more rapidly converted than low volatile fuels. This difference in fuel conversion rate is more pronounced in CLC than in CLOU. However, the fuel conversion is also, both for CLC and CLOU, increased by increasing temperature. Increased steam and SO2 fraction in the surrounding gas will also enhance the fuel conversion in CLC. CO2 gasification in CLC appears to be very slow in comparison to steam gasification. H2 can inhibit fuel gasification in CLC whereas CO did not seem to have any effect. Possible deactivation of oxygen carriers due to SO2 or ash also has to be considered. Lorsque l’on utilise des combustibles solides dans la boucle chimique (CLC pour Chemical Looping Combustion, il est nécessaire de gazéifier le char avant de faire la combustion du gaz de synthèse au contact du transporteur d’oxygène. Ces réactions peuvent s’effectuer dans le réacteur fuel, dans lequel le combustible et le transporteur d’oxygène sont bien mélangés. Cependant, la gazéification du charbon est lente et reste l’étape limitante du processus de combustion dans ces conditions. Une alternative

  9. Complete Sensitivity/Uncertainty Analysis of LR-0 Reactor Experiments with MSRE FLiBe Salt and Perform Comparison with Molten Salt Cooled and Molten Salt Fueled Reactor Models

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Mueller, Don [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Patton, Bruce W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-12-01

    In September 2016, reactor physics measurements were conducted at Research Centre Rez (RC Rez) using the FLiBe (2 7LiF + BeF2) salt from the Molten Salt Reactor Experiment (MSRE) in the LR-0 low power nuclear reactor. These experiments were intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems using FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL), in collaboration with RC Rez, performed sensitivity/uncertainty (S/U) analyses of these experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objectives of these analyses were (1) to identify potential sources of bias in fluoride salt-cooled and salt-fueled reactor simulations resulting from cross section uncertainties, and (2) to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a final report on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. In the future, these S/U analyses could be used to inform the design of additional FLiBe-based experiments using the salt from MSRE. The key finding of this work is that, for both solid and liquid fueled fluoride salt reactors, radiative capture in 7Li is the most significant contributor to potential bias in neutronics calculations within the FLiBe salt.

  10. Air scaling and modeling studies for the 1/5-scale mark I boiling water reactor pressure suppression experiment

    Energy Technology Data Exchange (ETDEWEB)

    Lai, W.; McCauley, E.W.

    1978-01-04

    Results of table-top model experiments performed to investigate pool dynamics effects due to a postulated loss-of-coolant accident (LOCA) for the Peach Bottom Mark I boiling water reactor containment system guided subsequent conduct of the 1/5-scale torus experiment and provided new insight into the vertical load function (VLF). Pool dynamics results were qualitatively correct. Experiments with a 1/64-scale fully modeled drywell and torus showed that a 90/sup 0/ torus sector was adequate to reveal three-dimensional effects; the 1/5-scale torus experiment confirmed this.

  11. Full-Scale 3-D Finite Element Modeling of a Two-Loop Pressurized Water Reactor for Heat Transfer, Thermal-Mechanical Cyclic Stress Analysis, and Environmental Fatigue Life Estimation

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-12-15

    This paper discusses a system-level finite element model of a two-loop pressurized water reactor (PWR). Based on this model, system-level heat transfer analysis and subsequent sequentially coupled thermal-mechanical stress analysis were performed for typical thermal-mechanical fatigue cycles. The in-air fatigue lives of example components, such as the hot and cold legs, were estimated on the basis of stress analysis results, ASME in-air fatigue life estimation criteria, and fatigue design curves. Furthermore, environmental correction factors and associated PWR environment fatigue lives for the hot and cold legs were estimated by using estimated stress and strain histories and the approach described in US-NRC report: NUREG-6909.

  12. Study of the Open Loop and Closed Loop Oscillator Techniques

    Energy Technology Data Exchange (ETDEWEB)

    Imel, George R. [Idaho State Univ., Pocatello, ID (United States); Baker, Benjamin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Riley, Tony [Knolls Atomic Power Lab. (KAPL), Schenectady, NY (United States); Langbehn, Adam [Puget Sound Naval Base, Bremerton, WA (United States); Aryal, Harishchandra [Idaho State Univ., Pocatello, ID (United States); Benzerga, M. Lamine [Idaho State Univ., Pocatello, ID (United States)

    2015-04-11

    This report presents the progress and completion of a five-year study undertaken at Idaho State University of the measurement of very small worth reactivity samples comparing open and closed loop oscillator techniques.The study conclusively demonstrated the equivalency of the two techniques with regard to uncertainties in reactivity values, i.e., limited by reactor noise. As those results are thoroughly documented in recent publications, in this report we will concentrate on the support work that was necessary. For example, we describe in some detail the construction and calibration of a pilot rod for the closed loop system. We discuss the campaign to measure the required reactor parameters necessary for inverse-kinetics. Finally, we briefly discuss the transfer of the open loop technique to other reactor systems.

  13. 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.

  14. A neutrino-induced deuteron disintegration experiment at the Krasnoyarsk nuclear reactor

    CERN Document Server

    Kozlov, Y V; Machulin, I N; Martemyanov, A V; Martemyanov, V P; Sabelnikov, A A; Tarasenkov, V G; Turbin, E V; Vyrodov, V N

    2002-01-01

    The results of studying antineutrino interactions with the nucleus of deuteron (CCD-and NCD reactions) and hydrogen (CCP) at the Krasnoyarsk underground reactor using Deuteron detector is presented. As a results, the cross sections for NCD and CCD were measured with 9% precision, and for the CCP precision is 3%: sigma sub e sub x sub p sup N sup C sup D = (3.35 +- 0.31) x 10 sup - sup 4 sup 4 cm sup 2 /fission sup 2 sup 3 sup 5 U, sigma sub e sub x sub p sup C sup C sup D = (1.08 +- 0.09) x 10 sup - sup 4 sup 4 cm sup 2 /fission sup 2 sup 3 sup 5 U, sigma sub e sub x sub p sup C sup C sup P = (6.39 +- 0.19) x 10 sup - sup 4 sup 3 cm sup 2 /fission sup 2 sup 3 sup 5 U. The precision of the experimental results is close to the theoretical one and is in a good agreement with the other experiments. The limit on the antineutrino oscillation parameters into the sterile state was obtained: DELTA m sup 2 <= 4.7 x 10 sup - sup 2 eV sup 2 , for sin sup 2 (2 theta) = 1.0 (68% C.L.). The comparison of the measured and...

  15. Thermal analysis to support decommissioning of the molten salt reactor experiment

    Energy Technology Data Exchange (ETDEWEB)

    Sulfredge, C.D.; Morris, D.G.; Park, J.E.; Williams, P.T.

    1996-06-01

    As part of the decommissioning process for the Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory, several thermal-sciences issues were addressed. Apparently a mixture of UF{sub 6} and F{sub 2} had diffused into the upper portion of one charcoal column in the MSRE auxiliary charcoal bed (ACB), leading to radiative decay heating and possible chemical reaction sources. A proposed interim corrective action was planned to remove the water from the ACB cell to reduce criticality and reactivity concerns and then fill the ACB cell with an inert material. This report describes design of a thermocouple probe to obtain temperature measurements for mapping the uranium deposit, as well as development of steady-state and transient numerical models for the heat transfer inside the charcoal column. Additional numerical modeling was done to support filling of the ACB cell. Results from this work were used to develop procedures for meeting the goals of the MSRE Remediation Project without exceeding appropriate thermal limits.

  16. A descriptive model of the molten salt reactor experiment after shutdown: Review of FY 1995 progress

    Energy Technology Data Exchange (ETDEWEB)

    Williams, D.F.; Del Cul, G.D.; Toth, L.M.

    1996-01-01

    During FY 1995 considerable progress was made toward gaining a better understanding of the chemistry and transport processes that continue to govern the behavior of the Molten Salt Reactor Experiment (MSRE). As measurements in the MSRE proceed, laboratory studies continue, and better analyses are available, our understanding of the state of the MSRE and the best path toward remediation improves. Because of the immediate concern about the deposit in the auxiliary charcoal bed (ACB), laboratory studies in the past year focused on carbon-fluorine chemistry. Secondary efforts were directed toward investigation of gas generation from MSRE salts by both radiolytic and nonradiolytic pathways. In addition to the laboratory studies, field measurements at the MSRE provided the basis for estimating the inventory of uranium and fluorine in the ACB. Analysis of both temperature and radiation measurements provided independent and consistent estimates of about 2.6 kg of uranium deposited in the top of the ACB. Further analysis efforts included a refinement in the estimates of the fuel- salt source term, the deposited decay energy, and the projected rate of radiolytic gas generation. This report also provides the background material necessary to explain new developments and to review areas of particular interest. The detailed history of the MSRE is extensively documented and is cited where appropriate. This work is also intended to update and complement the more recent MSRE assessment reports.

  17. Solvothermal recrystallization of α-calcium sulfate hemihydrate: Batch reactor experiments and kinetic modelling

    Science.gov (United States)

    Macedo Portela da Silva, Nayane; Rong, Yi; Espitalier, Fabienne; Baillon, Fabien; Gaunand, Alain

    2017-08-01

    Under appropriate temperature conditions, natural gypsum CaSO4·2H2O, dispersed in an aqueous solution, turns into calcium hemihydrate CaSO4·½H2O. This transformation is performed in a 2 L stirred baffled reactor, where the temperature increase is measured and controlled on line. The water content of the suspension and its size distribution are measured on samples during the transformation. Experiments are achieved at nominal temperature of 140 °C, with three initial solid mass fractions 0.5, 0.33 and 0.25. The transformation takes place through a dissolution followed by re-crystallization. A model is proposed which takes into account the size distribution of the particles of gypsum, their dissolution rate, primary and secondary nucleation and growth rates of calcium hemihydrate. The set of equations is solved with a MATLAB software, which allows to test the assumptions on the kinetics of the transformation and fit their parameters. A satisfying representation of the variations of the extent of transformation and of volume and surface mean diameters of the suspension is obtained.

  18. Localized fast neutron flux enhancement for damage experiments in a research reactor; Accroissement local du flux rapide pour des experiences de dommages dans un reacteur de recherche

    Energy Technology Data Exchange (ETDEWEB)

    Malouch, F

    2003-06-01

    In irradiation experiments on materials in the core of the Osiris reactor (CEA-Saclay) we seek to increase damage in irradiated samples and to reduce the duration of their stay in the core. Damage is essentially caused by fast neutrons (E {>=} 1 MeV); we have therefore pursued the possibility of a localized increase of their level in an irradiation experiment by using a flux converter device made up of fissile material arranged according to a suitable geometry that allows the converter to receive experiments. We have studied several parameters that are influential in the increase of fast neutron flux within the converter. We have also considered the problem of the converter's cooling in the core and its effect on the operation of the reactor. We have carried out a specific neutron calculation scheme based on the modular 2D-transport code APOLLO2 using a two-level transport method. Experimental validation of the flux calculation scheme was carried out in the ISIS reactor, the mock-up of OSIRIS, by optimizing the loading of fuel elements in the core. The experimental results show that the neutron calculation scheme computes the fluxes in close agreement with the measurements especially the fast flux. This study allows us to master the essential physical parameters needed for the design of a flux converter in an MTR reactor. (author)

  19. Treatment of low-strength wastewater using immobilized biomass in a sequencing batch external loop reactor: influence of the medium superficial velocity on the stability and performance

    Directory of Open Access Journals (Sweden)

    Camargo E.F.M.

    2002-01-01

    Full Text Available An anaerobic sequencing batch bioreactor with external circulation of the liquid phase wherein the biomass was immobilized on a polyurethane foam matrix was analyzed, focussing on the influence of the liquid superficial velocity on the reactor's stability and efficiency. Eight-hour cycles were carried out at 30ºC treating glucose-based synthetic wastewater around 500 mgDQO/L. The performance of the reactor was assessed without circulation and with circulating liquid superficial velocity between 0.034 and 0.188 cm/s. The reactor attained operating stability and a high organic matter removal was achieved when liquid was circulated. A first order model was used to evaluate the influence of the liquid superficial velocity (vS, resulting in an increase in the apparent first order parameter when vS increased from 0.034 to 0.094 cm/s. The parameter value remained unchangeable when 0.188 cm/s was applied, indicating that beyond this value no improvement on liquid mass transfer was observed. Moreover, the necessary time to reach the final removal efficiency decreased when liquid circulation was applied, indicating that a 3-hour cycle could be enough.

  20. 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

  1. Elucidating reactivity regimes in cyclopentane oxidation: Jet stirred reactor experiments, computational chemistry, and kinetic modeling

    KAUST Repository

    Rachidi, Mariam El

    2016-06-23

    This study is concerned with the identification and quantification of species generated during the combustion of cyclopentane in a jet stirred reactor (JSR). Experiments were carried out for temperatures between 740 and 1250K, equivalence ratios from 0.5 to 3.0, and at an operating pressure of 10atm. The fuel concentration was kept at 0.1% and the residence time of the fuel/O/N mixture was maintained at 0.7s. The reactant, product, and intermediate species concentration profiles were measured using gas chromatography and Fourier transform infrared spectroscopy. The concentration profiles of cyclopentane indicate inhibition of reactivity between 850-1000K for ϕ = 2.0 and ϕ = 3.0. This behavior is interesting, as it has not been observed previously for other fuel molecules, cyclic or non-cyclic. A kinetic model including both low- and high-temperature reaction pathways was developed and used to simulate the JSR experiments. The pressure-dependent rate coefficients of all relevant reactions lying on the PES of cyclopentyl+O, as well as the C-C and C-H scission reactions of the cyclopentyl radical were calculated at the UCCSD(T)-F12b/cc-pVTZ-F12//M06-2X/6-311++G(d,p) level of theory. The simulations reproduced the unique reactivity trend of cyclopentane and the measured concentration profiles of intermediate and product species. Sensitivity and reaction path analyses indicate that this reactivity trend may be attributed to differences in the reactivity of allyl radical at different conditions, and it is highly sensitive to the C-C/C-H scission branching ratio of the cyclopentyl radical decomposition.

  2. 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.

  3. An Integrated Management System (IMS) for JM-1 SLOWPOKE-2 research reactor in Jamaica: experiences in documentation

    Energy Technology Data Exchange (ETDEWEB)

    Warner, T., E-mail: traceyann.warner02@uwimona.edu.jm [Univ. of West Indies, Mona (Jamaica)

    2014-07-01

    Since the first criticality in March 1984, the Jamaica SLOWPOKE-2 research reactor at the University of the West Indies, Mona located in the department of the International Centre for Environmental and Nuclear Sciences (ICENS) has operated for approximately 52% of the lifetime of the existing core configuration. The 20kW pool type research reactor has been primarily used for neutron activation analysis in environmental, agricultural, geochemical, health-related studies and mineral exploration in Jamaica. The involvement of the JM-1 reactor for research and teaching activities has segued into commercial applications which, coupled with the current core conversion programme from HEU to LEU, has demanded the implementation of management systems to satisfy regulatory requirements and assure compliance with internationally defined quality standards. At ICENS, documentation related to the Quality Management System aspect of an Integrated Management System (IMS) is well underway. The quality system will incorporate operational and nuclear safety, training, maintenance, design, utilization, occupational health and safety, quality service, and environmental management for its Nuclear Analytical Laboratory, NAL. The IMS is being designed to meet the requirements of the IAEA GS-R-3 with additional controls from international standards including: ISO/IEC 17025:2005, ISO 9001:2008, ISO 14001:2004 and OHSAS 18001:2007. This paper reports on the experiences of the documentation process in a low power reactor facility characterized by limited human resource, where innovative mechanisms of system automation and modeling are included to increase productivity and efficiency. (author)

  4. Gap Size Uncertainty Quantification in Advanced Gas Reactor TRISO Fuel Irradiation Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Pham, Binh T.; Einerson, Jeffrey J.; Hawkes, Grant L.; Lybeck, Nancy J.; Petti, David A.

    2016-10-01

    The Advanced Gas Reactor (AGR)-3/4 experiment is the combination of the third and fourth tests conducted within the tristructural isotropic fuel development and qualification research program. The AGR-3/4 test consists of twelve independent capsules containing a fuel stack in the center surrounded by three graphite cylinders and shrouded by a stainless steel shell. This capsule design enables temperature control of both the fuel and the graphite rings by varying the neon/helium gas mixture flowing through the four resulting gaps. Knowledge of fuel and graphite temperatures is crucial for establishing the functional relationship between fission product release and irradiation thermal conditions. These temperatures are predicted for each capsule using the commercial finite-element heat transfer code ABAQUS. Uncertainty quantification reveals that the gap size uncertainties are among the dominant factors contributing to predicted temperature uncertainty due to high input sensitivity and uncertainty. Gap size uncertainty originates from the fact that all gap sizes vary with time due to dimensional changes of the fuel compacts and three graphite rings caused by extended exposure to high temperatures and fast neutron irradiation. Gap sizes are estimated using as-fabricated dimensional measurements at the start of irradiation and post irradiation examination dimensional measurements at the end of irradiation. Uncertainties in these measurements provide a basis for quantifying gap size uncertainty. However, lack of gap size measurements during irradiation and lack of knowledge about the dimension change rates lead to gap size modeling assumptions, which could increase gap size uncertainty. In addition, the dimensional measurements are performed at room temperature, and must be corrected to account for thermal expansion of the materials at high irradiation temperatures. Uncertainty in the thermal expansion coefficients for the graphite materials used in the AGR-3/4 capsules

  5. 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)

  6. 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

  7. Development of mechanical design technology for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Keun Bae; Choi, Suhn; Kim, Kang Soo; Kim, Tae Wan; Jeong, Kyeong Hoon; Lee, Gyu Mahn

    1999-03-01

    While Korean nuclear reactor strategy seems to remain focused on the large capacity power generation, it is expected that demand of small and medium size reactor will arise for multi-purpose application such as small capacity power generation, co-generation and sea water desalination. With this in mind, an integral reactor SMART is under development. Design concepts, system layout and types of equipment of integral reactor are significantly different from those of loop type reactor. Conceptual design development of mechanical structures of integral reactor SMART is completed through the first stage of the project. Efforts were endeavored for the establishment of design basis and evaluation of applicable codes and standards. Design and functional requirements of major structural components were setup, and three dimensional structural modelling of SMART reactor vessel assembly was prepared. Also, maintenance and repair scheme as well as preliminary fabricability evaluation were carried out. Since small integral reactor technology includes sensitive technologies and know-how's, it is hard to achieve systematic and comprehensive technology transfer from nuclear-advanced countries. Thus, it is necessary to develop the related design technology and to verify the adopted methodologies through test and experiments in order to assure the structural integrity of reactor system. (author)

  8. University of Illinois nuclear pumped laser program. [experiments with a TRIGA pulsed reactor with a broad pulse and a low peak flux

    Science.gov (United States)

    Miley, G. H.

    1979-01-01

    The development of nuclear pumped lasers with improved efficiency, energy storage capability, and UF6 volume pumping is reviewed. Results of nuclear pumped laser experiments using a TRIGA-type pulsed reactor are outlined.

  9. Blowing loop in the EL-4 reactor: CO{sub 2} flow control analogue study; Boucle de soufflage de la centrale EL-4 - regulation du debit CO{sub 2} - etude analogique

    Energy Technology Data Exchange (ETDEWEB)

    Chazal, G.; Merle, J.P.; Guillemard, B. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Leroy, C.; Robin, L.; Jacquin, J.C.; Cornudet, A. [Societe INDATOM, France (France)

    1966-07-01

    This report describes one study which contributed to the construction of the Monts d'Arree nuclear power station: EL-4. The reactor is cooled by a CO{sub 2} current provided by 3 turbo-blower groups. The priming vapour for the turbines is taken at the exit of the main CO{sub 2} - H{sub 2}O exchangers. The operation of EL 4 is based on a high degree of centralization of the controls which attributes an important role to the general regulation circuits. This general regulation includes in particular an internal blowing loop which controls the CO{sub 2} flow. The study of the control of this CO{sub 2} flow is made up of 3 parts: - analogue representation of the reactors cooling circuit and of the turbo blower unit. - first test campaign using the analogue computer describing the natural behaviour of the system in the absence of control. theoretical determination of the regulation factors; definition of the regulation using an analogue computer and second test campaign for recording the performances of the blowing loop. The 4. part of the report deals with the analogue study: analogue equations - development. (authors) [French] Ce rapport prend place parmi les etudes de realisation de la Centrale des Monts d'Arree EL-4. Le reacteur est refroidi par une circulation de CO{sub 2} assuree par 3 groupes turbosoufflantes. La vapeur d'entrainement des turbines est prelevee a la sortie des echangeurs principaux CO{sub 2} - H{sub 2}O. L'exploitation de EL-4 repose sur une centralisation poussee des moyens de controle-commande qui attribue un role essentiel aux circuits de regulation generale. Cette regulation generale comporte en particulier une boucle interne de soufflage qui realise un asservissement du debit de CO{sub 2}. L'etude de cette regulation du debit CO{sub 2} comprend 3 parties: - representation analogique du circuit de refroidissement du reacteur et de l'ensemble turbine-soufflante. - premiere campagne d'essais sur calculateur

  10. Experiments and modelling VOCs' removal in a DBD reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sarron, V.; Aubry, O.; Khacef, A.; Cormier, J.M. [Orleans Univ., Orleans Cedex (France). Polytech d' Orleans, Group for Research and Studies on Mediators of Inflamation

    2010-07-01

    Non-thermal plasma discharges are being considered as a means to convert volatile organic compounds (VOCs) diluted in air at atmospheric pressure. This study showed that the treatment of propane or ethane in a dielectric barrier discharge (DBD) reactor at a temperature of 800 K can be modeled from a chemical mechanism. The DBD reactor was simulated using consecutive elementary plug flow reactors (PFR). Streamer effects leading to active species production such as O-atoms in dry air from electronic dissociation of oxygen (O{sub 2}) were simulated by injection of O-atoms at the inlet of each elementary PFR. A good agreement was obtained for all the studied inlet mixtures, in which ethane concentrations and propane were varied in air. The concentration of O-atoms were found to play a role on carbon monoxide (CO) and carbon dioxide (CO{sub 2}) concentrations at a given energy density. An increase of O promoted CO{sub 2} concentration. In addition, the models made it possible to determine the concentrations levels of non measured by-products. The O-atom concentration was the main parameter of the developed model to simulate a DBD reactor. It was concluded that the obtained models can be efficient tools for predicting light hydrocarbons conversion in a non-thermal plasma. 7 refs., 10 figs.

  11. Measurement of neutrinos released in nuclear reactors through the Borexino experiment; Mesure des neutrinos de reacteurs nucleaires dans l'experience Borexino

    Energy Technology Data Exchange (ETDEWEB)

    Dadoun, O

    2003-06-01

    The main goal of the Borexino experiment is to measure in real time the solar neutrino flux from the beryllium (Be{sup 7}) line at 862 keV. Beyond this pioneer low energy neutrino detection, Borexino will be able to measure solar neutrinos above the MeV, (B{sup 8} neutrinos and pep neutrinos), nuclear reactor neutrinos (with an average energy of 3 MeV) and the supernova neutrinos (their spectrum goes up to some ten MeV). In this work I mainly focus on the study of the nuclear reactors neutrinos. This field has recently been enriched by the results of the KamLAND experiment, which have greatly improved the determination of the neutrino oscillation parameters. In order to measure these events which are above the MeV, the Borexino collaboration entrusted the PCC group at College de France, with the tasks of developing a fast digit system running at 400 MHz: the FADC cards. The PCC group designed the FADC cards and completed them at the beginning of 2002. The first cards which were introduced in the main electronic acquisition unit allowed us to control their functioning and that of the acquisition software. FADC cards were also installed in the Borexino prototype, CTF. The data are analysed in order to determine a limit to the expected background noise of Borexino in measuring the nuclear reactor neutrinos. (author)

  12. Application of Porous Nickel-Coated TiO2 for the Photocatalytic Degradation of Aqueous Quinoline in an Internal Airlift Loop Reactor

    Directory of Open Access Journals (Sweden)

    Mingxin Huo

    2012-02-01

    Full Text Available P25 film, prepared by a facile dip-coating method without any binder, was further developed in a recirculating reactor for quinoline removal from synthetic wastewater. Macroporous foam Ni, which has an open three-dimensional network structure, was utilized as a substrate to make good use of UV rays. Field emission scanning electron microscopy and X-ray diffraction analysis showed that the coated/calcinated P25 films consisted of two crystal phases, and had a number of uniform microcracks on the surface. The effects of initial quinoline concentration, light intensity, reaction temperature, aeration, and initial pH were studied. Increased reaction time, light intensity, environmental temperature, and gas aeration were found to significantly improve the quinoline removal efficiency. The aeration effect of oxygen dependency on the quinoline degradation had the trend pure oxygen > air > no gas > pure nitrogen with free O2. The solution pH crucially affected quinoline photodegradation; the high electrostatic adsorption of quinoline molecules on the TiO2 surface was strongly pH dependent. 2-Pyridine-carboxaldehyde, 3-pyridinecarboxaldehyde, and 2(1H-quinolinone were identified as the major intermediates of quinoline degradation. Based on these intermediates, a primary degradation mechanism was proposed. This reusable P25 film benefits the photodegradation of water contaminants and has potential in other various applications.

  13. INTEGRAL BENCHMARKS AVAILABLE THROUGH THE INTERNATIONAL REACTOR PHYSICS EXPERIMENT EVALUATION PROJECT AND THE INTERNATIONAL CRITICALITY SAFETY BENCHMARK EVALUATION PROJECT

    Energy Technology Data Exchange (ETDEWEB)

    J. Blair Briggs; Lori Scott; Enrico Sartori; Yolanda Rugama

    2008-09-01

    Interest in high-quality integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties accelerate to meet the demands of next generation reactor and advanced fuel cycle concepts. The International Reactor Physics Experiment Evaluation Project (IRPhEP) and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) continue to expand their efforts and broaden their scope to identify, evaluate, and provide integral benchmark data for method and data validation. Benchmark model specifications provided by these two projects are used heavily by the international reactor physics, nuclear data, and criticality safety communities. Thus far, 14 countries have contributed to the IRPhEP, and 20 have contributed to the ICSBEP. The status of the IRPhEP and ICSBEP is discussed in this paper, and the future of the two projects is outlined and discussed. Selected benchmarks that have been added to the IRPhEP and ICSBEP handbooks since PHYSOR’06 are highlighted, and the future of the two projects is discussed.

  14. 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...

  15. Applicable regulations and development of surveillance experiments of criticality approach in the TRIGA III Mark reactor; Normativa aplicable y desarrollo de experimentos de vigilancia de aproximacion a criticidad en el reactor Triga Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez M, J.L.; Aguilar H, F.; Rivero G, T.; Sainz M, E. [Instituto nacional de Investigaciones Nucleares, Departamento de Automatizacion, A.P. 18-1027, Col. Escandon, 11801 Mexico D.F. (Mexico)

    2000-07-01

    In the procedure elaborated to repair the vessel of TRIGA III Mark reactor is required to move toward two tanks of temporal storage the fuel elements which are in operation and the spent fuel elements which are in decay inside the reactor pool. The National Commission of Nuclear Safety and Safeguards (CNSNS) has requested as protection measure that it is carried out a surveillance of the criticality approach of the temporal storages. This work determines the main regulation aspects that entails an experiment of criticality approach, moreover, informing about the results obtained in the developing of this experiments. The regulation aspects are not exclusives for this work in the TRIGA Mark III reactor but they also apply toward any assembling of fissile material. (Author)

  16. Two-phase flow experiments in a model of the hot leg of a pressurised water reactor. Technical report

    Energy Technology Data Exchange (ETDEWEB)

    Seidel, Tobias; Vallee, Christophe; Lucas, Dirk; Beyer, Matthias; Deendarlianto

    2011-09-15

    In order to investigate the two-phase flow behaviour in a complex reactor-typical geometry and to supply suitable data for CFD code validation, a model of the hot leg of a pressurised water reactor was built at FZD. The hot leg model is operated in the pressure chamber of the TOPFLOW test facility, which is used to perform high-pressure experiments under pressure equilibrium with the inside atmosphere of the chamber. This technique makes it possible to visualise the two-phase flow through large windows, also at reactor-typical pressure levels. In order to optimise the optical observation possibilities, the test section was designed with a rectangular cross-section. Experiments were performed with air and water at 1.5 and 3.0 bar at room temperature as well as with steam and water at 15, 30 and 50 bar and the corresponding saturation temperature (i.e. up to 264 C). The total of 194 runs are divided into 4 types of experiments covering stationary co-current flow, counter-current flow, flow without water circulation and transient counter-current flow limitation (CCFL) experiments. This report provides a detailed documentation of the experiments including information on the experimental setup, experimental procedure, test matrix and on the calibration of the measuring devices. The available data is described and data sheets were arranged for each experiment in order to give an overview of the most important parameters. For the cocurrent flow experiments, water level histograms were arranged and used to characterise the flow in the hot leg. In fact, the form of the probability distribution was found to be sensitive to the boundary conditions and, therefore, is useful for the CFD comparison. Furthermore, the flooding characteristics of the hot leg model plotted in terms of the classical Wallis parameter or Kutateladze number were found to fail to properly correlate the data of the air/water and steam/water series. Therefore, a modified Wallis parameter is proposed, which

  17. High Temperature Fluoride Salt Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    Aaron, Adam M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Cunningham, Richard Burns [Univ. of Tennessee, Knoxville, TN (United States); Fugate, David L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Holcomb, David Eugene [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Kisner, Roger A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Peretz, Fred J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wilson, Dane F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yoder, Jr, Graydon L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-12-01

    with 3 cm diameter graphite-based fuel pebbles slowly circulating up through the core. Molten salt coolant (FLiBe) at 700°C flows concurrently (at significantly higher velocity) with the pebbles and is used to remove heat generated in the reactor core (approximately 1280 W/pebble), and supply it to a power conversion system. Refueling equipment continuously sorts spent fuel pebbles and replaces spent or damaged pebbles with fresh fuel. By combining greater or fewer numbers of pebble channel assemblies, multiple reactor designs with varying power levels can be offered. The PB-AHTR design is discussed in detail in Reference [1] and is shown schematically in Fig. 1. Fig. 1. PB-AHTR concept (drawing taken from Peterson et al., Design and Development of the Modular PB-AHTR Proceedings of ICApp 08). Pebble behavior within the core is a key issue in proving the viability of this concept. This includes understanding the behavior of the pebbles thermally, hydraulically, and mechanically (quantifying pebble wear characteristics, flow channel wear, etc). The experiment being developed is an initial step in characterizing the pebble behavior under realistic PB-AHTR operating conditions. It focuses on thermal and hydraulic behavior of a static pebble bed using a convective salt loop to provide prototypic fluid conditions to the bed, and a unique inductive heating technique to provide prototypic heating in the pebbles. The facility design is sufficiently versatile to allow a variety of other experimentation to be performed in the future. The facility can accommodate testing of scaled reactor components or sub-components such as flow diodes, salt-to-salt heat exchangers, and improved pump designs as well as testing of refueling equipment, high temperature instrumentation, and other reactor core designs.

