Characterization of the TRIGA Mark II reactor full-power steady state

Energy Technology Data Exchange (ETDEWEB)

Cammi, Antonio, E-mail: antonio.cammi@polimi.it [Politecnico di Milano – Department of Energy, CeSNEF (Enrico Fermi Center for Nuclear Studies), via La Masa 34, 20156 Milano (Italy); Zanetti, Matteo [Politecnico di Milano – Department of Energy, CeSNEF (Enrico Fermi Center for Nuclear Studies), via La Masa 34, 20156 Milano (Italy); Chiesa, Davide; Clemenza, Massimiliano; Pozzi, Stefano; Previtali, Ezio; Sisti, Monica [University of Milano-Bicocca, Physics Department “G. Occhialini” and INFN Section, Piazza dell’Ateneo Nuovo, 20126 Milan (Italy); Magrotti, Giovanni; Prata, Michele; Salvini, Andrea [University of Pavia, Applied Nuclear Energy Laboratory (L.E.N.A.), Via Gaspare Aselli 41, 27100 Pavia (Italy)

2016-04-15

Highlights: • Full-power steady state characterization of the TRIGA Mark II reactor. • Monte Carlo and Multiphysics simulation of the TRIGA Mark II reactor. • Sub-cooled boiling effects in the TRIGA Mark II reactor. • Thermal feedback effects in the TRIGA Mark II reactor. • Experimental data based validation. - Abstract: In this paper, the characterization of the full-power steady state of the TRIGA Mark II nuclear reactor at the University of Pavia is achieved by coupling the Monte Carlo (MC) simulation for neutronics with the “Multiphysics” model for thermal-hydraulics. Neutronic analyses have been carried out with a MCNP5 based MC model of the entire reactor system, already validated in fresh fuel and zero-power configurations (in which thermal effects are negligible) and using all available experimental data as a benchmark. In order to describe the full-power reactor configuration, the temperature distribution in the core must be established. To evaluate this, a thermal-hydraulic model has been developed, using the power distribution results from the MC simulation as input. The thermal-hydraulic model is focused on the core active region and takes into account sub-cooled boiling effects present at full reactor power. The obtained temperature distribution is then entered into the MC model and a benchmark analysis is carried out to validate the model in fresh fuel and full-power configurations. An acceptable correspondence between experimental data and simulation results concerning full-power reactor criticality proves the reliability of the adopted methodology of analysis, both from the perspective of neutronics and thermal-hydraulics.

Progress on advanced tokamak and steady-state scenario development on DIII-D and NSTX

Energy Technology Data Exchange (ETDEWEB)

Doyle, E J [Department of Electrical Engineering and PSTI, University of California, Los Angeles, California 90095 (United States); Garofalo, A M [Columbia University, New York, New York 10027 (United States); Greenfield, C M [General Atomics, San Diego, California 92186-5608 (United States); Kaye, S M [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543-0451 (United States); Menard, J E [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543-0451 (United States); Murakami, M [Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831 (United States); Sabbagh, S A [Columbia University, New York, New York 10027 (United States); Austin, M E [University of Texas-Austin, Austin, Texas 78712 (United States); Bell, R E [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543-0451 (United States); Burrell, K H [General Atomics, San Diego, California 92186-5608 (United States); Ferron, J R [General Atomics, San Diego, California 92186-5608 (United States); Gates, D A [Princeton Plasma Physics Lab., Princeton, New Jersey 08543-0451 (United States); Groebner, R J; Hyatt, A W; Luce, T C; Petty, C C; Wade, M R; Waltz, R E [General Atomics, San Diego, California 92186-5608 (United States); Jayakumar, R J [Lawrence Livermore National Lab., Livermore, California 94550 (United States); Kinsey, J E [Lehigh Univ., Bethlehem, Pennsylvania 18015 (United States); LeBlanc, B P [Princeton Plasma Physics Lab., Princeton, New Jersey 08543-0451 (United States); McKee, G R [Univ. of Wisconsin-Madison, Madison, Wisconsin 53706 (United States); Okabayashi, M [Princeton Plasma Physics Lab., Princeton, New Jersey 08543-0451 (United States); Peng, Y-K M [Oak Ridge National Lab., Oak Ridge, Tennessee 37831 (United States); Politzer, P A [General Atomics, San Diego, California 92186-5608 (United States); Rhodes, T L [Dept. of Electrical Engineering and PSTI, Univ. of California, Los Angeles, California 90095 (United States)

2006-12-15

Advanced tokamak (AT) research seeks to develop steady-state operating scenarios for ITER and other future devices from a demonstrated scientific basis. Normalized target parameters for steady-state operation on ITER are 100% non-inductive current operation with a bootstrap current fraction f{sub BS} {>=} 60%, q{sub 95} {approx} 4-5 and G {identical_to}{beta}{sub N}H{sub scaling}/q{sub 95}{sup 2} {>=}0.3. Progress in realizing such plasmas is considered in terms of the development of plasma control capabilities and scientific understanding, leading to improved AT performance. NSTX has demonstrated active resistive wall mode stabilization with low, ITER-relevant, rotation rates below the critical value required for passive stabilization. On DIII-D, experimental observations and GYRO simulations indicate that ion internal transport barrier (ITB) formation at rational-q surfaces is due to equilibrium zonal flows generating high local E ? B shear levels. In addition, stability modelling for DIII-D indicates a path to operation at {beta}{sub N} {>=} 4 with q{sub min} {>=} 2, using broad, hollow current profiles to increase the ideal wall stability limit. Both NSTX and DIII-D have optimized plasma performance and expanded AT operational limits. NSTX now has long-pulse, high performance discharges meeting the normalized targets for an spherical torus-based component test facility. DIII-D has developed sustained discharges combining high beta and ITBs, with performance approaching levels required for AT reactor concepts, e.g. {beta}{sub N} = 4, H{sub 89} = 2.5, with f{sub BS} > 60%. Most importantly, DIII-D has developed ITER steady-state demonstration discharges, simultaneously meeting the targets for steady-state Q {>=} 5 operation on ITER set out above, substantially increasing confidence in ITER meeting its steady-state performance objective.

New steady-state microbial community compositions and process performances in biogas reactors induced by temperature disturbances

DEFF Research Database (Denmark)

Luo, Gang; De Francisci, Davide; Kougias, Panagiotis

2015-01-01

that stochastic factors had a minor role in shaping the profile of the microbial community composition and activity in biogas reactors. On the contrary, temperature disturbance was found to play an important role in the microbial community composition as well as process performance for biogas reactors. Although...... three different temperature disturbances were applied to each biogas reactor, the increased methane yields (around 10% higher) and decreased volatile fatty acids (VFAs) concentrations at steady state were found in all three reactors after the temperature disturbances. After the temperature disturbance...... in shaping the profile of the microbial community composition and activity in biogas reactors. New steady-state microbial community profiles and reactor performances were observed in all the biogas reactors after the temperature disturbance....

Vulcan: A steady-state tokamak for reactor-relevant plasma–material interaction science

International Nuclear Information System (INIS)

Olynyk, G.M.; Hartwig, Z.S.; Whyte, D.G.; Barnard, H.S.; Bonoli, P.T.; Bromberg, L.; Garrett, M.L.; Haakonsen, C.B.; Mumgaard, R.T.; Podpaly, Y.A.

2012-01-01

Highlights: ► A new scaling for obtaining reactor similarity in the divertor of scaled tokamaks. ► Conceptual design for a tokamak (“Vulcan”) to implement this new scaling. ► Demountable superconducting coils and compact neutron shielding. ► Helium-cooled high-temperature vacuum vessel and first wall. ► High-field-side lower hybrid current drive for non-inductive operation. - Abstract: An economically viable magnetic-confinement fusion reactor will require steady-state operation and high areal power density for sufficient energy output, and elevated wall/blanket temperatures for efficient energy conversion. These three requirements frame, and couple to, the challenge of plasma–material interaction (PMI) for fusion energy sciences. Present and planned tokamaks are not designed to simultaneously meet these criteria. A new and expanded set of dimensionless figures of merit for PMI have been developed. The key feature of the scaling is that the power flux across the last closed flux surface P/S ≃ 1 MW m −2 is to be held constant, while scaling the core volume-averaged density weakly with major radius, n ∼ R −2/7 . While complete similarity is not possible, this new “P/S” or “PMI” scaling provides similarity for the most critical reactor PMI issues, compatible with sufficient current drive efficiency for non-inductive steady-state core scenarios. A conceptual design is developed for Vulcan, a compact steady-state deuterium main-ion tokamak which implements the P/S scaling rules. A zero-dimensional core analysis is used to determine R = 1.2 m, with a conventional reactor aspect ratio R/a = 4.0, as the minimum feasible size for Vulcan. Scoping studies of innovative fusion technologies to support the Vulcan PMI mission were carried out for three critical areas: a high-temperature, helium-cooled vacuum vessel and divertor design; a demountable superconducting toroidal field magnet system; and a steady-state lower hybrid current drive system

Steady-state and loss-of-pumping accident analyses of the Savannah River new production reactor representative design

International Nuclear Information System (INIS)

Pryor, R.J.; Maloney, K.J.

1990-10-01

This document contains the steady-state and loss-of-pumping accident analysis of the representative design for the Savannah River heavy water new production reactor. A description of the reactor system and computer input model, the results of the steady-state analysis, and the results of four loss-of-pumping accident calculations are presented. 5 refs., 37 figs., 4 tabs

Plasma control issues for an advanced steady state tokamak reactor

International Nuclear Information System (INIS)

Moreau, D.

2001-01-01

This paper deals with specific control issues related to the advanced tokamak scenarios in which rather accurate tailoring of the current density profile is a requirement in connection with the steady state operation of a reactor in a high confinement optimized shear mode. It is found that adequate current profile control can be performed if real-time magnetic flux reconstruction is available through a set of dedicated diagnostics and computers, with sufficient accuracy to deduce the radial profile of the safety factor and of the internal plasma loop voltage. It is also shown that the safety factor can be precisely controlled in the outer half of the plasma through the surface loop voltage and the off-axis current drive power, but that a compromise must be made between the accuracy of the core safety factor control and the total duration of the current and fuel density ramp-up phases, so that the demonstration of the steady state reactor potential of the optimized/reversed shear concept in the Next Step device will demand pulse lengths of the order of one thousand seconds (or more for an ITER-size machine). (author)

Plasma control issues for an advanced steady state tokamak reactor

International Nuclear Information System (INIS)

Moreau, D.; Voitsekhovitch, I.

1999-01-01

This paper deals with specific control issues related to the advanced tokamak scenarios in which rather accurate tailoring of the current density profile is a requirement in connection with the steady state operation of a reactor in a high confinement optimized shear mode. It is found that adequate current profile control can be performed if real-time magnetic flux reconstruction is available through a set of dedicated diagnostics and computers, with sufficient accuracy to deduce the radial profile of the safety factor and of the internal plasma loop voltage. It is also shown that the safety factor can be precisely controlled in the outer half of the plasma through the surface loop voltage and the off-axis current drive power, but that a compromise must be made between the accuracy of the core safety factor control and the total duration of the current and fuel density ramp-up phases, so that the demonstration of the steady state reactor potential of the optimized/reversed shear concept in the Next Step device will demand pulse lengths of the order of one thousand seconds (or more for an ITER-size machine). (author)

Parametric study of the primary and secondary systems of the CAREM-25 reactor on steady state

International Nuclear Information System (INIS)

Halpert, Silvia; Vazquez, Luis

2000-01-01

In the CAREM-25 reactor the primary coolant flows by natural convection that's why the flow is established when the balance between the buoyancy force and friction pressure drop through circuit is obtained. This paper presents a parametric study on primary and secondary systems of the reactor on steady state, for different values of some thermohydraulics parameters: safety factor on friction loss pressure calculations (f), steam generator heat transfer area (A T ) and primary pressure (P P ). The ESCAREM 2.08 thermohydraulic code, which calculates the primary system behavior for steady state conditions, was used for this study. The conclusions of this study are: -) There was a variation of the 15% on the primary coolant flow when the safety factor was changed a 50 %; -) The primary and secondary systems conditions do not change when the power is less than 100 MW; -) Between 100 and 110 MW the decrease of the heat transfer area produces an important change on the secondary systems conditions: the outlet steam generator temperature decrease and there is an important rice in the flow; -) The primary pressure could decrease up to 11.4 MPa without violating turbine requirements. (author)

Steady-state operation requirements of tokamak fusion reactor concepts

International Nuclear Information System (INIS)

Knobloch, A.F.

1991-06-01

In the last two decades tokamak conceptual reactor design studies have been deriving benefit from progressing plasma physics experiments, more depth in theory and increasing detail in technology and engineering. Recent full-scale reactor extrapolations such as the US ARIES-I and the EC Reference Reactor study provide information on rather advanced concepts that are called for when economic boundary conditions are imposed. The ITER international reactor design activity concentrated on defining the next step after the JET generation of experiments. For steady-state operation as required for any future commercial tokamak fusion power plants it is essential to have non-inductive current drive. The current drive power and other internal power requirements specific to magnetic confinement fusion have to be kept as low as possible in order to attain a competitive overall power conversion efficiency. A high plasma Q is primarily dependent on a high current drive efficiency. Since such conditions have not yet been attained in practice, the present situation and the degree of further development required are characterized. Such development and an appropriately designed next-step tokamak reactor make the gradual realization of high-Q operation appear feasible. (orig.)

Experimental and theoretical comparison of fuel temperature and bulk coolant characteristics in the Oregon State TRIGA reactor during steady state operation

Energy Technology Data Exchange (ETDEWEB)

Marcum, W.R., E-mail: marcumw@engr.orst.ed [Oregon State University, Department of Nuclear Engineering and Radiation Health Physics, 116 Radiation Center, Corvallis, OR 97330 (United States); Woods, B.G.; Reese, S.R. [Oregon State University, Department of Nuclear Engineering and Radiation Health Physics, 116 Radiation Center, Corvallis, OR 97330 (United States)

2010-01-15

In September of 2008 Oregon State University (OSU) completed its core conversion analysis as part of the Reduced Enrichment for Research and Test Reactors (RERTR) Program. Experimental bulk coolant temperatures were collected in various locations throughout the Oregon State TRIGA Reactor (OSTR) core in order to supplement the validity of the numerical thermal hydraulic results produced in RELAP5-3D Version 2.4.2. Axial bulk coolant temperature distributions were collected by acquiring discrete thermocouple measurements in individual subchannel locations during steady state operation at 1.0 MW{sub th}. The experimental axial temperature distribution collected was compared to one-channel, two-channel, and eight-channel RELAP5-3D models and found to match within 11.94%, 11.69%, and 8.78%, respectively, on average. Comparisons to similar studies were made based on a dimensional analysis of fluid body forces in the discrete core locations, indicating that the chosen approach produces conservative results for use in the OSTR safety analysis.

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

International Nuclear Information System (INIS)

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

1997-11-01

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

Tokamak Fusion Test Reactor D-T results

International Nuclear Information System (INIS)

Meade, D.M.

1995-01-01

Temperatures, densities and confinement of deuterium plasmas confined in tokamaks have been achieved within the last decade that are approaching those required for a D-T reactor. As a result, the unique phenomena present in a D-T reactor plasma (D-T plasma confinement, α confinement, α heating and possible α-driven instabilities) can now be studied in the laboratory. Recent experiments on the Tokamak Fusion Test Reactor (TFTR) have been the first magnetic fusion experiments to study plasmas with reactor fuel concentrations of tritium. The injection of about 20MW of tritium and 14MW of deuterium neutral beams into the TFTR produced a plasma with a T-to-D density ratio of about 1 and yielding a maximum fusion power of about 9.2MW. The fusion power density in the core of the plasma was about 1.8MWm -3 , approximating that expected in a D-T fusion reactor. A TFTR plasma with a T-to-D density ratio of about 1 was found to have about 20% higher energy confinement time than a comparable D plasma, indicating a confinement scaling with average ion mass A of τ E ∝A 0.6 . The core ion temperature increased from 30 to 37keV owing to a 35% improvement of ion thermal conductivity. Using the electron thermal conductivity from a comparable deuterium plasma, about 50% of the electron temperature increase from 9 to 10.6keV can be attributed to electron heating by the α particles. The approximately 5% loss of α particles, as observed on detectors near the bottom edge of the plasma, was consistent with classical first orbit loss without anomalous effects. Initial measurements have been made of the confined high energy α particles and the resultant α ash density. At fusion power levels of 7.5MW, fluctuations at the toroidal Alfven eigen-mode frequency were observed by the fluctuation diagnostics. However, no additional α loss due to the fluctuations was observed. (orig.)

Calculations of steady-state and reactivity insertion transients in a research reactor simulating the PWR

International Nuclear Information System (INIS)

Mladin, Mirea; Mladin, Daniela; Prodea, Ilie

2010-01-01

In 2008, IAEA started a Coordinated Research Project for benchmarking the thermalhydraulic and neutronic computer codes for research reactor analysis against the experimental data. In this framework, for the first year of research contract, the Institute for Nuclear Research engaged in steady-state analysis of SPERT-III reactor and also in the simulation of the reactivity insertion tests performed in this reactor during mid sixties. In the first part, the paper describes a Monte Carlo input model of the oxide core selected for investigation and the results of the steady-state neutronic calculations with respect to hot and cold core reactivity excess and control rods worth. Also, prompt neutron life and reactivity feed-back coefficients were examined. These results were compared with the data provided in the reactor specification document concerning neutronic design calculated data. The second part of the paper is dedicated to calculation of the reactivity insertion transients with RELAP5 and CATHARE2 thermalhydraulic codes, both including point reactor kinetics models, and to comparison with experimental data. (authors)

System and method for generating steady state confining current for a toroidal plasma fusion reactor

International Nuclear Information System (INIS)

Bers, A.

1981-01-01

A system for generating steady state confining current for a toroidal plasma fusion reactor providing steady-state generation of the thermonuclear power. A dense, hot toroidal plasma is initially prepared with a confining magnetic field with toroidal and poloidal components. Continuous wave rf energy is injected into said plasma to estalish a spectrum of traveling waves in the plasma, where the traveling waves have momentum components substantially either all parallel, or all anti-parallel to the confining magnetic field. The injected rf energy is phased to couple to said traveling waves with both a phase velocity component and a wave momentum component in the direction of the plasma traveling wave components. The injected rf energy has a predetermined spectrum selected so that said traveling waves couple to plasma electrons having velocities in a predetermined range delta . The velocities in the range are substantially greater than the thermal electron velocity of the plasma. In addition, the range is sufficiently broad to produce a raised plateau having width delta in the plasma electron velocity distribution so that the plateau electrons provide steady-state current to generate a poloidal magnetic field component sufficient for confining the plasma. In steady state operation of the fusion reactor, the fusion power density in the plasma exceeds the power dissipated inthe plasma

Steady-state resistive toroidal-field coils for tokamak reactors

International Nuclear Information System (INIS)

Kalnavarns, J.; Jassby, D.L.

1979-12-01

If spatially-averaged values of the beta ratio can reach 5 to 10% in tokamaks, as now seems likely, resistive toroidal-field coils may be advantageous for use in reactors intended for fusion-neutron applications. The present investigation has parameterized the design of steady-state water-cooled copper coils of rectangular cross section in order to maximize figures of merit such as the ratio of fusion neutron wall loading to coil power dissipation. Four design variations distinguished by different ohmic-heating coil configurations have been examined. For a wall loading of 0.5 MW/m 2 , minimum TF-coil lifetime costs (including capital and electricity costs) are found to occur with coil masses in the range 2400 to 4400 tons, giving 200 to 250 MW of resistive dissipation, which is comparable with the total power drain of the other reactor subsystems

A method for statistical steady state thermal analysis of reactor cores

International Nuclear Information System (INIS)

Whetton, P.A.

1981-01-01

In a previous publication the author presented a method for undertaking statistical steady state thermal analyses of reactor cores. The present paper extends the technique to an assessment of confidence limits for the resulting probability functions which define the probability that a given thermal response value will be exceeded in a reactor core. Establishing such confidence limits is considered an integral part of any statistical thermal analysis and essential if such analysis are to be considered in any regulatory process. In certain applications the use of a best estimate probability function may be justifiable but it is recognised that a demonstrably conservative probability function is required for any regulatory considerations. (orig.)

Continuous cryopump for steady state mirror fusion reactors

International Nuclear Information System (INIS)

Batzer, T.H.; Call, W.R.

1983-01-01

The characteristics of mirror fusion reactors, i.e., steady state operation, a low neutral gas density, and a large gas throughput require unique vacuum pumping capabilities. One approach that appears to meet these requirements is a liquid helium-cooled cryopump system in which a fixed portion can be isolated and degassed while the remainder continues to pump. The time to degas a rotating, fixed portion of the pumping area and the ratio of that area to the total area fixes the gas inventory in the chamber. It follows that the active pump area maintains the required neutral gas density and the time-averaged degassing rate equals the gas throughput. We have built such a cryopump whereby the gas condensed (deuterium) on the liquid helium-cooled panel can be transferred to a collector pump and subsequently to an exterior mechanical pump and exhausted. At panel loadings as high as 0.55 Torr-/lcm 2 the gas leakage during degassing is less than 8% and the degassing time is less than 10 min. Scaling to reactor size appears to be feasible

System and method for generating steady state confining current for a toroidal plasma fusion reactor

International Nuclear Information System (INIS)

Fisch, N.J.

1981-01-01

A system for generating steady state confining current for a toroidal plasma fusion reactor providing steady-state generation of the thermonuclear power. A dense, hot toroidal plasma is initially prepared with a confining magnetic field with toroidal and poloidal components. Continuous wave rf energy is injected into said plasma to establish a spectrum of traveling waves in the plasma, where the traveling waves have momentum components substantially either all parallel, or all anti-parallel to the confining magnetic field. The injected rf energy is phased to couple to said traveling waves with both a phase velocity component and a wave momentum component in the direction of the plasma traveling wave components. The injected rf energy has a predetermined spectrum selected so that said traveling waves couple to plasma electrons having velocities in a predetermined range delta . The velocities in the range are substantially greater than the thermal electron velocity of the plasma. In addition, the range is sufficiently broad to produce a raised plateau having width delta in the plasma electron velocity distribution so that the plateau electrons provide steady-state current to generate a poloidal magnetic field component sufficient for confining the plasma. In steady state operation of the fusion reactor, the fusion power density in the plasma exceeds the power dissipated in the plasma

Advantages of forced non-steady operated trickle-bed reactors

NARCIS (Netherlands)

Boelhouwer, J.G.; Piepers, H.W.; Drinkenburg, A.A.H.

2002-01-01

Trickle-bed reactors are usually operated in the steady state trickle flow regime. Uneven liquid distribution and the formation of hot spots are the most serious problems experienced during trickle flow operation. In this paper, we advocate the use of non-steady state operation of trickle-bed

COOLOD, Steady-State Thermal Hydraulics of Research Reactors

International Nuclear Information System (INIS)

Kaminaga, Masanori

1997-01-01

1 - Description of program or function: The COOLOD-N2 code provides a capability for the analyses of the steady-state thermal-hydraulics of research reactors. This code is a revised version of the COOLOD-N code, and is applicable not only for research reactors in which plate-type fuel is adopted, but also for research reactors in which rod-type fuel is adopted. In the code, subroutines to calculate temperature distribution in rod-type fuel have been newly added to the COOLOD-N code. The COOLOD-N2 code can calculate fuel temperatures under both forced convection cooling mode and natural convection cooling mode. A 'Heat Transfer package' is used for calculating heat transfer coefficient, DNB heat flux etc. The 'Heat Transfer package' is a subroutine program and is especially developed for research reactors in which plate-type fuel is adopted. In case of rod-type fuel, DNB heat flux is calculated by both the 'Heat Transfer package' and Lund DNB heat flux correlation which is popular for TRIGA reactor. The COOLOD-N2 code also has a capability of calculating ONB temperature, the heat flux at onset of flow instability as well as DNB heat flux. 2 - Method of solution: The 'Heat Transfer Package' is a subprogram for calculating heat transfer coefficients, ONB temperature, heat flux at onset of flow instability and DNB heat flux. The 'Heat transfer package' was especially developed for research reactors which are operated under low pressure and low temperature conditions using plate-type fuel, just like the JRR-3M. Heat transfer correlations adopted in the 'Heat Transfer Package' were obtained or estimated based on the heat transfer experiments in which thermal-hydraulic features of the upgraded JRR-3 core were properly reflected. The 'Heat Transfer Package' is applicable to upward and downward flow

Technology and physics in the Tokamak Program: The need for an integrated, steady-state RandD tokamak experiment

International Nuclear Information System (INIS)

1988-05-01

The Steady-state Tokamak (STE) Experiment is a proposed superconducting-coil, hydrogen-plasma tokamak device intended to address the integrated non-nuclear issues of steady state, high-power tokamak physics and technology. Such a facility has been called for in the US program plan for the mid 1990's, and will play a unique role in the world-wide fusion effort. Information from STE on steady-state current drive, plasma control, and high power technology will contribute significantly to the operating capabilities of future steady-state devices. This paper reviews preliminary designs and expected technological contributions to the US and world fusion reactor research from each of the above mentioned reactor systems. This document is intended as a proposal and feasibility discussion and does not include exhaustive technical reviews. 12 figs., 3 tabs

A highly efficient autothermal microchannel reactor for ammonia decomposition: Analysis of hydrogen production in transient and steady-state regimes

Science.gov (United States)

Engelbrecht, Nicolaas; Chiuta, Steven; Bessarabov, Dmitri G.

2018-05-01

The experimental evaluation of an autothermal microchannel reactor for H2 production from NH3 decomposition is described. The reactor design incorporates an autothermal approach, with added NH3 oxidation, for coupled heat supply to the endothermic decomposition reaction. An alternating catalytic plate arrangement is used to accomplish this thermal coupling in a cocurrent flow strategy. Detailed analysis of the transient operating regime associated with reactor start-up and steady-state results is presented. The effects of operating parameters on reactor performance are investigated, specifically, the NH3 decomposition flow rate, NH3 oxidation flow rate, and fuel-oxygen equivalence ratio. Overall, the reactor exhibits rapid response time during start-up; within 60 min, H2 production is approximately 95% of steady-state values. The recommended operating point for steady-state H2 production corresponds to an NH3 decomposition flow rate of 6 NL min-1, NH3 oxidation flow rate of 4 NL min-1, and fuel-oxygen equivalence ratio of 1.4. Under these flows, NH3 conversion of 99.8% and H2 equivalent fuel cell power output of 0.71 kWe is achieved. The reactor shows good heat utilization with a thermal efficiency of 75.9%. An efficient autothermal reactor design is therefore demonstrated, which may be upscaled to a multi-kW H2 production system for commercial implementation.

Advanced fuels for nuclear fusion reactors

International Nuclear Information System (INIS)

McNally, J.R. Jr.

1974-01-01

Should magnetic confinement of hot plasma prove satisfactory at high β (16 πnkT//sub B 2 / greater than 0.1), thermonuclear fusion fuels other than D.T may be contemplated for future fusion reactors. The prospect of the advanced fusion fuels D.D and 6 Li.D for fusion reactors is quite promising provided the system is large, well reflected and possesses a high β. The first generation reactions produce the very active, energy-rich fuels t and 3 He which exhibit a high burnup probability in very hot plasmas. Steady state burning of D.D can ensue in a 60 kG field, 5 m reactor for β approximately 0.2 and reflectivity R/sub mu/ = 0.9 provided the confinement time is about 38 sec. The feasibility of steady state burning of 6 Li.D has not yet been demonstrated but many important features of such systems still need to be incorporated in the reactivity code. In particular, there is a need for new and improved nuclear cross section data for over 80 reaction possibilities

An accelerator based steady state neutron source

International Nuclear Information System (INIS)

Burke, R.J.; Johnson, D.L.

1985-01-01

Using high current, c.w. linear accelerator technology, a spallation neutron source can achieve much higher average intensities than existing or proposed pulsed spallation sources. With about 100 mA of 300 MeV protons or deuterons, the Accelerator Based Neutron Research Facility (ABNR) would initially achieve the 10 16 n/cm 2 .s thermal flux goal of the advanced steady state neutron source, and upgrading could provide higher steady state fluxes. The relatively low ion energy compared to other spallation sources has an important impact on R and D requirements as well as capital cost, for which a range of $300-450M is estimated by comparison to other accelerator-based neutron source facilities. The source is similar to a reactor source in most respects. It has some higher energy neutrons but fewer gamma rays, and the moderator region is free of many of the design constraints of a reactor, which helps to implement sources for various neutron energy spectra, many beam tubes, etc. With the development of multi-beam concept and the basis for currents greater than 100 mA that is assumed in the R and D plan, the ABNR would serve many additional uses, such as fusion materials development, production of proton-rich isotopes, and other energy and defense program needs

Steady state and LOCA analysis of Kartini reactor using RELAP5/SCDAP code: The role of passive system

Science.gov (United States)

Antariksawan, Anhar R.; Wahyono, Puradwi I.; Taxwim

2018-02-01

Safety is the priority for nuclear installations, including research reactors. On the other hand, many studies have been done to validate the applicability of nuclear power plant based best estimate computer codes to the research reactor. This study aims to assess the applicability of the RELAP5/SCDAP code to Kartini research reactor. The model development, steady state and transient due to LOCA calculations have been conducted by using RELAP5/SCDAP. The calculation results are compared with available measurements data from Kartini research reactor. The results show that the RELAP5/SCDAP model steady state calculation agrees quite well with the available measurement data. While, in the case of LOCA transient simulations, the model could result in reasonable physical phenomena during the transient showing the characteristics and performances of the reactor against the LOCA transient. The role of siphon breaker hole and natural circulation in the reactor tank as passive system was important to keep reactor in safe condition. It concludes that the RELAP/SCDAP could be use as one of the tool to analyse the thermal-hydraulic safety of Kartini reactor. However, further assessment to improve the model is still needed.

3D steady-state MR cisternography in CSF rhinorrhoea

International Nuclear Information System (INIS)

Jayakumar, P.N.; Kovoor, J.M.E.; Srikanth, S.G.; Praharaj, S.S.

2001-01-01

Purpose: To determine the utility of 3D steady-state MR cisternography in the demonstration and localisation of cerebrospinal fluid (CSF) leak in patients with clinically suspected CSF rhinorrhoea. Material and Methods: Six consecutive patients with clinically suspected CSF rhinorrhoea were examined with routine MR evaluation and MR cisternography (MRC). All MR examinations included fast spin-echo (SE) T1WI in axial and sagittal planes, fast SE T2WI in axial and coronal planes and fluid attenuated inversion recovery (FLAIR) images in the axial plane. 3D evaluation was done using the CISS technique with 0.7-mm thickness in the sagittal and coronal planes. The site and extent of the defect, and any brain herniation detected on MRC were correlated with surgical findings. Results: In the 6 patients who underwent surgical exploration and repair, intraoperative findings correlated with the defect revealed by MRC in all cases. Conclusion: In clinically suspected CSF rhinorrhoea, MRC is highly accurate in localising the site and extent of CSF fistula and may be used as the first investigation due to its efficacy and non-invasive nature

Studies on energy gain of muon catalyzed hybrid D-D Reactor and it comparison to D-T system

International Nuclear Information System (INIS)

Eskandari, M.R.; Hoseine-Motlagh, S.N.; Faghihi, F.

1998-01-01

Regarding the advantages of hybrid fusion reactors, in most recent studies, the energy gain of muon catalyzed D-T hybrid reactors are studied. Knowing advantages of D-D fuel such as availability, not being radio-active, no tritium inventory requirement and transport problems, the muon catalyzed hybrid D-D reactor (μCHDDR) gain is calculated here for a given net reaction by solving its dynamical equations for various deuterium densities. It is shown theμCHDDR has advantages even for previously suggested similar D-T reactor

LANSCE steady state unperturbed thermal neutron fluxes at 100 μA

International Nuclear Information System (INIS)

Russell, G.J.

1989-01-01

The ''maximum'' unperturbed, steady state thermal neutron flux for LANSCE is calculated to be 2 /times/ 10 13 n/cm 2 -s for 100 μA of 800-MeV protons. This LANSCE neutron flux is a comparable entity to a steady state reactor thermal neutron flux. LANSCE perturbed steady state thermal neutron fluxes have also been calculated. Because LANSCE is a pulsed neutron source, much higher ''peak'' (in time) neutron fluxes can be generated than at a steady state reactor source. 5 refs., 5 figs

Steady-state leaching of tritiated water from silica gel

DEFF Research Database (Denmark)

Das, H.A.; Hou, Xiaolin

2009-01-01

Aqueous leaching of tritium from silica gel, loaded by absorption of water vapor, makes part of reactor de-commissioning. It is found to follow the formulation of steady-state diffusion.......Aqueous leaching of tritium from silica gel, loaded by absorption of water vapor, makes part of reactor de-commissioning. It is found to follow the formulation of steady-state diffusion....

BR2 reactor core steady state transient modeling

International Nuclear Information System (INIS)

Makarenko, A.; Petrova, T.

2000-01-01

A coupled neutronics/hydraulics/heat-conduction model of the BR2 reactor core is under development at SCK-CEN. The neutron transport phenomenon has been implemented as steady state and time dependent nodal diffusion. The non-linear heat conduction equation in-side fuel elements is solved with a time dependent finite element method. To allow coupling between functional modules and to simulate subcooled regimes, a simple single-phase hydraulics has been introduced, while the two-phase hydraulics is under development. Multiple tests, general benchmark cases as well as calculation/experiment comparisons demonstrated a good accuracy of both neutronic and thermal hydraulic models, numerical reliability and full code portability. A refinement methodology has been developed and tested for better neutronic representation in hexagonal geometry. Much effort is still needed to complete the development of an extended cross section library with kinetic data and two-phase flow representation. (author)

Joint DIII-D/EAST Experiments Toward Steady State AT Demonstration

Science.gov (United States)

Garofalo, A. M.; Meneghini, O.; Staebler, G. M.; van Zeeland, M. A.; Gong, X.; Ding, S.; Qian, J.; Ren, Q.; Xu, G.; Grierson, B. A.; Solomon, W. M.; Holcomb, C. T.

2015-11-01

Joint DIII-D/EAST experiments on fully noninductive operation at high poloidal beta have demonstrated several attractive features of this regime for a steady-state fusion reactor. Very large bootstrap fraction (>80 %) is desirable because it reduces the demands on external noninductive current drive. High bootstrap fraction with an H-mode edge results in a broad current profile and internal transport barriers (ITBs) at large minor radius, leading to high normalized energy confinement and high MHD stability limits. The ITB radius expands with higher normalized beta, further improving both stability and confinement. Electron density ITB and large Shafranov shift lead to low AE activity in the plasma core and low anomalous fast ion losses. Both the ITB and the current profile show remarkable robustness against perturbations, without external control. Supported by US DOE under DE-FC02-04ER54698, DE-AC02-09CH11466 & DE-AC52-07NA27344 & by NMCFSP under contracts 2015GB102000 and 2015GB110001.

Steady-State Characterization of Bacteriorhodopsin-D85N Photocycle

Science.gov (United States)

Timucin, Dogan A.; Downie, John D.; Norvig, Peter (Technical Monitor)

1999-01-01

An operational characterization of the photocycle of the genetic mutant D85N of bacteriorhodopsin, BR-D85N, is presented. Steady-state bleach spectra and pump-probe absorbance data are obtained with thick hydrated films containing BR-D85N embedded in a gelatin host. Simple two- and three-state models are used to analyze the photocycle dynamics and extract relevant information such as pure-state absorption spectra, photochemical-transition quantum efficiencies, and thermal lifetimes of dominant states appearing in the photocycle, the knowledge of which should aid in the analysis of optical recording and retrieval of data in films incorporating this photochromic material. The remarkable characteristics of this material and their implications from the viewpoint of optical data storage and processing are discussed.

Very high flux steady state reactor and accelerator based sources

International Nuclear Information System (INIS)

Ludewig, H.; Todosow, M.; Simos, N.; Shapiro, S.; Hastings, J.

2004-01-01

With the number of steady state neutron sources in the US declining (including the demise of the Bnl HFBR) the remaining intense sources are now in Europe (i.e. reactors - ILL and FMR, accelerator - PSI). The intensity of the undisturbed thermal flux for sources currently in operation ranges from 10 14 n/cm 2 *s to 10 15 n/cm 2 *s. The proposed Advanced Neutron Source (ANS) was to be a high power reactor (about 350 MW) with a projected undisturbed thermal flux of 7*10 15 n/cm 2 *s but never materialized. The objective of the current study is to explore the requirements and implications of two source concepts with an undisturbed flux of 10 16 n/cm 2 *s. The first is a reactor based concept operating at high power density (10 MW/l - 15 MW/l) and a total power of 100 MW - 250 MW, depending on fissile enrichment. The second is an accelerator based concept relying on a 1 GeV - 1.5 GeV proton Linac with a total beam power of 40 MW and a liquid lead-bismuth eutectic target. In the reactor source study, the effects of fissile material enrichment, coolant temperature and pressure drop, and estimates of pressure vessel stress levels will be investigated. The fuel form for the reactor will be different from all other operating source reactors in that it is proposed to use an infiltrated graphitic structure, which has been developed for nuclear thermal propulsion reactor applications. In the accelerator based source the generation of spallation products and their activation levels, and the material damage sustained by the beam window will be investigated. (authors)

S3C: EBT Steady-State Shooting code description and user's guide

International Nuclear Information System (INIS)

Downum, W.B.

1983-09-01

The Oak Ridge National Laboratory (ORNL) one-dimensional (1-D) Steady-State Shooting code (S3C) for ELMO Bumpy Torus (EBT) plasmas is described. Benchmark calculations finding the steady-state density and electron and ion temperature profiles for a known neutral density profile and known external energy sources are carried out. Good agreement is obtained with results from the ORNL Radially Resolved Time Dependent 1-D Transport code for an EBT-Q type reactor. The program logic is described, along with the physics models in each code block and the variable names used. Sample input and output files are listed, along with the main code

Recent results on steady state and confinement improvement research on JT-60U

International Nuclear Information System (INIS)

Ide, Shunsuke

2000-01-01

On the JT-60U tokamak, fusion plasma research for realization of a steady state tokamak reactor has been pursued. Towards that goal, confinement improved plasmas such as H-mode, high β p , reversed magnetic shear (RS) and latter two combined with H-mode edge pedestal have been developed and investigated intensively. A key issue to achieve non-inductive current drive relevant to a steady state fusion reactor is to increase the fraction of the bootstrap current and match the spatial profile to the optimum. In 1999, as the result of the optimization, the equivalent deuterium-tritium (D-T) fusion gain (Q DT eq ) of 0.5 was sustained for 0.8 s, which is roughly equal to the energy confinement time, in a RS plasma. In order to achieve a RS plasma in steady state two approach have been explored. One is to use external current driver such as lower hybrid current drive (LHCD), and by optimizing LHCD a quasi-steady RS discharge was obtained. The other approach is to utilize bootstrap current as much as possible, and with highly increased fraction of the bootstrap current, a confinement enhancement factor of 3.6 was maintained for 2.7 s in a RS plasma with H-mode edge. A heating and current drive system in the electron cyclotron range of frequency for localized heating and current drive has been installed on JT-60U, and in initial experiments a clear increase of the central electron temperature in a RS high density central region was confirmed only with injected power of 0.75 MW. (author)

D-3He fuel cycles for neutron lean reactors

International Nuclear Information System (INIS)

Kernbichler, W.; Miley, G.H.; Heindler, M.

1989-01-01

The intrinsic potential of D-3He as a reactor fuel is investigated for a large range of 3He to D density ratios. A steady-state zero-dimensional reactor model is developed in which much care is attributed to a proper treatment of fast fusion products. Useful ranges of reactor parameters as well as temperature-density windows for driven and ignited operation are identified. Various figures of merit are calculated, such as power densities, net power production, neutron production, tritium load and radiative power. These results suggest several optimistic conclusions about the performance of D-3He as a reactor fuel

Dynamical and technological consequences of multiple isolas of steady states in a catalytic fluidised-bed reactor

Directory of Open Access Journals (Sweden)

Bizon Katarzyna

2017-09-01

Full Text Available Steady-state characteristics of a catalytic fluidised bed reactor and its dynamical consequences are analyzed. The occurrence of an untypical steady-state structure manifesting in a form of multiple isolas is described. A two-phase bubbling bed model is used for a quantitative description of the bed of catalyst. The influence of heat exchange intensity and a fluidisation ratio onto the generation of isolated solution branches is presented for two kinetic schemes. Dynamical consequences of the coexistence of such untypical branches of steady states are presented. The impact of linear growth of the fluidisation ratio and step change of the cooling medium temperature onto the desired product yield is analyzed. The results presented in this study confirm that the identification of a region of the occurrence of multiple isolas is important due to their strong impact both on the process start-up and its control.

Steady State Advanced Tokamak (SSAT): The mission and the machine

International Nuclear Information System (INIS)

Thomassen, K.; Goldston, R.; Nevins, B.; Neilson, H.; Shannon, T.; Montgomery, B.

1992-03-01

Extending the tokamak concept to the steady state regime and pursuing advances in tokamak physics are important and complementary steps for the magnetic fusion energy program. The required transition away from inductive current drive will provide exciting opportunities for advances in tokamak physics, as well as important impetus to drive advances in fusion technology. Recognizing this, the Fusion Policy Advisory Committee and the US National Energy Strategy identified the development of steady state tokamak physics and technology, and improvements in the tokamak concept, as vital elements in the magnetic fusion energy development plan. Both called for the construction of a steady state tokamak facility to address these plan elements. Advances in physics that produce better confinement and higher pressure limits are required for a similar unit size reactor. Regimes with largely self-driven plasma current are required to permit a steady-state tokamak reactor with acceptable recirculating power. Reliable techniques of disruption control will be needed to achieve the availability goals of an economic reactor. Thus the central role of this new tokamak facility is to point the way to a more attractive demonstration reactor (DEMO) than the present data base would support. To meet the challenges, we propose a new ''Steady State Advanced Tokamak'' (SSAT) facility that would develop and demonstrate optimized steady state tokamak operating mode. While other tokamaks in the world program employ superconducting toroidal field coils, SSAT would be the first major tokamak to operate with a fully superconducting coil set in the elongated, divertor geometry planned for ITER and DEMO

Steady-State Thermal-Hydraulics Analyses for the Conversion of the BR2 Reactor to LEU

Energy Technology Data Exchange (ETDEWEB)

Licht, J. R. [Argonne National Lab. (ANL), Argonne, IL (United States); Bergeron, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Van den Branden, G. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Kalcheva, S [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Sikik, E [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Koonen, E [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium)

2016-09-01

BR2 is a research reactor used for radioisotope production and materials testing. It’s a tank-in-pool type reactor cooled by light water and moderated by beryllium and light water. The reactor core consists of a beryllium moderator forming a matrix of 79 hexagonal prisms in a hyperboloid configuration; each having a central bore that can contain a variety of different components such as a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Based on a series of tests, the BR2 operation is currently limited to a maximum allowable heat flux of 470 W/cm² to ensure fuel plate integrity during steady-state operation and after a loss-of-flow/loss-of-pressure accident. A feasibility study for the conversion of the BR2 reactor from highly-enriched uranium (HEU) to low-enriched uranium (LEU) fuel was previously performed to verify it can operate safely at the same maximum nominal steady-state heat flux. An assessment was also performed to quantify the heat fluxes at which the onset of flow instability and critical heat flux occur for each fuel type. This document updates and expands these results for the current representative core configuration (assuming a fresh beryllium matrix) by evaluating the onset of nucleate boiling (ONB), onset of fully developed nucleate boiling (FDNB), onset of flow instability (OFI) and critical heat flux (CHF).

Steady-state and transient simulations of gas cooled reactor with the computer code CATHARE

International Nuclear Information System (INIS)

Tauveron, N.; Saez, M.; Marchand, M.; Chataing, T.; Geffraye, G.; Cherel, J. M.

2003-01-01

This work concerns the design and safety analysis of Gas Cooled Reactors. The CATHARE code is used to test the design and safety of two different concepts, a High Temperature Gas Reactor concept (HTGR) and a Gas Fast Reactor concept (GFR). Relative to the HTGR concept, three transient simulations are performed and described in this paper: loss of electrical load without turbomachine trip, 10 inch cold duct break, 10 inch cold duct break combined with a tube rupture of a cooling exchanger. A second step consists in modelling a GFR concept. A nominal steady state situation at a power of 600 MW is obtained and first transient simulations are carried out to study decay heat removal situations after primary loop depressurisation

Numerical investigation of the 3-dimensional steady-state temperature- and flow distribution in the core of a pebble bed high temperature reactor

International Nuclear Information System (INIS)

Verfondern, K.

1983-01-01

This work presents a computer model determining the steady-state temperature- and flow field in 3 dimensions in the core of a pebble bed high temperature reactor. The numerical sprinkler method, basind on the Thermix-model, allows to describe the thermo-hydraulics of a non-rotational-symmetric core-geometry. The AVR-reactor in Juelich, in operation since 1967, represents a suitable investigation-object for the computer model of Thermix-3D. It is in a 3D-mesh-structure to reproduce very precisely the so called ''graphite noses'', in which the shut-down rods are conducted as well as the filling cones in the inner and outer area. The results of the final calculation of the normal operation condition for the AVR-reactor unambiguously show, that within the core reproduced in 3 dimensions there are evident deviations in the flow profile and in the temperatures of the cooling gas in contrast to a 2D-handling. (orig.) [de

Pellet injectors for steady state plasma fuelling

International Nuclear Information System (INIS)

Vinyar, I.; Geraud, A.; Yamada, H.; Lukin, A.; Sakamoto, R.; Skoblikov, S.; Umov, A.; Oda, Y.; Gros, G.; Krasilnikov, I.; Reznichenko, P.; Panchenko, V.

2005-01-01

Successful steady state operation of a fusion reactor should be supported by repetitive pellet injection of solidified hydrogen isotopes in order to produce high performance plasmas. This paper presents pneumatic pellet injectors and its implementation for long discharge on the LHD and TORE SUPRA, and a new centrifuge pellet injector test results. All injectors are fitted with screw extruders well suited for steady state operation

Tritium-management requirements for D-T fusion reactors (ETF, INTOR, FED)

International Nuclear Information System (INIS)

Finn, P.A.; Clemmer, R.G.; Misra, B.

1981-10-01

The successful operation of D-T fusion reactors will depend on the development of safe and reliable tritium-containment and fuel-recycle systems. The tritium handling requirements for D-T reactors were analyzed. The reactor facility was then designed from the viewpoint of tritium management. Recovery scenarios after a tritium release were generated to show the relative importance of various scenarios. A fusion-reactor tritium facility was designed which would be appropriate for all types of plants from the Engineering Test Facility (ETF), the International Tokamak Reactor (INTOR), and the Fusion Engineering Device (FED) to the full-scale power plant epitomized by the STARFIRE design

Steady-state thermal-hydraulic analysis of the Moroccan TRIGA MARK II reactor by using PARET/ANL and COOLOD-N2 codes

International Nuclear Information System (INIS)

Boulaich, Y.; Nacir, B.; El Bardouni, T.; Zoubair, M.; El Bakkari, B.; Merroun, O.; El Younoussi, C.; Htet, A.; Boukhal, H.; Chakir, E.

2011-01-01

Research highlights: → The COOLOD/N2 and PARET/ANL codes were used for a steady-state thermal-hydraulic and safety analysis of the 2 MW TRIGA MARK II reactor located at the Nuclear Studies Center of Maamora (CENM), Morocco. → The main objective of this study is to ensure the safety margins of different safety related parameters by steady-state calculations at full power level (2 MW). → The most important conclusion is that all obtained values of DNBR, fuel center and surface temperature, cladding surface temperature and coolant temperature across the hottest channel are largely far to compromise safety of the reactor. - Abstract: The COOLOD/N2 and PARET/ANL codes were used for a steady-state thermal-hydraulic and safety analysis of the 2 MW TRIGA MARK II reactor located at the Nuclear Studies Center of Maamora (CENM), Morocco. In order to validate our PARET/ANL and COOLOD-N2 models, the fuel center temperature as function of core power was calculated and compared with the corresponding experimental values. The comparison indicates that the calculated values are in satisfactory agreement with the measurement. The main objective of this study is to ensure the safety margins of different safety related parameters by steady-state calculations at full power level (2 MW). Therefore, we have calculated the departure from nucleate boiling ratio (DNBR), fuel center and surface temperature, cladding surface temperature and coolant temperature profiles across the hottest channel. The most important conclusion is that all obtained values are largely far to compromise safety of the reactor.

Steady-state Operational Characteristics of Ghana Research ...

African Journals Online (AJOL)

Steady state operational characteristics of the 30 kW tank-in-pool type reactor named Ghana Research Reactor-1 were investigated after a successful on-site zero power critical experiments. The steadystate operational character-istics determined were the thermal neutron fluxes, maximum period of operation at nominal ...

Effects of neutron source ratio on nuclear characteristics of D-D fusion reactor blankets and shields

International Nuclear Information System (INIS)

Nakashima, Hideki; Nakao, Yasuyuki; Ohta, Masao

1978-01-01

An examination is made of the dependence shown by the nuclear characteristics of the blanket and shield of D-D fusion reactors on S sub( d d)/S sub( d t), the ratio between the 2.45 MeV neutrons resulting from the D-D reaction and those of 14.06 MeV from the D-T reaction. Also, an estimate is presented of this neutron source ratio S sub( d d)/S sub( d t) for the case of D-D reactors, taken as an example. It is shown that an increase of S sub( d d)/S sub( d t) reduces the amount of nuclear heating per unit source neutron, while at the same time improving the shielding characteristics. This is accountable to lowering of the energy and penetrability of incident neutrons into the blanket brought about by the increase of S sub( d d)/S sub( d t). The value of S sub( d d)/S sub( d t) in a steady state D-D fusioning plasma core is estimated to be 1.46 -- 1.72 for an ion temperature ranging from 60 -- 180 keV. The reductions obtained on H sub( t)sup( b) (total heating in the blanket), H sub( t)sup( m g)/H sub( t)sup( b) (shielding indicator = ratio between total heating in superconducting magnet and that in the blanket) and phi sup( m g)/phi sup( w) (ratio of fast neutron fluxes between that at the magnet inner surface and that at the first wall inner surface) brought about by increasing S sub( d d)/S sub( d t) from unity to the value cited above do not differ to any appreciable extent, whichever is adopted among the design models considered here, the differences being at most about 10, 15 and 25%, respectively, for these three parameters. These results would broaden the validity of the conclusion derived in the previous paper for the case of S sub( d d)/S sub( d t) = 1.0, that the blanket-shield concept would appear to be the most suitable for D-D fusion reactors. (author)

Wildcat: A commercial deuterium-deuterium tokamak reactor

International Nuclear Information System (INIS)

Evans, K.; Baker, C.C.; Barry, K.M.

1983-01-01

WILDCAT is a conceptual design of a catalyzed deuterium-deuterium tokamak commercial fusion reactor. WILDCAT utilizes the beneficial features of no tritium breeding, while not extrapolating unnecessarily from existing deuterium-tritium (D-T) designs. The reactor is larger and has higher magnetic fields and plasma pressures than typical D-T devices. It is more costly, but eliminates problems associated with tritium breeding and has tritium inventories and throughputs approximately two orders of magnitude less than typical D-T reactors. There are both a steady-state version with Alfven-wave current drive and a pulsed version. Extensive comparison with D-T devices has been made, and cost and safety analyses have been included. All of the major reactor systems have been worked out to a level of detail appropriate to a complete conceptual design

Ultrahigh-resolution imaging of the human brain with phase-cycled balanced steady-state free precession at 7 T.

Science.gov (United States)

Zeineh, Michael M; Parekh, Mansi B; Zaharchuk, Greg; Su, Jason H; Rosenberg, Jarrett; Fischbein, Nancy J; Rutt, Brian K

2014-05-01

The objectives of this study were to acquire ultra-high resolution images of the brain using balanced steady-state free precession (bSSFP) at 7 T and to identify the potential utility of this sequence. Eight volunteers participated in this study after providing informed consent. Each volunteer was scanned with 8 phase cycles of bSSFP at 0.4-mm isotropic resolution using 0.5 number of excitations and 2-dimensional parallel acceleration of 1.75 × 1.75. Each phase cycle required 5 minutes of scanning, with pauses between the phase cycles allowing short periods of rest. The individual phase cycles were aligned and then averaged. The same volunteers underwent scanning using 3-dimensional (3D) multiecho gradient recalled echo at 0.8-mm isotropic resolution, 3D Cube T2 at 0.7-mm isotropic resolution, and thin-section coronal oblique T2-weighted fast spin echo at 0.22 × 0.22 × 2.0-mm resolution for comparison. Two neuroradiologists assessed image quality and potential research and clinical utility. The volunteers generally tolerated the scan sessions well, and composite high-resolution bSSFP images were produced for each volunteer. Rater analysis demonstrated that bSSFP had a superior 3D visualization of the microarchitecture of the hippocampus, very good contrast to delineate the borders of the subthalamic nucleus, and relatively good B1 homogeneity throughout. In addition to an excellent visualization of the cerebellum, subtle details of the brain and skull base anatomy were also easier to identify on the bSSFP images, including the line of Gennari, membrane of Liliequist, and cranial nerves. Balanced steady-state free precession had a strong iron contrast similar to or better than the comparison sequences. However, cortical gray-white contrast was significantly better with Cube T2 and T2-weighted fast spin echo. Balanced steady-state free precession can facilitate ultrahigh-resolution imaging of the brain. Although total imaging times are long, the individually short

Influence of longitudinal position on the evolution of steady-state signal in cardiac cine balanced steady-state free precession imaging.

