Sample records for br-2 zero power mock-up reactor

  1. The power distribution and neutron fluence measurements and calculations in the VVER-1000 Mock-Up on the LR-0 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kostal, M.; Juricek, V.; Rypar, V.; Svadlenkova, M. [Research Center Rez Ltd., 250 68 Husinec-Rez 130 (Czech Republic); Cvachovec, F. [Univ. of Defence, Kounicova 65, 662 10 Brno (Czech Republic)


    The power density distribution in a reactor has significant influence on core structures and pressure vessel mechanical resistance, as well as on the physical characteristics of nuclear fuel. This quantity also has an effect on the leakage neutron and photon field. This issue has become of increasing importance, as it touches on actual questions of the VVER nuclear power plant life time extension. This paper shows the comparison of calculated and experimentally determined pin by pin power distributions. The calculation has been performed with deterministic and Monte Carlo approaches. This quantity is accompanied by the neutron and photon flux density calculation and measurements at different points of the light water zero-power (LR-0) research reactor mock-up core, reactor built-in component (core barrel), and reactor pressure vessel and model. The effect of the different data libraries used for calculation is discussed. (authors)

  2. Absolute determination of power density in the VVER-1000 mock-up on the LR-0 research reactor. (United States)

    Košt'ál, Michal; Švadlenková, Marie; Milčák, Ján


    The work presents a detailed comparison of calculated and experimentally determined net peak areas of selected fission products gamma lines. The fission products were induced during a 2.5 h irradiation on the power level of 9.5 W in selected fuel pins of the VVER-1000 Mock-Up. The calculations were done with deterministic and stochastic (Monte Carlo) methods. The effects of different nuclear data libraries used for calculations are discussed as well. The Net Peak Area (NPA) may be used for the determination of fission density across the mock-up. This fission density is practically identical to power density.

  3. Irradiation capabilities of LR-0 reactor with VVER-1000 Mock-Up core. (United States)

    Košťál, Michal; Rypar, Vojtěch; Svadlenková, Marie; Cvachovec, František; Jánský, Bohumil; Milčák, Ján


    Even low power reactors, such as zero power reactors, are sufficient for semiconductor radiation hardness effect investigation. This reflects the fact that fluxes necessary for affecting semiconductor electrical resistance are much lower than fluxes necessary to affect material parameters. The paper aims to describe the irradiation possibilities of the LR-0 reactor with a special core arrangement corresponding to VVER-1000 dosimetry Mock-Up.

  4. Reactor BR2. Introduction

    Energy Technology Data Exchange (ETDEWEB)

    Gubel, P


    The BR2 is a materials testing reactor and is still one of SCK-CEN's important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. During the last three years, the availability of the installation was maintained at an average level of 97.6 percent. In the year 2000, the reactor was operated for a total of 104 days at a mean power of 56 MW. In 2000, most irradiation experiments were performed in the CALLISTO PWR loop. The report describes irradiations achieved or under preparation in 2000, including the development of advanced facilities and concept studies for new programmes. An overview of the scientific irradiation programmes as well as of the R and D programme of the BR2 reactor in 2000 is given.

  5. Comparison of various hours living fission products for absolute power density determination in VVER-1000 mock up in LR-0 reactor. (United States)

    Košťál, Michal; Švadlenková, Marie; Koleška, Michal; Rypar, Vojtěch; Milčák, Ján


    Measuring power level of zero power reactor is a quite difficult task. Due to the absence of measurable cooling media heating, it is necessary to employ a different method. The gamma-ray spectroscopy of fission products induced within reactor operation is one of possible ways of power determination. The method is based on the proportionality between fission product buildup and released power. The (92)Sr fission product was previously preferred as nuclide for LR-0 power determination for short-time irradiation experiments. This work aims to find more appropriate candidates, because the (92)Sr, however suitable, has a short half-life, which limits the maximal measurable amount of fuel pins within a single irradiation batch. The comparison of various isotopes is realized for (92)Sr, (97)Zr, (135)I, (91)Sr, and (88)Kr. The comparison between calculated and experimentally determined (C/E-1 values) net peak areas is assessed for these fission products. Experimental results show that studied fission products, except (88)Kr, are in comparable agreement with (92)Sr results. Since (91)Sr has notably higher half-life than (92)Sr, (91)Sr seems to be more appropriate marker in experiments with a large number of measured fuel pins.

  6. The application and comparison of 97Zr and 92Sr in the absolute determination of the contribution of power density and cladding activation in a VVER-1000 Mock-Up on the LR-0 Research Reactor (United States)

    Košťál, Michal; Švadlenková, Marie; Milčák, Ján


    97Zr is a relatively high-yield fission product that can be used for zero reactor power determination. The technique is not widely used because in the case of reactors that use zirconium metal in the fuel cladding, it is not only a fission product but is also produced by activation. In an appropriately chosen time interval, results obtained using 97Zr can be compared to those of power determination performed using 92Sr. The knowledge of the ratio between fission-induced 97Zr and the portion of 97Zr activated in the cladding can be used not only for power-density determination but also as an important indication of fuel failures.

  7. Numerical evaluation of weld overlay applied to a pressurized water reactor nozzle mock-up

    Energy Technology Data Exchange (ETDEWEB)

    Rabello, Emerson G.; Silva, Luiz L.; Gomes, Paulo T.V., E-mail: egr@cdtn.b, E-mail: silvall@cdtn.b, E-mail: gomespt@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil). Servico de Integridade Estrutural


    The primary water stress corrosion cracking (PWSCC) is a major mechanism of failure in the primary circuit of PWR type nuclear power plants. The PWSCC is associated with the presence of corrosive environment, the susceptibility to corrosion cracking of the materials involved and the tensile stresses presence. Residual stresses generated during dissimilar materials welding can contribute to PWSCC. An alternative to the PWSCC mitigation is the application of external weld layers in the regions of greatest susceptibility to corrosion cracking. This process, called Weld Overlay (WOL), has been widely used in regions of dissimilar weld (low alloy steel and stainless steel with nickel alloy addition) of nozzles and pipes on the primary circuit in order to promote internal compressive stresses on the wall of these components. This paper presents the steps required to the numerical stress evaluation (by finite element method) during the dissimilar materials welding as well as application of Weld Overlay process in a nozzle mock-up. Thus, one can evaluate the effectiveness of the application of weld overlay process to internal compressive stress generation on the wall nozzle. (author)

  8. High RF power test of a lower hybrid module mock-up in carbon fiber composite

    Energy Technology Data Exchange (ETDEWEB)

    Goniche, M.; Bibet, P.; Brossaud, J.; Cano, V.; Froissard, P.; Kazarian, F.; Rey, G. [Association Euratom-CEA Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee; Maebara, S.; Kiyono, K.; Seki, M.; Suganuma, K.; Ikeda, Y.; Imai, T. [Japan Atomic Energy Research Inst., Tokyo (Japan). Dept. of Fusion Facility


    A mock-up module of a Lower Hybrid Current Drive antenna module of a Carbon Fiber Composite (CFC) was fabricated for the development of heat resistive front facing the plasma. This module is made from CFC plates and rods which are copper coated to reduce the RF losses. The withstand-voltage, the RF properties and outgassing rates for long pulses and high RF power were tested at the Lower Hybrid test bed facility of Cadarache. After the short pulse conditioning, long pulses with a power density ranging between 50 and 150 MW/m{sup 2} were performed with no breakdowns. During these tests, the module temperaturewas increasing from 100-200 deg. C to 400-500 deg. C. It was also checked that high power density, up to 150 MW/m{sup 2}, could be transmitted when the waveguides are filled with H{sub 2} at a pressure of 5 x 10{sup -2} Pa. No significant change in the reflection coefficient is measured after the long pulse operation. During a long pulse, the power reflection increases during the pulse typically from 0.8% to 1.3%. It is concluded that the outgassing rate of Cu-plated CFC is about 6 times larger than of Dispersion Strengthened Copper (DSC) module at the module temperature of 300 deg. C. No significant increase of the global outgassing of the CFC module was measured after hydrogen pre-filling. After the test, visual inspection revealed that peeling of the copper coating occurred at one end of the module only on a very small area (0.2 cm{sup 2}). It is assessed that a CFC module is an attractive candidate for the hardening of the tip of the LHCD antenna. (authors)

  9. The Swedish Zero Power Reactor R0

    Energy Technology Data Exchange (ETDEWEB)

    Landergaard, Olof; Cavallin, Kaj; Jonsson, Georg


    The reactor R0 is a critical facility built for heavy water and natural uranium or fuel of low enrichment,, The first criticality was achieved September 25, 1959. During a first period of more than two years the R0 will be operated as a bare reactor in order to simplify interpretation of results. The reactor tank is 3. 2 m high and 2. 25 m in diameter. The fuel suspension system is quite flexible in order to facilitate fuel exchange and lattice variations. The temperature of the water can be varied between about 10 and 90 C by means of a heater and a cooler placed in the external circulating system. The instrumentation of the reactor has to meet the safety requirements not only during operation but also during rearrangements of the core in the shut-down state. Therefore, the shut-down state is always defined by a certain low 'safe' moderator level in the reactor tank. A number of safety rods are normally kept above the moderator ready for action. For manual or automatic control of the reactor power a specially designed piston pump is needed, by which the moderator level is varied. The pump speed is controlled from the reactor power error by means of a Ward-Leonard system. Moderator level measurement is made by means of a water gauge with an accuracy of {+-} 0. 1 mm.