  18. Asymptotic Behavior of a Viscoelastic Fluid in a Closed Loop Thermosyphon: Physical Derivation, Asymptotic Analysis, and Numerical Experiments

    Directory of Open Access Journals (Sweden)

    Justine Yasappan

    2013-01-01

    Full Text Available Fluids subject to thermal gradients produce complex behaviors that arise from the competition with gravitational effects. Although such sort of systems have been widely studied in the literature for simple (Newtonian fluids, the behavior of viscoelastic fluids has not been explored thus far. We present a theoretical study of the dynamics of a Maxwell viscoelastic fluid in a closed-loop thermosyphon. This sort of fluid presents elastic-like behavior and memory effects. We study the asymptotic properties of the fluid inside the thermosyphon and the exact equations of motion in the inertial manifold that characterizes the asymptotic behavior. We derive, for the first time, the mathematical derivations of the motion of a viscoelastic fluid in the interior of a closed-loop thermosyphon under the effects of natural convection and a given external temperature gradient.

  19. Experiments prior to construction of the Rapsodie reactor (1962); Experiences preliminaires a la construction de la pile rapsodie (1962)

    Energy Technology Data Exchange (ETDEWEB)

    Vautrey, L.; Zaleski, C.P. [Commissariat a l' Energie Atomique, Cadarache (France). Centre d' Etudes Nucleaires

    1962-07-01

    Before proceeding to the construction of the various reactor components described in the paper 'Fast Breeder Reactor Rapsodie', many experimental studies of a hydraulic, thermal and mechanical character have been carried out, or are under consideration, to test the validity of the principles adopted in the Preliminary Project. This paper deals with the most important of these: 1. Studies of coolant circuit components: sodium pumps (mechanical or electromagnetic), Na-NaK and NaK ir heat exchangers, measuring instruments (flow rates, temperatures), sodium purification circuits, etc. 2. Studies in cooling of fuel and fertile assemblies: a) study of the sodium cooling carried out by means of hydraulic mockups (scale of 1: 1 or over) reproducing the flow of the coolant fluid in the piping, upstream from and inside the fuel and fertile elements. b) study of the cooling by gas and by immersion in lead, employed during handling and storage operations. 3. Studies of special reactor devices: fusible rotating linkage, parts of the control rod mechanisms. 4. Study of the reactor block and coolant circuits as a whole. This study is to begin at the end of the year. The mock-up, now nearing completion, reproduces on a scale of 1: 1 the installation provided in the Preliminary Project and includes: the reactor block, to which is connected a high flow ate sodium circuit, permitting of long-term tests and thermal shocks, and also, a control rod testing circuit; complete installation of the 1 MW and 10 MW coolant circuits, the performances of which it will be possible to check under various operational conditions. 5. A safety study carried out on a 3: 10 scale mock p comprising the whole of the reactor block and shielding, with the object of limiting the effects of any accidental liberation of energy of an explosive character. (authors) [French] Avant d'entreprendre la realisation des divers elements du reacteur decrit dans le rapport 'Reacteur rapide

  20. TREAT Neutronics Analysis of Water-Loop Concept Accommodating LWR 9-rod Bundle

    Energy Technology Data Exchange (ETDEWEB)

    Hill, Connie M.; Woolstenhulme, Nicolas E.; Parry, James R.; Bess, John D.; Housley, Gregory K.

    2016-09-01

    Abstract. Simulation of a variety of transient conditions has been successfully achieved in the Transient Reactor Test (TREAT) facility during operation between 1959 and 1994 to support characterization and safety analysis of nuclear fuels and materials. A majority of previously conducted tests were focused on supporting sodium-cooled fast reactor (SFR) designs. Experiments evolved in complexity. Simulation of thermal-hydraulic conditions expected to be encountered by fuels and materials in a reactor environment was realized in the development of TREAT sodium loop experiment vehicles. These loops accommodated up to 7-pin fuel bundles and served to simulate more closely the reactor environment while safely delivering large quantities of energy into the test specimen. Some of the immediate TREAT restart operations will be focused on testing light water reactor (LWR) accident tolerant fuels (ATF). Similar to the sodium loop objectives, a water loop concept, developed and analyzed in the 1990’s, aimed at achieving thermal-hydraulic conditions encountered in commercial power reactors. The historic water loop concept has been analyzed in the context of a reactivity insertion accident (RIA) simulation for high burnup LWR 2-pin and 3-pin fuel bundles. Findings showed sufficient energy could be deposited into the specimens for evaluation. Similar results of experimental feasibility for the water loop concept (past and present) have recently been obtained using MCNP6.1 with ENDF/B-VII.1 nuclear data libraries. The old water loop concept required only two central TREAT core grid spaces. Preparation for future experiments has resulted in a modified water loop conceptual design designated the TREAT water environment recirculating loop (TWERL). The current TWERL design requires nine TREAT core grid spaces in order to place the water recirculating pump under the TREAT core. Due to the effectiveness of water moderation, neutronics analysis shows that removal of seven additional

  1. Controller design and experiment for autothermal reforming of methanol in miniature reactor.

    Science.gov (United States)

    Lu, Jiangang; Zhuang, Hong; Yang, Qinmin; Wang, Xuefei; Zheng, Jianfeng; Chen, Jinshui; Sun, Youxian

    2014-09-01

    In this paper, a miniature methanol fuel processor and its controller design is introduced for onboard hydrogen production. The hydrogen is generated via autothermal reforming of methanol. The control scheme consists of a hydrogen flow rate controller and a reforming temperature controller. To deal with uncertain system dynamics and external disturbance, an adaptive sliding mode control algorithm is adopted as the hydrogen flow rate controller for regulating hydrogen flow rate by manipulating methanol flow rate. Additionally, a high-gain observer is implemented to estimate the unmeasurable system state. The stability of closed-loop system is guaranteed by standard Lyapunov analysis. Furthermore, a variable ratio control law is employed as the reforming temperature controller to achieve steady reforming temperature by adjusting the reforming air flow rate. Finally, the effectiveness of the entire system is testified by experimental means. Copyright © 2013 ISA. Published by Elsevier Ltd. All rights reserved.

  2. 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

  3. Loop-to-loop coupling.

    Energy Technology Data Exchange (ETDEWEB)

    Warne, Larry Kevin; Lucero, Larry Martin; Langston, William L.; Salazar, Robert Austin; Coleman, Phillip Dale; Basilio, Lorena I.; Bacon, Larry Donald

    2012-05-01

    This report estimates inductively-coupled energy to a low-impedance load in a loop-to-loop arrangement. Both analytical models and full-wave numerical simulations are used and the resulting fields, coupled powers and energies are compared. The energies are simply estimated from the coupled powers through approximations to the energy theorem. The transmitter loop is taken to be either a circular geometry or a rectangular-loop (stripline-type) geometry that was used in an experimental setup. Simple magnetic field models are constructed and used to estimate the mutual inductance to the receiving loop, which is taken to be circular with one or several turns. Circuit elements are estimated and used to determine the coupled current and power (an equivalent antenna picture is also given). These results are compared to an electromagnetic simulation of the transmitter geometry. Simple approximate relations are also given to estimate coupled energy from the power. The effect of additional loads in the form of attached leads, forming transmission lines, are considered. The results are summarized in a set of susceptibility-type curves. Finally, we also consider drives to the cables themselves and the resulting common-to-differential mode currents in the load.

  4. Program management plan for the Molten Salt Reactor Experiment Remediation Project at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-09-01

    The primary mission of the Molten Salt Reactor Experiment (MSRE) Remediation Project is to effectively implement the risk-reduction strategies and technical plans to stabilize and prevent further migration of uranium within the MSRE facility, remove the uranium and fuel salts from the system, and dispose of the fuel and flush salts by storage in appropriate depositories to bring the facility to a surveillance and maintenance condition before decontamination and decommissioning. This Project Management Plan (PMP) for the MSRE Remediation Project details project purpose; technical objectives, milestones, and cost objectives; work plan; work breakdown structure (WBS); schedule; management organization and responsibilities; project management performance measurement planning, and control; conduct of operations; configuration management; environmental, safety, and health compliance; quality assurance; operational readiness reviews; and training.

  5. Experience with simulators for development and evaluation of operator support systems at the OECD Halden reactor project

    Energy Technology Data Exchange (ETDEWEB)

    Berg, O.; Holmstroem, C.O.B.; Volden, F. [Inst. for Energiteknikk, Halden (Norway)

    1994-12-31

    The OECD Halden Reactor Project carries out research and development of computer-based systems for nuclear power plants. The aim is to design, build and evaluate computer-based systems which can assist and support operators in their various cognitive tasks and through this improve the total performance and safety of complex plant operations. The operator support systems are tested and evaluated through experiments in the Halden Man-Machine Laboratory using the NORS pressurized water reactor simulator. An experiment to assess the impact on operator behaviour when using a rule-based expert system for fault diagnosis will be described. Two different computerised operator support systems utilising on-line simulation models are described. The first system is the core surveillance system SCORPIO which has been in operation at the Ringhals plant in Sweden since the end of 1987. This system performs core monitoring functions by logging and presenting measured data together with results from three-dimensional simulations of the core. In predictive mode the development during the coming two days may be simulated, and a strategy generator is available to facilitate transient planning. The second system is an early fault detection system for the feedwater system installed in the Loviisa plant in Finland. The method used is to run small, decoupled mathematical models which calculate the state of the process assuming no faults. The behaviour of these models is then compared with the behaviour of the real process, and if there is a deviation, an alarm is triggered. For both systems special emphasize has been put on making a user-friendly operator interface where simulator data are combined with measurements. (orig.) (8 refs., 10 figs.).

  6. QUARTERLY PROGRESS REPORT JANUARY, FEBRUARY, MARCH, 1968 REACTOR FUELS AND MATERIALS DEVELOPMENT PROGRAMS FOR FUELS AND MATERIALS BRANCH OF USAEC DIVISION OF REACTOR DEVELOPMENT AND TECHNOLOGY

    Energy Technology Data Exchange (ETDEWEB)

    Cadwell, J. J.; de Halas, D. R.; Nightingale, R. E.; Worlton, D. C.

    1968-06-01

    Progress is reported in these areas: nuclear graphite; fuel development for gas-cooled reactors; HTGR graphite studies; nuclear ceramics; fast-reactor nitrides research; non-destructive testing; metallic fuels; basic swelling studies; ATR gas and water loop operation and maintenance; reactor fuels and materials; fast reactor dosimetry and damage analysis; and irradiation damage to reactor metals.

  7. Model-Based Closed-Loop Glucose Control in Type 1 Diabetes: The DiaCon Experience

    DEFF Research Database (Denmark)

    Schmidt, Signe; Boiroux, Dimitri; Duun-Henriksen, Anne Katrine

    2013-01-01

    (hemoglobin A1c 7.2% ± 0.4%). In study 1, hypoglycemic events (plasma glucose glucose from 22:00–07:00 was 90 mg/dl [74–146 mg/dl; median (interquartile range)] during OL and 108 mg/dl (101–128 mg/dl) during CL (determined by continuous glucose......Background: To improve type 1 diabetes mellitus (T1DM) management, we developed a model predictive control (MPC) algorithm for closed-loop (CL) glucose control based on a linear second-order deterministic-stochastic model. The deterministic part of the model is specified by three patient...... crossover studies. Study 1 compared CL with open-loop (OL) control. Study 2 compared glucose control after CL initiation in the euglycemic (CL-Eu) and hyperglycemic (CL-Hyper) ranges, respectively. Patients were studied from 22:00–07:00 on two separate nights. Results: Each study included six T1DM patients...

  8. Model-based closed-loop glucose control in type 1 diabetes: the DiaCon experience.

    Science.gov (United States)

    Schmidt, Signe; Boiroux, Dimitri; Duun-Henriksen, Anne Katrine; Frøssing, Laurits; Skyggebjerg, Ole; Jørgensen, John Bagterp; Poulsen, Niels Kjølstad; Madsen, Henrik; Madsbad, Sten; Nørgaard, Kirsten

    2013-09-01

    To improve type 1 diabetes mellitus (T1DM) management, we developed a model predictive control (MPC) algorithm for closed-loop (CL) glucose control based on a linear second-order deterministic-stochastic model. The deterministic part of the model is specified by three patient-specific parameters: insulin sensitivity factor, insulin action time, and basal insulin infusion rate. The stochastic part is identical for all patients but identified from data from a single patient. Results of the first clinical feasibility test of the algorithm are presented. We conducted two randomized crossover studies. Study 1 compared CL with open-loop (OL) control. Study 2 compared glucose control after CL initiation in the euglycemic (CL-Eu) and hyperglycemic (CL-Hyper) ranges, respectively. Patients were studied from 22:00-07:00 on two separate nights. Each study included six T1DM patients (hemoglobin A1c 7.2% ± 0.4%). In study 1, hypoglycemic events (plasma glucose Hyper, respectively. This study suggests that our novel MPC algorithm can safely and effectively control glucose overnight, also when CL control is initiated during hyperglycemia. © 2013 Diabetes Technology Society.

  9. Scintillation light production, propagation and detection in the Stereo reactor antineutrino experiment

    Science.gov (United States)

    Buck, Christian; Lindner, Manfred; Roca, Christian

    2017-09-01

    The Stereo experiment’s detector has been optimized to observe reactor antineutrinos via inverse beta decay within a 1800 liter volume filled with Gadolinium-doped organic liquid scintillator (LS). The main requirements for the scintillator in Stereo are compatibility with detector materials as the acrylic vessels, transparency, light yield, pulse shape discrimination capabilities as well as chemical and optical stability over several years of data taking. With these conditions in mind, the composition of the LS is mainly a mix of 75% LAB, 20%PXE and 5% DIN combined with the two wavelength-shifters PPO and Bis-MSB. The final admixture after the full scale production lead to an attenuation length of more than 5 meters for optical photons of 430 nm. The scintillation light produced in the Gd-loaded target volume and the Gd-free outer crown is detected by 48 eight inch PMTs on top of the detector. A correct performance of the PMTs has been ensured through several tests. Common characteristics for PMTs as gain, single photoelectron peak, time behaviour, dark rate and afterpulse ratio were measured resulting in a complete agreement with the manufacturer values.

  10. Coolant mixing in pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hoehne, T.; Grunwald, G.

    1998-10-01

    The behavior of PWRs during cold water or boron dilution transients is strongly influenced by the distribution of coolant temperature and boron concentration at the core inlet. This distribution is the needed input to 3-dimensional neutron kinetics to calculate the power distribution in the core. It mainly depends on how the plugs of cold or unborated water formed in a single loop are mixed in the downcomer and in the lower plenum. To simulate such mixture phenomena requires the application of 3-dimensional CFD (computational fluid dynamics) codes. The results of the simulation have to be validated against mixture experiments at scaled facilities. Therefore, in the framework of a research project funded by BMBF, the institute creates a 1:5 mixture facility representing first the geometry of a German pressurized water reactor and later the European Pressurized Water Reactor (EPR) geometry. The calculations are based on the CFD Code CFX-4. (orig.)

  11. A Methodology for Loading the Advanced Test Reactor Driver Core for Experiment Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Cowherd, Wilson M.; Nielsen, Joseph W.; Choe, Dong O.

    2016-11-01

    In support of experiments in the ATR, a new methodology was devised for loading the ATR Driver Core. This methodology will replace the existing methodology used by the INL Neutronic Analysis group to analyze experiments. Studied in this paper was the as-run analysis for ATR Cycle 152B, specifically comparing measured lobe powers and eigenvalue calculations.

  12. Experience Gained during the Adaptation of Classical ChE Subjects to the Bologna Plan in Europe: The Case of Chemical Reactors

    Science.gov (United States)

    Ponsa, Sergio; Sanchez, Antoni

    2011-01-01

    At present, due to the overall adaptation to the European Higher Education Area (EHEA), a new concept regarding the teaching methodology was thought to be essential for engineering subjects. In this paper we describe our experience teaching the altered content of the courses on two classical subjects; Chemical Reactors (Chemical Engineering) and…

  13. Welding procedures used in the fabrication of fuel elements for the DON Reactor exponential experiment; La soldadura en la fabricacion de elementos combustibles destinados a una experiencia exponencial

    Energy Technology Data Exchange (ETDEWEB)

    Diaz Beltran, A.; Jaraiz Franco, E.; Rivas Diaz, M. de las

    1965-07-01

    This exponential experiment required 74 units (37 loaded with UO{sub 2} and 37 with UC) to simulate the Reactor fuel channels. Each unit was enclosed in a tube similar to the calandria ones. It contained the pressure tube, the shroud and the 19 rods cluster. Within the pressure tube, in touch with the elements, was the organic liquid. (Author)

  14. 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.

  15. Multiple Irradiation Capsule Experiment (MICE)-3B Irradiation Test of Space Fuel Specimens in the Advanced Test Reactor (ATR) - Close Out Documentation for Naval Reactors (NR) Information

    Energy Technology Data Exchange (ETDEWEB)

    M. Chen; CM Regan; D. Noe

    2006-01-09

    Few data exist for UO{sub 2} or UN within the notional design space for the Prometheus-1 reactor (low fission rate, high temperature, long duration). As such, basic testing is required to validate predictions (and in some cases determine) performance aspects of these fuels. Therefore, the MICE-3B test of UO{sub 2} pellets was designed to provide data on gas release, unrestrained swelling, and restrained swelling at the upper range of fission rates expected for a space reactor. These data would be compared with model predictions and used to determine adequacy of a space reactor design basis relative to fission gas release and swelling of UO{sub 2} fuel and to assess potential pellet-clad interactions. A primary goal of an irradiation test for UN fuel was to assess performance issues currently associated with this fuel type such as gas release, swelling and transient performance. Information learned from this effort may have enabled use of UN fuel for future applications.

  16. Argonne Liquid-Metal Advanced Burner Reactor : components and in-vessel system thermal-hydraulic research and testing experience - pathway forward.

    Energy Technology Data Exchange (ETDEWEB)

    Kasza, K.; Grandy, C.; Chang, Y.; Khalil, H.; Nuclear Engineering Division

    2007-06-30

    This white paper provides an overview and status report of the thermal-hydraulic nuclear research and development, both experimental and computational, conducted predominantly at Argonne National Laboratory. Argonne from the early 1970s through the early 1990s was the Department of Energy's (DOE's) lead lab for thermal-hydraulic development of Liquid Metal Reactors (LMRs). During the 1970s and into the mid-1980s, Argonne conducted thermal-hydraulic studies and experiments on individual reactor components supporting the Experimental Breeder Reactor-II (EBR-II), Fast Flux Test Facility (FFTF), and the Clinch River Breeder Reactor (CRBR). From the mid-1980s and into the early 1990s, Argonne conducted studies on phenomena related to forced- and natural-convection thermal buoyancy in complete in-vessel models of the General Electric (GE) Prototype Reactor Inherently Safe Module (PRISM) and Rockwell International (RI) Sodium Advanced Fast Reactor (SAFR). These two reactor initiatives involved Argonne working closely with U.S. industry and DOE. This paper describes the very important impact of thermal hydraulics dominated by thermal buoyancy forces on reactor global operation and on the behavior/performance of individual components during postulated off-normal accident events with low flow. Utilizing Argonne's LMR expertise and design knowledge is vital to the further development of safe, reliable, and high-performance LMRs. Argonne believes there remains an important need for continued research and development on thermal-hydraulic design in support of DOE's and the international community's renewed thrust for developing and demonstrating the Global Nuclear Energy Partnership (GNEP) reactor(s) and the associated Argonne Liquid Metal-Advanced Burner Reactor (LM-ABR). This white paper highlights that further understanding is needed regarding reactor design under coolant low-flow events. These safety-related events are associated with the transition

  17. R and D on fuzzy control applications to the BR1 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ruan, D.; Li, X. [Nuclear Research Center, Mol (Belgium)

    1998-12-31

    Fuzzy control applications in nuclear reactor operations present a tremendous challenge. The main reason for this is the public awareness of the risks of nuclear industry and the very strict safety regulations in force for nuclear power plants. The very same regulations prevent a researcher from quickly introducing novel fuzzy-logic methods into this field. On the other hand, the application of fuzzy logic has, despite the ominous sound of the word `fuzzy` to nuclear engineers, a number of very desirable advantages over classical methods, e.g., its robustness and the capability to include human experience into the controller. In this paper we report an on-going R and D project for controlling the power level of the Belgian Nuclear Reactor 1 (BR1) at the Belgian Nuclear Research Centre (SCK.CEN). The project started in 1995 and aims to investigate the added value of fuzzy control for nuclear reactors. We first review some relevant literature on fuzzy logic control in nuclear reactors, then present the state-of-the-art of the BR1 project. After experimenting fuzzy logic control under off-line tests at the BR1 reactor, we now foresee a new development for a closed-loop fuzzy control as an on-line operation of the BR1 reactor. Finally, we present the new development for a closed-loop fuzzy logic control at BR1 with an understanding of the safety requirements for this real fuzzy logic control application in nuclear rectors. (author) 18 refs.

  18. As-Run Physics Analysis for the UCSB-1 Experiment in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nielsen, Joseph Wayne [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

    The University of California Santa Barbara (UCSB) -1 experiment was irradiated in the A-10 position of the ATR. The experiment was irradiated during cycles 145A, 145B, 146A, and 146B. Capsule 6A was removed from the test train following Cycle 145A and replaced with Capsule 6B. This report documents the as-run physics analysis in support of Post-Irradiation Examination (PIE) of the test. This report documents the as-run fluence and displacements per atom (DPA) for each capsule of the experiment based on as-run operating history of the ATR. Average as-run heating rates for each capsule are also presented in this report to support the thermal analysis.

  19. 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.

  20. Design configurations of the methanol synthesis loop

    OpenAIRE

    Bøhn, Kristian

    2011-01-01

    In recent years the chemical industry has undergone considerable changes due to increased environmental regulations and energy costs. This master thesis has evaluated three different design considerations of the methanol synthesis loop using Honeywell's general purpose process simulator UniSim Design (R380 Build 14027) combined with MathWorks programming language MATLAB. The three configurations are Lurgis methanol reactor loop as built on Tjeldbergodden, the use of interstage methanol remova...

  1. 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...

  2. Health and safety plan for the Molten Salt Reactor Experiment remediation project at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    Energy Technology Data Exchange (ETDEWEB)

    Burman, S.N.; Uziel, M.S.

    1995-12-01

    The Lockheed Martin Energy Systems, Inc., (Energy Systems) policy is to provide a safe and healthful workplace for all employees and subcontractors. The accomplishment of the policy requires that operations at the Molten Salt Reactor Experiment (MSRE) facility at the Department of Energy (DOE) Oak Ridge National Laboratory (ORNL) are guided by an overall plan and consistent proactive approach to safety and health (S and H) issues. The policy and procedures in this plan apply to all MSRE operations. The provisions of this plan are to be carried out whenever activities are initiated at the MSRE that could be a threat to human health or the environment. This plan implements a policy and establishes criteria for the development of procedures for day-to-day operations to prevent or minimize any adverse impact to the environment and personnel safety and health and to meet standards that define acceptable management of hazardous and radioactive materials and wastes. The plan is written to utilize past experience and the best management practices to minimize hazards to human health or the environment from events such as fires, explosions, falls, mechanical hazards, or any unplanned release of hazardous or radioactive materials to the air.

  3. Review and future outlook of Dragon project/signatory organisations' collaborations in the integral reactor physics experiments - December 1973

    Energy Technology Data Exchange (ETDEWEB)

    Gutmann, H.

    1974-10-15

    The paper provides an overview of the collaborative reactor physics experiments conducted in the DRAGON Countries as of December 1973 summarizing those that have been conducted for high enriched uranium/thorium systems, those being conducted for low enriched uranium and plutonium systems, those conducted with irradiated fuel, and those on-going with integral fuel blocks. A list of relevant reports and papers on experiments is provided.

  4. Overview of the recovery and processing of {sup 233}U from the Oak Ridge molten salt reactor experiment (MSRE) remediation activities

    Energy Technology Data Exchange (ETDEWEB)

    Del Cul, G.D.; Icenhour, A.S.; Simmons, D.W.; Trowbridge, L.D.; Williams, D.F.; Toth, L.M.; Dai, S. [Oak Ridge National Lab., TN (United States)

    2001-07-01

    The Molten Salt Reactor Experiment (MSRE) was operated at Oak Ridge National Laboratory (ORNL) from 1965 to 1969 to test the concept of a high-temperature, homogeneous, fluid-fueled reactor. The discovery that UF{sub 6} and F{sub 2} migrated from the storage tanks into distant pipes and a charcoal bed resulted in significant activities to remove and recover the {sup 233}U and to decommission the reactor. The recovered fissile uranium will be converted into uranium oxide (U{sub 3}O{sub 8} ), which is a suitable form for long-term storage. This publication reports the research and several new developments that were needed to carry out these unique activities. (author)

  5. Protein distributions from a stochastic model of the lac operon of E. coli with DNA looping: analytical solution and comparison with experiments.

    Directory of Open Access Journals (Sweden)

    Krishna Choudhary

    Full Text Available Although noisy gene expression is widely accepted, its mechanisms are subjects of debate, stimulated largely by single-molecule experiments. This work is concerned with one such study, in which Choi et al., 2008, obtained real-time data and distributions of Lac permease in E. coli. They observed small and large protein bursts in strains with and without auxiliary operators. They also estimated the size and frequency of these bursts, but these were based on a stochastic model of a constitutive promoter. Here, we formulate and solve a stochastic model accounting for the existence of auxiliary operators and DNA loops. We find that DNA loop formation is so fast that small bursts are averaged out, making it impossible to extract their size and frequency from the data. In contrast, we can extract not only the size and frequency of the large bursts, but also the fraction of proteins derived from them. Finally, the proteins follow not the negative binomial distribution, but a mixture of two distributions, which reflect the existence of proteins derived from small and large bursts.

  6. High Fluency Low Flux Embrittlement Models of LWR Reactor Pressure Vessel Embrittlement and a Supporting Database from the UCSB ATR-2 Irradiation Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Odette, G. Robert [Univ. of California, Santa Barbara, CA (United States)

    2017-01-24

    Reactor pressure vessel embrittlement may limit the lifetime of light water reactors (LWR). Embrittlement is primarily caused by formation of nano-scale precipitates, which cause hardening and a subsequent increase in the ductile-to-brittle transition temperature of the steel. While the effect of Cu has historically been the largest research focus of RPV embrittlement, there is increasing evidence that Mn, Ni and Si are likely to have a large effect at higher fluence, where Mn-Ni-Si precipitates can form, even in the absence of Cu. Therefore, extending RPV lifetimes will require a thorough understanding of both precipitation and embrittlement at higher fluences than have ever been observed in a power reactor. To address this issue, test reactors that irradiate materials at higher neutron fluxes than power reactors are used. These experiments at high neutron flux can reach extended life neutron fluences in only months or several years. The drawback of these test irradiations is that they add additional complexity to interpreting the data, as the irradiation flux also plays a role into both precipitate formation and irradiation hardening and embrittlement. This report focuses on developing a database of both microstructure and mechanical property data to better understand the effect of flux. In addition, a previously developed model that enables the comparison of data taken over a range of neutron flux is discussed.

  7. Fission Product Transport and Source Terms in HTRs: Experience from AVR Pebble Bed Reactor

    OpenAIRE

    Rainer Moormann

    2008-01-01

    Fission products deposited in the coolant circuit outside of the active core play a dominant role in source term estimations for advanced small pebble bed HTRs, particularly in design basis accidents (DBA). The deposited fission products may be released in depressurization accidents because present pebble bed HTR concepts abstain from a gas tight containment. Contamination of the circuit also hinders maintenance work. Experiments, performed from 1972 to 88 on the AVR, an experimental ...

  8. Cooling system for a nuclear reactor

    Science.gov (United States)

    Amtmann, Hans H.

    1982-01-01

    A cooling system for a gas-cooled nuclear reactor is disclosed which includes at least one primary cooling loop adapted to pass coolant gas from the reactor core and an associated steam generator through a duct system having a main circulator therein, and at least one auxiliary cooling loop having communication with the reactor core and adapted to selectively pass coolant gas through an auxiliary heat exchanger and circulator. The main and auxiliary circulators are installed in a common vertical cavity in the reactor vessel, and a common return duct communicates with the reactor core and intersects the common cavity at a junction at which is located a flow diverter valve operative to effect coolant flow through either the primary or auxiliary cooling loops.

  9. High density LHRF experiments in Alcator C-Mod and implications for reactor scale devices

    Science.gov (United States)

    Baek, S. G.; Parker, R. R.; Bonoli, P. T.; Shiraiwa, S.; Wallace, G. M.; LaBombard, B.; Faust, I. C.; Porkolab, M.; Whyte, D. G.

    2015-04-01

    Parametric decay instabilities (PDI) appear to be an ubiquitous feature of lower hybrid current drive (LHCD) experiments at high density. In density ramp experiments in Alcator C-Mod and other machines the onset of PDI activity has been well correlated with a decrease in current drive efficiency and production of fast electron bremsstrahlung. However whether PDI is the primary cause of the ‘density limit’, and if so by exactly what mechanism (beyond the obvious one of pump depletion) has not been clearly established. In order to further understand the connection, the frequency spectrum of PDI activity occurring during Alcator C-Mod LHCD experiments has been explored in detail by means of a number of RF probes distributed around the periphery of the C-Mod tokamak including a probe imbedded in the inner wall. The results show that (i) the excited spectra consists mainly of a few discrete ion cyclotron (IC) quasi-modes, which have higher growth than the ion sound branch; (ii) PDI activity can begin either at the inner or outer wall, depending on magnetic configuration; (iii) the frequencies of the IC quasi-modes correspond to the magnetic field strength close to the low-field side (LFS) or high-field side separatrix; and (iv) although PDI activity may initiate near the inner separatrix, the loss in fast electron bremsstrahlung is best correlated with the appearance of IC quasi-modes characteristic of the magnetic field strength near the LFS separatrix. These data, supported by growth rate calculations, point to the importance of the LFS scrape-off layer (SOL) density in determining PDI onset and degradation in current drive efficiency. By minimizing the SOL density it is possible to extend the core density regime over which PDI can be avoided, thus potentially maximizing the effectiveness of LHCD at high density. Increased current drive efficiency at high density has been achieved in FTU and EAST through lithium coating and special fuelling methods, and in recent

  10. 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.

  11. 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.

  12. 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.

  13. Engineering Evaluation of Proposed Alternative Salt Transfer Method for the Molten Salt Reactor Experiement for the Oak Ridge National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Carlberg, Jon A.; Roberts, Kenneth T.; Kollie, Thomas G.; Little, Leslie E.; Brady, Sherman D.