Science.gov (United States)

Spear, Tyler J; Stromp, Tori A; Leung, Steve W; Vandsburger, Moriel H

2017-11-01

Emerging quantitative cardiac magnetic resonance imaging (CMRI) techniques use cine balanced steady-state free precession (bSSFP) to measure myocardial signal intensity and probe underlying physiological parameters. This correlation assumes that steady-state is maintained uniformly throughout the heart in space and time. To determine the effects of longitudinal cardiac motion and initial slice position on signal deviation in cine bSSFP imaging by comparing two-dimensional (2D) and three-dimensional (3D) acquisitions. Nine healthy volunteers completed cardiac MRI on a 1.5-T scanner. Short axis images were taken at six slice locations using both 2D and 3D cine bSSFP. 3D acquisitions spanned two slices above and below selected slice locations. Changes in myocardial signal intensity were measured across the cardiac cycle and compared to longitudinal shortening. For 2D cine bSSFP, 46% ± 9% of all frames and 84% ± 13% of end-diastolic frames remained within 10% of initial signal intensity. For 3D cine bSSFP the proportions increased to 87% ± 8% and 97% ± 5%. There was no correlation between longitudinal shortening and peak changes in myocardial signal. The initial slice position significantly impacted peak changes in signal intensity for 2D sequences ( P  cine bSSFP that is only restored at the center of a 3D excitation volume. During diastole, a transient steady-state is established similar to that achieved with 3D cine bSSFP regardless of slice location.

Long-term ammonia removal in a coconut fiber-packed biofilter: analysis of N fractionation and reactor performance under steady-state and transient conditions.

Science.gov (United States)

Baquerizo, Guillermo; Maestre, Juan P; Machado, Vinicius C; Gamisans, Xavier; Gabriel, David

2009-05-01

A comprehensive study of long-term ammonia removal in a biofilter packed with coconut fiber is presented under both steady-state and transient conditions. Low and high ammonia loads were applied to the reactor by varying the inlet ammonia concentration from 90 to 260 ppm(v) and gas contact times ranging from 20 to 36 s. Gas samples and leachate measurements were periodically analyzed and used for characterizing biofilter performance in terms of removal efficiency (RE) and elimination capacity (EC). Also, N fractions in the leachate were quantified to both identify the experimental rates of nitritation and nitratation and to determine the N leachate distribution. Results showed stratification in the biofilter activity and, thus, most of the NH(3) removal was performed in the lower part of the reactor. An average EC of 0.5 kg N-NH(3)m(-3)d(-1) was obtained for the whole reactor with a maximum local average EC of 1.7 kg N-NH(3)m(-3)d(-1). Leachate analyses showed that a ratio of 1:1 of ammonium and nitrate ions in the leachate was obtained throughout steady-state operation at low ammonia loads with similar values for nitritation and nitratation rates. Low nitratation rates during high ammonia load periods occurred because large amounts of ammonium and nitrite accumulated in the packed bed, thus causing inhibition episodes on nitrite-oxidizing bacteria due to free ammonia accumulation. Mass balances showed that 50% of the ammonia fed to the reactor was oxidized to either nitrite or nitrate and the rest was recovered as ammonium indicating that sorption processes play a fundamental role in the treatment of ammonia by biofiltration.

A method for statistical steady state thermal analysis of reactor cores

International Nuclear Information System (INIS)

Whetton, P.A.

1980-01-01

This paper presents a method for performing a statistical steady state thermal analysis of a reactor core. The technique is only outlined here since detailed thermal equations are dependent on the core geometry. The method has been applied to a pressurised water reactor core and the results are presented for illustration purposes. Random hypothetical cores are generated using the Monte-Carlo method. The technique shows that by splitting the parameters into two types, denoted core-wise and in-core, the Monte Carlo method may be used inexpensively. The idea of using extremal statistics to characterise the low probability events (i.e. the tails of a distribution) is introduced together with a method of forming the final probability distribution. After establishing an acceptable probability of exceeding a thermal design criterion, the final probability distribution may be used to determine the corresponding thermal response value. If statistical and deterministic (i.e. conservative) thermal response values are compared, information on the degree of pessimism in the deterministic method of analysis may be inferred and the restrictive performance limitations imposed by this method relieved. (orig.)

Extension of the reactor dynamics code MGT-3D for pebblebed and blocktype high-temperature-reactors

International Nuclear Information System (INIS)

Shi, Dunfu

2015-01-01

The High Temperature Gas cooled Reactor (HTGR) is an improved, gas cooled nuclear reactor. It was chosen as one of the candidates of generation IV nuclear plants [1]. The reactor can be shut down automatically because of the negative reactivity feedback due to the temperature's increasing in designed accidents. It is graphite moderated and Helium cooled. The residual heat can be transferred out of the reactor core by inactive ways as conduction, convection, and thermal radiation during the accident. In such a way, a fuel temperature does not go beyond a limit at which major fission product release begins. In this thesis, the coupled neutronics and fluid mechanics code MGT-3D used for the steady state and time-dependent simulation of HTGRs, is enhanced and validated [2]. The fluid mechanics part is validated by SANA experiments in steady state cases as well as transient cases. The fuel temperature calculation is optimized by solving the heat conduction equation of the coated particles. It is applied in the steady state and transient simulation of PBMR, and the results are compared to the simulation with the old overheating model. New approaches to calculate the temperature profile of the fuel element of block-type HTGRs, and the calculation of the homogeneous conductivity of composite materials are introduced. With these new developments, MGT-3D is able to simulate block-type HTGRs as well. This extended MGT-3D is used to simulate a cuboid ceramic block heating experiment in the NACOK-II facility. The extended MGT-3D is also applied to LOFC and DLOFC simulation of GT-MHR. It is a fluid mechanics calculation with a given heat source. This calculation result of MGT-3D is verified with the calculation results of other codes. The design of the Japanese HTTR is introduced. The deterministic simulation of the LOFC experiment of HTTR is conducted with the Monte-Carlo code Serpent and MGT-3D, which is the LOFC Project organized by OECD/NEA [3]. With Serpent the burnup

Theoretical investigation on the steady-state natural circulation characteristics of a new type of pressurized water reactor

International Nuclear Information System (INIS)

Gou Junli; Qiu Suizheng; Su Guanghui; Jia Dounan

2006-01-01

This article presents a theoretical investigation on the steady-state natural circulation characteristics of a new type of pressurized water reactor. Through numerically solving the one-dimensional steady-state single-phase conservative equations for the primary circuit and the steady-state two-phase drift-flux conservative equations for the secondary side of the steam generator, the natural circulation characteristics were studied. On the basis of the preliminary calculation analysis, it was found that natural circulation mass flow rate was proportional to the exponential function of the power and that the value of the exponent is related to the operating conditions of the secondary side of the steam generator. The higher the outlet pressure of the secondary side of the steam generator, the higher the primary natural circulation mass flow rate. The larger height difference between the core center and the steam generator center is favorable for the heat removal capacity of the natural circulation. (authors)

Fast-ion transport in qmin>2, high-β steady-state scenarios on DIII-D

International Nuclear Information System (INIS)

Holcomb, C. T.; Heidbrink, W. W.; Collins, C.; Ferron, J. R.; Van Zeeland, M. A.; Garofalo, A. M.; Bass, E. M.; Luce, T. C.; Pace, D. C.; Solomon, W. M.; Mueller, D.; Grierson, B.; Podesta, M.; Gong, X.; Ren, Q.; Park, J. M.; Kim, K.; Turco, F.

2015-01-01

Results from experiments on DIII-D [J. L. Luxon, Fusion Sci. Technol. 48, 828 (2005)] aimed at developing high β steady-state operating scenarios with high-q min confirm that fast-ion transport is a critical issue for advanced tokamak development using neutral beam injection current drive. In DIII-D, greater than 11 MW of neutral beam heating power is applied with the intent of maximizing β N and the noninductive current drive. However, in scenarios with q min >2 that target the typical range of q 95 = 5–7 used in next-step steady-state reactor models, Alfvén eigenmodes cause greater fast-ion transport than classical models predict. This enhanced transport reduces the absorbed neutral beam heating power and current drive and limits the achievable β N . In contrast, similar plasmas except with q min just above 1 have approximately classical fast-ion transport. Experiments that take q min >3 plasmas to higher β P with q 95 = 11–12 for testing long pulse operation exhibit regimes of better than expected thermal confinement. Compared to the standard high-q min scenario, the high β P cases have shorter slowing-down time and lower ∇β fast , and this reduces the drive for Alfvénic modes, yielding nearly classical fast-ion transport, high values of normalized confinement, β N , and noninductive current fraction. These results suggest DIII-D might obtain better performance in lower-q 95 , high-q min plasmas using broader neutral beam heating profiles and increased direct electron heating power to lower the drive for Alfvén eigenmodes

Solution of generalized control system equations at steady state

International Nuclear Information System (INIS)

Vilim, R.B.

1987-01-01

Although a number of reactor systems codes feature generalized control system models, none of the models offer a steady-state solution finder. Indeed, if a transient is to begin from steady-state conditions, the user must provide estimates for the control system initial conditions and run a null transient until the plant converges to steady state. Several such transients may have to be run before values for control system demand signals are found that produce the desired plant steady state. The intent of this paper is (a) to present the control system equations assumed in the SASSYS reactor systems code and to identify the appropriate set of initial conditions, (b) to describe the generalized block diagram approach used to represent these equations, and (c) to describe a solution method and algorithm for computing these initial conditions from the block diagram. The algorithm has been installed in the SASSYS code for use with the code's generalized control system model. The solution finder greatly enhances the effectiveness of the code and the efficiency of the user in running it

Small-angle scattering at a pulsed neutron source: comparison with a steady-state reactor

Energy Technology Data Exchange (ETDEWEB)

Borso, C S; Carpenter, J M; Williamson, F S; Holmblad, G L; Mueller, M H; Faber, J Jr; Epperson, J E; Danyluk, S S [Argonne National Lab., IL (USA)

1982-08-01

A time-of-flight small-angle diffractometer employing seven tapered collimator elements and a two-dimensional gas proportional counter was successfully utilized to collect small-angle scattering data from a solution sample of the lipid salt cetylpyridinium chloride, C/sub 21/H/sub 38/N/sup +/.Cl/sup -/, at the Argonne National Laboratory prototype pulsed spallation neutron source, ZING-P'. Comparison of the small-angle scattering observed from the same compound at the University of Missouri Research Reactor corroborated the ZING-P' results. The results are used to compare the neutron flux available from the ZING-P' source relative to the well characterized University of Missouri source. Calculations based on experimentally determined parameters indicated the time-averaged rate of detected neutrons at the ZING-P' pulsed spallation source to have been at least 33% higher than the steady-state count rate from the same sample. Differences between time-of-flight techniques and conventional steady-state techniques are discussed.

Measurement and Modelling of Tearing Mode Stability for Steady-State Plasmas in DIII-D

Energy Technology Data Exchange (ETDEWEB)

Turco, F; Luce, T; Ferron, J; Petty, C; Politzer, P; Turnbull, A; Brennan, D; Murakami, M; LoDestro, L; Pearlstein, L; Casper, T; Jayakumar, R; Holcomb, C

2009-06-23

High-beta, quasi-steady state scenarios represent a fundamental step towards the performance required for future fusion reactors. In DIII-D steady-state scenario discharges, the normalized beta {beta}{sub N} {triple_bond} {beta}(%) {center_dot} a(m) {center_dot} B{sub T}(T)/I{sub p}(MA) (where {beta} is the ratio of the plasma pressure to the magnetic field pressure, {alpha} the plasma minor radius, B{sub T} the toroidal magnetic field and I{sub p} the plasma current) exceeds the no-wall ideal kink beta limit. The performance of this scenario is limited by the onset of an n = 1 tearing mode, which appears on the resistive evolution time-scale (1-2 s) at constant pressure and causes both a loss of confinement and a radial redistribution of the current density from which the available current drive sources cannot recover. It is routinely observed that the injection of electron cyclotron current drive (ECCD), with a broad deposition localized around {rho} {approx} 0.35, can prevent the mode from appearing. It must be noted that this is not a case of a direct stabilization due to the interaction with the mode's rational surface. These variations of the scenario are illustrated in Fig. 1, where the total injected power [neutral beam injection (NBI) and ECCD], {beta}{sub N} and the n = 1 magnetic perturbation at the outer wall are shown. In case (a), the onset of the n = 1 mode is observed when the EC power is not present or if it is stopped before the end of the high {beta} phase, whereas in case (b) the difference is pointed out between broad and narrow current deposition (with the narrow deposition case becoming unstable). The current density profile evolution and the MHD modes of several sets of significant discharges with and without ECCD (at different locations) have been analyzed, using motional Stark effect (MSE) spectroscopy measurements for the former and edge magnetic probes measurements, toroidal rotation profiles and fast electron cyclotron emission

Measurement and Modelling of Tearing Mode Stability for Steady-State Plasmas in DIII-D

International Nuclear Information System (INIS)

Turco, F.; Luce, T.; Ferron, J.; Petty, C.; Politzer, P.; Turnbull, A.; Brennan, D.; Murakami, M.; LoDestro, L.; Pearlstein, L.; Casper, T.; Jayakumar, R.; Holcomb, C.

2009-01-01

High-beta, quasi-steady state scenarios represent a fundamental step towards the performance required for future fusion reactors. In DIII-D steady-state scenario discharges, the normalized beta β N (triple b ond) β(%) · a(m) · B T (T)/I p (MA) (where β is the ratio of the plasma pressure to the magnetic field pressure, α the plasma minor radius, B T the toroidal magnetic field and I p the plasma current) exceeds the no-wall ideal kink beta limit. The performance of this scenario is limited by the onset of an n = 1 tearing mode, which appears on the resistive evolution time-scale (1-2 s) at constant pressure and causes both a loss of confinement and a radial redistribution of the current density from which the available current drive sources cannot recover. It is routinely observed that the injection of electron cyclotron current drive (ECCD), with a broad deposition localized around ρ ∼ 0.35, can prevent the mode from appearing. It must be noted that this is not a case of a direct stabilization due to the interaction with the mode's rational surface. These variations of the scenario are illustrated in Fig. 1, where the total injected power (neutral beam injection (NBI) and ECCD), β N and the n = 1 magnetic perturbation at the outer wall are shown. In case (a), the onset of the n = 1 mode is observed when the EC power is not present or if it is stopped before the end of the high β phase, whereas in case (b) the difference is pointed out between broad and narrow current deposition (with the narrow deposition case becoming unstable). The current density profile evolution and the MHD modes of several sets of significant discharges with and without ECCD (at different locations) have been analyzed, using motional Stark effect (MSE) spectroscopy measurements for the former and edge magnetic probes measurements, toroidal rotation profiles and fast electron cyclotron emission (ECE) data for the latter. One equilibrium based on EFIT reconstruction (1) with kinetic

Steady-State Thermal-Hydraulics Analyses for the Conversion of the BR2 Reactor to LEU

Energy Technology Data Exchange (ETDEWEB)

Licht, J. R. [Argonne National Lab. (ANL), Argonne, IL (United States); Bergeron, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Van den Branden, G. [SCK CEN (Belgium); Kalcheva, S. [SCK CEN (Belgium); Sikik, E. [SCK CEN (Belgium); Koonen, E. [SCK CEN (Belgium)

2015-12-01

BR2 is a research reactor used for radioisotope production and materials testing. It’s a tank-in-pool type reactor cooled by light water and moderated by beryllium and light water (Figure 1). The reactor core consists of a beryllium moderator forming a matrix of 79 hexagonal prisms in a hyperboloid configuration; each having a central bore that can contain a variety of different components such as a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Based on a series of tests, the BR2 operation is currently limited to a maximum allowable heat flux of 470 W/cm2 to ensure fuel plate integrity during steady-state operation and after a loss-of-flow/loss-of-pressure accident.

A Review of Fusion and Tokamak Research Towards Steady-State Operation: A JAEA Contribution

Directory of Open Access Journals (Sweden)

Mitsuru Kikuchi

2010-11-01

Full Text Available Providing a historical overview of 50 years of fusion research, a review of the fundamentals and concepts of fusion and research efforts towards the implementation of a steady state tokamak reactor is presented. In 1990, a steady-state tokamak reactor (SSTR best utilizing the bootstrap current was developed. Since then, significant efforts have been made in major tokamaks, including JT-60U, exploring advanced regimes relevant to the steady state operation of tokamaks. In this paper, the fundamentals of fusion and plasma confinement, and the concepts and research on current drive and MHD stability of advanced tokamaks towards realization of a steady-state tokamak reactor are reviewed, with an emphasis on the contributions of the JAEA. Finally, a view of fusion energy utilization in the 21st century is introduced.

Steady State versus Pulsed Tokamak DEMO

Energy Technology Data Exchange (ETDEWEB)

Orsitto, F.P., E-mail: francesco.orsitto@enea.it [Associazione EURATOM-ENEA Unita Tecnica Fusione, Frascati (Italy); Todd, T. [CCFE/Fusion Association, Culham Science Centre, Abingdon (United Kingdom)

2012-09-15

Full text: The present report deals with a Review of problems for a Steady state(SS) DEMO, related argument is treated about the models and the present status of comparison between the characteristics of DEMO pulsed versus a Steady state device.The studied SS DEMO Models (SLIM CS, PPCS model C EU-DEMO, ARIES-RS) are analyzed from the point of view of the similarity scaling laws and critical issues for a steady state DEMO. A comparison between steady state and pulsed DEMO is therefore carried out: in this context a new set of parameters for a pulsed (6 - 8 hours pulse) DEMO is determined working below the density limit, peak temperature of 20 keV, and requiring a modest improvement in the confinement factor(H{sub IPBy2} = 1.1) with respect to the H-mode. Both parameters density and confinement parameter are lower than the DEMO models presently considered. The concept of partially non-inductive pulsed DEMO is introduced since a pulsed DEMO needs heating and current drive tools for plasma stability and burn control. The change of the main parameter design for a DEMO working at high plasma peak temperatures T{sub e} {approx} 35 keV is analyzed: in this range the reactivity increases linearly with temperature, and a device with smaller major radius (R = 7.5 m) is compatible with high temperature. Increasing temperature is beneficial for current drive efficiency and heat load on divertor, being the synchrotron radiation one of the relevant components of the plasma emission at high temperatures and current drive efficiency increases with temperature. Technology and engineering problems are examined including efficiency and availability R&D issues for a high temperature DEMO. Fatigue and creep-fatigue effects of pulsed operations on pulsed DEMO components are considered in outline to define the R&D needed for DEMO development. (author)

Conceptual design of fusion experimental reactor (FER)

International Nuclear Information System (INIS)

1984-01-01

Conceptual Design of Fusion Experimental Reactor (FER) of which the objective will be to realize self-ignition with D-T reaction is reported. Mechanical Configurations of FER are characterized with a noncircular plasma and a double-null divertor. The primary aim of design studies is to demonstrate fissibility of reactor structures as compact and simple as possible with removable torus sectors. The structures of each component such as a first-wall, blanket, shielding, divertor, magnet and so on have been designed. It is also discussed about essential reactor plant system requirements. In addition to the above, a brief concept of a steady-state reactor based on RF current drive is also discussed. The main aim, in this time, is to examine physical studies of a possible RF steady-state reactor. (author)

Computational fluid dynamic model for thermohydraulic calculation for the steady-state of the real scale HTR-1

Energy Technology Data Exchange (ETDEWEB)

Gamez, Abel; Rojas, Leorlen; Rosales, Jesus; Castro, Landy Y.; Gonzalez, Daniel; Garcia, Carlos, E-mail: agamezgmf@gmail.com, E-mail: leored1984@gmail.com, E-mail: jrosales@instec.cu, E-mail: lcastro@instec.cu, E-mail: danielgonro@gmail.com, E-mail: cgr@instec.cu [Instituto Superior de Tecnologias y Ciencias Aplicadas (InSTEC), La Habana (Cuba); Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil); Oliveira, Carlos B. de, E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil); Dominguez, Dany S., E-mail: dsdominguez@gmail.com [Universidade Estadual de Santa Cruz (UESC), Ilheus, BA (Brazil)

2015-07-01

The high temperature gas cooled reactor (HTGR) is one of candidates of next generation of nuclear reactor according to IAEA report 2013. Evaluation of thermohydraulic performance and an experimental comparison results were proposed to the international research community. In this article, the tree dimensional CFD thermohydraulic modelation of steady state of HTR-10 modular reactor, using ANSYS CFX v14.0, has been done. Code-to-code and Code-to-experiment benchmark analyses, related to the testing program of the HTR-10 plant including steady state temperature distribution with the reactor at full power, were developed. The 3D real scale representation of reflector zone and fluid path flow inner and outer reflector blocks and cold helium cavity were carried out. The porous medium model was used to simulate the core zone in the reactor. The power distribution of the initial core published by IAEA-TECDOC-1694 obtained by Chief Scientific Investigators (CSIs) from China was used as heat sources in the core zone. (author)

Multimode optical fibers: steady state mode exciter.

Science.gov (United States)

Ikeda, M; Sugimura, A; Ikegami, T

1976-09-01

The steady state mode power distribution of the multimode graded index fiber was measured. A simple and effective steady state mode exciter was fabricated by an etching technique. Its insertion loss was 0.5 dB for an injection laser. Deviation in transmission characteristics of multimode graded index fibers can be avoided by using the steady state mode exciter.

D-D tokamak reactor assessment

International Nuclear Information System (INIS)

Baxter, D.C.; Dabiri, A.E.

1983-01-01

A quantitative comparison of the physics and technology requirements, and the cost and safety performance of a d-d tokamak relative to a d-t tokamak has been performed. The first wall/blanket and energy recovery cycle for the d-d tokamak is simpler, and has a higher efficiency than the d-t tokamak. In most other technology areas (such as magnets, RF, vacuum, etc.) d-d requirements are more severe and the systems are more complex, expensive and may involve higher technical risk than d-t tokamak systems. Tritium technology for processing the plasma exhaust, and tritium refueling technology are required for d-d reactors, but no tritium containment around the blanket or heat transport system is needed. Cost studies show that for high plasma beta and high magnetic field the cost of electricity from d-d and d-t tokamaks is comparable. Safety analysis shows less radioactivity in a d-d reactor but larger amounts of stored energy and thus higher potential for energy release. Consequences of all postulated d-d accidents are significantly smaller than those from d-t reactor tritium releases

Experimental Study and steady state stability analysis of CLL-T Series Parallel Resonant Converter with Fuzzy controller using State Space Analysis

Directory of Open Access Journals (Sweden)

C. Nagarajan

2012-09-01

Full Text Available This paper presents a Closed Loop CLL-T (capacitor inductor inductor Series Parallel Resonant Converter (SPRC has been simulated and the performance is analysised. A three element CLL-T SPRC working under load independent operation (voltage type and current type load is presented in this paper. The Steady state Stability Analysis of CLL-T SPRC has been developed using State Space technique and the regulation of output voltage is done by using Fuzzy controller. The simulation study indicates the superiority of fuzzy control over the conventional control methods. The proposed approach is expected to provide better voltage regulation for dynamic load conditions. A prototype 300 W, 100 kHz converter is designed and built to experimentally demonstrate, dynamic and steady state performance for the CLL-T SPRC are compared from the simulation studies.

Tokamak burn cycle study: a data base for comparing long pulse and steady-state power reactors

International Nuclear Information System (INIS)

Ehst, D.A.; Brooks, J.N.; Cha, Y.; Evans, K. Jr.; Hassanein, A.; Kim, S.; Majumdar, S.; Misra, B.; Stevens, H.C.

1983-11-01

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

Progress Toward Steady State Tokamak Operation Exploiting the high bootstrap current fraction regime

Science.gov (United States)

Ren, Q.

2015-11-01

Recent DIII-D experiments have advanced the normalized fusion performance of the high bootstrap current fraction tokamak regime toward reactor-relevant steady state operation. The experiments, conducted by a joint team of researchers from the DIII-D and EAST tokamaks, developed a fully noninductive scenario that could be extended on EAST to a demonstration of long pulse steady-state tokamak operation. Fully noninductive plasmas with extremely high values of the poloidal beta, βp >= 4 , have been sustained at βT >= 2 % for long durations with excellent energy confinement quality (H98y,2 >= 1 . 5) and internal transport barriers (ITBs) generated at large minor radius (>= 0 . 6) in all channels (Te, Ti, ne, VTf). Large bootstrap fraction (fBS ~ 80 %) has been obtained with high βp. ITBs have been shown to be compatible with steady state operation. Because of the unusually large ITB radius, normalized pressure is not limited to low βN values by internal ITB-driven modes. βN up to ~4.3 has been obtained by optimizing the plasma-wall distance. The scenario is robust against several variations, including replacing some on-axis with off-axis neutral beam injection (NBI), adding electron cyclotron (EC) heating, and reducing the NBI torque by a factor of 2. This latter observation is particularly promising for extension of the scenario to EAST, where maximum power is obtained with balanced NBI injection, and to a reactor, expected to have low rotation. However, modeling of this regime has provided new challenges to state-of-the-art modeling capabilities: quasilinear models can dramatically underpredict the electron transport, and the Sauter bootstrap current can be insufficient. The analysis shows first-principle NEO is in good agreement with experiments for the bootstrap current calculation and ETG modes with a larger saturated amplitude or EM modes may provide the missing electron transport. Work supported in part by the US DOE under DE-FC02-04ER54698, DE-AC52-07NA

Steady-state thermal-hydraulic design analysis of the Advanced Neutron Source reactor

International Nuclear Information System (INIS)

Yoder, G.L. Jr.; Dixon, J.R.; Elkassabgi, Y.; Felde, D.K.; Giles, G.E.; Harrington, R.M.; Morris, D.G.; Nelson, W.R.; Ruggles, A.E.; Siman-Tov, M.; Stovall, T.K.

1994-05-01

The Advanced Neutron Source (ANS) is a research reactor that is planned for construction at Oak Ridge National Laboratory. This reactor will be a user facility with the major objective of providing the highest continuous neutron beam intensities of any reactor in the world. Additional objectives for the facility include providing materials irradiation facilities and isotope production facilities as good as, or better than, those in the High Flux Isotope Reactor. To achieve these objectives, the reactor design uses highly subcooled heavy water as both coolant and moderator. Two separate core halves of 67.6-L total volume operate at an average power density of 4.5 MW(t)/L, and the coolant flows upward through the core at 25 m/s. Operating pressure is 3.1 MPa at the core inlet with a 1.4-MPa pressure drop through the core region. Finally, in order to make the resources available for experimentation, the fuel is designed to provide a 17-d fuel cycle with an additional 4 d planned in each cycle for the refueling process. This report examines the codes and models used to develop the thermal-hydraulic design for ANS, as well as the correlations and physical data; evaluates thermal-hydraulic uncertainties; reports on thermal-hydraulic design and safety analysis; describes experimentation in support of the ANS reactor design and safety analysis; and provides an overview of the experimental plan

The Steady State Calculation for SMART with MIDAS/SMR

International Nuclear Information System (INIS)

Park, Jong Hwa; Kim, Dong Ha; Chung, Young Jong; Park, Sun Hee; Cho, Seong Won

2010-01-01

KAERI is developing a new concept of reactor that all the main components such as the steam generator, the coolant pumps and the pressurizer are located inside the reactor vessel. Before the severe accident sequences are estimated, it is prerequisite that MIDAS code predicts the steady state conditions properly. But MIDAS code does not include the heat transfer model for the helical tube. Therefore, the heat transfer models for the helical tube from TASS/SMR-S were implemented into MIDAS code. To estimate the validity of the implemented heat transfer correlations for the helical tube and the input data, the steady state was recalculated with MIDAS/SMR based on design level 2 and compared with the design values

Thermal response of a pin-type fusion reactor blanket during steady and transient reactor operation

International Nuclear Information System (INIS)

Grotz, S.; Ghoniem, N.M.

1986-02-01

The thermal analysis of the blanket examines both the steady-state and transient reactor operations. The steady-state analysis covers full power and fractional power operation whereas the transient analysis examines the effects of power ramps and blanket preheat. The blanket configuration chosen for this study is a helium cooled solid breeder design. We first discuss the full power, steady-state temperature fields in the first wall, beryllium rods, and breeder rods. Next we examine the effects of fractional power on coolant flow and temperature field distributions. This includes power plateaus of 10%, 20%, 50%, 80%, and 100% of full power. Also examined are the restrictions on the rates of power ramping between plateaus. Finally we discuss the power and time requirements for pre-heating the primary from cold iron conditions up to startup temperature (250 0 C)

CAT-D-T tokamaks

International Nuclear Information System (INIS)

Greenspan, E.; Blue, T.; Miley, G.H.

1981-01-01

The domains of plasma fuel cycles bounded by the D-T and Cat-D, and by the D-T and SCD modes of operation are examined. These domains, referred to as, respectively, the Cat-D-T and SCD-T modes of operation, are characterized by the number (γ) of tritons per fusion neutron available from external (to the plasma) sources. Two external tritium sources are considered - the blankets of the Cat-D-T (SCD-T) reactors and fission reactors supported by the Cat-D-T (SCD-T) driven hybrid reactors. It is found that by using 6 Li for the active material of the control elements of the fission reactors, it is possible to achieve γ values close to unity. Cat-D-T tokamaks could be designed to have smaller size, higher power density, lower magnetic field and even lower plasma temperature than Cat-D tokamaks; the difference becomes significant for γ greater than or equal to .75. The SCD-T mode of operation appears to be even more attractive. Promising applications identified for these Cat-D-T and SCD-T modes of operation include hybrid reactors, fusion synfuel factories and fusion reactors which have difficulty in providing all their tritium needs

An equation oriented approach to steady state flowsheeting of methanol synthesis loop

International Nuclear Information System (INIS)

Fathikalajahi, J.; Baniadam, M.; Rahimpour, M.R.

2008-01-01

An equation-oriented approach was developed for steady state flowsheeting of a commercial methanol plant. The loop consists of fixed bed reactor, flash separator, preheater, coolers, and compressor. For steady sate flowsheeting of the plant mathematical model of reactor and other units are needed. Reactor used in loop is a Lurgi type and its configuration is rather complex. Previously reactor and flash separator are modeled as two important units of plant. The model is based on mass and energy balances in each equipment and utilizing some auxiliary equations such as rate of reaction and thermodynamics model for activity coefficients of liquid. In order to validate the mathematical model for the synthesis loop, some simulation data were performed using operating conditions and characteristics of the commercial plant. The good agreement between the steady state simulation results and the plant data shows the validity of the model

Development of a 3D consistent 1D neutronics model for reactor core simulation

International Nuclear Information System (INIS)

Lee, Ki Bog; Joo, Han Gyu; Cho, Byung Oh; Zee, Sung Quun

2001-02-01

In this report a 3D consistent 1D model based on nonlinear analytic nodal method is developed to reproduce the 3D results. During the derivation, the current conservation factor (CCF) is introduced which guarantees the same axial neutron currents obtained from the 1D equation as the 3D reference values. Furthermore in order to properly use 1D group constants, a new 1D group constants representation scheme employing tables for the fuel temperature, moderator density and boron concentration is developed and functionalized for the control rod tip position. To test the 1D kinetics model with CCF, several steady state and transient calculations were performed and compared with 3D reference values. The errors of K-eff values were reduced about one tenth when using CCF without significant computational overhead. And the errors of power distribution were decreased to the range of one fifth or tenth at steady state calculation. The 1D kinetics model with CCF and the 1D group constant functionalization employing tables as a function of control rod tip position can provide preciser results at the steady state and transient calculation. Thus it is expected that the 1D kinetics model derived in this report can be used in the safety analysis, reactor real time simulation coupled with system analysis code, operator support system etc.

Prospects for Tokamak Fusion Reactors

International Nuclear Information System (INIS)

Sheffield, J.; Galambos, J.

1995-01-01

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

Thermal-hydraulic Fortran program for steady-state calculations of plate-type fuel research reactors

Directory of Open Access Journals (Sweden)

Khedr Ahmed

2008-01-01

Full Text Available The safety assessment of research and power reactors is a continuous process covering their lifespan and requiring verified and validated codes. Power reactor codes all over the world are well established and qualified against real measuring data and qualified experimental facilities. These codes are usually sophisticated, require special skills and consume a lot of running time. On the other hand, most research reactor codes still require much more data for validation and qualification. It is, therefore, of benefit to any regulatory body to develop its own codes for the review and assessment of research reactors. The present paper introduces a simple, one-dimensional Fortran program called THDSN for steady-state thermal-hydraulic calculations of plate-type fuel research reactors. Besides calculating the fuel and coolant temperature distributions and pressure gradients in an average and hot channel, the program calculates the safety limits and margins against the critical phenomena encountered in research reactors, such as the onset of nucleate boiling, critical heat flux and flow instability. Well known thermal-hydraulic correlations for calculating the safety parameters and several formulas for the heat transfer coefficient have been used. The THDSN program was verified by comparing its results for 2 and 10 MW benchmark reactors with those published in IAEA publications and a good agreement was found. Also, the results of the program are compared with those published for other programs, such as the PARET and TERMIC.

The Tokamak Fusion Test Reactor D-T modifications and operations

International Nuclear Information System (INIS)

1992-01-01

This Environmental Assessment (EA) was prepared in accordance with the National Environmental Policy Act (NEPA) of 1969, as amended, in support of the Department of Energy's proposal for the Tokamak Fusion Test Reactor (TFTR) D-T program. The objective of the proposed D-T program is to take the initial step in studying the effects of alpha particle heating and transport in a magnetic fusion device. These studies would enable the successful completion of the original TFTR program objectives, and would support the research and development needs of the Burning Plasma Experiment, BPX (formerly the Compact Ignition Tokamak (CIT)) and International Thermonuclear Experimental Reactor (ITER) in the areas of alpha particle physics, tritium retention, alpha particle diagnostic development, and tritium handling

Comparison of contrast-enhanced T1-weighted and 3D constructive interference in steady state images for predicting outcome after hearing-preservation surgery for vestibular schwannoma

Energy Technology Data Exchange (ETDEWEB)

Kocaoglu, M.; Bulakbasi, N.; Ucoz, T.; Ustunsoz, B.; Tayfun, C.; Somuncu, I. [GATA Department of Radiology, 06018, Etlik, Ankara (Turkey); Pabuscu, Y. [Department of Radiology, Celal Bayar University, Manisa (Turkey)

2003-07-01

We compared contrast-enhanced T1-weighted and 3D constructive interference in steady state (CISS) sequences for demonstrating possible prognostic factors in hearing-preservation surgery for vestibular schwannoma. We studied 22 patients with vestibular schwannomas having hearing-preservation surgery. Postoperatively six (27%) had a facial palsy and eight (36%) had hearing loss. There was a significant correlation between the size of the tumour and facial palsy (r=-0.72). Both techniques adequately demonstrated all tumours. Involvement of the fundus of the internal auditory canal (IAC) and a small distance between the lateral border of the tumour and the fundus were correlated significantly with hearing loss (r=-0.81 and -0.75, respectively). The 3D-CISS sequence, by virtue of its high contrast resolution was superior to T1-weighted images (P<0.05) for detection of the fundal involvement. The direction of displacement of the facial nerve did not correlate with facial palsy or hearing loss. We think that 3D-CISS images better show the features influencing surgical outcome, but that contrast-enhanced T1-weighted images are required for diagnosis. (orig.)

Comparison of contrast-enhanced T1-weighted and 3D constructive interference in steady state images for predicting outcome after hearing-preservation surgery for vestibular schwannoma

International Nuclear Information System (INIS)

Kocaoglu, M.; Bulakbasi, N.; Ucoz, T.; Ustunsoz, B.; Tayfun, C.; Somuncu, I.; Pabuscu, Y.

2003-01-01

We compared contrast-enhanced T1-weighted and 3D constructive interference in steady state (CISS) sequences for demonstrating possible prognostic factors in hearing-preservation surgery for vestibular schwannoma. We studied 22 patients with vestibular schwannomas having hearing-preservation surgery. Postoperatively six (27%) had a facial palsy and eight (36%) had hearing loss. There was a significant correlation between the size of the tumour and facial palsy (r=-0.72). Both techniques adequately demonstrated all tumours. Involvement of the fundus of the internal auditory canal (IAC) and a small distance between the lateral border of the tumour and the fundus were correlated significantly with hearing loss (r=-0.81 and -0.75, respectively). The 3D-CISS sequence, by virtue of its high contrast resolution was superior to T1-weighted images (P<0.05) for detection of the fundal involvement. The direction of displacement of the facial nerve did not correlate with facial palsy or hearing loss. We think that 3D-CISS images better show the features influencing surgical outcome, but that contrast-enhanced T1-weighted images are required for diagnosis. (orig.)

Proteotyping of laboratory-scale biogas plants reveals multiple steady-states in community composition.

Science.gov (United States)

Kohrs, F; Heyer, R; Bissinger, T; Kottler, R; Schallert, K; Püttker, S; Behne, A; Rapp, E; Benndorf, D; Reichl, U

2017-08-01

Complex microbial communities are the functional core of anaerobic digestion processes taking place in biogas plants (BGP). So far, however, a comprehensive characterization of the microbiomes involved in methane formation is technically challenging. As an alternative, enriched communities from laboratory-scale experiments can be investigated that have a reduced number of organisms and are easier to characterize by state of the art mass spectrometric-based (MS) metaproteomic workflows. Six parallel laboratory digesters were inoculated with sludge from a full-scale BGP to study the development of enriched microbial communities under defined conditions. During the first three month of cultivation, all reactors (R1-R6) were functionally comparable regarding biogas productions (375-625 NL L reactor volume -1 d -1 ), methane yields (50-60%), pH values (7.1-7.3), and volatile fatty acids (VFA, 1 gNH 3 L -1 ) showed an increase to pH 7.5-8.0, accumulation of acetate (>10 mM), and decreasing biogas production (<125 NL L reactor volume -1 d -1 ). Tandem MS (MS/MS)-based proteotyping allowed the identification of taxonomic abundances and biological processes. Although all reactors showed similar performances, proteotyping and terminal restriction fragment length polymorphisms (T-RFLP) fingerprinting revealed significant differences in the composition of individual microbial communities, indicating multiple steady-states. Furthermore, cellulolytic enzymes and cellulosomal proteins of Clostridium thermocellum were identified to be specific markers for the thermophilic reactors (R3, R4). Metaproteins found in R3 indicated hydrogenothrophic methanogenesis, whereas metaproteins of acetoclastic methanogenesis were identified in R4. This suggests not only an individual evolution of microbial communities even for the case that BGPs are started at the same initial conditions under well controlled environmental conditions, but also a high compositional variance of microbiomes under

Potential of incineration of long-life fission products from fission energy system by D-T and D-D fusion reactors

International Nuclear Information System (INIS)

Sekimoto, H.; Takashima, H.

2001-01-01

The incineration of LLFPs, all of which can not be incinerated with only the fast reactor without isotope separation is studied by employing the DT and DD fusion reactors. The requirement of production of tritium for the DT reactor is severe and the thickness of the blanket should be decreased considerably to incinerate the considerable amount of LLFPs. On the other hand the DD fusion reactor is free from the neutron economy constraint and can incinerate all LLFPs. The pure DD reactor can also show the excellent performance to reduce the first wall loading less than 1 MW/m 2 even for total LLFP incineration. By raising the wall loading to the design limit, the D-D reactor can incinerate the LLFPs from several fast reactors. When the fusion reactor is utilized as an energy producer, plasma confinement is very difficult problem, especially for the D-D reactor compared to the D-T reactor. However, when it is utilized as an incinerator of LLFP, this problem becomes considerably easier. Therefore, the incineration of LLFP is considered as an attractive subject for the D-D reactor. (author)

Potential of incineration of long-life fission products from fission energy system by D-T and D-D fusion reactors

International Nuclear Information System (INIS)

Sekimoto, Hiroshi; Takashima, Hiroaki

1999-01-01

The incineration of LLFPs, all of which can not be incinerated with only the fast reactor without isotope separation is studied by employing the DT and DD fusion reactors. The requirement of production of tritium for the DT reactor is severe and the thickness of the blanket should be decreased considerably to incinerate the considerable amount of LLFPs. On the other hand the DD fusion reactor is free from the neutron economy constraint and can incinerate all LLFPs. The pure DD reactor can also show the excellent performance to reduce the first wall loading less than 1 MW/m 2 even for total LLFP incineration. By raising the wall loading to the design limit, the D-D reactor can incinerate the LLFPs from several fast reactors. When the fusion reactor is utilized as an energy producer, plasma confinement is very difficult problem, especially for the D-D reactor compared to the D-T reactor. However, when it is utilized as an incinerator of LLFP, this problem becomes considerably easier. Therefore, the incineration of LLFP is considered as an attractive subject for the D-D reactor. (author)

Numerical method for three dimensional steady-state two-phase flow calculations

International Nuclear Information System (INIS)

Raymond, P.; Toumi, I.

1992-01-01

This paper presents the numerical scheme which was developed for the FLICA-4 computer code to calculate three dimensional steady state two phase flows. This computer code is devoted to steady state and transient thermal hydraulics analysis of nuclear reactor cores 1,3 . The first section briefly describes the FLICA-4 flow modelling. Then in order to introduce the numerical method for steady state computations, some details are given about the implicit numerical scheme based upon an approximate Riemann solver which was developed for calculation of flow transients. The third section deals with the numerical method for steady state computations, which is derived from this previous general scheme and its optimization. We give some numerical results for steady state calculations and comparisons on required CPU time and memory for various meshing and linear system solvers

Plasma flow reactor for steady state monitoring of physical and chemical processes at high temperatures.

Science.gov (United States)

Koroglu, Batikan; Mehl, Marco; Armstrong, Michael R; Crowhurst, Jonathan C; Weisz, David G; Zaug, Joseph M; Dai, Zurong; Radousky, Harry B; Chernov, Alex; Ramon, Erick; Stavrou, Elissaios; Knight, Kim; Fabris, Andrea L; Cappelli, Mark A; Rose, Timothy P

2017-09-01

We present the development of a steady state plasma flow reactor to investigate gas phase physical and chemical processes that occur at high temperature (1000 flow injector). We have modeled the system using computational fluid dynamics simulations that are bounded by measured temperatures. In situ line-of-sight optical emission and absorption spectroscopy have been used to determine the structures and concentrations of molecules formed during rapid cooling of reactants after they pass through the plasma. Emission spectroscopy also enables us to determine the temperatures at which these dynamic processes occur. A sample collection probe inserted from the open end of the reactor is used to collect condensed materials and analyze them ex situ using electron microscopy. The preliminary results of two separate investigations involving the condensation of metal oxides and chemical kinetics of high-temperature gas reactions are discussed.

Linking of FRAP-T, FRAPCON and RELAP-4 codes for transient analysis and accidents of light water reactors fuel rods

International Nuclear Information System (INIS)

Marra Neto, A.; Silva, A.T. e; Sabundjian, G.; Freitas, R.L.; Neves Conti, T. das.

1991-09-01

The computer codes FRAP-T, FRAPCON and RELAP-4 have been linked for the fuel rod behavior analysis under transients and hypothetical accidents in light water reactors. The results calculated by thermal hydraulic code RELAP-4 give input in file format into the transient fuel analysis code FRAP-T. If the effect of fuel burnup is taken into account, the fuel performance code FRAPCON should provide the initial steady state data for thhe transient analysis. With the thermal hydraulic boundary conditions provided by RELAP-4 (MOD3), FRAP-T6 is used to analyse pressurized water reactor fuel rod behavior during the blowdown phase under large break loss of coolant accident conditions. Two cases have been analysed: without and with initialization from FRAPCON-2 steady state data. (author)

Method for calculating the steady-state distribution of tritium in a molten-salt breeder reactor plant

International Nuclear Information System (INIS)

Briggs, R.B.; Nestor, C.W.

1975-04-01

Tritium is produced in molten salt reactors primarily by fissioning of uranium and absorption of neutrons by the constituents of the fuel carrier salt. At the operating temperature of a large power reactor, tritium is expected to diffuse from the primary system through pipe and vessel walls to the surroundings and through heat exchanger tubes into the secondary system which contains a coolant salt. Some tritium will pass from the secondary system into the steam power system. This report describes a method for calculating the steady state distribution of tritium in a molten salt reactor plant and a computer program for making the calculations. The method takes into account the effects of various processes for removing tritium, the addition of hydrogen or hydrogenous compounds to the primary and secondary systems, and the chemistry of uranium in the fuel salt. Sample calculations indicate that 30 percent or more of the tritium might reach the steam system in a large power reactor unless special measures are taken to confine the tritium. (U.S.)

Application of MCNPX 2.7.D for reactor core management at the research reactor BR2

International Nuclear Information System (INIS)

Kalcheva, Silva; Koonen, Edgar

2011-01-01

The paper discusses application of the Monte Carlo burn up code MCNPX 2.7.D for whole core criticality and depletion analysis of the Material Testing Research Reactor BR2 at SCK-CEN in Mol, Belgium. Two different approaches in the use of MCNPX 2.7.D are presented. The first methodology couples the evolution of fuel depletion, evaluated by MCNPX 2.7.D in an infinite lattice with a steady-state 3-D power distribution in the full core model. The second method represents fully automatic whole core depletion and criticality calculations in the detailed 3-D heterogeneous geometry model of the BR2 reactor. The accuracy of the method and computational time as function of the number of used unique burn up materials in the model are being studied. The depletion capabilities of MCNPX 2.7.D are compared vs. the developed at the BR2 reactor department MCNPX & ORIGEN-S combined method. Testing of MCNPX 2.7.D on the criticality measurements at the BR2 reactor is presented. (author)

Mathematical Modeling and Simulation of the Dehydrogenation of Ethyl Benzene to Form Styrene Using Steady-State Fixed Bed Reactor

Directory of Open Access Journals (Sweden)

Zaidon M. Shakoor

2013-05-01

Full Text Available In this research, two models are developed to simulate the steady state fixed bed reactor used for styrene production by ethylbenzene dehydrogenation. The first is one-dimensional model, considered axial gradient only while the second is two-dimensional model considered axial and radial gradients for same variables.The developed mathematical models consisted of nonlinear simultaneous equations in multiple dependent variables. A complete description of the reactor bed involves partial, ordinary differential and algebraic equations (PDEs, ODEs and AEs describing the temperatures, concentrations and pressure drop across the reactor was given. The model equations are solved by finite differences method. The reactor models were coded with Mat lab 6.5 program and various numerical techniques were used to obtain the desired solution.The simulation data for both models were validated with industrial reactor results with a very good concordance.

Thermal Hydraulic Fortran Program for Steady State Calculations of Plate Type Fuel Research Reactors

International Nuclear Information System (INIS)

Khedr, H.

2008-01-01

The safety assessment of Research and Power Reactors is a continuous process over their life and that requires verified and validated codes. Power Reactor codes all over the world are well established and qualified against a real measuring data and qualified experimental facilities. These codes are usually sophisticated, require special skills and consume much more running time. On the other hand, most of the Research Reactor codes still requiring more data for validation and qualification. Therefore it is benefit for a regulatory body and the companies working in the area of Research Reactor assessment and design to have their own program that give them a quick judgment. The present paper introduces a simple one dimensional Fortran program called THDSN for steady state best estimate Thermal Hydraulic (TH) calculations of plate type fuel RRs. Beside calculating the fuel and coolant temperature distribution and pressure gradient in an average and hot channel the program calculates the safety limits and margins against the critical phenomena encountered in RR such as the burnout heat flux and the onset of flow instability. Well known TH correlations for calculating the safety parameters are used. THDSN program is verified by comparing its results for 2 and 10 MW benchmark reactors with that published in IAEA publications and good agreement is found. Also the program results are compared with those published for other programs such as PARET and TERMIC. An extension for this program is underway to cover the transient TH calculations

Three-dimensional double-echo steady-state (3D-DESS) magnetic resonance imaging of the knee. Establishment of flip angles for evaluation of cartilage at 1.5 T and 3.0 T

International Nuclear Information System (INIS)

Moriya, Susumu; Miki, Yukio; Matsuno, Yukako; Okada, Masayo

2012-01-01

Background: The effect of flip angle (FA) on synovial fluid and cartilage signal and on image contrast using three-dimensional double-echo steady-state (3D-DESS) sequence have only been performed with 1.0-T but not with 1.5-T or 3.0-T scanners. Purpose: To identify the FA that gives the maximum synovial fluid and cartilage values, and to identify the FA at which maximum values of synovial fluid-cartilage contrast-to-noise ratio (CNR) in 3D-DESS sequences when 1.5-T and 3.0-T scanners are used. Material and Methods: Using 3D-DESS with water-excitation pulse, mid-sagittal plane images of the knees of 10 healthy volunteers (5 men, 5 women; age range, 21-42 years) were obtained with FA varying from 10 deg to 90 deg. Synovial fluid signals, cartilage signals, and background were measured at each FA, and the FA that gave the highest synovial fluid and cartilage values was obtained. Synovial fluid-cartilage CNR was also calculated, and the FA that gave the largest CNR was obtained. Results: At 1.5 T, the maximum synovial fluid signal was at FA 90 deg, and the maximum cartilage signal was at FA 30 deg. Synovial fluid-cartilage CNR was highest at FA 90 deg (P < 0.05). At 3.0 T, the maximum synovial fluid signal was at FA 90 deg, and the maximum cartilage signal was at FA 20 deg. Synovial fluid-cartilage CNR was highest at FA 90 deg (P < 0.05). Conclusion: In order to improve the visibility of cartilage itself, FA settings of 30 deg at 1.5 T and 20 deg at 3.0 T are apparently ideal. For observing the cartilage surface, the most effective FA setting is 90 deg for both 1.5 T and 3.0 T

Study on small long-life LBE cooled fast reactor with CANDLE burn-up. Part 1. Steady state research

International Nuclear Information System (INIS)

Yan, Mingyu; Sekimoto, Hiroshi

2008-01-01

Small long-life reactor is required for some local areas. CANDLE small long-life fast reactor which does not require control rods, mining, enrichment and reprocessing plants can satisfy this demand. In a CANDLE reactor, the shapes of neutron flux, nuclide number densities and power density distributions remain constant and only shift in axial direction. The core with 1.0 m radius, 2.0 m length can realize CANDLE burn-up with nitride (enriched N-15) natural uranium as fresh fuel. Lead-Bismuth is used as coolant. From steady state analysis, we obtained the burn-up velocity, output power distribution, core temperature distribution, etc. The burn-up velocity is less than 1.0 cm/year that enables a long-life design easily. The core averaged discharged fuel burn-up is about 40%. (author)

MARS input data for steady-state calculation of ATLAS

International Nuclear Information System (INIS)

Park, Hyun Sik; Euh, D. J.; Choi, K. Y.; Kwon, T. S.; Jeong, J. J.; Baek, W. P.

2004-12-01

An integral effect test loop for Pressurized Water Reactors (PWRs), the ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation), is under construction by Thermal-Hydraulics Safety Research Division in Korea Atomic Energy Research Institute (KAERI). This report includes calculation sheets of the input for the best-estimate system analysis code, the MARS code, based on the ongoing design features of ATLAS. The ATLAS facility has been designed to have the length scale of 1/2 and area scale of 1/144 compared with the reference plant, APR1400. The contents of this report are divided into three parts: (1) core and reactor vessel, (2) steam generator and steam line, and (3) primary piping, pressurizer and reactor coolant pump. The steady-state analysis for the ATLAS facility will be performed based on these calculation sheets, and its results will be applied to the detailed design of ATLAS. Additionally, the calculation results will contribute to getting optimum test conditions and preliminary operational test conditions for the steady-state and transient experiments

Pellet acceleration studies relating to the refuelling of a steady-state fusion reactor

International Nuclear Information System (INIS)

Dimock, D.; Jensen, K.; Jensen, V.O.; Joergensen, L.W.; Pecseli, H.L.; Soerensen, H.; Oester, F.

1975-11-01

Several methods for refuelling a steady state-fusion reactor have been proposed, and the pellet method seems advantageous if the pellet can be accelerated to the necessary velocity. A study group was formed to analyze this acceleration problem. Two pellet velocity values were considered: 10 4 m/s and 300 m/s. A pellet velocity of 10 4 m/s may be suitable in the case of a reactor, whereas 300 m/s is believed to be a reasonable velocity at which to perform realistic ablation experiments in the near future. A pneumatic acceleration method was found promising. The pressure is either supplied separately or by evaporation of a part of the pellet. In the latter case, a spark behind the pellet should provide the evoporation and the necessary heating of the driving gas. A preliminary test at room temperature with pellets made of beeswax (the density being ten times that of solid hydrogen, and plastic properties similar to those of solid hydrogen) resulted in a pellet velocity of 100 m/s at a modest value of the energy supplied to the spark. (Auth.)

Progress and prospect of true steady state operation with RF

Directory of Open Access Journals (Sweden)

Jacquinot Jean

2017-01-01

Full Text Available Operation of fusion confinement experiments in full steady state is a major challenge for the development towards fusion energy. Critical to achieving this goal is the availability of actively cooled plasma facing components and auxiliary systems withstanding the very harsh plasma environment. Equally challenging are physics issues related to achieving plasma conditions and current drive efficiency required by reactor plasmas. RF heating and current drive systems have been key instruments for obtaining the progress made until today towards steady state. They hold all the records of long pulse plasma operation both in tokamaks and in stellarators. Nevertheless much progress remains to be made in particular for integrating all the requirements necessary for maintaining in steady state the density and plasma pressure conditions of a reactor. This is an important stated aim of ITER and of devices equipped with superconducting magnets. After considering the present state of the art, this review will address the key issues which remain to be solved both in physics and technology for reaching this goal. They constitute very active subjects of research which will require much dedicated experimentation in the new generation of superconducting devices which are now in operation or becoming close to it.

Bioaccumulation factors and the steady state assumption for cesium isotopes in aquatic foodwebs near nuclear facilities.