  10. Fast Neutron Transport in the Biological Shielding Model and Other Regions of the VVER-1000 Mock-Up on the LR-0 Research Reactor

    Directory of Open Access Journals (Sweden)

    Košťál Michal


    Full Text Available A set of benchmark experiments was carried out in the full scale VVER-1000 mock-up on the reactor LR-0 in order to validate neutron transport calculation methodologies and to perform the optimization of the shape and locations of neutron flux operation monitors channels inside the shielding of the new VVER-1000 type reactors. Compared with previous experiments on the VVER-1000 mock-up on the reactor LR-0, the fast neutron spectra were measured in the extended neutron energy interval (0.1–10 MeV and new calculations were carried out with the MCNPX code using various nuclear data libraries (ENDF/B VII.0, JEFF 3.1, JENDL 3.3, JENDL 4, ROSFOND 2009, and CENDL 3.1. Measurements and calculations were carried out at different points in the mock-up. The calculation and experimental data are compared.

  11. Fast Neutron Transport in the Biological Shielding Model and Other Regions of the VVER-1000 Mock-Up on the LR-0 Research Reactor (United States)

    Košťál, Michal; Milčák, Ján; Cvachovec, František; Jánský, Bohumil; Rypar, Vojtěch; Juříček, Vlastimil; Novák, Evžen; Egorov, Alexander; Zaritskiy, Sergey


    A set of benchmark experiments was carried out in the full scale VVER-1000 mock-up on the reactor LR-0 in order to validate neutron transport calculation methodologies and to perform the optimization of the shape and locations of neutron flux operation monitors channels inside the shielding of the new VVER-1000 type reactors. Compared with previous experiments on the VVER-1000 mock-up on the reactor LR-0, the fast neutron spectra were measured in the extended neutron energy interval (0.1-10 MeV) and new calculations were carried out with the MCNPX code using various nuclear data libraries (ENDF/B VII.0, JEFF 3.1, JENDL 3.3, JENDL 4, ROSFOND 2009, and CENDL 3.1). Measurements and calculations were carried out at different points in the mock-up. The calculation and experimental data are compared.

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

    Energy Technology Data Exchange (ETDEWEB)

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


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

  13. Neutron and gamma field investigations in the VVER-1000 mock-up concrete shielding on the reactor LR-0

    Energy Technology Data Exchange (ETDEWEB)

    Zaritsky, S.; Egorov, A. [National Research Center, Kurchatov Inst., Moscow 123182 (Russian Federation); Osmera, B.; Marik, M.; Rypar, V. [Research Centre Rez Ltd., Rez 25068 (Czech Republic); Cvachovec, F. [Univ. of Defense, Brno 61200 (Czech Republic); Kolros, A. [Czech Technical Univ., Prague 18000 (Czech Republic)


    Two sets of neutron and gamma field investigations were carried out in the dismountable model of radiation shielding of the VVER-1000 mock-up on the LR-0 reactor. First, measurements and calculations of the {sup 3}He(n,p)T reaction rate and fast neutrons and gamma flux spectra in the operational neutron monitor channel inside a concrete shielding for different shapes and locations of the channel (cylindrical channel in a concrete, channels with collimator in a concrete, cylindrical channel in a graphite). In all cases measurements and calculations of the {sup 3}He(n,p)T reaction rate were done with and without an additional moderator-polyethylene insert inside the channel. Second, measurements and calculations of the {sup 3}He(n,p)T reaction rate spatial distribution inside a concrete. The {sup 3}He(n,p)T reaction rate measurements and calculations were carried out exploring the relative thermal neutron density in the channels and its space distribution in the concrete. Fast neutrons and gamma measurements were carried out with a stilbene (45 x 45 mm) scintillation spectrometer in the energy regions 0.5-10 MeV (neutrons) and 0.2-10 MeV (gammas). (authors)

  14. Recent activities at the zero-power teaching reactor CROCUS

    Energy Technology Data Exchange (ETDEWEB)

    Girardin, G.; Chawla, R., E-mail: [Swiss Polytechnical School of Lausanne, Lausanne (Switzerland)


    CROCUS is a zero-power critical facility used mainly for educational purposes at the Swiss Federal Institute of Technology (EPFL) in Lausanne, Switzerland. It is a low-enriched-uranium fuelled, light-water moderated reactor, with the fission power limited to 100 W. The presentation will discuss the crucial role of CROCUS in teaching -- both as framework for reactor practicals offered to physics students at EPFL and as key educational tool in the recently established Swiss Master of Science in Nuclear Engineering. Regular development work is needed for the various instruments and components associated with the facility. As illustration, the recently completed refurbishment of the control rod system and the related calibration experiments will also be discussed.

  15. Optimized Control Rods of the BR2 Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kalcheva, Silva; Koonen, E.


    At the present time the BR-2 reactor uses control elements with cadmium as neutron absorbing part. The lower section of the control element is a beryllium assembly cooled by light water. Due to the burn up of the lower end of the cadmium section during the reactor operation, the presently used rods for reactivity control of the BR-2 reactor have to be replaced by new ones. Considered are various types Control Rods with full active part of the following materials: cadmium (Cd), hafnium (Hf), europium oxide (Eu2O3) and gadolinium (Gd2O3). Options to decrease the burn up of the control rod material in the hot spot, such as use of stainless steel in the lower active part of the Control Rod are discussed. Comparison with the characteristics of the presently used Control Rods types is performed. The changing of the characteristics of different types Control Rods and the perturbation effects on the reactor neutronics during the BR-2 fuel cycle are investigated. The burn up of the Control Rod absorbing material, total and differential control rods worth, macroscopic and effective microscopic absorption cross sections, fuel and reactivity evolution are evaluated during approximately 30 operating cycles.

  16. Neutron dose estimation in a zero power nuclear reactor (United States)

    Triviño, S.; Vedelago, J.; Cantargi, F.; Keil, W.; Figueroa, R.; Mattea, F.; Chautemps, A.; Santibañez, M.; Valente, M.


    This work presents the characterization and contribution of neutron and gamma components to the absorbed dose in a zero power nuclear reactor. A dosimetric method based on Fricke gel was implemented to evaluate the separation between dose components in the mixed field. The validation of this proposed method was performed by means of direct measurements of neutron flux in different positions using Au and Mg-Ni activation foils. Monte Carlo simulations were conversely performed using the MCNP main code with a dedicated subroutine to incorporate the exact complete geometry of the nuclear reactor facility. Once nuclear fuel elements were defined, the simulations computed the different contributions to the absorbed dose in specific positions inside the core. Thermal/epithermal contributions of absorbed dose were assessed by means of Fricke gel dosimetry using different isotopic compositions aimed at modifying the sensitivity of the dosimeter for specific dose components. Clear distinctions between gamma and neutron capture dose were obtained. Both Monte Carlo simulations and experimental results provided reliable estimations about neutron flux rate as well as dose rate during the reactor operation. Simulations and experimental results are in good agreement in every positions measured and simulated in the core.

  17. Conformity Between LR0 Mock-Ups and Vvers Npp Rpv Neutron Flux Attenuation (United States)

    Belousov, Sergey; Ilieva, Krassimira; Kirilova, Desislava


    The conformity of the mock-up results and those for reactor pressure vessel (RPV) of nuclear power plants (NPP) has been evaluated in order to qualify if the mock-ups data could be used for benchmark's purpose only, or/and for simulating of the NPP irradiation conditions. Neutron transport through the vessel has been calculated by the three-dimensional discrete ordinate code TORT with problem oriented multigroup energy neutron cross-section library BGL. Neutron flux/fluence and spectrum shape represented by normalized group neutron fluxes in the multigroup energy structure, for neutrons with energy above 0.5 MeV, have been used for conformity analysis. It has been demonstrated that the relative difference of the attenuation factor as well as the group neutron fluxes did not exceed 10% at all considered positions for VVER-440. For VVER-1000, it has been obtained the same consistency, except for the location behind the RPV. The neutron flux attenuation behind the RPV is 18% higher than the mock-up attenuation. It has been shown that this difference arises from the dissimilarity of the biological shielding. The obtained results have demonstrated that the VVERs' mock-ups are appropriate for simulating the NPP irradiation conditions. The mock-up results for VVER-1000 have to be applied more carefully i.e. taking into account the existing peculiarity of the biological shielding and RPV attenuation azimuthal dependence.