    2009-09-30

    This evaluation was performed by Pro2Serve in accordance with the Technical Specification for an Engineering Evaluation of the Proposed Alternative Salt Transfer Method for the Molten Salt Reactor Experiment at the Oak Ridge National Laboratory (BJC 2009b). The evaluators reviewed the Engineering Evaluation Work Plan for Molten Salt Reactor Experiment Residual Salt Removal, Oak Ridge National Laboratory, Oak Ridge, Tennessee (DOE 2008). The Work Plan (DOE 2008) involves installing a salt transfer probe and new drain line into the Fuel Drain Tanks and Fuel Flush Tank and connecting them to the new salt transfer line at the drain tank cell shield. The probe is to be inserted through the tank ball valve and the molten salt to the bottom of the tank. The tank would then be pressurized through the Reactive Gas Removal System to force the salt into the salt canisters. The Evaluation Team reviewed the work plan, interviewed site personnel, reviewed numerous documents on the Molten Salt Reactor (Sects. 7 and 8), and inspected the probes planned to be used for the transfer. Based on several concerns identified during this review, the team recommends not proceeding with the salt transfer via the proposed alternate salt transfer method. The major concerns identified during this evaluation are: (1) Structural integrity of the tanks - The main concern is with the corrosion that occurred during the fluorination phase of the uranium removal process. This may also apply to the salt transfer line for the Fuel Flush Tank. Corrosion Associated with Fluorination in the Oak Ridge National Laboratory Fluoride Volatility Process (Litman 1961) shows that this problem is significant. (2) Continued generation of Fluorine - Although the generation of Fluorine will be at a lower rate than experienced before the uranium removal, it will continue to be generated. This needs to be taken into consideration regardless of what actions are taken with the salt. (3) More than one phase of material

  14. Calculation of a materials relocation experiment simulating a core disruptive accident condition in fast breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sawada, T. [Tokyo Inst. of Tech. (Japan). Research Lab. for Nuclear Reactors; Ninokata, H. [Tokyo Inst. of Tech. (Japan). Research Lab. for Nuclear Reactors; Shimizu, A. [Tokyo Inst. of Tech. (Japan). Research Lab. for Nuclear Reactors

    1995-07-01

    This paper describes an interpretation of the SIMBATH (Simulationsexperimente in Brennelementattrappen mit Thermit) experiments that use the SIMMER-II code. A series of SIMBATH experiments has aimed at simulating fuel pin disintegration and following materials relocation in the test sections of a single pin to 37-pin bundles. In the calculation, three modifications were incorporated into the SIMMER-II code. With these modifications, the calculation showed good agreement with the experimental measurements with respect to the void region propagation in sodium flow and the molten materials relocation leading to flow blockage. A set of parametric calculations has clarified the range of applicability of parameters for materials relocation and flow blockage formation. The particle radius r{sub p} in blockage regions and the mutiplier for particle viscosity (PARVIS) are recommended to be r{sub p}>or{approx}1/2D{sub h} and 0.001Pas

  15. Operating experience feedback report -- turbine-generator overspeed protection systems: Commercial power reactors. Volume 11

    Energy Technology Data Exchange (ETDEWEB)

    Ornstein, H.L.

    1995-04-01

    This report presents the results of the US Nuclear Regulatory Commission`s Office for Analysis and Evaluation of Operational Data (AEOD) review of operating experience of main turbine-generator overspeed and overspeed protection systems. It includes an indepth examination of the turbine overspeed event which occurred on November 9, 1991, at the Salem Unit 2 Nuclear Power Plant. It also provides information concerning actions taken by other utilities and the turbine manufacturers as a result of the Salem overspeed event. AEOD`s study reviewed operating procedures and plant practices. It noted differences between turbine manufacturer designs and recommendations for operations, maintenance, and testing, and also identified significant variations in the manner that individual plants maintain and test their turbine overspeed protection systems. AEOD`s study provides insight into the shortcomings in the design, operation, maintenance, testing, and human factors associated with turbine overspeed protection systems. Operating experience indicates that the frequency of turbine overspeed events is higher than previously thought and that the bases for demonstrating compliance with NRC`s General Design Criterion (GDC) 4, Environmental and dynamic effects design bases, may be nonconservative with respect to the assumed frequency.

  16. SBLOCA AND LOFW EXPERIMENTS IN A SCALED-DOWN IET FACILITY OF REX-10 REACTOR

    Directory of Open Access Journals (Sweden)

    YEON-GUN LEE

    2013-06-01

    Full Text Available This paper presents an experimental investigation of the small-break loss-of-coolant accident (SBLOCA and the loss-of-feedwater accident (LOFW in a scaled integral test facility of REX-10. REX-10 is a small integral-type PWR in which the coolant flow is driven by natural circulation, and the RCS is pressurized by the steam-gas pressurizer. The postulated accidents of REX-10 include the system depressurization initiated by the break of a nitrogen injection line connected to the steam-gas pressurizer and the complete loss of normal feedwater flow by the malfunction of control systems. The integral effect tests on SBLOCA and LOFW are conducted at the REX-10 Test Facility (RTF, a full-height full-pressure facility with reduced power by 1/50. The SBLOCA experiment is initiated by opening a flow passage out of the pressurizer vessel, and the LOFW experiment begins with the termination of the feedwater supply into the helical-coil steam generator. The experimental results reveal that the RTF can assure sufficient cooldown capability with the simulated PRHRS flow during these DBAs. In particular, the RTF exhibits faster pressurization during the LOFW test when employing the steam-gas pressurizer than the steam pressurizer. This experimental study can provide unique data to validate the thermal-hydraulic analysis code for REX-10.

  17. Construction of the blowdown and condensation loop

    Energy Technology Data Exchange (ETDEWEB)

    Park, Choon Kyung; Song, Chul Kyung; Cho, Seok; Chun, S. Y.; Chung, Moon Ki

    1997-12-01

    The blowdown and condensation loop (B and C loop) has been constructed to get experimental data for designing the safety depressurization system (SDS) and steam sparger which are considered to implement in the Korea Next Generation Reactor (KNGR). In this report, system description on the B and C loop is given in detail, which includes the drawings and technical specification of each component, instrumentation and control system, and the operational procedures and the results of the performance testing. (author). 7 refs., 11 tabs., 48 figs.

  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. Development of An Embedded FPGA-Based Data Acquisition System Dedicated to Zero Power Reactor Noise Experiments

    Directory of Open Access Journals (Sweden)

    Arkani Mohammad

    2014-08-01

    Full Text Available An embedded time interval data acquisition system (DAS is developed for zero power reactor (ZPR noise experiments. The system is capable of measuring the correlation or probability distribution of a random process. The design is totally implemented on a single Field Programmable Gate Array (FPGA. The architecture is tested on different FPGA platforms with different speed grades and hardware resources. Generic experimental values for time resolution and inter-event dead time of the system are 2.22 ns and 6.67 ns respectively. The DAS can record around 48-bit x 790 kS/s utilizing its built-in fast memory. The system can measure very long time intervals due to its 48-bit timing structure design. As the architecture can work on a typical FPGA, this is a low cost experimental tool and needs little time to be established. In addition, revisions are easily possible through its reprogramming capability. The performance of the system is checked and verified experimentally.

  20. PAR Loop Schedule Review

    Energy Technology Data Exchange (ETDEWEB)

    Schaffer, Jr.; W.F.

    1958-04-30

    The schedule for the installation of the PAR slurry loop experiment in the South Facility of the ORR has been reviewed and revised. The design, fabrications and Installation is approximately two weeks behind schedule at this time due to many factors; however, indications are that this time can be made up. Design is estimated to be 75% complete, fabrication 32% complete and installation 12% complete.

  1. Thermal hydraulic test for reactor safety system - Critical heat flux experiment and development of prediction models

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Soon Heung; Baek, Won Pil; Yang, Soo Hyung; No, Chang Hyun [Korea Advanced Institute of Science and Technology, Taejon (Korea)

    2000-04-01

    To acquire CHF data through the experiments and develop prediction models, research was conducted. Final objectives of research are as follows: 1) Production of tube CHF data for low and middle pressure and mass flux and Flow Boiling Visualization. 2) Modification and suggestion of tube CHF prediction models. 3) Development of fuel bundle CHF prediction methodology base on tube CHF prediction models. The major results of research are as follows: 1) Production of the CHF data for low and middle pressure and mass flux. - Acquisition of CHF data (764) for low and middle pressure and flow conditions - Analysis of CHF trends based on the CHF data - Assessment of existing CHF prediction methods with the CHF data 2) Modification and suggestion of tube CHF prediction models. - Development of a unified CHF model applicable for a wide parametric range - Development of a threshold length correlation - Improvement of CHF look-up table using the threshold length correlation 3) Development of fuel bundle CHF prediction methodology base on tube CHF prediction models. - Development of bundle CHF prediction methodology using correction factor. 11 refs., 134 figs., 25 tabs. (Author)

  2. Aging assessment of reactor instrumentation and protection system components. Aging-related operating experiences

    Energy Technology Data Exchange (ETDEWEB)

    Gehl, A.C.; Hagen, E.W. [Oak Ridge National Lab., TN (United States)

    1992-07-01

    A study of the aging-related operating experiences throughout a five-year period (1984--1988) of six generic instrumentation modules (indicators, sensors, controllers, transmitters, annunciators, and recorders) was performed as a part of the Nuclear Plant Aging Research Program. The effects of aging from operational and environmental stressors were characterized from results depicted in Licensee Event Reports (LERs). The data are graphically displayed as frequency of events per plant year for operating plant ages from 1 to 28 years to determine aging-related failure trend patterns. Three main conclusions were drawn from this study: (1) Instrumentation and control (I&C) modules make a modest contribution to safety-significant events: 17% of LERs issued during 1984--1988 dealt with malfunctions of the six I&C modules studied, and 28% of the LERs dealing with these I&C module malfunctions were aging related (other studies show a range 25--50%); (2) Of the six modules studied, indicators, sensors, and controllers account for the bulk (83%) of aging-related failures; and (3) Infant mortality appears to be the dominant aging-related failure mode for most I&C module categories (with the exception of annunciators and recorders, which appear to fail randomly).

  3. Naval Reactors Prime Contractor Team (NRPCT) Experiences and Considerations With Irradiation Test Performance in an International Environment

    Energy Technology Data Exchange (ETDEWEB)

    MH Lane

    2006-02-15

    This letter forwards a compilation of knowledge gained regarding international interactions and issues associated with Project Prometheus. The following topics are discussed herein: (1) Assessment of international fast reactor capability and availability; (2) Japanese fast reactor (JOYO) contracting strategy; (3) NRPCT/Program Office international contract follow; (4) Completion of the Japan Atomic Energy Agency (JAEA)/Pacific Northwest National Laboratory (PNNL) contract for manufacture of reactor test components; (5) US/Japanese Departmental interactions and required Treaties and Agreements; and (6) Non-technical details--interactions and considerations.

  4. NOMAGE4 activities 2011, Part II, Supercritical water loop

    DEFF Research Database (Denmark)

    Vierstraete, Pierre; Van Nieuwenhove, Rudi; Lauritzen, Bent

    and fuel tests. Working on this direction, the Halden Reactor Project has started an activity in collaboration with Risoe-DTU (with Mr. Rudi Van Nieuwenhove as the project leader) to study the feasibility of a SCW loop in the Halden Reactor, which is a Heavy Boiling Water Reactor (HBWR). The ultimate goal......The supercritical water reactor (SCWR) is one of the six different reactor technologies selected for research and development under the Generation IV program. Several countries have shown interest to this concept but up to now, there exist no in-pile facilities to perform the required material...

  5. Flow analyses for the LAVA-ERVC experiment and the KSNP under the external reactor vessel cooling using RELAP5/MOD3 code

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Kyoung-Ho; Park, Rae-Joon; Cho, Young-Ro; Kim, Sang-Baik

    2005-01-01

    Flow analyses were performed using RELAP5/MOD3 code to investigate and verify the steam binding phenomena in the LAVA-ERVC experiment and to investigate the occurrence and the effects of steam binding for the KSNP under the external reactor vessel cooling. Flow analyses for the LAVA-ERVC experiments confirmed the steam binding occurrence in case of the limited steam venting and represented the LAVA-ERVC experimental results quite well. The flow analyses results for the KSNP under the external reactor vessel cooling address that water ingression and steam ventilation through the insulator are crucial factors determining the effective cool down via boiling heat removal at the outer surface of the RPV lower plenum. The flow analyses results for the base cases of the SBO and the 9.6 inch LBLOCA imply that the limited steam venting through the insulator induced the steam binding and eventually prevented the effective cooling at the outer surface of the RPV lower plenum. From the sensitivity study on the additional flow area for the steam venting, it could be found that the RPV lower plenum experienced effective cooling by smooth water circulation. The current RELAP5 flow analyses results for the KSNP under the external reactor vessel cooling address that prevention of steam binding phenomena should be settled first for the in-vessel corium retention through the external reactor vessel cooling. Implementation of additional flow path for the effective steam ventilation is highly recommended as one of the most promising countermeasures to enhance the coolability through the external reactor vessel cooling.

  6. Progresses in the operation of large scale LBE loop: Helios

    Energy Technology Data Exchange (ETDEWEB)

    Jun, Lim; Seung Ho, Jeong; Young Jin, Oh; Hyo On, Nam; Chi Bum, Bahn; Seung Hee, Chang; Won Chang, Nam; Kyung Ha, Ryu; Tai Hyun, Lee; Seung Gi, Lee; Na Young, Lee; Il Soon, Hwang [Seoul National Univ., Nuclear Transmutation Energy Research Center of Korea (NUTRECK) (Korea, Republic of)

    2007-07-01

    A 12 m tall LBE (lead-bismuth eutectic) coolant loop, named as HELIOS (Heavy Eutectic liquid metal Loop for Investigation of Operability and Safety of PEACER), has been developed by thermal-hydraulic scaling of the PEACER-300-MWe (Proliferation-resistant, Environment-friendly, Accident-tolerant, Continual and Economical Reactor). HELIOS has 5 main vessels: mockup core, expansion tank, heat exchanger, sump tank (containing sump pump inside) and storage tank. The total power of the 4 four main heaters in the mockup core is 60 kW. Natural circulation experiment and material test are principal purposes of HELIOS operation. During preliminary tests, many practical difficulties arouse with sensors for pressure measurement and with the orifice flow meter for flow rate measurement. After modification and calibration of sensors, forced circulation and natural circulation tests were performed. For the long term material test, specimen cage was designed to control the flow-rate only by natural circulation. The required flow-rate of material test section is over 1 m/s under natural circulation condition. Natural circulation capability of HELIOS is expected to be sufficient to achieve flow velocity in test sections. Preliminary corrosion tests were performed in order to establish optimum process of LBE corrosion experiment in large loop.

  7. Ageing implementation and refurbishment development at the IEA-R1 nuclear research reactor: a 15 years experience

    Energy Technology Data Exchange (ETDEWEB)

    Cardenas, Jose Patricio N.; Ricci Filho, Walter; Carvalho, Marcos R. de; Berretta, Jose Roberto; Marra Neto, Adolfo, E-mail: ahiru@ipen.b, E-mail: wricci@ipen.b, E-mail: carvalho@ipen.b, E-mail: jrretta@ipen.b, E-mail: amneto@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    IPEN (Instituto de Pesquisas Energeticas e Nucleares) is a nuclear research center established into the Secretary of Science and Technology from the government of the state of Sao Paulo, and administered both technically and financially by Comissao Nacional de Energia Nuclear (CNEN), a federal government organization under the Ministry of Science and Technology. The institute is located inside the campus of the University of Sao Paulo, Sao Paulo city, Brazil. One of major nuclear facilities at IPEN is the IEA-R1 nuclear research reactor. It is the unique Brazilian research reactor with substantial power level suitable for application with research in physics, chemistry, biology and engineering, as well as radioisotope production for medical and other applications. Designed and built by Babcok-Wilcox, in accordance with technical specifications established by the Brazilian Nuclear Energy Commission, and financed by the US Atoms for Peace Program, it is a swimming pool type reactor, moderated and cooled by light water and uses graphite and beryllium as reflector elements. The first criticality was achieved on September 16, 1957 and the reactor is currently operating at 4.0 MW on a 64h per week cycle. Since 1996, an IEA-R1 reactor ageing study was established at the Research Reactor Center (CRPq) related with general deterioration of components belonging to some operational systems, as cooling towers from secondary cooling system, piping and pumps, sample irradiation devices, radiation monitoring system, fuel elements, rod drive mechanisms, nuclear and process instrumentation and safety operational system. Although basic structures are almost the same as the original design, several improvements and modifications in components, systems and structures had been made along reactor life. This work aims to show the development of the ageing program in the IEA-R1 reactor and the upgrading (modernization) that was carried out, concerning several equipment and system in the

  8. Experimental validation of packed bed chemical-looping combustion

    NARCIS (Netherlands)

    Noorman, S.; van Sint Annaland, M.; Kuipers, J.A.M.

    2010-01-01

    Chemical-looping combustion has emerged as a promising alternative technology, intrinsically integrating CO2 capture in power production. A novel reactor concept based on dynamically operated packed beds has been proposed [Noorman, S., van Sint Annaland, M., Kuipers, J.A.M., 2007. Packed bed reactor

  9. Geologic Investigation of a Potential Site for a Next-Generation Reactor Neutrino Oscillation Experiment -- Diablo Canyon, San Luis Obispo County, CA

    OpenAIRE

    Onishi, Celia Tiemi; Dobson, Patrick; Nakagawa, Seiji; Glaser, Steven; Galic, Dom

    2004-01-01

    This report provides information on the geology and selected physical and mechanical properties of surface rocks collected at Diablo Canyon, San Luis Obispo County, California as part of the design and engineering studies towards a future reactor neutrino oscillation experiment. The main objective of this neutrino project is to study the process of neutrino flavor transformation or neutrino oscillation by measuring neutrinos produced in the fission reactions of a nuclear power plant. Di...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-02-01

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

  11. Homogeneous Reactor Experiment (HRE) Pond cryogenic barrier technology demonstration: Pre-barrier subsurface hydrology and contaminant transport investigation

    Energy Technology Data Exchange (ETDEWEB)

    Moline, G.R.

    1998-03-01

    The Homogeneous Reactor Experiment (HRE) Pond is the site of a former impoundment for radioactive wastes that has since been drained, filled with soil, and covered with an asphalt cap. The site is bordered to the east and south by a tributary that empties into Melton Branch Creek and that contains significant concentrations of radioactive contaminants, primarily {sup 90}Sr. Because of the proximity of the tributary to the HRE disposal site and the probable flow of groundwater from the site to the tributary, it is hypothesized that the HRE Pond is a source of contamination to he creek. As a means for temporary containment of contaminants within the impoundment, a cryogenic barrier technology demonstration was initiated in FY96 with a background hydrologic investigation that continued through FY97. Cryogenic equipment installation was completed in FY97, and freezing was initiated in September of 1997. This report documents the results of a hydrologic and geologic investigation of the HRE Pond/cryogenic barrier site. The purpose of this investigation is to evaluate the hydrologic conditions within and around the impoundment in order to meet the following objectives: (1) to provide a pre-barrier subsurface hydrologic baseline for post-barrier performance assessment; (2) to confirm that the impoundment is hydraulically connected to the surrounding sediments; and (3) to determine the likely contaminant exit pathways from the impoundment. The methods of investigation included water level and temperature monitoring in a network of wells and standpipes in and surrounding the impoundment, a helium tracer test conducted under ambient flow conditions, and geologic logging during the drilling of boreholes for installation of cryogenic probes and temperature monitoring wells.

  12. 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

  13. Energy Conversion Loop: A Testbed for Nuclear Hybrid Energy Systems Use in Biomass Pyrolysis

    Science.gov (United States)

    Verner, Kelley M.

    Nuclear hybrid energy systems are a possible solution for contemporary energy challenges. Nuclear energy produces electricity without greenhouse gas emissions. However, nuclear power production is not as flexible as electrical grids demand and renewables create highly variable electricity. Nuclear hybrid energy systems are able to address both of these problems. Wasted heat can be used in processes such as desalination, hydrogen production, or biofuel production. This research explores the possible uses of nuclear process heat in bio-oil production via biomass pyrolysis. The energy conversion loop is a testbed designed and built to mimic the heat from a nuclear reactor. Small scale biomass pyrolysis experiments were performed and compared to results from the energy conversion loop tests to determine future pyrolysis experimentation with the energy conversion loop. Further improvements must be made to the energy conversion loop before more complex experiments may be performed. The current conditions produced by the energy conversion loop are not conducive for current biomass pyrolysis experimentation.tion.

  14. The neutronic analysis of opportunity of ITER blanket element tests in RF research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lopatkin, A.; Tocheniy, L. [ENTEK-RDIPE, Moscow (Russian Federation)

    1994-12-31

    In the framework of development of plan of in-pile radiative tests of ITER blanket elements the calculations are carried out of the models of tritium-producing elements in loop channels, placed in the number of Russian various type test reactors. There are presented: (1) The variants of models of blanket, on the base of which the set of experiment goal parameters and its ranges are formed; (2) Outline of loop channel; (3) The experimental opportunities of research reactors with thermal (SM-3, MIR, IVV-2M, RBT) and fast (BOR-60, EBR) spectra of neutrons; (4) The calculation procedures - settlement models, codes. The results are given: (1) power generation rates in components of channel; (2) the tritium breeding rate; (3) the helium production rate in beryllium; (4) the neutron group fluxes; (5) absorption rates in zones of loop channel. The possible reactivity effects due to experimental channel accommodation in reactor core and to radiated sample replace inside of the channel are shown. The last section includes the recommendations for the choice of reactor acceptable from the neutronics point of view, and for the next study directions and stages.

  15. Alternative loop rings

    CERN Document Server

    Goodaire, EG; Polcino Milies, C

    1996-01-01

    For the past ten years, alternative loop rings have intrigued mathematicians from a wide cross-section of modern algebra. As a consequence, the theory of alternative loop rings has grown tremendously. One of the main developments is the complete characterization of loops which have an alternative but not associative, loop ring. Furthermore, there is a very close relationship between the algebraic structures of loop rings and of group rings over 2-groups. Another major topic of research is the study of the unit loop of the integral loop ring. Here the interaction between loop rings and group ri

  16. 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.

  17. 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.

  18. In-air and pressurized water reactor environment fatigue experiments of 316 atainless ateel to study the effect of environment on cyclic hardening

    Energy Technology Data Exchange (ETDEWEB)

    Mohanty, Subhasish; Soppet, William; Majumdar, Saurindranath; Natesan, Krishnamurti

    2016-05-01

    Argonne National Laboratory (ANL), under the sponsorship of Department of Energy’s Light Water Reactor Sustainability (LWRS) program, is trying to develop a mechanistic approach for more accurate life estimation of LWR components. In this context, ANL has conducted many fatigue experiments under different test and environment conditions on type 316 stainless steel (316SS) material which is widely used in the US reactors. Contrary to the conventional S~N curve based empirical fatigue life estimation approach, the aim of the present DOE sponsored work is to develop an understanding of the material ageing issues more mechanistically (e.g. time dependent hardening and softening) under different test and environmental conditions. Better mechanistic understanding will help develop computer-based advanced modeling tools to better extrapolate stress-strain evolution of reactor components under multi-axial stress states and hence help predict their fatigue life more accurately. In this paper (part-I) the fatigue experiments under different test and environment conditions and related stress-strain results for 316 SS are discussed. In a second paper (part-II) the related evolutionary cyclic plasticity material modeling techniques and results are discussed.

  19. Conceptual design of a First Wall mock-up experiment in preparation for the qualification of breeding blanket technologies in the Helium Loop Karlsruhe (HELOKA) facility

    Energy Technology Data Exchange (ETDEWEB)

    Zeile, C., E-mail: christian.zeile@kit.edu [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Abou-Sena, A.; Boccaccini, L.V.; Ghidersa, B.E.; Kang, Q.; Kunze, A. [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Lamberti, L. [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Dipartimento Energia, Politecnico di Torino (Italy); Maione, I.A.; Rey, J.; Weth, A. von der [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2016-11-01

    Highlights: • Experiment in preparation for the qualification of Breeding Blanket technologies in HELOKA facility is proposed. • Experimental capabilities, instrumentation of the mock-up and experimental program are presented. • Design and manufacturing of the mock-up is described. • Design of modular attachment system to obtain different stress levels and distributions on the mock-up is discussed. - Abstract: An experimental program based on a First Wall mock-up is presented as preparation for the qualification of breeding blanket mock-ups at high heat flux in the Helium Loop Karlsruhe (HELOKA) facility. Two objectives of the experimental program have been defined: testing of the experimental setup and a first validation of FE models. The design and manufacturing of mock-up representing about 1/3 of the heated zone of an ITER Test Blanket Module (TBM) First Wall is discussed. A modular attachment system concept has been developed for the fixation of the mock-up in order to be able to generate different stress distributions and levels on the plate, which is confirmed by thermo-mechanical analyses. The HELOKA facility is able to provide a TBM relevant helium cooling system and to generate the required surface heat flux by an electron beam gun. An installed IR camera can be used to measure the temperature distribution on the surface.

  20. 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.

  1. Student Collaboration in a Series of Integrated Experiments to Study Enzyme Reactor Modeling with Immobilized Cell-Based Invertase

    Science.gov (United States)

    Taipa, M. A^ngela; Azevedo, Ana M.; Grilo, Anto´nio L.; Couto, Pedro T.; Ferreira, Filipe A. G.; Fortuna, Ana R. M.; Pinto, Ine^s F.; Santos, Rafael M.; Santos, Susana B.

    2015-01-01

    An integrative laboratory study addressing fundamentals of enzyme catalysis and their application to reactors operation and modeling is presented. Invertase, a ß-fructofuranosidase that catalyses the hydrolysis of sucrose, is used as the model enzyme at optimal conditions (pH 4.5 and 45 °C). The experimental work involves 3 h of laboratory time…

  2. Ignition of DME and DME/CH4 at High Pressure: Flow Reactor Experiments and Kinetic Modeling

    DEFF Research Database (Denmark)

    Hashemi, Hamid; Christensen, Jakob Munkholt; Glarborg, Peter

    The pyrolysis and oxidation of dimethyl ether (DME) and its mixtures with methane were investigated at high pressures (50 and 100 bar) and intermediate temperatures (450―900 K) in a laminar flow reactor. DME pyrolysis started at 825 K (at 50 bar). The onset of DME reaction was detected at 525―550 K...

  3. Operating experience feedback report: Reliability of safety-related steam turbine-driven standby pumps. Commercial power reactors, Volume 10

    Energy Technology Data Exchange (ETDEWEB)

    Boardman, J.R.

    1994-10-01

    This report documents a detailed analysis of failure initiators, causes and design features for steam turbine assemblies (turbines with their related components, such as governors and valves) which are used as drivers for standby pumps in the auxiliary feedwater systems of US commercial pressurized water reactor plants, and in the high pressure coolant injection and reactor core isolation cooling systems of US commercial boiling water reactor plants. These standby pumps provide a redundant source of water to remove reactor core heat as specified in individual plant safety analysis reports. The period of review for this report was from January 1974 through December 1990 for licensee event reports (LERS) and January 1985 through December 1990 for Nuclear Plant Reliability Data System (NPRDS) failure data. This study confirmed the continuing validity of conclusions of earlier studies by the US Nuclear Regulatory Commission and by the US nuclear industry that the most significant factors in failures of turbine-driven standby pumps have been the failures of the turbine-drivers and their controls. Inadequate maintenance and the use of inappropriate vendor technical information were identified as significant factors which caused recurring failures.

  4. Thermal-hydraulic and structural safety analysis of SLSF P3 experiment

    Energy Technology Data Exchange (ETDEWEB)

    Ragland, W A; Ariman, T; Tessier, J H

    1979-01-01

    The Sodium Loop Safety Facility (SLSF) P3 experiment was the fourth in a series of in-reactor tests in the Engineering Test Reactor and simulated an unprotected flow coastdown in a 37 pin bundle. A comprehensive thermal-hydraulic analysis of the SLSF loop was coupled with a structural analysis of the test section hexcan outer duct in order to provide assurance, before the transient, that the loop would safely contain the P3 experiment. This analysis was performed for both the expected transient and for an off-normal condition of failure of redundant logic circuits to provide the proper pump power demand signal. Results of the analysis showed the hexcan outer duct to safely contain the P3 experiment for both conditions. The analysis for the expected transient has been verified by the successful completion of the P3 experiment.

  5. Calculation of reactor antineutrino spectra in TEXONO

    CERN Document Server

    Chen Dong Liang; Mao Ze Pu; Wong, T H

    2002-01-01

    In the low energy reactor antineutrino physics experiments, either for the researches of antineutrino oscillation and antineutrino reactions, or for the measurement of abnormal magnetic moment of antineutrino, the flux and the spectra of reactor antineutrino must be described accurately. The method of calculation of reactor antineutrino spectra was discussed in detail. Furthermore, based on the actual circumstances of NP2 reactors and the arrangement of detectors, the flux and the spectra of reactor antineutrino in TEXONO were worked out

  6. Advanced Catalytic Hydrogenation Retrofit Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Reinaldo M. Machado

    2002-08-15

    Industrial hydrogenation is often performed using a slurry catalyst in large stirred-tank reactors. These systems are inherently problematic in a number of areas, including industrial hygiene, process safety, environmental contamination, waste production, process operability and productivity. This program proposed the development of a practical replacement for the slurry catalysts using a novel fixed-bed monolith catalyst reactor, which could be retrofitted onto an existing stirred-tank reactor and would mitigate many of the minitations and problems associated with slurry catalysts. The full retrofit monolith system, consisting of a recirculation pump, gas/liquid ejector and monolith catalyst, is described as a monolith loop reactor or MLR. The MLR technology can reduce waste and increase raw material efficiency, which reduces the overall energy required to produce specialty and fine chemicals.

  7. A Compilation of Boiling Water Reactor Operational Experience for the United Kingdom's Office for Nuclear Regulation's Advanced Boiling Water Reactor Generic Design Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Wheeler, Timothy A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Liao, Huafei [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2014-12-01

    United States nuclear power plant Licensee Event Reports (LERs), submitted to the United States Nuclear Regulatory Commission (NRC) under law as required by 10 CFR 50.72 and 50.73 were evaluated for reliance to the United Kingdom’s Health and Safety Executive – Office for Nuclear Regulation’s (ONR) general design assessment of the Advanced Boiling Water Reactor (ABWR) design. An NRC compendium of LERs, compiled by Idaho National Laboratory over the time period January 1, 2000 through March 31, 2014, were sorted by BWR safety system and sorted into two categories: those events leading to a SCRAM, and those events which constituted a safety system failure. The LERs were then evaluated as to the relevance of the operational experience to the ABWR design.