Science.gov (United States)

Rowan, D J

2013-07-01

Steady state approaches, such as transfer coefficients or bioaccumulation factors, are commonly used to model the bioaccumulation of (137)Cs in aquatic foodwebs from routine operations and releases from nuclear generating stations and other nuclear facilities. Routine releases from nuclear generating stations and facilities, however, often consist of pulses as liquid waste is stored, analyzed to ensure regulatory compliance and then released. The effect of repeated pulse releases on the steady state assumption inherent in the bioaccumulation factor approach has not been evaluated. In this study, I examine the steady state assumption for aquatic biota by analyzing data for two cesium isotopes in the same biota, one isotope in steady state (stable (133)Cs) from geologic sources and the other released in pulses ((137)Cs) from reactor operations. I also compare (137)Cs bioaccumulation factors for similar upstream populations from the same system exposed solely to weapon test (137)Cs, and assumed to be in steady state. The steady state assumption appears to be valid for small organisms at lower trophic levels (zooplankton, rainbow smelt and 0+ yellow perch) but not for older and larger fish at higher trophic levels (walleye). Attempts to account for previous exposure and retention through a biokinetics approach had a similar effect on steady state, upstream and non-steady state, downstream populations of walleye, but were ineffective in explaining the more or less constant deviation between fish with steady state exposures and non-steady state exposures of about 2-fold for all age classes of walleye. These results suggest that for large, piscivorous fish, repeated exposure to short duration, pulse releases leads to much higher (137)Cs BAFs than expected from (133)Cs BAFs for the same fish or (137)Cs BAFs for similar populations in the same system not impacted by reactor releases. These results suggest that the steady state approach should be used with caution in any

Comparison between temperature distributions of an annular fuel rod of circular cross-section and of a hemoglobin shaped cross-section rod for PWR reactors in steady state conditions

International Nuclear Information System (INIS)

Oliveira, Maria Vitória A. de; Alvim, Antônio Carlos Marques

2017-01-01

The objective of this work is to make a comparison between the temperature distributions of an annular fuel rod of circular cross-section and a hemoglobin shaped cross-section for PWR reactors in steady state conditions. The motivation for this article is due to the fact that the symmetric form of the red globules particles allows the O 2 gases to penetrate the center of the cell homogeneously and quickly. The diffusion equation of gases in any environment is very similar to the heat diffusion equation: Diffusion - Fick's Law; Heat Flow - Fourier; where, the temperature (T) replaces the concentration (c). In previous works the comparison between the shape of solid fuel rods with circular section, and a with hemoglobin-shaped cross-section has proved that this new format optimizes the heat transfer, decreasing the thermal resistance between the center of the UO 2 pellets and the clad. With this, a significant increase in the specific power of the reactor was made possible (more precisely a 23% increase). Currently, the advantages of annular fuel rods are being studied and recent works have shown that 12 x 12 arrays of annular fuel rods perform better, increasing the specific power of the reactor by at least 20% in relation to solid fuel rods, without affecting the safety of the reactor. Our proposal is analyzing the temperature distribution in annular fuel rods with cross sections with red blood cell shape and compare with the theoretical results of the annular fuel rods of circular cross section, initially in steady state. (author)

Comparison between temperature distributions of an annular fuel rod of circular cross-section and of a hemoglobin shaped cross-section rod for PWR reactors in steady state conditions

Energy Technology Data Exchange (ETDEWEB)

Oliveira, Maria Vitória A. de; Alvim, Antônio Carlos Marques, E-mail: moliveira@con.ufrj.br, E-mail: alvim@nuclear.ufrj.br [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

2017-07-01

The objective of this work is to make a comparison between the temperature distributions of an annular fuel rod of circular cross-section and a hemoglobin shaped cross-section for PWR reactors in steady state conditions. The motivation for this article is due to the fact that the symmetric form of the red globules particles allows the O{sub 2} gases to penetrate the center of the cell homogeneously and quickly. The diffusion equation of gases in any environment is very similar to the heat diffusion equation: Diffusion - Fick's Law; Heat Flow - Fourier; where, the temperature (T) replaces the concentration (c). In previous works the comparison between the shape of solid fuel rods with circular section, and a with hemoglobin-shaped cross-section has proved that this new format optimizes the heat transfer, decreasing the thermal resistance between the center of the UO{sub 2} pellets and the clad. With this, a significant increase in the specific power of the reactor was made possible (more precisely a 23% increase). Currently, the advantages of annular fuel rods are being studied and recent works have shown that 12 x 12 arrays of annular fuel rods perform better, increasing the specific power of the reactor by at least 20% in relation to solid fuel rods, without affecting the safety of the reactor. Our proposal is analyzing the temperature distribution in annular fuel rods with cross sections with red blood cell shape and compare with the theoretical results of the annular fuel rods of circular cross section, initially in steady state. (author)

TRANP - a computer code for digital simulation of steady - state and transient behavior of a pressurizer water reactor primary circuit

International Nuclear Information System (INIS)

Chalhoub, E.S.

1980-09-01

A digital computer code TRANP was developed to simulate the steady-state and transient behavior of a pressurizer water reactor primary circuit. The development of this code was based on the combining of three codes already developed for the simulation of a PWR core, a pressurizer, a steam generator and a main coolant pump, representing the primary circuit components. (Author) [pt

A pilot application of the RELAP file to the steady state and transient analysis of a test section inside the BR2 reactor

International Nuclear Information System (INIS)

Ferri, M. G.; D'Auria, F.; Forasassi, G.; Giot, M.

2000-01-01

BR2 is a material test reactor sited in the Belgian Nuclear Research Centre in Mol. The main research programs carried out in BR2 are related to the safety of nuclear reactor structural materials and fuels, in normal and accidental conditions, plant lifetime evaluation and ageing of components. In this framework, a computer program that allows the performance of detailed, steady state analysis of several kinds of in-pile sections with an axisymmetrical geometry has been developed. Furthermore, comparing its results with those of the well known, extensively used, Relap5/Mod 3.2 code on a test problem has validated this program. This was performed in three steps: 1. modalisation development of a subsystem of a typical in-pile section. 2. steady state analysis and comparison with the above-mentioned program. 3. transient simulation of the same system; the considered transient consists of a loss of coolant flow. (author)

The analysis of the annular fuel performance in steady state condition by using AFPAC code

International Nuclear Information System (INIS)

He Xiaojun; Ji Songtao; Zhang Yingchao

2012-01-01

The fuel performance code AFPAC v1.0 is used to analyze annular fuel's behavior under steady state conditions, including neutronics, thermal hydraulic, rod deformation, fission gas release and rod internal pressure. The calculation results show that: 1) Annular fuel has a good steady irradiation performance at 150% power level as current LWRs' with burnup up to 50 GWd/t, and all parameters, such as temperature, rod internal pressure and rod deformation, are meet the rod design criteria for current fuel of PWRs: 2) Compared to the solid fuel under the same irradiation condition. annular fuel has lower temperature, smaller deformation, lower fission gas release and lower pressure at EOL. From the point of view of steady irradiation performance, the safety of reactors can significantly improved by u sing the annular fuel. (authors)

Rapid 3D in vivo 1H human lung respiratory imaging at 1.5 T using ultra-fast balanced steady-state free precession.

Science.gov (United States)

Pusterla, Orso; Bauman, Grzegorz; Wielpütz, Mark O; Nyilas, Sylvia; Latzin, Philipp; Heussel, Claus P; Bieri, Oliver

2017-09-01

To introduce a reproducible, nonenhanced 1H MRI method for rapid in vivo functional assessment of the whole lung at 1.5 Tesla (T). At different respiratory volumes, the pulmonary signal of ultra-fast steady-state free precession (ufSSFP) follows an adapted sponge model, characterized by a respiratory index α. From the model, α reflects local ventilation-related information, is virtually independent from the lung density and thus from the inspiratory phase and breathing amplitude. Respiratory α-mapping is evaluated for healthy volunteers and patients with obstructive lung disease from a set of five consecutive 3D ultra-fast steady-state free precession (ufSSFP) scans performed in breath-hold and at different inspiratory volumes. For the patients, α-maps were compared with CT, dynamic contrast-enhanced MRI (DCE-MRI), and Fourier decomposition (FD). In healthy volunteers, respiratory α-maps showed good reproducibility and were homogeneous on iso-gravitational planes, but showed a gravity-dependent respiratory gradient. In patients with obstructive pulmonary disease, the functional impairment observed in respiratory α-maps was associated with emphysematous regions present on CT images, perfusion defects observable on DCE-MRI, and impairments visualized on FD ventilation and perfusion maps. Respiratory α-mapping derived from multivolumetric ufSSFP provides insights into functional lung impairment and may serve as a reproducible and normative measure for clinical studies. Magn Reson Med 78:1059-1069, 2017. © 2016 International Society for Magnetic Resonance in Medicine. © 2016 International Society for Magnetic Resonance in Medicine.

Mirror Fusion Test Facility: an intermediate device to a mirror fusion reactor

International Nuclear Information System (INIS)

Karpenko, V.N.

1983-01-01

The Mirror Fusion Test Facility (MFTF-B) now under construction at Lawrence Livermore National Laboratory represents more than an order-of-magnitude step from earlier magnetic-mirror experiments toward a future mirror fusion reactor. In fact, when the device begins operating in 1986, the Lawson criteria of ntau = 10 14 cm -3 .s will almost be achieved for D-T equivalent operation, thus signifying scientific breakeven. Major steps have been taken to develop MFTF-B technologies for tandem mirrors. Steady-state, high-field, superconducting magnets at reactor-revelant scales are used in the machine. The 30-s beam pulses, ECRH, and ICRH will also introduce steady-state technologies in those systems

Transient and steady-state currents in epoxy resin

International Nuclear Information System (INIS)

Guillermin, Christophe; Rain, Pascal; Rowe, Stephen W

2006-01-01

Charging and discharging currents have been measured in a diglycidyl ether of bisphenol-A epoxy resin with and without silica fillers, below and above its glass transition temperature T g = 65 deg. C. Both transient and steady-state current densities have been analysed. The average applied fields ranged from 3 to 35 kV mm -1 with a sample thickness of 0.5 mm. Above T g , transient currents suggested a phenomenon of charge injection forming trapped space charges even at low fields. Steady-state currents confirmed that the behaviour was not Ohmic and suggested Schottky-type injection. Below T g , the current is not controlled by the metal-dielectric interface but by the conduction in the volume: the current is Ohmic at low fields and both transient and steady-state currents suggest a phenomenon of space-charge limited currents at high fields. The field threshold is similar in the filler-free and the filled resin. Values in the range 12-17 kV mm -1 have been measured

Transient and steady-state currents in epoxy resin

Energy Technology Data Exchange (ETDEWEB)

Guillermin, Christophe [Schneider Electric Industries S.A.S., 37 quai Paul-Louis Merlin, 38050 Grenoble Cedex 9 (France); Rain, Pascal [Laboratoire d' Electrostatique et de Materiaux Dielectriques (LEMD), CNRS, 25 avenue des Martyrs, 38042 Grenoble Cedex 9 (France); Rowe, Stephen W [Schneider Electric Industries S.A.S., 37 quai Paul-Louis Merlin, 38050 Grenoble Cedex 9 (France)

2006-02-07

Charging and discharging currents have been measured in a diglycidyl ether of bisphenol-A epoxy resin with and without silica fillers, below and above its glass transition temperature T{sub g} = 65 deg. C. Both transient and steady-state current densities have been analysed. The average applied fields ranged from 3 to 35 kV mm{sup -1} with a sample thickness of 0.5 mm. Above T{sub g}, transient currents suggested a phenomenon of charge injection forming trapped space charges even at low fields. Steady-state currents confirmed that the behaviour was not Ohmic and suggested Schottky-type injection. Below T{sub g}, the current is not controlled by the metal-dielectric interface but by the conduction in the volume: the current is Ohmic at low fields and both transient and steady-state currents suggest a phenomenon of space-charge limited currents at high fields. The field threshold is similar in the filler-free and the filled resin. Values in the range 12-17 kV mm{sup -1} have been measured.

Physics design of advanced steady-state tokamak reactor A-SSTR2

International Nuclear Information System (INIS)

Nishio, Satoshi; Ushigusa, Kenkichi

2000-10-01

Based on design studies on the fusion power reactor such as the DEMO reactor SSTR, the compact power reactor A-SSTR and the DREAM reactor with a high environmental safety and high availability, a new concept of compact and economic fusion power reactor (A-SSTR2) with high safety and high availability is proposed. Employing high temperature superconductor, the toroidal filed coils supplies the maximum field of 23T on conductor which corresponds to 11T at the magnetic axis. A-SSTR2 (R p =6.2m, a p =1.5m, I p =12MA) has a fusion power of 4GW with β N =4. For an easy maintenance and for an enough support against a strong electromagnetic force on coils, a poloidal coils system has no center solenoid coils and consists of 6 coils located on top and bottom of the machine. Physics studies on the plasma equilibrium, controllability of the configuration, the plasma initiation and non-inductive current ramp-up, fusion power controllability and the diverter have shown the validity of the A-SSTR2 concept. (author)

SUPERENERGY-2: a multiassembly, steady-state computer code for LMFBR core thermal-hydraulic analysis

Energy Technology Data Exchange (ETDEWEB)

Basehore, K.L.; Todreas, N.E.

1980-08-01

Core thermal-hydraulic design and performance analyses for Liquid Metal Fast Breeder Reactors (LMFBRs) require repeated detailed multiassembly calculations to determine radial temperature profiles and subchannel outlet temperatures for various core configurations and subassembly structural analyses. At steady-state, detailed core-wide temperature profiles are required for core restraint calculations and subassembly structural analysis. In addition, sodium outlet temperatures are routinely needed for each reactor operating cycle. The SUPERENERGY-2 thermal-hydraulic code was designed specifically to meet these designer needs. It is applicable only to steady-state, forced-convection flow in LMFBR core geometries.

SUPERENERGY-2: a multiassembly, steady-state computer code for LMFBR core thermal-hydraulic analysis

International Nuclear Information System (INIS)

Basehore, K.L.; Todreas, N.E.

1980-08-01

Core thermal-hydraulic design and performance analyses for Liquid Metal Fast Breeder Reactors (LMFBRs) require repeated detailed multiassembly calculations to determine radial temperature profiles and subchannel outlet temperatures for various core configurations and subassembly structural analyses. At steady-state, detailed core-wide temperature profiles are required for core restraint calculations and subassembly structural analysis. In addition, sodium outlet temperatures are routinely needed for each reactor operating cycle. The SUPERENERGY-2 thermal-hydraulic code was designed specifically to meet these designer needs. It is applicable only to steady-state, forced-convection flow in LMFBR core geometries

Steady-state oxygen-solubility in niobium

International Nuclear Information System (INIS)

Schulze, K.; Jehn, H.

1977-01-01

During annealing of niobium in oxygen in certain temperature and pressure ranges steady states are established between the absorption of molecular oxygen and the evaporation of volatile oxides. The oxygen concentration in the niobium-oxygen α-solid solution is a function of oxygen pressure and temperature and has been redetermined in the ranges 10 -5 - 10 -2 Pa O 2 and 2,070 - 2,470 K. It follows differing from former results the equation csub(o) = 9.1 x 10 -6 x sub(po2) x exp (502000/RT) with csub(o) in at.-ppm, sub(po2) in Pa, T in K, R = 8.31 J x mol -1 x K -1 . The existence of steady states is limited to a temperature range from 1870 to 2470 K and to oxygen concentrations below the solubility limit given by solidus and solvus lines in the T-c diagram. In the experiments high-purity niobium wires with a specific electrical ratio rho (273 K)/rho(4.2 K) > 5,000 have been gassed under isothermal-isobaric conditions until the steady state has been reached. The oxygen concentration has been determined analytically by vacuum fusion extraction with platinum-flux technique as well as by electrical residual resistivity measurements at 4.2 K. (orig.) [de

Development of multi-physics code systems based on the reactor dynamics code DYN3D

Energy Technology Data Exchange (ETDEWEB)

Kliem, Soeren; Gommlich, Andre; Grahn, Alexander; Rohde, Ulrich [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany); Schuetze, Jochen [ANSYS Germany GmbH, Darmstadt (Germany); Frank, Thomas [ANSYS Germany GmbH, Otterfing (Germany); Gomez Torres, Armando M.; Sanchez Espinoza, Victor Hugo [Karlsruher Institut fuer Technologie (KIT), Eggenstein-Leopoldshafen (Germany)

2011-07-15

The reactor dynamics code DYN3D has been coupled with the CFD code ANSYS CFX and the 3D thermal hydraulic core model FLICA4. In the coupling with ANSYS CFX, DYN3D calculates the neutron kinetics and the fuel behavior including the heat transfer to the coolant. The physical data interface between the codes is the volumetric heat release rate into the coolant. In the coupling with FLICA4 only the neutron kinetics module of DYN3D is used. Fluid dynamics and related transport phenomena in the reactor's coolant and fuel behavior is calculated by FLICA4. The correctness of the coupling of DYN3D with both thermal hydraulic codes was verified by the calculation of different test problems. These test problems were set-up in such a way that comparison with the DYN3D stand-alone code was possible. This included steady-state and transient calculations of a mini-core consisting of nine real-size PWR fuel assemblies with ANSYS CFX/DYN3D as well as mini-core and a full core steady-state calculation using FLICA4/DYN3D. (orig.)

Development of multi-physics code systems based on the reactor dynamics code DYN3D

International Nuclear Information System (INIS)

Kliem, Soeren; Gommlich, Andre; Grahn, Alexander; Rohde, Ulrich; Schuetze, Jochen; Frank, Thomas; Gomez Torres, Armando M.; Sanchez Espinoza, Victor Hugo

2011-01-01

The reactor dynamics code DYN3D has been coupled with the CFD code ANSYS CFX and the 3D thermal hydraulic core model FLICA4. In the coupling with ANSYS CFX, DYN3D calculates the neutron kinetics and the fuel behavior including the heat transfer to the coolant. The physical data interface between the codes is the volumetric heat release rate into the coolant. In the coupling with FLICA4 only the neutron kinetics module of DYN3D is used. Fluid dynamics and related transport phenomena in the reactor's coolant and fuel behavior is calculated by FLICA4. The correctness of the coupling of DYN3D with both thermal hydraulic codes was verified by the calculation of different test problems. These test problems were set-up in such a way that comparison with the DYN3D stand-alone code was possible. This included steady-state and transient calculations of a mini-core consisting of nine real-size PWR fuel assemblies with ANSYS CFX/DYN3D as well as mini-core and a full core steady-state calculation using FLICA4/DYN3D. (orig.)

Simultaneous multislice triple-echo steady-state (SMS-TESS) T1 , T2 , PD, and off-resonance mapping in the human brain.

Science.gov (United States)

Heule, Rahel; Celicanin, Zarko; Kozerke, Sebastian; Bieri, Oliver

2018-02-21

To investigate the ability of simultaneous multislice triple-echo steady-state (SMS-TESS) imaging to provide quantitative maps of multiple tissue parameters, i.e., longitudinal and transverse relaxation times (T 1 and T 2 ), proton density (PD), and off-resonance (ΔB 0 ), in the human brain at 3T from a single scan. TESS acquisitions were performed in 2D mode to reduce motion sensitivity and accelerated by an SMS excitation scheme (CAIPIRINHA) with SENSE reconstruction. SMS-acceleration factors (R) of 2 and 4 were evaluated. The in vitro and in vivo validation process included standard reference scans to analyze the accuracy of T 1 , T 2 , and ΔB 0 estimates, as well as single-slice TESS measurements. For R = 2, the quantification of T 1 , T 2 , PD, and ΔB 0 was overall reliable with marginal noise enhancement. T 1 and T 2 values were in good agreement with the reference measurements and single-slice TESS. For R = 4, the agreement of ΔB 0 with the standard reference was excellent and the determination of T 1 , T 2 , and PD was reproducible; however, increased variations in T 1 and T 2 values with respect to single-slice TESS were observed. SMS-TESS has shown potential to offer rapid simultaneous T 1 , T 2 , PD, and ΔB 0 mapping of human brain tissues. © 2018 International Society for Magnetic Resonance in Medicine.

3-D steady analysis of flow in CRDM sewerage system

International Nuclear Information System (INIS)

Sun Yan; Liang Tiebo; Chen Zhihui; Zhao Jing; Zhang Yulong

2014-01-01

In order to obtain the flow state during sewer condition in Reactor and CRDM Sewerage system (RSE), this paper analyzes the 3-D steady flow in RSE by using Computational Fluid Dynamics (CFD) method. In the premise that the pressure drop of the RSE is known, the mass flow rate, the velocity and the type of flow in the system is obtained with the inverse method, which is proposed and validated to be applicable in the paper. The result shows that in the sewerage conditions, the type of flow in the RSE is turbulence flow, which is helpful to sewer drain. The study results give an reference for the design of RSE. (authors)

COOLOD-N2: a computer code, for the analyses of steady-state thermal-hydraulics in research reactors

International Nuclear Information System (INIS)

Kaminaga, Masanori

1994-03-01

The COOLOD-N2 code provides a capability for the analyses of the steady-state thermal-hydraulics of research reactors. This code is revised version of the COOLOD-N code, and is applicable not only for research reactors in which plate-type fuel is adopted, but also for research reactors in which rod-type fuel is adopted. In the code, subroutines to calculate temperature distribution in rod-type fuel have been newly added to the COOLOD-N code. The COOLOD-N2 code can calculate fuel temperatures under both forced convection cooling mode and natural convection cooling mode as well as COOLOD-N code. In the COOLOD-N2 code, a 'Heat Transfer package' is used for calculating heat transfer coefficient, DNB heat flux etc. The 'Heat Transfer package' is subroutine program and is especially developed for research reactors in which plate-type fuel is adopted. In case of rod-type fuel, DNB heat flux is calculated by both the 'Heat Transfer package' and Lund DNB heat flux correlation which is popular for TRIGA reactor. The COOLOD-N2 code also has a capability of calculating ONB temperature, the heat flux at onset of flow instability as well as DNB heat flux. (author)

Steady-state heat transfer in an inverted U-tube steam generator

International Nuclear Information System (INIS)

Boucher, T.J.

1986-01-01

Experimental results are presented involving U-tube steam generator tube bundle local heat transfer and fluid conditions during steady-state, full-power operations performed at high temperatures and pressures with conditions typical of a pressurized water reactor (15.0 MPa primary pressure, 600 K hot-leg fluid temperatures, 6.2 MPa secondary pressure). The MOD-2C facility represents the state-of-the-art in measurement of tube local heat transfer data and average tube bundle secondary fluid density at several elevations, which allows an estimate of the axial heat transfer and void distributions during steady-state and transient operations. The method of heat transfer data reduction is presented and the heat flux, secondary convective heat transfer coefficient, and void fraction distributions are quantified for steady-state, full-power operations

Coupled MCNP - SAS-SFR calculations for sodium fast reactor core at steady-state - 15460

International Nuclear Information System (INIS)

Ponomarev, A.; Travleev, A.; Pfrang, W.; Sanchez, V.

2015-01-01

The prediction of core parameters at steady state is the first step when studying core accident transient behaviour. At this step thermal hydraulics (TH) and core geometry parameters are calculated corresponding to initial operating conditions. In this study we present the coupling of the SAS-SFR code to the Monte-Carlo neutron transport code MCNP at steady state together with application to the European Sodium Fast Reactor (ESFR). The SAS-SFR code employs a multi-channel core representation where each channel represents subassemblies with similar power, thermal-hydraulics and pin mechanics conditions. For every axial node of every channel the individual geometry and material compositions parameters are calculated in accord with power and cooling conditions. This requires supplying the SAS-SFR-code with nodal power values which should be calculated by neutron physics code with given realistic core parameters. In the conventional approach the neutron physics model employs some core averaged TH and geometry data (fuel temperature, coolant density, core axial and radial expansion). In this study we organize a new approach coupling the MCNP neutron physics models and the SAS-SFR models, so that calculations of power can be improved by using distributed core parameters (TH and geometry) taken from SAS-SFR. The MCNP code is capable to describe cores with distributed TH parameters and even to model non-uniform axial expansion of fuel subassemblies. In this way, core TH and geometrical data calculated by SAS-SFR are taken into account accurately in the neutronics model. The coupling implementation is done by data exchange between two codes with help of processing routines managed by driver routine. Currently it is model-specific and realized for the ESFR 'Reference Oxide' core. The Beginning-Of-Life core state is considered with 10 channel representation for fuel subassemblies. For this model several sets of coupled calculations are performed, in which different

Design and manufacture of a D-shape coil-based toroid-type HTS DC reactor using 2nd generation HTS wire

Energy Technology Data Exchange (ETDEWEB)

Kim, Kwangmin, E-mail: kwangmin81@gmail.com [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of); Go, Byeong-Soo; Sung, Hae-Jin; Park, Hea-chul; Kim, Seokho [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of); Lee, Sangjin [Uiduk University, Gyeongju 780-713 (Korea, Republic of); Jin, Yoon-Su; Oh, Yunsang [Vector Fields Korea Inc., Pohang 790-834 (Korea, Republic of); Park, Minwon [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of); Yu, In-Keun, E-mail: yuik@changwon.ac.kr [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of)

2014-09-15

Highlights: • The authors designed and fabricated a D-shape coil based toroid-type HTS DC reactor using 2G GdBCO HTS wires. • The toroid-type magnet consisted of 30 D-shape double pancake coil (DDC)s. The total length of the wire was 2.32 km. • The conduction cooling method was adopted for reactor magnet cooling. • The maximum cooling temperature of reactor magnet is 5.5 K. • The inductance was 408 mH in the steady-state condition (300 A operating). - Abstract: This paper describes the design specifications and performance of a real toroid-type high temperature superconducting (HTS) DC reactor. The HTS DC reactor was designed using 2G HTS wires. The HTS coils of the toroid-type DC reactor magnet were made in the form of a D-shape. The target inductance of the HTS DC reactor was 400 mH. The expected operating temperature was under 20 K. The electromagnetic performance of the toroid-type HTS DC reactor magnet was analyzed using the finite element method program. A conduction cooling method was adopted for reactor magnet cooling. Performances of the toroid-type HTS DC reactor were analyzed through experiments conducted under the steady-state and charge conditions. The fundamental design specifications and the data obtained from this research will be applied to the design of a commercial-type HTS DC reactor.

Exploiting multicompartment effects in triple-echo steady-state T2 mapping for fat fraction quantification.

Science.gov (United States)

Liu, Dian; Steingoetter, Andreas; Curcic, Jelena; Kozerke, Sebastian

2018-01-01

To investigate and exploit the effect of intravoxel off-resonance compartments in the triple-echo steady-state (TESS) sequence without fat suppression for T 2 mapping and to leverage the results for fat fraction quantification. In multicompartment tissue, where at least one compartment is excited off-resonance, the total signal exhibits periodic modulations as a function of echo time (TE). Simulated multicompartment TESS signals were synthesized at various TEs. Fat emulsion phantoms were prepared and scanned at the same TE combinations using TESS. In vivo knee data were obtained with TESS to validate the simulations. The multicompartment effect was exploited for fat fraction quantification in the stomach by acquiring TESS signals at two TE combinations. Simulated and measured multicompartment signal intensities were in good agreement. Multicompartment effects caused erroneous T 2 offsets, even at low water-fat ratios. The choice of TE caused T 2 variations of as much as 28% in cartilage. The feasibility of fat fraction quantification to monitor the decrease of fat content in the stomach during digestion is demonstrated. Intravoxel off-resonance compartments are a confounding factor for T 2 quantification using TESS, causing errors that are dependent on the TE. At the same time, off-resonance effects may allow for efficient fat fraction mapping using steady-state imaging. Magn Reson Med 79:423-429, 2018. © 2017 International Society for Magnetic Resonance in Medicine. © 2017 International Society for Magnetic Resonance in Medicine.

The code DYN3DR for steady-state and transient analyses of light water reactor cores with Cartesian geometry

International Nuclear Information System (INIS)

Grundmann, U.

1995-11-01

The code DYN3D/M2 was developed for 3-dimensional steady-state and transient analyses of reactor cores with hexagonal fuel assemblies. The neutron kinetics of the new version DYN3DR is based on a nodal method for the solution of the 3-dimensional 2-group neutron diffusion equation for Cartesian geometry. The thermal-hydraulic model FLOCAL simulating the two phase flow of coolant and the fuel rod behaviour is used in the two versions. The fundamentals for the solution of the neutron diffusion equations in DYN3DR are described. The 3-dimensional NEACRP benchmarks for rod ejections in LWR with quadratic fuel assemblies were calculated and the results were compared with the published solutions. The developed algorithm for neutron kinetics are suitable for using parallel processing. The behaviour of speed-up versus the number of processors is demonstrated for calculations of a static neutron flux distribution using a workstation with 4 processors. (orig.) [de

Simulation of steady states of an integral PWR and power change transients using RELAP5 MOD3

International Nuclear Information System (INIS)

Aronne, Ivan Dionysio Aronne; Palmieri, Elcio Tadeu; Azwvedo, Carlos Vicente Goulart de; Baptista Filho, Benedito Dias; Barroso, Antonio Carlos de Oliveira

2005-01-01

An integral pressurized water reactor presents several differences in relation to conventional PWRs. The metal and cooling fluid masses of integral reactors are larger than those of a conventional reactor and, on the other hand, bombs tend to be smaller and the pressurizer should present characteristics proper of that arrangement. These characteristics, representing inertias different from the usual ones, makes obtaining the stationary state of the integral reactor a task with particularities that demand strategies different from the usually employed. This paper presents, initially, the main physical characteristics of the reactor in study and then the options adopted in developing the model and that were used to obtain the simulation of stationary states with the code RELAP5-MOD3. The results of the simulation of the steady state show the effects of the fore mentioned differences, where the times lags are significantly larger, as well as the suitability and efficiency of the defined approach. Two transients were simulated for changing the reactor power from steady state power of 100% to steady state power of 90%. The power change of these transients were one in step and the other in ramp with a rate of 5%/min. These calculations represent a first step for the definition and tests of parts of a preliminary control system for this reactor. The two transient simulated were based on plausible control hypotheses whose results are presented and commented. The final objective of this study is the use of results of simulations of steady states as much as of transients in support to the development of a transient identification and classification system, based on a neural network using self organizing maps whose basic proposition is presented in this paper. (author)

SYNTH-C, Steady-State and Time-Dependent 3-D Neutron Diffusion with Thermohydraulic Feedback

Energy Technology Data Exchange (ETDEWEB)

Brega, E [ENEL-CRTN, Bastioni di Porta Volta 10, Milan (Italy); Salina, E [A.R.S. Spa, Viale Maino 35, Milan (Italy)

1980-04-01

1 - Description of problem or function: SYNTH-C-STEADY and SYNTH-C- TRANS solve respectively the steady-state and time-dependent few- group neutron diffusion equations in three dimensions x,y,z in the presence of fuel temperature and thermal-hydraulic feedback. The neutron diffusion and delayed precursor equations are approximated by a space-time (z,t) synthesis method with axially discontinuous trial functions. Three thermal-hydraulic and fuel heat transfer models are available viz. COBRA-3C/MIT model, lumped parameter (WIGL) model and adiabatic fuel heat-up model. 2 - Method of solution: The steady-state and time-dependent synthesis equations are solved respectively by the Wielandt's power method and by the theta-difference method (in time), both coupled with a block factorization technique and double precision arithmetic. The thermal-hydraulic model equations are solved by fully implicit finite differences (WIGL) or explicit-implicit difference techniques with iterations (COBRA-EC/MIT). 3 - Restrictions on the complexity of the problem: Except for the few- group limitation, the programs have no other fixed limitation so the ability to run a problem depends only on the available computer storage.

Steady-state and accident analyses of PBMR with the computer code SPECTRA

International Nuclear Information System (INIS)

Stempniewicz, Marek M.

2002-01-01

The SPECTRA code is an accident analysis code developed at NRG. It is designed for thermal-hydraulic analyses of nuclear or conventional power plants. The code is capable of analysing the whole power plant, including reactor vessel, primary system, various control and safety systems, containment and reactor building. The aim of the work presented in this paper was to prepare a preliminary thermal-hydraulic model of PBMR for SPECTRA, and perform steady state and accident analyses. In order to assess SPECTRA capability to model the PBMR reactors, a model of the INCOGEN system has been prepared first. Steady state and accident scenarios were analyzed for INCOGEN configuration. Results were compared to the results obtained earlier with INAS and OCTOPUS/PANTHERMIX. A good agreement was obtained. Results of accident analyses with PBMR model showed qualitatively good results. It is concluded that SPECTRA is a suitable tool for analyzing High Temperature Reactors, such as INCOGEN or for example PBMR (Pebble Bed Modular Reactor). Analyses of INCOGEN and PBMR systems showed that in all analyzed cases the fuel temperatures remained within the acceptable limits. Consequently there is no danger of release of radioactivity to the environment. It may be concluded that those are promising designs for future safe industrial reactors. (author)

Reactor D and D at Argonne National Laboratory - lessons learned

International Nuclear Information System (INIS)

Fellhauer, C. R.

1998-01-01

This paper focuses on the lessons learned during the decontamination and decommissioning (D and D) of two reactors at Argonne National Laboratory-East (ANL-E). The Experimental Boiling Water Reactor (EBWR) was a 100 MW(t), 5 MSV(e) proof-of-concept facility. The Janus Reactor was a 200 kW(t) reactor located at the Biological Irradiation Facility and was used to study the effects of neutron radiation on animals

Computational multiple steady states for enzymatic esterification of ethanol and oleic acid in an isothermal CSTR.

Science.gov (United States)

Ho, Pang-Yen; Chuang, Guo-Syong; Chao, An-Chong; Li, Hsing-Ya

2005-05-01

The capacity of complex biochemical reaction networks (consisting of 11 coupled non-linear ordinary differential equations) to show multiple steady states, was investigated. The system involved esterification of ethanol and oleic acid by lipase in an isothermal continuous stirred tank reactor (CSTR). The Deficiency One Algorithm and the Subnetwork Analysis were applied to determine the steady state multiplicity. A set of rate constants and two corresponding steady states are computed. The phenomena of bistability, hysteresis and bifurcation are discussed. Moreover, the capacity of steady state multiplicity is extended to the family of the studied reaction networks.

Maximum Entropy Production Is Not a Steady State Attractor for 2D Fluid Convection

Directory of Open Access Journals (Sweden)

Stuart Bartlett

2016-12-01

Full Text Available Multiple authors have claimed that the natural convection of a fluid is a process that exhibits maximum entropy production (MEP. However, almost all such investigations were limited to fixed temperature boundary conditions (BCs. It was found that under those conditions, the system tends to maximize its heat flux, and hence it was concluded that the MEP state is a dynamical attractor. However, since entropy production varies with heat flux and difference of inverse temperature, it is essential that any complete investigation of entropy production allows for variations in heat flux and temperature difference. Only then can we legitimately assess whether the MEP state is the most attractive. Our previous work made use of negative feedback BCs to explore this possibility. We found that the steady state of the system was far from the MEP state. For any system, entropy production can only be maximized subject to a finite set of physical and material constraints. In the case of our previous work, it was possible that the adopted set of fluid parameters were constraining the system in such a way that it was entirely prevented from reaching the MEP state. Hence, in the present work, we used a different set of boundary parameters, such that the steady states of the system were in the local vicinity of the MEP state. If MEP was indeed an attractor, relaxing those constraints of our previous work should have caused a discrete perturbation to the surface of steady state heat flux values near the value corresponding to MEP. We found no such perturbation, and hence no discernible attraction to the MEP state. Furthermore, systems with fixed flux BCs actually minimize their entropy production (relative to the alternative stable state, that of pure diffusive heat transport. This leads us to conclude that the principle of MEP is not an accurate indicator of which stable steady state a convective system will adopt. However, for all BCs considered, the quotient of

Steady state and transient critical heat flux examinations

International Nuclear Information System (INIS)

Szabados, L.

1978-02-01

In steady state conditions within the P.W.R. parameter range the critical heat flux correlations based on local parameters reproduce the experimental data with less deviations than those based on system parameters. The transient experiments were restricted for the case of power transients. A data processing method for critical heat flux measurements has been developed and the applicability of quasi steady state calculation has been verified. (D.P.)

Comparison of preliminary D-T and ''catalyzed'' D-D system studies

International Nuclear Information System (INIS)

Usher, J.L.; Powell, J.R.; Fillo, J.A.; Lazareth, O.W.

1976-01-01

The purpose of the research currently underway is to provide technological and eventual economic comparison of a reference D-T reactor to a ''catalyzed'' D-D reactor. Two separate reactor designs are delineated and examined for this purpose. These systems include plasma parameters, blanket and shield configurations, magnetic coil configurations, and power conversion systems, including a divertor-direct convertor system for the D-D design. The initial conclusions reached are as follows: (a) no extraordinary requirements in the D-D reactor in the areas of blanket or magnet technology, (b) advantageous use of minimum activity blankets and shields, (c) increased overall efficiency via introduction of divertor-direct convertor subsystem in D-D design and (d) 65 percent increase in the toroidal radius of the D-D design compared to the D-T reference value

Steady State and Transient Fuel Rod Performance Analyses by Pad and Transuranus Codes

International Nuclear Information System (INIS)

Slyeptsov, O.; Slyeptsov, S.; Kulish, G.; Ostapov, A.; Chernov, I.

2013-01-01

The report performed under IAEA research contract No.15370/L2 describes the analysis results of WWER and PWR fuel rod performance at steady state operation and transients by means of PAD and TRANSURANUS codes. The code TRANSURANUS v1m1j09 developed by Institute for of Transuranium Elements (ITU) was used based on the Licensing Agreement N31302. The code PAD 4.0 developed by Westinghouse Electric Company was utilized in the frame of the Ukraine Nuclear Fuel Qualification Project for safety substantiation for the use of Westinghouse fuel assemblies in the mixed core of WWER-1000 reactor. The experimental data for the Russian fuel rod behavior obtained during the steady-state operation in the WWER-440 core of reactor Kola-3 and during the power transients in the core of MIR research reactor were taken from the IFPE database of the OECD/NEA and utilized for assessing the codes themselves during simulation of such properties as fuel burnup, fuel centerline temperature (FCT), fuel swelling, cladding strain, fission gas release (FGR) and rod internal pressure (RIP) in the rod burnup range of (41 - 60) GWD/MTU. The experimental data of fuel behavior at steady-state operation during seven reactor cycles presented by AREVA for the standard PWR fuel rod design were used to examine the code FGR model in the fuel burnup range of (37 - 81) GWD/MTU. (author)

Status of fusion reactor concept development in Japan

International Nuclear Information System (INIS)

Tsuji-Iio, Shunji

1996-01-01

Fusion power reactor studies in Japan based on magnetic confinement schemes are reviewed. As D-T fusion reactors, a steady-state tokamak reactor (SSTR) was proposed and extensively studied at the Japan Atomic Energy Research Institute (JAERI) and an inductively operated day-long tokamak reactor (IDLT) was proposed by a group at the University of Tokyo. The concept of a drastically easy maintenance (DREAM) tokamak reactor is being developed at JAERI. A high-field tokamak reactor with force-balanced coils as a volumetric neutron source is being studied by our group at Tokyo Institute of Technology. The conceptual design of a force-free helical reactor (FFHR) is under way at the National Institute for Fusion Science. A design study of a D- 3 He field-reversed configuration (FRC) fusion reactor called ARTEMIS was conducted by the FRC fusion working group of research committee of lunar base an lunar resources. (author)

Radioactivity computation of steady-state and pulsed fusion reactors operation

International Nuclear Information System (INIS)

Attaya, H.

1994-06-01

Different mathematical methods are used to calculate the nuclear transmutation in steady-state and pulsed neutron irradiation. These methods are the Schuer decomposition, the eigenvector decomposition, and the Pade approximation of the matrix exponential function. In the case of the linear decay chain approximation, a simple algorithm is used to evaluate the transition matrices

HEATHYD, Steady-State Thermal Hydraulic Analysis of Low-Enriched U Fuel Reactor

International Nuclear Information System (INIS)

NABBI, R.

1989-01-01

1 - Description of program or function: HEATHYD is a code for the steady-state heat transfer calculation of research nuclear reactors with forced convection. It models heat transfer and coolant flow for assemblies of parallel fuel plates of MTR type with any axial power distribution. The thermodynamic model accounts for single phase cooling and sub- cooled boiling condition using the transition criterion of Bergeles-Rosenow. In addition to the calculation of the channel flow velocities and coolant pressure drops, HEATHYD calculates axial distribution of the coolant and clad-surface temperatures. Safety margins to the critical heat flux as a result of burnout condition or flow instability are determined. 2 - Method of solution: Applying the finite difference method, HEATHYD solves the equations of heat conduction and heat transfer to the coolant. For the physical properties of the coolant as a function of the coolant temperature polynomials of degree 6 are used. Depending on the coolant condition, different correlations for the heat transfer coefficient can be applied. The analysis of the critical cooling conditions resulting in burnout or flow instability, is performed according to the correlations developed by Mirshak/ Labuntsov and Forgan/Whittle

Cranial nerve assessment in cavernous sinus tumors with contrast-enhanced 3D fast-imaging employing steady-state acquisition MR imaging

Energy Technology Data Exchange (ETDEWEB)

Amemiya, Shiori; Aoki, Shigeki; Ohtomo, Kuni [University of Tokyo, Department of Radiology, Graduate School of Medicine, Bunkyo-ku, Tokyo (Japan)

2009-07-15

The purpose of this study is to apply contrast-enhanced 3D fast-imaging employing steady-state acquisition (3D-FIESTA) imaging to the evaluation of cranial nerves (CN) in patients with cavernous sinus tumors. Contrast-enhanced 3D-FIESTA images were acquired from ten patients with cavernous sinus tumors with a 3-T unit. In all cases, the trigeminal nerve with tumor involvement was easily identified in the cavernous portions. Although oculomotor and abducens nerves were clearly visualized against the tumor area with intense contrast enhancement, they were hardly identifiable within the area lacking contrast enhancement. The trochlear nerve was visualized in part, but not delineated as a linear structure outside of the lesion. Contrast-enhanced 3D-FIESTA can be useful in the assessment of cranial nerves in and around the cavernous sinus with tumor involvement. (orig.)

Cranial nerve assessment in cavernous sinus tumors with contrast-enhanced 3D fast-imaging employing steady-state acquisition MR imaging

International Nuclear Information System (INIS)

Amemiya, Shiori; Aoki, Shigeki; Ohtomo, Kuni

2009-01-01

The purpose of this study is to apply contrast-enhanced 3D fast-imaging employing steady-state acquisition (3D-FIESTA) imaging to the evaluation of cranial nerves (CN) in patients with cavernous sinus tumors. Contrast-enhanced 3D-FIESTA images were acquired from ten patients with cavernous sinus tumors with a 3-T unit. In all cases, the trigeminal nerve with tumor involvement was easily identified in the cavernous portions. Although oculomotor and abducens nerves were clearly visualized against the tumor area with intense contrast enhancement, they were hardly identifiable within the area lacking contrast enhancement. The trochlear nerve was visualized in part, but not delineated as a linear structure outside of the lesion. Contrast-enhanced 3D-FIESTA can be useful in the assessment of cranial nerves in and around the cavernous sinus with tumor involvement. (orig.)

LAVENDER: A steady-state core analysis code for design studies of accelerator driven subcritical reactors

Energy Technology Data Exchange (ETDEWEB)

Zhou, Shengcheng; Wu, Hongchun; Cao, Liangzhi; Zheng, Youqi, E-mail: yqzheng@mail.xjtu.edu.cn; Huang, Kai; He, Mingtao; Li, Xunzhao

2014-10-15

Highlights: • A new code system for design studies of accelerator driven subcritical reactors (ADSRs) is developed. • S{sub N} transport solver in triangular-z meshes, fine deletion analysis and multi-channel thermal-hydraulics analysis are coupled in the code. • Numerical results indicate that the code is reliable and efficient for design studies of ADSRs. - Abstract: Accelerator driven subcritical reactors (ADSRs) have been proposed and widely investigated for the transmutation of transuranics (TRUs). ADSRs have several special characteristics, such as the subcritical core driven by spallation neutrons, anisotropic neutron flux distribution and complex geometry etc. These bring up requirements for development or extension of analysis codes to perform design studies. A code system named LAVENDER has been developed in this paper. It couples the modules for spallation target simulation and subcritical core analysis. The neutron transport-depletion calculation scheme is used based on the homogenized cross section from assembly calculations. A three-dimensional S{sub N} nodal transport code based on triangular-z meshes is employed and a multi-channel thermal-hydraulics analysis model is integrated. In the depletion calculation, the evolution of isotopic composition in the core is evaluated using the transmutation trajectory analysis algorithm (TTA) and fine depletion chains. The new code is verified by several benchmarks and code-to-code comparisons. Numerical results indicate that LAVENDER is reliable and efficient to be applied for the steady-state analysis and reactor core design of ADSRs.

DNA Polymerase Fidelity: Comparing Direct Competition of Right and Wrong dNTP Substrates with Steady State and Presteady State Kinetics†

Science.gov (United States)

Bertram, Jeffrey G.; Oertell, Keriann; Petruska, John; Goodman, Myron F.

2009-01-01

DNA polymerase fidelity is defined as the ratio of right (R) to wrong (W) nucleotide incorporations when dRTP and dWTP substrates compete at equal concentrations for primer extension at the same site in the polymerase-primer-template DNA complex. Typically, R incorporation is favored over W by 103 – 105, even in the absence of 3′-exonuclease proofreading. Straightforward in principal, a direct competition fidelity measurement is difficult to perform in practice because detection of a small amount of W is masked by a large amount of R. As an alternative, enzyme kinetics measurements to evaluate kcat/Km for R and W in separate reactions are widely used to measure polymerase fidelity indirectly, based on a steady-state derivation by Fersht. A systematic comparison between direct competition and kinetics has not been made until now. By separating R and W products using electrophoresis, we have successfully made accurate fidelity measurements for directly competing R and W dNTP substrates for 9 of the 12 natural base mispairs. We compare our direct competition results with steady state and presteady state kinetic measurements of fidelity at the same template site, using the proofreading-deficient mutant of Klenow Fragment (KF−) DNA polymerase. All the data are in quantitative agreement. PMID:20000359

Engineering aspects of a D-D commercial tokamak reactor

International Nuclear Information System (INIS)

Evans, K. Jr.; Baker, C.C.; Brooks, J.N.

1981-01-01

This paper presents some of the engineering aspects of WILDCAT, a conceptual design of a D-D tokamak, fusion reactor. This conceptual design has evolved from initial studies of D-D tokamak reactors, and is intended to be a study of a later-model, commerical fusion reactor in the same sense that STARFIRE was such a study for D-T fuel cycle. The major guidelines of the study have been to utilize as fully as possible the advantages of the D-D fuel cycle but to avoid unnecessary extrapolations of parameters from existing D-T designs, in particular STARFIRE. The paper consists of an overview of the reference design, a description of each of the major engineering systems (rf current drive, burn cycle, impurity control, first wall, blanket/shield, TF magnets, and tritium system, and a summary of conclusions)

Homogeneity of Continuum Model of an Unsteady State Fixed Bed Reactor for Lean CH4 Oxidation

Directory of Open Access Journals (Sweden)

Subagjo

2014-07-01

Full Text Available In this study, the homogeneity of the continuum model of a fixed bed reactor operated in steady state and unsteady state systems for lean CH4 oxidation is investigated. The steady-state fixed bed reactor system was operated under once-through direction, while the unsteady-state fixed bed reactor system was operated under flow reversal. The governing equations consisting of mass and energy balances were solved using the FlexPDE software package, version 6. The model selection is indispensable for an effective calculation since the simulation of a reverse flow reactor is time-consuming. The homogeneous and heterogeneous models for steady state operation gave similar conversions and temperature profiles, with a deviation of 0.12 to 0.14%. For reverse flow operation, the deviations of the continuum models of thepseudo-homogeneous and heterogeneous models were in the range of 25-65%. It is suggested that pseudo-homogeneous models can be applied to steady state systems, whereas heterogeneous models have to be applied to unsteady state systems.

TRIO a general computer code for reactor 3-D flows analysis. Application to a LMFBR hot plenum

International Nuclear Information System (INIS)

Magnaud, J.P.; Rouzaud, P.

1985-09-01

TRIO is a code developed at CEA to investigate general incompressible 2D and 3D viscous flows. Two calculations are presented: the lid driven cubic cavity at Re=400; steady state (velocity and temperature field) of a LMFBR hot plenum, carried out in order to prepare the calculation of a cold shock consecutive to a reactor scram. 8 refs., 26 figs.

Cryogenic distillation: a fuel enrichment system for near-term tokamak-type D-T fusion reactors

International Nuclear Information System (INIS)

Misra, B.; Davis, J.F.

1980-02-01

The successful operation and economic viability of deuterium-tritium- (D-T-) fueled tokamak-type commercial power fusion reactors will depend to a large extent on the development of reliable tritium-containment and fuel-recycle systems. Of the many operating steps in the fuel recycle scheme, separation or enrichment of the isotropic species of hydrogen by cryogenic distillation is one of the most important. A parametric investigation was carried out to study the effects of the various operating conditions and the composition of the spent fuel on the degree of separation. A computer program was developed for the design and analysis of a system of interconnected distillation columns for isotopic separation such that the requirements of near-term D-T-fueled reactors are met. The analytical results show that a distillation cascade consisting of four columns is capable of reprocessing spent fuel varying over a wide range of compositions to yield reinjection-grade fuel with essentially unlimited D/T ratio

Preoperative mapping of arterial spinal supply using 3.0-T MR angiography with an intravasal contrast medium and high-spatial-resolution steady-state

International Nuclear Information System (INIS)

Mordasini, Pasquale; El-Koussy, Marwan; Schmidli, Jürg; Bonel, Harald Marcel; Ith, Michael; Gralla, Jan; Schroth, Gerhard; Hoppe, Hanno

2012-01-01

Introduction: Preoperative mapping of the arterial spinal supply prior to thoracoabdominal aortic aneurysm repair is highly relevant because of high risk for postoperative ischemic spinal cord injuries such as paraparesis or paraplegia. Methods: Twenty-four consecutive patients prior to surgical thoracoabdominal aortic aneurysm repair were investigated. All patients underwent steady-state MR angiography (MRA) of the spinal vasculature with 3-T MRI. The sequence used was a steady-state coronary 3D FLASH with 0.7-mm isotropic voxels. MRA was performed using an intravasal contrast agent. Studies were evaluated by three readers including delineation of arterial spinal supply including both aortic origin and spinal canal entry by three readers. Results: Identification and localization of the Adamkiewicz artery and its spinal canal entry was successful in all patients. Overall depiction of the vascular anatomy was graded as very good in 3 (12.5%), good in 14 (58.4%), sufficient in 5 (20.8%), and poor in 2 (8.3%) patients. Depiction of segmental artery aortic exit level was graded as good in 6 (25.0%), sufficient in 10 (41.7%), poor in 4 (16.7%) and not identifiable in 4 (16.7%) patients. Delineation of segmental artery entry level into the spinal canal was graded as very good in 4 (16.7%), good in 11 (45.8%), sufficient in 6 (25.0%), and poor in 3 (12.5%) patients. Conclusions: The use of 3-T MRA with an intravascular contrast agent and steady-state enables AKA localization including its segmental arteries with regard to the level of aortic origin and spinal canal entry in most patients.

Steady-state heat transfer in an inverted U-tube steam generator

International Nuclear Information System (INIS)

Boucher, T.J.

1987-01-01

Experimental results are presented involving U-tube steam generator tube bundle local heat transfer and fluid conditions during stead-state, full-power operations performed at high temperatures and pressures with conditions typical of a pressurized water reactor (15.0 MPa primary pressure, 600 K steam generator inlet plenum fluid temperatures, 6.2 MPa secondary pressure). The Semiscale (MOD-2C facility represents the state-of-the-art in measurement of tube local heat transfer data and average tube bundle secondary fluid density at several elevations, which allows an estimate of the axial heat transfer and void distributions during steady-state and transient operations. The method of heat transfer data reduction is presented and the heat flux, secondary convective heat transfer coefficient, and void fraction distributions are quantified for steady-state, full-power operations

Modelling and thermal hydraulic analysis of the Angra-2 nuclear reactor using RELAP5-3D code

International Nuclear Information System (INIS)

González Mantecón, Javier

2015-01-01

The evaluation of Nuclear Power Plants (NPPs) performance during steady-state and accident conditions has been one of the main research subjects in the nuclear field. In order to simulate the behavior of water-cooled reactors, several complex thermal-hydraulic codes systems have been developed. Particularly, the RELAP5 code, developed by the Idaho National Laboratory, is a best-estimate thermal-hydraulic analysis tool and one of the most used in nuclear industry. The RELAP5-3D 3.0.0 code was used to develop a detailed model of Angra 2 nuclear reactor using reference data from the Final Safety Analysis Report. Angra 2 is the second Brazilian NPP, which began commercial operation in 2001. The plant is equipped with a Pressurized Water Reactor (PWR) type with 3771.0 MWt. Simulations of the reactor behavior during normal operation conditions and postulated accident conditions were performed. Results achieved in the reactor steady-state simulation were compared with nominal parameters of the NPP. These results proved to be in good agreement, with relative errors less than 1%. In the transient simulation, the obtained results were coherent and satisfactory. This study demonstrates that the RELAP5-3D model is capable to reproduce the thermal-hydraulic behavior of the Angra-2 PWR during diverse operation conditions and it can contribute for the process of the plant safety analysis. (author)

TFTR D-T results

International Nuclear Information System (INIS)

Meade, D.M.

1994-01-01

Temperatures, densities and confinement of deuterium plasmas confined in tokamaks have been achieved within the last decade that are approaching those required for a D-T reactor. As a result, the unique phenomena present in a D-T reactor plasma (D-T plasma confinement, alpha confinement, alpha heating and possible alpha driven instabilities) can now be studied in the laboratory. Recent experiments on the Tokamak Fusion Test Reactor (TFTR) have been the first magnetic fusion experiments to study plasmas with reactor fuel concentrations of tritium. The injection of ∼ 20 MW of tritium and 14 MW of deuterium neutral beams into the TFTR produced a plasma with a T/D density ratio of ∼1 and yielded a maximum fusion power of ∼ 9.2 MW. The fusion power density in the core of the plasma was ∼ 1.8 MW m -3 approximating that expected in a D-T fusion reactor. A TFTR plasma with T/D density ratio of ∼ 1 was found to have ∼ 20% higher energy confinement time than a comparable D plasma, indicating a confinement scaling with average ion mass, A, of τ E ∼ A 0.6 . The core ion temperature increased from 30 keV to 37 keV due to a 35% improvement of ion thermal conductivity. Using the electron thermal conductivity from a comparable deuterium plasma, about 50% of the electron temperature increase from 9 keV to 10.6 keV can be attributed to electron heating by the alpha particles. The ∼ 5% loss of alpha particles, as observed on detectors near the bottom edge of the plasma, was consistent with classical first orbit loss without anomalous effects. Initial measurements have been made of the confined energetic alphas and the resultant alpha ash density. At fusion power levels of 7.5 MW, fluctuations at the Toroidal Alfven Eigenmode frequency were observed by the fluctuation diagnostics. However, no additional alpha loss due to the fluctuations was observed

Hybrid Reactor designs in the United States

International Nuclear Information System (INIS)

Wolkenhauer, W.C.