  18. The State of the Art Report on the Development of Manufacturing Technology of Fusion Reactor FW Blanket and Mock-up in USA

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyung Gil; Jeong, Y. H.; Baek, J. H.; Park, J. Y.; Kim, J. H


    Researches on the development of the bonding method between Cu/SS and Be/Cu had been done early in 1990 in the USA. To introduce the optimized bonding method, the bonding of Cu/SS had been tried with HIP and EXP methods and its bonding properties was evaluated by means of the mechanical tests for such as tensile, sear, fatigue etc. Although a small mock-up sample had not been carried out to identify the Cu/SS bonding characteristics, it was found out that HIP or EXP was one of promising bonding methods. Especially, HIP method to bond Be/Cu was confirmed to be more favorable method than EXP from the study on the brazing and HIP. It was also tried to increase the bonding effectiveness of Be/Cu by appling various interplay to the bonding. The HIP application condition to bond Be/Cu was studied in the USA from the test of small mock-up. It appears that the HIP bonding condition can be applied to manufacture the FW blanket.

  19. Design and Analysis of the Power Control System of the Fast Zero Energy Reactor FR-0

    Energy Technology Data Exchange (ETDEWEB)

    Schuh, N.J.H.


    This report describes the power control by means of the fine-control rod and the design of the control system of the fast zero energy reactor FR-0 located in Studsvik, Sweden. System requirements and some operational conditions were used as design criteria. Manual and automatic control is possible. Variable electronic end-stops for the control rod have been designed, because of the special construction of the reactor and control rod. Noise in the control system caused by the reactor, detector and electronics caused disturbances of the control system at the lower power levels. The noise power-spectrum was measured. Statistical design methods, using the measured noise power spectrum, were used to design filters, which will reduce the influence of the noise at the lower power levels. Root Loci sketches and Bode diagrams were used for stability analyses. The system was simulated on an analogue computer, taking into account even nonlinearities of the control system and noise. Typical cases of reactor operation were simulated and stability analysis performed.

  20. Strength of Mock-up Trial Grout

    DEFF Research Database (Denmark)

    Sørensen, Eigil V.

    The present report describes tests carried out on samples taken and cast during the execution of a mock-up trial placement of the high performance grout MASTERFLOW 9500 on January 21, 2009.......The present report describes tests carried out on samples taken and cast during the execution of a mock-up trial placement of the high performance grout MASTERFLOW 9500 on January 21, 2009....

  1. Feasibility study of the thermo-siphon mock-up test

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jung Woon; Kim, Young Jin; Lee, Kye Hong; Kim, Young Ki; Jeong, Sang Kwon


    Described is the feasibility of the thermo-siphon mock-up test for the HANARO-CNS facility. The purposes of the mock-up tests are discussed in detail as the three concepts: for the detailed design, for the operation of the CNS facility, for the safety assurance of itself. This report considers the two stages of mock-up tests in terms of the experimental schedule and plan. As the first stage, the small-size mock-up test using Argon will be implemented to obtain the experience in the cryogenic fluid and to understand the basic concept of the CNS thermo-siphon. In the second stage, two kinds of mock-up tests are discussed: the full-scale mock-up test using liquid hydrogen or the integrated final test using hydrogen outside the reactor after the full-scale mock-up test using Freon gas. The contents discussed in this report will be the basis or the guide lines for the mock-up test. In addition, the results of the mock-up test will be the foundation for the safe operation of the HANARO-CNS facility.

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

    Hashimoto, Toshiyuki; Takatsu, Hideyuki; Sato, Satoshi


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

  3. The making of a mock-up

    DEFF Research Database (Denmark)

    Rosenqvist, Tanja Schultz; Heimdal, Elisabeth Jacobsen


    As part of a research project about user involvement in textile design we have carried out two Design:Labs (Binder & Brandt 2008) engaging different stakeholders in designing textile products for Danish hospital environments. In this paper we follow a mock-up session done as part of the second...

  4. How tangible mock-ups support design collaboration

    DEFF Research Database (Denmark)

    Brandt, Eva


    This paper is a contribution to a more conscious use of tangible mock-ups in collaborative design processes. It describes a design team's use of mock-ups in a series of workshops involving potential customers and users. Focus is primarily on the use of three-dimensional design mock-ups and how...... differences in these affected the dialogue. Reflective conversations were established by using tangible mock-ups as 'things-to-think with'. They served as boundary objects that spanned the gap between the different competencies and interests of participants in design. The design mock-ups evoked different...... things for different participants whereas the challenge for the design team was to find boundaries upon which everybody could agree. The level of details represented in a mock-up affected the communication so that a mock-up with few details evoked different issues whereas a very detailed mock-up evoked...

  5. 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)


    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.


    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess; Nozomu Fujimoto; James W. Sterbentz; Luka Snoj; Atsushi Zukeran


    The High Temperature Engineering Test Reactor (HTTR) of the Japan Atomic Energy Agency (JAEA) is a 30 MWth, graphite-moderated, helium-cooled reactor that was constructed with the objectives to establish and upgrade the technological basis for advanced high-temperature gas-cooled reactors (HTGRs) as well as to conduct various irradiation tests for innovative high-temperature research. The core size of the HTTR represents about one-half of that of future HTGRs, and the high excess reactivity of the HTTR, necessary for compensation of temperature, xenon, and burnup effects during power operations, is similar to that of future HTGRs. During the start-up core physics tests of the HTTR, various annular cores were formed to provide experimental data for verification of design codes for future HTGRs. The experimental benchmark performed and currently evaluated in this report pertains to the data available for two zero-power, warm-critical measurements with the fully-loaded HTTR core. Six isothermal temperature coefficients for the fully-loaded core from approximately 340 to 740 K have also been evaluated. These experiments were performed as part of the power-up tests (References 1 and 2). Evaluation of the start-up core physics tests specific to the fully-loaded core (HTTR-GCR-RESR-001) and annular start-up core loadings (HTTR-GCR-RESR-002) have been previously evaluated.

  7. 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)


    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).

  8. SoLid: Search for Oscillation with a 6Li Detector at the BR2 research reactor

    CERN Document Server

    Michiels, Ianthe


    In the past decades, various nuclear reactor neutrino experiments have measured a deficit in the flux of antineutrinos coming from the reactor at short reactor-detector distances, when compared to theoretical calculations. One of the experiments designed to investigate this reactor antineutrino anomaly is the SoLid experiment. It uses the compact BR2 research reactor from the SCK-CEN in Mol, Belgium, to perform reactor antineutrino flux measurements at very short baseline. These proceedings discuss the general detection concepts of the SoLid experiment and its novel detector technology. The performance of the SoLid design is demonstrated with some results of the analysis of the data gathered with the experiment's first large scale test module, SM1.

  9. First results of the deployment of a SoLid detector module at the SCK-CEN BR2 reactor

    CERN Document Server

    Ryder, Nick


    The SoLid experiment aims to resolve the reactor neutrino anomaly by searching for electron-to-sterile anti-neutrino oscillations. The search will be performed between 5.5 and 10 m from the highly enriched uranium core of the BR2 reactor at SCK-CEN. The experiment utilises a novel approach to anti-neutrino detection based on a highly segmented, composite scintillator detector design. High experimental sensitivity can be achieved using a combination of high neutron-gamma discrimination using 6 LiF:ZnS(Ag) and precise localisation of the inverse beta decay products. This compact detector system requires limited passive shielding as it relies on spacial topology to determine the different classes of backgrounds. The first full scale, 288 kg, detector module was deployed at the BR2 reactor in November 2014. A phased three tonne experimental deployment will begin in the second half of 2016, allowing a precise search for oscillations that will resolve the reactor anomaly using a three tonne detector running for thr...

  10. Loss-of-Flow and Loss-of-Pressure Simulations of the BR2 Research Reactor with HEU and LEU Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Licht, J. [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); Sikik, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Van den Branden, G. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Koonen, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium)


    Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The reactor core of BR2 is located inside a pressure vessel that contains 79 channels in a hyperboloid configuration. The core configuration is highly variable as each channel can contain a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Because of this variability, a representative core configuration, based on current reactor use, has been defined for the fuel conversion analyses. The code RELAP5/Mod 3.3 was used to perform the transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. The input model has been modernized relative to that historically used at BR2 taking into account the best modeling practices developed by Argonne National Laboratory (ANL) and BR2 engineers.

  11. Loss-of-Flow and Loss-of-Pressure Simulations of the BR2 Research Reactor with HEU and LEU Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Licht, J. [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); Sikik, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Van den Branden, G. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Koonen, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium)


    Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The reactor core of BR2 is located inside a pressure vessel that contains 79 channels in a hyperboloid configuration (Figure 1). The core configuration is highly variable as each channel can contain a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Because of this variability, a representative core configuration (Figure 2), based on current reactor use, has been defined for the fuel conversion analyses [1]. The code RELAP5/Mod 3.3 [2] was used to perform the transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. The input model has been modernized relative to that historically used at BR2 taking into account the best modeling practices developed by Argonne National Laboratory (ANL) and BR2 engineers.

  12. Computation of a core disruptive accident in the MARS mock-up

    Energy Technology Data Exchange (ETDEWEB)

    Robbe, M.F. [CEA Saclay, Bat 118, 91191 Gif sur Yvette Cedex (France)]. E-mail:; Lepareux, M. [CEA Saclay, Bat 118, 91191 Gif sur Yvette Cedex (France); Seinturier, E. [Socotec Industrie, 1 av. du Parc, 78180 Montigny le Bretonneux (France)


    A hypothetical core disruptive accident in a liquid metal fast breeder reactor (LMFBR) results from the interaction between molten fuel and liquid sodium, which creates a high-pressure bubble of gas in the core. The violent expansion of this bubble loads and deforms the vessel and the internal structures. The MARS experimental test simulates a HCDA in a small-scale mock-up containing all the significant internal components of a fast breeder reactor. The mock-up is filled with water, topped by an argon blanket, and the explosion is generated by an explosive charge. This paper presents a numerical simulation of the test with the EUROPLEXUS code. The top closure is represented by massive structures and the main internal structures are described by shells. The current numerical results are described and compared with the experimental ones, and previous computations with the CASTEM-PLEXUS code.