  8. Crystal Plasticity Model of Reactor Pressure Vessel Embrittlement in GRIZZLY

    Energy Technology Data Exchange (ETDEWEB)

    Chakraborty, Pritam [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Biner, Suleyman Bulent [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Zhang, Yongfeng [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Spencer, Benjamin Whiting [Idaho National Laboratory (INL), Idaho Falls, ID (United States)

    2015-07-01

    The integrity of reactor pressure vessels (RPVs) is of utmost importance to ensure safe operation of nuclear reactors under extended lifetime. Microstructure-scale models at various length and time scales, coupled concurrently or through homogenization methods, can play a crucial role in understanding and quantifying irradiation-induced defect production, growth and their influence on mechanical behavior of RPV steels. A multi-scale approach, involving atomistic, meso- and engineering-scale models, is currently being pursued within the GRIZZLY project to understand and quantify irradiation-induced embrittlement of RPV steels. Within this framework, a dislocation-density based crystal plasticity model has been developed in GRIZZLY that captures the effect of irradiation-induced defects on the flow stress behavior and is presented in this report. The present formulation accounts for the interaction between self-interstitial loops and matrix dislocations. The model predictions have been validated with experiments and dislocation dynamics simulation.

  9. 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.)

  10. Introduction to the study of an optimal control for irradiation loops of the reactor Pegase; Introduction a l'etude d'une commande optimale des boucles d'irradiation du reacteur Pegase

    Energy Technology Data Exchange (ETDEWEB)

    Guintrand, C. [Commissariat a l' Energie Atomique, Cadarache (France). Centre d' Etudes Nucleaires

    1968-07-01

    The control system under consideration is made up of: a regulation unit consisting of a conventional nonlinear looped circuit for static tests, a cycling unit operating in open loop for dynamic tests. After a definition of a mathematical model for an irradiation loop, the behaviour of the regulation unit is studied, first of all theoretically using three-dimensional topological methods, and then by analogue simulation. A prototype unit is under construction and its principal characteristics are given. Finally, as far as the cycling unit is concerned, the first tests involving self-instruction technique, are described. (author) [French] Le systeme de commande envisage se compose de r: une unite de regulation se presentant sous la forme d'un circuit boucle classique non-lineaire, pour les essais statiques, une unite de cyclage fonctionnant en boucle ouverte, pour les essais dynamiques. Apres avoir defini un modele mathematique d'une boucle d'irradiation, on etudie le comportement de l'unite de regulation d'abord de facon theorique par les methodes topologiques a trois dimensions, puis par une simulation analogique. Une unite prototype est en cours de realisation, les principales caracteristiques en sont donnees. Enfin, en ce qui concerne l'unite de cyclage, les premiers essais effectues faisant appel aux techniques d'autoapprentissage, sont decrits. (auteur)

  11. Identification and evaluation of alternatives for the disposition of fluoride fuel and flush salts from the molten salt reactor experiment at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-08-15

    This document presents an initial identification and evaluation of the alternatives for disposition of the fluoride fuel and flush salts stored in the drain tanks at the Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory (ORNL). It will serve as a resource for the U.S. Department of Energy contractor preparing the feasibility study for this activity under the Comprehensive Environmental Response, Compensation and Liability Act (CERCLA). This document will also facilitate further discussion on the range of credible alternatives, and the relative merits of alternatives, throughout the time that a final alternative is selected under the CERCLA process.

  12. Chemical Looping Combustion of Rice Husk

    OpenAIRE

    Rashmeet Singh Monga; Ganesh R. Kale

    2015-01-01

    A thermodynamic investigation of direct chemical looping combustion (CLC) of rice husk is presented in this paper. Both steam and CO2 are used for gasification within the temperature range of 500–1200˚C and different amounts of oxygen carriers. Chemical equilibrium model was considered for the CLC fuel reactor. The trends in product compositions of the fuel reactor, were determined. Rice husk gasification using 3 moles H2O and 0 moles CO2 per mole carbon (in rice husk) at 1 bar pr...

  13. The HLMA project: determination of high Δm2 LMA mixing parameters and constraint on |Ue3| with a new reactor neutrino experiment

    Science.gov (United States)

    Schönert, Stefan; Lasserre, Thierry; Oberauer, Lothar

    2003-03-01

    In the forthcoming months, the KamLAND experiment will probe the parameter space of the solar large mixing angle MSW solution as the origin of the solar neutrino deficit with ν¯e's from distant nuclear reactors. If however the solution realized in nature is such that Δm2sol>~2×10-4 eV2 (thereafter named the HLMA region), KamLAND will only observe a rate suppression but no spectral distortion and hence it will not have the optimal sensitivity to measure the mixing parameters. In this case, we propose a new medium baseline reactor experiment located at Heilbronn (Germany) to pin down the precise value of the solar mixing parameters. In this paper, we present the Heilbronn detector site, we calculate the ν¯e interaction rate and the positron spectrum expected from the surrounding nuclear power plants. We also discuss the sensitivity of such an experiment to |Ue3| in both normal and inverted neutrino mass hierarchy scenarios. We then outline the detector design, estimate background signals induced by natural radioactivity as well as by in situ cosmic ray muon interaction, and discuss a strategy to detect the anti-neutrino signal `free of background'.

  14. Transient modelling of a natural circulation loop under variable pressure

    Energy Technology Data Exchange (ETDEWEB)

    Vianna, Andre L.B.; Faccini, Jose L.H.; Su, Jian, E-mail: avianna@nuclear.ufrj.br, E-mail: sujian@nuclear.ufrj.br, E-mail: faccini@ien.gov.br [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear; Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil). Lab. de Termo-Hidraulica Experimental

    2017-07-01

    The objective of the present work is to model the transient operation of a natural circulation loop, which is one-tenth scale in height to a typical Passive Residual Heat Removal system (PRHR) of an Advanced Pressurized Water Nuclear Reactor and was designed to meet the single and two-phase flow similarity criteria to it. The loop consists of a core barrel with electrically heated rods, upper and lower plena interconnected by hot and cold pipe legs to a seven-tube shell heat exchanger of countercurrent design, and an expansion tank with a descending tube. A long transient characterized the loop operation, during which a phenomenon of self-pressurization, without self-regulation of the pressure, was experimentally observed. This represented a unique situation, named natural circulation under variable pressure (NCVP). The self-pressurization was originated in the air trapped in the expansion tank and compressed by the loop water dilatation, as it heated up during each experiment. The mathematical model, initially oriented to the single-phase flow, included the heat capacity of the structure and employed a cubic polynomial approximation for the density, in the buoyancy term calculation. The heater was modelled taking into account the different heat capacities of the heating elements and the heater walls. The heat exchanger was modelled considering the coolant heating, during the heat exchanging process. The self-pressurization was modelled as an isentropic compression of a perfect gas. The whole model was computationally implemented via a set of finite difference equations. The corresponding computational algorithm of solution was of the explicit, marching type, as for the time discretization, in an upwind scheme, regarding the space discretization. The computational program was implemented in MATLAB. Several experiments were carried out in the natural circulation loop, having the coolant flow rate and the heating power as control parameters. The variables used in the

  15. 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.

  16. 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.

  17. Growth and Expansion of the International Criticality Safety Benchmark Evaluation Project and the Newly Organized International Reactor Physics Experiment Evaluation Project

    Energy Technology Data Exchange (ETDEWEB)

    J. Blair Briggs; Lori Scott; Yolanda Rugama; Enrico Satori

    2007-05-01

    Since ICNC 2003, the International Criticality Safety Benchmark Evaluation Project (ICSBEP) has continued to expand its efforts and broaden its scope. Criticality-alarm / shielding type benchmarks and fundamental physics measurements that are relevant to criticality safety applications are not only included in the scope of the project, but benchmark data are also included in the latest version of the handbook. A considerable number of improvements have been made to the searchable database, DICE and the criticality-alarm / shielding benchmarks and fundamental physics measurements have been included in the database. There were 12 countries participating on the ICSBEP in 2003. That number has increased to 18 with recent contributions of data and/or resources from Brazil, Czech Republic, Poland, India, Canada, and China. South Africa, Germany, Argentina, and Australia have been invited to participate. Since ICNC 2003, the contents of the “International Handbook of Evaluated Criticality Safety Benchmark Experiments” have increased from 350 evaluations (28,000 pages) containing benchmark specifications for 3070 critical or subcritical configurations to 442 evaluations (over 38,000 pages) containing benchmark specifications for 3957 critical or subcritical configurations, 23 criticality-alarm-placement / shielding configurations with multiple dose points for each, and 20 configurations that have been categorized as fundamental physics measurements that are relevant to criticality safety applications in the 2006 Edition of the ICSBEP Handbook. Approximately 30 new evaluations and 250 additional configurations are expected to be added to the 2007 Edition of the Handbook. Since ICNC 2003, a reactor physics counterpart to the ICSBEP, The International Reactor Physics Experiment Evaluation Project (IRPhEP) was initiated. Beginning in 1999, the IRPhEP was conducted as a pilot activity by the by the Organization of Economic Cooperation and Development (OECD) Nuclear Energy

  18. LBB application in the US operating and advanced reactors

    Energy Technology Data Exchange (ETDEWEB)

    Wichman, K.; Tsao, J.; Mayfield, M.

    1997-04-01

    The regulatory application of leak before break (LBB) for operating and advanced reactors in the U.S. is described. The U.S. Nuclear Regulatory Commission (NRC) has approved the application of LBB for six piping systems in operating reactors: reactor coolant system primary loop piping, pressurizer surge, safety injection accumulator, residual heat removal, safety injection, and reactor coolant loop bypass. The LBB concept has also been applied in the design of advanced light water reactors. LBB applications, and regulatory considerations, for pressurized water reactors and advanced light water reactors are summarized in this paper. Technology development for LBB performed by the NRC and the International Piping Integrity Research Group is also briefly summarized.

  19. Loop quantum gravity and observations

    CERN Document Server

    Barrau, A

    2014-01-01

    Quantum gravity has long been thought to be completely decoupled from experiments or observations. Although it is true that smoking guns are still missing, there are now serious hopes that quantum gravity phenomena might be tested. We review here some possible ways to observe loop quantum gravity effects either in the framework of cosmology or in astroparticle physics.

  20. In-air and pressurized water reactor environment fatigue experiments of 316 stainless steel to study the effect of environment on cyclic hardening

    Energy Technology Data Exchange (ETDEWEB)

    Mohanty, Subhasish, E-mail: smohanty@anl.gov; Soppet, William K., E-mail: soppet@anl.gov; Majumdar, Saurindranath, E-mail: majumdar@anl.gov; Natesan, Krishnamurti, E-mail: natesan@anl.gov

    2016-05-15

    Argonne National Laboratory (ANL), under the sponsorship of Department of Energy's Light Water Reactor Sustainability (LWRS) program, is trying to develop a mechanistic approach for more accurate life estimation of LWR components. In this context, ANL has conducted many fatigue experiments under different test and environment conditions on type 316 stainless steel (316 SS) material which is widely used in the US reactors. Contrary to the conventional S ∼ N curve based empirical fatigue life estimation approach, the aim of the present DOE sponsored work is to develop an understanding of the material ageing issues more mechanistically (e.g. time dependent hardening and softening) under different test and environmental conditions. Better mechanistic understanding will help develop computer-based advanced modeling tools to better extrapolate stress-strain evolution of reactor components under multi-axial stress states and hence help predict their fatigue life more accurately. Mechanics-based modeling of fatigue such as by using finite element (FE) tools requires the time/cycle dependent material hardening properties. Presently such time-dependent material hardening properties are hardly available in fatigue modeling literature even under in-air conditions. Getting those material properties under PWR environment, are even harder. Through this work we made preliminary attempt to generate time/cycle dependent stress-strain data both under in-air and PWR water conditions for further study such as for possible development of material models and constitutive relations for FE model implementation. Although, there are open-ended possibility to further improve the discussed test methods and related material estimation techniques we anticipate that the data presented in this paper will help the metal fatigue research community particularly, the researchers who are dealing with mechanistic modeling of metal fatigue such as using FE tools. In this paper the fatigue

  1. NOMAGE4 activities 2011. Part II, Supercritical water loop

    Energy Technology Data Exchange (ETDEWEB)

    Vierstraete, P. (Ecole Nationale Superieure des mines, Paris (France)); Van Nieuwenhove, R. (Institutt for Energiteknikk, OECD Halden Reactor Project (HRP), Kjeller (Norway)); Lauritzen, B. (Technical Univ. of Denmark, Risoe National Lab. for Sustainable Energy, Roskilde (Denmark))

    2012-01-15

    The supercritical water reactor (SCWR) is one of the six different reactor technologies selected for research and development under the Generation IV program. Several countries have shown interest to this concept but up to now, there exist no in-pile facilities to perform the required material and fuel tests. Working on this direction, the Halden Reactor Project has started an activity in collaboration with Risoe-DTU (with Mr. Rudi Van Nieuwenhove as the project leader) to study the feasibility of a SCW loop in the Halden Reactor, which is a Heavy Boiling Water Reactor (HBWR). The ultimate goal of the project is to design a loop allowing material and fuel test studies at significant mass flow with in-core instrumentation and chemistry control possibilities. The present report focusses on the main heat exchanger required for such a loop in the Halden Reactor. The goal of this heat exchanger is to assure a supercritical flow state inside the test section (the core side) and a subcritical flow state inside the pump section. The objective is to design the heat exchanger in order to optimize the efficiency of the heat transfer and to respect several requirements as the room available inside the reactor hall, the maximal total pressure drop allowed and so on. (Author)

  2. Electrically Heated Testing of the Kilowatt Reactor Using Stirling Technology (KRUSTY) Experiment Using a Depleted Uranium Core

    Science.gov (United States)

    Briggs, Maxwell H.; Gibson, Marc A.; Sanzi, James

    2017-01-01

    The Kilopower project aims to develop and demonstrate scalable fission-based power technology for systems capable of delivering 110 kW of electric power with a specific power ranging from 2.5 - 6.5 Wkg. This technology could enable high power science missions or could be used to provide surface power for manned missions to the Moon or Mars. NASA has partnered with the Department of Energys National Nuclear Security Administration, Los Alamos National Labs, and Y-12 National Security Complex to develop and test a prototypic reactor and power system using existing facilities and infrastructure. This technology demonstration, referred to as the Kilowatt Reactor Using Stirling TechnologY (KRUSTY), will undergo nuclear ground testing in the summer of 2017 at the Nevada Test Site. The 1 kWe variation of the Kilopower system was chosen for the KRUSTY demonstration. The concept for the 1 kWe flight system consist of a 4 kWt highly enriched Uranium-Molybdenum reactor operating at 800 degrees Celsius coupled to sodium heat pipes. The heat pipes deliver heat to the hot ends of eight 125 W Stirling convertors producing a net electrical output of 1 kW. Waste heat is rejected using titanium-water heat pipes coupled to carbon composite radiator panels. The KRUSTY test, based on this design, uses a prototypic highly enriched uranium-molybdenum core coupled to prototypic sodium heat pipes. The heat pipes transfer heat to two Advanced Stirling Convertors (ASC-E2s) and six thermal simulators, which simulate the thermal draw of full scale power conversion units. Thermal simulators and Stirling engines are gas cooled. The most recent project milestone was the completion of non-nuclear system level testing using an electrically heated depleted uranium (non-fissioning) reactor core simulator. System level testing at the Glenn Research Center (GRC) has validated performance predictions and has demonstrated system level operation and control in a test configuration that replicates the one

  3. 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

  4. Approximate Loop Unrolling

    OpenAIRE

    Rodriguez-Cancio, Marcelino; Combemale, Benoit; Baudry, Benoit

    2016-01-01

    We introduce Approximate Unrolling, a loop optimization that reduces execution time and energy consumption, exploiting the existence of code regions that can endure some degree of approximation while still producing acceptable results. This work focuses on a specific kind of forgiving region: counted loops that map a given functions over the elements of an array. Approximate Unrolling transforms loops in a similar way Loop Unrolling does. However, unlike its exact counterpart, our optimizatio...

  5. Mechanistic Model for Ash Deposit Formation in Biomass Suspension Firing. Part 1: Model Verification by Use of Entrained Flow Reactor Experiments

    DEFF Research Database (Denmark)

    Hansen, Stine Broholm; Jensen, Peter Arendt; Jappe Frandsen, Flemming

    2017-01-01

    stronger influence of this parameter. Model #2 was able to provide a reasonable description of the influence of temperature on the deposit buildup rates observed in the EFR experiments. A parametric study was conducted to examine the influence of some physical parameters, including ash concentration...... used to describe the deposit formation rates and deposit chemistry observed in a series of entrained flow reactor (EFR) experiments using straw and wood as fuels. It was found that model #1 was not able to describe the observed influence of temperature on the deposit buildup rates, predicting a much......, viscosity of ash and deposits, surface tension, Young’s modulus, and porosity. On the basis of this model evaluation, where a wide range of temperatures (700–1000 °C) and fuels (straw and wood) were applied, model #2 can be regarded as a promising tool for the description of deposit formation from biomass...

  6. Blind Loop Syndrome

    Science.gov (United States)

    ... breeding ground for bacteria. The bacteria may produce toxins as well as block the absorption of nutrients. The greater the length of small bowel involved in the blind loop, the greater the chance of bacterial overgrowth. What triggers blind loop syndrome? Blind loop ...

  7. Reanalysis of the Gas-cooled fast reactor experiments at the zero power facility Proteus - Spectral indices

    Science.gov (United States)

    Perret, G.; Pattupara, R. M.; Girardin, G.; Chawla, R.

    2013-03-01

    PROTEUS is a zero power reactor at the Paul Scherrer Institute which has been employed during the 1970's to study experimentally the physics of the gas-cooled fast reactor. Reaction rate distributions, flux spectrum and reactivity effects have been measured in several configurations featuring PuO2/UO2 fuel, absorbers, large iron shields, and thorium oxide and thorium metal fuel either distributed quasihomogeneously in the reference PuO2/UO2 lattice or introduced in the form of radial and axial blanket zones. This papers focus on the spectral indices - including fission and capture in 232Th and 237Np - measured in the reference PuO2/UO2 lattices and their predictions with an MCNPX model specially developed for the PROTEUS-GCFR core. Predictions were obtained with JEFF-3.1 and -3.11, ENDF/B-VII.0 and VII.1, and JENDL-3.3 and -4.0. A general good agreement was demonstrated. The ratio of 232Th fission to 239Pu fission, however, was under-predicted by 8.7±2.1% and 6.5±2.1% using ENDF/B-VII.0 and VII.1, respectively. Finally, the capture rates in 237Np tended to be underpredicted by the JEFF and JENDL libraries, although the new cross section in JEFF-3.1.1 slightly improved the 237Np capture to 239Pu fission results (3.4±2.4%).

  8. Droplet flows through periodic loop networks

    Science.gov (United States)

    Jeanneret, Raphael; Schindler, Michael; Bartolo, Denis

    2010-11-01

    Numerous microfluidic experiments have revealed non-trivial traffic dynamics when droplets flow through a channel including a single loop. A complex encoding of the time intervals between the droplets is achieved by the binary choices they make as they enter the loop. Very surprisingly, another set of experiments has demonstrated that the addition of a second loop does not increase the complexity of the droplet pattern. Conversely, the second loop decodes the temporal signal encrypted by the first loop [1]. In this talk we show that no first principle argument based on symmetry or conservation laws can account for this unexpected decoding process. Then, to better understand how a loop maps time intervals between droplets, we consider a simplified model which has proven to describe accurately microfluidic droplet flows. Combining numerical simulations and analytical calculations for the dynamic of three droplets travelling through N loops: (i) We show that three different traffic regimes exist, yet none of them yields exact decoding. (ii) We uncover that for a wide class of loop geometry, the coding process is analogous to a Hamiltonian mapping: regular orbits are destabilized in island chains and separatrix. (iii) Eventually, we propose a simple explanation to solve the apparent paradox with the coding/decoding dynamics observed in experiments. [1] M.J. Fuerstman, P. Garstecki, and G.M. Whitesides, Science, 315:828, 2007.

  9. Report detailing comparative analysis of results from high flux isotope reactor and national institute of standards technology small-angle neutron scattering experiments

    Energy Technology Data Exchange (ETDEWEB)

    Sokolov, Mikhail A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Littrell, Ken [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wells, Peter [Univ. of California, Santa Barbara, CA (United States); Cunningham, Nicholas J. [Univ. of California, Santa Barbara, CA (United States)

    2015-09-01

    The major issues regarding irradiation effects are discussed in [1-3] and have also been discussed in previous progress and milestone reports. As noted previously, of the many significant issues discussed, the issue considered to have the most impact on the current regulatory process is that associated with effects of neutron irradiation on RPV steels at high fluence, for long irradiation times, and as affected by neutron flux. It is clear that embrittlement of RPV steels is a critical issue that may limit LWR plant life extension. The primary objective of the LWRSP RPV task is to develop robust predictions of transition temperature shifts (TTS) at high fluence ( t) to at least 1020 n/cm2 (>1 MeV) pertinent to plant operation of some pressurized water reactors (PWR) for 80 full power years. Correlations between the high flux test reactor results and low flux surveillance specimens must be established for proper RPV embrittlement predictions of the current nuclear power fleet. Additionally, a complete understanding of defect evolution for high nickel RPV steels is needed to characterize the embrittlement potential of Mn-Ni-enriched precipitates (MNPs), particularly for the high fluence regime. While understanding of copper-enriched precipitates (CRPs) have been fully developed, the recent discovery and experimental verification [4] of late blooming MNPs with little to no copper for nucleation has stimulated research efforts to understand the evolution of these phases. New and existing databases will be combined to support developing physically based models of TTS for high fluence-low flux ( < 10 11n/cm2-s) conditions, beyond the existing surveillance database, to neutron fluences of at least 1 1020 n/cm2 (>1 MeV). Moreover, large number of various RPV materials have been irradiated in ATR-2 experiment and will be jointly studied by University of California Santa Barbara (UCSB) and ORNL to address majority of microstructural characteristics

  10. Report of blind start-up experiments carried out on the reactor Cabri between 4. and 8. July 1966; Compte rendu des experiences de demarrage en aveugle effectuees sur le reacteur cabri du 4 au 8 juillet 1966

    Energy Technology Data Exchange (ETDEWEB)

    Filipczak, N.; Filipczak, W. [Institut Basan Jadrowych Swerk pta, Otwock (Poland); Furet, J.; Kaiser, J. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1967-07-01

    The blind start-up of a reactor without any neutronic data concerning a relatively wide range of power dynamics can be necessary when difficulties arise in the positioning of the detector or in neutron-gamma discrimination near the multiplying medium. The object of the experiments carried out on the reactor Cabri was to check the very complete analysis of the start-up accident which was studied on an analogue computer. The number of experiments carried out (12) is not sufficient to allow a definite conclusion. Nevertheless the blind start-up method advocated by N. FILIPCZAK and W. FILIPCZAK does not appear to be incompatible with the security during the operational phase (on condition that its dynamic characteristics are close to that of the reactor Cabri). (authors) [French] Le demarrage en aveugle d'un reacteur sans aucune information neutronique sur une dynamique de puissance relativement etendue peut se presenter lorsqu'il y a des difficultes de positionnement de detecteur ou de discrimination neutrons-gamma a proximite du milieu multiplicateur. Les experiences effectuees sur le reacteur CABRI avaient pour but de verifier l'analyse tres poussee de l'accident de demarrage etudie sur calculateur analogique. Le nombre d'experiences effectuees (12) n'est pas suffisant pour tirer des conclusions bien nettes. Neanmoins la methode de demarrage aveugle preconisee par N. PILIPCZAK et W. FILIPCZAK ne semble pas incompatible avec la securite pendant la phase d'exploitation du reacteur (a condition que ses caracteristiques dynamiques soient voisines de celles du reacteur CABRI). (auteurs)

  11. Depth distribution of Frank loop defects formed in ion-irradiated stainless steel and its dependence on Si addition

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Dongyue, E-mail: dychen@safety.n.t.u-tokyo.ac.jp [The University of Tokyo, Department of Nuclear Engineering and Management, School of Engineering, 7-3-1 Hongo Bunkyo-ku, Tokyo 113-8656 (Japan); Murakami, Kenta [The University of Tokyo, Nuclear Professional School, School of Engineering, 2-22 Shirakata-Shirane, Tokai-mura, Ibaraki 319-1188 (Japan); Dohi, Kenji; Nishida, Kenji; Soneda, Naoki [Central Research Institute of Electric Power Industry, 2-11-1 Iwado-kita, Komae, Tokyo 201-8511 (Japan); Li, Zhengcao, E-mail: zcli@tsinghua.edu.cn [Tsinghua University, School of Materials Science and Engineering, Beijing 100084 (China); Liu, Li; Sekimura, Naoto [The University of Tokyo, Department of Nuclear Engineering and Management, School of Engineering, 7-3-1 Hongo Bunkyo-ku, Tokyo 113-8656 (Japan)

    2015-12-15

    Although heavy ion irradiation is a good tool to simulate neutron irradiation-induced damages in light water reactor, it produces inhomogeneous defect distribution. Such difference in defect distribution brings difficulty in comparing the microstructure evolution and mechanical degradation between neutron and heavy ion irradiation, and thus needs to be understood. Stainless steel is the typical structural material used in reactor core, and could be taken as an example to study the inhomogeneous defect depth distribution in heavy ion irradiation and its influence on the tested irradiation hardening by nano-indentation. In this work, solution annealed stainless steel model alloys are irradiated by 3 MeV Fe{sup 2+} ions at 400 °C to 3 dpa to produce Frank loops that are mainly interstitial in nature. The silicon content of the model alloys is also tuned to change point defect diffusion, so that the loop depth distribution influenced by diffusion along the irradiation beam direction could be discussed. Results show that in low Si (0% Si) and base Si (0.42% Si) samples the depth distribution of Frank loop density quite well matches the dpa profile calculated by the SRIM code, but in high Si sample (0.95% Si), the loop number density in the near-surface region is very low. One possible explanation could be Si’s role in enhancing the effective vacancy diffusivity, promoting recombination and thus suppressing interstitial Frank loops, especially in the near-surface region, where vacancies concentrate. By considering the loop depth distribution, the tested irradiation hardening is successfully explained by the Orowan model. A hardening coefficient of around 0.30 is obtained for all the three samples. This attempt in interpreting hardening results may make it easier to compare the mechanical degradation between different irradiation experiments.

  12. Steady-state CFD simulations of an EPR™ reactor pressure vessel: A validation study based on the JULIETTE experiments

    Energy Technology Data Exchange (ETDEWEB)

    Puragliesi, R., E-mail: riccardo.puragliesi@psi.ch [Laboratory for Reactor Physics and Systems Behaviour, PSI, 5232 Villigen (Switzerland); Zhou, L. [Science and Technology on Reactor System Design Technology Laboratory, NPIC, Chengdu (China); Zerkak, O.; Pautz, A. [Laboratory for Reactor Physics and Systems Behaviour, PSI, 5232 Villigen (Switzerland)

    2016-04-15

    Highlights: • CFD validation of k–ε (RANS model of EPR RPV. • Flat inlet velocity profile is not sufficient to correctly predict the pressure drops. • Swirl is responsible for asymmetric loads at the core barrel. • Parametric study to the turbulent Schmidt number for better predictions of passive-scalar transport. • The optimal turbulent Schmidt number was found to be one order of magnitude smaller than the standard value. - Abstract: Validating computational fluid dynamics (CFD) models against experimental measurements is a fundamental step towards a broader acceptance of CFD as a tool for reactor safety analysis when best-estimate one-dimensional thermal-hydraulic codes present strong modelling limitations. In the present paper numerical results of steady-state RANS analyses are compared to pressure, volumetric flow rate and concentration distribution measurements in different locations of an Areva EPR™ reactor pressure vessel (RPV) mock-up named JULIETTE. Several flow configurations are considered: Three different total volumetric flow rates, cold leg velocity field with or without swirl, three or four reactor coolant pumps functioning. Investigations on the influence of two types of inlet boundary profiles (i.e. flat or 1/7th power-law) and the turbulent Schmidt number have shown that the first affects sensibly the pressure loads at the core barrel whereas the latter parameter strongly affects the transport and the mixing of the tracer (passive scalar) and consequently its distribution at the core inlet. Furthermore, the introduction of an integral parameter as the swirl number has helped to decrease the large epistemic uncertainty associated with the swirling device. The swirl is found to be the cause of asymmetric loads on the walls of the core barrel and also asymmetries are enhanced for the tracer concentration distribution at the core inlet. The k–ϵ CFD model developed with the commercial code STAR-CCM+ proves to be able to predict

  13. Reactor Simulator Testing

    Science.gov (United States)

    Schoenfeld, Michael P.; Webster, Kenny L.; Pearson, Boise J.

    2013-01-01

    As part of the Nuclear Systems Office Fission Surface Power Technology Demonstration Unit (TDU) project, a reactor simulator test loop (RxSim) was design & built to perform integrated testing of the TDU components. In particular, the objectives of RxSim testing was to verify the operation of the core simulator, the instrumentation and control system, and the ground support gas and vacuum test equipment. In addition, it was decided to include a thermal test of a cold trap purification design and a pump performance test at pump voltages up to 150 V since the targeted mass flow rate of 1.75 kg/s was not obtained in the RxSim at the originally constrained voltage of 120 V. This paper summarizes RxSim testing. The gas and vacuum ground support test equipment performed effectively in NaK fill, loop pressurization, and NaK drain operations. The instrumentation and control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings. The cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained which was lower than the predicted 750 K but 156 K higher than the cold temperature indicating the design provided some heat regeneration. The annular linear induction pump (ALIP) tested was able to produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz.

  14. Reanalysis of the Gas-cooled fast reactor experiments at the zero power facility Proteus – Spectral indices

    Directory of Open Access Journals (Sweden)

    Girardin G.

    2013-03-01

    Full Text Available PROTEUS is a zero power reactor at the Paul Scherrer Institute which has been employed during the 1970’s to study experimentally the physics of the gas-cooled fast reactor. Reaction rate distributions, flux spectrum and reactivity effects have been measured in several configurations featuring PuO2/UO2 fuel, absorbers, large iron shields, and thorium oxide and thorium metal fuel either distributed quasihomogeneously in the reference PuO2/UO2 lattice or introduced in the form of radial and axial blanket zones. This papers focus on the spectral indices – including fission and capture in 232Th and 237Np - measured in the reference PuO2/UO2 lattices and their predictions with an MCNPX model specially developed for the PROTEUS-GCFR core. Predictions were obtained with JEFF-3.1 and -3.11, ENDF/B-VII.0 and VII.1, and JENDL-3.3 and -4.0. A general good agreement was demonstrated. The ratio of 232Th fission to 239Pu fission, however, was under-predicted by 8.7±2.1% and 6.5±2.1% using ENDF/B-VII.0 and VII.1, respectively. Finally, the capture rates in 237Np tended to be underpredicted by the JEFF and JENDL libraries, although the new cross section in JEFF-3.1.1 slightly improved the 237Np capture to 239Pu fission results (3.4±2.4%.