1978-01-01

This paper reviews the current, active, interrelated Hybrid Reactor development programs in the United States, and offers a probable future course of action for the technology. The Department of Energy (DOE) program primarily emphasizes development of Hybrid Reactors that are optimized for proliferation resistance. The Electric Power Research Institute (EPRI) program concentrates on avenues for Hybrid Reactor commercialization. The history of electrical generation technology has been one of steady movement toward higher power densities and higher quality fuels. An apparent advantage of the Hybrid Reactor option is that it follows this trend

MyD88 Signaling Regulates Steady-State Migration of Intestinal CD103+ Dendritic Cells Independently of TNF-α and the Gut Microbiota

DEFF Research Database (Denmark)

Hägerbrand, Karin; Westlund, Jessica; Yrlid, Ulf

2015-01-01

Intestinal homeostasis and induction of systemic tolerance to fed Ags (i.e., oral tolerance) rely on the steady-state migration of smallintestinal lamina propria dendritic cells (DCs) into draining mesenteric lymph nodes (MLN). The majority of these migratory DCs express the α integrin chain CD103......, and in this study we demonstrate that the steady-state mobilization of CD103+ DCs into the MLN is in part governed by the IL-1R family/TLR signaling adaptor molecule MyD88. Similar to mice with complete MyD88 deficiency, specific deletion of MyD88 in DCs resulted in a 50–60% reduction in short-term accumulation......, the proportion and phenotypic composition of DCs were similar in mesentericlymph from germ-free and conventionally housed mice. Although TNF-α was required for CD103+ DC migration to the MLN after oral administration of the TLR7 agonist R848, it was not required for the steady-state migration of these cells...

Design and cost evaluation of generic magnetic fusion reactor using the D-D fuel cycle

International Nuclear Information System (INIS)

Shannon, T.E.

1988-01-01

A fusion reactor systems code has been developed to evaluate the economic potential of power generation from a toroidal magnetic fusion reactor using deuterium-deuterium (D-D) fuel. A method similar to that developed by J. Sheffield, of the Oak Ridge National Laboratory, for deuterium-tritium (D-T) fuel was used to model the generic aspects of magnetic fusion reactors. The results of the systems study and cost evaluation show that the cost of electricity produced by a D-D reactor is two times higher than that produced by an equivalent D-T reactor design. The significant finding of the study is that the cost ratio between the D-D and D-T systems can potentially be reduced to 1.5 by improved engineering design and even lower by better physics performance. The absolute costs for both systems at this level are close to the costs for nuclear fission and fossil fuel plants. A design for a magnet reinforced with advanced composite materials is presented as an example of an engineering improvement that could reduce the cost of electricity produced by both reactors. However, since the magnets in the D-D reactor are much larger than in the K-T reactor, the cost ratio of the two systems is significantly reduced

Standard mirror fusion reactor design study

International Nuclear Information System (INIS)

Moir, R.W.

1978-01-01

This report covers the work of the Magnetic Fusion Energy Division's reactor study group during FY 1976 on the standard mirror reactor. The ''standard'' mirror reactor is characterized as a steady state, neutral beam sustained, D-T fusioning plasma confined by a Yin-Yang magnetic mirror field. The physics parameters are obtained from the same physics model that explains the 2XIIB experiment. The model assumes that the drift cyclotron loss cone mode occurs on the boundary of the plasma, and that it is stabilized by warm plasma with negligible energy investment. The result of the study was a workable mirror fusion power plant, steady-state blanket removal made relatively simple by open-ended geometry, and no impurity problem due to the positive plasma potential. The Q (fusion power/injected beam power) turns out to be only 1.1 because of loss out the ends from Coulomb collisions, i.e., classical losses. This low Q resulted in 77% of the gross electrical power being used to power the injectors, thereby causing the net power cost to be high. The low Q stimulated an intensive search for Q-enhancement concepts, resulting in the LLL reactor design effort turning to the field reversal mirror and the tandem mirror, each having Q of order 5

IBIS, FBR 3-D Steady-State and Kinetics with Thermohydraulic Feedback

International Nuclear Information System (INIS)

Konomura, Mamoru; Tada, Nobuo; Oka, Yoshiaki; An, Shigehiro

1987-01-01

1 - Description of program or function: The IBIS code performs steady state and kinetics calculations based on a three-dimensional nuclear diffusion kinetics with thermal hydraulic feedback. It can calculate the following values in hexagonal-Z geometry of a fast breeder reactor core through the progress of transient: (1) Net reactivity; (2) Total and group-wise delayed neutron fraction; (3) Group-wise delayed neutron precursor concentration; (4) Total power and energy; (5) Space dependent neutron flux in each energy group; (6) Space dependent temperature of each material; (7) Maximum temperature of each material and its location. 2 - Method of solution: The quasi-static method is adopted to solve the three-dimensional nuclear diffusion kinetics problem. The method is the same as employed in the code QX1. The shape function equation is solved with the finite difference treatment as used in the codes CITATION and HONEYCOMB. One-dimensional thermo-hydraulics is solved with a model similar to that given in the code SASLA. Sodium boiling can be taken into account. 3 - Restrictions on the complexity of the problem: The number of neutron energy groups is fixed to 3 groups in the present version of the code

A comparison of multi-echo spin-echo and triple-echo steady-state T2 mapping for in vivo evaluation of articular cartilage.

Science.gov (United States)

Juras, Vladimir; Bohndorf, Klaus; Heule, Rahel; Kronnerwetter, Claudia; Szomolanyi, Pavol; Hager, Benedikt; Bieri, Oliver; Zbyn, Stefan; Trattnig, Siegfried

2016-06-01

To assess the clinical relevance of T2 relaxation times, measured by 3D triple-echo steady-state (3D-TESS), in knee articular cartilage compared to conventional multi-echo spin-echo T2-mapping. Thirteen volunteers and ten patients with focal cartilage lesions were included in this prospective study. All subjects underwent 3-Tesla MRI consisting of a multi-echo multi-slice spin-echo sequence (CPMG) as a reference method for T2 mapping, and 3D TESS with the same geometry settings, but variable acquisition times: standard (TESSs 4:35min) and quick (TESSq 2:05min). T2 values were compared in six different regions in the femoral and tibial cartilage using a Wilcoxon signed ranks test and the Pearson correlation coefficient (r). The local ethics committee approved this study, and all participants gave written informed consent. The mean quantitative T2 values measured by CPMG (mean: 46±9ms) in volunteers were significantly higher compared to those measured with TESS (mean: 31±5ms) in all regions. Both methods performed similarly in patients, but CPMG provided a slightly higher difference between lesions and native cartilage (CPMG: 90ms→61ms [31%],p=0.0125;TESS 32ms→24ms [24%],p=0.0839). 3D-TESS provides results similar to those of a conventional multi-echo spin-echo sequence with many benefits, such as shortening of total acquisition time and insensitivity to B1 and B0 changes. • 3D-TESS T 2 mapping provides clinically comparable results to CPMG in shorter scan-time. • Clinical and investigational studies may benefit from high temporal resolution of 3D-TESS. • 3D-TESS T 2 values are able to differentiate between healthy and damaged cartilage.

High-β steady-state advanced tokamak regimes for ITER and FIRE

International Nuclear Information System (INIS)

Meade, D.M.; Sauthoff, N.R.; Kessel, C.E.; Budny, R.V.; Gorelenkov, N.; Jardin, S.C.; Schmidt, J.A.; Navratil, G.A.; Bialek, J.; Ulrickson, M.A.; Rognlein, T.; Mandrekas, J.

2005-01-01

An attractive tokamak-based fusion power plant will require the development of high-β steady-state advanced tokamak regimes to produce a high-gain burning plasma with a large fraction of self-driven current and high fusion-power density. Both ITER and FIRE are being designed with the objective to address these issues by exploring and understanding burning plasma physics both in the conventional H-mode regime, and in advanced tokamak regimes with β N ∼ 3 - 4, and f bs ∼50-80%. ITER has employed conservative scenarios, as appropriate for its nuclear technology mission, while FIRE has employed more aggressive assumptions aimed at exploring the scenarios envisioned in the ARIES power-plant studies. The main characteristics of the advanced scenarios presently under study for ITER and FIRE are compared with advanced tokamak regimes envisioned for the European Power Plant Conceptual Study (PPCS-C), the US ARIES-RS Power Plant Study and the Japanese Advanced Steady-State Tokamak Reactor (ASSTR). The goal of the present work is to develop advanced tokamak scenarios that would fully exploit the capability of ITER and FIRE. This paper will summarize the status of the work and indicate critical areas where further R and D is needed. (author)

A comparison of multi-echo spin-echo and triple-echo steady-state T2 mapping for in vivo evaluation of articular cartilage

International Nuclear Information System (INIS)

Juras, Vladimir; Szomolanyi, Pavol; Bohndorf, Klaus; Kronnerwetter, Claudia; Hager, Benedikt; Zbyn, Stefan; Heule, Rahel; Bieri, Oliver; Trattnig, Siegfried

2016-01-01

To assess the clinical relevance of T 2 relaxation times, measured by 3D triple-echo steady-state (3D-TESS), in knee articular cartilage compared to conventional multi-echo spin-echo T 2 -mapping. Thirteen volunteers and ten patients with focal cartilage lesions were included in this prospective study. All subjects underwent 3-Tesla MRI consisting of a multi-echo multi-slice spin-echo sequence (CPMG) as a reference method for T 2 mapping, and 3D TESS with the same geometry settings, but variable acquisition times: standard (TESSs 4:35min) and quick (TESSq 2:05min). T 2 values were compared in six different regions in the femoral and tibial cartilage using a Wilcoxon signed ranks test and the Pearson correlation coefficient (r). The local ethics committee approved this study, and all participants gave written informed consent. The mean quantitative T 2 values measured by CPMG (mean: 46±9ms) in volunteers were significantly higher compared to those measured with TESS (mean: 31±5ms) in all regions. Both methods performed similarly in patients, but CPMG provided a slightly higher difference between lesions and native cartilage (CPMG: 90ms→61ms [31%],p=0.0125;TESS 32ms→24ms [24%],p=0.0839). 3D-TESS provides results similar to those of a conventional multi-echo spin-echo sequence with many benefits, such as shortening of total acquisition time and insensitivity to B 1 and B 0 changes. (orig.)

Conceptual design of D-3He FRC reactor 'ARTEMIS'

International Nuclear Information System (INIS)

Momota, H.; Ishida, A.; Kohzaki, Y.

1991-07-01

A comprehensive design study of the D- 3 He fueled field-reversed configuration (FRC) reactor 'ARTEMIS' is carried out for the purpose of proving its attractive characteristics and clarifying the critical issues for a commercial fusion reactor. The FRC burning plasma is stabilized and sustained in a steady equilibrium by means of a preferential trapping of D- 3 He fusion-produced energetic protons. A novel direct energy converter for 15MeV protons is also presented. On the bases of a consistent scenario of the fusion plasma production and simple engineering, a compact and simple reactor concept is presented. The design of the D- 3 He FRC power plant definitely offers the most attractive prospect for energy development. It is environmentally acceptable in view of radio-activity and fuel resources; and the estimated cost of electricity is low compared to a light water reactor. Critical issues concerning physics or engineering for the development of the D- 3 He FRC reactor are clarified. (author)

40 CFR Appendix II to Part 1039 - Steady-State Duty Cycles

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 32 2010-07-01 2010-07-01 false Steady-State Duty Cycles II Appendix... Appendix II to Part 1039—Steady-State Duty Cycles (a) The following duty cycles apply for constant-speed engines: (1) The following duty cycle applies for discrete-mode testing: D2 mode number Engine speed...

Steady-state and transient fission gas release and swelling model for LIFE-4

International Nuclear Information System (INIS)

Villalobos, A.; Liu, Y.Y.; Rest, J.

1984-06-01

The fuel-pin modeling code LIFE-4 and the mechanistic fission gas behavior model FASTGRASS have been coupled and verified against gas release data from mixed-oxide fuels which were transient tested in the TREAT reactor. Design of the interface between LIFE-4 and FASTGRASS is based on an earlier coupling between an LWR version of LIFE and the GRASS-SST code. Fission gas behavior can significantly affect steady-state and transient fuel performance. FASTGRASS treats fission gas release and swelling in an internally consistent manner and simultaneously includes all major mechanisms thought to influence fission gas behavior. The FASTGRASS steady-state and transient analysis has evolved through comparisons of code predictions with fission-gas release and swelling data from both in- and ex-reactor experiments. FASTGRASS was chosen over other fission-gas behavior models because of its availability, its compatibility with the LIFE-4 calculational framework, and its predictive capability

Prokaryotic diversity and dynamics in a full-scale municipal solid waste anaerobic reactor from start-up to steady-state conditions.

Science.gov (United States)

Cardinali-Rezende, Juliana; Colturato, Luís F D B; Colturato, Thiago D B; Chartone-Souza, Edmar; Nascimento, Andréa M A; Sanz, José L

2012-09-01

The prokaryotic diversity of an anaerobic reactor for the treatment of municipal solid waste was investigated over the course of 2 years with the use of 16S rDNA-targeted molecular approaches. The fermentative Bacteroidetes and Firmicutes predominated, and Proteobacteria, Actinobacteria, Tenericutes and the candidate division WWE1 were also identified. Methane production was dominated by the hydrogenotrophic Methanomicrobiales (Methanoculleus sp.) and their syntrophic association with acetate-utilizing and propionate-oxidizing bacteria. qPCR demonstrated the predominance of the hydrogenotrophic over aceticlastic Methanosarcinaceae (Methanosarcina sp. and Methanimicrococcus sp.), and Methanosaetaceae (Methanosaeta sp.) were measured in low numbers in the reactor. According to the FISH and CARD-FISH analyses, Bacteria and Archaea accounted for 85% and 15% of the cells, respectively. Different cell counts for these domains were obtained by qPCR versus FISH analyses. The use of several molecular tools increases our knowledge of the prokaryotic community dynamics from start-up to steady-state conditions in a full-scale MSW reactor. Copyright © 2012 Elsevier Ltd. All rights reserved.

KIM, Steady-State Transport for Fixed Source in 2-D Thermal Reactor by Monte-Carlo

International Nuclear Information System (INIS)

Cupini, E.; De Matteis, A.; Simonini, R.

1980-01-01

1 - Description of problem or function: KIM (K-infinite Monte Carlo) is a program which solves the steady-state linear transport equation for a fixed-source problem (or, by successive fixed-source runs, for the eigenvalue problem) in a two-dimensional infinite thermal reactor lattice. The main quantities computed in some broad energy groups are the following: - Fluxes and cross sections averaged over the region (i.e. a space portion that can be unconnected but contains everywhere the same homogeneous material), grouping of regions, the whole element. - Average absorption and fission rates per nuclide. - Average flux, absorption and production distributions versus energy. 2 - Method of solution: Monte Carlo simulation is used by tracing particle histories from fission birth down through the resonance region until absorption in the thermal range. The program is organised in three sections for fast, epithermal and thermal simulation, respectively; each section implements a particular model for both numerical techniques and cross section representation (energy groups in the fast section, groups or resonance parameters in the epithermal section, points in the thermal section). During slowing down (energy above 1 eV) nuclei are considered as stationary, with the exception of some resonance nuclei whose spacing between resonances is much greater than the resonance width. The Doppler broadening of s-wave resonances of these nuclides is taken into account by computing cross sections at the current neutron energy and at the temperature of the nucleus hit. During thermalization (energy below 1 eV) the thermal motion of some nuclides is also considered, by exploiting scattering kernels provided by the library for light water, heavy water and oxygen at several temperatures. KIM includes splitting and Russian roulette. A characteristic feature of the program is its approach to the lattice geometry. In fact, besides the usual continuous treatment of the geometry using the well

The behaviour of impurities in a steady-state DT gas-blanket reactor

International Nuclear Information System (INIS)

Markvoort, J.A.

1975-11-01

A four-fluid model of a cylindrical steady-state DT gas-blanket reactor is analysed. The four fluids are electrons, deuterium-tritium, helium and a high -Z impurity. The behaviour of the plasma is described by the multifluid MHD-equations which are numerically solved with the aid of a Runge Kutta method. Whether impurities tend to concentrate on the axis is found to depend on how, in the collision term, the Nernst effect is taken into account. In order to show the influence of the Nernst terms arising from electron-ion collisions and the Nernst terms due to ion-ion collisions separately, the thermal force is dealt with in two ways. In model A, only the contribution from electron-ion collisions was considered. The computer calculations show that the impurities have their maximum concentration on the axis. A theoretical analysis explains this result. In model B, which is more realistic, these ion-ion collisions are included. The computer calculations as well as the theoretical analysis show that the influence of the thermoforce due to ion-ion collisions on the density profiles dominates over the force due to electron collisions, and lead to a minimum in the impurity density on the axis. As in model A, the analytical analysis yields relationships between the various density profiles and the temperature profile

Research on steady-state visual evoked potentials in 3D displays

Science.gov (United States)

Chien, Yu-Yi; Lee, Chia-Ying; Lin, Fang-Cheng; Huang, Yi-Pai; Ko, Li-Wei; Shieh, Han-Ping D.

2015-05-01

Brain-computer interfaces (BCIs) are intuitive systems for users to communicate with outer electronic devices. Steady state visual evoked potential (SSVEP) is one of the common inputs for BCI systems due to its easy detection and high information transfer rates. An advanced interactive platform integrated with liquid crystal displays is leading a trend to provide an alternative option not only for the handicapped but also for the public to make our lives more convenient. Many SSVEP-based BCI systems have been studied in a 2D environment; however there is only little literature about SSVEP-based BCI systems using 3D stimuli. 3D displays have potentials in SSVEP-based BCI systems because they can offer vivid images, good quality in presentation, various stimuli and more entertainment. The purpose of this study was to investigate the effect of two important 3D factors (disparity and crosstalk) on SSVEPs. Twelve participants participated in the experiment with a patterned retarder 3D display. The results show that there is a significant difference (p-value<0.05) between large and small disparity angle, and the signal-to-noise ratios (SNRs) of small disparity angles is higher than those of large disparity angles. The 3D stimuli with smaller disparity and lower crosstalk are more suitable for applications based on the results of 3D perception and SSVEP responses (SNR). Furthermore, we can infer the 3D perception of users by SSVEP responses, and modify the proper disparity of 3D images automatically in the future.

Thermal hydraulic core simulation of the MYRRHA Reactor in steady state operation

International Nuclear Information System (INIS)

Ferandes, Gustavo H.N.; Ramos, Mário C.; Carvalho, Athos M.S.S.; Cabrera, Carlos E.V.; Costa, Antonella L.; Pereira, Claubia

2017-01-01

MYRRHA (Multi-purpose Hybrid Research Reactor for High-tech Applications) is a prototype nuclear subcritical reactor driven by a particle accelerator. As a special property, the reactor maintains the nuclear fission chain reaction by means of an external neutron source provided by a particle accelerator. The main aim of this work is to study two types of coolants, LBE (Lead-Bismuth Eutectic) and Na (Sodium) that are two strong candidates to be used in ADS systems as well as in Generation IV (GEN-IV) reactors. Firstly, it was developed a thermal hydraulic model of the MYRRHA core using the RELAP5-3D, considering LBE as coolant (original project). After this, the LBE was substituted by Na coolant to investigate the reactor behavior in such case. Results have demonstrated the high heat transfer capacity of the LBE coolant in this type of system. (author)

Thermal hydraulic core simulation of the MYRRHA Reactor in steady state operation

Energy Technology Data Exchange (ETDEWEB)

Ferandes, Gustavo H.N.; Ramos, Mário C.; Carvalho, Athos M.S.S.; Cabrera, Carlos E.V.; Costa, Antonella L.; Pereira, Claubia, E-mail: ghnfernandes@gmail.com, E-mail: marc5663@gmail.com, E-mail: athos1495@yahoo.com.br, E-mail: carlosvelcab@hotmail.com, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciência e Tecnologia de Reatores Nucleares Inovadores/CNPq (Brazil)

2017-07-01

MYRRHA (Multi-purpose Hybrid Research Reactor for High-tech Applications) is a prototype nuclear subcritical reactor driven by a particle accelerator. As a special property, the reactor maintains the nuclear fission chain reaction by means of an external neutron source provided by a particle accelerator. The main aim of this work is to study two types of coolants, LBE (Lead-Bismuth Eutectic) and Na (Sodium) that are two strong candidates to be used in ADS systems as well as in Generation IV (GEN-IV) reactors. Firstly, it was developed a thermal hydraulic model of the MYRRHA core using the RELAP5-3D, considering LBE as coolant (original project). After this, the LBE was substituted by Na coolant to investigate the reactor behavior in such case. Results have demonstrated the high heat transfer capacity of the LBE coolant in this type of system. (author)

A comparison of multi-echo spin-echo and triple-echo steady-state T2 mapping for in vivo evaluation of articular cartilage

Energy Technology Data Exchange (ETDEWEB)

Juras, Vladimir; Szomolanyi, Pavol [Medical University of Vienna, High Field MR Centre, Department of Biomedical Imaging and Image-Guided Therapy, Vienna (Austria); Institute of Measurement Science, Department of Imaging Methods, Bratislava (Slovakia); Bohndorf, Klaus; Kronnerwetter, Claudia; Hager, Benedikt; Zbyn, Stefan [Medical University of Vienna, High Field MR Centre, Department of Biomedical Imaging and Image-Guided Therapy, Vienna (Austria); Heule, Rahel; Bieri, Oliver [University of Basel Hospital, Division of Radiological Physics, Department of Radiology, Basel (Switzerland); Trattnig, Siegfried [Medical University of Vienna, High Field MR Centre, Department of Biomedical Imaging and Image-Guided Therapy, Vienna (Austria); Christian Doppler Laboratory for Clinical Molecular MR Imaging, Vienna (Austria); Ludwig Boltzmann Institute for Experimental and Clinical Traumatology, Austrian Cluster for Tissue Regeneration, Vienna (Austria)

2016-06-15

To assess the clinical relevance of T{sub 2} relaxation times, measured by 3D triple-echo steady-state (3D-TESS), in knee articular cartilage compared to conventional multi-echo spin-echo T{sub 2}-mapping. Thirteen volunteers and ten patients with focal cartilage lesions were included in this prospective study. All subjects underwent 3-Tesla MRI consisting of a multi-echo multi-slice spin-echo sequence (CPMG) as a reference method for T{sub 2} mapping, and 3D TESS with the same geometry settings, but variable acquisition times: standard (TESSs 4:35min) and quick (TESSq 2:05min). T{sub 2} values were compared in six different regions in the femoral and tibial cartilage using a Wilcoxon signed ranks test and the Pearson correlation coefficient (r). The local ethics committee approved this study, and all participants gave written informed consent. The mean quantitative T{sub 2} values measured by CPMG (mean: 46±9ms) in volunteers were significantly higher compared to those measured with TESS (mean: 31±5ms) in all regions. Both methods performed similarly in patients, but CPMG provided a slightly higher difference between lesions and native cartilage (CPMG: 90ms→61ms [31%],p=0.0125;TESS 32ms→24ms [24%],p=0.0839). 3D-TESS provides results similar to those of a conventional multi-echo spin-echo sequence with many benefits, such as shortening of total acquisition time and insensitivity to B{sub 1} and B{sub 0} changes. (orig.)

Physics of high performance JET plasmas in D-T

International Nuclear Information System (INIS)

2001-01-01

JET has recently operated with deuterium-tritium (D-T) mixtures, carried out an ITER physics campaign in hydrogen, deuterium, D-T and tritium, installed the Mark IIGB ''Gas Box'' divertor fully by remote handling and started physics experiments with this more closed divertor. The D-T experiments set records for fusion power (16.1 MW), ratio of fusion power to plasma input power (0.62, and 0.95±0.17 if a similar plasma could be obtained in steady-state) and fusion duration (4 MW for 4 s). A large scale tritium supply and processing plant, the first of its kind, allowed the repeated use of the 20 g tritium on site to supply 99.3 g of tritium to the machine. The H-mode threshold power is significantly lower in D-T, but the global energy confinement time is practically unchanged (no isotope effect). Dimensionless scaling ''Wind Tunnel'' experiments in D-T extrapolate to ignition with ITER parameters. The scaling is close to gyroBohm, but the mass dependence is not correct. Separating the thermal plasma energy into core and pedestal contributions could resolve this discrepancy (leading to proper gyroBohm scaling for the core) and also account for confinement degradation at high density and at high radiated power. Four radio frequency heating schemes have been tested successfully in D-T, showing good agreement with calculations. Alpha particle heating has been clearly observed and is consistent with classical expectations. Internal transport barriers have been established in optimised magnetic shear discharges for the first time in D-T and steady-state conditions have been approached with simultaneous internal and edge transport barriers. First results with the newly installed Mark IIGB divertor show that the in/out symmetry of the divertor plasma can be modified using differential gas fuelling, that optimised shear discharges can be produced, and that krypton gas puffing is effective in restoring L-mode edge conditions and establishing an internal transport barrier in

Physics of high performance jet plasmas in D-T

International Nuclear Information System (INIS)

1999-01-01

JET has recently operated with deuterium-tritium (D-T) mixtures, carried out an ITER physics campaign in hydrogen, deuterium, D-T and tritium, installed the Mark IIGB 'Gas Box' divertor fully by remote handling and started physics experiments with this more closed divertor. The D-T experiments set records for fusion power (16.1 MW), ratio of fusion power to plasma input power (0.62, and 0.95±0.17 if a similar plasma could be obtained in steady-state) and fusion duration (4 MW for 4 s). A large scale tritium supply and processing plant, the first of its kind, allowed the repeated use of the 20 g tritium on site to supply 99.3 g of tritium to the machine. The H-mode threshold power is significantly lower in D-T, but the global energy confinement time is practically unchanged (no isotope effect). Dimensionless scaling 'Wind Tunnel' experiments in D-T extrapolate to ignition with ITER parameters. The scaling is close to gyroBohm, but the mass dependence is not correct. Separating the thermal plasma energy into core and pedestal contributions could resolve this discrepancy (leading to proper gyroBohm scaling for the core) and also account for confinement degradation at high density and at high radiated power. Four radio frequency heating schemes have been tested successfully in D-T, showing good agreement with calculations. Alpha particle heating has been clearly observed and is consistent with classical expectations. Internal transport barriers have been established in optimised magnetic shear discharges for the first time in D-T and steady-state conditions have been approached with simultaneous internal and edge transport barriers. First results with the newly installed Mark IIGB divertor show that the in/out symmetry of the divertor plasma can be modified using differential gas fuelling, that optimised shear discharges can be produced, and that krypton gas puffing is effective in restoring L-mode edge conditions and establishing an internal transport barrier in such

Three-dimensional balanced steady state free precession myocardial perfusion cardiovascular magnetic resonance at 3T using dual-source parallel RF transmission: initial experience.

Science.gov (United States)

Jogiya, Roy; Schuster, Andreas; Zaman, Arshad; Motwani, Manish; Kouwenhoven, Marc; Nagel, Eike; Kozerke, Sebastian; Plein, Sven

2014-11-28

The purpose of this study was to establish the feasibility of three-dimensional (3D) balanced steady-state-free-precession (bSSFP) myocardial perfusion cardiovascular magnetic resonance (CMR) at 3T using local RF shimming with dual-source RF transmission, and to compare it with spoiled gradient echo (TGRE) acquisition. Dynamic contrast-enhanced 3D bSSFP perfusion imaging was performed on a 3T MRI scanner equipped with dual-source RF transmission technology. Images were reconstructed using k-space and time broad-use linear acquisition speed-up technique (k-t BLAST) and compartment based principle component analysis (k-t PCA). In phantoms and volunteers, local RF shimming with dual source RF transmission significantly improved B1 field homogeneity compared with single source transmission (P=0.01). 3D bSSFP showed improved signal-to-noise, contrast-to-noise and signal homogeneity compared with 3D TGRE (29.8 vs 26.9, P=0.045; 23.2 vs 21.6, P=0.049; 14.9% vs 12.4%, p=0.002, respectively). Image quality was similar between bSSFP and TGRE but there were more dark rim artefacts with bSSFP. k-t PCA reconstruction reduced artefacts for both sequences compared with k-t BLAST. In a subset of five patients, both methods correctly identified those with coronary artery disease. Three-dimensional bSSFP myocardial perfusion CMR using local RF shimming with dual source parallel RF transmission at 3T is feasible and improves signal characteristics compared with TGRE. Image artefact remains an important limitation of bSSFP imaging at 3T but can be reduced with k-t PCA.

Steady state HTTR model in the RELAP5-3D

Energy Technology Data Exchange (ETDEWEB)

Scari, Maria E.; Reis, Patrícia A. L.; Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A. F., E-mail: melizabethscari@yahoo.com, E-mail: patricialire@yahoo.com.br, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciências e Tecnologia de Reatores Nucleares Inovadores/CNPQ (Brazil)

2017-07-01

The HTTR (High Temperature Engineering Test Reactor) is a high temperature gas-cooled reactor (HTGR). It is a helium-cooled and graphite-moderated with a thermal power of 30 MW, developed and operated by the Japan Atomic Energy Research Institute (JAERI). In a previous work, a thermal model for the HTTR has been developed using the RELAP5-3D code. In the present work, additional thermal analyses have been presented. Studies about the core neutronic simulation were also initiated and are described here. The code WIMSD-5B was used to generate the cross sections and the group constants necessary to the NESTLE neutron kinetic code incorporated in the RELAP5-3D. (author)

Steady state investigation on neutronics of a molten salt reactor considering the flow effect of fuel salt

Institute of Scientific and Technical Information of China (English)

ZHANG Da-Lin; QIU Sui-Zheng; LIU Chang-Liang; SU Guang-Hui

2008-01-01

The Molten Salt Reactor (MSR),one of the‘Generation Ⅳ'concepts,is a liquid-fuel reactor,which is different from the conventional reactors using solid fissile materials due to the flow effect of fuel salt.The study on its neutronice considering the fuel salt flow,which is the base of the thermal-hydraulic calculation and safety analysis,must be done.In this paper,the theoretical model on neutronics under steady condition for a single-liquid-fueled MSR is conducted and calculated by numerical method.The neutronics model consists of two group neutron diffusion equations for fast and thermal neutron fluxes,and balance equations for six-group delayed neutron precursors considering the flow effect of fuel salt. The spatial discretization of the above models is based on the finite volume method,and the discretization equations are computed by the source iteration method.The distributions of neutron fluxes and the distributions of the delayed neutron precursors in the core are obtained.The numerical calculated results show that,the fuel salt flow has little effect on the distribution of fast and thermal neutron fluxes and the effective multiplication factor;however,it affects the distribution of the delayed neutron precursors significantly,especially the long-lived one.In addition,it could be found that the delayed neutron precursors influence the nentronics slightly under the steady condition.

Steady state investigation on neutronics of a molten salt reactor considering the flow effect of fuel salt

International Nuclear Information System (INIS)

Zhang Dalin; Qiu Suizheng; Su Guanghui; Liu Changliang

2008-01-01

The Molten Salt Reactor (MSR), one of the 'Generation IV' concepts, is a liquid-fuel reactor, which is different from the conventional reactors using solid fissile materials due to the flow effect of fuel salt. The study on its neutronics considering the fuel salt flow, which is the base of the thermal-hydraulic calculation and safety analysis, must be done. In this paper, the theoretical model on neutronics under steady condition for a single-liquid-fueled MSR is conducted and calculated by numerical method. The neutronics model consists of two group neutron diffusion equations for fast and thermal neutron fluxes, and balance equations for six-group delayed neutron precursors considering the flow effect of fuel salt. The spatial discretization of the above models is based on the finite volume method, and the discretization equations are computed by the source iteration method. The distributions of neutron fluxes and the distributions of the delayed neutron precursors in the core are obtained. The numerical calculated results show that, the fuel salt flow has little effect on the distribution of fast and thermal neutron fluxes and the effective multiplication factor; however, it affects the distribution of the delayed neutron precursors significantly, especially the long-lived one. In addition, it could be found that the delayed neutron precursors influence the neutronics slightly under the steady condition. (authors)

FRAPCON-2: A Computer Code for the Calculation of Steady State Thermal-Mechanical Behavior of Oxide Fuel Rods

Energy Technology Data Exchange (ETDEWEB)

Berna, G. A; Bohn, M. P.; Rausch, W. N.; Williford, R. E.; Lanning, D. D.

1981-01-01

FRAPCON-2 is a FORTRAN IV computer code that calculates the steady state response of light Mater reactor fuel rods during long-term burnup. The code calculates the temperature, pressure, deformation, and tai lure histories of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include (a) heat conduction through the fuel and cladding, (b) cladding elastic and plastic deformation, (c) fuel-cladding mechanical interaction, (d) fission gas release, (e} fuel rod internal gas pressure, (f) heat transfer between fuel and cladding, (g) cladding oxidation, and (h) heat transfer from cladding to coolant. The code contains necessary material properties, water properties, and heat transfer correlations. FRAPCON-2 is programmed for use on the CDC Cyber 175 and 176 computers. The FRAPCON-2 code Is designed to generate initial conditions for transient fuel rod analysis by either the FRAP-T6 computer code or the thermal-hydraulic code, RELAP4/MOD7 Version 2.

Study of impurity effects on CFETR steady-state scenario by self-consistent integrated modeling

Science.gov (United States)

Shi, Nan; Chan, Vincent S.; Jian, Xiang; Li, Guoqiang; Chen, Jiale; Gao, Xiang; Shi, Shengyu; Kong, Defeng; Liu, Xiaoju; Mao, Shifeng; Xu, Guoliang

2017-12-01

Impurity effects on fusion performance of China fusion engineering test reactor (CFETR) due to extrinsic seeding are investigated. An integrated 1.5D modeling workflow evolves plasma equilibrium and all transport channels to steady state. The one modeling framework for integrated tasks framework is used to couple the transport solver, MHD equilibrium solver, and source and sink calculations. A self-consistent impurity profile constructed using a steady-state background plasma, which satisfies quasi-neutrality and true steady state, is presented for the first time. Studies are performed based on an optimized fully non-inductive scenario with varying concentrations of Argon (Ar) seeding. It is found that fusion performance improves before dropping off with increasing {{Z}\\text{eff}} , while the confinement remains at high level. Further analysis of transport for these plasmas shows that low-k ion temperature gradient modes dominate the turbulence. The decrease in linear growth rate and resultant fluxes of all channels with increasing {{Z}\\text{eff}} can be traced to impurity profile change by transport. The improvement in confinement levels off at higher {{Z}\\text{eff}} . Over the regime of study there is a competition between the suppressed transport and increasing radiation that leads to a peak in the fusion performance at {{Z}\\text{eff}} (~2.78 for CFETR). Extrinsic impurity seeding to control divertor heat load will need to be optimized around this value for best fusion performance.

A Comparison between Gadofosveset Trisodium and Gadobenate Dimeglumine for Steady State MRA of the Thoracic Vasculature

OpenAIRE

Camren, G. Paul; Wilson, Gregory J.; Bamra, Vikram R.; Nguyen, Khahn Q.; Hippe, Daniel S.; Maki, Jeffrey H.

2014-01-01

Purpose. Retrospective comparison between gadofosveset trisodium and gadobenate dimeglumine steady state magnetic resonance angiography (SS-MRA) of the thoracic vasculature at 1.5T using signal-to-noise ratio (SNR) and vessel edge sharpness (ES) as markers of image quality. Materials and Methods. IRB approval was obtained. Twenty separate patients each underwent SS-MRA using high-resolution 3D ECG-triggered coronal IR-TFE at 1.5T approximately 3-4 minutes following 10 or 15 mL gadofosveset or...

Simulations of KSTAR high performance steady state operation scenarios

International Nuclear Information System (INIS)

Na, Yong-Su; Kessel, C.E.; Park, J.M.; Yi, Sumin; Kim, J.Y.; Becoulet, A.; Sips, A.C.C.

2009-01-01

We report the results of predictive modelling of high performance steady state operation scenarios in KSTAR. Firstly, the capabilities of steady state operation are investigated with time-dependent simulations using a free-boundary plasma equilibrium evolution code coupled with transport calculations. Secondly, the reproducibility of high performance steady state operation scenarios developed in the DIII-D tokamak, of similar size to that of KSTAR, is investigated using the experimental data taken from DIII-D. Finally, the capability of ITER-relevant steady state operation is investigated in KSTAR. It is found that KSTAR is able to establish high performance steady state operation scenarios; β N above 3, H 98 (y, 2) up to 2.0, f BS up to 0.76 and f NI equals 1.0. In this work, a realistic density profile is newly introduced for predictive simulations by employing the scaling law of a density peaking factor. The influence of the current ramp-up scenario and the transport model is discussed with respect to the fusion performance and non-inductive current drive fraction in the transport simulations. As observed in the experiments, both the heating and the plasma current waveforms in the current ramp-up phase produce a strong effect on the q-profile, the fusion performance and also on the non-inductive current drive fraction in the current flattop phase. A criterion in terms of q min is found to establish ITER-relevant steady state operation scenarios. This will provide a guideline for designing the current ramp-up phase in KSTAR. It is observed that the transport model also affects the predictive values of fusion performance as well as the non-inductive current drive fraction. The Weiland transport model predicts the highest fusion performance as well as non-inductive current drive fraction in KSTAR. In contrast, the GLF23 model exhibits the lowest ones. ITER-relevant advanced scenarios cannot be obtained with the GLF23 model in the conditions given in this work

Fusion reactor start-up without an external tritium source

Energy Technology Data Exchange (ETDEWEB)

Zheng, S., E-mail: Shanliang.Zheng@ccfe.ac.uk; King, D.B.; Garzotti, L.; Surrey, E.; Todd, T.N.

2016-02-15

Highlights: • Investigated the feasibility (including plasma physics, neutronics and economics) of starting a fusion reactor from running pure D–D fusion reactor to gradually move towards the D–T operation. • Proposed building up tritium from making use of neutrons generated by D–D fusion reactions. • Studied plasma physics feasibility for pure D–D operation and provided consistent fusion power and neutron yield in the plasma with different mixture of deuterium and tritium. • Discussed the economics aspect for operating a pure D–D fusion reactor towards a full-power D–T fusion reactor. - Abstract: It has long been recognised that the shortage of external tritium sources for fusion reactors using D–T, the most promising fusion fuel, requires all such fusion power plants (FPP) to breed their own tritium. It is also recognised that the initial start-up of a fusion reactor will require several kilograms of tritium within a scenario in which radioactive decay, ITER and subsequent demonstrator reactors are expected to have consumed most of the known tritium stockpile. To circumvent this tritium fuel shortage and ultimately achieve steady-state operation for a FPP, it is essential to first accumulate sufficient tritium to compensate for loss due to decay and significant retention in the materials in order to start a new FPP. In this work, we propose to accumulate tritium starting from D–D fusion reactions, since D exists naturally in water, and to gradually build up the D–T plasma targeted in fusion reactor designs. There are two likely D–D fusion reaction channels, (1) D + D → T + p, and (2) D + D → He3 + n. The tritium can be generated via the reaction channel ‘(1)’ and the 2.45 MeV neutrons from ‘(2)’ react with lithium-6 in the breeding blanket to produce more tritium to be fed back into plasma fuel. Quantitative evaluations are conducted for two blanket concepts to assess the feasibility and suitability of this approach to FPP

COOLOD-N: a computer code, for the analyses of steady-state thermal-hydraulics in plate-type research reactors

International Nuclear Information System (INIS)

Kaminaga, Masanori

1990-02-01

The COOLOD-N code provides a capability for the analysis of the steady-state thermal-hydraulics of research reactors in which plate-type fuel is employed. This code is revised version of the COOLOD code, and is applicable not only to a forced convection cooling mode, but also to a natural convection cooling mode. In the code, a function to calculate flow rate under a natural convection, and a heat transfer package which was a subroutine program to calculate heat transfer coefficient, ONB temperature and DNB heat flux, and was especially developed for the upgraded JRR-3, have been newly added to the COOLOD code. The COOLOD-N code also has a capability of calculating the heat flux at onset of flow instability as well as DNB heat flux. (author)

Development of steady thermal-hydraulic analysis code for China advanced research reactor

International Nuclear Information System (INIS)

Tian Wenxi; Qiu Suizheng; Guo Yun; Su Guanghui; Jia Dounan; Liu Tiancai; Zhang Jianwei

2006-01-01

A multi-channel model steady-state thermal-hydraulic analysis code was developed for China Advanced Research Reactor (CARR). By simulating the whole reactor core, the detailed flow distribution in the core was obtained. The result shows that the structure size plays the most important role in flow distribution and the influence of core power could be neglected under single-phase flow. The temperature field of fuel element under unsymmetrical cooling condition was also obtained, which is necessary for the further study such as stress analysis etc. of the fuel element. At the same time, considering the hot channel effect including engineering factor and nuclear factor, calculation of hot channel was carried out and it is proved that all thermal-hydraulic parameters accord with the Safety Regulation of CARR. (authors)

STEADY STATE MODELING OF THE MINIMUM CRITICAL CORE OF THE TRANSIENT REACTOR TEST FACILITY

Energy Technology Data Exchange (ETDEWEB)

Anthony L. Alberti; Todd S. Palmer; Javier Ortensi; Mark D. DeHart

2016-05-01

With the advent of next generation reactor systems and new fuel designs, the U.S. Department of Energy (DOE) has identified the need for the resumption of transient testing of nuclear fuels. The DOE has decided that the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory (INL) is best suited for future testing. TREAT is a thermal neutron spectrum, air-cooled, nuclear test facility that is designed to test nuclear fuels in transient scenarios. These specific scenarios range from simple temperature transients to full fuel melt accidents. DOE has expressed a desire to develop a simulation capability that will accurately model the experiments before they are irradiated at the facility. It is the aim for this capability to have an emphasis on effective and safe operation while minimizing experimental time and cost. The multi physics platform MOOSE has been selected as the framework for this project. The goals for this work are to identify the fundamental neutronics properties of TREAT and to develop an accurate steady state model for future multiphysics transient simulations. In order to minimize computational cost, the effect of spatial homogenization and angular discretization are investigated. It was found that significant anisotropy is present in TREAT assemblies and to capture this effect, explicit modeling of cooling channels and inter-element gaps is necessary. For this modeling scheme, single element calculations at 293 K gave power distributions with a root mean square difference of 0.076% from those of reference SERPENT calculations. The minimum critical core configuration with identical gap and channel treatment at 293 K resulted in a root mean square, total core, radial power distribution 2.423% different than those of reference SERPENT solutions.

steady – state performance of induction and transfer state

African Journals Online (AJOL)

eobe

This paper presents paper presents paper presents the steady the steady the steady–state performance state performance state performance comparison comparison comparison between polyphase induction motor and polyphase between polyphase induction motor and polyphase. TF motor operating in. TF motor ...

Understanding void fraction in steady state and dynamic environments

International Nuclear Information System (INIS)

Chexal, B.; Maulbetsch, J.; Harrison, J.; Petersen, C.; Jensen, P.; Horowitz, J.

1997-01-01

Understanding void fraction behavior in steady-state and dynamic environments is important to accurately predict the thermal-hydraulic behavior of two-phase or two-component systems. The Chexal-Lellouche (C-L) void fraction mode described herein covers the full range of pressures, flows, void fractions, and fluid types (steam-water, air-water, and refrigerants). A drift flux model formulation is used which covers the complete range of concurrent and countercurrent flows. The (1996) model revises the earlier C-L void fraction correlation, improves the capability of the model in countercurrent flow based on the incorporation of additional data, and improves the characteristics of the correlation that are important in transient programs. The model has been qualified with data from a number of steady state two-phase and two-component tests, and has been incorporated into the transient analysis code RELAP5 and RETRAN-3D and evaluated with a variety of transient and steady state tests. A 'plug-in' module for the void fraction correlation has been developed and implemented in RELAP5 and RETRAN-3D. The module is available as source code for inclusion into other thermal-hydraulic programs and can be used in any program that utilizes the same interface variables

The technology and science of steady-state operation in magnetically confined plasmas

International Nuclear Information System (INIS)

Becoulet, A; Hoang, G T

2008-01-01

fusion device. The first includes specific additional heating and current drive methods (through externally launched radio-frequency waves or energetic atoms), fuelling and pumping methods, dedicated plasma diagnostics as well as software technologies required for mandatory real time control loops, involving such actuators and sensors. The second class of technologies, generic to any magnetic fusion device, includes the superconducting magnet technologies, in order to provide a stationary confinement magnetic field, the actively cooled plasma facing components (PFCs) handling either radiated or convected power fluxes (often in excess of several tens of MW m -2 ), dedicated diagnostics monitoring the interfaces (e.g. infrared survey of PFCs), etc. The detailed specifications of all elements must comply with a reactor-relevant environment, in terms of operational parameters as well as lifetime. The paper presents a summary of the present status and understanding of the technology and science of steady-state operation in magnetically confined plasmas, as well as the forthcoming work programme dedicated to the vast R and D programme undertaken in this domain, in particular within the European fusion framework.

Thermal hydraulic and neutron kinetic coupled simulation of the IPR-R1 Triga reactor

Energy Technology Data Exchange (ETDEWEB)

Reis, Patricia A.L.; Costa, Antonella L.; Pereira, Claubia; Silva, Clarysson A.M. da; Veloso, Maria Auxiliadora F.; Soares, Humbero V., E-mail: patricialire@yahoo.com.br, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br, E-mail: clarysson@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br, E-mail: betovitor@ig.com.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores (INCT/CNPq Rede), Rio de Janeiro, RJ (Brazil)

2013-07-01

The nuclear industry and the scientific community have turned the attention for the development of coupled 3D neutron kinetics (NK) and thermal-hydraulic (TH) system codes to investigate specific nuclear reactor transients. Improving in theoretical investigations of complex phenomena in nuclear reactor technology have been increased thanks to numerical methods and computational resources incorporated in nuclear codes. This paper presents a model for the IPR-R1 TRIGA research reactor using the RELAP5-3D 3.0 code. The development and the assessment of the thermal-hydraulic RELAP5 code model for the IPR-R1 have been validated for steady state and transient situations and the results were published in preceding works. Results of RELAP5-3D steady state and a transient case presented in this paper show good agreement with experimental data, validating then this model for point kinetic calculations. To supply adequate cross sections to the NK code, the WIMSD5 is being used. First results of steady state calculation using the 3D neutron modeling are being presented in this paper. (author)

Steady state and linear stability analysis of a supercritical water natural circulation loop

International Nuclear Information System (INIS)

Sharma, Manish; Pilkhwal, D.S.; Vijayan, P.K.; Saha, D.; Sinha, R.K.

2010-01-01

Supercritical water (SCW) has excellent heat transfer characteristics as a coolant for nuclear reactors. Besides it results in high thermal efficiency of the plant. However, the flow can experience instabilities in supercritical water reactors, as the density change is very large for the supercritical fluids. A computer code SUCLIN using supercritical water properties has been developed to carry out the steady state and linear stability analysis of a SCW natural circulation loop. The conservation equations of mass, momentum and energy have been linearized by imposing small perturbation in flow rate, enthalpy, pressure and specific volume. The equations have been solved analytically to generate the characteristic equation. The roots of the equation determine the stability of the system. The code has been qualitatively assessed with published results and has been extensively used for studying the effect of diameter, height, heater inlet temperature, pressure and local loss coefficients on steady state and stability behavior of a Supercritical Water Natural Circulation Loop (SCWNCL). The present paper describes the linear stability analysis model and the results obtained in detail.

Local transport in Joint European Tokamak edge-localized, high-confinement mode plasmas with H, D, DT, and T isotopes

International Nuclear Information System (INIS)

Budny, R. V.; Ernst, D. R.; Hahm, T. S.; McCune, D. C.; Christiansen, J. P.; Cordey, J. G.; Gowers, C. G.; Guenther, K.; Hawkes, N.; Jarvis, O. N.

2000-01-01

The edge-localized, high-confinement mode regime is of interest for future Tokamak reactors since high performance has been sustained for long durations. Experiments in the Joint European Tokamak [M. Keilhacker , Nuclear Fusion 39, 209 (1999)] have studied this regime using scans with the toroidal field and plasma current varied together in H, D, DT, and T isotopes. The local energy transport in more than fifty of these plasmas is analyzed, and empirical scaling relations are derived for energy transport coefficients during quasi-steady state conditions using dimensionless parameters. Neither the Bohm nor gyro-Bohm expressions give the shapes of the profiles. The scalings with β and ν * are in qualitative agreement with Ion Temperature Gradient theory

High Beta Steady State Research and Future Directions on JT-60U and JFT-2M

Science.gov (United States)

Ishida, Shinichi

2003-10-01

JT-60U and JFT-2M research is focused on high beta steady state operation towards economically and environmentally attractive reactors. In JT-60U, a high-βp H-mode plasma was sustained with βN 2.7 for 7.4 s in which neoclassical tearing modes (NTMs) limited the attainable β_N. Real-time tracking NTM stabilization system using ECCD demonstrated complete suppression of NTM leading to recovery of βN before onset of NTM. Performance in a fully non-inductive H-mode plasma was improved up to n_i(0) τE T_i(0) = 3.1 x 10^20 keV s m-3 using N-NBCD with βN 2.4, HH_y,2=1.2 and bootstrap fraction f_BS 0.5. ECH experiments extended the confinement enhancement for dominantly electron heated reversed shear plasmas up to HH_y,2 2 at T_e/Ti 1.25. A world record ECCD efficiency, 4.2 x 10^18 A/W/m^2, was achieved at Te 23 keV with a highly localized central current density. Innovative initiation and current build-up without center solenoid currents were established by LHCD/ECH and bootstrap current up to f_BS 0.9. In JFT-2M, the inside of the vacuum vessel wall was fully covered with low-activation ferritic steel plates to investigate their use in plasmas near fusion conditions. High βN plasmas were produced up to βN = 3.3 with an internal transport barrier (ITB) and a steady H-mode edge. A new H-mode regime with steady high recycling (HRS) and an ITB was exploited leading to βN H_89P 6.2 at n_e/nG 0.7. In 2003, JT-60U will be able to operate for the duration up to 65 s at 1 MA/2.7 T and the heating/current-drive duration up to 30 s at 17 MW to prolong high-βN and/or high-f_BS discharges with feedback controls. JFT-2M is planning to implement wall stabilization experiments in 2004 to pursue plasmas above the ideal no-wall limit using a ferritic wall. The modification of JT-60 to a fully superconducting tokamak is under discussion to explore high-β steady state operation in collision-less plasmas well above no-wall limit with ferritic wall in a steady state.

Competing States in the t-J Model: Uniform d-Wave State versus Stripe State versus Stripe State

NARCIS (Netherlands)

Corboz, P.R.; Rice, T.M.; Troyer, M.

2014-01-01

Variational studies of the t-J model on the square lattice based on infinite projected-entangled pair states confirm an extremely close competition between a uniform d-wave superconducting state and different stripe states. The site-centered stripe with an in-phase d-wave order has an equal or only

Steady-state and pre-steady-state kinetic analysis of halopropane conversion by a Rhodococcus haloalkane dehalogenase

NARCIS (Netherlands)

Bosma, T; Pikkemaat, MG; Kingma, Jacob; Dijk, J; Janssen, DB

2003-01-01

Haloalkane dehalogenase from Rhodococcus rhodochrous NCIMB 13064 (DhaA) catalyzes the hydrolysis of carbon-halogen bonds in a wide range of haloalkanes. We examined the steady-state and pre-steady-state kinetics of halopropane conversion by DhaA to illuminate mechanistic details of the

Pseudo Steady-State Free Precession for MR-Fingerprinting.

Science.gov (United States)

Assländer, Jakob; Glaser, Steffen J; Hennig, Jürgen

2017-03-01

This article discusses the signal behavior in the case the flip angle in steady-state free precession sequences is continuously varied as suggested for MR-fingerprinting sequences. Flip angle variations prevent the establishment of a steady state and introduce instabilities regarding to magnetic field inhomogeneities and intravoxel dephasing. We show how a pseudo steady state can be achieved, which restores the spin echo nature of steady-state free precession. Based on geometrical considerations, relationships between the flip angle, repetition and echo time are derived that suffice to the establishment of a pseudo steady state. The theory is tested with Bloch simulations as well as phantom and in vivo experiments. A typical steady-state free precession passband can be restored with the proposed conditions. The stability of the pseudo steady state is demonstrated by comparing the evolution of the signal of a single isochromat to one resulting from a spin ensemble. As confirmed by experiments, magnetization in a pseudo steady state can be described with fewer degrees of freedom compared to the original fingerprinting and the pseudo steady state results in more reliable parameter maps. The proposed conditions restore the spin-echo-like signal behavior typical for steady-state free precession in fingerprinting sequences, making this approach more robust to B 0 variations. Magn Reson Med 77:1151-1161, 2017. © 2016 International Society for Magnetic Resonance in Medicine. © 2016 International Society for Magnetic Resonance in Medicine.

Modeling steady state and transient fission gas behaviour with the Karlsruhe code LAKU

International Nuclear Information System (INIS)

Vaeth, L.

1984-08-01

The programme LAKU models the behaviour of gaseous fission products in reactor fuel under steady state and transient conditions, including molten fuel. A presentation of the full model is given, starting with gas behaviour in the grains and on grain faces and including the treatment of release from porosity. The results of some recent calculations are presented. (orig.) [de

Progress towards Steady State at Low Aspect Ratio on the National Spherical Torus Experiment (NSTX)

International Nuclear Information System (INIS)

Gates, D.A.; Menard, J.; Maingi, R.; Kaye, S.; Sabbagh, S.A.; Diem, S.; Wilson, J.R.; Bell, M.G.; Bell, R.E.; Ferron, J.; Fredrickson, E.D.; Kessel, C.E.; LeBlanc, B.P.; Levinton, F.; Manickam, J.; Mueller, D.; Raman, R.; Stevenson, T.; Stutman, D.; Taylor, G.; Tritz, K.; Yu, H.