  13. SoLid: An innovative anti-neutrino detector for searching oscillations at the SCK•CEN BR2 reactor (United States)

    Abreu, Yamiel


    The SoLid experiment intends to search for active-to-sterile anti-neutrino oscillations at a very short baseline from the SCK•CEN BR2 research reactor (Mol, Belgium). A novel detector approach to measure reactor anti-neutrinos was developed based on an innovative sandwich of composite polyvinyl-toluene and 6LiF:ZnS(Ag) scintillators. The system is highly segmented and read out by a network of wavelength shifting fibers and SiPM. High experimental sensitivity can be achieved compared to other standard technologies thanks to the combination of high granularity, good neutron-gamma discrimination using 6LiF:ZnS(Ag) scintillator and precise localisation of the Inverse Beta Decay products. This technology can be considered as a new generation of an anti-neutrino detector. This compact system requires limited passive shielding and relies on spatial topology to determine the different classes of backgrounds. We will describe the principle of detection and the detector design. Particular focus on the neutron discrimination will be made, as well as on the capability to use cosmic muons for channel equalisation and energy calibration. The performance of the first 288 kg SoLid module (SM1), based on the data taken at BR2 from February to September 2015, will be presented. We will conclude with the next phase, which will start in 2016, and the future plans of the experiment.

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

    Directory of Open Access Journals (Sweden)

    Girardin G.


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

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

    Directory of Open Access Journals (Sweden)

    Arkani Mohammad


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

  16. Design and optimization of W/Cu divertor mock-ups

    Institute of Scientific and Technical Information of China (English)

    Qiong Li; Weiping Shen


    Tungsten is a promising candidate for plasma-facing materials to cover the surface of the divertor plate in the design of an international thermonuclear experimental reactor (ITER). Copper as a heat sink material serves to transfer heat excellently. Divertor mock-ups with W/Cu graded interlayers were designed to reduce thermal stresses. Thermally induced stresses and temperature in a W/Cu divertor mock-up were analyzed using the finite element method. The graded structures with different exponents p and thicknesses were designed and discussed. The conclusions drawn from these analyses are that thermal stresses reach the minimum and the temperature is suitable when exponent p is 1.5 and the thickness of five graded interlayers is 5 mm.

  17. Demonstration of the LHC Safety Training Tunnel Mock-Up

    CERN Multimedia

    Brice, Maximilien


    Members of CERN's management visit the LHC tunnel mock-up at the Safety Training Centre on the Prévessin site. The facility is used to train personnel in emergency responses including the use of masks and safe evacuation.

  18. Digital Mock-up Technology in Product Development and Research

    Institute of Scientific and Technical Information of China (English)

    CHEN Xu; ZHANG Pandeng; YANG Cheng; XU Zhongming


    After introducing the present status of digital mock-up (DMU) technology in product development and research, the modeling and its key technologies in product design are described. The architecture of digital design platform system for main DMU model is developed. Based on the architecture, a method of skeleton design has been applied to the development of digital design system.

  19. Validation of SCALE4.4a for Calculation of Xe-Sm Transients After a Scram of the BR2 Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kalcheva, S.; Ponsard, B.; Koonen, E.


    The aim of this report is to validate the computational modules system SCALE4.4a for evaluation of reactivity changes, macroscopic absorption cross sections and calculations of the positions of the Control Rods during their motion in Xe-Sm transient after a scram of the BR-2 reactor. The rapid shutting down of the reactor by inserting of negative reactivity by the Control Rods is known as a reactor scram. Following reactor scram, a large xenon and samarium buildup occur in the reactor, which may appreciably affect the multiplication factor of the core due to enormous neutron absorption. The validation of the calculations of Xe-Sm transients by SCALE4.4a has been performed on the measurements of the positions of the Control Rods during their motion in Xe-Sm transients of the BR-2 reactor and on comparison with the calculations by the standard procedure XESM, developed at the BR-2 reactor. A final conclusion is made that the SCALE4.4a modules system can be used for evaluation of Xe-Sm transients of the BR-2 reactor. The utilization of the code is simple, the computational time takes from few seconds.

  20. A new magnet for the LHC mock-up

    CERN Multimedia

    HSE Unit


    This year, the safety training centre on the Prévessin site acquired a mock-up of the LHC, which simulates the work and safety conditions in the tunnel.   Photo: Christoph Balle. A new dummy quadrupole has just been added to the magnet chain, making the mock-up even more realistic. The new facility, which was a joint endeavour by the TE, GS, BE and EN Departments, will significantly improve the quality of the various training courses held at the centre, particularly the course on the use of self-rescue masks. To consult the safety training catalogue and/or sign up for radiation protection training, please go to: For further information, please contact the Safety Training and Awareness service by telephone on 73811 or 79935 or by e-mail to  

  1. China Overseas Plaza Mock-up Floor Grand Openning

    Institute of Scientific and Technical Information of China (English)


    @@ Global Leasing launched on Full Scale Beijing, November 6, 2008- China Overseas Plaza, located at the heart of the Beijing CBD with Chang'an Avenue to the south and China World Trade Center PhasenⅢto the east, is constructed including two international Grade-A office towers with a commercial podium. The developer held a grand opening ceremony on the completed mock-up floor the sixth floor of the South Tower.

  2. A Coupled THMC model of FEBEX mock-up test

    Energy Technology Data Exchange (ETDEWEB)

    Zheng, Liange; Samper, Javier


    FEBEX (Full-scale Engineered Barrier EXperiment) is a demonstration and research project for the engineered barrier system (EBS) of a radioactive waste repository in granite. It includes two full-scale heating and hydration tests: the in situ test performed at Grimsel (Switzerland) and a mock-up test operating at CIEMAT facilities in Madrid (Spain). The mock-up test provides valuable insight on thermal, hydrodynamic, mechanical and chemical (THMC) behavior of EBS because its hydration is controlled better than that of in situ test in which the buffer is saturated with water from the surrounding granitic rock. Here we present a coupled THMC model of the mock-up test which accounts for thermal and chemical osmosis and bentonite swelling with a state-surface approach. The THMC model reproduces measured temperature and cumulative water inflow data. It fits also relative humidity data at the outer part of the buffer, but underestimates relative humidities near the heater. Dilution due to hydration and evaporation near the heater are the main processes controlling the concentration of conservative species while surface complexation, mineral dissolution/precipitation and cation exchanges affect significantly reactive species as well. Results of sensitivity analyses to chemical processes show that pH is mostly controlled by surface complexation while dissolved cations concentrations are controlled by cation exchange reactions.

  3. Digital mock-up for the spent fuel disassembly processes

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J. Y.; Kim, S. H.; Song, T. G.; Kim, Y. H.; Hong, D. H.; Yoon, J. S


    In this study, the graphical design system is developed and the digital mock-up is implemented for designing the spent fuel handling and disassembly processes. The system consists of a 3D graphical modeling system, a devices assembling system, and a motion simulation system. This system is used throughout the design stages from the conceptual design to the motion analysis. By using this system, all the process involved in the spent fuel handling and disassembly processes are analyzed and optimized. Also, this system is used in developing the on-line graphic simulator which synchronously simulates the motion of the equipment in a real time basis by connecting the device controllers with the graphic server through the TCP/IP network. This simulator can be effectively used for detecting the malfunctions of the process equipment which is remotely operated. Thus, the simulator enhances the reliability and safety of the spent fuel handling process by providing the remote monitoring function of the process. The graphical design system and the digital mock-up system can be effectively used for designing the process equipment, as well as the optimized process and maintenance process. And the on-line graphic simulator can be an alternative of the conventional process monitoring system which is a hardware based system.

  4. Estimation of steady-state and transcient power distributions for the RELAP analyses of the 1963 loss-of-flow and loss-of-pressure tests at BR2.

    Energy Technology Data Exchange (ETDEWEB)

    Dionne, B.; Tzanos, C. P. (Nuclear Engineering Division)


    To support the safety analyses required for the conversion of the Belgian Reactor 2 (BR2) from highly-enriched uranium (HEU) to low-enriched uranium (LEU) fuel, the simulation of a number of loss-of-flow tests, with or without loss of pressure, has been undertaken. These tests were performed at BR2 in 1963 and used instrumented fuel assemblies (FAs) with thermocouples (TC) imbedded in the cladding as well as probes to measure the FAs power on the basis of their coolant temperature rise. The availability of experimental data for these tests offers an opportunity to better establish the credibility of the RELAP5-3D model and methodology used in the conversion analysis. In order to support the HEU to LEU conversion safety analyses of the BR2 reactor, RELAP simulations of a number of loss-of-flow/loss-of-pressure tests have been undertaken. Preliminary analyses showed that the conservative power distributions used historically in the BR2 RELAP model resulted in a significant overestimation of the peak cladding temperature during the transient. Therefore, it was concluded that better estimates of the steady-state and decay power distributions were needed to accurately predict the cladding temperatures measured during the tests and establish the credibility of the RELAP model and methodology. The new approach ('best estimate' methodology) uses the MCNP5, ORIGEN-2 and BERYL codes to obtain steady-state and decay power distributions for the BR2 core during the tests A/400/1, C/600/3 and F/400/1. This methodology can be easily extended to simulate any BR2 core configuration. Comparisons with measured peak cladding temperatures showed a much better agreement when power distributions obtained with the new methodology are used.