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

    Energy Technology Data Exchange (ETDEWEB)

    1976-07-01

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

  16. Experience feedback examination in PWR type reactors operating for the 1997-1999 period; Examen du retour d'experience en exploitation des reacteurs a eau sous pression pour la periode 1997-1999

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2004-07-01

    The present report is relative to the examination that the permanent group has made on the experience feedback in operation for PWR type reactors for the period 1997-1999 that was on eleven themes chosen by the Nuclear Safety and Radiation Protection Authority. It used analysis reports made by I.R.S.N. in support of four meetings of the permanent group devoted to this examination from April 2001 to June 2002. The different themes were operating uncertainties, machining to vibrations, analysis of incidents and gaseous releases, circuits, human factors, behaviour of electric batteries, risk of cold source loss. (N.C.)

  17. NUCLEAR REACTOR

    Science.gov (United States)

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

    1958-01-21

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

  18. 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.

  19. Behavior of deuterium and rare gases in thermal convection loops with molten Pb-17Li

    Energy Technology Data Exchange (ETDEWEB)

    Feuerstein, H.; Graebner, H.; Horn, S.; Oschinski, J. (Kernforschungszentrum Karlsruhe GmbH (Germany, F.R.))

    1991-04-01

    The behaviour of deuterium in thermal convection loops with molten Pb-17Li{sup *} was investigated in the temperature range from 300 to 610deg C, and in the range of deuterium partial pressures from 0.05 to 1000 mbar. Dissolution and desorption are controlled by diffusion through a 0.002 mm thick LM boundary layer at the interface, no chemical reactions are involved in the rate determining step. This boundary layer is also effective in case of permeation through membranes, if one side is covered by the LM. The permeation through 0.6 mm iron was reduced by a factor of 100. However in case of a fusion reactor blanket this boundary layer will not be important, because the wall thickness of the components is much larger. For the 2 mm stainless steel of the thermal convection loops with a downstream oxide layer, no effect of the boundary layer could be seen. The amount of oxides in the loop had no influence on the results. Furthermore an excess of H{sub 2} at low pressures did not change transfer rates of deuterium. The solubility of deuterium in the LM was determined from the kinetics of loading and degassing. The found values are one order of magnitude smaller than the lowest values so far published. The transport behavior of the rare gases He, Ne, Ar, Kr and Xe was investigated. The solubility of helium was found five orders of magnitude lower than that of deuterium, those for Ne, Ar, Kr and Xe even lower than that for helium. Helium-bubble formation has to be considered if the flow rate of the LM in a blanket is small, or in case of static irradiation experiments. On the other hand argon can be used as covergas for a fusion reactor blanket. Because of the low solubility in the LM, the Ar-41 activity will be much smaller than in sodium cooled reactors. (orig.).

  20. Gas Test Loop Booster Fuel Hydraulic Testing

    Energy Technology Data Exchange (ETDEWEB)

    Gas Test Loop Hydraulic Testing Staff

    2006-09-01

    The Gas Test Loop (GTL) project is for the design of an adaptation to the Advanced Test Reactor (ATR) to create a fast-flux test space where fuels and materials for advanced reactor concepts can undergo irradiation testing. Incident to that design, it was found necessary to make use of special booster fuel to enhance the neutron flux in the reactor lobe in which the Gas Test Loop will be installed. Because the booster fuel is of a different composition and configuration from standard ATR fuel, it is necessary to qualify the booster fuel for use in the ATR. Part of that qualification is the determination that required thermal hydraulic criteria will be met under routine operation and under selected accident scenarios. The Hydraulic Testing task in the GTL project facilitates that determination by measuring flow coefficients (pressure drops) over various regions of the booster fuel over a range of primary coolant flow rates. A high-fidelity model of the NW lobe of the ATR with associated flow baffle, in-pile-tube, and below-core flow channels was designed, constructed and located in the Idaho State University Thermal Fluids Laboratory. A circulation loop was designed and constructed by the university to provide reactor-relevant water flow rates to the test system. Models of the four booster fuel elements required for GTL operation were fabricated from aluminum (no uranium or means of heating) and placed in the flow channel. One of these was instrumented with Pitot tubes to measure flow velocities in the channels between the three booster fuel plates and between the innermost and outermost plates and the side walls of the flow annulus. Flow coefficients in the range of 4 to 6.5 were determined from the measurements made for the upper and middle parts of the booster fuel elements. The flow coefficient for the lower end of the booster fuel and the sub-core flow channel was lower at 2.3.

  1. Multi-physic simulations of irradiation experiments in a technological irradiation reactor; Modelisation pluridisciplinaire d'experiences d'irradiation dans un reacteur d'irradiation technologique

    Energy Technology Data Exchange (ETDEWEB)

    Bonaccorsi, Th

    2007-09-15

    A Material Testing Reactor (MTR) makes it possible to irradiate material samples under intense neutron and photonic fluxes. These experiments are carried out in experimental devices localised in the reactor core or in periphery (reflector). Available physics simulation tools only treat, most of the time, one physics field in a very precise way. Multi-physic simulations of irradiation experiments therefore require a sequential use of several calculation codes and data exchanges between these codes: this corresponds to problems coupling. In order to facilitate multi-physic simulations, this thesis sets up a data model based on data-processing objects, called Technological Entities. This data model is common to all of the physics fields. It permits defining the geometry of an irradiation device in a parametric way and to associate information about materials to it. Numerical simulations are encapsulated into interfaces providing the ability to call specific functionalities with the same command (to initialize data, to launch calculations, to post-treat, to get results,... ). Thus, once encapsulated, numerical simulations can be re-used for various studies. This data model is developed in a SALOME platform component. The first application case made it possible to perform neutronic simulations (OSIRIS reactor and RJH) coupled with fuel behavior simulations. In a next step, thermal hydraulics could also be taken into account. In addition to the improvement of the calculation accuracy due to the physical phenomena coupling, the time spent in the development phase of the simulation is largely reduced and the possibilities of uncertainty treatment are under consideration. (author)

  2. Numerical simulation of a natural circulation loop

    Energy Technology Data Exchange (ETDEWEB)

    Verissimo, Gabriel L.; Moreira, Maria de Lourdes; Faccini, Jose Luiz H., E-mail: gabrielverissimo@poli.ufrj.b, E-mail: malu@ien.gov.b, E-mail: faccini@ien.gov.b [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    This work presents a numerical simulation of a natural circulation loop using computational fluid dynamics. The simulated loop is an experimental model in a reduced scale of 1:10 of a passive heat removal system typical of advanced PWR reactors. The loop is composed of a heating vessel containing 52 electric heaters, a vertical shell-tube heat exchanger and a column of expansion. The working fluid is distilled water. Initially it was created a tridimensional geometric model of the loop components. After that, it was generated a tridimensional mesh of finite elements in order to calculate the variables of the problem. The boundaries of the numerical simulation were the power of the electric resistances and the cooling flow in the secondary side of the heat exchanger. The initial conditions were the temperature, the pressure and the fluid velocity at the time just before the power has been switched on. The results of this simulation were compared with the experimental data, in terms of the evolution of the temperatures in different locations of the loop, and of the average natural circulation flow as a function of time for a given power. (author)

  3. Osmotic mechanism of the loop extrusion process

    Science.gov (United States)

    Yamamoto, Tetsuya; Schiessel, Helmut

    2017-09-01

    The loop extrusion theory assumes that protein factors, such as cohesin rings, act as molecular motors that extrude chromatin loops. However, recent single molecule experiments have shown that cohesin does not show motor activity. To predict the physical mechanism involved in loop extrusion, we here theoretically analyze the dynamics of cohesin rings on a loop, where a cohesin loader is in the middle and unloaders at the ends. Cohesin monomers bind to the loader rather frequently and cohesin dimers bind to this site only occasionally. Our theory predicts that a cohesin dimer extrudes loops by the osmotic pressure of cohesin monomers on the chromatin fiber between the two connected rings. With this mechanism, the frequency of the interactions between chromatin segments depends on the loading and unloading rates of dimers at the corresponding sites.

  4. Thermal-hydraulic study of the LBE-cooled fuel assembly in the MYRRHA reactor: Experiments and simulations

    Energy Technology Data Exchange (ETDEWEB)

    Pacio, J., E-mail: Julio.pacio@kit.edu [Karlsruhe Institute of Technology (KIT), Institute for Nuclear and Energy Technologies (IKET), Hermann-von-Helmholtz Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Wetzel, T. [Karlsruhe Institute of Technology (KIT), Institute for Nuclear and Energy Technologies (IKET), Hermann-von-Helmholtz Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Doolaard, H.; Roelofs, F. [Nuclear Research and Consultancy Group (NRG), Westerduinweg 3, 1755 ZG Petten (Netherlands); Van Tichelen, K. [Belgian Nuclear Reseach Center (SCK-CEN), Boeretang 200, Mol (Belgium)

    2017-02-15

    Heavy liquid metals (HLMs), such as lead-bismuth eutectic (LBE) and pure lead are prominent candidate coolants for many advanced systems based on fast neutrons. In particular, LBE is used in the first-of-its-kind MYRRHA fast reactor, to be built in Mol (Belgium), which can be operated either in critical mode or as a sub-critical accelerator-driven system. With a strong focus on safety, key thermal-hydraulic aspects of these systems, such as the proper cooling of fuel assemblies, must be assessed. Considering the complex geometry and low Prandtl number of LBE (Pr ∼ 0.025), this flow scenario is challenging for the models used in Computational Fluid Dynamics (CFD), e.g. for relating the turbulent transport of momentum and heat. Thus, reliable experimental data for the relevant scenario are needed for validation. In this general context, this topic is studied both experimentally and numerically in the framework of the European FP7 project SEARCH (2011–2015). An experimental campaign, including a 19-rod bundle with wire spacers, cooled by LBE is undertaken at KIT. With prototypical geometry and operating conditions, it is intended to evaluate the validity of current empirical correlations for the MYRRHA conditions and, at the same time, to provide validation data for the CFD simulations performed at NRG. The results of one benchmarking case are presented in this work. Moreover, this validated approach is then used for simulating a complete MYRRHA fuel assembly (127 rods).

  5. Preparation of actinide specimens for the US/UK joint experiment in the Dounreay Prototype Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Quinby, T C; Adair, H L; Kobisk, E H

    1982-05-01

    A joint research program involving the United States and the United Kingdom was initiated about four years ago for the purpose of studying the fuel behavior of higher actinides using in-core irradiation in the fast reactor at Dounreay, Scotland. Simultaneously, determination of integral cross sections of a wide variety of higher actinide isotopes (physics specimens) was proposed. Coincidental neutron flux and energy spectral measurements were to be made using vanadium encapsulated dosimetry materials in the immediate region of the fuel pellets and physics samples. The higher actinide samples chosen for the fuel study were /sup 241/Am and /sup 244/Cm in the forms of Am/sub 2/O/sub 3/, Cm/sub 2/O/sub 3/, and Am/sub 6/Cm(RE)/sub 7/O/sub 21/, where (RE) represents a mixture of lanthanides. Milligram quantities of actinide oxides of /sup 248/Cm, /sup 246/Cm, /sup 244/Cm, /sup 243/Cm, /sup 243/Am, /sup 241/Am, /sup 244/Pu, /sup 242/Pu, /sup 241/Pu, /sup 240/Pu, /sup 239/Pu, /sup 238/Pu, /sup 237/Np, /sup 238/U, /sup 236/U, /sup 235/U, /sup 234/U, /sup 233/U, /sup 232/Th, /sup 230/Th, and /sup 231/Pa were encapsulated to obtain nuclear cross section and reaction rate data for these materials.

  6. Reactor monitoring and safeguards using antineutrino detectors

    CERN Document Server

    Bowden, N S

    2008-01-01

    Nuclear reactors have served as the antineutrino source for many fundamental physics experiments. The techniques developed by these experiments make it possible to use these very weakly interacting particles for a practical purpose. The large flux of antineutrinos that leaves a reactor carries information about two quantities of interest for safeguards: the reactor power and fissile inventory. Measurements made with antineutrino detectors could therefore offer an alternative means for verifying the power history and fissile inventory of a reactors, as part of International Atomic Energy Agency (IAEA) and other reactor safeguards regimes. Several efforts to develop this monitoring technique are underway across the globe.

  7. 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.

  8. 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.

  9. Two-phase flow experiments on Counter-Current Flow Limitation in a model of the hot leg of a pressurized water reactor (2015 test series)

    Energy Technology Data Exchange (ETDEWEB)

    Beyer, Matthias; Lucas, Dirk; Pietruske, Heiko; Szalinski, Lutz

    2016-12-15

    Counter-Current Flow Limitation (CCFL) is of importance for PWR safety analyses in several accident scenarios connected with loss of coolant. Basing on the experiences obtained during a first series of hot leg tests now new experiments on counter-current flow limitation were conducted in the TOPFLOW pressure vessel. The test series comprises air-water tests at 1 and 2 bar as well as steam-water tests at 10, 25 and 50 bar. During the experiments the flow structure was observed along the hot leg model using a high-speed camera and web-cams. In addition pressure was measured at several positions along the horizontal part and the water levels in the reactor-simulator and steam-generator-simulator tanks were determined. This report documents the experimental setup including the description of operational and special measuring techniques, the experimental procedure and the data obtained. From these data flooding curves were obtained basing on the Wallis parameter. The results show a slight shift of the curves in dependency of the pressure. In addition a slight decrease of the slope was found with increasing pressure. Additional investigations concern the effects of hysteresis and the frequencies of liquid slugs. The latter ones show a dependency on pressure and the mass flow rate of the injected water. The data are available for CFD-model development and validation.

  10. Short Baseline Reactor Antineutrino-Electron Scattering Experiments and Non-Standard Neutrino Interactions at Source and Detector

    OpenAIRE

    Khan, Amir N.; McKay, Douglas W.; Tahir, F

    2014-01-01

    We investigate non-standard interaction effects in antineutrino-electron scattering experiments with baselines short enough to ignore standard oscillation phenomena. The setup is free of ambiguities from the interference between new physics and oscillation effects and is sensitive to both semileptonic new physics at the source and purely leptonic new physics in the weak interaction scattering at the detector. We draw on the TEXONO experiment as the model system, extending its analysis of non-...

  11. Model predictive control of a solar-thermal reactor

    Science.gov (United States)

    Saade Saade, Maria Elizabeth

    Solar-thermal reactors represent a promising alternative to fossil fuels because they can harvest solar energy and transform it into storable and transportable fuels. The operation of solar-thermal reactors is restricted by the available sunlight and its inherently transient behavior, which affects the performance of the reactors and limits their efficiency. Before solar-thermal reactors can become commercially viable, they need to be able to maintain a continuous high-performance operation, even in the presence of passing clouds. A well-designed control system can preserve product quality and maintain stable product compositions, resulting in a more efficient and cost-effective operation, which can ultimately lead to scale-up and commercialization of solar thermochemical technologies. In this work, we propose a model predictive control (MPC) system for a solar-thermal reactor for the steam-gasification of biomass. The proposed controller aims at rejecting the disturbances in solar irradiation caused by the presence of clouds. A first-principles dynamic model of the process was developed. The model was used to study the dynamic responses of the process variables and to identify a linear time-invariant model used in the MPC algorithm. To provide an estimation of the disturbances for the control algorithm, a one-minute-ahead direct normal irradiance (DNI) predictor was developed. The proposed predictor utilizes information obtained through the analysis of sky images, in combination with current atmospheric measurements, to produce the DNI forecast. In the end, a robust controller was designed capable of rejecting disturbances within the operating region. Extensive simulation experiments showed that the controller outperforms a finely-tuned multi-loop feedback control strategy. The results obtained suggest that our controller is suitable for practical implementation.

  12. 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.

  13. Numerical analysis of the fluid dynamics in a natural circulation loop; Analise numerica da dinamica do escoamento em circuitos de circulacao natural

    Energy Technology Data Exchange (ETDEWEB)

    Angelo, Gabriel

    2013-07-01

    Natural circulation loops apply to many engineering applications such as: water heating solar energy system (thermo-siphons), thermal management of electrical components (voltage converter), geothermal energy, nuclear reactors, etc. In pressurized water nuclear reactors, known as PWR's, the natural circulation loops are employed to ensure passive safety. In critical situations, the heat transfer will occur only by natural convection, without any external control or mechanical devices. This feature is desired and has been considered in modern nuclear reactor projects. This work consists of a numerical study of the natural circulation loop, located at the Instituto de Pesquisas Energeticas e Nucleares / Comissao Nacional de Energia Nuclear in Sao Paulo, Brazil, in order to establish the flow pattern in single phase conditions. The comparison of numerical results to experiments in transient condition revealed significant deviations for the Zero Equation turbulence model. Intermediate deviations for the Eddy Viscosity Turbulence Equation (EVTE), k - {omega}, SST e SSG models. And the best results are obtained by the k - {epsilon} e DES models (with better results for the k - {epsilon} model). (author)

  14. Modelling and characterization of an airlift-loop bioreactor

    NARCIS (Netherlands)

    Verlaan, P.

    1987-01-01

    An airlift-loop reactor is a bioreactor for aerobic biotechnological processes. The special feature of the ALR is the recirculation of the liquid through a downcomer connecting the top and the bottom of the main bubbling section. Due to the high circulation-flow rate, efficient mixing and

  15. The Tokamak Fusion Test Reactor decontamination and decommissioning project and the Tokamak Physics Experiment at the Princeton Plasma Physics Laboratory. Environmental Assessment

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-05-27

    If the US is to meet the energy needs of the future, it is essential that new technologies emerge to compensate for dwindling supplies of fossil fuels and the eventual depletion of fissionable uranium used in present-day nuclear reactors. Fusion energy has the potential to become a major source of energy for the future. Power from fusion energy would provide a substantially reduced environmental impact as compared with other forms of energy generation. Since fusion utilizes no fossil fuels, there would be no release of chemical combustion products to the atmosphere. Additionally, there are no fission products formed to present handling and disposal problems, and runaway fuel reactions are impossible due to the small amounts of deuterium and tritium present. The purpose of the TPX Project is to support the development of the physics and technology to extend tokamak operation into the continuously operating (steady-state) regime, and to demonstrate advances in fundamental tokamak performance. The purpose of TFTR D&D is to ensure compliance with DOE Order 5820.2A ``Radioactive Waste Management`` and to remove environmental and health hazards posed by the TFTR in a non-operational mode. There are two proposed actions evaluated in this environmental assessment (EA). The actions are related because one must take place before the other can proceed. The proposed actions assessed in this EA are: the decontamination and decommissioning (D&D) of the Tokamak Fusion Test Reactor (TFTR); to be followed by the construction and operation of the Tokamak Physics Experiment (TPX). Both of these proposed actions would take place primarily within the TFTR Test Cell Complex at the Princeton Plasma Physics Laboratory (PPPL). The TFTR is located on ``D-site`` at the James Forrestal Campus of Princeton University in Plainsboro Township, Middlesex County, New Jersey, and is operated by PPPL under contract with the United States Department of Energy (DOE).

  16. System design description for GCFR-core flow test loop

    Energy Technology Data Exchange (ETDEWEB)

    Huntley, W.R.; Grindell, A.G.

    1980-12-01

    The Core Flow Test Loop is a high-pressure, high-temperature, out-of-reactor helium circulation system that is being constructed to permit detailed study of the thermomechanical and thermal performance at prototypic steady-state and transient operating conditions of simulated segments of core assemblies for a GCFR Demonstration Plant, as designed by General Atomic Company. It will also permit the expermental verification of predictive analytical models of the GCFR core assemblies needed to reduce operational and safety uncertainties of the GCFR. Full-sized blanket assemblies and segments of fuel rod and control rod fuel assemblies will be simulated with test bundles of electrically powered fuel rod or blanket rod simulators. The loop will provide the steady-state and margin test requirements of bundle power and heat removal, and of helium coolant flow rate, pressure, and temperature for test bundles having up to 91 rods; these requirements set the maximum power, coolant helium flow, and thermal requirements for the loop. However, the size of the test vessel that contains the test bundles will be determined by the bundles that simulate a full-sized GCFR blanket assembly. The loop will also provide for power and coolant transients to simulate transient operation of GCFR core assemblies, including the capability for rapid helium depressurization to simulate the depressurization class of GCFR accidents. In addition, the loop can be used as an out-of-reactor test bed for characterizing in-reactor test bundle configurations.

  17. Reactor lifetime under control: the contribution of the VENUS neutron dosimetry programme

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, H.; Minsart, G.; Dhondt, P. [SCK-CEN, Boeretang 200, B-2400 Mol (Belgium)

    1998-07-01

    Lifetime management of PWR reactors is an important stake concerning safety and competitiveness of these electricity production systems. The embrittlement of the pressure vessel is due to neutron exposure is the main limiting phenomenon of the PWR lifetime. Therefore, an accurate assessment of the pressure vessel fast fluency contributes deeply to a better management of the reactor lifetime. The Light Water Reactor Pressure Vessel Surveillance Dosimetry Improvement Program (LWR-PV-SDIP) was established in 1977 and sponsored by the US NRC. Its objectives were: improve, maintain and standardize neutron dosimetry, damage correlation and associated reactor analysis procedures used for predicting integrated effects of neutron exposure on LWR-PV. The VENUS PWR-Engineering Mock-up Experiment is part of the Belgian PWR- Pressure Vessel Surveillance Programme sponsored by the Belgian utilities. It contributed also, as one of the benchmark fields to the LWR-PV-SDIP, the VENUS mock-ups simulate the reflector geometry and the core boundary shape of a generic 3-loop PWR reactor. In this paper we will report on the three configurations considered in this programme namely: VENUS-1 simulating a PWR fresh core loading, VENUS-2 simulating a low-Ieakage core loading, and VENUS-3 simulating the PLSA (Partial Length Shielded Assembly) concept. Thanks to this programme, the PWR PV fast fluency can be assessed with an uncertainty ranging between 15 and 20%. (author)

  18. The Analysis of Loop Seal Purge Time for the KHNP Pressurizer Safety Valve Test Facility Using the GOTHIC Code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Ae; Kim, Chang Hyun; Kweon, Gab Joo; Park, Jong Woon [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of)

    2007-10-15

    The pressurizer safety valves (PSV) in Pressurized Water Reactors are required to provide the overpressure protection for the Reactor Coolant System (RCS) during the overpressure transients. Korea Hydro and Nuclear Power Company (KHNP) plans to build the PSV test facility for the purpose of providing the PSV pop-up characteristics and the loop seal dynamics for the new safety analysis. When the pressurizer safety valve is mounted in a loop seal configuration, the valve must initially pass the loop seal water prior to popping open on steam. The loop seal in the upstream of PSV prevents leakage of hydrogen gas or steam through the safety valve seat. This paper studies on the loop seal clearing dynamics using GOTHIC-7.2a code to verify the effects of loop seal purge time on the reactor coolant system overpressure.

  19. 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.

  20. Status of ITER task T213 collaborative irradiation screening experiment on Cu/SS joints in the Russian Federation SM-2-reactor

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, D.J. [Pacific Northwest National Lab., Richland, WA (United States); Fabritsiev, S.A. [D.V. Efremov Inst., St. Petersburg (Russian Federation); Pokrovsky, A.S. [SRIAR, Dimitrovgrad (Russian Federation); Zinkle, S.J. [Oak Ridge National Lab., TN (United States)] [and others

    1996-04-01

    Specimen fabrication is underway for an irradiation screening experiment planned to start in January 1996 in the SM-2 reactor in Dimitrovgrad, Russia. The purpose of the experiment is to evaluate the effects of neutron irradiation at ITER-relevant temperatures on the bond integrity performance of Cu/SS and Be/Cu joints, as well as to further investigate the base metal properties of irradiated copper alloys. Specimens from each of the four ITER parties (U.S., EU, japan, and RF) will be irradiated to a dose of {approx}0.2 dpa at two different temperatures, 150 and 300{degrees}C. The specimens will consist of Cu/SS and Be/Cu joints in several different geometries, as well as a large number of specimens from the base materials. Fracture toughness data on base metal and Cu/SS bonded specimens will be obtained from specimens supplied by the U.S. Due to lack of material, the Be/Cu specimens supplied by the U.S will only be irradiated as TEM disks.

  1. Testing the {rho}* scaling of thermal transport models: predicted and measured temperatures in the Tokamak Fusion Test Reactor dimensionless scaling experiments

    Energy Technology Data Exchange (ETDEWEB)

    Mikkelsen, D.R.; Scott, S.D. [Princeton Univ., NJ (United States). Plasma Physics Lab.; Dorland, W. [Texas Univ., Austin, TX (United States). Inst. for Fusion Studies

    1997-04-01

    Theoretical predictions of ion and electron thermal diffusivities are tested by comparing calculated and measured temperatures in low (L) mode plasmas from the Tokamak Fusion Test Reactor [D. J. Grove and D. M. Meade, Nucl. Fusion 25 , 1167 (1985)] nondimensional scaling experiments. The DIII-D [J. L. Luxon and L. G. Davis, Fusion Technol. 8 , 441 (1985)] L-mode {rho}* scalings, the transport models of Rebut-Lallia-Watkins (RLW), Boucher`s modification of RLW, and the Institute for Fusion Studies-Princeton Plasma Physics Laboratory (IFS-PPPL) model for transport due to ion temperature gradient modes are tested. The predictions use the measured densities in order to include the effects of density profile shape variations on the transport models. The uncertainties in the measured and predicted temperatures are discussed. The predictions based on the DIII- D scalings are within the measurement uncertainties. All the theoretical models predict a more favorable {rho}* dependence for the ion temperatures than is seen. Preliminary estimates indicate that sheared ow stabilization is important for some discharges, and that inclusion of its effects may bring the predictions of the IFS-PPPL model into agreement with the experiments.

  2. Environmental health and safety plan for the Molten Salt Reactor Experiment Remediation Project at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    Energy Technology Data Exchange (ETDEWEB)

    Burman, S.N.; Tiner, P.F.; Gosslee, R.C.

    1998-01-01

    The Lockheed Martin Energy Systems, Inc. (Energy Systems) policy is to provide a safe and healthful workplace for all employees and subcontractors. The accomplishment of this policy requires that operations at the Molten Salt Reactor Experiment (MSRE) facility at the Department of Energy (DOE) Oak Ridge National Laboratory (ORNL) are guided by an overall plan and consistent proactive approach to environmental protection and safety and health (S and H) issues. The policy and procedures in this plan apply to all MSRE operations. The provisions of this plan are to be carried out whenever activities are initiated at the MSRE that could be a threat to human health or the environment. This plan implements a policy and establishes criteria for the development of procedures for day-to-day operations to prevent or minimize any adverse impact to the environment and personnel safety and health and to meet standards that define acceptable management of hazardous and radioactive materials and wastes. The plan is written to utilize past experience and the best management practices to minimize hazards to human health or the environment from events such as fires, explosions, falls, mechanical hazards, or any unplanned release of hazardous or radioactive materials to the air.

  3. 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.

  4. 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.

  5. Reactor Simulator Testing Overview

    Science.gov (United States)

    Schoenfeld, Michael P.

    2013-01-01

    Test Objectives Summary: a) Verify operation of the core simulator, the instrumentation & control system, and the ground support gas and vacuum test equipment. b) Examine cooling & heat regeneration performance of the cold trap purification. c) Test the ALIP pump at voltages beyond 120V to see if the targeted mass flow rate of 1.75 kg/s can be obtained in the RxSim. Testing Highlights: a) Gas and vacuum ground support test equipment performed effectively for operations (NaK fill, loop pressurization, and NaK drain). b) Instrumentation & Control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings and ramped within prescribed constraints. It effectively interacted with reactor simulator control model and defaulted back to temperature control mode if the transient fluctuations didn't dampen. c) Cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained which was lower than the predicted 750 K but 156 K higher than the minimum temperature indicating the design provided some heat regeneration. d) ALIP produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz.

  6. Oxygen Carriers for Chemical Looping Combustion - 4 000 h of Operational Experience Transporteurs d’oxygène pour la combustion en boucle chimique : expérience accumulée pendant 4 000 h d’opération

    Directory of Open Access Journals (Sweden)

    Lyngfelt A.

    2011-04-01

    Full Text Available Chemical Looping Combustion (CLC is a new combustion technology with inherent separation of the greenhouse gas CO2. The technology involves the use of a metal oxide as an oxygen carrier which transfers oxygen from combustion air to the fuel, and hence a direct contact between air and fuel is avoided. Two interconnected fluidized beds, a fuel reactor and an air reactor, are used in the process. The outlet gas from the fuel reactor consists of CO2 and H2O, and the latter is easily removed by condensation. Considerable research has been conducted on CLC in the last years with respect to oxygen carrier development, reactor design, system efficiencies and prototype testing. Today, more than 700 materials have been tested and the technology has been successfully demonstrated in chemical looping combustors in the size range 0.3-140 kW, using different types of oxygen carriers based on oxides of the metals Ni, Co, Fe, Cu and Mn. The total time of operational experience is more than 4 000 hours. From these tests, it can be established that almost complete conversion of the fuel can be obtained and 100% CO2 capture is possible. Most work so far has been focused on gaseous fuels, but the direct application to solid fuels is also being studied. This paper presents an overview of operational experience with oxygen carriers in chemical looping combustors. La combustion en boucle chimique (CLC est une nouvelle technique de combustion permettant la séparation intrinsèque du CO2. Dans ce procédé, un oxyde métallique est utilisé comme transporteur d’oxygène pour véhiculer l’oxygène de l’air vers le combustible, ce qui permet d’éviter un contact direct entre le combustible et l’air. Deux lits fluidisés interconnectés sont utilisés, le réacteur air et le réacteur de combustion. Les fumées du réacteur de combustion contiennent le CO2 et la vapeur d’eau qui peut être facilement éliminée par condensation. Des recherches consid

  7. Control of SHARON reactor for autotrophic nitrogen removal in two-reactor configuration

    DEFF Research Database (Denmark)

    Valverde Perez, Borja; Mauricio Iglesias, Miguel; Sin, Gürkan

    2012-01-01

    With the perspective of investigating a suitable control design for autotrophic nitrogen removal, this work explores the control design for a SHARON reactor. With this aim, a full model is developed, including the pH dependency, in order to simulate the reactor and determine the optimal operating...... conditions. Then, the screening of controlled variables and pairing is carried out by an assessment of the effect of the disturbances based on the closed loop disturbance gain plots. Two controlled structures are obtained and benchmarked by their capacity to reject the disturbances before the Anammox reactor....