2007-01-01

Modifications to the plasma control capabilities and poloidal field coils of the National Spherical Torus Experiment (NSTX) have enabled a significant enhancement in shaping capability which has led to the transient achievement of a record shape factor (S (triple b ond) q 95 (I p /aB t )) of ∼ 41 (MA m -1 T -1 ) simultaneous with a record plasma elongation of κ (triple b ond) b/a ∼ 3. This result was obtained using isoflux control and real-time equilibrium reconstruction. Achieving high shape factor together with tolerable divertor loading is an important result for future ST burning plasma experiments as exemplified by studies for future ST reactor concepts, as well as neutron producing devices, which rely on achieving high shape factors in order to achieve steady state operation while maintaining MHD stability. Statistical evidence is presented which demonstrates the expected correlation between increased shaping and improved plasma performance.

Time Reversibility, Correlation Decay and the Steady State Fluctuation Relation for Dissipation

Directory of Open Access Journals (Sweden)

Denis J. Evans

2013-04-01

Full Text Available Steady state fluctuation relations for nonequilibrium systems are under intense investigation because of their important practical implications in nanotechnology and biology. However the precise conditions under which they hold need clarification. Using the dissipation function, which is related to the entropy production of linear irreversible thermodynamics, we show time reversibility, ergodic consistency and a recently introduced form of correlation decay, called T-mixing, are sufficient conditions for steady state fluctuation relations to hold. Our results are not restricted to a particular model and show that the steady state fluctuation relation for the dissipation function holds near or far from equilibrium subject to these conditions. The dissipation function thus plays a comparable role in nonequilibrium systems to thermodynamic potentials in equilibrium systems.

The simplified P3 approach on a trigonal geometry in the nodal reactor code DYN3D

International Nuclear Information System (INIS)

Duerigen, S.; Fridman, E.

2011-01-01

DYN3D is a three-dimensional nodal diffusion code for steady-state and transient analyses of Light-Water Reactors with square and hexagonal fuel assembly geometries. Currently, several versions of the DYN3D code are available including a multi-group diffusion and a simplified P 3 (SP 3 ) neutron transport option. In this work, the multi-group SP 3 method based on trigonal-z geometry was developed. The method is applicable to the analysis of reactor cores with hexagonal fuel assemblies and allows flexible mesh refinement, which is of particular importance for WWER-type Pressurized Water Reactors as well as for innovative reactor concepts including block type High-Temperature Reactors and Sodium Fast Reactors. In this paper, the theoretical background for the trigonal SP 3 methodology is outlined and the results of a preliminary verification analysis are presented by means of a simplified WWER-440 core test example. The accordant cross sections and reference solutions were produced by the Monte Carlo code SERPENT. The DYN3D results are in good agreement with the reference solutions. The average deviation in the nodal power distribution is about 1%. (Authors)

Finite element modelling of creep process - steady state stresses and strains

Directory of Open Access Journals (Sweden)

Sedmak Aleksandar S.

2014-01-01

Full Text Available Finite element modelling of steady state creep process has been described. Using an analogy of visco-plastic problem with a described procedure, the finite element method has been used to calculate steady state stresses and strains in 2D problems. An example of application of such a procedure have been presented, using real life problem - cylindrical pipe with longitudinal crack at high temperature, under internal pressure, and estimating its residual life, based on the C*integral evaluation.

Three-dimensional constructive interference in steady-state magnetic resonance imaging in syringomyelia: advantages over conventional imaging.

Science.gov (United States)

Roser, Florian; Ebner, Florian H; Danz, Søren; Riether, Felix; Ritz, Rainer; Dietz, Klaus; Naegele, Thomas; Tatagiba, Marcos S

2008-05-01

Neuroradiology has become indispensable in detecting the pathophysiology in syringomyelia. Constructive interference in steady-state (CISS) magnetic resonance (MR) imaging can provide superior contrast at the sub-arachnoid tissue borders. As this region is critical in preoperative evaluation, the authors hypothesized that CISS imaging would provide superior assessment of syrinx pathology and surgical planning. Based on records collected from a database of 130 patients with syringomyelia treated at the authors' institution, 59 patients were prospectively evaluated with complete neuroradiological examinations. In addition to routine acquisitions with FLAIR, T1- and T2-weighted, and contrast-enhanced MR imaging series, the authors obtained sagittal cardiac-gated sequences to visualize cerebrospinal fluid (CSF) pulsations and axial 3D CISS MR sequences to detect focal arachnoid webs. Statistical qualitative and quantitative evaluations of spinal cord/CSF contrast, spinal cord/CSF delineation, motion artifacts, and artifacts induced by pulsatile CSF flow were performed. The 3D CISS MR sequences demonstrated a contrast-to-noise ratio significantly better than any other routine imaging sequence (p CSF flow voids. Constructive interference in steady-state MR imaging enables the neurosurgeon to accurately identify cases requiring decompression for obstructed CSF. Motion artifacts can be eliminated with technical variations.

Dissipative dark matter halos: The steady state solution. II.

Science.gov (United States)

Foot, R.

2018-05-01

Within the mirror dark matter model and dissipative dark matter models in general, halos around galaxies with active star formation (including spirals and gas-rich dwarfs) are dynamical: they expand and contract in response to heating and cooling processes. Ordinary type II supernovae (SNe) can provide the dominant heat source, which is possible if kinetic mixing interaction exists with strength ɛ ˜10-9- 10-10 . Dissipative dark matter halos can be modeled as a fluid governed by Euler's equations. Around sufficiently isolated and unperturbed galaxies the halo can relax to a steady state configuration, where heating and cooling rates locally balance and hydrostatic equilibrium prevails. These steady state conditions can be solved to derive the physical properties, including the halo density and temperature profiles, for model galaxies. Here, we consider idealized spherically symmetric galaxies within the mirror dark particle model, as in our earlier paper [Phys. Rev. D 97, 043012 (2018), 10.1103/PhysRevD.97.043012], but we assume that the local halo heating in the SN vicinity dominates over radiative sources. With this assumption, physically interesting steady state solutions arise which we compute for a representative range of model galaxies. The end result is a rather simple description of the dark matter halo around idealized spherically symmetric systems, characterized in principle by only one parameter, with physical properties that closely resemble the empirical properties of disk galaxies.

User's manual for ASTERIX-2: A two-dimensional modular code system for the steady state and xenon transient analysis of a pebble bed high temperature reactor

International Nuclear Information System (INIS)

Wu, T.; Cowan, C.L.; Lauer, A.; Schwiegk, H.J.

1982-03-01

The ASTERIX modular code package was developed at KFA Laboratory-Juelich for the steady state and xenon transient analysis of a pebble bed high temperature reactor. The code package was implemented on the Stanford Linear Accelerator Center Computer in August, 1980, and a user's manual for the current version of the code, identified as ASTERIX-2, was prepared as a cooperative effort by KFA Laboratory and GE-ARSD. The material in the manual includes the requirements for accessing the program, a description of the major subroutines, a listing of the input options, and a listing of the input data for a sample problem. The material is provided in sufficient detail for the user to carry out a wide range of analysis from steady state operations to the xenon induced power transients in which the local xenon, temperature, buckling and control feedback effects have been incorporated in the problem solution. (orig.)

User's manual for ASTERIX-2: a two-dimensional modular-code system for the steady-state and xenon-transient analysis of a pebble-bed high-temperature reactor

International Nuclear Information System (INIS)

Lauer, A.; Schwiegk, H.J.; Wu, T.; Cowan, C.L.

1982-03-01

The ASTERIX modular code package was developed at KFA Laboratory-Juelich for the steady state and xenon transient analysis of a pebble bed high temperature reactor. The code package was implemented on the Stanford Linear Accelerator Center Computer in August, 1980, and a user's manual for the current version of the code, identified as ASTERIX-2, was prepared as a cooperative effort by KFA Laboratory and GE-ARSD. The material in the manual includes the requirements for accessing the program, a description of the major subroutines, a listing of the input options, and a listing of the input data for a sample problem. The material is provided in sufficient detail for the user to carry out a wide range of analyses from steady state operations to the xenon induced power transients in which the local xenon, temperature, buckling and control feedback effects have been incorporated in the problem solution

Steady-State Creep of Asphalt Concrete

Directory of Open Access Journals (Sweden)

Alibai Iskakbayev

2017-02-01

Full Text Available This paper reports the experimental investigation of the steady-state creep process for fine-grained asphalt concrete at a temperature of 20 ± 2 °С and under stress from 0.055 to 0.311 MPa under direct tension and was found to occur at a constant rate. The experimental results also determined the start, the end point, and the duration of the steady-state creep process. The dependence of these factors, in addition to the steady-state creep rate and viscosity of the asphalt concrete on stress is satisfactorily described by a power function. Furthermore, it showed that stress has a great impact on the specific characteristics of asphalt concrete: stress variation by one order causes their variation by 3–4.5 orders. The described relations are formulated for the steady-state of asphalt concrete in a complex stressed condition. The dependence is determined between stress intensity and strain rate intensity.

Comparison of two single-breath-held 3-D acquisitions with multi-breath-held 2-D cine steady-state free precession MRI acquisition in children with single ventricles

Energy Technology Data Exchange (ETDEWEB)

Atweh, Lamya A.; Dodd, Nicholas A.; Krishnamurthy, Ramkumar; Chu, Zili D. [Texas Children' s Hospital, EB Singleton Department of Pediatric Radiology, Cardiovascular Imaging, Houston, TX (United States); Pednekar, Amol [Philips Healthcare, Houston, TX (United States); Krishnamurthy, Rajesh [Texas Children' s Hospital, EB Singleton Department of Pediatric Radiology, Cardiovascular Imaging, Houston, TX (United States); Baylor College of Medicine, Department of Radiology, Houston, TX (United States); Baylor College of Medicine, Department of Pediatrics, Houston, TX (United States)

2016-05-15

Breath-held two-dimensional balanced steady-state free precession cine acquisition (2-D breath-held SSFP), accelerated with parallel imaging, is the method of choice for evaluating ventricular function due to its superior blood-to-myocardial contrast, edge definition and high intrinsic signal-to-noise ratio throughout the cardiac cycle. The purpose of this study is to qualitatively and quantitatively compare the two different single-breath-hold 3-D cine SSFP acquisitions using 1) multidirectional sensitivity encoding (SENSE) acceleration factors (3-D multiple SENSE SSFP), and 2) k-t broad-use linear acceleration speed-up technique (3-D k-t SSFP) with the conventional 2-D breath-held SSFP in non-sedated asymptomatic volunteers and children with single ventricle congenital heart disease. Our prospective study was performed on 30 non-sedated subjects (9 healthy volunteers and 21 functional single ventricle patients), ages 12.5 +/- 2.8 years. Two-dimensional breath-held SSFP with SENSE acceleration factor of 2, eight-fold accelerated 3-D k-t SSFP, and 3-D multiple SENSE SSFP with total parallel imaging factor of 4 were performed to evaluate ventricular volumes and mass in the short-axis orientation. Image quality scores (blood myocardial contrast, edge definition and interslice alignment) and volumetric analysis (end systolic volume, end diastolic volume and ejection fraction) were performed on the data sets by experienced users. Paired t-test was performed to compare each of the 3-D k-t SSFP and 3-D multiple SENSE SSFP clinical scores against 2-D breath-held SSFP. Bland-Altman analysis was performed on left ventricle (LV) and single ventricle volumetry. Interobserver and intraobserver variability in volumetric measurements were determined using intraclass coefficients. The clinical scores were highest for the 2-D breath-held SSFP images. Between the two 3-D sequences, 3-D multiple SENSE SSFP performed better than 3-D k-t SSFP. Bland-Altman analysis for volumes

Tritium permeation in fusion reactors: INTOR

International Nuclear Information System (INIS)

Baskes, M.I.; Bauer, W.; Kerst, R.A.; Swansiger, W.A.; Wilson, K.L.

1981-12-01

Tritium permeation through the first wall of advanced fusion reactors is examined. A fraction of the D-T which bombards the first wall as charge exchange neutral particles will permeate through the first wall and enter the coolant. Calculations of the steady state permeation rate for the US INTOR Tokamak design result in values of less than or equal to 0.002 grams of tritium per day under the most favorable conditions. For unfavorable surface conditions the rate is greater than or equal to 0.1 g/day. The magnitude of these permeation rates is critically dependent on the temperatures and surface conditions of the wall. The introduction of permeation barriers at the wall-coolant interface can significantly reduce permeation rates and hence may be desirable for reactor applications

HESTER: a hot-electron superconducting tokamak experimental reactor at M.I.T

International Nuclear Information System (INIS)

Schultz, J.H.; Montgomery, D.B.

1983-04-01

HESTER is an experimental tokamak, designed to resolve many of the central questions in the tokamak development program in the 1980's. It combines several unique features with new perspectives on the other major tokamak experiments scheduled for the next decade. The overall objectives of HESTER, in rough order of their presently perceived importance, are the achievement of reactor-like wall-loadings and plasma parameters for long pulse periods, determination of a good, reactor-relevant method of steady-state or very long pulse tokamak current drive, duplication of the planned very high temperature neutral injection experiments using only radio frequency heating, a demonstration of true steady-state tokamak operation, integration of a high-performance superconducting magnet system into a tokamak experiment, determination of the best methods of long term impurity control, and studies of transport and pressure limits in high field, high aspect ratio tokamak plasmas. These objectives are described

Preliminary Evaluation of the Adequacy of Lithium Resources of the World and China for D-T Fusion Reactors

Science.gov (United States)

Wang, Yongliang; Ni, Muyi; Jiang, Jieqiong; Wu, Yican; FDS-Team

2012-07-01

This paper studied the adequacy of the World and China lithium resources, considering the most promising uses in the future, involving nuclear fusion and electric-vehicles. The lithium recycle model for D-T fusion power plant and electric-vehicles, and the logistic growth prediction model of the primary energy for the World and China were constructed. Based on these models, preliminary evaluation of lithium resources adequacy of the World and China for D-T fusion reactors was presented under certain assumptions. Results show that: a. The world terrestrial reserves of lithium seems too limited to support a significant D-T power program, but the lithium reserves of China are relatively abundant, compared with the world case. b. The lithium resources contained in the oceans can be called the “permanent" energy. c. The change in 6Li enrichment has no obvious effect on the availability period of the lithium resources using FDS-II (Liquid Pb-17Li breeder blanket) type of reactors, but it has a stronger effect when PPCS-B (Solid Li4 SiO4 ceramics breeder blanket) is used.

Parameter study toward economical magnetic fusion power reactors

International Nuclear Information System (INIS)

Yoshida, Tomoaki; Okano, Kunihiko; Nanahara, Toshiya; Hatayama, Akiyoshi; Yamaji, Kenji; Takuma, Tadashi.

1996-01-01

Although the R and D of nuclear fusion reactors has made a steady progress as seen in ITER project, it has become of little doubt that fusion power reactors require hugeness and enormous amount of construction cost as well as surmounting the physics and engineering difficulties. Therefore, it is one of the essential issues to investigate the prospect of realizing fusion power reactors. In this report we investigated the effects of physics and engineering improvements on the economics of ITER-like steady state tokamak fusion reactors using our tokamak system and costing analysis code. With the results of this study, we considered what is the most significant factor for realizing economical competitive fusion reactors. The results show that with the conventional TF coil maximum field (12T), physics progress in β-value (or Troyon coefficient) has the most considerable effect on the reduction of fusion plant COE (Cost of Electricity) while the achievement of H factor = 2-3 and neutron wall load =∼5MW/m 2 is necessary. The results also show that with the improvement of TF coil maximum field, reactors with a high aspect ratio are economically advantageous because of low plasma current driving power while the improvement of current density in the conductors and yield strength of support structures is indispensable. (author)

The total quasi-steady-state approximation for complex enzyme reactions

DEFF Research Database (Denmark)

Pedersen, Morten Gram; Bersani, A. M.; Bersani, E.

2008-01-01

) approximation (or standard quasi-steady-state approximation (sQSSA)), which is valid when the enzyme concentration is sufficiently small. This condition is usually fulfilled for in vitro experiments, but often breaks down in vivo. The total QSSA (tQSSA), which is valid for a broader range of parameters covering...

Experiences in the D ampersand D of the EBWR reactor complex at Argonne National Laboratory

International Nuclear Information System (INIS)

Bhattacharyya, S.K.; Boing, L.E.; Fellhauer, C.R.

1995-02-01

EBWR went critical in Dec 1957 at 20 MW(t), was upgraded to 100 MW(t) operation. EBWR was shut down July 1967 and placed in dry lay-up. In 1986, the D ampersand D work was planned in 4 phases: final planning and preparations for D ampersand D, removal of reactor systems, removal of reactor vessel complex, and final decontamination and project closeout. Despite precautions, there was an uptake of 241 Am by D ampersand D workers following underwater plasma arc cutting within the pool; the cause was traced to an experimental 241 Pu foil (200 μg) that was lost in the mid-1960s in the reactor vessel. Several major lessons were learned from this episode, among which is the fact that research facilities often involve unusual experiments which may not be recorded. Safety analysis and review procedure for D ampersand D operations need to be carefully considered since they represent considerably different situations than reactor operations. EBWR is one of the very few cases of a prototypic reactor facility designed, operated, tested and now D ampersand D'd by one organization

Steady-state pattern electroretinogram and short-duration transient visual evoked potentials in glaucomatous and healthy eyes.

Science.gov (United States)

Amarasekera, Dilru C; Resende, Arthur F; Waisbourd, Michael; Puri, Sanjeev; Moster, Marlene R; Hark, Lisa A; Katz, L Jay; Fudemberg, Scott J; Mantravadi, Anand V

2018-01-01

This study evaluates two rapid electrophysiological glaucoma diagnostic tests that may add a functional perspective to glaucoma diagnosis. This study aimed to determine the ability of two office-based electrophysiological diagnostic tests, steady-state pattern electroretinogram and short-duration transient visual evoked potentials, to discern between glaucomatous and healthy eyes. This is a cross-sectional study in a hospital setting. Forty-one patients with glaucoma and 41 healthy volunteers participated in the study. Steady-state pattern electroretinogram and short-duration transient visual evoked potential testing was conducted in glaucomatous and healthy eyes. A 64-bar-size stimulus with both a low-contrast and high-contrast setting was used to compare steady-state pattern electroretinogram parameters in both groups. A low-contrast and high-contrast checkerboard stimulus was used to measure short-duration transient visual evoked potential parameters in both groups. Steady-state pattern electroretinogram parameters compared were MagnitudeD, MagnitudeD/Magnitude ratio, and the signal-to-noise ratio. Short-duration transient visual evoked potential parameters compared were amplitude and latency. MagnitudeD was significantly lower in glaucoma patients when using a low-contrast (P = 0.001) and high-contrast (P state pattern electroretinogram stimulus. MagnitudeD/Magnitude ratio and SNR were significantly lower in the glaucoma group when using a high-contrast 64-bar-size stimulus (P state pattern electroretinogram was effectively able to discern between glaucomatous and healthy eyes. Steady-state pattern electroretinogram may thus have a role as a clinically useful electrophysiological diagnostic tool. © 2017 Royal Australian and New Zealand College of Ophthalmologists.

Practical steady-state enzyme kinetics.

Science.gov (United States)

Lorsch, Jon R

2014-01-01

Enzymes are key components of most biological processes. Characterization of enzymes is therefore frequently required during the study of biological systems. Steady-state kinetics provides a simple and rapid means of assessing the substrate specificity of an enzyme. When combined with site-directed mutagenesis (see Site-Directed Mutagenesis), it can be used to probe the roles of particular amino acids in the enzyme in substrate recognition and catalysis. Effects of interaction partners and posttranslational modifications can also be assessed using steady-state kinetics. This overview explains the general principles of steady-state enzyme kinetics experiments in a practical, rather than theoretical, way. Any biochemistry textbook will have a section on the theory of Michaelis-Menten kinetics, including derivations of the relevant equations. No specific enzymatic assay is described here, although a method for monitoring product formation or substrate consumption over time (an assay) is required to perform the experiments described. © 2014 Elsevier Inc. All rights reserved.

Tore-Supra infrared thermography system, a real steady-state diagnostic

International Nuclear Information System (INIS)

Guilhem, D.; Bondil, J.L.; Bertrand, B.; Desgranges, C.; Lipa, M.; Messina, P.; Missirlian, M.; Portafaix, C.; Reichle, R.; Roche, H.; Saille, A.

2005-01-01

Tore-Supra Tokamak (I p = 1.5 MA, B t = 4 T) has been constructed with a steady-state magnetic field using super-conducting magnets and water-cooled plasma facing components (PFCs) for high-performance long pulse plasma discharges. When not actively cooled, plasma facing components can only accumulate a limited amount of energy since the temperature increases continuously during the discharge until radiation cooling equals the incoming heat flux. Such an environment is found in the JET Tokamak [JET Team, IAEA-CN-60/A1-3, Seville, 1994] and on TRIAM [M. Sakamoto, H. Nakashima, S. Kawasaki, A. Iyomasa, S.V. Kulkarni, M. Hasegawa, E. Jotaki, H. Zushi, K. Nakamura, K. Hanada, S. Itoh, Static and dynamic properties of wall recycling in TRIAM-1M, J. Nucl. Mater. 313-316 (2003) 519-523] [Y. Kamada, et al., Nucl. Fusion 3 (1999) 1845]. In Tore-Supra, the surface temperature of the actively cooled plasma facing components reach steady state within a second. We present here the Tore-Supra thermographic system, made of seven endoscope bodies equipped so far with eight infrared (IR) cameras. It has to be noted that this diagnostic is the first diagnostic to be actively cooled, as required for steady state. The main purpose of such a diagnostic is to prevent the plasma to damage the actively cooled plasma facing components (ACPFCs), which consist of the toroidal pumped limiter (TPL), 7 m 2 , and of five radio-frequency antennae, 1.5 m 2 each

Effects of governing parameters on steady-state inter-wrapper flow in an LMFBR

International Nuclear Information System (INIS)

Moriya, Shoichi

2001-01-01

Hydraulic experiments were performed using a 1/8th scale rectangular model, based on a Japanese demonstration fast breeder reactor design, in order to study fundamental characteristics of interwrapper flows occurring under steady state conditions in an LMFBR. The steady state interwrapper flow of which direction was downward in the center region and upward in the peripheral region of a core barrel was observed because of the radial static pressure gradient in the upper part of the core barrel, produced by a core blockage effect resulting from an above core structure with a perforated skirt. Thermal stratification phenomena were moreover observed in the interwrapper region, created by the hot steady state interwrapper flow from an upper plenum and the cold leakage flow through the separated plate of the core barrel. The thermal interface was generated in higher part of the core barrel when the core blockage effect was smaller and Richardson number and the leakage flow rate ratio were larger. Significant temperature fluctuations occurred in the peripheral region of the core barrel, when the difference between the interface elevations in the center and peripheral regions of the core barrel was enough large. (author)

Theory of minimum dissipation of energy for the steady state

International Nuclear Information System (INIS)

Chu, T.K.

1992-02-01

The magnetic configuration of an inductively driven steady-state plasma bounded by a surface (or two adjacent surfaces) on which B·n = 0 is force-free: ∇xB = 2αB, where α is a constant, in time and in space. α is the ratio of the Poynting flux to the magnetic helicity flux at the boundary. It is also the ratio of the dissipative rates of the magnetic energy to the magnetic helicity in the plasma. The spatial extent of the configuration is noninfinitesimal. This global constraint is a result of the requirement that, for a steady-state plasma, the rate of change of the vector potential, ∂A/∂t, is constant in time and uniform in space

Requirements for charged-particle reaction cross sections in the d-d, d-t, t-t, and d-3He fuel cycles

International Nuclear Information System (INIS)

Jarmie, N.

1986-12-01

This paper reviews the status of experimental data and data evaluations for charged-particle reactions of interest in fusion-reactor design. In particular, the 2 H(t,α)n, 2 H(d,p) 3 H, 2 H(d, 3 He)n, 3 H(t,α)nn and 3 He(d,p) 4 He reactions at low energies are studied. Other secondary reactions are considered. The conclusion is that such cross sections are well known for the near and medium term, and that no crucial experimental lack exists. There is a serious lack of standard evaluations of these reactions, which should be in an internationally acceptable format and easily accessible. Support for generating such evaluations should be given serious consideration

Steady-State Performance of Kalman Filter for DPLL

Institute of Scientific and Technical Information of China (English)

QIAN Yi; CUI Xiaowei; LU Mingquan; FENG Zhenming

2009-01-01

For certain system models, the structure of the Kalman filter is equivalent to a second-order vari-able gain digital phase-locked loop (DPLL). To apply the knowledge of DPLLs to the design of Kalman filters, this paper studies the steady-state performance of Kalman filters for these system models. The results show that the steady-state Kalman gain has the same form as the DPLL gain. An approximate simple form for the steady-state Kalman gain is used to derive an expression for the equivalent loop bandwidth of the Kalman filter as a function of the process and observation noise variances. These results can be used to analyze the steady-state performance of a Kalman filter with DPLL theory or to design a Kalman filter model with the same steady-state performance as a given DPLL.

Steady-state transport equation resolution by particle methods, and numerical results

International Nuclear Information System (INIS)

Mercier, B.

1985-10-01

A method to solve steady-state transport equation has been given. Principles of the method are given. The method is studied in two different cases; estimations given by the theory are compared to numerical results. Results got in 1-D (spherical geometry) and in 2-D (axisymmetric geometry) are given [fr

Development of a steady thermal-hydraulic analysis code for the China Advanced Research Reactor

Institute of Scientific and Technical Information of China (English)

TIAN Wenxi; QIU Suizheng; GUO Yun; SU Guanghui; JIA Dounan; LIU Tiancai; ZHANG Jianwei

2007-01-01

A multi-channel model steady-state thermalhydraulic analysis code was developed for the China Advanced Research Reactor (CARR). By simulating the whole reactor core, the detailed mass flow distribution in the core was obtained. The result shows that structure size plays the most important role in mass flow distribution, and the influence of core power could be neglected under singlephase flow. The temperature field of the fuel element under unsymmetrical cooling condition was also obtained, which is necessary for further study such as stress analysis, etc. Of the fuel element. At the same time, considering the hot channel effect including engineering factor and nuclear factor, calculation of the mean and hot channel was carried out and it is proved that all thermal-hydraulic parameters satisfy the "Safety design regulation of CARR".

Triple echo steady-state (TESS) relaxometry.

Science.gov (United States)

Heule, Rahel; Ganter, Carl; Bieri, Oliver

2014-01-01

Rapid imaging techniques have attracted increased interest for relaxometry, but none are perfect: they are prone to static (B0 ) and transmit (B1 ) field heterogeneities, and commonly biased by T2 /T1 . The purpose of this study is the development of a rapid T1 and T2 relaxometry method that is completely (T2 ) or partly (T1 ) bias-free. A new method is introduced to simultaneously quantify T1 and T2 within one single scan based on a triple echo steady-state (TESS) approach in combination with an iterative golden section search. TESS relaxometry is optimized and evaluated from simulations, in vitro studies, and in vivo experiments. It is found that relaxometry with TESS is not biased by T2 /T1 , insensitive to B0 heterogeneities, and, surprisingly, that TESS-T2 is not affected by B1 field errors. Consequently, excellent correspondence between TESS and reference spin echo data is observed for T2 in vitro at 1.5 T and in vivo at 3 T. TESS offers rapid T1 and T2 quantification within one single scan, and in particular B1 -insensitive T2 estimation. As a result, the new proposed method is of high interest for fast and reliable high-resolution T2 mapping, especially of the musculoskeletal system at high to ultra-high fields. Copyright © 2013 Wiley Periodicals, Inc.

The STAT7 Code for Statistical Propagation of Uncertainties In Steady-State Thermal Hydraulics Analysis of Plate-Fueled Reactors

Energy Technology Data Exchange (ETDEWEB)

Dunn, Floyd E. [Argonne National Lab. (ANL), Argonne, IL (United States); Hu, Lin-wen [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States). Nuclear Reactor Lab.; Wilson, Erik [Argonne National Lab. (ANL), Argonne, IL (United States)

2016-12-01

The STAT code was written to automate many of the steady-state thermal hydraulic safety calculations for the MIT research reactor, both for conversion of the reactor from high enrichment uranium fuel to low enrichment uranium fuel and for future fuel re-loads after the conversion. A Monte-Carlo statistical propagation approach is used to treat uncertainties in important parameters in the analysis. These safety calculations are ultimately intended to protect against high fuel plate temperatures due to critical heat flux or departure from nucleate boiling or onset of flow instability; but additional margin is obtained by basing the limiting safety settings on avoiding onset of nucleate boiling. STAT7 can simultaneously analyze all of the axial nodes of all of the fuel plates and all of the coolant channels for one stripe of a fuel element. The stripes run the length of the fuel, from the bottom to the top. Power splits are calculated for each axial node of each plate to determine how much of the power goes out each face of the plate. By running STAT7 multiple times, full core analysis has been performed by analyzing the margin to ONB for each axial node of each stripe of each plate of each element in the core.

WRAP: a water reactor analysis package

International Nuclear Information System (INIS)

Anderson, M.M.

1977-06-01

The modular computational system known as the Water Reactor Analysis Package (WRAP) has been developed at the Savannah River Laboratory. WRAP is essentially a reprogrammed version of the RELAP4 computer code with an extensively restructured input format, a dynamic dimensioning capability and additional computational capabilities such as an automatic steady-state option for pressurized water reactors and an automatic restart capability with provision for renodalization. The report describes the capabilities of WRAP at its current stage of development. The addition of new capabilities (e.g., a BWR steady-state capability), the inclusion of improved models (e.g., models in RELAP4/M0D8) and the development of improved numerical techniques to reduce execution time are being planned at this time

Steady-state thermal hydraulic analysis and flow channel blockage accident analysis of JRR-3 silicide core

International Nuclear Information System (INIS)

Kaminaga, Masanori

1997-03-01

JRR-3 is a light water moderated and cooled, beryllium and heavy water reflected pool type research reactor using low enriched uranium (LEU) plate-type fuels. Its thermal power is 20 MW. The core conversion program from uranium-aluminum (UAl x -Al) dispersion type fuel (aluminide fuel) to uranium-silicon-aluminum (U 3 Si 2 -Al) dispersion type fuel (silicide fuel) is currently conducted at the JRR-3. This report describes about the steady-state thermal hydraulic analysis results and the flow channel blockage accident analysis result. In JRR-3, there are two operation mode. One is high power operation mode up to 20 MW, under forced convection cooling using the primary and the secondary cooling systems. The other is low power operation mode up to 200 kW, under natural circulation cooling between the reactor core and the reactor pool without the primary and the secondary cooling systems. For the analysis of the flow channel blockage accident, COOLOD code was used. On the other hand, steady-state thermal hydraulic analysis for both of the high power operation mode under forced convection cooling and low power operation under natural convection cooling, COOLOD-N2 code was used. From steady-state thermal hydraulic analysis results of both forced and natural convection cooling, fuel temperature, minimum DNBR etc. meet the design criteria and JRR-3 LEU silicide core has enough safety margin under normal operation conditions. Furthermore, flow channel blockage accident analysis results show that one channel flow blockage accident meet the safety criteria for accident conditions which have been established for JRR-3 LEU silicide core. (author)

Recalculating the steady state conditions of the V-1000 zero-power facility at Kurchatov Institute using Monte Carlo and nodal diffusion codes

Energy Technology Data Exchange (ETDEWEB)

Sahlberg, Ville [VTT Technical Research Centre of Finland Ltd, VTT (Finland)

2017-09-15

Continuous-energy Monte Carlo reactor physics code Serpent 2 was used to model the critical steady state conditions measured in V-1000 zero-power critical facility at Kurchatov Institute (KI), Moscow in 1990-1992. The Serpent 2 results were compared to measurements and Serpent 2 was used to generate group constants for reactor dynamics code HEXTRAN. The results of a HEXTRAN calculation of the steady state were compared to Serpent 2. The relative power density distribution of the SERPENT2 calculations compared with the measurements was within the statistical accuracy. The comparison of HEXTRAN and Serpent 2 node-wise relative power density distributions showed an accuracy of ±10%.

MR fingerprinting using fast imaging with steady state precession (FISP) with spiral readout.

Science.gov (United States)

Jiang, Yun; Ma, Dan; Seiberlich, Nicole; Gulani, Vikas; Griswold, Mark A

2015-12-01

This study explores the possibility of using gradient echo-based sequences other than balanced steady-state free precession (bSSFP) in the magnetic resonance fingerprinting (MRF) framework to quantify the relaxation parameters . An MRF method based on a fast imaging with steady-state precession (FISP) sequence structure is presented. A dictionary containing possible signal evolutions with physiological range of T1 and T2 was created using the extended phase graph formalism according to the acquisition parameters. The proposed method was evaluated in a phantom and a human brain. T1 , T2 , and proton density were quantified directly from the undersampled data by the pattern recognition algorithm. T1 and T2 values from the phantom demonstrate that the results of MRF FISP are in good agreement with the traditional gold-standard methods. T1 and T2 values in brain are within the range of previously reported values. MRF-FISP enables a fast and accurate quantification of the relaxation parameters. It is immune to the banding artifact of bSSFP due to B0 inhomogeneities, which could improve the ability to use MRF for applications beyond brain imaging. © 2014 Wiley Periodicals, Inc.

A comparison of the steady-state facility in the RELAP-UK code with the CUSH code and with collant flow in the Winfrith SGHWR

International Nuclear Information System (INIS)

Roberts, H.A.; Smith, C.P.

1976-02-01

Provision of capability for performing steady-state calculations in RELAP-UK has led to the possibility of the wider use of this code for steady-state assessments of the behaviour of commercial systems with complicated circuit arrangements. In the studies discussed in this report, the first objective is to demonstrate the performance of the RELAP-UK code in a steady-state role, and to make comparisons with the CUSH code, and with measurements obtained on the Winfrith Steam-Generating Heavy Water Reactor. (U.K.)

New Tore Supra steady state operating scenario

International Nuclear Information System (INIS)

Martin, G.; Parlange, F.; van Houtte, D.; Wijnands, T.

1995-01-01

This document deals with plasma control in steady state conditions. A new plasma control systems enabling feedback control of global plasma equilibrium parameters has been developed. It also enables to operate plasma discharge in steady state regime. (TEC). 4 refs., 5 figs

HEXNOD23, 2-D, 3-D Coarse Mesh Solution of Steady State Diffusion Equation in Hexagonal Geometry

International Nuclear Information System (INIS)

Grundmann, Ulrich

1986-01-01

1 - Description of program or function: Two- or three dimensional coarse mesh solution of steady state two group neutron diffusion equation in arrays of regular hexagons or hexagonal subassemblies. 2 - Method of solution: The neutron flux in a hexagonal node is expanded in a series of Bessel functions in the hexagonal plane. Polynomials up to the 4. order are used for the approximation of neutron flux in axial direction of three dimensional cases. Resulting relations between node averaged fluxes and mean partial currents of node faces in connection with the neutron balance of nodes are used to calculate the eigenvalue Keff, mean fluxes and mean powers of nodes. The iterations process is divided into inner and outer iterations. The iterations are accelerated by Ljusternik and Tschebyscheff extrapolation schemes. The power densities in the nodes and subassembly powers are computed for given reactor power in three dimensional cases. 30 degree reflectional, 60 and 120 degree rotational core symmetry and the whole core can be treated. 3 - Restrictions on the complexity of the problem: If the problem size designated by LIAR and LRAR exceeds 3000 and 50000 respectively, the lengths of the working array MIAR and MRAR in the main program can be increased. External sources are not permitted

TRUMP, Steady-State and Transient 1-D, 2-D and 3-D Potential Flow, Temperature Distribution

International Nuclear Information System (INIS)

Elrod, D.C.; Turner, W.D.

1981-01-01

1 - Description of problem or function: TRUMP solves a general non- linear parabolic partial differential equation describing flow in various kinds of potential fields, such as fields of temperature, pressure, or electricity and magnetism; simultaneously, it will solve two additional equations representing, in thermal problems, heat production by decomposition of two reactants having rate constants with a general Arrhenius temperature dependence. Steady- state and transient flow in one, two, or three dimensions are considered in geometrical configurations having simple or complex shapes and structures. Problem parameters may vary with spatial position, time, or primary dependent variables--temperature, pressure, or field strength. Initial conditions may vary with spatial position, and among the criteria that may be specified for ending a problem are upper and lower limits on the size of the primary dependent variable, upper limits on the problem time or on the number of time-steps or on the computer time, and attainment of steady state. 2 - Method of solution: Solutions may be obtained by use of explicit- or implicit-difference equations, or by an optimized combination of both. 3 - Restrictions on the complexity of the problem: The program currently provides for maxima of: 40 materials, 5 reactants, 105 surface conditions, 20 boundary nodes, 16 entries per tabulated function (table-length)

Steady state neutral beam injector

International Nuclear Information System (INIS)

Mattoo, S.K.; Bandyopadhyay, M.; Baruah, U.K.; Bisai, N.; Chakbraborty, A.K.; Chakrapani, Ch.; Jana, M.R.; Bajpai, M.; Jaykumar, P.K.; Patel, D.; Patel, G.; Patel, P.J.; Prahlad, V.; Rao, N.V.M.; Rotti, C.; Singh, N.P.; Sridhar, B.

2000-01-01

Learning from operational reliability of neutral beam injectors in particular and various heating schemes including RF in general on TFTR, JET, JT-60, it has become clear that neutral beam injectors may find a greater role assigned to them for maintaining the plasma in steady state devices under construction. Many technological solutions, integrated in the present day generation of injectors have given rise to capability of producing multimegawatt power at many tens of kV. They have already operated for integrated time >10 5 S without deterioration in the performance. However, a new generation of injectors for steady state devices have to address to some basic issues. They stem from material erosion under particle bombardment, heat transfer > 10 MW/m 2 , frequent regeneration of cryopanels, inertial power supplies, data acquisition and control of large volume of data. Some of these engineering issues have been addressed to in the proposed neutral beam injector for SST-1 at our institute; the remaining shall have to wait for the inputs of the database generated from the actual experience with steady state injectors. (author)

Fusion performances and alpha heating in future JET D-T plasmas

Energy Technology Data Exchange (ETDEWEB)

Balet, B; Cordey, J G; Gibson, A; Lomas, P; Stubberfield, P M; Thomas, P [Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking

1994-07-01

The new pump divertor installed at JET should allow high performance pulses of a few seconds duration by both preventing the impurity influx and controlling the density evolution. The TRANSP code has been used in a predictive mode to assess the possible fusion performance of such plasmas fuelled with a 50:50 mixture of D and T, and the effect of alpha particles heating on Te and Ti. Several cases are considered: 50:50 D-T mix; 50:50 D-T mix, no C bloom; 50:50 D-T mix, VH phase, density control; 50:50 D-T mix, VH phase, density control, 6 Ma. The predictions show that if the the bloom and MHD instabilities can be controlled at higher plasma currents using a higher toroidal field to keep a reasonable beta value, then a higher fusion performance steady state plasma with Q{sub DT} superior to 2.5 should be possible. The alpha heating power of 4.9 MW would lead to a 74% increase in Te. 4 refs., 4 figs., 1 tab.

Measurement of non-steady-state free fatty acid turnover

International Nuclear Information System (INIS)

Jensen, M.D.; Heiling, V.; Miles, J.M.

1990-01-01

The accuracy of non-steady-state equations for measuring changes in free fatty acid rate of appearance (Ra) is unknown. In the present study, endogenous lipolysis (traced with [ 14 C]-linoleate) was pharmacologically suppressed in six conscious mongrel dogs. A computer-responsive infusion pump was then used to deliver an intravenous oleic acid emulsion in both constant and linear gradient infusion modes. Both non-steady-state equations with various effective volumes of distribution (V) and steady-state equations were used to measure oleate Ra [( 14 C]oleate). Endogenous lipolysis did not change during the experiment. When oleate Ra increased in a linear gradient fashion, only non-steady-state equations with a large (150 ml/kg) V resulted in erroneous values (9% overestimate, P less than 0.05). In contrast, when oleate Ra decreased in a similar fashion, steady-state and standard non-steady-state equations (V = plasma volume = 50 ml/kg) overestimated total oleate Ra (18 and 7%, P less than 0.001 and P less than 0.05, respectively). Overall, non-steady-state equations with an effective V of 90 ml/kg (1.8 x plasma volume) allowed the most accurate estimates of oleate Ra

Steady-state operation of spheromaks by inductive techniques

International Nuclear Information System (INIS)

Janos, A.

1984-04-01

A method to maintain a steady-state spheromak configuration inductively using the S-1 Spheromak device is described. The S-1 Spheromak formation apparatus can be utilized to inject magnetic helicity continuously (C.W., not pulsed or D.C.) into the spheromak configuration after equilibrium is achieved in the linked mode of operation. Oscillation of both poloidal- and toroidal-field currents in the flux core (psi-phi Pumping), with proper phasing, injects a net time-averaged helicity into the plasma. Steady-state maintenance relies on flux conversion, which has been earlier identified. Relevant experimental data from the operation of S-1 are described. Helicity flow has been measured and the proposed injection scheme simulated. In a reasonable time practical voltages and frequencies can inject an amount of helicity comparable to that in the initial plasma. Plasma currents can be maintained or increased. This pumping technique is similar to F-THETA Pumping of a Reversed-Field-Pinch but is applied to this inverse-pinch formation

Steady-state and transient hydrocarbon production in graphite by low energy impact of atomic and molecular deuterium projectiles

International Nuclear Information System (INIS)

Zhang, H.; Meyer, F.W.

2009-01-01

We report measurements of steady-state yields of methyl, methane and heavier hydrocarbons for deuterium atomic and molecular ions incident on ATJ graphite, HOPG, and a-C:D thin films in the energy range 10-200 eV/D. The yields were determined using a QMS technique in conjunction with calibrated hydrocarbon leaks. We have also studied transient hydrocarbon production and hydrogen (deuterium) re-emission for 80 and 150 eV/D D + , D 2 + , and D 3 + projectiles incident on ATJ graphite surfaces pre-loaded to steady state by 20 eV/D beams of the corresponding species. Immediately after starting the higher-energy beams, transient hydrocarbon and D 2 re-emission yields significantly larger than steady-state values were observed, which exponentially decayed as a function of beam fluence. The initial yield values were related to the starting hydrocarbon and deuterium densities in the prepared sample, while the exponential decay constants provided information on the hydrocarbon kinetic release and hydrogen (deuterium) detrapping cross-sections.

Tritium contamination experience in an operational D-T fusion reactor

International Nuclear Information System (INIS)

Gentile, C.A.; Ascione, G.

1994-01-01

During December 1993, the Tokamak Fusion Test Reactor (TFTR) injected a mixture of deuterium and tritium in the TFTR vacuum vessel for the purpose of creating D-T plasmas. The tritium used in these D-T plasmas was stored, delivered and processed in the TFTR tritium facility that includes the tritium vault, waste handling area, clean-up area, and gas holding tank room. During this time period, several components in the tritium process system were found to have tritium leaks which led to tritium deposition on process skids, components and floor area. Radiological surveys of surfaces contaminated with tritium oxide indicate a decrease in surface contamination in time (on the order of 12 to 36 hours) as the result of room ventilation. In instances where the facility HVAC system was maintained in the purge mode, a dramatic decrease in surface contamination was observed. Areas contaminated with tritium oxide (> 16.6 Bq/100 cm 2 ) were found to be clean ( 2 ) after several hours of continuous purging by the facility HVAC system. In instances where relative humidity was not decreased, the tritium surface contamination was found to be attenuated. During the months of December 1993, January and February 1994 tritium leaking components were either replaced, redesigned or repaired. During this time period, data were collected in the form of contamination surveys, real time tritium monitor output, and HVAC configuration indicating the correlation of purge ventilation leading to a decrease in tritium oxide surface contamination

Experimental Observations of Natural Circulation Flow in the NSTF at Steady-State Conditions

International Nuclear Information System (INIS)

Lisowski, Darius D.; Farmer, Mitch T.

2014-01-01

A ½ scale test facility has been constructed at Argonne National Laboratory (ANL) to study the heat removal performance and natural circulation flow patterns in a Reactor Cavity Cooling System (RCCS). Our test facility, the Natural convection Shutdown heat removal Test Facility (NSTF), supports the broader goal of developing an inherently safe and fully passive ex-vessel decay heat removal for advanced reactor designs. The project, initiated in 2010 to support the Advanced Reactor Concepts (ARC), Small Modular Reactor (SMR), and Next Generation Nuclear Plant (NGNP) programs, has been conducting experimental operations since early 2014. The following paper provides a summary of some primary design features of the 26-m tall test facility along with a description of the data acquisition suite that guides our experimental practices. Specifics of the distributed fiber optic temperature measurements will be discussed, which introduces an unparalleled level of data density that has never before been implemented in a large scale natural circulation test facility. Results from our first test series will then be presented, which provide insight into the thermal hydraulic behavior at steady-state conditions for varying heat flux levels and exhaust chimney configuration states. (author)

RelB+ Steady-State Migratory Dendritic Cells Control the Peripheral Pool of the Natural Foxp3+ Regulatory T Cells

Directory of Open Access Journals (Sweden)

Anja Döhler

2017-06-01

Full Text Available Thymus-derived natural Foxp3+ CD4+ regulatory T cells (nTregs play a key role in maintaining immune tolerance and preventing autoimmune disease. Several studies indicate that dendritic cells (DCs are critically involved in the maintenance and proliferation of nTregs. However, the mechanisms how DCs manage to keep the peripheral pool at constant levels remain poorly understood. Here, we describe that the NF-κB/Rel family transcription factor RelB controls the frequencies of steady-state migratory DCs (ssmDCs in peripheral lymph nodes and their numbers control peripheral nTreg homeostasis. DC-specific RelB depletion was investigated in CD11c-Cre × RelBfl/fl mice (RelBDCko, which showed normal frequencies of resident DCs in lymph nodes and spleen while the subsets of CD103− Langerin− dermal DCs (dDCs and Langerhans cells but not CD103+ Langerin+ dDC of the ssmDCs in skin-draining lymph nodes were increased. Enhanced frequencies and proliferation rates were also observed for nTregs and a small population of CD4+ CD44high CD25low memory-like T cells (Tml. Interestingly, only the Tml but not DCs showed an increase in IL-2-producing capacity in lymph nodes of RelBDCko mice. Blocking of IL-2 in vivo reduced the frequency of nTregs but increased the Tml frequencies, followed by a recovery of nTregs. Taken together, by employing RelBDCko mice with increased frequencies of ssmDCs our data indicate a critical role for specific ssmDC subsets for the peripheral nTreg and IL-2+ Tml frequencies during homeostasis.

Review of tokamak power reactor and blanket designs in the United States

International Nuclear Information System (INIS)

Baker, C.; Brooks, J.; Ehst, D.; Gohar, Y.; Smith, D.; Sze, D.

1986-01-01

The last major conceptual design study of a tokamak power reactor in the United States was STARFIRE which was carried out in 1979-1980. Since that time US studies have concentrated on engineering test reactors, demonstration reactors, parametric systems studies, scoping studies, and studies of selected critical issues such as pulsed vs. steady-state operation and blanket requirements. During this period, there have been many advancements in tokamak physics and reactor technology, and there has also been a recognition that it is desirable to improve the tokamak concept as a commercial power reactor candidate. During 1984-1985 several organizations participated in the Tokamak Power Systems Study (TPSS) with the objective of developing ideas for improving the tokamak as a power reactor. Also, the US completed a comprehensive Blanket Comparison and Selection Study which formed the basis for further studies on improved blankets for fusion reactors

3D double-echo steady-state sequence assessment of hip joint cartilage and labrum at 3 Tesla: comparative analysis of magnetic resonance imaging and intraoperative data

Energy Technology Data Exchange (ETDEWEB)

Schleich, Christoph; Antoch, Gerald [University of Dusseldorf, Department of Diagnostic and Interventional Radiology, Medical Faculty, Duesseldorf (Germany); Hesper, Tobias; Rettegi, Fanni; Zilkens, Christoph; Krauspe, Ruediger; Bittersohl, Bernd [University of Dusseldorf, Department of Orthopedic Surgery, Medical Faculty, Duesseldorf (Germany); Hosalkar, Harish S. [Paradise Valley Hospital, Joint Preservation and Deformity Correction, San Diego, CA (United States); Tri-city Medical Center, Hip Preservation, San Diego, CA (United States)

2017-10-15

To assess the diagnostic accuracy of a high-resolution, three-dimensional (3D) double-echo steady-state (DESS) sequence with radial imaging at 3 Tesla (T) for evaluating cartilage and labral alterations in the hip. Magnetic resonance imaging (MRI) data obtained at 3 T, including radially reformatted DESS images and intraoperative data of 45 patients (mean age 42 ± 13.7 years) who underwent hip arthroscopy, were compared. The acetabular cartilage and labrum of the upper hemisphere of the acetabulum and the central femoral head cartilage were evaluated. Sensitivity, specificity, accuracy, and negative and positive predictive values were determined. Sensitivity, specificity and accuracy of the DESS technique were 96.7%, 75% and 93.7% for detecting cartilage lesions and 98%, 76.2% and 95.9% for detecting labral lesions. The positive and negative predictive values for detecting or ruling out cartilage lesions were 96% and 78.9%. For labral lesions, the positive and negative predictive values were 97.5% and 80%. A high-resolution, 3D DESS technique with radial imaging at 3 T demonstrated high accuracy for detecting hip cartilage and labral lesions with excellent interobserver agreement and moderate correlation between MRI and intraoperative assessment. (orig.)

Conceptual designs of tokamak reactor and R D

International Nuclear Information System (INIS)

Fukai, Yuzo; Yamato, Harumi; Sawada, Yoshio

1983-01-01

The conceptual design of both FER (Fusion Experimental Reactor) and R-project is now under way as the new step of JT-60. From the engineering viewpoint, these reactors, requiring D-T operation, have the challenge, such as the handling of tritium and components irradiated by neutron bombardment. Toshiba's design team is participating to these projects in order to realize the reactor and plant concept coping with the above objectives. This paper represents the conceptual design contributions of the FER and R-project as well as R D technology which are now under development, such as tritium handling app aratus, reactor materials, etc. (author)

Efficient decoding with steady-state Kalman filter in neural interface systems.

Science.gov (United States)

Malik, Wasim Q; Truccolo, Wilson; Brown, Emery N; Hochberg, Leigh R

2011-02-01

The Kalman filter is commonly used in neural interface systems to decode neural activity and estimate the desired movement kinematics. We analyze a low-complexity Kalman filter implementation in which the filter gain is approximated by its steady-state form, computed offline before real-time decoding commences. We evaluate its performance using human motor cortical spike train data obtained from an intracortical recording array as part of an ongoing pilot clinical trial. We demonstrate that the standard Kalman filter gain converges to within 95% of the steady-state filter gain in 1.5±0.5 s (mean ±s.d.). The difference in the intended movement velocity decoded by the two filters vanishes within 5 s, with a correlation coefficient of 0.99 between the two decoded velocities over the session length. We also find that the steady-state Kalman filter reduces the computational load (algorithm execution time) for decoding the firing rates of 25±3 single units by a factor of 7.0±0.9. We expect that the gain in computational efficiency will be much higher in systems with larger neural ensembles. The steady-state filter can thus provide substantial runtime efficiency at little cost in terms of estimation accuracy. This far more efficient neural decoding approach will facilitate the practical implementation of future large-dimensional, multisignal neural interface systems.

Implications of steady-state operation on divertor design

International Nuclear Information System (INIS)

Sevier, D.L.; Reis, E.E.; Baxi, C.B.; Silke, G.W.; Wong, C.P.C.; Hill, D.N.

1996-01-01

As fusion experiments progress towards long pulse or steady state operation, plasma facing components are undergoing a significant change in their design. This change represents the transition from inertially cooled pulsed systems to steady state designs of significant power handling capacity. A limited number of Plasma Facing Component (PFC) systems are in operation or planning to address this steady state challenge at low heat flux. However in most divertor designs components are required to operate at heat fluxes at 5 MW/m 2 or above. The need for data in this area has resulted in a significant amount of thermal/hydraulic and thermal fatigue testing being done on prototypical elements. Short pulse design solutions are not adequate for longer pulse experiments and the areas of thermal design, structural design, material selection, maintainability, and lifetime prediction are undergoing significant changes. A prudent engineering approach will guide us through the transitional phase of divertor design to steady-state power plant components. This paper reviews the design implications in this transition to steady state machines and the status of the community efforts to meet evolving design requirements. 54 refs., 5 figs., 2 tabs

A twin study of the trough plasma steady-state concentration of metformin

DEFF Research Database (Denmark)

Stage, Tore B; Damkier, Per; Pedersen, Rasmus S

2015-01-01

OBJECTIVE: The aim of this study was to determine the intrapair similarity in trough steady-state plasma concentrations of metformin in monozygotic and dizygotic twin pairs. METHODS: We included 16 twin pairs (eight monozygotic and eight dizygotic twin pairs) for this study after contacting 524 t...

Tritium contamination experience in an operational D-T fusion reactor

Energy Technology Data Exchange (ETDEWEB)

Gentile, C.A.; Ascione, G. [Princeton Univ., NJ (United States). Plasma Physics Lab.; Anderson, J.L. [Los Alamos National Lab., NM (United States)] [and others

1994-09-01

During December 1993, the Tokamak Fusion Test Reactor (TFTR) injected a mixture of deuterium and tritium in the TFTR vacuum vessel for the purpose of creating D-T plasmas. The tritium used in these D-T plasmas was stored, delivered and processed in the TFTR tritium facility that includes the tritium vault, waste handling area, clean-up area, and gas holding tank room. During this time period, several components in the tritium process system were found to have tritium leaks which led to tritium deposition on process skids, components and floor area. Radiological surveys of surfaces contaminated with tritium oxide indicate a decrease in surface contamination in time (on the order of 12 to 36 hours) as the result of room ventilation. In instances where the facility HVAC system was maintained in the purge mode, a dramatic decrease in surface contamination was observed. Areas contaminated with tritium oxide (> 16.6 Bq/100 cm{sup 2}) were found to be clean (< 16.6 Bq/100 cm{sub 2}) after several hours of continuous purging by the facility HVAC system. In instances where relative humidity was not decreased, the tritium surface contamination was found to be attenuated. During the months of December 1993, January and February 1994 tritium leaking components were either replaced, redesigned or repaired. During this time period, data were collected in the form of contamination surveys, real time tritium monitor output, and HVAC configuration indicating the correlation of purge ventilation leading to a decrease in tritium oxide surface contamination.