  5. Engineering and manufacturing of ITER first mirror mock-ups. (United States)

    Joanny, M; Travère, J M; Salasca, S; Corre, Y; Marot, L; Thellier, C; Gallay, G; Cammarata, C; Passier, B; Fermé, J J


    Most of the ITER optical diagnostics aiming at viewing and monitoring plasma facing components will use in-vessel metallic mirrors. These mirrors will be exposed to a severe plasma environment and lead to an important tradeoff on their design and manufacturing. As a consequence, investigations are carried out on diagnostic mirrors toward the development of optimal and reliable solutions. The goals are to assess the manufacturing feasibility of the mirror coatings, evaluate the manufacturing capability and associated performances for the mirrors cooling and polishing, and finally determine the costs and delivery time of the first prototypes with a diameter of 200 and 500 mm. Three kinds of ITER candidate mock-ups are being designed and manufactured: rhodium films on stainless steel substrate, molybdenum on TZM substrate, and silver films on stainless steel substrate. The status of the project is presented in this paper.

  6. X-15 mock-up with test pilot Milt Thompson (United States)


    NASA research pilot Milt Thompson is seen here with the mock-up of X-15 #3 that was later installed at the NASA Dryden Flight Research Center, Edwards, California. Milton 0. Thompson was a research pilot, Chief Engineer and Director of Research Projects during a long career at the NASA Dryden Flight Research Center. Thompson was hired as an engineer at the flight research facility on 19 March 1956, when it was still under the auspices of NACA. He became a research pilot on 25 May 1958. Thompson was one of the 12 NASA, Air Force, and Navy pilots to fly the X-15 rocket-powered research aircraft between 1959 and 1968. He began flying X-15s on 29 October 1963. He flew the aircraft 14 times during the following two years, reaching a maximum speed of 3723 mph (Mach 5.42) and a peak altitude of 214,100 feet on separate flights. (On a different flight, he reached a Mach number of 5.48 but his mph was only 3712.) Thompson concluded his active flying career in 1968, becoming Director of Research Projects. In 1975 he was appointed Chief Engineer and retained the position until his death on 8 August 1993. The X-15 was a rocket powered aircraft 50 ft long with a wingspan of 22 ft. It was a missile-shaped vehicle with an unusual wedge-shaped vertical tail, thin stubby wings, and unique side fairings that extended along the side of the fuselage. The X-15 weighed about 14,000 lb empty and approximately 34,000 lb at launch. The XLR-99 rocket engine, manufactured by Thiokol Chemical Corp., was pilot controlled and was capable of developing 57,000 lb of thrust. North American Aviation built three X-15 aircraft for the program. The X-15 research aircraft was developed to provide in-flight information and data on aerodynamics, structures, flight controls, and the physiological aspects of high-speed, high-altitude flight. A follow on program used the aircraft as a testbed to carry various scientific experiments beyond the Earth's atmosphere on a repeated basis. For flight in the dense

  7. Project W-314 performance mock-up test procedure

    Energy Technology Data Exchange (ETDEWEB)



    The purpose of this Procedure is to assist construction in the pre-operational fabrication and testing of the pit leak detection system and the low point drain assembly by: (1) Control system testing of the pit leak detection system will be accomplished by actuating control switches and verifying that the control signal is initiated, liquid testing and overall operational requirements stated in HNF-SD-W314-PDS-003, ''Project Development Specification for Pit Leak Detection''. (2) Testing of the low point floor drain assembly by opening and closing the drain to and from the ''retracted'' and ''sealed'' positions. Successful operation of this drain will be to verify that the seal does not leak on the ''sealed'' position, the assembly holds liquid until the leak detector actuates and the assembly will operate from on top of the mock-up cover block.

  8. Design study of a fast spectrum zero-power reactor dedicated to source driven sub-critical experiments

    Energy Technology Data Exchange (ETDEWEB)

    Mercatali, L.; Serikov, A. [Forschungszentrum Karlsruhe, Institute for Neutron Physics and Reactor Technology, P.O. Box 3640, 76021 Karlsruhe (Germany); Baeten, P.; Uyttenhove, W. [SCK-CEN, Boeretang 200, B-2400 Mol (Belgium); Lafuente, A. [Univerisdad Politecnica de Madrid, 28006 Madrid (Spain); Teles, P. [Instituto Tecnologico e Nuclear, EN 10, 2680-953 Sacavem (Portugal)


    In the framework of the European P and T program (IFP6-EUROTRANS), the Generation of Uninterrupted Intense NEutrons pulses at the lead VEnus REactor (GUINEVERE) project consists of an Accelerator Driven System (ADS) that is composed by a fast lead simulated-cooled reactor operated in sub-critical conditions, coupled with an updated version of the GENEPI neutron generator previously used for the MUSE experiments. The GUINEVERE facility aims at developing and improving different techniques for the reactivity monitoring of sub-critical ADS's. As such, the GUINEVERE project will comprise a series of major experiments that will be performed in the near future. The GUINEVERE facility will be located at the VENUS light water moderated research reactor at the SCK-CEN site of Mol (Belgium), which needs to be modified in order to accommodate a completely different and new type of core. A series of constraints were taken into account in the technical design of the GUINEVERE core, in order to properly conjugate the technical feasibility of this facility and the necessity to comply with the envisioned experimental program and its associated scientific outcome. The complete design study of the GUINEVERE core is the subject of this paper. The final design of the fuel assemblies, safety and control rods is provided. Also, the critical core configuration, to be used as reference for absolute reactivity measurements, is presented along with its associated reactor physics parameters, calculated by means of Monte Carlo methodologies. Finally, for licensing purposes, the GUINEVERE facility must satisfy the required nuclear safety criteria of the Belgian safety authorities, and in this paper, an overview of the safety analysis that has been performed with regard to the core physics, thermal assessment and shielding issues is also provided. (author)

  9. Simulation in full-scale mock-ups: an ergonomics evaluation method?

    DEFF Research Database (Denmark)

    Andersen, Simone Nyholm; Broberg, Ole


    This paper presents and exploratory study of four simulation sessions in full-scale mock-ups of future hospital facilities.......This paper presents and exploratory study of four simulation sessions in full-scale mock-ups of future hospital facilities....

  10. Characterization of flaws in a tube bundle mock-up for reliability studies

    Energy Technology Data Exchange (ETDEWEB)

    Kupperman, D.S.; Bakhtiari, S. [Argonne National Lab., IL (United States)


    As part of an assessment of in-service inspection of steam generator tubes, the authors will assemble a steam generator mock-up for round robin studies and use as a test bed in evaluating emerging technologies. Progress is reported on the characterization of flaws that will be part of the mock-up. Eddy current and ultrasonic techniques are being evaluated as a means to characterize the flaws in the mock-up tubes before final assembly. Twenty Inconel 600 tubes with laboratory-grown cracks, typical of those to be used in the mock-up, were provided by Pacific Northwest National Laboratory for laboratory testing. After the tubes were inspected with eddy current and ultrasonic techniques, they were destructively analyzed to establish the actual depths, lengths, and profiles of the cracks. The analysis of the results will allow the best techniques to be used for characterizing the flaws in the mock-up tubes.

  11. Residual stress measurement inside a dissimilar metal weld mock-up of the pressurizer safety and relief nozzle

    Energy Technology Data Exchange (ETDEWEB)

    Campos, Wagner R.C.; Rabello, Emerson G.; Silva, Luiz L.; Mansur, Tanius R., E-mail:, E-mail:, E-mail:, E-mail: [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte (Brazil). Servico de Integridade Estrutural; Martins, Ketsia S., E-mail: [Universidade Federal de Minas Gerais (UFMG), Nelo Horizonte (Brazil). Departamento de Engenharia Metalurgica


    Residual stresses are present in materials or structural component in the absence of external loads or changes in temperatures. The most common causes of residual stresses being present are the manufacturing or assembling processes. All manufacturing processes, such as casting, welding, machining, molding, heat treatment, among others, introduces residual stresses into the manufactured object. The residual stresses effects could be beneficial or detrimental, depending on its distribution related to the component or structure, its load service and if it is compressive or tensile. In this work, the residual strains and stresses inside a mock-up that simulates the safety and relief nozzle of Angra 1 Nuclear Power Plant pressurizer were studied. The current paper presents a blind hole-drilling method residual stress measurements both at the inner surface of dissimilar metal welds of dissimilar metal weld nozzle mock-up. (author)

  12. Control strategies for managing energy in a building mock-up


    Eynard, Julien; Paris, Benjamin; Grieu, Stéphane; Talbert, Thierry; Thiéry, Frédérik


    8 pages; International audience; The present work focuses on investigating ways to enhance the energetic performance of buildings i.e. on proposing control strategies for managing energy in buildings. Therefore, control algorithms were tested using a prototype, composed of a building mock-up, a monitoring system and a data post-treatment software. The data acquisition system allows recording both mock-up indoor/outdoor temperatures and energy consumption. Two resistors serve as renewable and ...