  8. CFD simulation of a four-loop PWR at asymmetric operation conditions

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, Jian-Ping; Yan, Li-Ming; Li, Feng-Chen, E-mail: lifch@hit.edu.cn

    2016-04-15

    Highlights: • A CFD numerical simulation procedure was established for simulating RPV of VVER-1000. • The established CFD approach was validated by comparing with available data. • Thermal hydraulic characteristics under asymmetric operation condition were investigated. • Apparent influences of the shutdown loop on its neighboring loops were obtained. - Abstract: The pressurized water reactor (PWR) with multiple loops may have abnormal working conditions with coolant pumps out of running in some loops. In this paper, a computational fluid dynamics (CFD) numerical study of the four-loop VVER-1000 PWR pressure vessel model was presented. Numerical simulations of the thermohydrodynamic characteristics in the pressure vessel were carried out at different inlet conditions with four and three loops running, respectively. At normal stead-state condition (four-loop running), different parameters were obtained for the full fluid domain, including pressure losses across different parts, pressure, velocity and temperature distributions in the reactor pressure vessel (RPV) and mass flow distribution of the coolant at the inlet of reactor core. The obtained results for pressure losses matched with the experimental reference values of the VVER-1000 PWR at Tianwan nuclear power plant (NPP). For most fuel assemblies (FAs), the inlet flow rates presented a symmetrical distribution about the center under full-loop operation conditions, which accorded with the practical distribution. These results indicate that it is now possible to study the dynamic transition process between different asymmetric operation conditions in a multi-loop PWR using the established CFD method.

  9. 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.

  10. Characterizing oxidative flow reactor SOA production and OH radical exposure from laboratory experiments of complex mixtures (engine exhaust) and simple precursors (monoterpenes)

    Science.gov (United States)

    Michael Link, M. L.; Friedman, B.; Ortega, J. V.; Son, J.; Kim, J.; Park, G.; Park, T.; Kim, K.; Lee, T.; Farmer, D.

    2016-12-01

    Recent commercialization of the Oxidative Flow Reactor (OFR, occasionally described in the literature as a "Potential Aerosol Mass") has created the opportunity for many researchers to explore the mechanisms behind OH-driven aerosol formation on a wide range of oxidative timescales (hours to weeks) in both laboratory and field measurements. These experiments have been conducted in both laboratory and field settings, including simple (i.e. single component) and complex (multi-component) precursors. Standard practices for performing OFR experiments, and interpreting data from the measurements, are still being developed. Measurement of gas and particle phase chemistry, from oxidation products generated in the OFR, through laboratory studies on single precursors and the measurement of SOA from vehicle emissions on short atmospheric timescales represent two very different experiments in which careful experimental design is essential for exploring reaction mechanisms and SOA yields. Two parameters essential in experimental design are (1) the role of seed aerosol in controlling gas-particle partitioning and SOA yields, and (2) the accurate determination of OH exposure during any one experiment. We investigated the role of seed aerosol surface area in controlling the observed SOA yields and gas/particle composition from the OH-initiated oxidation of four monoterpenes using an aerosol chemical ionization time-of-flight mass spectrometer and scanning mobility particle sizer. While the OH exposure during laboratory experiments is simple to constrain, complex mixtures such as diesel exhaust have high estimated OH reactivity values, and thus require careful consideration. We developed methods for constraining OH radical exposure in the OFR during vehicle exhaust oxidation experiments. We observe changes in O/C ratios and highly functionalized species over the temperature gradient employed in the aerosol-CIMS measurement. We relate this observed, speciated chemistry to the

  11. 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

  12. Introduction to Loop Heat Pipes

    Science.gov (United States)

    Ku, Jentung

    2015-01-01

    This is the presentation file for the short course Introduction to Loop Heat Pipes, to be conducted at the 2015 Thermal Fluids and Analysis Workshop, August 3-7, 2015, Silver Spring, Maryland. This course will discuss operating principles and performance characteristics of a loop heat pipe. Topics include: 1) pressure profiles in the loop; 2) loop operating temperature; 3) operating temperature control; 4) loop startup; 4) loop shutdown; 5) loop transient behaviors; 6) sizing of loop components and determination of fluid inventory; 7) analytical modeling; 8) examples of flight applications; and 9) recent LHP developments.

  13. A study on the use of the reactor basic experiments in the U-D2O lattices of the RB critical assembly for validation of modern nuclear data libraries

    Directory of Open Access Journals (Sweden)

    Pešić Milan P.

    2016-01-01

    Full Text Available Demand on the availability of well-defined reactor experiments for validation of computer codes for use in nuclear industry and nuclear technology is everlasting. Users must be confident of the results obtained by the proven computer codes and nuclear data libraries chosen in the models. The well-defined (mostly historical and evaluated reactor experiments (about 5000 in 2015 were collected continuously as the benchmarks within the frame of the OECD/NEA international projects ICSBEP (since 1995 and IRPhEP (since 2003. The Handbooks of the Projects are published in electronic forms (at the NEA web site of the OECD and at a DVD media every year. This study is aimed to (a examine and evaluate reactor basic experiments, carried out in the lattice of the natural uranium metal fuel in the heavy water of the RB critical assembly first core in 1958, and (b demonstrate their possibility for validation of modern nuclear data libraries. These RB reactor basic experiments include: (1 approach to criticality, (2 determination of the reactivity gradient at the D2O critical level, (3 measurement of the dependence of the D2O critical level on the D2O temperature, i. e. dependence of the reactivity with change in the D2O temperature; (4 the critical reactor geometrical parameter (buckling measurements, (5 the migration length measurements, (6 determination of the neutron multiplication factor in the infinite lattice, and (7 the safety rods reactivity measurements. Results of the experiments are compared to the results obtained using modern nuclear data libraries of the ACE type by applying the MCNP6.1, a well-known and proven computer code based on the Monte Carlo method. A short overview of these experiments (done at the RB assembly is shown. A brief description of the neutron ACE type nuclear data libraries (created in the LANL, based on the ENDF/B-VII.0 and ENDF/B-VII.1 files, or created in the OECD/NEA, based on the JEFF-3.2 evaluated nuclear data files

  14. Some particular aspects of control in nuclear power reactors; Conception de la surete en france et influence des imperatifs de surete sur la conception des reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    Vathaire, F. de; Vernier, Ph.; Pascouet, A. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    This paper reviews the experience acquired in France on the question, of reactor safety. Since a special paper is being presented on reactors of the graphite gas type, the safety of the other types studied in France is discussed here: - heavy water-gas reactors, - fast neutron reactors, - water research reactors of the swimming-pool and tank types. The safety rules peculiar to the different types are explained, with emphasis on their influence on the reactor designs and on the power limits they impose. The corresponding safety studies are presented, particular stress being placed on the original work developed in these fields. Special mention is made of the experimental systems constructed for these studies: the reactor CABRI, pile loop for depressurization tests, loops outside the pile, mock-ups etc. (authors) [French] La presente communication propose une synthese de l'experience acquise en France en matiere de surete des reacteurs. Les reacteurs de la filiere graphite-gaz faisant l'objet d'une communication particuliere, on examine ici la surete des autres types de reacteurs etudies en France: - reacteurs eau lourde-gaz, - reacteurs a neutrons rapides, - reacteurs de recherche a eau des types piscines et tank. Les imperatifs de surete propres aux differentes filieres sont developpes, en mettant l'accent sur leur influence sur la conception des reacteurs et sur les limitations de puissance qu'ils entrainent. Les etudes de surete correspondantes sont presentees, en insistant plus particulierement sur les travaux originaux developpes dans ces domaines. On indique notamment les moyens d'essais qui ont ete construits pour ces etudes: le reacteur CABRI, boucle en pile pour essais de depressurisation, boucles hors pile, maquettes, etc. (auteurs)

  15. Analysis of space reactor system components: Investigation through simulation and non-nuclear testing

    Science.gov (United States)

    Bragg-Sitton, Shannon M.

    The use of fission energy in space power and propulsion systems offers considerable advantages over chemical propulsion. Fission provides over six orders of magnitude higher energy density, which translates to higher vehicle specific impulse and lower specific mass. These characteristics enable ambitious space exploration missions. The natural space radiation environment provides an external source of protons and high energy, high Z particles that can result in the production of secondary neutrons through interactions in reactor structures. Applying the approximate proton source in geosynchronous orbit during a solar particle event, investigation using MCNPX 2.5.b for proton transport through the SAFE-400 heat pipe cooled reactor indicates an incoming secondary neutron current of (1.16 +/- 0.03) x 107 n/s at the core-reflector interface. This neutron current may affect reactor operation during low power maneuvers (e.g., start-up) and may provide a sufficient reactor start-up source. It is important that a reactor control system be designed to automatically adjust to changes in reactor power levels, maintaining nominal operation without user intervention. A robust, autonomous control system is developed and analyzed for application during reactor start-up, accounting for fluctuations in the radiation environment that result from changes in vehicle location or to temporal variations in the radiation field. Development of a nuclear reactor for space applications requires a significant amount of testing prior to deployment of a flight unit. High confidence in fission system performance can be obtained through relatively inexpensive non-nuclear tests performed in relevant environments, with the heat from nuclear fission simulated using electric resistance heaters. A series of non-nuclear experiments was performed to characterize various aspects of reactor operation. This work includes measurement of reactor core deformation due to material thermal expansion and

  16. Energy Conversion Advanced Heat Transport Loop and Power Cycle

    Energy Technology Data Exchange (ETDEWEB)

    Oh, C. H.

    2006-08-01

    operating conditions as well as trade offs between efficiency and capital cost. Prametric studies were carried out on reactor outlet temperature, mass flow, pressure, and turbine cooling. Recommendations on the optimal working fluid for each configuration were made. A steady state model comparison was made with a Closed Brayton Cycle (CBC) power conversion system developed at Sandia National Laboratory (SNL). A preliminary model of the CBC was developed in HYSYS for comparison. Temperature and pressure ratio curves for the Capstone turbine and compressor developed at SNL were implemented into the HYSYS model. A comparison between the HYSYS model and SNL loop demonstrated power output predicted by HYSYS was much larger than that in the experiment. This was due to a lack of a model for the electrical alternator which was used to measure the power from the SNL loop. Further comparisons of the HYSYS model and the CBC data are recommended. Engineering analyses were performed for several configurations of the intermediate heat transport loop that transfers heat from the nuclear reactor to the hydrogen production plant. The analyses evaluated parallel and concentric piping arrangements and two different working fluids, including helium and a liquid salt. The thermal-hydraulic analyses determined the size and insulation requirements for the hot and cold leg pipes in the different configurations. Economic analyses were performed to estimate the cost of the va

  17. Reactor antineutrinos and nuclear physics

    Energy Technology Data Exchange (ETDEWEB)

    Balantekin, A.B. [University of Wisconsin, Department of Physics, Madison, WI (United States)

    2016-11-15

    Short-baseline reactor neutrino experiments successfully measured the neutrino parameters they set out to measure, but they also identified a shape distortion in the 5-7 MeV range as well as a reduction from the predicted value of the flux. Nuclear physics input into the calculations of reactor antineutrino spectra needs to be better refined if this anomaly is to be interpreted as due to sterile neutrino states. (orig.)

  18. Blind Loop Syndrome

    Science.gov (United States)

    ... of tissue that protrude through the intestinal wall (diverticulosis) Certain medical conditions, including Crohn's disease, radiation enteritis, ... History of radiation therapy to the abdomen Diabetes Diverticulosis of the small intestine A blind loop can ...

  19. 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.

  20. Thermal-hydraulic interfacing code modules for CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Liu, W.S.; Gold, M.; Sills, H. [Ontario Hydro Nuclear, Toronto (Canada)] [and others

    1997-07-01

    The approach for CANDU reactor safety analysis in Ontario Hydro Nuclear (OHN) and Atomic Energy of Canada Limited (AECL) is presented. Reflecting the unique characteristics of CANDU reactors, the procedure of coupling the thermal-hydraulics, reactor physics and fuel channel/element codes in the safety analysis is described. The experience generated in the Canadian nuclear industry may be useful to other types of reactors in the areas of reactor safety analysis.

  1. Phenol biodegradation in a batch jet loop bioreactor (JLB): kinetics study and pH variation.

    Science.gov (United States)

    Ucun, Handan; Yildiz, Ergun; Nuhoglu, Alper

    2010-05-01

    Phenol biodegradation in a batch jet loop bioreactor (JLB) using activated sludge was investigated. The biodegradation experiments were conducted at different phenol concentrations (S(0)) from 50 to 1000 mg/l. The results of the biodegradation of phenol by JLB show that a good phenol removal of 100%. The biodegradation capacity of the JLB was higher than that of the stirred tank reactor reported in literatures. The Haldane equation was adopted in order to describe the relation between the specific growth rates (micro) and S(0). Kinetic constants of Haldane equation were micro(m) = 0.119 1/h, K(s) = 11.13 mg/l and K(i) = 250.88 mg/l. Model equations were simulated using the MATHCAD 7.0 software's ordinary differential equation solver. Simulations were performed at each experiment with different initial phenol concentrations. Copyright 2009 Elsevier Ltd. All rights reserved.

  2. Counter-part Test and Code Analysis of the Integral Test Loop, SNUF

    Energy Technology Data Exchange (ETDEWEB)

    Park, Goon Cherl; Bae, B. U.; Lee, K. H.; Cho, Y. J. [Seoul National Univ., Seoul (Korea, Republic of)

    2007-02-15

    The thermal-hydraulic phenomena of Direct Vessel Injection (DVI) line Small-Break Loss-of-Coolant Accident (SBLOCA) in pressurized water reactor, APR1400, were investigated. The reduced-height and reduced-pressure integral test loop, SNUF (Seoul National University Facility), was constructed with scaling down the prototype. For the appropriate test conditions in the experiment of SNUF, the energy scaling methodology was suggested as scaling the coolant mass inventory and thermal power for the reduced-pressure condition. From the MARS code analysis, the energy scaling methodology was confirmed to show the reasonable transient when ideally scaled-down SNUF model was compared to the prototype model. In the experiments according to the conditions determined by energy scaling methodology, the phenomenon of downcomer seal clearing had a dominant role in decrease of the system pressure and increase of the coolant level of core. The experimental results was utilized to validate the calculation capability of MARS.

  3. New reactor concepts. An analysis of the actual research status; Neue Reaktorkonzepte. Eine Analyse des aktuellen Forschungsstands

    Energy Technology Data Exchange (ETDEWEB)

    Pistner, Christoph; Englert, Matthias

    2017-04-15

    The report on new reactor concepts covers the following issues: characterization and survey of new reactor concepts; evaluation criteria: safety, resources for fuel supply, waste problems, economy and proliferation; comprehensive relevant aspects: thorium as alternative resource, partitioning and transmutation; actual developments and preliminary experiences for fast breeding reactor (FBR), high-temperature reactor (HTR), molten salt reactor (MSR), small modular reactor (SMR).

  4. Technical Information on the Carbonation of the EBR-II Reactor, Summary Report Part 1: Laboratory Experiments and Application to EBR-II Secondary Sodium System

    Energy Technology Data Exchange (ETDEWEB)

    Steven R. Sherman

    2005-04-01

    Residual sodium is defined as sodium metal that remains behind in pipes, vessels, and tanks after the bulk sodium metal has been melted and drained from such components. The residual sodium has the same chemical properties as bulk sodium, and differs from bulk sodium only in the thickness of the sodium deposit. Typically, sodium is considered residual when the thickness of the deposit is less than 5-6 cm. This residual sodium must be removed or deactivated when a pipe, vessel, system, or entire reactor is permanently taken out of service, in order to make the component or system safer and/or to comply with decommissioning regulations. As an alternative to the established residual sodium deactivation techniques (steam-and-nitrogen, wet vapor nitrogen, etc.), a technique involving the use of moisture and carbon dioxide has been developed. With this technique, sodium metal is converted into sodium bicarbonate by reacting it with humid carbon dioxide. Hydrogen is emitted as a by-product. This technique was first developed in the laboratory by exposing sodium samples to humidified carbon dioxide under controlled conditions, and then demonstrated on a larger scale by treating residual sodium within the Experimental Breeder Reactor II (EBR-II) secondary cooling system, followed by the primary cooling system, respectively. The EBR-II facility is located at the Idaho National Laboratory (INL) in southeastern Idaho, U.S.A. This report is Part 1 of a two-part report. It is divided into three sections. The first section describes the chemistry of carbon dioxide-water-sodium reactions. The second section covers the laboratory experiments that were conducted in order to develop the residual sodium deactivation process. The third section discusses the application of the deactivation process to the treatment of residual sodium within the EBR-II secondary sodium cooling system. Part 2 of the report, under separate cover, describes the application of the technique to residual sodium

  5. Reactor safety research programs. Quarterly progress report, January 1--March 31, 1977

    Energy Technology Data Exchange (ETDEWEB)

    Romano, A.J. (comp.)

    1977-05-01

    The projects reported each quarter are the following: Gas Reactor Safety Evaluation, THOR Code Development, SSC Code Development, LMFBR and LWR Safety Experiments, Fast Reactor Safety Code Validation, Technical Coordination of Structural Integrity, and Fast Reactor Safety Reliability Assessment.

  6. Simulation of the flow obstruction of a jet pump in a BWR reactor with the code RELAP/SCDAPSIM; Simulacion de la obstruccion de flujo de una bomba jet en un reactor BWR con el codigo RELAP/SCDAPSIM

    Energy Technology Data Exchange (ETDEWEB)

    Cardenas V, J.; Filio L, C., E-mail: jaime.cardenas@cnsns.gob.mx [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose M. Barragan 779, Col. Narvarte, 03020 Ciudad de Mexico (Mexico)

    2016-09-15

    This work simulates the flow obstruction of a jet pump in one of the recirculation loops of a nuclear power plant with a reactor of type BWR at 100% of operating power, in order to analyze the behavior of the total flow of the refrigerant passing through the reactor core, the total flow in each recirculation loop of the reactor, together with the 10 jet pumps of each loop. The behavior of the power and the reactivity insertion due to the change of the refrigerant flow pattern is also analyzed. The simulation was carried out using the RELAP/SCDAPSIM version 3.5 code, using a reactor model with 10 jet pumps in each recirculation loop and a core consisting of 6 radial zones and 25 axial zones. The scenario postulates the flow obstruction in a jet pump in a recirculation loop A when the reactor operates at 100% rated power, causing a change in the total flow of refrigerant in the reactor core, leading to a decrease in power. Once the reactor conditions are established to its new power, the operator tries to recover the nominal power using the flow control valve of the recirculation loop A, opening stepwise as a strategy to safely recover the reactor power. In this analysis is assumed that the intention of the nuclear plant operator is to maintain the operation of the reactor during the established cycle. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-02-01

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

  8. Chemical Looping Combustion of Methane: A Technology Development View

    Directory of Open Access Journals (Sweden)

    Rutuja Bhoje

    2013-01-01

    Full Text Available Methane is a reliable and an abundantly available energy source occurring in nature as natural gas, biogas, landfill gas, and so forth. Clean energy generation using methane can be accomplished by using chemical looping combustion. This theoretical study for chemical looping combustion of methane was done to consider some key technology development points to help the process engineer choose the right oxygen carrier and process conditions. Combined maximum product (H2O + CO2 generation, weight of the oxygen carrier, net enthalpy of CLC process, byproduct formation, CO2 emission from the air reactor, and net energy obtainable per unit weight (gram of oxygen carrier in chemical looping combustion can be important parameters for CLC operation. Carbon formed in the fuel reactor was oxidised in the air reactor and that increased the net energy obtainable from the CLC process but resulted in CO2 emission from the air reactor. Use of CaSO4 as oxygen carrier generated maximum energy (−5.3657 kJ, 800°C per gram of oxygen carrier used in the CLC process and was found to be the best oxygen carrier for methane CLC. Such a model study can be useful to identify the potential oxygen carriers for different fuel CLC systems.

  9. Process/Equipment Co-Simulation on Syngas Chemical Looping Process

    Energy Technology Data Exchange (ETDEWEB)

    Zeng, Liang; Zhou, Qiang; Fan, Liang-Shih

    2012-09-30

    The chemical looping strategy for fossil energy applications promises to achieve an efficient energy conversion system for electricity, liquid fuels, hydrogen and/or chemicals generation, while economically separate CO{sub 2} by looping reaction design in the process. Chemical looping particle performance, looping reactor engineering, and process design and applications are the key drivers to the success of chemical looping process development. In order to better understand and further scale up the chemical looping process, issues such as cost, time, measurement, safety, and other uncertainties need to be examined. To address these uncertainties, advanced reaction/reactor modeling and process simulation are highly desired and the modeling efforts can accelerate the chemical looping technology development, reduce the pilot-scale facility design time and operating campaigns, as well as reduce the cost and technical risks. The purpose of this work is thus to conduct multiscale modeling and simulations on the key aspects of chemical looping technology, including particle reaction kinetics, reactor design and operation, and process synthesis and optimization.

  10. Sabatier Methanation Reactor for Space Exploration

    Science.gov (United States)

    Murdoch, Karen; Goldblatt, Loel; Carrasquillo, Robyn; Harris, Danny

    2005-01-01

    The Sabatier Methanation Reactor technology is of vital importance to the success of the human and robotic exploration program. In order to achieve an affordable program, the logistics supply to support the mission must be minimized to the fullest extent possible. One area of potential reduction with high return on investment is the closure of life support loops, particularly oxygen and water. The Sabatier system accomplishes this by utilizing hydrogen and carbon dioxide, waste products from the life support system, to produce water and methane. The recovered water is then recycled back into the life support system to provide oxygen; while the methane can be used for propulsion, or can be broken down further to recover the hydrogen. This technology is applicable not only to transit phases of exploration, but surface habitats as well as in-situ propellant production. The Sabatier Reactor system has been developed for ground based demonstration experiments extensively over the past 30 years. Over the past three years, NASA has funded development of the Sabatier Carbon Dioxide Reduction Assembly (CRA) for use on the ISS. Currently this system is at TRL 5 and it is expected that the system will be flown on the ISS as a flight experiment, The purpose of the flight experiment is to integrate the Sabatier CRA into a synchronized system with the oxygen generation system and the carbon dioxide concentrator. The flight experiment will verify the integration of the different systems working together plus it will verify the capability of the system to operate, and effectively separate its products in a micro-gravity environment. Subsequent to design validation, the flight experiment can remain onboard the ISS providing valuable water to offset logistics re-supply requirements. Some of the challenges facing the development of the Sabatier system include handling vibration induced particulates, microgravity phase separation and containment of hazardous gases. Plans for

  11. Thermal coupling within LTP dynamics control loop

    Energy Technology Data Exchange (ETDEWEB)

    Nofrarias, M; Garcia Marin, A F; Heinzel, G; Hewitson, M; Danzmann, K [Max-Planck-Institut fuer Gravitationsphysik, Albert Einstein Institut (AEI), Callinstrasse 38, 30167 Hannover (Germany); Lobo, A; Sanjuan, J [Institut de Ciencies de l' Espai (ICE-CSIC), Facultat de Ciencies, Torre C5, 08193 Bellaterra (Spain); Ramos-Castro, J, E-mail: miquel.nofrarias@aei.mpg.d [Departament d' Enginyeria Electronica, UPC, Campus Nord, Edifici C4, Jordi Girona 1-3, 08034 Barcelona (Spain)

    2009-03-01

    The Diagnostic Subsytem in the LISA Technology Package (LTP) on board the LISA Pathfinder mission (LPF) will characterise those external disturbances with a potential impact on the performance of the experiment coming from either thermal, magnetic or charged particles perturbations. A correct design of the experiments to measure these effects in flight requires a closed loop analysis that takes into account the dynamics of the test masses, the force applied by the controllers and those noisy terms (coming from sensing or force noise) that enters into the loop. We describe this analysis in the thermal case and we give a first numerical example of the instrument response to controlled thermal inputs.

  12. 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…

  13. Experimental investigation of the chemical looping method on a 1 MW pilot plant; Experimentelle Untersuchung des Chemical Looping Verfahrens an einer 1 MW Versuchsanlage

    Energy Technology Data Exchange (ETDEWEB)

    Orth, Matthias

    2014-08-27

    Attempting to counteract the consequences of climate change, leading industrial nations have agreed on reducing their CO{sub 2} emissions significantly. To reach these reduction goals, it is essential to reduce the CO{sub 2} emissions in the field of energy conversion. This PHD thesis covers the field of chemical looping combustion, a technology that uses fossil fuels for energy conversion with inherent capture of CO{sub 2}. Since the research regarding chemical looping had so far focused mainly on lab scale or small scale experiments, a 1 MW pilot plant has been erected at Technische Universitaet Darmstadt in order to investigate the process in a semi-industrial scale and to check the process efficiency with commercially usable equipment. This pilot consists of two interconnected fluidized bed reactors and has an overall height of more than 11 m. In this thesis, some experiments with ilmenite - used as the oxygen carrier - are explained. Furthermore, the design, erection and commissioning of the pilot plant are presented as well as the results of the first test campaigns. The evaluation of the latter proves that the process can be handled in the design configuration and that CO{sub 2} can be safely captured in a pilot plant of this scale.

  14. Autonomous Control of Space Nuclear Reactors

    Science.gov (United States)

    Merk, John

    2013-01-01

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

  15. The Cinderella loop project

    Science.gov (United States)

    Schmelz, J. T.; Beene, J.; Coyle, T.; Douglass, J.; Nasraoui, K.; O'Connor, J.; Roames, J.; Scott, M.

    2006-01-01

    The solar loop that formed off the northeast limb of the Sun on 1999 November 6 (a.k.a. the Cinderella loop) is one of the few examples of a loop on the limb observed with all three of the following imaging instruments: the Transition Region and Coronal Explorer (TRACE), the SOHO Extreme-ultraviolet Imaging Telescope (EIT), and the Yohkoh Soft X-ray Telescope (SXT). In this project we investigate the temperature differences that result when examining the Cinderella loop with one instrument compared with another. For example, what temperature differences result from the increased spatial resolution between the two EUV imagers? More specifically, given that TRACE and EIT have almost identical temperature response to coronal plasma, does the different spatial resolution of TRACE (with 0.5″ pixels) and EIT (with 2.6″ pixels) produce statistically different results? We find that the answer is no, and that our results do not change after background subtraction. In addition, the spatial resolution of EIT and SXT is similar, but the temperature responses of the two instruments are quite different. The two instruments do not seem to be viewing the same loop strands, and the plasma temperature differences are significant.

  16. 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.

  17. Adaptive LQ Cascade Control of a Tubular Chemical Reactor

    Directory of Open Access Journals (Sweden)

    Petr Dostal

    2016-01-01

    Full Text Available The paper deals with adaptive LQ cascade control design of a tubular chemical reactor with an exothermic consecutive reaction. The control is performed in primary and secondary control-loops where the primary controlled output of the reactor is the concentration of a main reaction product and the secondary output is the mean temperature of the reactant. A common control input is the coolant flow rate. The controller in the primary control-loop is a nonlinear P-controller with the gain calculated using simulated or measured steady-state characteristics of the reactor. The controller in the secondary control-loop is a LQ adaptive controller. The proposed method is verified by control simulations.

  18. Effect of fuel particle size on reaction rate in chemical looping combustion

    Energy Technology Data Exchange (ETDEWEB)

    Iggland, M.; Leion, H.; Mattisson, T.; Lyngfelt, A. [ETH, Zurich (Switzerland). Inst. of Process Engineering

    2010-11-15

    Chemical looping combustion (CLC) uses an oxygen carrier circulating between an air and a fuel reactor to replace direct burning of fuels in air. The very low energy penalty for CO{sub 2} separation in CLC gives it the potential to become an important technology on the way to a CO{sub 2} neutral energy supply. In this work, the influence of the particle size of coal on the rate of reaction of the coal was investigated in a bed of oxygen carrier. In order to do this, a method to quench the reaction of coal with oxygen carriers at a specified time and measure the particle size distribution of the remaining coal was developed. Three size fractions of coal were used in the experiments: 90-125, 180-212 and 250-355 mu m. Particle size distributions of the fuel show a decrease in particle size with time. The influence of devolatilisation of the coal on the coal particle size was measured, showing that coal particles do not break in the fluidized bed reactor used for the experiments. Reaction rates based on measurements of gas phase concentrations of CO{sub 2}, CO and CH{sub 4} showed that the reaction rate is independent of the particle size. These results are in line with literature findings, as studies have shown that carbon gasification is size-independent at conditions similar to those in the performed CLC experiments.

  19. Evaluation of the Molten Salt Reactor Experiment drain tanks for reuse in salt disposal, Oak Ridge National Laboratory, Oak Ridge, Tennessee

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-05-01

    This report was prepared to identify the source documentation used to evaluate the drain tanks in the Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory (ORNL). The evaluation considered the original quality of the tanks, their service history, and their intended use during the removal of fluoride salts. It also includes recommendations for a quality verification plan. The estimates of corrosion damage to the salt containing system at the MSRE are low enough to lend optimism that the system will be fit for its intended use, which is disposal of the salt by transferring it to transport containers. The expected corrosion to date is estimated between 10 and 50 mil, or 2 to 10% of the shell wall. The expected corrosion rate when the tanks are used to remove the salt at 110 F is estimated to be .025 to 0.1 mil per hour of exposure to HF and molten salt. To provide additional assurance that the estimates of corrosion damage are accurate, cost effective nondestructive examination (NDE) has been recommended. The NDE procedures are compared with industry standards and give a perspective for the extent of additional measures taken in the recommendation. A methodology for establishing the remaining life has been recommended, and work is progressing towards providing an engineering evaluation based upon thickness and design conditions for the future use of the tanks. These extra measures and the code based analysis will serve to define the risk of salt or radioactive gases leaking during processing and transfer of the salt as acceptable.

  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. 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.

  3. Closing global material loops

    DEFF Research Database (Denmark)

    Prosman, Ernst-Jan; Wæhrens, Brian Vejrum; Liotta, Giacomo

    2017-01-01

    Replacing virgin materials with waste materials, a practice known as Industrial Symbiosis (IS), has been identified as a key strategy for closing material loops. This article adopts a critical view on geographic proximity and external coordinators – two key enablers of IS. By ‘uncovering’ a case...... for geographic proximity and external coordinators. In doing so, our insights into firm-level challenges of long-distance IS exchanges contribute to closing global material loops by increasing the number of potential circular pathways....

  4. Hidden-loop colostomy.

    Science.gov (United States)

    Rombeau, J L; Turnbul, R B

    1978-04-01

    Records of 15 patients having hidden-loop colostomies were reviewed. All patients had metastatic colonic cancers with impending obstructions. Six colostomies were subsequently opened because of obstructions due to cancer. All colostomy openings were done using local anesthesia in the emergency room. This technique prevented six major celiotomies and provided additional time of living without a stoma. There were two postoperative stomal prolapses, one of which necessitated reoperation. A hidden-loop colostomy is easily constructed and readily opened. It should be considered at celiotomy for selected patients who have metastatic colonic cancer with impending obstruction.

  5. QUARTERLY PROGRESS REPORT JANUARY, FEBRUARY, MARCH, 1967 REACTOR FUELS AND MATERIALS DEVELOPMENT PROGRAMS FOR FUELS AND MATERIALS BRANCH OF USAEC DEVISION OF REACTOR DEVELOPMENT AND TECHNOLOGY

    Energy Technology Data Exchange (ETDEWEB)

    Albaugh, F. W.; Bush, S. H.; Cadwell, J. J.; de Halas, D. R.; Worlton, D. C.