Single-channel model for steady thermal-hydraulic analysis in nuclear reactor

International Nuclear Information System (INIS)

Zhang Xiaoying; Huang Yuanyuan

2010-01-01

This article established a single-channel model for steady analysis in the reactor and an example of thermal-hydraulic analysis was made by using this model, including the Maximum heat flux density of fuel element, enthalpy, Coolant flow, various kinds of pressure drop, enthalpy increase in average tube and thermal tube. I also got the Coolant temperature distribution and the fuel element temperature distribution and analysis of the final result. The results show that some relevant parameters which we got in this paper are well coincide with the actual operating parameters. It is also show that the single-channel model can be used to the steady thermal-hydraulic analysis. (authors)

Steady-State Process Modelling

DEFF Research Database (Denmark)

Cameron, Ian; Gani, Rafiqul

2011-01-01

illustrate the “equation oriented” approach as well as the “sequential modular” approach to solving complex flowsheets for steady state applications. The applications include the Williams-Otto plant, the hydrodealkylation (HDA) of toluene, conversion of ethylene to ethanol and a bio-ethanol process....

Steady-state coupled transport of HNO3 through a hollow-fiber supported liquid membrane

International Nuclear Information System (INIS)

Noble, R.D.; Danesi, P.R.

1987-01-01

Nitric acid removal from an aqueous stream was accomplished by continuously passing the fluid through a hollow fiber supported liquid membrane (SLM). The nitric acid was extracted through the membrane wall by coupled transport. The system was modeled as a series of (SLM)-continuous stirred tank reactor (CSTR) pairs. An approximate technique was used to predict the steady state nitric acid concentration in the system. The comparison with experimental data was very good

Comparison between steady-state and dynamic I-V measurements from a single-cell thermionic fuel element

International Nuclear Information System (INIS)

Wernsman, Bernard

1997-01-01

A comparison between steady-state and dynamic I-V measurements from a single-cell thermionic fuel element (TFE) is made. The single-cell TFE used in this study is the prototype for the 40 kW e space nuclear power system that is similar to the 6 kW e TOPAZ-II. The steady-state I-V measurements influence the emitter temperature due to electron cooling. Therefore, to eliminate the steady-state I-V measurement influence on the TFE and provide a better understanding of the behavior of the thermionic energy converter and TFE characteristics, dynamic I-V measurements are made. The dynamic I-V measurements are made at various input power levels, cesium pressures, collector temperatures, and steady-state current levels. From these measurements, it is shown that the dynamic I-V's do not change the TFE characteristics at a given operating point. Also, the evaluation of the collector work function from the dynamic I-V measurements shows that the collector optimization is not due to a minimum in the collector work function but due to an emission optimization. Since the dynamic I-V measurements do not influence the TFE characteristics, it is believed that these measurements can be done at a system level to understand the influence of TFE placement in the reactor as a function of the core thermal distribution

Steady-State Ion Beam Modeling with MICHELLE

Science.gov (United States)

Petillo, John

2003-10-01

There is a need to efficiently model ion beam physics for ion implantation, chemical vapor deposition, and ion thrusters. Common to all is the need for three-dimensional (3D) simulation of volumetric ion sources, ion acceleration, and optics, with the ability to model charge exchange of the ion beam with a background neutral gas. The two pieces of physics stand out as significant are the modeling of the volumetric source and charge exchange. In the MICHELLE code, the method for modeling the plasma sheath in ion sources assumes that the electron distribution function is a Maxwellian function of electrostatic potential over electron temperature. Charge exchange is the process by which a neutral background gas with a "fast" charged particle streaming through exchanges its electron with the charged particle. An efficient method for capturing this is essential, and the model presented is based on semi-empirical collision cross section functions. This appears to be the first steady-state 3D algorithm of its type to contain multiple generations of charge exchange, work with multiple species and multiple charge state beam/source particles simultaneously, take into account the self-consistent space charge effects, and track the subsequent fast neutral particles. The solution used by MICHELLE is to combine finite element analysis with particle-in-cell (PIC) methods. The basic physics model is based on the equilibrium steady-state application of the electrostatic particle-in-cell (PIC) approximation employing a conformal computational mesh. The foundation stems from the same basic model introduced in codes such as EGUN. Here, Poisson's equation is used to self-consistently include the effects of space charge on the fields, and the relativistic Lorentz equation is used to integrate the particle trajectories through those fields. The presentation will consider the complexity of modeling ion thrusters.

Coal conversion rate in 1t/d PSU liquefaction reactor; 1t/d PSU ekika hannoto ni okeru sekitan tenka sokudo no kento

Energy Technology Data Exchange (ETDEWEB)

Ikeda, K.; Imada, K. [Nippon Steel Corp., Tokyo (Japan); Nogami, Y.; Inokuchi, K. [Mitsui SRC Development Co. Ltd., Tokyo (Japan)

1996-10-28

To investigate the coal liquefaction characteristics, coal slurry samples were taken from the outlets of the reactors and slurry preheater of NEDOL process 1 t/d process supporting unit (PSU), and were analyzed. Tanito Harum coal was used for liquefaction, and the slurry was prepared with recycle solvent. Liquefaction was performed using synthetic iron sulfide catalyst at reaction temperatures, 450 and 465{degree}C. Solubility of various solid samples was examined against n-hexane, toluene, and tetrahydrofuran (THF). When considering the decrease of IMO (THF-insoluble and ash) as a characteristic of coal conversion reaction, around 20% at the outlet of the slurry preheater, around 70% within the first reactor, and several percents within the successive second and third reactors were converted against supplied coal. Increase of reaction temperature led to the increase of evaporation of oil fraction, which resulted in the decrease of actual slurry flow rate and in the increase of residence time. Thus, the conversion of coal was accelerated by the synergetic effect of temperature and time. Reaction rate constant of the coal liquefaction was around 2{times}10{sup -1} [min{sup -1}], which increased slightly with increasing the reaction temperature from 450 to 465{degree}C. 3 refs., 5 figs., 1 tab.

A model for steady-state and transient determination of subcooled boiling for calculations coupling a thermohydraulic and a neutron physics calculation program for reactor core calculation

International Nuclear Information System (INIS)

Mueller, R.G.

1987-06-01

Due to the strong influence of vapour bubbles on the nuclear chain reaction, an exact calculation of neutron physics and thermal hydraulics in light water reactors requires consideration of subcooled boiling. To this purpose, in the present study a dynamic model is derived from the time-dependent conservation equations. It contains new methods for the time-dependent determination of evaporation and condensation heat flow and for the heat transfer coefficient in subcooled boiling. Furthermore, it enables the complete two-phase flow region to be treated in a consistent manner. The calculation model was verified using measured data of experiments covering a wide range of thermodynamic boundary conditions. In all cases very good agreement was reached. The results from the coupling of the new calculation model with a neutron kinetics program proved its suitability for the steady-state and transient calculation of reactor cores. (orig.) [de

Steady state compact toroidal plasma production

Science.gov (United States)

Turner, William C.

1986-01-01

Apparatus and method for maintaining steady state compact toroidal plasmas. A compact toroidal plasma is formed by a magnetized coaxial plasma gun and held in close proximity to the gun electrodes by applied magnetic fields or magnetic fields produced by image currents in conducting walls. Voltage supply means maintains a constant potential across the electrodes producing an increasing magnetic helicity which drives the plasma away from a minimum energy state. The plasma globally relaxes to a new minimum energy state, conserving helicity according to Taylor's relaxation hypothesis, and injecting net helicity into the core of the compact toroidal plasma. Controlling the voltage so as to inject net helicity at a predetermined rate based on dissipative processes maintains or increases the compact toroidal plasma in a time averaged steady state mode.

Dynamic stabilization of D—T burn in Tokamak reactors

Institute of Scientific and Technical Information of China (English)

ShiBing－Ren; LongYong－Xing

1997-01-01

A simple,engineeringly feasible dynamic method is supposed to control the deuterium-tritium burn process in Tokamak reactors operated in an advanced scenario.The thermal transport of the D-T plasma is described by an anomalous thermal conduction which is a radially increasing function and the central conduction value is proportional to the central temperature of the plasma.The dynamic external heating power is selected to be inversely proportional to certain power function of this temperature,As a result,the D-T burn can undergo in controllable way in different temperature regimes with different power output.Anomalous alpha particle transport effect is taken into account.It can affect the resultant plasma equilibrium ,the reactor efficency,the operation mode and so on.

Steady State Shift Damage Localization

DEFF Research Database (Denmark)

Sekjær, Claus; Bull, Thomas; Markvart, Morten Kusk

2017-01-01

The steady state shift damage localization (S3DL) method localizes structural deterioration, manifested as either a mass or stiffness perturbation, by interrogating the damage-induced change in the steady state vibration response with damage patterns cast from a theoretical model. Damage is, thus...... the required accuracy when examining complex structures, an extensive amount of degrees of freedom (DOF) must often be utilized. Since the interrogation matrix for each damage pattern depends on the size of the system matrices constituting the FE-model, the computational time quickly becomes of first......-order importance. The present paper investigates two sub-structuring approaches, in which the idea is to employ Craig-Bampton super-elements to reduce the amount of interrogation distributions while still providing an acceptable localization resolution. The first approach operates on a strict super-element level...

The steady-state ECRH-system at Wendelstein7-X

International Nuclear Information System (INIS)

Laqua, H.P.; Erckmann, V.; Brakel, R.; Braune, H.; Maassberg, H.; Marushchenko, N.; Michel, G.; Turkin, Y.; Ullrich, S.; Dammertz, G.; Thumm, M.; Brand, P.; Gantenbein, G.; Kasparek, W.

2005-01-01

Electron Cyclotron Resonance Heating (ECRH) is the main heating system for the Wendelstein7-X (W7-X) stellarator and the only one for CW-operation in the first stage. The mission of W7-X, which is presently under construction at IPP-Greifswald, is to demonstrate the inherent steady state capability of stellarators at reactor relevant plasma parameters. A modular 10 MW ECRH plant at 140 GHz with 1 MW CW-capability power for each module is under construction to meet the scientific objectives. Simulations of different ECRH scenarios, which are foreseen for W7-X operation and base on ray- tracing calculations and confinement studies, will be presented. A steady state ECRH has specific requirements on the stellarator machine itself, on the ECRH-sources, transmissions elements and on the experimental environment. In particular all elements have to be sufficiently cooled, screened and armoured against microwaves. The commissioning of the ECRH plant is well under way, the strategy and status of the project will be reported. First full power, CW integral tests of one ECRH module have been performed. A large microwave stray radiation chamber for integrated in-vessel component tests had been brought into operation. A bi-axially movable, motor driven ECRH antenna mock-up was build and is tested for reliability now. A strategy for the commissioning and the first experimental campaign at W7-X has been developed. (author)

Quantum thermodynamics of nanoscale steady states far from equilibrium

Science.gov (United States)

Taniguchi, Nobuhiko

2018-04-01

We develop an exact quantum thermodynamic description for a noninteracting nanoscale steady state that couples strongly with multiple reservoirs. We demonstrate that there exists a steady-state extension of the thermodynamic function that correctly accounts for the multiterminal Landauer-Büttiker formula of quantum transport of charge, energy, or heat via the nonequilibrium thermodynamic relations. Its explicit form is obtained for a single bosonic or fermionic level in the wide-band limit, and corresponding thermodynamic forces (affinities) are identified. Nonlinear generalization of the Onsager reciprocity relations are derived. We suggest that the steady-state thermodynamic function is also capable of characterizing the heat current fluctuations of the critical transport where the thermal fluctuations dominate. Also, the suggested nonequilibrium steady-state thermodynamic relations seemingly persist for a spin-degenerate single level with local interaction.

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

International Nuclear Information System (INIS)

Puragliesi, R.; Zhou, L.; Zerkak, O.; Pautz, A.

2016-01-01

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

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

Steady states in conformal theories

CERN Multimedia

CERN. Geneva

2015-01-01

A novel conjecture regarding the steady state behavior of conformal field theories placed between two heat baths will be presented. Some verification of the conjecture will be provided in the context of fluid dynamics and holography.

Pu recycling in a full Th-MOX PWR core. Part I: Steady state analysis

International Nuclear Information System (INIS)

Fridman, E.; Kliem, S.

2011-01-01

Research highlights: → Detailed 3D 100% Th-MOX PWR core design is developed. → Pu incineration increased by a factor of 2 as compared to a full MOX PWR core. → The core controllability under steady state conditions is demonstrated. - Abstract: Current practice of Pu recycling in existing Light Water Reactors (LWRs) in the form of U-Pu mixed oxide fuel (MOX) is not efficient due to continuous Pu production from U-238. The use of Th-Pu mixed oxide (TOX) fuel will considerably improve Pu consumption rates because virtually no new Pu is generated from thorium. In this study, the feasibility of Pu recycling in a typical pressurized water reactor (PWR) fully loaded with TOX fuel is investigated. Detailed 3-dimensional 100% TOX and 100% MOX PWR core designs are developed. The full MOX core is considered for comparison purposes. The design stages included determination of Pu loading required to achieve 18-month fuel cycle assuming three-batch fuel management scheme, selection of poison materials, development of the core loading pattern, optimization of burnable poison loadings, evaluation of critical boron concentration requirements, estimation of reactivity coefficients, core kinetic parameters, and shutdown margin. The performance of the MOX and TOX cores under steady-state condition and during selected reactivity initiated accidents (RIAs) is compared with that of the actual uranium oxide (UOX) PWR core. Part I of this paper describes the full TOX and MOX PWR core designs and reports the results of steady state analysis. The TOX core requires a slightly higher initial Pu loading than the MOX core to achieve the target fuel cycle length. However, the TOX core exhibits superior Pu incineration capabilities. The significantly degraded worth of control materials in Pu cores is partially addressed by the use of enriched soluble boron and B 4 C as a control rod absorbing material. Wet annular burnable absorber (WABA) rods are used to flatten radial power distribution

Validation of Contrast Enhanced Cine Steady-State Free Precession and T2-Weigthed CMR for Assessment of Ischemic Myocardial Area- At-Risk

DEFF Research Database (Denmark)

Søvsø Szocska Hansen, Esben; Pedersen, Steen Fjord; Pedersen, Steen Bønløkke

2017-01-01

-CINE) has recently been used to quantify AAR and validated against myocardial perfusion SPECT. In this study we sought to determine how well T2-STIR and CE-CINE depicts AAR in an experimental porcine model of myocardial ischemia-reperfusion injury using histopathology as the reference for infarct size......Measuring myocardial salvage is important to evaluate the possible cardioprotective effects of adjunctive cardioprotective intervention in patients with myocardial infarction undergoing primary percutaneous intervention. Contrast-enhanced steady-state free precession magnetic resonance imaging (CE...

Steady- and transient-state analyses of fully ceramic microencapsulated fuel loaded reactor core via two-temperature homogenized thermal-conductivity model

International Nuclear Information System (INIS)

Lee, Yoonhee; Cho, Nam Zin

2015-01-01

Highlights: • Fully ceramic microencapsulated fuel-loaded core is analyzed via a two-temperature homogenized thermal-conductivity model. • The model is compared to harmonic- and volumetric-average thermal conductivity models. • The three thermal analysis models show ∼100 pcm differences in the k eff eigenvalue. • The three thermal analysis models show more than 70 K differences in the maximum temperature. • There occur more than 3 times differences in the maximum power for a control rod ejection accident. - Abstract: Fully ceramic microencapsulated (FCM) fuel, a type of accident-tolerant fuel (ATF), consists of TRISO particles randomly dispersed in a SiC matrix. In this study, for a thermal analysis of the FCM fuel with such a high heterogeneity, a two-temperature homogenized thermal-conductivity model was applied by the authors. This model provides separate temperatures for the fuel-kernels and the SiC matrix. It also provides more realistic temperature profiles than those of harmonic- and volumetric-average thermal conductivity models, which are used for thermal analysis of a fuel element in VHTRs having a composition similar to the FCM fuel, because such models are unable to provide the fuel-kernel and graphite matrix temperatures separately. In this study, coupled with a neutron diffusion model, a FCM fuel-loaded reactor core is analyzed via a two-temperature homogenized thermal-conductivity model at steady- and transient-states. The results are compared to those from harmonic- and volumetric-average thermal conductivity models, i.e., we compare k eff eigenvalues, power distributions, and temperature profiles in the hottest single-channel at steady-state. At transient-state, we compare total powers, reactivity, and maximum temperatures in the hottest single-channel obtained by the different thermal analysis models. The different thermal analysis models and the availability of fuel-kernel temperatures in the two-temperature homogenized thermal

Joint DIII-D/EAST research on the development of a high poloidal beta scenario for the steady state missions of ITER and CFETR

Science.gov (United States)

Garofalo, A. M.; Gong, X. Z.; Ding, S. Y.; Huang, J.; McClenaghan, J.; Pan, C. K.; Qian, J.; Ren, Q. L.; Staebler, G. M.; Chen, J.; Cui, L.; Grierson, B. A.; Hanson, J. M.; Holcomb, C. T.; Jian, X.; Li, G.; Li, M.; Pankin, A. Y.; Peysson, Y.; Zhai, X.; Bonoli, P.; Brower, D.; Ding, W. X.; Ferron, J. R.; Guo, W.; Lao, L. L.; Li, K.; Liu, H.; Lyv, B.; Xu, G.; Zang, Q.

2018-01-01

Experimental and modeling investigations on the DIII-D and EAST tokamaks show the attractive transport and stability properties of fully noninductive, high poloidal-beta (β P ) plasmas, and their suitability for steady-state operating scenarios in ITER and CFETR. A key feature of the high-β P regime is the large-radius (ρ > 0.6) internal transport barrier (ITB), often observed in all channels (ne, Te, Ti, rotation), and responsible for both excellent energy confinement quality and excellent stability properties. Experiments on DIII-D have shown that, with a large-radius ITB, very high β N and β P values (both ≥ 4) can be reached by taking advantage of the stabilizing effect of a nearby conducting wall. Synergistically, higher plasma pressure provides turbulence suppression by Shafranov shift, leading to ITB sustainment independent of the plasma rotation. Experiments on EAST have been used to assess the long pulse potential of the high-β P regime. Using RF-only heating and current drive, EAST achieved minute-long fully noninductive steady state H-mode operation with strike points on an ITER-like tungsten divertor. Improved confinement (relative to standard H-mode) and steady state ITB features are observed with a monotonic q-profile with q min ˜ 1.5. Separately, experiments have shown that increasing the density in plasmas driven by lower hybrid wave broadens the q-profile, a technique that could enable a large radius ITB. These experimental results have been used to validate MHD, current drive, and turbulent transport models, and to project the high-β P regime to a burning plasma. These projections suggest the Shafranov shift alone will not suffice to provide improved confinement (over standard H-mode) without rotation and rotation shear. However, increasing the negative magnetic shear (higher q on axis) provides a similar turbulence suppression mechanism to Shafranov shift, and can help devices such as ITER and CFETR achieve their steady-state fusion

A Novel ARX-Based Approach for the Steady-State Identification Analysis of Industrial Depropanizer Column Datasets

Directory of Open Access Journals (Sweden)

Franklin D. Rincón

2015-04-01

Full Text Available This paper introduces a novel steady-state identification (SSI method based on the auto-regressive model with exogenous inputs (ARX. This method allows the SSI with reduced tuning by analyzing the identifiability properties of the system. In particular, the singularity of the model matrices is used as an index for steady-state determination. In this contribution, the novel SSI method is compared to other available techniques, namely the F-like test, wavelet transform and a polynomial-based approach. These methods are implemented for SSI of three different case studies. In the first case, a simulated dataset is used for calibrating the output-based SSI methods. The second case corresponds to a literature nonlinear continuous stirred-tank reactor (CSTR example running at different steady states in which the ARX-based approach is tuned with the available input-output data. Finally, an industrial case with real data of a depropanizer column from PETROBRAS S.A. considering different pieces of equipment is analyzed. The results for a reflux drum case indicate that the wavelet and the F-like test can satisfactorily detect the steady-state periods after careful tuning and when respecting their hypothesis, i.e., smooth data for the wavelet method and the presence of variance in the data for the F-like test. Through a heat exchanger case with different measurement frequencies, we demonstrate the advantages of using the ARX-based method over the other techniques, which include the aspect of online implementation.

Development of steady-state scenarios compatible with ITER-like wall conditions

Energy Technology Data Exchange (ETDEWEB)

Litaudon, X [Association Euratom-CEA, CEA/DSM/DRFC-Cadarache 13108, St Paul Durance (France); Arnoux, G [Association Euratom-CEA, CEA/DSM/DRFC-Cadarache 13108, St Paul Durance (France); Beurskens, M [Euratom/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom)] (and others)

2007-12-15

A key issue for steady-state tokamak operation is to determine the edge conditions that are compatible both with good core confinement and with the power handling and plasma exhaust capabilities of the plasma facing components (PFCs) and divertor systems. A quantitative response to this open question will provide a robust scientific basis for reliable extrapolation of present regimes to an ITER compatible steady-state scenario. In this context, the JET programme addressing steady-state operation is focused on the development of non-inductive, high confinement plasmas with the constraints imposed by the PFCs. A new beryllium main chamber wall and tungsten divertor together with an upgrade of the heating/fuelling capability are currently in preparation at JET. Operation at higher power with this ITER-like wall will impose new constraints on non-inductive scenarios. Recent experiments have focused on the preparation for this new phase of JET operation. In this paper, progress in the development of advanced tokamak (AT) scenarios at JET is reviewed keeping this long-term objective in mind. The approach has consisted of addressing various critical issues separately during the 2006-2007 campaigns with a view to full scenario integration when the JET upgrades are complete. Regimes with internal transport barriers (ITBs) have been developed at q{sub 95} {approx} 5 and high triangularity, {delta} (relevant to the ITER steady-state demonstration) by applying more than 30 MW of additional heating power reaching {beta}{sub N} {approx} 2 at B{sub o} {approx} 3.1 T. Operating at higher {delta} has allowed the edge pedestal and core densities to be increased pushing the ion temperature closer to that of the electrons. Although not yet fully integrated into a performance enhancing ITB scenario, Neon seeding has been successfully explored to increase the radiated power fraction (up to 60%), providing significant reduction of target tile power fluxes (and hence temperatures) and

Differences between automatically detected and steady-state fractional flow reserve.

Science.gov (United States)

Härle, Tobias; Meyer, Sven; Vahldiek, Felix; Elsässer, Albrecht

2016-02-01

Measurement of fractional flow reserve (FFR) has become a standard diagnostic tool in the catheterization laboratory. FFR evaluation studies were based on pressure recordings during steady-state maximum hyperemia. Commercially available computer systems detect the lowest Pd/Pa ratio automatically, which might not always be measured during steady-state hyperemia. We sought to compare the automatically detected FFR and true steady-state FFR. Pressure measurement traces of 105 coronary lesions from 77 patients with intermediate coronary lesions or multivessel disease were reviewed. In all patients, hyperemia had been achieved by intravenous adenosine administration using a dosage of 140 µg/kg/min. In 42 lesions (40%) automatically detected FFR was lower than true steady-state FFR. Mean bias was 0.009 (standard deviation 0.015, limits of agreement -0.02, 0.037). In 4 lesions (3.8%) both methods lead to different treatment recommendations, in all 4 cases instantaneous wave-free ratio confirmed steady-state FFR. Automatically detected FFR was slightly lower than steady-state FFR in more than one-third of cases. Consequently, interpretation of automatically detected FFR values closely below the cutoff value requires special attention.

Modeling of Phenix End-of-Life control rod withdrawal benchmark with DYN3D SFR version

Energy Technology Data Exchange (ETDEWEB)

Nikitin, Evgeny; Fridman, Emil [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Reactor Safety

2017-06-01

The reactor dynamics code DYN3D is currently under extension for Sodium cooled Fast Reactor applications. The control rod withdrawal benchmark from the Phenix End-of-Life experiments was selected for verification and validation purposes. This report presents some selected results to demonstrate the feasibility of using DYN3D for steady-state Sodium cooled Fast Reactor analyses.

Steady-state and time-dependent modelling of parallel transport in the scrape-off layer

DEFF Research Database (Denmark)

Havlickova, E.; Fundamenski, W.; Naulin, Volker

2011-01-01

The one-dimensional fluid code SOLF1D has been used for modelling of plasma transport in the scrape-off layer (SOL) along magnetic field lines, both in steady state and under transient conditions that arise due to plasma turbulence. The presented work summarizes results of SOLF1D with attention...... given to transient parallel transport which reveals two distinct time scales due to the transport mechanisms of convection and diffusion. Time-dependent modelling combined with the effect of ballooning shows propagation of particles along the magnetic field line with Mach number up to M ≈ 1...... temperature calculated in SOLF1D is compared with the approximative model used in the turbulence code ESEL both for steady-state and turbulent SOL. Dynamics of the parallel transport are investigated for a simple transient event simulating the propagation of particles and energy to the targets from a blob...

Safety in the ARIES-III D-3He tokamak reactor design

International Nuclear Information System (INIS)

Herring, J.S.; Dolan, T.J.

1992-01-01

This paper reports on the ARIES-III reactor study, an extensive examination of the viability of a D- 3 He-fueled commercial tokamak powder reactor. Because neutrons are produced only through side reactions (D+D- 3 HE+N; and D+D-T+p followed by D+T- 4 He+n), the reactor has the significant advantages of reduced activation of the first wall and shield, low afterheat and Class A or C low level waste disposal. Since no tritium is required for operation, no lithium-containing breeding blanket is necessary. A ferritic steel shield behind the first wall protects the magnets from gamma and neutron heating and from radiation damage. The authors explored the potential for isotopically tailoring the 4 mm tungsten layer on the divertor in order to reduce the offsite doses should a tungsten aerosol be released from the reactor after an accident. The authors also modeled a loss-of-cooling accident (LOCA) in which the organic coolant was burning in order to estimate the amount of radionuclides released from the first wall. Because the maximum temperature is low, degree C, release fractions are small. The authors analyzed the disposition of the 20 g/day of tritium that is produced by D-D reactions and removed by the vacuum pumps

WWER reactor physics code applications

International Nuclear Information System (INIS)

Gado, J.; Kereszturi, A.; Gacs, A.; Telbisz, M.

1994-01-01

The coupled steady-state reactor physics and thermohydraulic code system KARATE has been developed and applied for WWER-1000 and WWER-440 operational calculations. The 3 D coupled kinetic code KIKO3D has been developed and validated for WWER-440 accident analysis applications. The coupled kinetic code SMARTA developed by VTT Helsinki has been applied for WWER-440 accident analysis. The paper gives a summary of the experience in code development and application. (authors). 10 refs., 2 tabs., 5 figs

Steady-State Serum T3 Concentrations for 48 Hours Following the Oral Administration of a Single Dose of 3,5,3'-Triiodothyronine Sulfate (T3S).

Science.gov (United States)

Santini, Ferruccio; Giannetti, Monica; Ricco, Ilaria; Querci, Giorgia; Saponati, Giorgio; Bokor, Daniela; Rivolta, Giovanni; Bussi, Simona; Braverman, Lewis E; Vitti, Paolo; Pinchera, Aldo

2014-07-01

absorbed following oral administration in hypothyroid humans; (2) after a single oral dose, T3S is converted to T3 in a dose-dependent manner, resulting in steady-state serum T3 concentrations for 48 hours; (3) T3S may represent a new agent in combination with T4 in the therapy of hypothyroidism, if similar conversion of T3S to T3 can be demonstrated in euthyroid patients who are already taking T4.

Application of RELAP5-3D code for thermal analysis of the ADS reactor core; Aplicação do código RELAP5-3D para análise térmica do núcleo de um reator ADS

Energy Technology Data Exchange (ETDEWEB)

Fernandes, Gustavo Henrique Nazareno

2018-04-01

Nuclear power is essential to supply global energy demand. Therefore, in order to use nuclear fuel more efficiently, more efficient nuclear reactors technologies researches have been intensified, such as hybrid systems, composed of particle accelerators coupled into nuclear reactors. In order to add knowledge to such studies, an innovative reactor design was considered where the RELAP5-3D thermal-hydraulic analysis code was used to perform a thermal analysis of the core, either in stationary operation or in situations transitory. The addition of new kind of coolants, such as, liquid salts, among them Flibe, lead, lead-bismuth, sodium, lithium-bismuth and lithium-lead was an important advance in this version of the code, making possible to do the thermal simulation of reactors that use these types of coolants. The reactor, object of study in this work, is an innovative reactor, due to its ability to operate in association with an Accelerator Driven System (ADS), considered a predecessor system of the next generation of nuclear reactors (GEN IV). The reactor selected was the MYRRHA (Multi-purpose Hybrid Research Reactor for High tech Applications) due to the availability of data to perform the simulation. In the modeling of the reactor with the code RELAP5-3D, the core was simulated using nodules with 1, 7, 15 and 51 thermohydraulic channels and eutectic lead-bismuth (LBE) as coolant. The parameters, such as, pressure, mass flow and coolant and heat structure temperature were analyzed. In addition, the thermal behavior of the core was evaluated by varying the type of coolant (sodium) in substitution for the LBE of the original design using the model with 7 thermohydraulic channels. The results of the steady-state calculations were compared with data from the literature and the proposed models were verified certifying the ability of the RELAP5-3D code to simulate this innovative reactor. After this step, it was analysed cases of transients with loss of coolant flow

Calculation analysis on steady state natural circulation characteristics

International Nuclear Information System (INIS)

Wang Fei; Nie Changhua; Huang Yanping

2005-01-01

The calculation results of single-phase steady state natural circulation characteristics by using Retran02 code have been presented, good agreement is achieved between the verified calculation result and the experimental data which were conducted at a test facility. Based on the calculation model, some sensibility analyses were made and much deeper understanding for single-phase steady state natural circulation characteristics was obtained. (author)

Selection of steady states in planar Darcy convection

International Nuclear Information System (INIS)

Tsybulin, V.G.; Karasoezen, B.; Ergenc, T.

2006-01-01

The planar natural convection of an incompressible fluid in a porous medium is considered. We study the selection of steady states under temperature perturbations on the boundary. A selection map is introduced in order to analyze the selection of a steady state from a continuous family of equilibria which exists under zero boundary conditions. The results of finite-difference modeling for a rectangular enclosure are presented

Experiments on steady state particle control in Tore Supra and DIII-D

International Nuclear Information System (INIS)

Mioduszewski, P.K.; Hogan, J.T.; Owen, L.W.; Maingi, R.; Lee, D.K.; Hillis, D.L.; Klepper, C.C.; Menon, M.M.; Thomas, C.E.; Uckan, T.; Wade, M.R.; Chatelier, M.; Grisolia, C.; Ghendrih, P.; Grosman, A.; Hutter, T.; Loarer, T.; Pegourie, B.; Mahdavi, M.A.; Schaffer, M.

1995-01-01

Particle control is playing an increasingly important role in tokamak plasma performance. The present paper discusses particle control of hydogen/deuterium by wall pumping on graphite or carbonized surfaces, as well as by external exhaust with pumped limiters and pumped divertors. Wall pumping is ultimately a transient effect and by itself not suitable for steady state particle exhaust. Therefore, external exhaust techniques with pumped divertors and limiters are being developed. How wall pumping phenomena interact and correlate with these inherently steady state, external exhaust techniques, is not well known to date. In the present paper, the processes involved in wall pumping and in external pumping are investigated in an attempt to evaluate the effect of external exhaust on wall pumping. Some of the key elements of this analysis are: (1) charge-exchange fluxes to the wall play a crucial role in the core-wall particle dynamics, (2) the recycling fluxes of thermal molecules have a high probability of ionization in the scrape-off layer, (3) thermal particles originating from the wall, which are ionized within the scrape-off layer, can be directly exhausted, thus providing a direct path between wall and exhaust which can be used to control the wall inventory. This way, the wall can be kept in a continuous pumping state in the sense that it continuously absorbs energetic particles and releases thermal molecules which are then removed by the external exhaust mechanism. While most of the ingredients of this analysis have been observed individually before, the present evaluation is an attempt to correlate effects of wall recycling and external exhaust. ((orig.))

Myocardial tagging with steady state free precession techniques and semi-automatic postprocessing--impact on diagnostic value

DEFF Research Database (Denmark)

Johnson, Thorsten R C; Bayrhof, Nicole; Huber, Armin

2007-01-01

Our aim was to determine the diagnostic value of myocardial tagging sequences with regard to the evaluable share of the cardiac cycle. Thirty-three patients were examined at 1.5 T using tagging sequences with gradient-echo (GRE) readout, 18 patients at 1.5 T with steady-state free precession (SSF...

Steady-state entanglement activation in optomechanical cavities

Science.gov (United States)

Farace, Alessandro; Ciccarello, Francesco; Fazio, Rosario; Giovannetti, Vittorio

2014-02-01

Quantum discord, and related indicators, are raising a relentless interest as a novel paradigm of nonclassical correlations beyond entanglement. Here, we discover a discord-activated mechanism yielding steady-state entanglement production in a realistic continuous-variable setup. This comprises two coupled optomechanical cavities, where the optical modes (OMs) communicate through a fiber. We first use a simplified model to highlight the creation of steady-state discord between the OMs. We show next that such discord improves the level of stationary optomechanical entanglement attainable in the system, making it more robust against temperature and thermal noise.

Development of synchronous generator saturation model from steady-state operating data

Energy Technology Data Exchange (ETDEWEB)

Jadric, Martin; Despalatovic, Marin; Terzic, Bozo [FESB University of Split, Faculty of Electrical Engineering, Mechanical Engineering and Naval Architecture, Split (Croatia)

2010-11-15

A new method to estimate and model the saturated synchronous reactances of hydroturbine generators from operating data is presented. For the estimation process, measurements of only the generator steady-state variables are required. First, using a specific procedure, the field to armature turns ratio is estimated from measured steady-state variables at constant power generation and various excitation conditions. Subsequently, for each set of steady-state operating data, saturated synchronous reactances are identified. Fitting surfaces, defined as polynomial functions in two variables, are later used to model these saturated reactances. It is shown that the simpler polynomial functions may be used to model saturation at the steady-state than at the dynamic conditions. The developed steady-state model is validated with measurements performed on the 34 MVA hydroturbine generator. (author)

A novel solar multifunctional PV/T/D system for green building roofs

International Nuclear Information System (INIS)

Feng, Chaoqing; Zheng, Hongfei; Wang, Rui; Yu, Xu; Su, Yuehong

2015-01-01

Highlights: • A novel transparent roof combines the solar PV/T/D system with green building design. • Novel photovoltaic-thermal roofing design can achieve excellent light control at noon. • The roof has no obvious influence on indoor light intensity in morning and afternoon. • Higher efficiency of solar energy utilization could be achieved with new roofing. - Abstract: A novel transparent roof which is made of solid CPC (Compound Parabolic Concentrator) PV/T/D (Photovoltaic/Thermal/Day lighting) system is presented. It combines the solar PV/T/D system with green building design. The PV/T/D system can achieve excellent light control at noon and adjust the thermal environment in the building, such that high efficiency utilization of solar energy could be achieved in modern architecture. This kind of roof can increase the visual comfort for building occupants; it can also avoid the building interior from overheating and dazzling at noon which is caused by direct sunlight through transparent roof. Optical simulation software is used to track the light path in different incidence angles. CFD (Computational Fluid Dynamics) simulation and steady state experiment have been taken to investigate the thermal characteristic of PV/T/D device. Finally, the PV/T/D experimental system was built; and the PV efficiency, light transmittance and air heating power of the system are tested under real sky conditions

Analysis of steady-state creep of Fe-Mo alloys from the viewpoint of recovery

International Nuclear Information System (INIS)

Maruyama, K.; Karashima, S.; Oikawa, H.

1979-01-01

A theoretical equation to d evaluate the steady-state creep-rates, d epsilon/dtsub(s), based on a recovery creep model is derived: epsilonsub(s)/dt proportional to r/sigma 2 sub(a) x lambda 2 , where r is the recovery rate, which can be determined from results of stress-reduction tests, deltasub(a) the applied stress, and lambda the dislocation link-length. Two cases of recovery are considered, i.e., recovery of dislocation networks at sub-boundaries and that of three-dimensional networks within subgrains. The high-temperature steady-state creep of Fe-Mo solid solutions, creep characteristics of which have been reported to be well rationalized as viscous glide creep, is analyzed using this equation. It is shown that stress dependence of d epsilon/dtsub(s) is well explained from the viewpoint of recovery, in which the activation and the annihilation of dislocations at sub-boundaries are considered to take place. (orig.) [de

Steady-State VEP-Based Brain-Computer Interface Control in an Immersive 3D Gaming Environment

Directory of Open Access Journals (Sweden)

Burke R

2005-01-01

Full Text Available This paper presents the application of an effective EEG-based brain-computer interface design for binary control in a visually elaborate immersive 3D game. The BCI uses the steady-state visual evoked potential (SSVEP generated in response to phase-reversing checkerboard patterns. Two power-spectrum estimation methods were employed for feature extraction in a series of offline classification tests. Both methods were also implemented during real-time game play. The performance of the BCI was found to be robust to distracting visual stimulation in the game and relatively consistent across six subjects, with 41 of 48 games successfully completed. For the best performing feature extraction method, the average real-time control accuracy across subjects was 89%. The feasibility of obtaining reliable control in such a visually rich environment using SSVEPs is thus demonstrated and the impact of this result is discussed.

D-T axicell magnet system for MFTF-α+T

International Nuclear Information System (INIS)

Srivastava, V.C.

1983-01-01

The configuration and design of the deuterium-tritium (D-T) axicell superconducting magnets for the Mirror Fusion Test Facility (MFTF-α+T) are described. The MFTF-α+T is an upgrade of the MFTF-B, with new end-plug magnets and a neutron-producing central D-T axicell section. The 4-m long axicell - its length defined by the 12-T peaks in the mirror field - is beam fueled and heated by two beam lines, each with four neutral beam injection ports. Two large superconducting coils (means diameter approx. 3.8 m) located at Z = +-2.40 m, in conjunction with a small copper coil located outside the test volume region, produce the 4.5-T mirror midplane field. This background field is augmented by two copper coils to create the 12-T peak mirror fields at Z = +-2 m. The central region of the axicell accommodates a 1-m-long, replaceable blanket test module. The length (4 m) of the axicell was chosen to provide relatively uniform neutron wall loading over the test module. In many respects, this axicell is less than full scale, but it could be viewed as a short section of a reactor, complete with the support systems and technologies associated with a mirror reactor. The peak field at the superconducting coils is 10.8 T. The coils employ hybrid superconducting winding - Nb 3 Sn conductor in the 8- to 12-T region and NbTi in the 0- to 8-T region. The winding is cryostable and is cooled by a 4.2 K liquid helium bath. The conductor design, the winding design, and the performance analyses for these superconducting coils are described

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

International Nuclear Information System (INIS)

Strupczewski, A.; Marks, P.

1997-01-01

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

Steady-state capabilities for hydroturbines with OpenFOAM

Science.gov (United States)

Page, M.; Beaudoin, M.; Giroux, A. M.

2010-08-01

The availability of a high quality Open Source CFD simulation platform like OpenFOAM offers new R&D opportunities by providing direct access to models and solver implementation details. Efforts have been made by Hydro-Québec to adapt OpenFOAM to hydroturbines for the development of steady-state capabilities. The paper describes the developments that have been made to implement new turbomachinery related capabilities: Multiple Frame of Reference solver, domain coupling interfaces (GGI, cyclicGGI and mixing plane) and specialized boundary conditions. Practical use of the new turbomachinery capabilities are demonstrated for the analysis of a 195-MW Francis hydroturbine.

Steady-state capabilities for hydroturbines with OpenFOAM

Energy Technology Data Exchange (ETDEWEB)

Page, M; Beaudoin, M; Giroux, A M, E-mail: page.maryse@ireq.c [Hydro-Quebec, Institut de recherche 1800 Lionel-Boulet, Varennes, Quebec J3X 1S1 (Canada)

2010-08-15

The availability of a high quality Open Source CFD simulation platform like OpenFOAM offers new R and D opportunities by providing direct access to models and solver implementation details. Efforts have been made by Hydro-Quebec to adapt OpenFOAM to hydroturbines for the development of steady-state capabilities. The paper describes the developments that have been made to implement new turbomachinery related capabilities: Multiple Frame of Reference solver, domain coupling interfaces (GGI, cyclicGGI and mixing plane) and specialized boundary conditions. Practical use of the new turbomachinery capabilities are demonstrated for the analysis of a 195-MW Francis hydroturbine.

Steady-state capabilities for hydroturbines with OpenFOAM

International Nuclear Information System (INIS)

Page, M; Beaudoin, M; Giroux, A M

2010-01-01

The availability of a high quality Open Source CFD simulation platform like OpenFOAM offers new R and D opportunities by providing direct access to models and solver implementation details. Efforts have been made by Hydro-Quebec to adapt OpenFOAM to hydroturbines for the development of steady-state capabilities. The paper describes the developments that have been made to implement new turbomachinery related capabilities: Multiple Frame of Reference solver, domain coupling interfaces (GGI, cyclicGGI and mixing plane) and specialized boundary conditions. Practical use of the new turbomachinery capabilities are demonstrated for the analysis of a 195-MW Francis hydroturbine.

Steady-State Calculation of the ATLAS Test Facility Using the SPACE Code

International Nuclear Information System (INIS)

Kim, Hyoung Tae; Choi, Ki Yong; Kim, Kyung Doo

2011-01-01

The Korean nuclear industry is developing a thermalhydraulic analysis code for safety analysis of pressurized water reactors (PWRs). The new code is called the Safety and Performance Analysis Code for Nuclear Power Plants (SPACE). Several research and industrial organizations including KAERI (Korea Atomic Energy Research Institute) are participating in the collaboration for the development of the SPACE code. One of the main tasks of KAERI is to carry out separate effect tests (SET) and integral effect tests (IET) for code verification and validation (V and V). The IET has been performed with ATLAS (Advanced Thermalhydraulic Test Loop for Accident Simulation) based on the design features of the APR1400 (Advanced Power Reactor of 1400MWe). In the present work the SPACE code input-deck for ATLAS is developed and used for simulation of the steady-state conditions of ATLAS as a preliminary work for IET V and V of the SPACE code

The Markov process admits a consistent steady-state thermodynamic formalism

Science.gov (United States)

Peng, Liangrong; Zhu, Yi; Hong, Liu

2018-01-01

The search for a unified formulation for describing various non-equilibrium processes is a central task of modern non-equilibrium thermodynamics. In this paper, a novel steady-state thermodynamic formalism was established for general Markov processes described by the Chapman-Kolmogorov equation. Furthermore, corresponding formalisms of steady-state thermodynamics for the master equation and Fokker-Planck equation could be rigorously derived in mathematics. To be concrete, we proved that (1) in the limit of continuous time, the steady-state thermodynamic formalism for the Chapman-Kolmogorov equation fully agrees with that for the master equation; (2) a similar one-to-one correspondence could be established rigorously between the master equation and Fokker-Planck equation in the limit of large system size; (3) when a Markov process is restrained to one-step jump, the steady-state thermodynamic formalism for the Fokker-Planck equation with discrete state variables also goes to that for master equations, as the discretization step gets smaller and smaller. Our analysis indicated that general Markov processes admit a unified and self-consistent non-equilibrium steady-state thermodynamic formalism, regardless of underlying detailed models.

Steady-state cerebral glucose concentrations and transport in the human brain

OpenAIRE

Gruetter, R.; Ugurbil, K.; Seaquist, E. R.

1998-01-01

Understanding the mechanism of brain glucose transport across the blood- brain barrier is of importance to understanding brain energy metabolism. The specific kinetics of glucose transport nave been generally described using standard Michaelis-Menten kinetics. These models predict that the steady- state glucose concentration approaches an upper limit in the human brain when the plasma glucose level is well above the Michaelis-Menten constant for half-maximal transport, K(t). In experiments wh...

CFD Model Development and validation for High Temperature Gas Cooled Reactor Cavity Cooling System (RCCS) Applications

International Nuclear Information System (INIS)

Hassan, Yassin; Corradini, Michael; Tokuhiro, Akira; Wei, Thomas Y.C.

2014-01-01

The Reactor Cavity Cooling Systems (RCCS) is a passive safety system that will be incorporated in the VTHR design. The system was designed to remove the heat from the reactor cavity and maintain the temperature of structures and concrete walls under desired limits during normal operation (steady-state) and accident scenarios. A small scale (1:23) water-cooled experimental facility was scaled, designed, and constructed in order to study the complex thermohydraulic phenomena taking place in the RCCS during steady-state and transient conditions. The facility represents a portion of the reactor vessel with nine stainless steel coolant risers and utilizes water as coolant. The facility was equipped with instrumentation to measure temperatures and flow rates and a general verification was completed during the shakedown. A model of the experimental facility was prepared using RELAP5-3D and simulations were performed to validate the scaling procedure. The experimental data produced during the steady-state run were compared with the simulation results obtained using RELAP5-3D. The overall behavior of the facility met the expectations. The facility capabilities were confirmed to be very promising in performing additional experimental tests, including flow visualization, and produce data for code validation.

CFD Model Development and validation for High Temperature Gas Cooled Reactor Cavity Cooling System (RCCS) Applications

Energy Technology Data Exchange (ETDEWEB)

Hassan, Yassin [Univ. of Wisconsin, Madison, WI (United Texas A & M Univ., College Station, TX (United States); Corradini, Michael; Tokuhiro, Akira; Wei, Thomas Y.C.

2014-07-14

The Reactor Cavity Cooling Systems (RCCS) is a passive safety system that will be incorporated in the VTHR design. The system was designed to remove the heat from the reactor cavity and maintain the temperature of structures and concrete walls under desired limits during normal operation (steady-state) and accident scenarios. A small scale (1:23) water-cooled experimental facility was scaled, designed, and constructed in order to study the complex thermohydraulic phenomena taking place in the RCCS during steady-state and transient conditions. The facility represents a portion of the reactor vessel with nine stainless steel coolant risers and utilizes water as coolant. The facility was equipped with instrumentation to measure temperatures and flow rates and a general verification was completed during the shakedown. A model of the experimental facility was prepared using RELAP5-3D and simulations were performed to validate the scaling procedure. The experimental data produced during the steady-state run were compared with the simulation results obtained using RELAP5-3D. The overall behavior of the facility met the expectations. The facility capabilities were confirmed to be very promising in performing additional experimental tests, including flow visualization, and produce data for code validation.

Over Twenty Years Of Experience In ITU TRIGA MARK-II Reactor

International Nuclear Information System (INIS)

Yavuz, Hasbi

2008-01-01

I.T.U. TRIGA MARK-II Training and Research Reactor, rated at 250 kW steady-state and 1200 MW pulsing power is the only research and training reactor owned and operated by a university in Turkey. Reactor has been operating since March 11, 1979; therefore the reactor has been operating successfully for more than twenty years. Over the twenty years of operation: - The tangential beam tube was equipped with a neutron radiography facility, which consists of a divergent collimator and exposure room; - A computerized data acquisition system was designed and installed such that all parameters of the reactor, which are observed from the console, could be monitored both in normal and pulse operations; - An electrical power calibration system was built for the thermal power calibration of the reactor; - Publications related with I.T.U. TRIGA MARK-II Training and Research Reactor are listed in Appendix; - Two majors undesired shutdown occurred; - The I.T.U. TRIGA MARK-II Training and Research Reactor is still in operation at the moment. (authors)

Comparing a Dynamic Fed-Batch and a Continuous Steady-State Simulation of Ethanol Fermentation in a Distillery to a Stoichiometric Conversion Simulation

OpenAIRE

Fonseca, G.C.; Costa, C.B.B.; Cruz, A.J.G.

2017-01-01

Abstract An autonomous sugarcane bioethanol plant was simulated in EMSO software, an equation oriented process simulator. Three types of fermentation units were simulated: a six parallel fed-batch reactor system, a set of four CSTR in steady state and one consisting of a single stoichiometric reactor. Stoichiometric models are less accurate than kinetic-based fermentation models used for fed-batch and continuous fermenter simulations, since they do not account for inhibition effects and depen...

Direct energy conversion and neutral beam injection for catalyzed D and D-3He tokamak reactors

International Nuclear Information System (INIS)

Blum, A.S.; Moir, R.W.

1977-01-01

The calculated performance of single stage and Venetian blind direct energy converters for Catalyzed D and D- 3 He Tokamak reactors are discussed. Preliminary results on He pumping are outlined. The efficiency of D and T neutral beam injection is reviewed

Progress towards Steady State on NSTX

International Nuclear Information System (INIS)

Gates, D.A.; Kessel, C.; Menard, J.; Taylor, G.; Wilson, J.R.

2005-01-01

In order to reduce recirculating power fraction to acceptable levels, the spherical torus concept relies on the simultaneous achievement of high toroidal β and high bootstrap fraction in steady state. In the last year, as a result of plasma control system improvements, the achievable plasma elongation on the National Spherical Torus Experiment (NSTX) has been raised from κ ∼ 2.1 to κ ∼ 2.6--approximately a 25% increase. This increase in elongation has lead to a doubling increase in the toroidal β for long-pulse discharges. The increase in β is associated with an increase in plasma current at nearly fixed poloidal β, which enables higher β t with nearly constant bootstrap fraction. As a result, for the first time in a spherical torus, a discharge with a plasma current of 1 MA has been sustained for 1 second. Data is presented from NSTX correlating the increase in performance with increased plasma shaping capability. In addition to improved shaping, H-modes induced during the current ramp phase of the plasma discharge have been used to reduce flux consumption during and to delay the onset of MHD instabilities. A modeled integrated scenario, which has 100% non-inductive current drive with very high toroidal β, will also be presented. The NSTX poloidal field coils are currently being modified to produce the plasma shape which is required for this scenario, which requires high triangularity ((delta) ∼ 0.8) at elevated elongation (κ ∼ 2.5). The other main requirement for steady state on NSTX is the ability to drive a fraction of the total plasma current with radio-frequency waves. The results of High Harmonic Fast Wave heating and current drive studies as well as electron Bernstein Wave emission studies will be presented

Economically attractive features of steady-state neoclassical reversed field pinch equilibrium with low aspect ratio

International Nuclear Information System (INIS)

Shiina, S.; Yagi, Y.; Sugimoto, H.; Ashida, H.; Hirano, Y.; Koguchi, H.; Sakakita, H.; Taguchi, M.; Nagamine, Y.; Osanai, Y.; Saito, K.; Watanabe, M.; Aizawa, M.

2005-01-01

Dominant plasma self-induced current equilibrium is achieved together with the high β for the steady-state neoclassical reversed field pinch (RFP) equilibrium with low aspect ratio by broadening the plasma pressure profile. The RF-driven current, when the safety factor is smaller than unity, is much less than the self-induced current, which dominates (96%) the toroidal current. This neoclassical RFP equilibrium has strong magnetic shear or a high-stability beta (β t = 63%) due to its hollow current profile. It is shown that the obtained equilibrium is close to the relaxed-equilibrium state with a minimum energy, and is also robust against microinstabilities. These attractive features allow the economical design of compact steady-state fusion power plants with low cost of electricity (COE). (author)

Evaluation of cartilage surface injuries using 3D-double echo steady state (3D-DESS): Effect of changing flip angle from 40 deg to 90 deg

International Nuclear Information System (INIS)

Moriya, Susumu; Yokobayashi, Tsuneo; Ishikawa, Mitsunori; Miki, Yukio; Kanagaki, Mitsunori; Yamamoto, Akira; Okudaira, Shuzo; Nakamura, Shinichiro

2011-01-01

Background. In magnetic resonance imaging (MRI) with 3D-double-echo steady-state (3D-DESS) sequences, the cartilage-synovial fluid contrast is reported to be better with a flip angle of 90 deg than with the conventional flip angle of 40 deg, and the detection rate of knee cartilage injury may be improved. Purpose. To compare the diagnostic performance and certainty of diagnosis with 3D-DESS images made using two flip angle settings, 40 deg and 90 deg, for knee cartilage surface lesions of Grade 2 or above confirmed by arthroscopy. Material and Methods. Images were obtained with 3D-DESS flip angles of 40 deg and 90 deg at 1.0T in 13 consecutive patients (2 men, 11 women, age range 18-68 years) evaluated for superficial cartilage injury by arthroscopy. Two radiologists classified the presence or absence of cartilage damage of ≥Grade 2 as 'positive (p)' or 'negative (n)', respectively. The rate of agreement with arthroscopic diagnosis was then examined, and the diagnostic performance compared. Diagnostic confidence was assessed scoring the presence or absence of cartilage damage into three categories: 3 = can diagnose with absolute confidence; 2 = can diagnose with a level of certainty as probably present or probably absent; and 1 = cannot make a diagnosis. Results. In a comparison of the rate of agreement between diagnosis by 3D-DESS images and arthroscopy, the rate of agreement was significantly higher and diagnostic performance was better in 90 deg images for the medial femoral condyle only. Diagnostic confidence was significantly better with 90 deg flip angle images than with 40 deg flip angle images for all six cartilage surfaces. Conclusion. In evaluating knee cartilage surface lesions with 3D-DESS sequences, a flip angle setting of 90 deg is more effective than the conventional setting of 40 deg

Evaluation of cartilage surface injuries using 3D-double echo steady state (3D-DESS): Effect of changing flip angle from 40 deg to 90 deg

Energy Technology Data Exchange (ETDEWEB)

Moriya, Susumu; Yokobayashi, Tsuneo; Ishikawa, Mitsunori (Ishikawa Clinic, Kyoto (Japan)), email: smoyari@yahoo.co.jp; Miki, Yukio (Dept. of Radiology, Osaka City Univ. Graduate School of Medicine, Osaka (Japan)); Kanagaki, Mitsunori; Yamamoto, Akira (Dept. of Diagnostic Imaging and Nuclear Medicine, Kyoto Univ., Kyoto (Japan)); Okudaira, Shuzo (Dept. of Orthopaedics, Kyoto Police Hospital, Kyoto (Japan)); Nakamura, Shinichiro (Center for Musculoskeletal Research, Univ. of Tennessee, Knoxville, TN (United States))

2011-12-15

Background. In magnetic resonance imaging (MRI) with 3D-double-echo steady-state (3D-DESS) sequences, the cartilage-synovial fluid contrast is reported to be better with a flip angle of 90 deg than with the conventional flip angle of 40 deg, and the detection rate of knee cartilage injury may be improved. Purpose. To compare the diagnostic performance and certainty of diagnosis with 3D-DESS images made using two flip angle settings, 40 deg and 90 deg, for knee cartilage surface lesions of Grade 2 or above confirmed by arthroscopy. Material and Methods. Images were obtained with 3D-DESS flip angles of 40 deg and 90 deg at 1.0T in 13 consecutive patients (2 men, 11 women, age range 18-68 years) evaluated for superficial cartilage injury by arthroscopy. Two radiologists classified the presence or absence of cartilage damage of >=Grade 2 as 'positive (p)' or 'negative (n)', respectively. The rate of agreement with arthroscopic diagnosis was then examined, and the diagnostic performance compared. Diagnostic confidence was assessed scoring the presence or absence of cartilage damage into three categories: 3 = can diagnose with absolute confidence; 2 = can diagnose with a level of certainty as probably present or probably absent; and 1 = cannot make a diagnosis. Results. In a comparison of the rate of agreement between diagnosis by 3D-DESS images and arthroscopy, the rate of agreement was significantly higher and diagnostic performance was better in 90 deg images for the medial femoral condyle only. Diagnostic confidence was significantly better with 90 deg flip angle images than with 40 deg flip angle images for all six cartilage surfaces. Conclusion. In evaluating knee cartilage surface lesions with 3D-DESS sequences, a flip angle setting of 90 deg is more effective than the conventional setting of 40 deg

Temperature stabilisation in Fischer–Tropsch reactors using phase change material (PCM)

International Nuclear Information System (INIS)

Odunsi, Ademola O.; O'Donovan, Tadhg S.; Reay, David A.