  13. Thermo-mechanical tests of a CFC divertor mock-up (United States)

    Cardella, A.; Akiba, M.; Duwe, R.; Di Pietro, E.; Suzuki, S.; Satoh, K.; Reheis, N.


    Thermo-mechanical tests have been performed on a divertor mock-up consisting of a metallic tube armoured with five carbon fibre composite tiles. The tube is inserted inside the tiles and brazed with TiCuSil braze (monoblock concept). The tube material is TZM, a molybdenum alloy, and the armour material is SEP CARB N112, a high conductivity carbon-carbon composite. Using special surface preparation consisting of laser drilling, small (˜- 500 μm) holes in the composite have been made to increase the surface wetted by the braze and the resistance. The mock-up has been tested at the JAERI 400 kW electron beam test facility JEBIS. The aim of the test was to assess the performance of the mock-up in screening and thermal fatigue tests with particular attention to the behaviour of the armour to heat sink joint.

  14. Structural-hydraulic test of the liquid metal EURISOL target mock-up

    Energy Technology Data Exchange (ETDEWEB)

    Milenkovic, Rade Z. [Paul Scherrer Institut, CH-5232 Villigen PSI (Switzerland)], E-mail:; Dementjevs, Sergejs; Samec, Karel [Paul Scherrer Institut, CH-5232 Villigen PSI (Switzerland); Platacis, Ernests; Zik, Anatolij; Flerov, Aleksej [Institute of Physics of University of Latvia, LV-2156 Salaspils (Latvia); Manfrin, Enzo; Thomsen, Knud [Paul Scherrer Institut, CH-5232 Villigen PSI (Switzerland)


    Structural-hydraulic tests of the European Isotope Separation On-Line (EURISOL) neutron converter target mock-up, named MErcury Target EXperiment 1 (METEX 1), have been conducted by Paul Scherrer Institut (PSI, Switzerland) in cooperation with Institute of Physics of University of Latvia (IPUL, Latvia). PSI proceeded with extensive thermal-hydraulic and structural computational studies, followed by the target mock-up tests carried out on the mercury loop at IPUL. One of the main goals of the METEX 1 test is to investigate the hydraulic and structural behaviour of the EURISOL target mock-up for various inlet flow conditions (i.e. mass flow rates) and, in particular, for nominal operating flow rate and pressure in the system. The experimental results were analysed by advanced time-frequency methods such as Short-Time Fourier Transform in order to check the vibration characteristics of the mock-up and the resonance risk. The experimental results (obtained in METEX 1), which include inlet flow rate, pressure of the cover gas, total pressure loss, structural acceleration, sound and strain data, were jointly analysed together with numerical data obtained from Computational Fluid Dynamics (CFD)

  15. Structural-hydraulic test of the liquid metal EURISOL target mock-up (United States)

    Milenković, Rade Ž.; Dementjevs, Sergejs; Samec, Karel; Platacis, Ernests; Zik, Anatolij; Flerov, Aleksej; Manfrin, Enzo; Thomsen, Knud


    Structural-hydraulic tests of the European Isotope Separation On-Line (EURISOL) neutron converter target mock-up, named MErcury Target EXperiment 1 (METEX 1), have been conducted by Paul Scherrer Institut (PSI, Switzerland) in cooperation with Institute of Physics of the University of Latvia (IPUL, Latvia). PSI proceeded with extensive thermal-hydraulic and structural computational studies, followed by the target mock-up tests carried out on the mercury loop at IPUL. One of the main goals of the METEX 1 test is to investigate the hydraulic and structural behaviour of the EURISOL target mock-up for various inlet flow conditions (i.e. mass flow rates) and, in particular, for nominal operating flow rate and pressure in the system. The experimental results were analysed by advanced time-frequency methods such as Short-Time Fourier Transform in order to check the vibration characteristics of the mock-up and the resonance risk. The experimental results (obtained in METEX 1), which include inlet flow rate, pressure of the cover gas, total pressure loss, structural acceleration, sound and strain data, were jointly analysed together with numerical data obtained from Computational Fluid Dynamics (CFD).

  16. High heat flux testing of first wall mock-ups with and without neutron irradiation

    Directory of Open Access Journals (Sweden)

    G. Pintsuk


    For qualification, all three flat-tile mock-ups were exposed to cyclic steady state heat loads in the electron beam facility JUDITH-1 up to a maximum of 3.0MW/m2. Thereby, each tile was loaded individually as the full loading area exceeds the limits of the facility.

  17. High heat flux testing of EU tungsten monoblock mock-ups for the ITER divertor

    Energy Technology Data Exchange (ETDEWEB)

    Gavila, P., E-mail: [Fusion for Energy, 08019 Barcelona (Spain); Riccardi, B. [Fusion for Energy, 08019 Barcelona (Spain); Pintsuk, G. [Forschungszentrum Juelich, 52425 Juelich (Germany); Ritz, G. [AREVA NP, Centre Technique France, 71205 Le Creusot (France); Kuznetsov, V. [JCS “Efremov Institute”, Doroga na Metallostroy 3, Metallostroy, Saint-Petersburg 196641 (Russian Federation); Durocher, A. [ITER Organization, Route de Vinon sur Verdon, CS 90 046, 13067 Saint Paul-lez-Durance (France)


    Highlights: • All the tested items sustained the ITER Full W divertor qualification program requirements. This confirms that the technology for the manufacturing of the first set of the ITER Divertor is available in Europe. • The surface roughening and local melting of the W surface under high heat flux was proven to be significantly reduced for an armour thickness lower or equal to 6 mm. • However, this campaign highlighted some specific areas of improvement to be implemented ideally before the upcoming ITER Divertor IVT serial production. • The issue of the self-castellation of the W monoblocks, which typically appears after a few tenths of cycles at 20 MW/m{sup 2}, is critical because it generates some uncontrolled defects at the amour to heat sink joints. Besides, they create a gap which exposure is almost perpendicular to the magnetic field lines and which might lead to local W melting in the strike point region. • This campaign also evidenced that the minimum IO requirements on the CuCrZr ductility could be revised to avoid the occurrence of rather early fatigue failures. Although the W material characterization program has been set up by the IO, the strategy on the CuCrZr still needs to be defined. - Abstract: With the aim to assess the option to start the ITER operation with a full tungsten divertor, an R&D program was launched in order to evaluate the performances of tungsten (W) armoured plasma facing components (PFCs) under high heat flux. The F4E program consisted in the manufacturing and high heat flux (HHF) testing of W monoblock mock-ups and medium scale prototypes up to 20 MW/m{sup 2}. During the test campaign, 26 W mock-ups and two medium scale prototypes manufactured by Plansee SE (Austria) and by Ansaldo Nucleare (Italy) have been tested at the FE200 (AREVA, Le Creusot, France) and ITER Divertor Test Facility (IDTF) (Efremov Institute Saint Petersburg, Russian Federation) electron beam test facilities. The high heat flux (HHF) testing

  18. High heat flux test with HIP-bonded Ferritic Martensitic Steel mock-up for the first wall of the KO HCML TBM

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Won, E-mail: [Korea Atomic Energy Research Institute, Deokjin-dong, Yuseong-gu, Daejeon 305-600 (Korea, Republic of); Bae, Young Dug; Kim, Suk Kwon; Shin, Hee Shin; Hong, Bong Guen [Korea Atomic Energy Research Institute, Deokjin-dong, Yuseong-gu, Daejeon 305-600 (Korea, Republic of); Bang, In Cheol, E-mail: [Ulsan National Institute of Science and Technology, 100 Banyeon-ri, Eonyang-eup, Ulju-gun, Ulsan 689-798 (Korea, Republic of)


    In order for a Korean Helium Cooled Molten Lithium (HCML) Test Blanket Module (TBM) to be tested in the International Thermonuclear Experimental Reactor (ITER), fabrication method for the TBM FW such as Hot Isostatic Pressing (HIP, 1050 deg. C, 100 MPa, 2 h) has been developed including post HIP heat treatment (PHHT, normalizing at 950 deg. C for 2 h and tempering at 750 deg. C for 2 h) with Ferritic Martensitic Steel (FMS). Several mock-ups were fabricated using the developed methods and one of them, three-channel mock-up, was used for performing a High Heat Flux (HHF) test to verify the joint integrity. Test conditions were determined using the commercial code, ANSYS-11, and the test was performed in the Korea Heat Load Test (KoHLT) facility, which was used a radiation heating with a graphite heater. The mock-up survived up to 1000 cycles under 1.0 MW/m{sup 2} heat flux and there is no delamination or failure during the test.