    1967-06-01

    Work is reported in the areas of: fast fuels oxides and nitrides; nuclear ceramics; nuclear graphite; basic swelling studies; irradiation damage to reactor metals; ATR gas loop operation and maintenance; metallic fuels; nondestructive testing research; and fast reactor dosimetry and damage analysis.

  6. PACTEL: Experiments on the behaviour of the new horizontal steam generator

    Energy Technology Data Exchange (ETDEWEB)

    Kouhia, J.; Riikonen, V.; Purhonen, H. [VTT Energy, Lappeenranta (Finland)

    1995-12-31

    Experiments were performed to study the behaviour of the PACTEL facility, a medium scale integral test loop simulating VVER 440 pressurized water reactors. The study focused on the operation of the new horizontal steam generator model installed in PACTEL. Three experiments were conducted: a small-break test to observe the steam generator behaviour over a range of primary coolant inventories, a hot leg loop seal experiment to study the cyclic behaviour of a loop seal and a loss of secondary side feedwater test to examine the effect of uncovered tubes in the steam generator. A reverse flow was observed in the lower part of the U-tube bundle of the steam generator during natural circulation. The flow reversal point dropped when the tubes uncovered, during secondary inventory reduction. (orig.). 5 refs.

  7. The Energy Landscape of Hyperstable LacI-DNA Loops

    Science.gov (United States)

    Kahn, Jason

    2009-03-01

    The Escherichia coli LacI protein represses transcription of the lac operon by blocking access to the promoter through binding at a promoter-proximal DNA operator. The affinity of tetrameric LacI (and therefore the repression efficiency) is enhanced by simultaneous binding to an auxiliary operator, forming a DNA loop. Hyperstable LacI-DNA loops were previously shown to be formed on DNA constructs that include a sequence-directed bend flanked by operators. Biochemical experiments showed that two such constructs (9C14 and 11C12) with different helical phasing between the operators and the DNA bend form different DNA loop shapes. The geometry and topology of the loops and the relevance of alternative conformations suggested by probable flexible linkers in LacI remain unclear. Bulk and single molecule fluorescence resonance energy transfer (SM-FRET, with D. English) experiments on a dual fluorophore-labeled 9C14-LacI loop demonstrate that it adopts a single, stable, rigid closed-form loop conformation. Here, we characterize the LacI-9C14 loop by SM-FRET as a function of inducer isopropyl-β,D-thiogalactoside (IPTG) concentration. Energy transfer measurements reveal partial but incomplete destabilization of loop formation by IPTG. Surprisingly, there is no change in the energy transfer efficiency of the remaining looped population. Models for the regulation of the lac operon often assume complete disruption of LacI-operator complexes upon inducer binding to LacI. Our work shows that even at saturating IPTG there is still a significant population of LacI-DNA complexes in a looped state, in accord with previous in vivo experiments that show incomplete induction (with J. Maher). Finally, we will report progress on characterizing the ``energy landscape'' for DNA looping upon systematic variation of the DNA linkers between the operators and the bending locus. Rod mechanics simulations (with N. Perkins) provide testable predictions on loop stability, topology, and FRET.

  8. Reference design for the standard mirror hybrid reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bender, D.J.; Fink, J.H.; Galloway, T.R.; Kastenberg, W.E.; Lee, J.D.; Devoto, R.S.; Neef, W.S. Jr.; Schultz, K.R.; Culver, D.W.; Rao, S.B.; Rao, S.R.

    1978-05-22

    This report describes the results of a two-year study by Lawrence Livermore Laboratory and General Atomic Co. to develop a conceptual design for the standard (minimum-B) mirror hybrid reactor. The reactor parameters have been chosen to minimize the cost of producing nuclear fuel (/sup 239/Pu) for consumption in fission power reactors (light water reactors). The deuterium-tritium plasma produces approximately 400 MW of fusion power with a plasma Q of 0.64. The fast-fission blanket, which is fueled with depleted uranium and lithium, generates sufficient tritium to run the reactor, has a blanket energy multiplication of M = 10.4, and has a net fissile breeding ratio of Pu/n = 1.51. The reactor has a net electrical output of 600 MWe, a fissile production of 2000 kg of plutonium per year (at a capacity factor of 0.74), and a net plant efficiency of 0.18. The plasma-containment field is generated by a Yin-Yang magnet using NbTi superconductor, and the neutral beam system uses positive-ion acceleration with beam direct conversion. The spherical blanket is based on gas-cooled fast reactor technology. The fusion components, blanket, and primary heat-transfer loop components are all contained within a prestressed-concrete reactor vessel, which provides magnet restraint and supports the primary heat-transfer loop and the blanket.

  9. Loop Quantum Gravity

    Directory of Open Access Journals (Sweden)

    Rovelli Carlo

    1998-01-01

    Full Text Available The problem of finding the quantum theory of the gravitational field, and thus understanding what is quantum spacetime, is still open. One of the most active of the current approaches is loop quantum gravity. Loop quantum gravity is a mathematically well-defined, non-perturbative and background independent quantization of general relativity, with its conventional matter couplings. Research in loop quantum gravity today forms a vast area, ranging from mathematical foundations to physical applications. Among the most significant results obtained are: (i The computation of the physical spectra of geometrical quantities such as area and volume, which yields quantitative predictions on Planck-scale physics. (ii A derivation of the Bekenstein-Hawking black hole entropy formula. (iii An intriguing physical picture of the microstructure of quantum physical space, characterized by a polymer-like Planck scale discreteness. This discreteness emerges naturally from the quantum theory and provides a mathematically well-defined realization of Wheeler's intuition of a spacetime ``foam''. Long standing open problems within the approach (lack of a scalar product, over-completeness of the loop basis, implementation of reality conditions have been fully solved. The weak part of the approach is the treatment of the dynamics: at present there exist several proposals, which are intensely debated. Here, I provide a general overview of ideas, techniques, results and open problems of this candidate theory of quantum gravity, and a guide to the relevant literature.

  10. LAPAROSCOPIC ILEAL LOOP CONDUIT

    African Journals Online (AJOL)

    Then the right ureter was laparoscopically spa- tulated and anastomosed to the ileostomy opening using interrupted 4/0 vicryl sutures. After finishing half the circumference of the anastomotic line, a 4 Fr. ureteric catheter was introduced through the external stoma of the loop up to the site of the anastomosis with the aid of a ...

  11. Improving Loop Dependence Analysis

    DEFF Research Database (Denmark)

    Jensen, Nicklas Bo; Karlsson, Sven

    2017-01-01

    Programmers can no longer depend on new processors to have significantly improved single-thread performance. Instead, gains have to come from other sources such as the compiler and its optimization passes. Advanced passes make use of information on the dependencies related to loops. We improve th...

  12. 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.

  13. Phosphorus removal in aerated stirred tank reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ghigliazza, R.; Lodi, A.; Rovatti, M. [Inst. of Chemical and Process Engineering ``G.B. Bonino``, Univ. of Genoa (Italy)

    1999-03-01

    The possibility to obtain biological phosphorus removal in strictly aerobic conditions has been investigated. Experiments, carried out in a continuous stirred tank reactor (CSTR), show the feasibility to obtain phosphorus removal without the anaerobic phase. Reactor performance in terms of phosphorus abatement kept always higher then 65% depending on adopted sludge retention time (SRT). In fact increasing SRT from 5 days to 8 days phosphorus removal and reactor performance increase but overcoming this SRT value a decreasing in reactor efficiency was recorded. (orig.) With 6 figs., 3 tabs., 18 refs.

  14. Helium-cooled high temperature reactors

    Energy Technology Data Exchange (ETDEWEB)

    Trauger, D.B.

    1985-01-01

    Experience with several helium cooled reactors has been favorable, and two commercial plants are now operating. Both of these units are of the High Temperature Graphite Gas Cooled concept, one in the United States and the other in the Federal Republic of Germany. The initial helium charge for a reactor of the 1000 MW(e) size is modest, approx.15,000 kg.

  15. Utilisation of British University Research Reactors.

    Science.gov (United States)

    Duncton, P. J.; And Others

    British experience relating to the employment of university research reactors and subcritical assemblies in the education of nuclear scientists and technologists, in the training of reactor operators and for fundamental pure and applied research in this field is reviewed. The facilities available in a number of British universities and the uses…

  16. SoLid: Search for Oscillation with a 6Li Detector at the BR2 research reactor

    OpenAIRE

    Michiels, Ianthe

    2016-01-01

    In the past decades, various nuclear reactor neutrino experiments have measured a deficit in the flux of antineutrinos coming from the reactor at short reactor-detector distances, when compared to theoretical calculations. One of the experiments designed to investigate this reactor antineutrino anomaly is the SoLid experiment. It uses the compact BR2 research reactor from the SCK-CEN in Mol, Belgium, to perform reactor antineutrino flux measurements at very short baseline. These proceedings d...

  17. Microchannel Reactors for ISRU Applications

    Science.gov (United States)

    Carranza, Susana; Makel, Darby B.; Blizman, Brandon; Ward, Benjamin J.

    2005-02-01

    Affordable planning and execution of prolonged manned space missions depend upon the utilization of local resources and the waste products which are formed in manned spacecraft and surface bases. Successful in-situ resources utilization (ISRU) will require component technologies which provide optimal size, weight, volume, and power efficiency. Microchannel reactors enable the efficient chemical processing of in situ resources. The reactors can be designed for the processes that generate the most benefit for each mission. For instance, propellants (methane) can be produced from carbon dioxide from the Mars atmosphere using the Sabatier reaction and ethylene can be produced from the partial oxidation of methane. A system that synthesizes ethylene could be the precursor for systems to synthesize ethanol and polyethylene. Ethanol can be used as a nutrient for Astrobiology experiments, as well as the production of nutrients for human crew (e.g. sugars). Polyethylene can be used in the construction of habitats, tools, and replacement parts. This paper will present recent developments in miniature chemical reactors using advanced Micro Electro Mechanical Systems (MEMS) and microchannel technology to support ISRU of Mars and lunar missions. Among other applications, the technology has been demonstrated for the Sabatier process and for the partial oxidation of methane. Microchannel reactors were developed based on ceramic substrates as well as metal substrates. In both types of reactors, multiple layers coated with catalytic material are bonded, forming a monolithic structure. Such reactors are readily scalable with the incorporation of extra layers. In addition, this reactor structure minimizes pressure drop and catalyst settling, which are common problems in conventional packed bed reactors.

  18. Reactor Simulator Integration and Testing

    Science.gov (United States)

    Schoenfield, M. P.; Webster, K. L.; Pearson, J. B.

    2013-01-01

    As part of the Nuclear Systems Office Fission Surface Power Technology Demonstration Unit (TDU) project, a reactor simulator (RxSim) test loop was designed and built to perform integrated testing of the TDU components. In particular, the objectives of RxSim testing were to verify the operation of the core simulator, the instrumentation and control system, and the ground support gas and vacuum test equipment. In addition, it was decided to include a thermal test of a cold trap purification design and a pump performance test at pump voltages up to 150 V because the targeted mass flow rate of 1.75 kg/s was not obtained in the RxSim at the originally constrained voltage of 120 V. This Technical Memorandum summarizes RxSim testing. The gas and vacuum ground support test equipment performed effectively in NaK fill, loop pressurization, and NaK drain operations. The instrumentation and control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings. The cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained, which was lower than the predicted 750 K but 156 K higher than the cold temperature, indicating the design provided some heat regeneration. The annular linear induction pump tested was able to produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz.

  19. Physics of reactor safety. Quarterly report, April--June 1977. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    None

    1977-09-01

    The work in the Applied Physics Division includes reports on reactor safety program by members of the Reactor Safety Appraisals Group, Monte Carlo analysis of safety-related critical assembly experiments by members of the Theoretical Fast Reactor Physics Group, and planning of safety-related critical experiments by members of the Zero Power Reactor (ZPR) Planning and Experiments Group. Work on Reactor core thermal-hydraulic code development performed in the Components Technology Division is also included in the report.

  20. Natural Ores as Oxygen Carriers in Chemical Looping Combustion

    Energy Technology Data Exchange (ETDEWEB)

    Tian, Hanjing; Siriwardane, Ranjani; Simonyi, Thomas; Poston, James

    2013-08-01

    Chemical looping combustion (CLC) is a combustion technology that utilizes oxygen from oxygen carriers (OC), such as metal oxides, instead of air to combust fuels. The use of natural minerals as oxygen carriers has advantages, such as lower cost and availability. Eight materials, based on copper or iron oxides, were selected for screening tests of CLC processes using coal and methane as fuels. Thermogravimetric experiments and bench-scale fixed-bed reactor tests were conducted to investigate the oxygen transfer capacity, reaction kinetics, and stability during cyclic reduction/oxidation reaction. Most natural minerals showed lower combustion capacity than pure CuO/Fe{sub 2}O{sub 3} due to low-concentrations of active oxide species in minerals. In coal CLC, chryscolla (Cu-based), magnetite, and limonite (Fe-based) demonstrated better reaction performances than other materials. The addition of steam improved the coal CLC performance when using natural ores because of the steam gasification of coal and the subsequent reaction of gaseous fuels with active oxide species in the natural ores. In methane CLC, chryscolla, hematite, and limonite demonstrated excellent reactivity and stability in 50-cycle thermogravimetric analysis tests. Fe{sub 2}O{sub 3}-based ores possess greater oxygen utilization but require an activation period before achieving full performance in methane CLC. Particle agglomeration issues associated with the application of natural ores in CLC processes were also studied by scanning electron microscopy (SEM).

  1. Korea advanced liquid metal reactor development - Development of measuring techniques of the sodium two-phase flow

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Moo Hwan; Cha, Jae Eun [Pohang University of Science and Technology, Pohang (Korea)

    2000-04-01

    The technology which models and measures the behavior of bubble in liquid sodium is very important to insure the safety of the liquid metal reactor. In this research, we designed/ manufactured each part and loop of experimental facility for sodium two phase flow, and applied a few possible methods, measured characteristic of two phase flow such as bubbly flow. A air-water loop similar to sodium loop on each measuring condition was designed/manufactured. This air-water loop was utilized to acquire many informations which were necessary in designing the two phase flow of sodium and manufacturing experimental facility. Before the manufacture of a electromagnetic flow meter for sodium, the experiment using each electromagnetic flow mete was developed and the air-water loop was performed to understand flow characteristics. Experiments for observing the signal characteristics of flow were performed by flowing two phase mixture into the electromagnetic flow mete. From these experiments, the electromagnetic flow meter was designed and constructed by virtual electrode, its signal processing circuit and micro electro magnet. It was developed to be applicable to low conductivity fluid very successfully. By this experiment with the electromagnetic flow meter, we observed that the flow signal was very different according to void fraction in two phase flow and that probability density function which was made by statistical signal treatment is also different according to flow patterns. From this result, we confirmed that the electromagnetic flow meter could be used to understand the parameters of two phase flow of sodium. By this study, the experimental facility for two phase flow of sodium was constricted. Also the new electromagnetic flow meter was designed/manufactured, and experimental apparatus for two phase flow of air-water. Finally, this study will be a basic tool for measurement of two phase flow of sodium. As the fundamental technique for the applications of sodium at

  2. Closed Brayton cycle power conversion systems for nuclear reactors :

    Energy Technology Data Exchange (ETDEWEB)

    Wright, Steven A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Lipinski, Ronald J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Vernon, Milton E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sanchez, Travis [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2006-04-01

    This report describes the results of a Sandia National Laboratories internally funded research program to study the coupling of nuclear reactors to gas dynamic Brayton power conversion systems. The research focused on developing integrated dynamic system models, fabricating a 10-30 kWe closed loop Brayton cycle, and validating these models by operating the Brayton test-loop. The work tasks were performed in three major areas. First, the system equations and dynamic models for reactors and Closed Brayton Cycle (CBC) systems were developed and implemented in SIMULINKTM. Within this effort, both steady state and dynamic system models for all the components (turbines, compressors, reactors, ducting, alternators, heat exchangers, and space based radiators) were developed and assembled into complete systems for gas cooled reactors, liquid metal reactors, and electrically heated simulators. Various control modules that use proportional-integral-differential (PID) feedback loops for the reactor and the power-conversion shaft speed were also developed and implemented. The simulation code is called RPCSIM (Reactor Power and Control Simulator). In the second task an open cycle commercially available Capstone C30 micro-turbine power generator was modified to provide a small inexpensive closed Brayton cycle test loop called the Sandia Brayton test-Loop (SBL-30). The Capstone gas-turbine unit housing was modified to permit the attachment of an electrical heater and a water cooled chiller to form a closed loop. The Capstone turbine, compressor, and alternator were used without modification. The Capstone systems nominal operating point is 1150 K turbine inlet temperature at 96,000 rpm. The annular recuperator and portions of the Capstone control system (inverter) and starter system also were reused. The rotational speed of the turbo-machinery is controlled by adjusting the alternator load by using the electrical grid as the load bank. The SBL-30 test loop was operated at

  3. Chemical looping reforming of generator gas

    Energy Technology Data Exchange (ETDEWEB)

    Mendiara, T.; Jensen, Anker; Glarborg, P.

    2010-02-15

    The main objective of this work is to investigate the carbon deposition during reforming of hydrocarbons in a Chemical Looping Reformer (CLR). This knowledge is needed to asses the viability of the CLR technology in reforming tar from biomass gasification preserving lighter hydrocarbons and minimizing the carbon formation during the process. Two different setups were used to test the reactivity of the different samples in the conditions of interest for the tar reforming process: 1) Fixed bed flow reactor (FR), and 2) Thermogravimetric analyzer (TGA). In the experiments, the gas atmosphere was switched from reducing to oxidizing atmosphere in every cycle. During the oxidizing cycle, the carrier was regenerated using a mixture of oxygen and nitrogen. Four different oxygen carriers based on nickel (Ni40 and Ni60), manganese (Mn) and ilmenite (Fe) were tested. In the tests, toluene was used to simulate the tars. The Fe and the Mn carrier reacted to a small extent with methane at the highest temperature studied, 800 degrees C. The Ni-carriers did not react at 600 degrees C at first, but they showed some reactivity after having been activated at the higher temperature. Carbon formation occurred with the Ni-carriers, more so with the Ni60 than the Ni40. Ni40, Mn and Fe were activated at the higher temperature. However, Fe showed only low capacity. Ni60 showed no capability of tar reforming. Ni40 showed a high tendency to carbon formation at 800 degrees C, but the formation could be lowered by changing some parameters. Mn formed almost no carbon. Ni40 and Mn were chosen for further studies. Carbon deposition occurred for both Ni40 and Mn, but the amount deposited for Ni40 was about 10 times bigger. Ni40 reacted with the methane and toluene only at 800 degrees C. The conversion over Mn was not as big as for toluene alone. Carbon was formed from carbon monoxide on the Ni40 carrier and on the Mn, but to a much less extent on the latter one. The presence of hydrogen decreased

  4. LOSS-OF-COOLANT ACIDENT SIMULATIONS IN THE NATIONAL RESEARCH UNIVERSAL REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    Bennett, W D; Goodman, R L; Heaberlin, S W; Hesson, G M; Nealley, C; Kirg, L L; Marshall, R K; McNair, G W; Meitzler, W D; Neally, G W; Parchen, L J; Pilger, J P; Rausch, W N; Russcher, G E; Schreiber, R E; Wildung, N J

    1981-02-01

    Pressurized water reactor loss-of-coolant accident (LOCA) phenomena are being simulated with a series of experiments in the U-2 loop of the National Research Universal Reactor at Chalk River, Ontario, Canada. The first of these experiments includes up to 45 parametric thermal-hydraulic tests to establish the relationship among the reflood delay time of emergency coolant, the reflooding rate, and the resultant fuel rod cladding peak temperature. Subsequent experiments establish the fuel rod failure characteristics at selected peak cladding temperatures. Fuel rod cladding pressurization simulates high burnup fission gas pressure levels of modern PWRs. This document contains both an experiment overview of the LOCA simulation program and a review of the safety analyses performed by Pacific Northwest Laboratory (PNL) to define the expected operating conditions as well as to evaluate the worst case operating conditions. The primary intent of this document is to supply safety information required by the Chalk River Nuclear Laboratories (CRNL), to establish readiness to proceed from one test phase to the next and to establish the overall safety of the experiment. A hazards review summarizes safety issues, normal operation and three worst case accidents that have been addressed during the development of the experiment plan.

  5. MODELLING AND CONTROL OF CONTINUOUS STIRRED TANK REACTOR WITH PID CONTROLLER

    Directory of Open Access Journals (Sweden)

    Artur Wodołażski

    2016-09-01

    Full Text Available This paper presents a model of dynamics control for continuous stirred tank reactor (CSTR in methanol synthesis in a three-phase system. The reactor simulation was carried out for steady and transient state. Efficiency ratio to achieve maximum performance of the product per reactor unit volume was calculated. Reactor dynamics simulation in closed loop allowed to received data for tuning PID controller (proportional-integral-derivative. The results of the regulation process allow to receive data for optimum reactor production capacity, along with local hot spots eliminations or temperature runaway.

  6. Post test calculation of the experiment `small break loss-of- coolant test` SBL-22 at the Finnish integral test facility PACTEL with the thermohydraulic code ATHLET

    Energy Technology Data Exchange (ETDEWEB)

    Lischke, W.; Vandreier, B. [Univ. for Applied Sciences, Zittau/Goerlitz (Germany). Dept. of Nuclear Technology

    1997-12-31

    At the University for Applied Sciences Zittau/Goerlitz (FH) calculations for the verification of the ATHLET-code for reactors of type VVER are carried out since 1991, sponsored by the German Ministry for Education, Science and Technology (BMBF). The special features of these reactors in comparison to reactors of western countries are characterized by the duct route of reactor coolant pipes and the horizontal steam generators. Because of these special features, a check of validity of the ATHLET-models is necessary. For further verification of the ATHLET-code the post test calculation of the experiment SBL-22 (Small break loss-of-coolant test) realized at the finnish facility PACTEL was carried out. The experiment served for the examination of the natural circulation behaviour of the loop over a continuous range of primary side water inventory. 5 refs.

  7. Parameterizing loop fusion for automated empirical tuning

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Y; Yi, Q; Kennedy, K; Quinlan, D; Vuduc, R

    2005-12-15

    Traditional compilers are limited in their ability to optimize applications for different architectures because statically modeling the effect of specific optimizations on different hardware implementations is difficult. Recent research has been addressing this issue through the use of empirical tuning, which uses trial executions to determine the optimization parameters that are most effective on a particular hardware platform. In this paper, we investigate empirical tuning of loop fusion, an important transformation for optimizing a significant class of real-world applications. In spite of its usefulness, fusion has attracted little attention from previous empirical tuning research, partially because it is much harder to configure than transformations like loop blocking and unrolling. This paper presents novel compiler techniques that extend conventional fusion algorithms to parameterize their output when optimizing a computation, thus allowing the compiler to formulate the entire configuration space for loop fusion using a sequence of integer parameters. The compiler can then employ an external empirical search engine to find the optimal operating point within the space of legal fusion configurations and generate the final optimized code using a simple code transformation system. We have implemented our approach within our compiler infrastructure and conducted preliminary experiments using a simple empirical search strategy. Our results convey new insights on the interaction of loop fusion with limited hardware resources, such as available registers, while confirming conventional wisdom about the effectiveness of loop fusion in improving application performance.

  8. 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

  9. Degradation Mechanisms of Colloidal Organic Matter in Biofilm Reactors

    DEFF Research Database (Denmark)

    Larsen, Tove; Harremoës, Poul

    1994-01-01

    The degradation mechanisms of colloidal organic matter in biofilm reactors have been studied in an idealized laboratory reactor system with soluble starch as a model substrate. Batch tests and experiments with different reactor configurations have shown that for this specific substrate, bulk liquid...

  10. 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...

  11. Dynamic feedback characteristics of Ghana Research Reactor-1 ...

    African Journals Online (AJOL)

    Dynamic experiments were performed to investigate the effects of insertions of step and ramp reactivities on Ghana Research Reactor-1. These safety performance tests of the reactor show that the reactor is inherently safe. The peak powers were found to be low and could not lead to damage of fuel meat and cladding.

  12. Loop Quantum Cosmology.

    Science.gov (United States)

    Bojowald, Martin

    2008-01-01

    Quantum gravity is expected to be necessary in order to understand situations in which classical general relativity breaks down. In particular in cosmology one has to deal with initial singularities, i.e., the fact that the backward evolution of a classical spacetime inevitably comes to an end after a finite amount of proper time. This presents a breakdown of the classical picture and requires an extended theory for a meaningful description. Since small length scales and high curvatures are involved, quantum effects must play a role. Not only the singularity itself but also the surrounding spacetime is then modified. One particular theory is loop quantum cosmology, an application of loop quantum gravity to homogeneous systems, which removes classical singularities. Its implications can be studied at different levels. The main effects are introduced into effective classical equations, which allow one to avoid the interpretational problems of quantum theory. They give rise to new kinds of early-universe phenomenology with applications to inflation and cyclic models. To resolve classical singularities and to understand the structure of geometry around them, the quantum description is necessary. Classical evolution is then replaced by a difference equation for a wave function, which allows an extension of quantum spacetime beyond classical singularities. One main question is how these homogeneous scenarios are related to full loop quantum gravity, which can be dealt with at the level of distributional symmetric states. Finally, the new structure of spacetime arising in loop quantum gravity and its application to cosmology sheds light on more general issues, such as the nature of time. Supplementary material is available for this article at 10.12942/lrr-2008-4.

  13. Loop Quantum Cosmology

    Directory of Open Access Journals (Sweden)

    Bojowald Martin

    2008-07-01

    Full Text Available Quantum gravity is expected to be necessary in order to understand situations in which classical general relativity breaks down. In particular in cosmology one has to deal with initial singularities, i.e., the fact that the backward evolution of a classical spacetime inevitably comes to an end after a finite amount of proper time. This presents a breakdown of the classical picture and requires an extended theory for a meaningful description. Since small length scales and high curvatures are involved, quantum effects must play a role. Not only the singularity itself but also the surrounding spacetime is then modified. One particular theory is loop quantum cosmology, an application of loop quantum gravity to homogeneous systems, which removes classical singularities. Its implications can be studied at different levels. The main effects are introduced into effective classical equations, which allow one to avoid the interpretational problems of quantum theory. They give rise to new kinds of early-universe phenomenology with applications to inflation and cyclic models. To resolve classical singularities and to understand the structure of geometry around them, the quantum description is necessary. Classical evolution is then replaced by a difference equation for a wave function, which allows an extension of quantum spacetime beyond classical singularities. One main question is how these homogeneous scenarios are related to full loop quantum gravity, which can be dealt with at the level of distributional symmetric states. Finally, the new structure of spacetime arising in loop quantum gravity and its application to cosmology sheds light on more general issues, such as the nature of time.

  14. Loop Quantum Cosmology

    Directory of Open Access Journals (Sweden)

    Bojowald Martin

    2005-12-01

    Full Text Available Quantum gravity is expected to be necessary in order to understand situations where classical general relativity breaks down. In particular in cosmology one has to deal with initial singularities, i.e., the fact that the backward evolution of a classical space-time inevitably comes to an end after a finite amount of proper time. This presents a breakdown of the classical picture and requires an extended theory for a meaningful description. Since small length scales and high curvatures are involved, quantum effects must play a role. Not only the singularity itself but also the surrounding space-time is then modified. One particular realization is loop quantum cosmology, an application of loop quantum gravity to homogeneous systems, which removes classical singularities. Its implications can be studied at different levels. Main effects are introduced into effective classical equations which allow to avoid interpretational problems of quantum theory. They give rise to new kinds of early universe phenomenology with applications to inflation and cyclic models. To resolve classical singularities and to understand the structure of geometry around them, the quantum description is necessary. Classical evolution is then replaced by a difference equation for a wave function which allows to extend space-time beyond classical singularities. One main question is how these homogeneous scenarios are related to full loop quantum gravity, which can be dealt with at the level of distributional symmetric states. Finally, the new structure of space-time arising in loop quantum gravity and its application to cosmology sheds new light on more general issues such as time.

  15. Cosmic string loop microlensing

    Science.gov (United States)

    Bloomfield, Jolyon K.; Chernoff, David F.

    2014-06-01

    Cosmic superstring loops within the galaxy microlens background point sources lying close to the observer-string line of sight. For suitable alignments, multiple paths coexist and the (achromatic) flux enhancement is a factor of two. We explore this unique type of lensing by numerically solving for geodesics that extend from source to observer as they pass near an oscillating string. We characterize the duration of the flux doubling and the scale of the image splitting. We probe and confirm the existence of a variety of fundamental effects predicted from previous analyses of the static infinite straight string: the deficit angle, the Kaiser-Stebbins effect, and the scale of the impact parameter required to produce microlensing. Our quantitative results for dynamical loops vary by O(1) factors with respect to estimates based on infinite straight strings for a given impact parameter. A number of new features are identified in the computed microlensing solutions. Our results suggest that optical microlensing can offer a new and potentially powerful methodology for searches for superstring loop relics of the inflationary era.

  16. Digital computer study of nuclear reactor thermal transients during startup of 60-kWe Brayton power conversion system

    Science.gov (United States)

    Jefferies, K. S.; Tew, R. C.

    1974-01-01

    A digital computer study was made of reactor thermal transients during startup of the Brayton power conversion loop of a 60-kWe reactor Brayton power system. A startup procedure requiring the least Brayton system complication was tried first; this procedure caused violations of design limits on key reactor variables. Several modifications of this procedure were then found which caused no design limit violations. These modifications involved: (1) using a slower rate of increase in gas flow; (2) increasing the initial reactor power level to make the reactor respond faster; and (3) appropriate reactor control drum manipulation during the startup transient.

  17. Determination and Fabrication of New Shield Super Alloys Materials for Nuclear Reactor Safety by Experiments and Cern-Fluka Monte Carlo Simulation Code, Geant4 and WinXCom

    Science.gov (United States)

    Aygun, Bünyamin; Korkut, Turgay; Karabulut, Abdulhalik

    2016-05-01

    Despite the possibility of depletion of fossil fuels increasing energy needs the use of radiation tends to increase. Recently the security-focused debate about planned nuclear power plants still continues. The objective of this thesis is to prevent the radiation spread from nuclear reactors into the environment. In order to do this, we produced higher performanced of new shielding materials which are high radiation holders in reactors operation. Some additives used in new shielding materials; some of iron (Fe), rhenium (Re), nickel (Ni), chromium (Cr), boron (B), copper (Cu), tungsten (W), tantalum (Ta), boron carbide (B4C). The results of this experiments indicated that these materials are good shields against gamma and neutrons. The powder metallurgy technique was used to produce new shielding materials. CERN - FLUKA Geant4 Monte Carlo simulation code and WinXCom were used for determination of the percentages of high temperature resistant and high-level fast neutron and gamma shielding materials participated components. Super alloys was produced and then the experimental fast neutron dose equivalent measurements and gamma radiation absorpsion of the new shielding materials were carried out. The produced products to be used safely reactors not only in nuclear medicine, in the treatment room, for the storage of nuclear waste, nuclear research laboratories, against cosmic radiation in space vehicles and has the qualities.