2016-01-01

The Fischer–Tropsch (FT) reaction is highly exothermic. The exothermicity combined with a high sensitivity of product selectivity to temperature constitute the main challenges in the design of FT reactors. Temperature control is particularly critical to the process in order to ensure longevity of the catalyst, optimise the product distribution, and to ensure thermo-mechanical reliability of the entire process. The use of encapsulated, Phase Change Material (PCM), in conjunction with a supervisory temperature control mechanism, could help mitigate these challenges and intensify the heat transport from the reactor. A 2D-axisymmetric, pseudo-homogeneous, steady-state model, with the dissipation of the enthalpy of reaction into an isothermal PCM sink, in a wall-cooled, single-tube fixed bed reactor is presented. Effective temperature control shows a shift in thermodynamic equilibrium, favouring the selectivity of longer chain hydrocarbons (C_5_+) to the disadvantage of CH_4 selectivity-a much desired outcome in the hydrocarbon Gas-to-Liquid (GTL) industry. - Highlights: • Phase change material is used to control temperature in a Fischer–Tropsch reactor. • Effective temperature control favours the production of C_5_+ over CH_4. • A 2D-axisymmetric, steady-state model is presented. • The model is verified against similar experimental work done in literature.

The steady-state and time-resolved photophysical properties of a dimeric indium phthalocyanine complex

International Nuclear Information System (INIS)

Chen Yu; Araki, Yasuyuki; Dini, Danilo; Liu Ying; Ito, Osamu; Fujitsuka, Mamoru

2006-01-01

The steady-state and time-resolved photophysical properties and some molecular orbital calculation results of a dimeric indium phthalocyanine complex with an indium-indium bond, i.e., [tBu 4 PcIn] 2 .2tmed, have been described. The results regarding triplet excited state lifetimes can be ascribed to strong intramolecular interactions existing only in the excited state of this dimer because no significant difference in the absorption spectra of the tBu 4 PcInCl monomer and the [tBu 4 PcIn] 2 .2tmed dimer is observed, suggesting that no ground-state interaction can be assessed. The deactivation processes of the excited singlet state of [tBu 4 PcIn] 2 .2tmed are apparently faster than that of μ-oxo-bridged PcIn dimer [tBu 4 PcIn] 2 O. Molecular orbital calculation on the PcIn dimer shows no node between two indium atoms was found in the HOMO - 2 of the PcIn-InPc dimer, suggesting that bonding electrons distribute between two indium atoms

Steady state of tapped granular polygons

International Nuclear Information System (INIS)

Carlevaro, Carlos M; Pugnaloni, Luis A

2011-01-01

The steady state packing fraction of a tapped granular bed is studied for different grain shapes via a discrete element method. Grains are monosized regular polygons, from triangles to icosagons. Comparisons with disc packings show that the steady state packing fraction as a function of the tapping intensity presents the same general trends in polygon packings. However, better packing fractions are obtained, as expected, for shapes that can tessellate the plane (triangles, squares and hexagons). In addition, we find a sharp transition for packings of polygons with more than 13 vertices signaled by a discontinuity in the packing fraction at a particular tapping intensity. Density fluctuations for most shapes are consistent with recent experimental findings in disc packing; however, a peculiar behavior is found for triangles and squares

Application of RELAP5-3D code for thermal analysis of the ADS reactor core

International Nuclear Information System (INIS)

Fernandes, Gustavo Henrique Nazareno

2018-01-01

Nuclear power is essential to supply global energy demand. Therefore, in order to use nuclear fuel more efficiently, more efficient nuclear reactors technologies researches have been intensified, such as hybrid systems, composed of particle accelerators coupled into nuclear reactors. In order to add knowledge to such studies, an innovative reactor design was considered where the RELAP5-3D thermal-hydraulic analysis code was used to perform a thermal analysis of the core, either in stationary operation or in situations transitory. The addition of new kind of coolants, such as, liquid salts, among them Flibe, lead, lead-bismuth, sodium, lithium-bismuth and lithium-lead was an important advance in this version of the code, making possible to do the thermal simulation of reactors that use these types of coolants. The reactor, object of study in this work, is an innovative reactor, due to its ability to operate in association with an Accelerator Driven System (ADS), considered a predecessor system of the next generation of nuclear reactors (GEN IV). The reactor selected was the MYRRHA (Multi-purpose Hybrid Research Reactor for High tech Applications) due to the availability of data to perform the simulation. In the modeling of the reactor with the code RELAP5-3D, the core was simulated using nodules with 1, 7, 15 and 51 thermohydraulic channels and eutectic lead-bismuth (LBE) as coolant. The parameters, such as, pressure, mass flow and coolant and heat structure temperature were analyzed. In addition, the thermal behavior of the core was evaluated by varying the type of coolant (sodium) in substitution for the LBE of the original design using the model with 7 thermohydraulic channels. The results of the steady-state calculations were compared with data from the literature and the proposed models were verified certifying the ability of the RELAP5-3D code to simulate this innovative reactor. After this step, it was analysed cases of transients with loss of coolant flow

Quantitative Comparison of 2D and 3D MRI Techniques for the Evaluation of Chondromalacia Patellae in 3.0T MR Imaging of the Knee

Directory of Open Access Journals (Sweden)

Ali Özgen

2016-09-01

Full Text Available INTRODUCTION: Chondromalacia patellae is a very common disorder of patellar cartilage. Magnetic resonance imaging (MRI is a powerful non-invasive tool to investigate patellar cartilage lesions. Although many MRI sequences have been used in MR imaging of the patellar cartilage and the optimal pulse sequence is controversial, fat-saturated proton density images have been considered very valuable to evaluate patellar cartilage. The purpose of this study is to quantitatively compare the diagnostic performance of various widely used 2D and 3D MRI techniques for the evaluation of chondromalacia patellae in 3.0T MR imaging of the knee using T2 mapping images as the reference standard. METHODS: Sevety-five knee MRI exams of 69 adult consecutive were included in the study. Fat-saturated T2-weighted (FST2, fat-saturated proton density (FSPD, water-only T2-weighted DIXON (T2mD, T2-weighted 3 dimensional steady state (3DT2FFE, merged multi-echo steady state (3DmFFE, and water selective T1-weighted fat-supressed (WATSc images were acquired. Quantitative comparison of grade 1 and grade 5 lesions were made using contrast-to-noise (CNR ratios. Grade 2-4 lesions were scored qualitatively and scorings of the lesions were compared statistically. Analysis of variance and Tukey’s tests were used to compare CNR data. Two sample z-test was used to compare the ratio of MR exams positive for grade 1 lesions noted on T2-mapping and other conventional sequences. Paired samples t-test was used to compare two different pulse sequences. RESULTS: In detecting grade 1 lesions, FSPD, FST2 and T2mD images were superior in comparison to other sequences. FSPD and FST2 images were statistically superior in detecting grade 2-4 lesions. Although all grade 5 lesions were noted in every single sequence, FST2 images have the highest mean CNR followed by 3DT2FFE images. DISCUSSION AND CONCLUSION: FST2 sequence is equal or superior in detecting every grade of patellar chondromalacia in

Review of fusion DEMO reactor study

International Nuclear Information System (INIS)

Seki, Yasushi

1996-01-01

Fusion DEMO Reactor is defined and the Steady State Tokamak Reactor (SSTR) concept is introduced as a typical example of a DEMO reactor. Recent DEMO reactor studies in Japan and abroad are introduced. The DREAM Reactor concept is introduced as an ultimate target of fusion research. (author)

Steady-state propagation of interface corner crack

DEFF Research Database (Denmark)

Veluri, Badrinath; Jensen, Henrik Myhre

2013-01-01

Steady-state propagation of interface cracks close to three-dimensional corners has been analyzed. Attention was focused on modeling the shape of the interface crack front and calculating the critical stress for steady-state propagation of the crack. The crack propagation was investigated...... on the finite element method with iterative adjustment of the crack front to estimate the critical delamination stresses as a function of the fracture criterion and corner angles. The implication of the results on the delamination is discussed in terms of crack front profiles and the critical stresses...... for propagation and the angle of intersection of the crack front with the free edge....

Quasi steady-state aerodynamic model development for race vehicle simulations

Science.gov (United States)

Mohrfeld-Halterman, J. A.; Uddin, M.

2016-01-01

Presented in this paper is a procedure to develop a high fidelity quasi steady-state aerodynamic model for use in race car vehicle dynamic simulations. Developed to fit quasi steady-state wind tunnel data, the aerodynamic model is regressed against three independent variables: front ground clearance, rear ride height, and yaw angle. An initial dual range model is presented and then further refined to reduce the model complexity while maintaining a high level of predictive accuracy. The model complexity reduction decreases the required amount of wind tunnel data thereby reducing wind tunnel testing time and cost. The quasi steady-state aerodynamic model for the pitch moment degree of freedom is systematically developed in this paper. This same procedure can be extended to the other five aerodynamic degrees of freedom to develop a complete six degree of freedom quasi steady-state aerodynamic model for any vehicle.

Numerical Simulation of Steady State Conduction Heat Transfer During the Solidification of Aluminum Casting in Green Sand Mould

Directory of Open Access Journals (Sweden)

Victor ANJO

2012-08-01

Full Text Available The solidification of molten metal during the casting process involves heat transfer from the molten metal to the mould, then to the atmosphere. The mechanical properties and grain size of metals are determined by the heat transfer process during solidification. The aim of this study is to numerically stimulate the steady conduction heat transfer during the solidification of aluminum in green sand mould using finite difference analysis 2D. The properties of materials used are industrial AI 50/60 AFS green sand mould, pure aluminum and MATLAB 7.0.1. for the numerical simulation. The method includes; the finite difference analysis of the heat conduction equation in steady (Laplace’s and transient states and using MATLAB to numerically stimulate the thermal flow and cooling curve. The results obtained are: the steady state thermal flow in 2D and transient state cooling curve of casting. The results obtain were consider relevant in the control of the grain size and mechanical properties of the casting.

Analysis of the steady state hydraulic behaviour of the ITER blanket cooling system

Energy Technology Data Exchange (ETDEWEB)

Di Maio, P.A., E-mail: pietroalessandro.dimaio@unipa.it [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, 90128 Palermo (Italy); Dell’Orco, G.; Furmanek, A. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Garitta, S. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, 90128 Palermo (Italy); Merola, M.; Mitteau, R.; Raffray, R. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Spagnuolo, G.A.; Vallone, E. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, 90128 Palermo (Italy)

2015-10-15

Highlights: • Nominal steady state hydraulic behaviour of ITER blanket standard sector cooling system has been investigated. • Numerical simulations have been run adopting a qualified thermal-hydraulic system code. • Hydraulic characteristic functions and coolant mass flow rates, velocities and pressure drops have been assessed. • Most of the considered circuits are able to effectively cool blanket modules, meeting ITER requirements. - Abstract: The blanket system is the ITER reactor component devoted to providing a physical boundary for plasma transients and contributing to thermal and nuclear shielding of vacuum vessel, magnets and external components. It is expected to be subjected to significant heat loads under nominal conditions and its cooling system has to ensure an adequate cooling, preventing any risk of critical heat flux occurrence while complying with pressure drop limits. At the University of Palermo a study has been performed, in cooperation with the ITER Organization, to investigate the steady state hydraulic behaviour of the ITER blanket standard sector cooling system. A theoretical–computational approach based on the finite volume method has been followed, adopting the RELAP5 system code. Finite volume models of the most critical blanket cooling circuits have been set-up, realistically simulating the coolant flow domain. The steady state hydraulic behaviour of each cooling circuit has been investigated, determining its hydraulic characteristic function and assessing the spatial distribution of coolant mass flow rates, velocities and pressure drops under reference nominal conditions. Results obtained have indicated that the investigated cooling circuits are able to provide an effective cooling to blanket modules, generally meeting ITER requirements in term of pressure drop and velocity distribution, except for a couple of circuits that are being revised.

Realizing steady-state tokamak operation for fusion energy

International Nuclear Information System (INIS)

Luce, T. C.

2011-01-01

Continuous operation of a tokamak for fusion energy has clear engineering advantages but requires conditions beyond those sufficient for a burning plasma. The fusion reactions and external sources must support both the pressure and the current equilibrium without inductive current drive, leading to demands on stability, confinement, current drive, and plasma-wall interactions that exceed those for pulsed tokamaks. These conditions have been met individually, and significant progress has been made in the past decade to realize scenarios where the required conditions are obtained simultaneously. Tokamaks are operated routinely without disruptions near pressure limits, as needed for steady-state operation. Fully noninductive sustainment with more than half of the current from intrinsic currents has been obtained for a resistive time with normalized pressure and confinement approaching those needed for steady-state conditions. One remaining challenge is handling the heat and particle fluxes expected in a steady-state tokamak without compromising the core plasma performance.

Assessing Quasi-Steady State in Evaporation of Sessile Drops by Diffusion Models

Science.gov (United States)

Martin, Cameron; Nguyen, Hoa; Kelly-Zion, Peter; Pursell, Chris

2017-11-01

The vapor distributions surrounding sessile drops of methanol are modeled as the solutions of the steady-state and transient diffusion equations using Matlab's PDE Toolbox. The goal is to determine how quickly the transient diffusive transport reaches its quasi-steady state as the droplet geometry is varied between a Weber's disc, a real droplet shape, and a spherical cap with matching thickness or contact angle. We assume that the only transport mechanism at work is diffusion. Quasi-steady state is defined using several metrics, such as differences between the transient and steady-state solutions, and change in the transient solution over time. Knowing the vapor distribution, the gradient is computed to evaluate the diffusive flux. The flux is integrated along the surface of a control volume surrounding the drop to obtain the net rate of diffusion out of the volume. Based on the differences between the transient and steady-state diffusive fluxes at the discrete points along the control-volume surface, the time to reach quasi-steady state evaporation is determined and is consistent with other proposed measurements. By varying the dimensions of the control volume, we can also assess what regimes have equivalent or different quasi-steady states for different droplet geometries. Petroleum Research Fund.

3 T MR 3D fast imaging employing steady state acquisition demonstrating branches of intraparotid facial nerve, parotid duct, and relation with parotid tumors

International Nuclear Information System (INIS)

Li Yan; Li Chuanting; Zhang Dongsheng; Ai Bin; Zhang Weidong; Wu Lebin

2010-01-01

Objective: To investigate the usefulness of 3 T MRI 3D-FIESTA in the evaluation of the intraparotid components of the facial nerve and parotid duct, and compare them with surgical findings. Methods: Twenty-two cases with parotid benign tumors were scanned with conventional and 3D-FIESTA sequences on 3 T MRI scanner. Postprocessed multiplanar images were obtained with the workstation. Parotid ducts and facial nerves and tumors were identified on these images. The relationship of the tumors to the facial nerves and Parotid ducts was confirmed at surgery. Results: Various types of parotid benign tumors had their characteristics on 3 T MR imaging. Parotid benign tumors mainly showed hypo-intensity on T 1 WI in 21 cases, and hyper-intensity on T 2 WI in 22 cases. But on 3D-FIESTA images, they appeared hypointensity (10 cases) or high intensity (12 cases) due to different types. Facial nerves in parotid appeared as linear structures with hypo-inteusity. The indication of the main trunks were 16 and 18 cases for T 1 WI and T 2 WI images, while on 3D-FIESTA images, the main trunks and cervicofacial and temporofacial divisions of the facial nerves were found in 22, 21,22 cases. Parotid ducts appeared as structures with hypo-intensity on T 1 WI and hyper-intensity on multiplanar images (14, 20, 22 cases). Compared with surgical results, the main trunks of the facial nerve were correctly showed by 3D-FIESTA images in 20 cases. However, in 2 cases they were not located in the operation because of shifting. Conclusion: 3 T MR 3D-FIESTA imaging could depict the extracranial facial nerve and the parotid duct in the parotid gland, which is useful for preoperative evaluation of parotid gland tumors. (authors)

Reactor Dosimetry State of the Art 2008

Science.gov (United States)

Voorbraak, Wim; Debarberis, Luigi; D'Hondt, Pierre; Wagemans, Jan

2009-08-01

. Williams, A. P. Ribaric and T. Schnauber. Agile high-fidelity MCNP model development techniques for rapid mechanical design iteration / J. A. Kulesza.Extension of Raptor-M3G to r-8-z geometry for use in reactor dosimetry applications / M. A. Hunter, G. Longoni and S. L. Anderson. In vessel exposure distributions evaluated with MCNP5 for Atucha II / J. M. Longhino, H. Blaumann and G. Zamonsky. Atucha I nuclear power plant azimutal ex-vessel flux profile evaluation / J. M. Longhino ... [et al.]. UFTR thermal column characterization and redesign for maximized thermal flux / C. Polit and A. Haghighat. Activation counter using liquid light-guide for dosimetry of neutron burst / M. Hayashi ... [et al.]. Control rod reactivity curves for the annular core research reactor / K. R. DePriest ... [et al.]. Specification of irradiation conditions in VVER-440 surveillance positions / V. Kochkin ... [et al.]. Simulations of Mg-Ar ionisation and TE-TE ionisation chambers with MCNPX in a straightforward gamma and beta irradiation field / S. Nievaart ... [et al.]. The change of austenitic stainless steel elements content in the inner parts of VVER-440 reactor during operation / V. Smutný, J. Hep and P. Novosad. Fast neutron environmental spectrometry using disk activation / G. Lövestam ... [et al.]. Optimization of the neutron activation detector location scheme for VVER-lOOO ex-vessel dosimetry / V. N. Bukanov ... [et al.]. Irradiation conditions for surveillance specimens located into plane containers installed in the WWER-lOOO reactor of unit 2 of the South-Ukrainian NPP / O. V. Grytsenko. V. N. Bukanov and S. M. Pugach. Conformity between LRO mock-ups and VVERS NPP RPV neutron flux attenuation / S. Belousov. Kr. Ilieva and D. Kirilova. FLUOLE: a new relevant experiment for PWR pressure vessel surveillance / D. Beretz ... [et al.]. Transport of neutrons and photons through the iron and water layers / M. J. Kost'ál ... [et al.]. Condition evaluation of spent nuclear fuel assemblies

Development of a real-time simulation tool towards self-consistent scenario of plasma start-up and sustainment on helical fusion reactor FFHR-d1

Science.gov (United States)

Goto, T.; Miyazawa, J.; Sakamoto, R.; Suzuki, Y.; Suzuki, C.; Seki, R.; Satake, S.; Huang, B.; Nunami, M.; Yokoyama, M.; Sagara, A.; the FFHR Design Group

2017-06-01

This study closely investigates the plasma operation scenario for the LHD-type helical reactor FFHR-d1 in view of MHD equilibrium/stability, neoclassical transport, alpha energy loss and impurity effect. In 1D calculation code that reproduces the typical pellet discharges in LHD experiments, we identify a self-consistent solution of the plasma operation scenario which achieves steady-state sustainment of the burning plasma with a fusion gain of Q ~ 10 was found within the operation regime that has been already confirmed in LHD experiment. The developed calculation tool enables systematic analysis of the operation regime in real time.

Nuclear characteristics of D-D fusion reactor blankets

International Nuclear Information System (INIS)

Nakashima, Hideki; Ohta, Masao

1978-01-01

Fusion reactors operating on deuterium (D-D) cycle are considered to be of long range interest for their freedom from tritium breeding in the blanket. The present paper discusses the various possibilities of D-D fusion reactor blanket designs mainly from the standpoint of the nuclear characteristics. Neutronic and photonic calculations are based on presently available data to provide a basis of the optimal blanket design in D-D fusion reactors. It is found that it appears desirable to design a blanket with blanket/shield (BS) concept in D-D fusion reactors. The BS concept is designed to obtain reasonable shielding characteristics for superconducting magnet (SCM) by using shielding materials in the compact blanket. This concept will open the possibility of compact radiation shield design based on assured technology, and offer the advantage from the system economics point of view. (auth.)

Pre-steady state transients in the Drosophila alcohol dehydrogenase catalyzed reaction: isotope effects and stereospecificity

International Nuclear Information System (INIS)

Place, A.R.; Eccleston, J.F.

1987-01-01

The alcohol dehydrogenase (ADH) isolated from Drosophila is unique among alcohol metabolizing enzymes by not requiring metals for catalysis, by showing 4-pro-S (B-sided) hydride transfer stereospecificity, and by possessing a greater catalytic turnover rate for secondary alcohols than for primary alcohols. They have extended their studies on the kinetic mechanism for this enzyme by examining the pre-steady state transients of ternary complex interconversion using stopped-flow fluorescence methods. When enzyme and a 30-fold molar excess of NADH is mixed with excess acetadehyde, methyl ethyl ketone (MEK), or cyclohexanone a rapid (> 100 s -1 ) transient is observe before the steady-state. The rates are insensitive to isotope substitution. With the substrate MEK, the rate and amplitude suggests a single turnover of the enzyme. Similar pre-steady state transients are observed when enzyme and a 50-fold molar excess of NAD + is mixed with ethanol, 2-propanol, and cyclohexanol. The rates show a hyperbolic concentration dependence and a deuterium isotope effect. With d 6 -deuteroethanol the transient no longer occurs in the pre-steady state. When the optical isomers of secondary alcohols are used as substrates, transients are observed only in the R-(-) isomers for all chain lengths. With 2-S(+)-heptanol and 2-S(+)-octanol no transients occur

On the validity of travel-time based nonlinear bioreactive transport models in steady-state flow.

Science.gov (United States)

Sanz-Prat, Alicia; Lu, Chuanhe; Finkel, Michael; Cirpka, Olaf A

2015-01-01

Travel-time based models simplify the description of reactive transport by replacing the spatial coordinates with the groundwater travel time, posing a quasi one-dimensional (1-D) problem and potentially rendering the determination of multidimensional parameter fields unnecessary. While the approach is exact for strictly advective transport in steady-state flow if the reactive properties of the porous medium are uniform, its validity is unclear when local-scale mixing affects the reactive behavior. We compare a two-dimensional (2-D), spatially explicit, bioreactive, advective-dispersive transport model, considered as "virtual truth", with three 1-D travel-time based models which differ in the conceptualization of longitudinal dispersion: (i) neglecting dispersive mixing altogether, (ii) introducing a local-scale longitudinal dispersivity constant in time and space, and (iii) using an effective longitudinal dispersivity that increases linearly with distance. The reactive system considers biodegradation of dissolved organic carbon, which is introduced into a hydraulically heterogeneous domain together with oxygen and nitrate. Aerobic and denitrifying bacteria use the energy of the microbial transformations for growth. We analyze six scenarios differing in the variance of log-hydraulic conductivity and in the inflow boundary conditions (constant versus time-varying concentration). The concentrations of the 1-D models are mapped to the 2-D domain by means of the kinematic (for case i), and mean groundwater age (for cases ii & iii), respectively. The comparison between concentrations of the "virtual truth" and the 1-D approaches indicates extremely good agreement when using an effective, linearly increasing longitudinal dispersivity in the majority of the scenarios, while the other two 1-D approaches reproduce at least the concentration tendencies well. At late times, all 1-D models give valid approximations of two-dimensional transport. We conclude that the

NATBWR: a steady-state model for natural circulation in boiling-water reactors

International Nuclear Information System (INIS)

Healzer, J.M.; Abdollahian, D.

1983-02-01

This report documents the NATBWR steady-state BWR natural-circulation model and activities under EPRI Project RP1561-1 to gather data and predict the natural-circulation operation of the BWR. The report is organized into two parts, with the first part describing the NATBWR model and applications of the model to available BWR natural-circulation data and the second part providing user and programming information about the model. Five different operating BWR's were selected to demonstrate the application of the NATBWR model, one of each type from BWR/1 through BWR/4. For each operating plant, the available natural circulation data has been compared to model predictions. In addition to the data predictions, the behavior of the BWR system at reduced inventory, where the system is isolated and scrammed, and cooling provided by natural circulation has been analyzed. Finally, included as an appendix to Part 1 of this report is a discussion of the stability of the BWR system at natural-circulation conditions

A nodally condensed SUPG formulation for free-surface computation of steady-state flows constrained by unilateral contact - Application to rolling

Science.gov (United States)

Arora, Shitij; Fourment, Lionel

2018-05-01

In the context of the simulation of industrial hot forming processes, the resultant time-dependent thermo-mechanical multi-field problem (v →,p ,σ ,ɛ ) can be sped up by 10-50 times using the steady-state methods while compared to the conventional incremental methods. Though the steady-state techniques have been used in the past, but only on simple configurations and with structured meshes, and the modern-days problems are in the framework of complex configurations, unstructured meshes and parallel computing. These methods remove time dependency from the equations, but introduce an additional unknown into the problem: the steady-state shape. This steady-state shape x → can be computed as a geometric correction t → on the domain X → by solving the weak form of the steady-state equation v →.n →(t →)=0 using a Streamline Upwind Petrov Galerkin (SUPG) formulation. There exists a strong coupling between the domain shape and the material flow, hence, a two-step fixed point iterative resolution algorithm was proposed that involves (1) the computation of flow field from the resolution of thermo-mechanical equations on a prescribed domain shape and (2) the computation of steady-state shape for an assumed velocity field. The contact equations are introduced in the penalty form both during the flow computation as well as during the free-surface correction. The fact that the contact description is inhomogeneous, i.e., it is defined in the nodal form in the former, and in the weighted residual form in the latter, is assumed to be critical to the convergence of certain problems. Thus, the notion of nodal collocation is invoked in the weak form of the surface correction equation to homogenize the contact coupling. The surface correction algorithm is tested on certain analytical test cases and the contact coupling is tested with some hot rolling problems.

A cubic autocatalytic reaction in a continuous stirred tank reactor

Energy Technology Data Exchange (ETDEWEB)

Yakubu, Aisha Aliyu; Yatim, Yazariah Mohd [School of Mathematical Sciences, Universiti Sains Malaysia, 11800 USM, Penang Malaysia (Malaysia)

2015-10-22

In the present study, the dynamics of the cubic autocatalytic reaction model in a continuous stirred tank reactor with linear autocatalyst decay is studied. This model describes the behavior of two chemicals (reactant and autocatalyst) flowing into the tank reactor. The behavior of the model is studied analytically and numerically. The steady state solutions are obtained for two cases, i.e. with the presence of an autocatalyst and its absence in the inflow. In the case with an autocatalyst, the model has a stable steady state. While in the case without an autocatalyst, the model exhibits three steady states, where one of the steady state is stable, the second is a saddle point while the last is spiral node. The last steady state losses stability through Hopf bifurcation and the location is determined. The physical interpretations of the results are also presented.

Short TR imaging with refocusing of the steady-state transverse magnetization

International Nuclear Information System (INIS)

Zur, Y.; Stokar, S.; Bendel, P.

1987-01-01

Repetitive application of a sequence with repetition time (TR) shorter than T2 results in a steady state in which the transverse magnetization Mt reaches a nonzero value at the end of the sequence. This value depends on the TR and flip angle as well as on the frequency offset ν of each spin isochromat. The authors present a detailed analysis of the time domain and image domain signals for sequences with short TR that employ gradient reversal echoes. Because of the dependence of Mt on ν, two distinct echos appear in the time domain. With proper adjustment of the view gradients, each echo can be sampled separately. Image intensities derived for spins in a liquid (i.e., T1 -- T2) suggest enhanced signal intensity for the cerebrospinal fluid. This was confirmed experimentally

Measurement of TFTR D-T radiation shielding efficiency

International Nuclear Information System (INIS)

Kugel, H.W.; Ascione G.; Elwood, S.

1994-01-01

High power D-T fusion reactor designs presently exhibit complex geometric and material density configurations. Simulations of the radiation shielding required for safe operation and full compliance with all regulatory requirements must include sufficient margin to accommodate uncertainties in material properties and distributions, uncertainties in the final configurations, and uncertainties in approximations employing the homogenization of complex geometries. Measurements of radiation shielding efficiency performed in a realistic D-T tokamak environment can provide empirical guidance for simulating safe, efficient, and cost effective shielding systems for future high power fusion reactors. In this work, the authors present the results of initial measurements of the TFTR radiation shielding efficiency during high power D-T operations with record neutron yields. The TFTR design objective is to limit the total dose-equivalent at the nearest PPPL property lines from all radiation pathways to 10 mrem per calendar year. Compliance with this design objective over a calendar year requires measurements in the presence of typical site backgrounds of about 80 mrem per year

Technical specifications (replaces note T.62). Irradiation of graphite at ambient temperature, Note T. 76; Specification technique, (Annule et remplace la note T. 62), Irradiation de graphite a temperature ambiante, Note T. 76

Energy Technology Data Exchange (ETDEWEB)

Reseau, R A [Services des grandes piles experimentales, Section ' Physique et Experimentation, Saclay (France)

1962-12-15

The objective is to study the effects of fast neutron irradiation of different graphite samples. The irradiation conditions should be as follows: integral fast neutron flux should be higher than 10{sup 20} neutrons/cm{sup 2}, the reactor should operate at steady state for 15 days, the temperature od samples should not be higher than 100 deg C, preferably 80 deg C. Note T. 62 which is replaced by this Note is attached.

Steady state theta pinch concept for slow formation of FRC

International Nuclear Information System (INIS)

Hirano, K.

1987-05-01

A steady state high beta plasma flow through a channel along the magnetic field increasing downstream can be regarded as a ''steady state theta pinch'', because if we see the plasma riding on the flow we should observe very similar process taking place in a theta pinch. Anticipating to produce an FRC without using very high voltage technics such as the ones required in a conventional theta pinch, we have studied after the analogy a ''steady state reversed field theta pinch'' which is brought about by steady head-on collision of counter plasma streams along the channel as ejected from two identical co-axial plasma sources mounted at the both ends of the apparatus. The ideal Poisson and shock adiabatic flow models are employed for the analysis of the steady colliding process. It is demonstrated that an FRC involving large numbers of particles is produced only by the weak shock mode which is achieved in case energetic plasma flow is decelerated almost to be stagnated through Poisson adiabatic process before the streams are collided. (author)

A new perspective on steady-state cosmology: from Einstein to Hoyle

OpenAIRE

O'Raifeartaigh, Cormac; Mitton, Simon

2015-01-01

We recently reported the discovery of an unpublished manuscript by Albert Einstein in which he attempted a 'steady-state' model of the universe, i.e., a cosmic model in which the expanding universe remains essentially unchanged due to a continuous formation of matter from empty space. The manuscript was apparently written in early 1931, many years before the steady-state models of Fred Hoyle, Hermann Bondi and Thomas Gold. We compare Einstein’s steady-state cosmology with that of Hoyle, Bondi...

Fission multipliers for D-D/D-T neutron generators

International Nuclear Information System (INIS)

Lou, T.P.; Vujic, J.L.; Koivunoro, H.; Reijonen, J.; Leung, K.-N.

2003-01-01

A compact D-D/D-T fusion based neutron generator is being designed at the Lawrence Berkeley National Laboratory to have a potential yield of 10 12 D-D n/s and 10 14 D-T n/s. Because of its high neutron yield and compact size (∼20 cm in diameter by 4 cm long), this neutron generator design will be suitable for many applications. However, some applications required higher flux available from nuclear reactors and spallation neutron sources operated with GeV proton beams. In this study, a subcritical fission multiplier with k eff of 0.98 is coupled with the compact neutron generators in order to increase the neutron flux output. We have chosen two applications to show the gain in flux due to the use of fission multipliers--in-core irradiation and out-of-core irradiation. For the in-core irradiation, we have shown that a gain of ∼25 can be achieved in a positron production system using D-T generator. For the out-of-core irradiation, a gain of ∼17 times is obtained in Boron Neutron Capture Therapy (BNCT) using a D-D neutron generator. The total number of fission neutrons generated by a source neutron in a fission multiplier with k eff is ∼50. For the out-of-core irradiation, the theoretical maximum net multiplication is ∼30 due to the absorption of neutrons in the fuel. A discussion of the achievable multiplication and the theoretical multiplication will be presented in this paper

JANUS reactor d and d project

International Nuclear Information System (INIS)

Fellhauer, C. R.

1998-01-01

Argonne National Laboratory (ANL-E) has recently completed the decontamination and decommissioning (D and D) of the JANUS Reactor Facility located in Building 202. The 200 KW reactor operated from August 1963 to March 1992. The facility was used to study the effects of both high and low doses of fission neutrons in animals. There were two exposure rooms on opposite sides of the reactor and the reactor was therefore named after the two-faced Roman god. The High Dose Room was capable of specimen exposure at a dose rate of 3,600 rads per hour. During calendar year 1996 a detailed characterization of the facility was performed by ANL-E Health Physics personnel. ANL-E Analytical Services performed the required sample analysis. An Auditable Safety Analysis and an Environmental Assessment were completed. D and D plans, procedures and procurement documents were prepared and approved. A D and D subcontractor was selected and a firm, fixed price contract awarded for the field work and final survey effort. The D and D subcontractor was mobilized to ANL-E in January 1997. Electrical isolation of all reactor equipment and control panels was accomplished and the equipment removed. A total of 207,230 pounds (94,082 Kg) of lead shielding was removed, surveyed and sampled, and free-released for recycle. All primary and secondary piping was removed, size reduced and packaged for disposal or recycled as appropriate. The reactor vessel was removed, sized reduced and packaged as radioactive waste in April. The activated graphite block reflector was removed next, followed by the bioshield concrete and steel. All of this material was packaged as low level waste. Total low level radioactive waste generation was 4002.1 cubic feet (113.3 cubic meters). Mixed waste generation was 538 cubic feet (15.2 cubic meters). The Final Release Survey was completed in September. The project field work was completed in 38 weeks without any lost-time accidents, personnel contaminations or unplanned

Evaluation of Torsatrons as reactors

International Nuclear Information System (INIS)

Lyon, J.F.; Gulec, K.; Miller, R.L.; El-Guebaly, L.

1994-03-01

Stellarators have significant operational advantages over tokamaks as ignited steady-state reactors. This scoping study, which uses an integrated cost-minimization code that incorporates costing and reactor component models self-consistently with a 1-D energy transport calculation, shows that a torsatron reactor could also be economically competitive with a tokamak reactor. The projected cost of electricity (COE) estimated using the Advanced Reactor Innovation and Evaluation Studies (ARIES) costing algorithms is 65.6 mill/kW(e)h in constant 1992 dollars for a reference 1-GW(e) Compact Torsatron reactor case. The COE is relatively insensitive (<10% variation) over a wide range of assumptions, including variations in the maximum field allowed on the coils, the coil elongation, the shape of the density profile, the beta limit, the confinement multiplier, and the presence of a large loss region for alpha particles. The largest variations in the COE occur for variations in the electrical power output demanded and the plasma-coil separation ratio

Toroidal visco-resistive magnetohydrodynamic steady states contain vortices

International Nuclear Information System (INIS)

Bates, J.W.; Montgomery, D.C.

1998-01-01

Poloidal velocity fields seem to be a fundamental feature of resistive toroidal magnetohydrodynamic (MHD) steady states. They are a consequence of force balance in toroidal geometry, do not require any kind of instability, and disappear in the open-quotes straight cylinderclose quotes (infinite aspect ratio) limit. If a current density j results from an axisymmetric toroidal electric field that is irrotational inside a torus, it leads to a magnetic field B such that ∇x(jxB) is nonvanishing, so that the Lorentz force cannot be balanced by the gradient of any scalar pressure in the equation of motion. In a steady state, finite poloidal velocity fields and toroidal vorticity must exist. Their calculation is difficult, but explicit solutions can be found in the limit of low Reynolds number. Here, existing calculations are generalized to the more realistic case of no-slip boundary conditions on the velocity field and a circular toroidal cross section. The results of this paper strongly suggest that discussions of confined steady states in toroidal MHD must include flows from the outset. copyright 1998 American Institute of Physics

Efficient steady-state solver for hierarchical quantum master equations

Science.gov (United States)

Zhang, Hou-Dao; Qiao, Qin; Xu, Rui-Xue; Zheng, Xiao; Yan, YiJing

2017-07-01

Steady states play pivotal roles in many equilibrium and non-equilibrium open system studies. Their accurate evaluations call for exact theories with rigorous treatment of system-bath interactions. Therein, the hierarchical equations-of-motion (HEOM) formalism is a nonperturbative and non-Markovian quantum dissipation theory, which can faithfully describe the dissipative dynamics and nonlinear response of open systems. Nevertheless, solving the steady states of open quantum systems via HEOM is often a challenging task, due to the vast number of dynamical quantities involved. In this work, we propose a self-consistent iteration approach that quickly solves the HEOM steady states. We demonstrate its high efficiency with accurate and fast evaluations of low-temperature thermal equilibrium of a model Fenna-Matthews-Olson pigment-protein complex. Numerically exact evaluation of thermal equilibrium Rényi entropies and stationary emission line shapes is presented with detailed discussion.

FARST: A computer code for the evaluation of FBR fuel rod behavior under steady-state/transient conditions

International Nuclear Information System (INIS)

Nakamura, M.; Sakagami, M.

1984-01-01

FARST, a computer code for the evaluation of fuel rod thermal and mechanical behavior under steady-state/transient conditions has been developed. The code characteristics are summarized as follows: (I) FARST evaluates the fuel rod behavior under the transient conditions. The code analyzes thermal and mechanical phenomena within a fuel rod, taking into account the temperature change in coolant surrounding the fuel rod. (II) Permanent strains such as plastic, creep and swelling strains as well as thermoelastic deformations can be analyzed by using the strain increment method. (III) Axial force and contact pressure which act on the fuel stack and cladding are analyzed based on the stick/slip conditions. (IV) FARST used a pellet swelling model which depends on the contact pressure between pellet and cladding, and an empirical pellet relocation model, designated as 'jump relocation model'. The code was successfully applied to analyses of the fuel rod irradiation data from pulse reactor for nuclear safety research in Cadarache (CABRI) and pulse reactor for nuclear safety research in Japan Atomic Energy Research Institute (NSRR). The code was further applied to stress analysis of a 1000 MW class large FBR plant fuel rod during transient conditions. The steady-state model which was used so far gave the conservative results for cladding stress during overpower transient, but underestimated the results for cladding stress during a rapid temperature decrease of coolant. (orig.)

Steady state flow evaluations for passive auxiliary feedwater system of APR

International Nuclear Information System (INIS)

Park, Jongha; Kim, Jaeyul; Seong, Hoje; Kang, Kyoungho

2012-01-01

This paper briefly introduces a methodology to evaluate steady state flow of APR+ Passive Auxiliary Feedwater System (PAFS). The PAFS is being developed as a safety grade passive system to completely replace the existing active Auxiliary Feedwater System (AFWS). Natural circulation cooling can be generally classified into the single-phase, two-phase, and boiling-condensation modes. The PAF is designed to be operated in a boiling-condensation natural circulation mode. The steady-state flow rate should be equal to the steady-state boiling/condensation rate determined by the steady-state energy and momentum balances in the PAFS. The determined steady-state flow rate can be used in the design optimization for the natural circulation loop of the PAFS through the steady-state momentum balance. Since the retarding force, which is to be balanced by the driving force in the natural circulation system design depends on the reliable evaluation of the success of a natural circulation system design depends on the reliable evaluation of the pressure loss coefficients. In PAFS, the core decay heat is released by natural circulation flow between the S G secondary side and the Passive Condensation Heat Exchanger (PCHX) that is immersed in the Passive Condensation Cooling Tank (PCCT). The PCCT is located on the top of Auxiliary building The driving force is determined by the difference between the S/G (heat Source) secondary water level and condensation liquid (heat sink) level. It will overcome retarding force at flowrate in the system, which is determined by vaporization and condensation of the steam which is generated at the S/G by the latent heat in system. In this study, the theoretical method to estimate the steady state flow rate in boiling-condensation natural circulation system is developed and compared with test results

Einstein's steady-state theory: an abandoned model of the cosmos

Science.gov (United States)

O'Raifeartaigh, Cormac; McCann, Brendan; Nahm, Werner; Mitton, Simon

2014-09-01

We present a translation and analysis of an unpublished manuscript by Albert Einstein in which he attempted to construct a `steady-state' model of the universe. The manuscript, which appears to have been written in early 1931, demonstrates that Einstein once explored a cosmic model in which the mean density of matter in an expanding universe is maintained constant by the continuous formation of matter from empty space. This model is very different to previously known Einsteinian models of the cosmos (both static and dynamic) but anticipates the later steady-state cosmology of Hoyle, Bondi and Gold in some ways. We find that Einstein's steady-state model contains a fundamental flaw and suggest that it was abandoned for this reason. We also suggest that he declined to explore a more sophisticated version because he found such theories rather contrived. The manuscript is of historical interest because it reveals that Einstein debated between steady-state and evolving models of the cosmos decades before a similar debate took place in the cosmological community.

LOFT shield tank steady state temperatures with addition of gamma and neutron shielding

International Nuclear Information System (INIS)

Kyllingstad, G.

1977-01-01

The effect of introducing a neutron and gamma shield into the annulus between the reactor vessel and the shield tank is analyzed. This addition has been proposed in order to intercept neutron streaming up the annulus during nuclear operations. Its installation will require removal of approximately 20- 1 / 2 inches of stainless steel foil insulation at the top of the annulus. The resulting conduction path is believed to result in increased water temperatures within the shield tank, possibly beyond the 150 0 F limit, and/or cooling of the reactor vessel nozzles such that adverse thermal stresses would be generated. A two dimensional thermal analysis using the finite element code COUPLE/MOD2 was done for the shield tank system illustrated in the figure (1). The reactor was assumed to be at full power, 55 MW (th), with a loop flow rate of 2.15 x 10 6 lbm/hr (268.4 kg/s) at 2250 psi (15.51 MPa). Calculations indicate a steady state shield tank water temperature of 140 0 F (60 0 C). This is below the 150 0 F (65.56 0 C) limit. Also, no significant changes in thermal gradients within the nozzle or reactor vessel wall are generated. A spacer between the gamma shield and the shield tank is recommended, however, in order to ensure free air circulation through the annulus

Flt3 ligand-receptor interaction is important for maintenance of early thymic progenitor numbers in steady-state thymopoiesis.

Science.gov (United States)

Kenins, Linda; Gill, Jason W; Holländer, Georg A; Wodnar-Filipowicz, Aleksandra

2010-01-01

T-cell production throughout life depends on efficient colonization and intrathymic expansion of BM-derived hematopoietic precursors. After irradiation-induced thymic damage, thymic recovery is facilitated by Flt3 ligand (FL), expressed by perivascular fibroblasts surrounding the thymic entry site of Flt3 receptor-positive progenitor cells. Whether intrathymic FL-Flt3 interactions play a role in steady-state replenishment of T cells remains unknown. Here, using competitive BM transplantation studies and fetal thymic organ cultures we demonstrated the continued numerical advantage of Flt3+ intrathymic T-cell precursors. Sub-kidney capsule thymic transplantation experiments, in which WT and FL-/- thymic lobes were grafted into FL-/- recipients, revealed that FL expression by the thymic microenvironment plays a role in steady-state thymopoiesis. The deficiency of the most immature thymic T-cell precursors correlated to upregulation of FL by thymic MTS15+ fibroblasts, suggesting that the number of Flt3+ progenitor cells may regulate the thymic expression of this cytokine. Together, these results show that FL expression by thymic stromal fibroblasts interacting with Flt3+ T-cell progenitors is important for the physiological maintenance of early T-cell development.

Primary side crud sampling and characterization: how it may help to better maintain CANDU reactors

International Nuclear Information System (INIS)

Sawicki, J.A.; Dymarski, M.; Sawicka, B.D.

1997-01-01

The paper reports a corrosion-product transport study in the primary-side of CANDU reactors. The study is based on systematic crud sampling examinations at Darlington NGS. Emphasis is placed on the corrosion-product transport and oxidation state as a monitor of primary water chemistry state, and as a monitor of system corrosion effects. The D 2 O was sampled at reactor outlet headers and the deposits collected on 0.45 μm membrane filters were analyzed by using γ-ray spectroscopy, X-ray fluorescence and Moessbauer effect techniques. In the beginning of this study in 1994 and 1995, the steady state crude concentration used to be as high as 20 to 30 μg/kg D 2 O and the crud contained large fractions of highly oxidized species (40 to 60% Fe), in the form of ferrihydrite, lepidocrocite, goethite and hematite. During start-up operations, the crud concentrations were as high as 8 mg/kg D 2 O and contained 70 to 80% Fe in the form of oxidized iron species. However, as a result of various improvements, and especially because of better oxygen control during lay-ups (nitrogen blanketing) and better pH control (lower and more stable values), in recent steady-state runs, the crud concentrations were found to be lower than 3 to 5 μg/kg D 2 O. At present, the steady-state crud consists mostly of Fe, largely in the form of magnetite (60 to 80% Fe), and hematite (20 to 40% Fe). The steady-state concentration of crud was found to increase markedly with pH a in the range from 10.3 to 10.7. This seems to be in accord with the trend predicted by flow-assisted solubility of magnetite in lithiated water at ∼ 300 degrees C, and is generally in accord with the model of dissolution, transport, and redeposition of iron in the primary heat transport system of CANDU reactors. The crud contains also significant amounts of Zr (∼ 5 to 30 wt. % Fe), apparently in the form of zirconium oxide. Zirconium oxide particles may originate largely from the wear of fuel bundle pads and pressure tube

CFD Analysis of the Primary Cooling System for the Small Modular Natural Circulation Lead Cooled Fast Reactor SNRLFR-100

Directory of Open Access Journals (Sweden)

Pengcheng Zhao

2016-01-01

Full Text Available Small modular reactor (SMR has drawn wide attention in the past decades, and Lead cooled fast reactor (LFR is one of the most promising advanced reactors which are able to meet the safety economic goals of Gen-IV nuclear energy systems. A small modular natural circulation lead cooled fast reactor-100 MWth (SNRLFR-100 is being developed by University of Science and Technology of China (USTC. In the present work, a 3D CFD model, primary heat exchanger model, fuel pin model, and point kinetic model were established based on some reasonable simplifications and assumptions, the steady-state natural circulation characteristics of SNCLFR-100 primary cooling system were discussed and illustrated, and some reasonable suggestions were proposed for the reactor’s thermal-hydraulic and structural design. Moreover, in order to have a first evaluation of the system behavior in accident conditions, an unprotected loss of heat sink (ULOHS transient simulation at beginning of the reactor cycle (BOC has been analyzed and discussed based on the steady-state simulation results. The key temperatures of the reactor core are all under the safety limits at transient state; the reactor has excellent thermal-hydraulic performance.

Superconducting magnets and cryogenics for the steady state superconducting tokamak SST-1

International Nuclear Information System (INIS)

Saxena, Y.C.

2000-01-01

SST-1 is a steady state superconducting tokamak for studying the physics of the plasma processes in tokamak under steady state conditions and to learn technologies related to the steady state operation of the tokamak. SST-1 will have superconducting magnets made from NbTi based conductors operating at 4.5 K temperature. The design of the superconducting magnets and the cryogenic system of SST-1 tokamak are described. (author)

On steady-state concentrations of ammonia and molecular hydrogen in the primary circuit of the WWER-1000 reactors

International Nuclear Information System (INIS)

Arkhipov, O.P.; Bugaenko, V.L.; Kamakchi, S.A.

1997-01-01

It is shown that the MORAVA-N2 software package describes well the coolant state in the primary circuit of an actual reactor facility with the WWER-1000 during on-load operation. It permits using the package for analysis of process perturbation effect on the coolant composition. Specific feature of ammonia radiation chemistry in the primary circuit of a reactor facility with the WWER-1000, assuring the rates hydrogen concentration in the coolant with ammonia concentration variation in the coolant within wide limits, when reactor operates on power, can be mentioned by way of example, the fact being ascertained in this study

Quasi-steady State Reduction of Molecular Motor-Based Models of Directed Intermittent Search

KAUST Repository

Newby, Jay M.

2010-02-19

We present a quasi-steady state reduction of a linear reaction-hyperbolic master equation describing the directed intermittent search for a hidden target by a motor-driven particle moving on a one-dimensional filament track. The particle is injected at one end of the track and randomly switches between stationary search phases and mobile nonsearch phases that are biased in the anterograde direction. There is a finite possibility that the particle fails to find the target due to an absorbing boundary at the other end of the track. Such a scenario is exemplified by the motor-driven transport of vesicular cargo to synaptic targets located on the axon or dendrites of a neuron. The reduced model is described by a scalar Fokker-Planck (FP) equation, which has an additional inhomogeneous decay term that takes into account absorption by the target. The FP equation is used to compute the probability of finding the hidden target (hitting probability) and the corresponding conditional mean first passage time (MFPT) in terms of the effective drift velocity V, diffusivity D, and target absorption rate λ of the random search. The quasi-steady state reduction determines V, D, and λ in terms of the various biophysical parameters of the underlying motor transport model. We first apply our analysis to a simple 3-state model and show that our quasi-steady state reduction yields results that are in excellent agreement with Monte Carlo simulations of the full system under physiologically reasonable conditions. We then consider a more complex multiple motor model of bidirectional transport, in which opposing motors compete in a "tug-of-war", and use this to explore how ATP concentration might regulate the delivery of cargo to synaptic targets. © 2010 Society for Mathematical Biology.

α-particle D-state effects in (d,α) reactions

International Nuclear Information System (INIS)

Santos, F.D.; Tonsfelt, S.A.; Clegg, T.B.; Ludwig, E.J.; Tagishi, Y.; Wilkerson, J.F.

1982-01-01

It is shown that the tensor analyzing powers for (d,α) reactions are sensitive to the D-state component in the α-particle wave function. The D to S-state asymptotic ratio extracted from T 20 and T 22 data in J = L +- 1 transitions is discussed using an α-particle D state generated with the Jackson and Riska model

Steady-state models in electrophoresis: from isotachophoresis to capillary zone electrophoresis

NARCIS (Netherlands)

Beckers, J.L.

1995-01-01

Although all electrophoretic techniques are closely allied and controlled by the same rules, we often distinguish between steady-state and dynamic models in the modeling of electrophoretic processes, whereby steady-state models are applied for isotachophoresis (ITP) and dynamic models are applied

Indirect MRI of 17 o-labeled water using steady-state sequences: Signal simulation and preclinical experiment.

Science.gov (United States)

Kudo, Kohsuke; Harada, Taisuke; Kameda, Hiroyuki; Uwano, Ikuko; Yamashita, Fumio; Higuchi, Satomi; Yoshioka, Kunihiro; Sasaki, Makoto

2018-05-01

Few studies have been reported for T 2 -weighted indirect 17 O imaging. To evaluate the feasibility of steady-state sequences for indirect 17 O brain imaging. Signal simulation, phantom measurements, and prospective animal experiments were performed in accordance with the institutional guidelines for animal experiments. Signal simulations of balanced steady-state free precession (bSSFP) were performed for concentrations of 17 O ranging from 0.037-1.600%. Phantom measurements with concentrations of 17 O water ranging from 0.037-1.566% were also conducted. Six healthy beagle dogs were scanned with intravenous administration of 20% 17 O-labeled water (1 mL/kg). Dynamic 3D-bSSFP scans were performed at 3T MRI. 17 O-labeled water was injected 60 seconds after the scan start, and the total scan duration was 5 minutes. Based on the result of signal simulation and phantom measurement, signal changes in the beagle dogs were measured and converted into 17 O concentrations. The 17 O concentrations were averaged for every 15 seconds, and compared to the baseline (30-45 sec) with Dunnett's multiple comparison tests. Signal simulation revealed that the relationships between 17 O concentration and the natural logarithm of relative signals were linear. The intraclass correlation coefficient between relative signals in phantom measurement and signal simulations was 0.974. In the animal experiments, significant increases in 17 O concentration (P O. At the end of scanning, mean respective 17 O concentrations of 0.084 ± 0.026%, 0.117 ± 0.038, 0.082 ± 0.037%, and 0.049 ± 0.004% were noted for the cerebral cortex, cerebellar cortex, cerebral white matter, and ventricle. Dynamic steady-state sequences were feasible for indirect 17 O imaging, and absolute quantification was possible. This method can be applied for the measurement of permeability and blood flow in the brain, and for kinetic analysis of cerebrospinal fluid. 2 Technical Efficacy: Stage 1 J. Magn. Reson

Calculation of power density with MCNP in TRIGA reactor

International Nuclear Information System (INIS)

Snoj, L.; Ravnik, M.