  19. Wax-up and mock-up. A guide for anterior periodontal and restorative treatments. (United States)

    Gurrea, Jon; Bruguera, August


    When starting a case, having the end result in mind is the basis in any kind of treatment, even more so in those where the anterior teeth morphology, size and proportion will be changed. Here is where a good treatment plan based on a diagnostic wax-up that is tried in with a mock-up and approved by the patient becomes crucial. This case report exemplifies how transferring the information from the diagnostic wax up to the patient's mouth is of help not only to the restorative dentist and the laboratory technician, but also to the surgeon when performing the crown lengthening. This treatment plan cannot be seen as a sequence of isolated procedures but as a single workflow. The wax-up/mock-up binomial is a guide even for the periodontist in a novel approach to surgical crown lengthening.

  20. A zero-power radio receiver.

    Energy Technology Data Exchange (ETDEWEB)

    Brocato, Robert Wesley


    This report describes both a general methodology and some specific examples of passive radio receivers. A passive radio receiver uses no direct electrical power but makes sole use of the power available in the radio spectrum. These radio receivers are suitable as low data-rate receivers or passive alerting devices for standard, high power radio receivers. Some zero-power radio architectures exhibit significant improvements in range with the addition of very low power amplifiers or signal processing electronics. These ultra-low power radios are also discussed and compared to the purely zero-power approaches.

  1. Clinically based diagnostic wax-up for optimal esthetics: the diagnostic mock-up. (United States)

    Simon, Harel; Magne, Pascal


    A diagnostic wax-up can enhance the predictability of treatment by modeling the desired result in wax prior to treatment. It is critical to correlate the wax-up to the patient to avoid a result that appears optimal on the casts but does not correspond to the patient's smile. This article reviews the applications and techniques for clinically based diagnostic wax-up, and focuses on the diagnostic mock-up philosophy as a means to obtain predictable esthetics and function.

  2. An experimental study of the characteristics of a mock up of a centrifugal conduction magnetohydrodynamic pump

    Energy Technology Data Exchange (ETDEWEB)

    Gorbunov, V.A.; Frolov, V.V.; Kolesnikov, Yu.B.; Kolokolov, V.Ye.; Polyakov, N.N.


    The design of a mock up of a centrifugal conduction magnetohydrodynamic (MGD) pump is described. The dependences of the pressure developed by the pump in a locked mode on the magnetic induction and the operational current are cited, along with the flow rate and pressure characteristics of the pump. The dependences of the characteristics of the pump on the dimensions of the operational zone and the conductivity of the facial walls are experimentally studied.

  3. Thermo-mechanical study of high heat flux component mock-ups for ITER TBM

    Energy Technology Data Exchange (ETDEWEB)

    Bonelli, Flavia [Institute for Neutron Physics and Reactor Technology, Karlsruhe Institute of Technology (Germany); Dipartimento Energia, Politecnico di Torino (Italy); Boccaccini, Lorenzo Virgilio, E-mail: [Institute for Neutron Physics and Reactor Technology, Karlsruhe Institute of Technology (Germany); Kunze, André; Maione, Ivan Alessio [Institute for Neutron Physics and Reactor Technology, Karlsruhe Institute of Technology (Germany); Savoldi, Laura; Zanino, Roberto [Dipartimento Energia, Politecnico di Torino (Italy)


    Highlights: • Infrared radiation heaters for test of plasma facing component available at KIT. • Numerical model developed and validated to check uniformity of heat flux. • Thermo-mechanical calculations performed on a mock-up of the HCPB TBM FW. • Assessment done of representativity of stress conditions for the ITER TBMs. - Abstract: Commercial infrared heaters have been proposed to be used in the HELOKA facility under construction at Karlsruhe Institute of Technology (KIT) to test a mock-up of the first wall (FW), called thermo-cycle mock-up (TCM) plate, under stress loading comparable to those experienced by the test blanket modules (TBMs) in ITER. Two related issues are analyzed in this paper, in relation to the ongoing European project aimed at the design of the two EU TBMs: (1) the possibility to reproduce, by means of those heaters, high heat flux loading conditions on the TCM plate similar to those expected on the ITER TBMs, and (2) the thermo-mechanical analysis of the TCM itself, in order to define a suitable choice of experimental parameters and mechanical constraints leading to a relevant stress condition. A suitable heater model is developed and validated against experimental data from an ad-hoc test campaign. A thermo-mechanical study of the TCM plate is presented, showing that the structure is able to withstand high thermal loads, even in the most constrained case, reaching stress levels comparable to the ITER TBM.

  4. Performance test results of helium gas circulator of mock-up test facility with full-scale reaction tube for HTTR hydrogen production system. Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Shimizu, Akira; Kato, Michio; Hayashi, Koji [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment] [and others


    Hydrogen production system by steam reforming of methane will be connected to the High Temperature Engineering Test Reactor (HTTR) of the Japan Atomic Energy Research Institute (JAERI) against development of nuclear heat utilization system. To obtain design and safety database of the HTTR hydrogen production system, mock-up test facility with full-scale reaction was constructed in FY 2001 and hydrogen of 120m{sup 3}N{sub /}h was successfully produced in overall performance test. This report describes performance test results of a helium gas circulator in this facility. The circulator performance curves regarding to pressure-rise, input power and adiabatic thermal efficiency at standard revolution number were made based on the measured flow-rate, temperature and pressure data in overall performance test. The circulator performance prediction code was made based on these performance curves. The code can calculate revolution number, electric power and temperature-rise of the circulator using flow-rate, inlet temperature, inlet pressure and pressure-rise data. The verification of the code was carried out with the test data in FY 2002. Total pressure loss of the helium gas circulation loop was also evaluated. The circulator should be operated in conditions such as pressure from 2.7MPa to 4.0MPa and flow-rate from 250g/s to 400g/s and at maximum pressure-rise of 250 kPa in test operation. It was confirmed in above verification and evaluations that the circulator had performance to satisfy above conditions within operation limitation of the circulator such as maximum input-power of 150 kW and maximum revolution number of 12,000 rpm. (author)

  5. Thermodynamic analysis of the advanced zero emission power plant

    Directory of Open Access Journals (Sweden)

    Kotowicz Janusz


    Full Text Available The paper presents the structure and parameters of advanced zero emission power plant (AZEP. This concept is based on the replacement of the combustion chamber in a gas turbine by the membrane reactor. The reactor has three basic functions: (i oxygen separation from the air through the membrane, (ii combustion of the fuel, and (iii heat transfer to heat the oxygen-depleted air. In the discussed unit hot depleted air is expanded in a turbine and further feeds a bottoming steam cycle (BSC through the main heat recovery steam generator (HRSG. Flue gas leaving the membrane reactor feeds the second HRSG. The flue gas consist mainly of CO2 and water vapor, thus, CO2 separation involves only the flue gas drying. Results of the thermodynamic analysis of described power plant are presented.

  6. Thermodynamic analysis of the advanced zero emission power plant (United States)

    Kotowicz, Janusz; Job, Marcin


    The paper presents the structure and parameters of advanced zero emission power plant (AZEP). This concept is based on the replacement of the combustion chamber in a gas turbine by the membrane reactor. The reactor has three basic functions: (i) oxygen separation from the air through the membrane, (ii) combustion of the fuel, and (iii) heat transfer to heat the oxygen-depleted air. In the discussed unit hot depleted air is expanded in a turbine and further feeds a bottoming steam cycle (BSC) through the main heat recovery steam generator (HRSG). Flue gas leaving the membrane reactor feeds the second HRSG. The flue gas consist mainly of CO2 and water vapor, thus, CO2 separation involves only the flue gas drying. Results of the thermodynamic analysis of described power plant are presented.

  7. NDE of explosion welded copper stainless steel first wall mock-up

    Energy Technology Data Exchange (ETDEWEB)

    Taehtinen, S.; Kauppinen, P.; Jeskanen, H.; Lahdenperae, K.; Ehrnsten, U. [VTT Manufacturing Technology, Espoo (Finland). Materials and Structural Integrity


    The study showed that reflection type C-mode scanning acoustic microscope (C-SAM) and internal ultrasonic inspection (IRIS) equipment can be applied for ultrasonic examination of copper stainless steel compound structures of ITER first wall mock-ups. Explosive welding can be applied to manufacture fully bonded copper stainless steel compound plates. However, explosives can be applied only for mechanical tightening of stainless steel cooling tubes within copper plate. If metallurgical bonding between stainless steel tubes and copper plate is required Hot Isostatic Pressing (HIP) method can be applied. (orig.)