  18. Effects of Steam and CO2 in the Fluidizing Gas when Using Bituminous Coal in Chemical-Looping Combustion

    Science.gov (United States)

    Leion, H.; Lyngfelt, A.; Mattisson, T.

    Chemical-looping combustion (CLC) is a combustion technology where an oxygen carrier is used to transfer oxygen from the combustion air to the fuel in order to avoid direct contact between air and fuel. Thus, the CO2 is inherently separated from the flue gases with a potential for considerably lower energy penalty and cost compared to other techniques for CO2 separation. The oxygen carrier is circulated between two reactors, a fuel and an air reactor, where the flue gas from the air reactor contains oxygen depleted air and the flue gas from the fuel reactor contains mainly CO2 and H2O. The water can easily be condensed and the remaining CO2 can be transported for underground storage. Most of the prior work with CLC has focused on using natural gas and syngas as fuel and oxygen carrying material normally produced from pure chemicals. However, recent work on adapting the CLC process for solid fuels with ores and natural minerals as oxygen carrier shows promising results. This paper will present results from reactivity investigations in a laboratory fluidized-bed reactor system using previously investigated natural mineral ilmenite as oxygen carrier and a bituminous Colombian coal as fuel. Experiments were conducted at a temperature of 970°C with N2, steam, and/or CO2 in the fluidizing gas. Synergy effects between steam and CO2 on fuel conversion was noted. The results show that the fuel conversion was a roughly a factor 5 faster with steam as compared to CO2 in the fluidizing gas.

  19. LoopIng: a template-based tool for predicting the structure of protein loops.

    KAUST Repository

    Messih, Mario Abdel

    2015-08-06

    Predicting the structure of protein loops is very challenging, mainly because they are not necessarily subject to strong evolutionary pressure. This implies that, unlike the rest of the protein, standard homology modeling techniques are not very effective in modeling their structure. However, loops are often involved in protein function, hence inferring their structure is important for predicting protein structure as well as function.We describe a method, LoopIng, based on the Random Forest automated learning technique, which, given a target loop, selects a structural template for it from a database of loop candidates. Compared to the most recently available methods, LoopIng is able to achieve similar accuracy for short loops (4-10 residues) and significant enhancements for long loops (11-20 residues). The quality of the predictions is robust to errors that unavoidably affect the stem regions when these are modeled. The method returns a confidence score for the predicted template loops and has the advantage of being very fast (on average: 1 min/loop).www.biocomputing.it/loopinganna.tramontano@uniroma1.itSupplementary data are available at Bioinformatics online.

  20. Development of Electrical Capacitance Sensors for Accident Tolerant Fuel (ATF) Testing at the Transient Reactor Test (TREAT) Facility

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Maolong; Ryals, Matthew; Ali, Amir; Blandford, Edward; Jensen, Colby; Condie, Keith; Svoboda, John; O' Brien, Robert

    2016-08-01

    A variety of instruments are being developed and qualified to support the Accident Tolerant Fuels (ATF) program and future transient irradiations at the Transient Reactor Test (TREAT) facility at Idaho National Laboratory (INL). The University of New Mexico (UNM) is working with INL to develop capacitance-based void sensors for determining the timing of critical boiling phenomena in static capsule fuel testing and the volume-averaged void fraction in flow-boiling in-pile water loop fuel testing. The static capsule sensor developed at INL is a plate-type configuration, while UNM is utilizing a ring-type capacitance sensor. Each sensor design has been theoretically and experimentally investigated at INL and UNM. Experiments are being performed at INL in an autoclave to investigate the performance of these sensors under representative Pressurized Water Reactor (PWR) conditions in a static capsule. Experiments have been performed at UNM using air-water two-phase flow to determine the sensitivity and time response of the capacitance sensor under a flow boiling configuration. Initial measurements from the capacitance sensor have demonstrated the validity of the concept to enable real-time measurement of void fraction. The next steps include designing the cabling interface with the flow loop at UNM for Reactivity Initiated Accident (RIA) ATF testing at TREAT and further characterization of the measurement response for each sensor under varying conditions by experiments and modeling.

  1. Dynamics and control of molten-salt breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-08-15

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

  2. Neutron imaging on the VR-1 reactor

    Science.gov (United States)

    Crha, J.; Sklenka, L.; Soltes, J.

    2016-09-01

    Training reactor VR-1 is a low power research reactor with maximal thermal power of 1 kW. The reactor is operated by the Faculty of Nuclear Science and Physical Engineering of the Czech Technical University in Prague. Due to its low power it suits as a tool for education of university students and training of professionals. In 2015, as part of student research project, neutron imaging was introduced as another type of reactor utilization. The low available neutron flux and the limiting spatial and construction capabilities of the reactor's radial channel led to the development of a special filter/collimator insertion inside the channel and choosing a nonstandard approach by placing a neutron imaging plate inside the channel. The paper describes preliminary experiments carried out on the VR-1 reactor which led to first radiographic images. It seems, that due to the reactor construction and low reactor power, the neutron imaging technique on the VR-1 reactor is feasible mainly for demonstration or educational and training purposes.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-02-01

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

  4. Simulation of German PKL refill/reflood experiment K9A using RELAP4/MOD7. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Hsu, M.T.; Davis, C.B.; Behling, S.R.

    1981-11-01

    This paper describes a RELAP4/MOD7 simulation of West Germany's Kraftwerk Union (KWU) Primary Coolant Loop (PKL) refill/reflood experiment K9A. RELAP4/MOD7, a best-estimate computer program for the calculation of thermal and hydraulic phenomena in a nuclear reactor or related system, is the latest version in the RELAP4 code development series. This study was the first major simulation using RELAP4/MOD7 since its release by the Idaho National Engineering Laboratory (INEL). The PKL facility is a reduced scale (1:134) representation of a typical West German four-loop 1300 MW pressurized water reactor (PWR). A prototypical scale of the total volume to power ratio was maintained. The test facility was designed specifically for an experiment simulating the refill/reflood phase of a Loss-of-Coolant Accident (LOCA).

  5. MANTA. An Integral Reactor Physics Experiment to Infer the Neutron Capture Cross Sections of Actinides and Fission Products in Fast and Epithermal Spectra

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-01

    Neutron cross-sections characterize the way neutrons interact with matter. They are essential to most nuclear engineering projects and, even though theoretical progress has been made as far as the predictability of neutron cross-section models, measurements are still indispensable to meet tight design requirements for reduced uncertainties. Within the field of fission reactor technology, one can identify the following specializations that rely on the availability of accurate neutron cross-sections: (1) fission reactor design, (2) nuclear fuel cycles, (3) nuclear safety, (4) nuclear safeguards, (5) reactor monitoring and neutron fluence determination and (6) waste disposal and transmutation. In particular, the assessment of advanced fuel cycles requires an extensive knowledge of transuranics cross sections. Plutonium isotopes, but also americium, curium and up to californium isotope data are required with a small uncertainty in order to optimize significant features of the fuel cycle that have an impact on feasibility studies (e.g. neutron doses at fuel fabrication, decay heat in a repository, etc.). Different techniques are available to determine neutron cross sections experimentally, with the common denominator that a source of neutrons is necessary. It can either come from an accelerator that produces neutrons as a result of interactions between charged particles and a target, or it can come from a nuclear reactor. When the measurements are performed with an accelerator, they are referred to as differential since the analysis of the data provides the cross-sections for different discrete energies, i.e. σ(Ei), and for the diffusion cross sections for different discrete angles. Another approach is to irradiate a very pure sample in a test reactor such as the Advanced Test Reactor (ATR) at INL and, after a given time, determine the amount of the different transmutation products. The precise characterization of the nuclide densities before and after

  6. Selection of support structure materials for irradiation experiments in the HFIR (High Flux Isotope Reactor) at temperatures up to 500 degrees C

    Energy Technology Data Exchange (ETDEWEB)

    Farrell, K.; Longest, A.W.

    1990-01-01

    The key factor in the design of capsules for irradiation of test specimens in the High Flux Isotope Reactor at preselected temperatures up to 500{degree}C utilizing nuclear heating is a narrow gas-filled gap which surrounds the specimens and controls the transfer of heat from the specimens through the wall of a containment tube to the reactor cooling water. Maintenance of this gap to close tolerances is dependent on the characteristics of the materials used to support the specimens and isolate them from the water. These support structure materials must have low nuclear heating rates, high thermal conductivities, and good dimensional stabilities under irradiation. These conditions are satisfied by certain aluminum alloys. One of these alloys, a powder metallurgy product containing a fine dispersion of aluminum oxide, is no longer manufactured. A new alloys of this type, with the trade name DISPAL, is determined to be a suitable substitute. 23 refs., 13 figs., 3 tabs.

  7. Coupled dual loop absorption heat pump

    Science.gov (United States)

    Sarkisian, Paul H.; Reimann, Robert C.; Biermann, Wendell J.

    1985-01-01

    A coupled dual loop absorption system which utilizes two separate complete loops. Each individual loop operates at three temperatures and two pressures. This low temperature loop absorber and condenser are thermally coupled to the high temperature loop evaporator, and the high temperature loop condenser and absorber are thermally coupled to the low temperature generator.

  8. PREFACE: SANS-YuMO User Meeting at the Start-up of Scientific Experiments on the IBR-2M Reactor: Devoted to the 75th anniversary of Yu M Ostanevich's birth

    Science.gov (United States)

    Gordely, Valentin; Kuklin, Alexander; Balasoiu, Maria

    2012-03-01

    The Second International Workshop 'SANS-YuMO User Meeting at the Start-up of Scientific Experiments on the IBR-2M Reactor', devoted to the 75th anniversary of the birth of Professor Yu M Ostanevich (1936-1992), an outstanding neutron physicist and the founder of small-angle neutron scattering (field, group, and instrument) at JINR FLNPh, was held on 27-30 May at the Frank Laboratory of Neutron Physics. The first Workshop was held in October 2006. Research groups from different neutron centers, universities and research institutes across Europe presented more than 35 oral and poster presentations describing scientific and methodological results. Most of them were obtained with the help of the YuMO instrument before the IBR-2 shutdown in 2006. For the last four years the IBR-2 reactor has been shut down for refurbishment. At the end of 2010 the physical launch of the IBR-2M reactor was finally realized. Nowadays the small-angle neutron scattering (SANS) technique is applied to a wide range of scientific problems in condensed matter, soft condensed matter, biology and nanotechnology, and despite the fact that there are currently over 30 SANS instruments in operation worldwide at both reactor and spallation sources, the demand for beam-time is considerably higher than the time available. It must be remembered, however, that as the first SANS machine on a steady-state reactor was constructed at the Institute Laue Langevin, Grenoble, the first SANS instrument on a 'white' neutron pulsed beam was accomplished at the Joint Institute for Nuclear Research at the IBR-30 reactor, beamline N5. During the meeting Yu M Ostanevich's determinative and crucial contribution to the construction of spectrometers at the IBR-2 high-pulsed reactor was presented, as well as his contribution to the development of the time-of-flight (TOF) small-angle scattering technique, and a selection of other scientific areas. His leadership and outstanding scientific achievements in applications of the

  9. Modeling of pulsed heat load in a cryogenic SHe loop using Artificial Neural Networks

    Science.gov (United States)

    Savoldi Richard, L.; Bonifetto, R.; Carli, S.; Grand Blanc, M.; Zanino, R.

    2013-10-01

    The pulsed heat load to the cryoplant is an important issue in the design and operation of tokamaks adopting superconducting (SC) magnets for the magnetic confinement, as the International Thermonuclear Experimental Reactor (ITER). The smoothing of the heat load during plasma operation is being addressed by experiments, e.g. in the HELIOS facility at CEA Grenoble, and simulations. The assessment of the operation of the cryoplant mainly requires the knowledge of the evolution of the heat load to the liquid helium (LHe) baths that are used as interfaces/buffers between the magnets cooling loops and the cryoplant itself. In this paper, an innovative approach based on Artificial Neural Networks (ANNs) is presented, leading to a simplified but fast model of the transient heat load from the magnets to the LHe baths. An ANN model is developed for the HELIOS loop and the resulting network is trained using detailed transient simulations performed with the 4C code, which was previously extensively validated against experimental data from HELIOS. The predictive capability of the (simplified) ANN model is then demonstrated by considering another, independent dataset, not used during the ANN training, and comparing the evolution of the heat load to the LHe bath computed by the ANNs with that obtained from the (detailed) 4C model.

  10. Capabilities of computer materials science and irradiation experiments for irradiation materials database and design methodology development (based on discussions at the {sup I}EA symposium on fusion reactor materials development)

    Energy Technology Data Exchange (ETDEWEB)

    Jitsukawa, S.; Suzuki, K.; Kaburaki, H. [Japan Atomic Energy Agency, Tokai-mura, Naga-gun, Ibaraki-ken (Japan); Wiffen, F.W.; Stoller, R. [ORNL - Oak Ridge National Laboratory, Materials Science and Technology Div., AK TN (United States); Sharafat, S. [UCLA, Los Angeles, Mechanical and Aerospace Engineering Dept., AK CA (United States)

    2007-07-01

    Full text of publication follows: Irradiation by high-energy fusion neutrons of the first wall of a DEMO blanket introduces transmutation produced helium atoms and displacement damage in the structural material to levels of greater than 1000 appm and 100 displacement per atom during typical service lifetimes, respectively. To simulate high levels of helium atoms in materials, doping techniques using species with large helium producing cross sections are often used in fission reactor irradiation experiments. However, the capability of these techniques is rather limited due to the geometric accumulation of dopants and helium along grain boundaries. In recent years, significant progress in modeling and simulation studies on these irradiation effects using large-scale computational techniques has been achieved. However, further improvements in accuracy and reliability of modeling results are needed prior to application of these results to structural design analyses and licensing. Therefore, the community is anticipating the construction of an intense neutron source, such as the d-Li stripping reaction neutron source of the International Fusion Materials Irradiation Facility (IFMIF). At the recent 'IEA Symposium on Fusion Reactor Materials Development' held Tokyo, Japan 2006, accomplishments of modeling and simulation studies, results of fission neutron and ion irradiation experiments, and a gap between the knowledge of these activities and design methodology of fusion reactor components were presented and discussed. One of the major conclusions of the meeting was that 'IFMIF is an essential facility in the pursuit of a low-risk path to the rapid development of commercially-viable fusion energy (DEMO). To facilitate highly productive IFMIF and supporting efforts (theory, modeling, and computer materials science), it is also important to enhance the activities of materials development, by continuing irradiation experiments in fission reactors and at other

  11. Wilson loops as precursors

    Energy Technology Data Exchange (ETDEWEB)

    Susskind, Leonard [Department of Physics, Stanford University, Stanford, California 94305-4060 (United States); Toumbas, Nicolaos [Department of Physics, Stanford University, Stanford, California 94305-4060 (United States)

    2000-02-15

    There is substantial evidence that string theory on AdS{sub 5}xS{sub 5} is a holographic theory in which the number of degrees of freedom scales as the area of the boundary in Planck units. Precisely how the theory can describe bulk physics using only surface degrees of freedom is not well understood. A particularly paradoxical situation involves an event deep in the interior of the bulk space. The event must be recorded in the (Schroedinger picture) state vector of the boundary theory long before a signal, such as a gravitational wave, can propagate from the event to the boundary. In a previous paper with Polchinski, we argued that the ''precursor'' operators which carry information stored in the wave during the time when it vanishes in a neighborhood of the boundary are necessarily non-local. In this paper we argue that the precursors cannot be products of local gauge invariant operators such as the energy momentum tensor. In fact gauge theories have a class of intrinsically non-local operators which cannot be built from local gauge invariant objects. These are the Wilson loops. We show that the precursors can be identified with Wilson loops whose spatial size is dictated by the UV-IR connection. (c) 2000 The American Physical Society.

  12. High temperature storage loop :

    Energy Technology Data Exchange (ETDEWEB)

    Gill, David Dennis; Kolb, William J.

    2013-07-01

    A three year plan for thermal energy storage (TES) research was created at Sandia National Laboratories in the spring of 2012. This plan included a strategic goal of providing test capability for Sandia and for the nation in which to evaluate high temperature storage (>650ÀC) technology. The plan was to scope, design, and build a flow loop that would be compatible with a multitude of high temperature heat transfer/storage fluids. The High Temperature Storage Loop (HTSL) would be reconfigurable so that it was useful for not only storage testing, but also for high temperature receiver testing and high efficiency power cycle testing as well. In that way, HTSL was part of a much larger strategy for Sandia to provide a research and testing platform that would be integral for the evaluation of individual technologies funded under the SunShot program. DOEs SunShot program seeks to reduce the price of solar technologies to 6/kWhr to be cost competitive with carbon-based fuels. The HTSL project sought to provide evaluation capability for these SunShot supported technologies. This report includes the scoping, design, and budgetary costing aspects of this effort

  13. High-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1980

    Energy Technology Data Exchange (ETDEWEB)

    1981-08-01

    Research activities are described concerning HTGR chemistry; fueled graphite development; prestressed concrete pressure vessel development; structural materials; HTGR graphite studies; HTR core evaluation; reactor physics; shielding; application and project assessments; and HTR Core Flow Test Loop studies.

  14. 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.

  15. 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.

  16. 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.

  17. Liquid metal reactor development -Studies on safety measure of LMR coolant

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Sung Tae; Choi, Yoon Dong; Park, Jin Hoh; Kwon, Sun Kil; Choi, Jong Hyun; Cho, Byung Ryul; Kim, Tae Joon; Kwon, Sang Woon; Jung, Kyung Chae; Kim, Byung Hoh; Hong, Soon Bok; Jung, Ji Yung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    A study on the safety measures of LMR coolant showed the results as follows; 1. LMR coolant safety measure. A. Analysis and improvement of sodium fire code. B. Analysis of sodium fire phenomena. 2. Sodium fire aerosol characteristics. It was carried out conceptual design and basic design for sodium fire facility of medium size composed of sodium supply tank, sodium reactor vessel, sodium fire aerosol filter system and scrubbing column, and drain tank etc. 3. Sodium purification technology. A. Construction of calibration loop. (1) Design of sodium loop for the calibration of the equipment. (2) Construction of sodium loop including test equipments and other components. B. Na-analysis technology. (1) Oxygen concentration determination by the wet method. (2) Cover gas purification preliminary experiment. 4. The characteristics of sodium-water reaction. A. Analysis of the micro and small leak phenomena. (1) Manufacture of the micro-leak test apparatus. B. Analysis of large leak events. (1) Development of preliminary code for analysis of initial spike pressure. (2) Sample calculation and comparison with previous works. C. Development of test facility for large leak event evaluation. (1) Conceptional and basic design for the water and sodium-water test facility. D. Technology development for water leak detection system. (1) Investigations for the characteristics of active acoustic detection system. (2) Testing of the characteristics of hydrogen leak detection system. 171 figs, 29 tabs, 3 refs. (Author).

  18. Gasification Coupled Chemical Looping Combustion of Coal: A Thermodynamic Process Design Study

    OpenAIRE

    Borkhade, Sonali A.; Shriwas, Preksha A.; Ganesh R. Kale

    2013-01-01

    A thermodynamic investigation of gasification coupled chemical looping combustion (CLC) of carbon (coal) is presented in this paper. Both steam and CO2 are used for gasification within the temperature range of 500–1200°C. Chemical equilibrium model was considered for the gasifier and CLC fuel reactor. The trends in product compositions and energy requirements of the gasifier, fuel reactor, and air reactor were determined. Coal (carbon) gasification using 1.5 mol H2O and 1.5 mol CO2 per mole c...

  19. Experiment and modeling of CO{sub 2} capture from flue gases at high temperature in a fluidized bed reactor with Ca-based sorbents

    Energy Technology Data Exchange (ETDEWEB)

    Fan Fang; Zhen-Shan Li; Ning-Sheng Cai [Tsinghua University, Beijing (China). Key Laboratory for Thermal Science and Power Engineering of the Ministry of Education (MOE)

    2009-01-15

    The cyclic CO{sub 2} capture and CaCO{sub 3} regeneration characteristics in a small fluidized bed reactor were experimentally investigated with limestone and dolomite sorbents. Kinetic rate constants for carbonation and calcination were determined using thermogravimetric analysis (TGA) data. Mathematical models developed to model the Ca-based sorbent multiple cycles of CO{sub 2} capture and calcination in the bubbling fluidized bed reactor agreed with the experimental data. The experimental and simulated results showed that the CO{sub 2} in flue gases could be absorbed efficiently by limestone and dolomite. The time for high-efficiency CO{sub 2} capture decreased with an increasing number of cycles because of the loss of sorbent activity, and the final CO{sub 2} capture efficiency remained nearly constant as the sorbent reached its final residual capture capacity. In a continuous carbonation and calcination system, corresponding to the sorbent activity loss, the carbonation kinetic rates of sorbent undergoing various cycles are different, and the carbonation kinetic rates of sorbent circulating N times in the carbonation/calcination cycles are also different because of the different residence time of sorbent in the carbonator. Therefore, the average carbonation rate was given based on the mass balance and exit age distribution for sorbent in the carbonator. The CO{sub 2} capture characteristics in a continuous carbonation/calcination system were predicted, taking into consideration the mass balance, sorbent circulation rate, sorbent activity loss, and average carbonation kinetic rate, to give useful information for the reactor design and operation of multiple carbonation/calcination reaction cycles. 27 refs., 15 figs., 1 tab.

  20. Investigation of molten salt fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-06-01

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

  1. 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.

  2. Open-loop and closed-loop control of flying qubits

    Energy Technology Data Exchange (ETDEWEB)

    Lucamarini, M; Di Giuseppe, G; Vitali, D; Tombesi, P, E-mail: marco.lucamarini@unicam.it [School of Science and Technology, Physics Division, University of Camerino, I-62032 Camerino (Italy)

    2011-08-14

    We describe two recent techniques, along with related experiments, to control and reduce the noise affecting a photon polarization qubit. The first is based on the open-loop 'bang-bang' method, where suitably tailored pulses are implemented on the system to prevent polarization decoherence. This requires only passive elements when the physical system is a photon and the operation is performed in space rather than in time. The second technique is based on closed-loop 'asymmetric feedback', where some quantities are measured and used for a real-time correction of the system dynamics. This technique necessarily requires active electronics to work.

  3. Structure prediction of loops with fixed and flexible stems.

    Science.gov (United States)

    Subramani, A; Floudas, C A

    2012-06-14

    The prediction of loop structures is considered one of the main challenges in the protein folding problem. Regardless of the dependence of the overall algorithm on the protein data bank, the flexibility of loop regions dictates the need for special attention to their structures. In this article, we present algorithms for loop structure prediction with fixed stem and flexible stem geometry. In the flexible stem geometry problem, only the secondary structure of three stem residues on either side of the loop is known. In the fixed stem geometry problem, the structure of the three stem residues on either side of the loop is also known. Initial loop structures are generated using a probability database for the flexible stem geometry problem, and using torsion angle dynamics for the fixed stem geometry problem. Three rotamer optimization algorithms are introduced to alleviate steric clashes between the generated backbone structures and the side chain rotamers. The structures are optimized by energy minimization using an all-atom force field. The optimized structures are clustered using a traveling salesman problem-based clustering algorithm. The structures in the densest clusters are then utilized to refine dihedral angle bounds on all amino acids in the loop. The entire procedure is carried out for a number of iterations, leading to improved structure prediction and refined dihedral angle bounds. The algorithms presented in this article have been tested on 3190 loops from the PDBSelect25 data set and on targets from the recently concluded CASP9 community-wide experiment.

  4. Loop Quantum Gravity.

    Science.gov (United States)

    Rovelli, Carlo

    2008-01-01

    The problem of describing the quantum behavior of gravity, and thus understanding quantum spacetime, is still open. Loop quantum gravity is a well-developed approach to this problem. It is a mathematically well-defined background-independent quantization of general relativity, with its conventional matter couplings. Today research in loop quantum gravity forms a vast area, ranging from mathematical foundations to physical applications. Among the most significant results obtained so far are: (i) The computation of the spectra of geometrical quantities such as area and volume, which yield tentative quantitative predictions for Planck-scale physics. (ii) A physical picture of the microstructure of quantum spacetime, characterized by Planck-scale discreteness. Discreteness emerges as a standard quantum effect from the discrete spectra, and provides a mathematical realization of Wheeler's "spacetime foam" intuition. (iii) Control of spacetime singularities, such as those in the interior of black holes and the cosmological one. This, in particular, has opened up the possibility of a theoretical investigation into the very early universe and the spacetime regions beyond the Big Bang. (iv) A derivation of the Bekenstein-Hawking black-hole entropy. (v) Low-energy calculations, yielding n-point functions well defined in a background-independent context. The theory is at the roots of, or strictly related to, a number of formalisms that have been developed for describing background-independent quantum field theory, such as spin foams, group field theory, causal spin networks, and others. I give here a general overview of ideas, techniques, results and open problems of this candidate theory of quantum gravity, and a guide to the relevant literature.

  5. Loop Quantum Gravity

    Directory of Open Access Journals (Sweden)

    Rovelli Carlo

    2008-07-01

    Full Text Available The problem of describing the quantum behavior of gravity, and thus understanding quantum spacetime, is still open. Loop quantum gravity is a well-developed approach to this problem. It is a mathematically well-defined background-independent quantization of general relativity, with its conventional matter couplings. Today research in loop quantum gravity forms a vast area, ranging from mathematical foundations to physical applications. Among the most significant results obtained so far are: (i The computation of the spectra of geometrical quantities such as area and volume, which yield tentative quantitative predictions for Planck-scale physics. (ii A physical picture of the microstructure of quantum spacetime, characterized by Planck-scale discreteness. Discreteness emerges as a standard quantum effect from the discrete spectra, and provides a mathematical realization of Wheeler’s “spacetime foam” intuition. (iii Control of spacetime singularities, such as those in the interior of black holes and the cosmological one. This, in particular, has opened up the possibility of a theoretical investigation into the very early universe and the spacetime regions beyond the Big Bang. (iv A derivation of the Bekenstein–Hawking black-hole entropy. (v Low-energy calculations, yielding n-point functions well defined in a background-independent context. The theory is at the roots of, or strictly related to, a number of formalisms that have been developed for describing background-independent quantum field theory, such as spin foams, group field theory, causal spin networks, and others. I give here a general overview of ideas, techniques, results and open problems of this candidate theory of quantum gravity, and a guide to the relevant literature.

  6. Three-dimensional reactor dynamics code for VVER type nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kyrki-Rajamaeki, R. [VTT Energy, Espoo (Finland)

    1995-10-01

    A three-dimensional reactor dynamics computer code has been developed, validated and applied for transient and accident analyses of VVER type nuclear reactors. This code, HEXTRAN, is a part of the reactor physics and dynamics calculation system of the Technical Research Centre of Finland, VTT. HEXTRAN models accurately the VVER core with hexagonal fuel assemblies. The code uses advanced mathematical methods in spatial and time discretization of neutronics, heat transfer and the two-phase flow equations of hydraulics. It includes all the experience of VTT from 20 years on the accurate three-dimensional static reactor physics as well as on the one-dimensional reactor dynamics. The dynamic coupling with the thermal hydraulic system code SMABRE also allows the VVER circuit-modelling experience to be included in the analyses. (79 refs.).

  7. Experience in Remote Demolition of the Activated Biological Shielding of the Multi Purpose Research Reactor (MZFR) on the German Karlsruhe Site - 12208

    Energy Technology Data Exchange (ETDEWEB)

    Eisenmann, Beata; Fleisch, Joachim; Prechtl, Erwin; Suessdorf, Werner; Urban, Manfred [WAK Rueckbau- und Entsorgungs- GmbH, P.O.Box 12 63, 76339 Eggenstein-Leopoldshafen (Germany)

    2012-07-01

    In 2009, WAK Decommissioning and Waste Management GmbH (WAK) became owner and operator of the waste treatment facilities of Karlsruhe Institute of Technology (KIT) as well as of the prototype reactors, the Compact Sodium-Cooled Fast Reactor (KNK) and Multi-Purpose Reactor (MZFR), both being in an advanced stage of dismantling. Together with the dismantling and decontamination activities of the former WAK reprocessing facility since 1990, the envisaged demolishing of the R and D reactor FR2 and a hot cell facility, all governmentally funded nuclear decommissioning projects on the Karlsruhe site are concentrated under the WAK management. The small space typical of prototype research reactors represented a challenge also during the last phase of activated dismantling, dismantling of the activated biological shield of the MZFR. Successful demolition of the biological shield required detailed planning and extensive testing in the years before. In view of the limited space and the ambient dose rate that was too high for manual work, it was required to find a tool carrier system to take up and control various demolition and dismantling tools in a remote manner. The strategy formulated in the concept of dismantling the biological shield by means of a modified electro-hydraulic demolition excavator in an adaptable working scaffolding turned out to be feasible. The following boundary conditions were essential: - Remote exchange of the dismantling and removal tools in smallest space. - Positioning of various supply facilities on the working platform. - Avoiding of interfering edges. - Optimization of mass flow (removal of the dismantled mass from the working area). - Maintenance in the surroundings of the dismantling area (in the controlled area). - Testing and qualification of the facilities and training of the staff. Both the dismantling technique chosen and the proceeding selected proved to be successful. Using various designs of universal cutters developed on the basis of

  8. The AFR. An approved network of research reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-10-15

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

  9. 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.

  10. 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

  11. Facility for a Low Power Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chalker, R. G.

    1949-09-14

    Preliminary investigation indicates that a reactor facility with ample research provisions for use by University or other interested groups, featuring safety in design, can be economically constructed in the Los Angeles area. The complete installation, including an underground gas-tight reactor building, with associated storage and experiment assembly building, administration offices, two general laboratory buildings, hot latoratory and lodge, can be constructed for approxinately $1,500,000. This does not include the cost of the reactor itself or of its auxiliary equipment,

  12. 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.

  13. Loop quantum cosmology: Recent progress

    Indian Academy of Sciences (India)

    Aspects of the full theory of loop quantum gravity can be studied in a simpler context by reducing to symmetric models like cosmological ones. This leads to several applications where loop effects play a significant role when one is sensitive to the quantum regime. As a consequence, the structure of and the approach to ...

  14. The primary loop confinement and pressure boundary system of the HTR-10

    Energy Technology Data Exchange (ETDEWEB)

    He Shuyan; Zhang Zhengming E-mail: zhengmin@inet.tsinghua.edu.cn; Qin Zhenya

    2002-10-01

    The design philosophy and requirements of the HTR-10 reactor building and the primary loop confinement are introduced in this paper. Also introduced are the design, fabrication and the installation of the HTR-10 primary loop pressure boundary system. The primary loop confinement comprises the sealed cavities of the reinforced concrete structure. The main components and the connected gas systems of the primary loop pressure boun