2006-01-01

Modern Monte Carlo codes (e.g. MCNP) allow calculation of power density distribution in 3-D geometry assuming detailed geometry without unit-cell homogenization. To normalize MCNP calculation by the steady-state thermal power of a reactor, one must use appropriate scaling factors. The description of the scaling factors is not adequately described in the MCNP manual and requires detailed knowledge of the code model. As the application of MCNP for power density calculation in TRIGA reactors has not been reported in open literature, the procedure of calculating power density with MCNP and its normalization to the power level of a reactor is described in the paper. (author)

On possibility of vibrational stabilization in nuclear instable stationary regimes in nuclear reactors

International Nuclear Information System (INIS)

Trakhtenberg, A.M.

1987-01-01

A principle possibility of applying the vibrational stabilization method to nuclear reactors is studied. The problem of securing the stability of nuclear reactor operation steady-state regimes is one of the central ones in dynamics theory and nuclear reaction operation experience. In particular, the problem of xenon oscillation suppressing in a reactor, occuring as a result of steady-state regime instability is urgent. Investigation is conducted using the simpliest reactor model, repesenting it as a non-linear object with concentrated parameters. It is proved that vibrational stabilization is achieved by periodic fluctuations of the control rod positions in the reactor core and boric acid concentration in the coolant with period 1s 4 s. In practice stabilization is effective, when the steady-state regime is located near the stability boundary, which appears to be dangerous, i.e. self-oscillations with inadmissibly high amplitude occure in the reactor

Steady-State Clozapine and Norclozapine Pharmacokinetics in Maori and European Patients.

Science.gov (United States)

Menkes, David B; Glue, Paul; Gale, Christopher; Lam, Frederic; Hung, Cheung-Tak; Hung, Noelyn

2018-01-01

Clozapine is the most effective drug for treatment-resistant schizophrenia, but its use is limited by toxicity. Because ethnicity has been reported to affect clozapine metabolism, we compared its steady state pharmacokinetics in New Zealand Maori and European patients. Clozapine and norclozapine steady state bioavailability was assessed over 24h under fasting and fed conditions in 12 Maori and 16 European patients treated for chronic psychotic illnesses with stable once-daily clozapine doses. Plasma clozapine and norclozapine concentrations were assessed using liquid chromatography with tandem mass spectrometry; pharmacokinetic parameters were calculated using standard non-compartmental methods, and compared using unpaired t-tests. Mean pharmacokinetic parameters (AUC, C max and C min ) for clozapine and norclozapine were virtually identical in Maori and European subjects, under both fed and fasted conditions. Clozapine bioavailability does not vary between Maori and European patients, and thus does not need to be considered in prescribing decisions. Additional studies are needed to identify if there are differences between Maori and European populations for drugs metabolized by other enzyme pathways. Copyright © 2017 The Authors. Published by Elsevier B.V. All rights reserved.

X-Ray Spectral Analysis of the Steady States of GRS1915+105

Science.gov (United States)

Peris, Charith S.; Remillard, Ronald A.; Steiner, James F.; Vrtilek, Saeqa D.; Varnière, Peggy; Rodriguez, Jerome; Pooley, Guy

2016-05-01

We report on the X-ray spectral behavior within the steady states of GRS1915+105. Our work is based on the full data set of the source obtained using the Proportional Counter Array (PCA) on the Rossi X-ray Timing Explorer (RXTE) and 15 GHz radio data obtained using the Ryle Telescope. The steady observations within the X-ray data set naturally separated into two regions in the color-color diagram and we refer to these regions as steady-soft and steady-hard. GRS1915+105 displays significant curvature in the coronal component in both the soft and hard data within the RXTE/PCA bandpass. A majority of the steady-soft observations displays a roughly constant inner disk radius ({R}{{in}}), while the steady-hard observations display an evolving disk truncation which is correlated to the mass accretion rate through the disk. The disk flux and coronal flux are strongly correlated in steady-hard observations and very weakly correlated in the steady-soft observations. Within the steady-hard observations, we observe two particular circumstances when there are correlations between the coronal X-ray flux and the radio flux with log slopes η ˜ 0.68+/- 0.35 and η ˜ 1.12+/- 0.13. They are consistent with the upper and lower tracks of Gallo et al. (2012), respectively. A comparison of the model parameters to the state definitions shows that almost all of the steady-soft observations match the criteria of either a thermal or steep power-law state, while a large portion of the steady-hard observations match the hard-state criteria when the disk fraction constraint is neglected.

Implications of rf current drive theory for next step steady-state tokamak design

International Nuclear Information System (INIS)

Schultz, J.H.

1985-06-01

Two missions have been identified for a next-step tokamak experiment in the United States. The more ambitious Mission II device would be a superconducting tokamak, capable of doing long-pulse ignition demonstrations, and hopefully capable of also being able to achieve steady-state burn. A few interesting lines of approach have been identified, using a combination of logical design criteria and parametric system scans [SC85]. These include: (1) TIBER: A point-design suggested by Lawrence Livermore, that proposes a machine with the capability of demonstrating ignition, high beta (10%) and high Q (=10), using high frequency, fast-wave current drive. The TIBER topology uses moderate aspect ratio and high triangularity to achieve high beta. (2) JET Scale-up. (3) Magic5: It is argued here that an aspect ratio of 5 is a magic number for a good steady-state current drive experiment. A moderately-sized machine that achieves ignition and is capable of high Q, using either fast wave or slow wave current drive is described. (4) ET-II: The concept of a highly elongated tokamak (ET) was first proposed as a low-cost approach to Mission I, because of the possibility of achieving ohmic ignition with low-stress copper magnets. We propose that its best application is really for commercial tokamaks, using fast-wave current drive, and suggest a Mission II experiment that would be prototypical of such a reactor

Synergism of the method of characteristic, R-functions and diffusion solution for accurate representation of 3D neutron interactions in research reactors using the AGENT code system

International Nuclear Information System (INIS)

Hursin, Mathieu; Xiao Shanjie; Jevremovic, Tatjana

2006-01-01

This paper summarizes the theoretical and numerical aspects of the AGENT code methodology accurately applied for detailed three-dimensional (3D) multigroup steady-state modeling of neutron interactions in complex heterogeneous reactor domains. For the first time we show the fine-mesh neutron scalar flux distribution in Purdue research reactor (that was built over forty years ago). The AGENT methodology is based on the unique combination of the three theories: the method of characteristics (MOC) used to simulate the neutron transport in two-dimensional (2D) whole core heterogeneous calculation, the theory of R-functions used as a mathematical tool to describe the true geometry and fuse with the MOC equations, and one-dimensional (1D) higher-order diffusion correction of 2D transport model to account for full 3D heterogeneous whole core representation. The synergism between the radial 2D transport and the 1D axial transport (to take into account the axial neutron interactions and leakage), called the 2D/1D method (used in DeCART and CHAPLET codes), provides a 3D computational solution. The unique synergism between the AGENT geometrical algorithm capable of modeling any current or future reactor core geometry and 3D neutron transport methodology is described in details. The 3D AGENT accuracy and its efficiency are demonstrated showing the eigenvalues, point-wise flux and reaction rate distributions in representative reactor geometries. The AGENT code, comprising this synergism, represents a building block of the computational system, called the virtual reactor. Its main purpose is to perform 'virtual' experiments and demonstrations of various mainly university research reactor experiments

Left ventricular function assessment using a fast 3D gradient echo pulse sequence: comparison to standard multi-breath hold 2D steady state free precession imaging and accounting for papillary muscles and trabeculations.

Science.gov (United States)

Sievers, Burkhard; Schrader, Sebastian; Rehwald, Wolfgang; Hunold, Peter; Barkhausen, Joerg; Erbel, Raimund

2011-06-01

Papillary muscles and trabeculae for ventricular function analysis are known to significantly contribute to accurate volume and mass measurements. Fast imaging techniques such as three-dimensional steady-state free precession (3D SSFP) are increasingly being used to speed up imaging time, but sacrifice spatial resolution. It is unknown whether 3D SSFP, despite its reduced spatial resolution, allows for exact delineation of papillary muscles and trabeculations. We therefore compared 3D SSFP ventricular function measurements to those measured from standard multi-breath hold two-dimensional steady-state free precession cine images (standard 2D SSFP). 14 healthy subjects and 14 patients with impaired left ventricularfunction underwent 1.5 Tesla cine imaging. A stack of short axis images covering the left ventricle was acquired with 2D SSFP and 3D SSFP. Left ventricular volumes, ejection fraction, and mass were determined. Analysis was performed by substracting papillary muscles and trabeculae from left ventricular volumes. In addition, reproducibility was assessed. EDV, ESV, EF, and mass were not significantly different between 2D SSFP and 3D SSFP (mean difference healthy subjects: -0.06 +/- 3.2 ml, 0.54 +/- 2.2 ml, -0.45 +/- 1.8%, and 1.13 +/- 0.8 g, respectively; patients: 1.36 +/- 2.8 ml, -0.15 3.5 ml, 0.86 +/- 2.5%, and 0.91 +/- 0.9 g, respectively; P > or = 0.095). Intra- and interobserver variability was not different for 2D SSFP (P > or = 0.64 and P > or = 0.397) and 3D SSFP (P > or = 0.53 and P > or = 0.47). Differences in volumes, EF, and mass measurements between 3D SSFP and standard 2D SSFP are very small, and not statistically significant. 3D SSFP may be used for accurate ventricular function assessment when papillary muscles and trabeculations are to be taken into account.

PROMETHEUS - a code system for dynamic 3-D analysis of nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Khotylev, V.A.; Hoogenboom, J.E.; Leege, P.F.A. de [Technische Univ. Delft (Netherlands). Interfacultair Reactor Inst.

1996-09-01

The paper presents a multidimensional, general-purpose neutronics code system. It solves a number of steady-state and/or transient problems with coupled thermal hydraulics in one-, two-, or three-dimensional geometry. Due to a number of specialized features such as cavity treatment, automated convergence control, burnup treatment using the full isotopic transition matrix, the code system can be applied for the analysis of fast and slow transients in small, large, and innovative reactor cores. (author)

Tokamak reactor studies

International Nuclear Information System (INIS)

Baker, C.C.

1981-01-01

This paper presents an overview of tokamak reactor studies with particular attention to commercial reactor concepts developed within the last three years. Emphasis is placed on DT fueled reactors for electricity production. A brief history of tokamak reactor studies is presented. The STARFIRE, NUWMAK, and HFCTR studies are highlighted. Recent developments that have increased the commercial attractiveness of tokamak reactor designs are discussed. These developments include smaller plant sizes, higher first wall loadings, improved maintenance concepts, steady-state operation, non-divertor particle control, and improved reactor safety features

Physical design of MW-class steady-state spherical tokamak, QUEST

International Nuclear Information System (INIS)

Hanada, K.; Sato, K.N.; Zushi, H.; Nakamura, K.; Sakamoto, M.; Idei, H.; Hasegawa, M.; Kawasaki, S.; Nakashima, H.; Higashijima, A.; Higashizono, Y.; Yoshida, N.; Takase, Y.; Ejiri, A.; Ogawa, Y.; Ono, Y.; Yoshida, Z.; Mitarai, O.; Maekawa, T.; Kishimoto, Y.; Ishiguro, M.; Yoshinaga, T.; Igami, H.; Hirooka, Y.; Komori, A.; Motojima, O.; Sudo, S.; Yamada, H.; Ando, A.; Asakura, Nobuyuki; Matsukawa, Makoto; Ishida, A.; Ohno, N.; Peng, M.

2008-10-01

QUEST (R=0.68 m, a=0.4 m) focuses on the steady state operation of the spherical tokamak (ST) by controlled PWI and electron Bernstain wave (EBW) current drive (CD). The QUEST project will be developed along two phases, phase I: steady state operation with plasma current, I p =20-30 kA on open divertor configuration and phase II: steady state operation with I p = 100 kA and β of 10% in short pulse on closed divertor configuration. Feasibility of the missions on QUEST was investigated and the suitable machine size of QUEST was decided based on the physical view of plasma parameters. Electron Bernstein wave (EBW) current drive are planned to establish the maintenance of plasma current in steady state. Mode conversion efficiency to EBW was calculated and the conversion of 95% will be expected. A new type antenna for QUEST has been fabricated to excite EBW effectively. The situation of heat and particle handling is challenging, and W and high temperature wall is adopted. The start-up scenario of plasma current was investigated based on the driven current by energetic electron and the most favorable magnetic configuration for start-up is proposed. (author)

Steady-State Core Temperature Prediction Based on GAMMA+/CAPP Coupling

International Nuclear Information System (INIS)

Tak, Nam-il; Lee, Hyun-Chul; Lim, Hong-Sik

2015-01-01

In spite of sizable applications of the GAMMA+ code for the thermo-fluid analysis and design of a prismatic VHTR, the existing works are limited to stand-alone calculations. In the stand-alone calculations, information from the neutronic analysis (e.g., reactor power density profile) was considered only once i.e., when the calculations get started. For the neutronic analysis and design of a VHTR, the CAPP code, which is also under development at KAERI, is used. The main objective of this paper is to investigate the capability of GAMMA+ and CAPP coupling and to examine the results of the coupled analysis. Based on the coupling of GAMMA+ and CAPP, the steady-state core temperature was investigated in this work. It is found that the communication of data was successful. And the results of the GAMMA+ and CAPP coupling are found to be reasonable. The design modification of PMR200 is required to satisfy the design limit for the hot spot fuel temperature

Development of 'low activation superconducting wire' for an advanced fusion reactor

International Nuclear Information System (INIS)

Hishinuma, Y.; Yamada, S.; Sagara, A.; Kikuchi, A.; Takeuchi, T.; Matsuda, K.; Taniguchi, H.

2011-01-01

In the D-T burning plasma reactor beyond ITER such as DEMO and fusion power plants assuming the steady-state and long time operation, it will be necessary to consider carefully induced radioactivity and neutron irradiation properties on the all components for fusion reactors. The decay time of the induced radioactivity can control the schedule and scenarios of the maintenance and shutdown on the fusion reactor. V 3 Ga and MgB 2 compound have shorter decay time within 1 years and they will be desirable as a candidate material to realize 'low activation and high magnetic field superconducting magnet' for advanced fusion reactor. However, it is well known that J c -B properties of V 3 Ga and MgB 2 wires are lower than that of the Nb-based A15 compound wires, so the J c -B enhancements on the V 3 Ga and MgB 2 wires are required in order to apply for an advanced fusion reactor. We approached and succeeded to developing the new process in order to improve J c properties of V 3 Ga and MgB 2 wires. In this paper, the recent activities for the J c improvements and detailed new process in V 3 Ga and MgB 2 wires are investigated. (author)

Basin stability measure of different steady states in coupled oscillators

Science.gov (United States)

Rakshit, Sarbendu; Bera, Bidesh K.; Majhi, Soumen; Hens, Chittaranjan; Ghosh, Dibakar

2017-04-01

In this report, we investigate the stabilization of saddle fixed points in coupled oscillators where individual oscillators exhibit the saddle fixed points. The coupled oscillators may have two structurally different types of suppressed states, namely amplitude death and oscillation death. The stabilization of saddle equilibrium point refers to the amplitude death state where oscillations are ceased and all the oscillators converge to the single stable steady state via inverse pitchfork bifurcation. Due to multistability features of oscillation death states, linear stability theory fails to analyze the stability of such states analytically, so we quantify all the states by basin stability measurement which is an universal nonlocal nonlinear concept and it interplays with the volume of basins of attractions. We also observe multi-clustered oscillation death states in a random network and measure them using basin stability framework. To explore such phenomena we choose a network of coupled Duffing-Holmes and Lorenz oscillators which are interacting through mean-field coupling. We investigate how basin stability for different steady states depends on mean-field density and coupling strength. We also analytically derive stability conditions for different steady states and confirm by rigorous bifurcation analysis.

Molecular control of steady-state dendritic cell maturation and immune homeostasis.

Science.gov (United States)

Hammer, Gianna Elena; Ma, Averil

2013-01-01

Dendritic cells (DCs) are specialized sentinels responsible for coordinating adaptive immunity. This function is dependent upon coupled sensitivity to environmental signs of inflammation and infection to cellular maturation-the programmed alteration of DC phenotype and function to enhance immune cell activation. Although DCs are thus well equipped to respond to pathogens, maturation triggers are not unique to infection. Given that immune cells are exquisitely sensitive to the biological functions of DCs, we now appreciate that multiple layers of suppression are required to restrict the environmental sensitivity, cellular maturation, and even life span of DCs to prevent aberrant immune activation during the steady state. At the same time, steady-state DCs are not quiescent but rather perform key functions that support homeostasis of numerous cell types. Here we review these functions and molecular mechanisms of suppression that control steady-state DC maturation. Corruption of these steady-state operatives has diverse immunological consequences and pinpoints DCs as potent drivers of autoimmune and inflammatory disease.

Development of a version of the reactor dynamics code DYN3D applicable for High Temperature Reactors; Entwicklung einer Version des Reaktordynamikcodes DYN3D fuer Hochtemperaturreaktoren. Abschlussbericht

Energy Technology Data Exchange (ETDEWEB)

Rohde, Ulrich; Apanasevich, Pavel; Baier, Silvio; Duerigen, Susan; Fridman, Emil; Grahn, Alexander; Kliem, Soeren; Merk, Bruno

2012-07-15

Based on the reactor dynamics code DYN3D for the simulation of transient processes in Light Water Reactors, a code version DYN3D-HTR for application to graphitemoderated, gas-cooled block-type high temperature reactors has been developed. This development comprises: - the methodical improvement of the 3D steady-state neutron flux calculation for the hexagonal geometry of the HTR fuel element blocks - the development of methods for the generation of homogenised cross section data taking into account the double heterogeneity of the fuel element block structure - the implementation of a 3D model for heat conduction and heat transport in the graphite matrix. The nodal method for neutron flux calculation based on SP3 transport approximation was extended to hexagonal fuel element geometry, where the hexagons are subdivided into triangles, thus the method had finally to be derived for triangular geometry. In triangular geometry, a subsequent subdivision of the hexagonal elements can be considered, and therefore, the effect of systematic mesh refinement can be studied. The algorithm was verified by comparison with Monte Carlo reference solutions, on the node-wise level, as well as also on the pin-wise level. New procedures were developed for the homogenization of the double-heterogeneous fuel element structures. One the one hand, the so-called Reactivity equivalent Physical Transformation (RPT), the two-step homogenization method based on 2D deterministic lattice calculations, was extended to cells with different temperatures of the materials. On the other hand, the progress in development of Monte Carlo methods for spectral calculations, in particular the development of the code SERPENT, opened a new, fully consistent 3D approach, where all details of the structures on fuel particle, fuel compact and fuel block level can be taken into account within one step. Moreover, a 3D heat conduction and heat transport model was integrated into DYN3D to be able to simulate radial

Dark Entangled Steady States of Interacting Rydberg Atoms

DEFF Research Database (Denmark)

Dasari, Durga; Mølmer, Klaus

2013-01-01

their short-lived excited states lead to rapid, dissipative formation of an entangled steady state. We show that for a wide range of physical parameters, this entangled state is formed on a time scale given by the strengths of coherent Raman and Rabi fields applied to the atoms, while it is only weakly...

Tailored parameter optimization methods for ordinary differential equation models with steady-state constraints.

Science.gov (United States)

Fiedler, Anna; Raeth, Sebastian; Theis, Fabian J; Hausser, Angelika; Hasenauer, Jan

2016-08-22

Ordinary differential equation (ODE) models are widely used to describe (bio-)chemical and biological processes. To enhance the predictive power of these models, their unknown parameters are estimated from experimental data. These experimental data are mostly collected in perturbation experiments, in which the processes are pushed out of steady state by applying a stimulus. The information that the initial condition is a steady state of the unperturbed process provides valuable information, as it restricts the dynamics of the process and thereby the parameters. However, implementing steady-state constraints in the optimization often results in convergence problems. In this manuscript, we propose two new methods for solving optimization problems with steady-state constraints. The first method exploits ideas from optimization algorithms on manifolds and introduces a retraction operator, essentially reducing the dimension of the optimization problem. The second method is based on the continuous analogue of the optimization problem. This continuous analogue is an ODE whose equilibrium points are the optima of the constrained optimization problem. This equivalence enables the use of adaptive numerical methods for solving optimization problems with steady-state constraints. Both methods are tailored to the problem structure and exploit the local geometry of the steady-state manifold and its stability properties. A parameterization of the steady-state manifold is not required. The efficiency and reliability of the proposed methods is evaluated using one toy example and two applications. The first application example uses published data while the second uses a novel dataset for Raf/MEK/ERK signaling. The proposed methods demonstrated better convergence properties than state-of-the-art methods employed in systems and computational biology. Furthermore, the average computation time per converged start is significantly lower. In addition to the theoretical results, the

A design of steady state fusion burner

International Nuclear Information System (INIS)

Hasegawa, Akira; Hatori, Tadatsugu; Itoh, Kimitaka; Ikuta, Takashi; Kodama, Yuji.

1975-01-01

We present a brief design of a steady state fusion burner in which a continuous burning of nuclear fuel may be achieved with output power of a gigawatt. The laser fusion is proposed to ignite the fuel. (auth.)

Test-bench for characterization of steady state magnetic sensors parameters in wide temperature range

International Nuclear Information System (INIS)

Kovařík, Karel; Ďuran, Ivan; Sentkerestiová, Jana; Šesták, David

2013-01-01

Highlights: •Prepared test bench for calibration of steady state magnetic sensors. •Test-bench design optimized for calibration up to 300 °C. •Test-bench is remotely controllable and allows long term measurements. •Construction allows easy manipulation with even irradiated samples. -- Abstract: Magnetic sensors in ITER tokamak and in other future fusion devices will face an environment with temperature often elevated well above 200 °C. Dedicated test benches are needed to allow characterization of performance of magnetic sensors at such elevated temperatures. This contribution describes realization of test bench for calibration of steady state magnetic sensors based on Hall effect. The core of the set-up is the coil providing DC calibration magnetic field. Optimization of coils design to ensure its compatibility with elevated temperature up to 300 °C is described. Optimized coil was manufactured, and calibrated both at room temperature and at temperature of 250 °C. Measured calibration magnetic field of the coil biased by a 30 A commercial laboratory power supplies is 224 mT. The coil is supplemented by PID regulated air cooling system for fine control of sensors temperature during measurements. Data acquisition system is composed from PC A/D converter boards with resolution below 1 μV. The key parameters of the test bench are remotely controllable and the system allows long term continuous measurements including tests of irradiated samples. The performance of the test bench is demonstrated on recent measurements with metal Hall sensors based on thin copper sensing layers

The OECD/NEA/NSC PBMR400 MW coupled neutronics thermal hydraulics transient benchmark - Steady-state results and status

International Nuclear Information System (INIS)

Reitsma, F.; Han, J.; Ivanov, K.; Sartori, E.

2008-01-01

The PBMR is a High-Temperature Gas-cooled Reactor (HTGR) concept developed to be built in South Africa. The analysis tools used for core neutronic design and core safety analysis need to be verified and validated. Since only a few pebble-bed HTR experimental facilities or plant data are available the use of code-to-code comparisons are an essential part of the V and V plans. As part of this plan the PBMR 400 MW design and a representative set of transient cases is defined as an OECD benchmark. The scope of the benchmark is to establish a series of well-defined multi-dimensional computational benchmark problems with a common given set of cross-sections, to compare methods and tools in coupled neutronics and thermal hydraulics analysis with a specific focus on transient events. The OECD benchmark includes steady-state and transients cases. Although the focus of the benchmark is on the modelling of the transient behaviour of the PBMR core, it was also necessary to define some steady-state cases to ensure consistency between the different approaches before results of transient cases could be compared. This paper describes the status of the benchmark project and shows the results for the three steady state exercises defined as a standalone neutronics calculation, a standalone thermal-hydraulic core calculation, and a coupled neutronics/thermal-hydraulic simulation. (authors)

Complex nonlinear behaviour of a fixed bed reactor with reactant recycle

DEFF Research Database (Denmark)

Recke, Bodil; Jørgensen, Sten Bay

1999-01-01

The fixed bed reactor with reactant recycle investigated in this paper can exhibit periodic solutions. These solutions bifurcate from the steady state in a Hopf bifurcation. The Hopf bifurcation encountered at the lowest value of the inlet concentration turns the steady state unstable and marks......,that the dynamic behaviour of a fixed bed reactor with reactant recycle is much more complex than previously reported....

Diffusion in coronas around clinopyroxene: modelling with local equilibrium and steady state, and a non-steady-state modification to account for zoned actinolite-hornblende

Science.gov (United States)

Ashworth, J. R.; Birdi, J. J.; Emmett, T. F.

1992-01-01

evolution of actinolite to more stable aluminous compositions. This is modelled by a non-steady-state modification of the theory, retaining local equilibrium in grain boundaries while relatively steep zoning profiles develop in grain interiors through slow intracrystalline diffusion. Replacement of actinolite by hornblende does not require a change in P- T conditions if actinolite is a kinetically determined, non-equilibrium product. The common preservation of a sharp contact between hornblende and actionolite layers may be explained by ineffectiveness of intracrystalline diffusion: according to the theory, given sufficient grain-boundary Al flux, a metastable actinolite + quartz layer in contact with hornblende may be diffusionally stable and may continue to grow in a steady state.

Steady State Dynamic Operating Behavior of Universal Motor

Directory of Open Access Journals (Sweden)

Muhammad Khan Burdi

2015-01-01

Full Text Available A detailed investigation of the universal motor is developed and used for various dynamic steady state and transient operating conditions of loads. In the investigation, output torque, motor speed, input current, input/output power and efficiency are computed, compared and analyzed for different loads. While this paper discusses the steady-state behavior of the universal motor, another companion paper, ?Transient dynamic behavior of universal motor?, will discuss its transient behavior in detail. A non-linear generalized electric machine model of the motor is considered for the analysis. This study was essential to investigate effect of output load on input current, power, speed and efficiency of the motor during operations. Previously such investigation is not known

Quasi-steady state natural convection in a tilted porous layer

Energy Technology Data Exchange (ETDEWEB)

Robillard, L.; Vasseur, P. (Ecole Polytechnique, Montreal, PQ (Canada))

1992-12-01

Natural convection in an inclined porous layer heated or cooled on one side, when its other walls are insulated, has several important engineering applications. These include solar power collection, regenerative heat exchangers, and high performance insulation for buildings and cold storage. Although the problem is basically an unsteady state one, it is known that if the heating (or cooling) process is maintained for a sufficiently long time, a quasi-steady state is approached. Quasi-steady state laminar natural convection in an inclined porous layer is studied analytically and numerically. On the basis of the Darcy-Oberbeck-Boussinesq equations, the problem is solved analytically in the limit of a thin porous layer heated on one side by a heat flux while the other boundaries are maintained adiabatic. For quasi-steady state, the flow and temperature fields overall heat transfer rates are obtained in terms of the controlling parameters and the onset of convection in a bottom heated horizantal system is predicted. It is also demonstrated for the case of a bottom-heated layer that for sufficiently small inclinations, multiple unicellular quasi-steady states exist, some of which are unstable. A numerical study of the same phenomenon, obtained by solving the complete set of governing equations, is conducted. Good agreement is found between the analytical predictions and the numerical simulation. 22 refs., 6 figs.

On Steady-State Tropical Cyclones

Science.gov (United States)

2014-01-01

Press: London. Marks FD, Black PG, Montgomery MT, Burpee RW. 2008. Structure of the eye and eyewall of Hurricane Hugo (1989). Mon. Weather Rev. 136: 1237... hurricanes ; tropical cyclones; typhoons; steady-state Received 18 April 2013; Revised 25 November 2013; Accepted 29 December 2013; Published online in Wiley...the concept of the ‘mature stage’ of a hurricane vortex. The definition of the ‘mature stage’ is commonly based on the time period in which the maximum

STEADY-STATE RELATIVISTIC STELLAR DYNAMICS AROUND A MASSIVE BLACK HOLE

Energy Technology Data Exchange (ETDEWEB)

Bar-Or, Ben; Alexander, Tal [Department of Particle Physics and Astrophysics, Weizmann Institute of Science, P.O. Box 26, Rehovot 76100 (Israel)

2016-04-01

A massive black hole (MBH) consumes stars whose orbits evolve into the small phase-space volume of unstable orbits, the “loss cone,” which take them into the MBH, or close enough to interact strongly with it. The resulting phenomena, e.g., tidal heating and disruption, binary capture and hyper-velocity star ejection, gravitational wave (GW) emission by inspiraling compact remnants, or hydrodynamical interactions with an accretion disk, can produce observable signatures and thereby reveal the MBH, affect its mass and spin evolution, test strong gravity, and probe stars and gas near the MBH. These continuous stellar loss and resupply processes shape the central stellar distribution. We investigate relativistic stellar dynamics near the loss cone of a non-spinning MBH in steady state, analytically and by Monte Carlo simulations of the diffusion of the orbital parameters. These take into account Newtonian mass precession due to enclosed stellar mass, in-plane precession due to general relativity, dissipation by GW, uncorrelated two-body relaxation, correlated resonant relaxation (RR), and adiabatic invariance due to secular precession, using a rigorously derived description of correlated post-Newtonian dynamics in the diffusion limit. We argue that general maximal entropy considerations strongly constrain the orbital diffusion in steady state, irrespective of the relaxation mechanism. We identify the exact phase-space separatrix between plunges and inspirals, and predict their steady-state rates. We derive the dependence of the rates on the mass of the MBH, show that the contribution of RR in steady state is small, and discuss special cases where unquenched RR in restricted volumes of phase-space may affect the steady state substantially.

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

International Nuclear Information System (INIS)

Ehst, D.A.; Brooks, J.N.; Cha, Y.; Evans, K.; Hassanein, A.M.; Kim, S.; Majumdar, S.; Misra, B.; Stevens, H.C.

1984-05-01

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

Analysis of steady state creep of southeastern New Mexico bedded salt

International Nuclear Information System (INIS)

Herrmann, W.; Wawersik, W.R.; Lauson, H.S.

1980-03-01

Steady state creep rates have been obtained from a large suite of existing experimental creep data relating to bedded rock salt from the Salado formation of S.E. New Mexico. Experimental conditions covered an intermediate temperature range from 22 0 C to 200 0 C, and shear stresses from 1000 psi (7 MPa) to 6000 psi (31 MPa). An expression, based on a single diffusion controlled dislocation climb mechanism, has been found to fit the observed dependence of steady state creep rate on shear stress and temperature, yielding an activation energy of 12 kcal/mole (50 kJ/mole) and a stress exponent of 4.9. Multiple regression analysis revealed a dependence on stratigraphy, but no statistically significant dependence on pressure of specimen size. No consistent dilatancy or compaction associated with steady state creep was found, although some individual specimens dilated or compacted during creep. The steady state creep data were found to agree very well with creep data for both bedded and dome salt from a variety of other locations

Resistive wall mode feedback control in EXTRAP T2R with improved steady-state error and transient response

Science.gov (United States)

Brunsell, P. R.; Olofsson, K. E. J.; Frassinetti, L.; Drake, J. R.

2007-10-01

Experiments in the EXTRAP T2R reversed field pinch [P. R. Brunsell, H. Bergsåker, M. Cecconello et al., Plasma Phys. Control. Fusion 43, 1457 (2001)] on feedback control of m =1 resistive wall modes (RWMs) are compared with simulations using the cylindrical linear magnetohydrodynamic model, including the dynamics of the active coils and power amplifiers. Stabilization of the main RWMs (n=-11,-10,-9,-8,+5,+6) is shown using modest loop gains of the order G ˜1. However, other marginally unstable RWMs (n=-2,-1,+1,+2) driven by external field errors are only partially canceled at these gains. The experimental system stability limit is confirmed by simulations showing that the latency of the digital controller ˜50μs is degrading the system gain margin. The transient response is improved with a proportional-plus-derivative controller, and steady-state error is improved with a proportional-plus-integral controller. Suppression of all modes is obtained at high gain G ˜10 using a proportional-plus-integral-plus-derivative controller.

Chlorine decay under steady and unsteady-state hydraulic conditions

DEFF Research Database (Denmark)

Stoianov, Ivan; Aisopou, Angeliki

2014-01-01

This paper describes a simulation framework for the scale-adaptive hydraulic and chlorine decay modelling under steady and unsteady-state flows. Bulk flow and pipe wall reaction coefficients are replaced with steady and unsteady-state reaction coefficients. An unsteady decay coefficient is defined...... which depends upon the absolute value of shear stress and the rate of change of shear stress for quasi-unsteady and unsteady-state flows. A preliminary experimental and analytical investigation was carried out in a water transmission main. The results were used to model monochloramine decay...... and these demonstrate that the dynamic hydraulic conditions have a significant impact on water quality deterioration and the rapid loss of disinfectant residual. © 2013 The Authors....

Steady States in SIRS Epidemical Model of Mobile Individuals

International Nuclear Information System (INIS)

Zhang Duanming; He Minhua; Yu Xiaoling; Pan Guijun; Sun Hongzhang; Su Xiangying; Sun Fan; Yin Yanping; Li Rui; Liu Dan

2006-01-01

We consider an epidemical model within socially interacting mobile individuals to study the behaviors of steady states of epidemic propagation in 2D networks. Using mean-field approximation and large scale simulations, we recover the usual epidemic behavior with critical thresholds δ c and p c below which infectious disease dies out. For the population density δ far above δ c , it is found that there is linear relationship between contact rate λ and the population density δ in the main. At the same time, the result obtained from mean-field approximation is compared with our numerical result, and it is found that these two results are similar by and large but not completely the same.

Reduction of vascular artifact on T1-weighted images of the brain by using three-dimensional double IR fast spoiled gradient echo recalled acquisition in the steady state (FSPGR) at 3.0 Tesla

International Nuclear Information System (INIS)

Fujiwara, Yasuhiro; Yamaguchi, Isao; Ookoshi, Yusuke; Ootani, Yuriko; Matsuda, Tsuyoshi; Ishimori, Yoshiyuki; Hayashi, Hiroyuki; Miyati, Tosiaki; Kimura, Hirohiko

2007-01-01

The purpose of this study was to decrease vascular artifacts caused by the in-flow effect in three-dimensional inversion recovery prepared fast spoiled gradient recalled acquisition in the steady state (3D IR FSPGR) at 3.0 Tesla. We developed 3D double IR FSPGR and investigated the signal characteristics of the new sequence. The 3D double IR FSPGR sequence uses two inversion pulses, the first for obtaining tissue contrast and the second for nulling vascular signal, which is applied at the time of the first IR period at the neck region. We have optimized scan parameters based on both phantom and in-vivo study. As a result, optimized parameters (1st TI=700 ms, 2nd TI=400 ms) successfully have produced much less vessel signal at reduction than conventional 3D IR FSPGR over a wide imaging range, while preserving the signal-to-noise ratio (SNR) and gray/white matter contrast. Moreover, the decreased artifact was also confirmed by visual inspection of the images obtained in vivo using those parameters. Thus, 3D double IR FSPGR was a useful sequence for the acquisition of T1-weighted images at 3.0 Tesla. (author)

Development of an automated core model for nuclear reactors

International Nuclear Information System (INIS)

Mosteller, R.D.

1998-01-01

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

Thermal properties of reactors and some instabilities

Energy Technology Data Exchange (ETDEWEB)

Hearfield, F.

1979-03-01

A discussion covers the thermal properties of adiabatic reactors and the failure of the reaction rate to increase with increasing temperature due to depletion of reagents, transition to mass transfer control, or reduction of adsorption at catalytic surfaces; non-adiabatic reactors and factors upsetting the balance between heat generation and removal and possibly causing a runaway reaction, including loss of agitation loop circulation, and cooling or heating media; multiple steady states, i.e. multiple balances between heat generation and removal, for a continuous stirred tank reactor and the conditions necessary for stability of a steady state; and the temperature distribution in a tubular reactor, including mechanisms for feedback of heat from downstream to upstream in the reactor, e.g. heat conduction and radiation from hot catalyst, or an added heat exchanger. Three case histories are presented in which reactants accumulated in the reactors and cooling was decreased, permitting the occurrence of violent runaway reactions.

40 CFR Appendix II to Part 1042 - Steady-State Duty Cycles

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 32 2010-07-01 2010-07-01 false Steady-State Duty Cycles II Appendix..., App. II Appendix II to Part 1042—Steady-State Duty Cycles (a) The following duty cycles apply as specified in § 1042.505(b)(1): (1) The following duty cycle applies for discrete-mode testing: E3 mode No...

Developing a Steady-state Kinetic Model for Industrial Scale Semi-Regenerative Catalytic Naphtha Reforming Process

Directory of Open Access Journals (Sweden)

Seif Mohaddecy, R.

2014-05-01

Full Text Available Due to the demand for high octane gasoline as a transportation fuel, the catalytic naphtha reformer has become one of the most important processes in petroleum refineries. In this research, the steady-state modelling of a catalytic fixed-bed naphtha reforming process to predict the momentous output variables was studied. These variables were octane number, yield, hydrogen purity, and temperature of all reforming reactors. To do such a task, an industrial scale semi-regenerative catalytic naphtha reforming unit was studied and modelled. In addition, to evaluate the developed model, the predicted variables i.e. outlet temperatures of reactors, research octane number, yield of gasoline and hydrogen purity were compared against actual data. The results showed that there is a close mapping between the actual and predicted variables, and the mean relative absolute deviation of the mentioned process variables were 0.38 %, 0.52 %, 0.54 %, 0.32 %, 4.8 % and 3.2 %, respectively.

Progress Towards High Performance, Steady-state Spherical Torus

International Nuclear Information System (INIS)

Ono, M.; Bell, M.G.; Bell, R.E.; Bigelow, T.; Bitter, M.; Blanchard, W.; Boedo, J.; Bourdelle, C.; Bush, C.; Choe, W.; Chrzanowski, J.; Darrow, D.S.; Diem, S.J.; Doerner, R.; Efthimion, P.C.; Ferron, J.R.; Fonck, R.J.; Fredrickson, E.D.; Garstka, G.D.; Gates, D.A.; Gray, T.; Grisham, L.R.; Heidbrink, W.; Hill, K.W.; Hoffman, D.; Jarboe, T.R.; Johnson, D.W.; Kaita, R.; Kaye, S.M.; Kessel, C.; Kim, J.H.; Kissick, M.W.; Kubota, S.; Kugel, H.W.; LeBlanc, B.P.; Lee, K.; Lee, S.G.; Lewicki, B.T.; Luckhardt, S.; Maingi, R.; Majeski, R.; Manickam, J.; Maqueda, R.; Mau, T.K.; Mazzucato, E.; Medley, S.S.; Menard, J.; Mueller, D.; Nelson, B.A.; Neumeyer, C.; Nishino, N.; Ostrander, C.N.; Pacella, D.; Paoletti, F.; Park, H.K.; Park, W.; Paul, S.F.; Peng, Y.-K. M.; Phillips, C.K.; Pinsker, R.; Probert, P.H.; Ramakrishnan, S.; Raman, R.; Redi, M.; Roquemore, A.L.; Rosenberg, A.; Ryan, P.M.; Sabbagh, S.A.; Schaffer, M.; Schooff, R.J.; Seraydarian, R.; Skinner, C.H.; Sontag, A.C.; Soukhanovskii, V.; Spaleta, J.; Stevenson, T.; Stutman, D.; Swain, D.W.; Synakowski, E.; Takase, Y.; Tang, X.; Taylor, G.; Timberlake, J.; Tritz, K.L.; Unterberg, E.A.; Von Halle, A.; Wilgen, J.; Williams, M.; Wilson, J.R.; Xu, X.; Zweben, S.J.; Akers, R.; Barry, R.E.; Beiersdorfer, P.; Bialek, J.M.; Blagojevic, B.; Bonoli, P.T.; Carter, M.D.; Davis, W.; Deng, B.; Dudek, L.; Egedal, J.; Ellis, R.; Finkenthal, M.; Foley, J.; Fredd, E.; Glasser, A.; Gibney, T.; Gilmore, M.; Goldston, R.J.; Hatcher, R.E.; Hawryluk, R.J.; Houlberg, W.; Harvey, R.; Jardin, S.C.; Hosea, J.C.; Ji, H.; Kalish, M.; Lowrance, J.; Lao, L.L.; Levinton, F.M.; Luhmann, N.C.; Marsala, R.; Mastravito, D.; Menon, M.M.; Mitarai, O.; Nagata, M.; Oliaro, G.; Parsells, R.; Peebles, T.; Peneflor, B.; Piglowski, D.; Porter, G.D.; Ram, A.K.; Rensink, M.; Rewoldt, G.; Roney, P.; Shaing, K.; Shiraiwa, S.; Sichta, P.; Stotler, D.; Stratton, B.C.; Vero, R.; Wampler, W.R.; Wurden, G.A.

2003-01-01

Research on the Spherical Torus (or Spherical Tokamak) is being pursued to explore the scientific benefits of modifying the field line structure from that in more moderate aspect-ratio devices, such as the conventional tokamak. The Spherical Tours (ST) experiments are being conducted in various U.S. research facilities including the MA-class National Spherical Torus Experiment (NSTX) at Princeton, and three medium-size ST research facilities: Pegasus at University of Wisconsin, HIT-II at University of Washington, and CDX-U at Princeton. In the context of the fusion energy development path being formulated in the U.S., an ST-based Component Test Facility (CTF) and, ultimately a Demo device, are being discussed. For these, it is essential to develop high-performance, steady-state operational scenarios. The relevant scientific issues are energy confinement, MHD stability at high beta (B), noninductive sustainment, ohmic-solenoid-free start-up, and power and particle handling. In the confinement area, the NSTX experiments have shown that the confinement can be up to 50% better than the ITER-98-pby2 H-mode scaling, consistent with the requirements for an ST-based CTF and Demo. In NSTX, CTF-relevant average toroidal beta values bT of up to 35% with the near unity central betaT have been obtained. NSTX will be exploring advanced regimes where bT up to 40% can be sustained through active stabilization of resistive wall modes. To date, the most successful technique for noninductive sustainment in NSTX is the high beta-poloidal regime, where discharges with a high noninductive fraction (∼60% bootstrap current + neutral-beam-injected current drive) were sustained over the resistive skin time. Research on radio-frequency-based heating and current drive utilizing HHFW (High Harmonic Fast Wave) and EBW (Electron Bernstein Wave) is also pursued on NSTX, Pegasus, and CDX-U. For noninductive start-up, the Coaxial Helicity Injection (CHI), developed in HIT/HIT-II, has been adopted

Emergence of advance waves in a steady-state universe

Energy Technology Data Exchange (ETDEWEB)

Hobart, R.H.

1979-10-01

In standard Wheeler-Feynman electrodynamics advanced waves from any source are absolutely canceled by the advanced waves from the absorber responding to that source. The present work shows this cancellation fails over cosmic distances in a steady-state universe. A test of the view proposed earlier, in a paper which assumed failure of cancellation ad hoc, that zero-point fluctuations of the electromagnetic field are such emergent advanced waves, is posed. The view entails anomalous slowing of spontaneous transition rates at longer emission wavelengths; available data go against this, furnishing additional argument against the suspect assumption that the universe is steady-state.

Emergence of advance waves in a steady-state universe

International Nuclear Information System (INIS)

Hobart, R.H.

1979-01-01

In standard Wheeler-Feynman electrodynamics advanced waves from any source are absolutely canceled by the advanced waves from the absorber responding to that source. The present work shows this cancellation fails over cosmic distances in a steady-state universe. A test of the view proposed earlier, in a paper which assumed failure of cancellation ad hoc, that zero-point fluctuations of the electromagnetic field are such emergent advanced waves, is posed. The view entails anomalous slowing of spontaneous transition rates at longer emission wavelengths; available data go against this, furnishing additional argument against the suspect assumption that the universe is steady-state

Study on Off-Design Steady State Performances of Helium Gas Turbo-compressor for HTGR-GT

International Nuclear Information System (INIS)

Qisen Ren; Xiaoyong Yang; Zhiyong Huang; Jie Wang

2006-01-01

The high temperature gas-cooled reactor (HTGR) coupled with direct gas turbine cycle is a promising concept in the future of nuclear power development. Both helium gas turbine and compressor are key components in the cycle. Under normal conditions, the mode of power adjustment is to control total helium mass in the primary loop using gas storage vessels. Meanwhile, thermal power of reactor core is regulated. This article analyzes off-design performances of helium gas turbine and compressors for high temperature gas-cooled reactor with gas turbine cycle (HTGR-GT) at steady state level of electric power adjustment. Moreover, performances of the cycle were simply discussed. Results show that the expansion ratio of turbine decreases as electric power reduces but the compression ratios of compressors increase, efficiencies of both turbine and compressors decrease to some extent. Thermal power does not vary consistently with electric power, the difference between these two powers increases as electric power reduces. As a result of much thermal energy dissipated in the temperature modulator set at core inlet, thermal efficiency of the cycle has a widely reduction under partial load conditions. (authors)

TIBER (Tokamak Ignition/Burn Experimental Reactor) II as a precursor to an international thermonuclear experimental reactor

International Nuclear Information System (INIS)

Henning, C.D.; Gilleland, J.R.

1988-01-01

The Tokamak Ignition/Burn Experimental Reactor (TIBER) was pursued in the US as one option for an International Thermonuclear Experimental Reactor (ITER). This concept evolved from earlier work on the Tokamak Fusion Core Experiment (TFCX) to develop a small, ignited tokamak. While the copper-coil versions of TFCX became the short-pulsed, 1.23-m radius, Compact Ignition Tokamak (CIT), the superconducting TIBER with long pulse or steady state and a 2.6-m radius was considered for international collaboration. Recently the design was updated to TIBER II, to accommodate more conservative confinement scaling, double-poloidal divertors for impurity control, steady-state current drive, and nuclear testing. 18 refs., 1 fig

Assessment of torsatrons as reactors

International Nuclear Information System (INIS)

Lyon, J.F.; Painter, S.L.

1992-12-01

Stellarators have significant operational advantages over tokamaks as ignited steady-state reactors because stellarators have no dangerous disruptions and no need for continuous current drive or power recirculated to the plasma, both easing the first wall, blanket, and shield design; less severe constraints on the plasma parameters and profiles; and better access for maintenance. This study shows that a reactor based on the torsatron configuration (a stellarator variant) could also have up to double the mass utilization efficiency (MUE) and a significantly lower cost of electricity (COE) than a conventional tokamak reactor (ARIES-I) for a range of assumptions. Torsatron reactors can have much smaller coil systems than tokamak reactors because the coils are closer to the plasma and they have a smaller cross section (higher average current density because of the lower magnetic field). The reactor optimization approach and the costing and component models are those used in the current stage of the ARIES-I tokamak reactor study. Typical reactor parameters for a 1-GW(e) Compact Torsatron reactor example are major radius R 0 = 6.6-8.8 m, on-axis magnetic field B 0 = 4.8-7.5 T, B max (on coils) = 16 T, MUE 140-210 kW(e)/tonne, and COE (in constant 1990 dollars) = 67-79 mill/kW(e)h. The results are relatively sensitive to assumptions on the level of confinement improvement and the blanket thickness under the inboard half of the helical windings but relatively insensitive to other assumptions

Mechanism for multiplicity of steady states with distinct cell concentration in continuous culture of mammalian cells.

Science.gov (United States)

Yongky, Andrew; Lee, Jongchan; Le, Tung; Mulukutla, Bhanu Chandra; Daoutidis, Prodromos; Hu, Wei-Shou

2015-07-01

Continuous culture for the production of biopharmaceutical proteins offers the possibility of steady state operations and thus more consistent product quality and increased productivity. Under some conditions, multiplicity of steady states has been observed in continuous cultures of mammalian cells, wherein with the same dilution rate and feed nutrient composition, steady states with very different cell and product concentrations may be reached. At those different steady states, cells may exhibit a high glycolysis flux with high lactate production and low cell concentration, or a low glycolysis flux with low lactate and high cell concentration. These different steady states, with different cell concentration, also have different productivity. Developing a mechanistic understanding of the occurrence of steady state multiplicity and devising a strategy to steer the culture toward the desired steady state is critical. We establish a multi-scale kinetic model that integrates a mechanistic intracellular metabolic model and cell growth model in a continuous bioreactor. We show that steady state multiplicity exists in a range of dilution rate in continuous culture as a result of the bistable behavior in glycolysis. The insights from the model were used to devise strategies to guide the culture to the desired steady state in the multiple steady state region. The model provides a guideline principle in the design of continuous culture processes of mammalian cells. © 2015 Wiley Periodicals, Inc.

Sensitivity and Uncertainty Analysis of Coupled Reactor Physics Problems : Method Development for Multi-Physics in Reactors

NARCIS (Netherlands)

Perkó, Z.

2015-01-01

This thesis presents novel adjoint and spectral methods for the sensitivity and uncertainty (S&U) analysis of multi-physics problems encountered in the field of reactor physics. The first part focuses on the steady state of reactors and extends the adjoint sensitivity analysis methods well

RELAP5-3D Results for Phase I (Exercise 2) of the OECD/NEA MHTGR-350 MW Benchmark

Energy Technology Data Exchange (ETDEWEB)

Gerhard Strydom

2012-06-01

The coupling of the PHISICS code suite to the thermal hydraulics system code RELAP5-3D has recently been initiated at the Idaho National Laboratory (INL) to provide a fully coupled prismatic Very High Temperature Reactor (VHTR) system modeling capability as part of the NGNP methods development program. The PHISICS code consists of three modules: INSTANT (performing 3D nodal transport core calculations), MRTAU (depletion and decay heat generation) and a perturbation/mixer module. As part of the verification and validation activities, steady state results have been obtained for Exercise 2 of Phase I of the newly-defined OECD/NEA MHTGR-350 MW Benchmark. This exercise requires participants to calculate a steady-state solution for an End of Equilibrium Cycle 350 MW Modular High Temperature Reactor (MHTGR), using the provided geometry, material, and coolant bypass flow description. The paper provides an overview of the MHTGR Benchmark and presents typical steady state results (e.g. solid and gas temperatures, thermal conductivities) for Phase I Exercise 2. Preliminary results are also provided for the early test phase of Exercise 3 using a two-group cross-section library and the Relap5-3D model developed for Exercise 2.

Customized Steady-State Constraints for Parameter Estimation in Non-Linear Ordinary Differential Equation Models.

Science.gov (United States)

Rosenblatt, Marcus; Timmer, Jens; Kaschek, Daniel

2016-01-01

Ordinary differential equation models have become a wide-spread approach to analyze dynamical systems and understand underlying mechanisms. Model parameters are often unknown and have to be estimated from experimental data, e.g., by maximum-likelihood estimation. In particular, models of biological systems contain a large number of parameters. To reduce the dimensionality of the parameter space, steady-state information is incorporated in the parameter estimation process. For non-linear models, analytical steady-state calculation typically leads to higher-order polynomial equations for which no closed-form solutions can be obtained. This can be circumvented by solving the steady-state equations for kinetic parameters, which results in a linear equation system with comparatively simple solutions. At the same time multiplicity of steady-state solutions is avoided, which otherwise is problematic for optimization. When solved for kinetic parameters, however, steady-state constraints tend to become negative for particular model specifications, thus, generating new types of optimization problems. Here, we present an algorithm based on graph theory that derives non-negative, analytical steady-state expressions by stepwise removal of cyclic dependencies between dynamical variables. The algorithm avoids multiple steady-state solutions by construction. We show that our method is applicable to most common classes of biochemical reaction networks containing inhibition terms, mass-action and Hill-type kinetic equations. Comparing the performance of parameter estimation for different analytical and numerical methods of incorporating steady-state information, we show that our approach is especially well-tailored to guarantee a high success rate of optimization.

Comparison of FSE T2 W PROPELLER and 3D-FIESTA of 3 T MR for the internal auditory canal.

Science.gov (United States)

Wu, Hai-Bo; Yuan, Hui-Shu; Ma, Furong; Zhao, Qiang

The study compared the use of periodically rotated overlapping parallel lines with enhanced reconstruction (PROPELLER) technique fast spin echo (FSE) T2 W and the sequence of three-dimensional fast imaging employing steady-state acquisition (3D-FIESTA) technique in the MRI of the internal auditory canal for overall image quality improvement. One hundred thirty-two patients undergoing FSE T2 W PROPELLER and 3D-FIESTA examinations of the internal auditory canal were included. All examinations were performed at 3.0 T with comparison of a sagittal oblique FSE T2 W sequence with the PROPELLER technique to 3D-FIESTA in the same reconstructed orientation with PROPELLER. Image quality was evaluated by two radiologists using a 4-point scale. The Wilcoxon signed rank test was used to compare the data of the two techniques. The image quality of FSE T2 W PROPELLER was significantly improved compared to the reconstructed images of 3D-FIESTA. Observer 1: median FSE T2 W with PROPELLER, 4 [mean, 3.455] versus median reconstructed 3D-FIESTA, 3 [mean, 3.15], (PW with PROPELLER, 4 [mean, 3.47] versus median reconstructed 3D-FIESTA, 3 [mean, 3.25], (PW PROPELLER technique for MRI of internal auditory canal reduced uncertainty caused by motion artifact and improved the quality of the image compared to the reconstructed 3D-FIESTA. It was affected by different parameters including the blade width, echo train length (ETL). This is explained by data oversampling at the center region of k-space, which requires additional imaging time over conventional MRI techniques. Increasing blade was expected to improve motion correction effects but also the signal-to-noise ratio. ETL increases the image sharpness and the overall image quality. Copyright © 2016. Published by Elsevier Inc.