  8. Structural-hydraulic test of the liquid metal EURISOL target mock-up

    CERN Document Server

    Milenković, Rade Ž; Platacis, Ernests; Flerov, Aleksej; Samec, Karel; Dementjevs, Sergejs; Manfrin, Enzo; Thomsen, Knud


    Structural-hydraulictestsoftheEuropeanIsotopeSeparationOn-Line(EURISOL)neutronconverter target mock-up,namedMErcuryTargetEXperiment1(METEX1),havebeenconductedbyPaulScherrer Institut(PSI,Switzerland)incooperationwithInstituteofPhysicsoftheUniversityofLatvia(IPUL, Latvia).PSIproceededwithextensivethermal-hydraulicandstructuralcomputationalstudies,followed by thetargetmock-uptestscarriedoutonthemercuryloopatIPUL. One ofthemaingoalsoftheMETEX1testistoinvestigatethehydraulicandstructuralbehaviour of theEURISOLtargetmock-upforvariousinletflowconditions(i.e.massflowrates)and,inparticular, for nominaloperatingflowrateandpressureinthesystem.Theexperimentalresultswereanalysedby advancedtime–frequencymethodssuchasShort-TimeFourierTransforminordertocheckthe vibration characteristicsofthemock-upandtheresonancerisk.Theexperimentalresults(obtainedin METEX 1),whichincludeinletflowrate,pressureofthecovergas,totalpressureloss,structural acceleration,soundandstraindata,werejointlyanalysedtogetherwithnumericaldataobtainedfrom ...

  9. Parallel Execution of Functional Mock-up Units in Buildings Modeling

    Energy Technology Data Exchange (ETDEWEB)

    Ozmen, Ozgur [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Nutaro, James J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); New, Joshua Ryan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)


    A Functional Mock-up Interface (FMI) defines a standardized interface to be used in computer simulations to develop complex cyber-physical systems. FMI implementation by a software modeling tool enables the creation of a simulation model that can be interconnected, or the creation of a software library called a Functional Mock-up Unit (FMU). This report describes an FMU wrapper implementation that imports FMUs into a C++ environment and uses an Euler solver that executes FMUs in parallel using Open Multi-Processing (OpenMP). The purpose of this report is to elucidate the runtime performance of the solver when a multi-component system is imported as a single FMU (for the whole system) or as multiple FMUs (for different groups of components as sub-systems). This performance comparison is conducted using two test cases: (1) a simple, multi-tank problem; and (2) a more realistic use case based on the Modelica Buildings Library. In both test cases, the performance gains are promising when each FMU consists of a large number of states and state events that are wrapped in a single FMU. Load balancing is demonstrated to be a critical factor in speeding up parallel execution of multiple FMUs.

  10. Laser vibrometry vibration measurements on vehicle cabins in running conditions: helicopter mock-up application (United States)

    Revel, Gian Marco; Castellini, Paolo; Chiariotti, Paolo; Tomasini, Enrico Primo; Cenedese, Fausto; Perazzolo, Alessandro


    The present work deals with the analysis of problems and potentials of laser vibrometer measurements inside vehicle cabins in running conditions, with particular reference to helicopters where interior vibro-acoustic issues are very important. This paper describes the results of a systematic measurement campaign performed on an Agusta A109MKII mock-up. The aim is to evaluate the applicability of scanning laser Doppler vibrometer (SLDV) for tests in simulated flying conditions and to understand how performances of the technique are affected when the laser head is placed inside the cabin, thus being subjected to interfering inputs. First a brief description of the performed test cases and the used measuring set-ups are given. Comparative tests between the SLDV and accelerometers are presented, analyzing the achievable performances for the specific application. Results obtained measuring with the SLDV placed inside the helicopter cabin during operative excitation conditions are compared with those performed with the laser lying outside the mock-up, these last being considered as ``reference measurements.'' Finally, in order to give an estimate of the uncertainty level on measured signals, a study linking the admitted percentage of noise content on vibrometer signals due to laser head vibration levels will be introduced.


    Energy Technology Data Exchange (ETDEWEB)

    Stefanko, D.; Langton, C.


    The United States Department of Energy (US DOE) has determined that Tanks 18-F and 19-F have met the F-Tank Farm (FTF) General Closure Plan Requirements and are ready to be permanently closed. The high-level waste (HLW) tanks have been isolated from FTF facilities. To complete operational closure they will be filled with grout for the purpose of: (1) physically stabilizing the tanks, (2) limiting/eliminating vertical pathways to residual waste, (3) discouraging future intrusion, and (4) providing an alkaline, chemical reducing environment within the closure boundary to control speciation and solubility of select radionuclides. Bulk waste removal and heel removal equipment remain in Tanks 18-F and 19-F. This equipment includes the Advance Design Mixer Pump (ADMP), transfer pumps, transfer jets, standard slurry mixer pumps, equipment-support masts, sampling masts, dip tube assemblies and robotic crawlers. The present Tank 18 and 19-F closure strategy is to grout the equipment in place and eliminate vertical pathways by filling voids in the equipment to vertical fast pathways and water infiltration. The mock-up tests described in this report were intended to address placement issues identified for grouting the equipment that will be left in Tank 18-F and Tank 19-F. The Tank 18-F and 19-F closure strategy document states that one of the Performance Assessment (PA) requirements for a closed tank is that equipment remaining in the tank be filled to the extent practical and that vertical flow paths 1 inch and larger be grouted. The specific objectives of the Tier 1A equipment grout mock-up testing include: (1) Identifying the most limiting equipment configurations with respect to internal void space filling; (2) Specifying and constructing initial test geometries and forms that represent scaled boundary conditions; (3) Identifying a target grout rheology for evaluation in the scaled mock-up configurations; (4) Scaling-up production of a grout mix with the target rheology


    Energy Technology Data Exchange (ETDEWEB)

    Ronald Bischoff; Stephen Doyle


    Clean Energy Systems (CES) was previously funded by DOE's ''Vision 21'' program. This program provided a proof-of-concept demonstration that CES' novel gas generator (combustor) enabled production of electrical power from fossil fuels without pollution. CES has used current DOE funding for additional design study exercises which established the utility of the CES-cycle for retrofitting existing power plants for zero-emission operations and for incorporation in zero-emission, ''green field'' power plant concepts. DOE funding also helped define the suitability of existing steam turbine designs for use in the CES-cycle and explored the use of aero-derivative turbines for advanced power plant designs. This work is of interest to the California Energy Commission (CEC) and the Norwegian Ministry of Petroleum & Energy. California's air quality districts have significant non-attainment areas in which CES technology can help. CEC is currently funding a CES-cycle technology demonstration near Bakersfield, CA. The Norwegian government is supporting conceptual studies for a proposed 40 MW zero-emission power plant in Stavager, Norway which would use the CES-cycle. The latter project is called Zero-Emission Norwegian Gas (ZENG). In summary, current engineering studies: (1) supported engineering design of plant subsystems applicable for use with CES-cycle zero-emission power plants, and (2) documented the suitability and availability of steam turbines for use in CES-cycle power plants, with particular relevance to the Norwegian ZENG Project.

  13. Mock-up stands for a rotating target for CSNS project (United States)

    Jia, X. J.; Bauer, G. S.; He, W.; Sun, Y. L.; Liang, T. J.; Yin, W.; Zhao, D.


    This paper summarises pre-conceptual solutions for all sub-units of a potential rotating target system for the CSNS project. In order to test the validity of this concept and to gain first experience with a rotating target, the CSNS project has decided to embark on the construction of a mock-up test stand. The purpose is to provide first demonstration of the viability of the above concept by using a full model of the target head components and shaft and a dummy target with the right diameter and weight; confirm that acceptance criteria can be reached; gain experience in running a rotating target; verify certain parameters obtained by calculations. By carrying out a development program, it should be possible to produce a sound basis for a decision as to whether or not CSNS wants to adopt a rotating target as the preferred solution.

  14. Comparison of TOFD and Radiographic Testing for a Mock-up Specimen

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chung Jick; Jeon, Jong Gun; Kim, Jin Taek [Seoul Inspection and Testing Company, Seoul (Korea, Republic of)


    In order to detect the internal defects which occur in welding parts of pressure vessel and structures, radiographic testing and ultrasonic testing is applied. However, because of the risks of radiation exposure and film processing, radiographic testing takes a relatively long time to verify the test results and it has affected in the production process. Typically, the manual ultrasonic testing is not easy to reproduce the result and it is highly dependent on the tester's skills. The TOFD technique, one of the automatic ultrasonic testings is spreading alternatively. This research describes the comparing test results by applying radiographic testing and TOFD technique to a mock-up specimen including the flaws. The TOFD technique will contribute to improve the objective reliability of the ultrasonic technique.

  15. Designing a zero emissions power switch locomotive

    Energy Technology Data Exchange (ETDEWEB)

    Shapiro, J.; Hines, J. [National Instruments, Austin, TX (United States)


    In addition to providing electric power and drinking water in manned spacecraft, fuel cell power plants have provided safe, clean electric power to hospitals, universities and other facilities since the early 1990s. This paper described a zero emissions hydrogen and battery-powered hybrid switching locomotive designed for use in rail, port and military base applications. Designed in partnership with a consortium, the prototype hybrid switching locomotive is comprised of a number of proven commercial technologies and includes a control system developed by National Instruments. New applications for hydrogen fuel cell use in industrial vehicles were also discussed. The new design was scheduled for field testing at the end of 2008.

  16. Multimegawatt space power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dearien, J.A.; Whitbeck, J.F.


    In response to the need of the Strategic Defense Initiative (SDI) and long range space exploration and extra-terrestrial basing by the National Air and Space Administration (NASA), concepts for nuclear power systems in the multi-megawatt levels are being designed and evaluated. The requirements for these power systems are being driven primarily by the need to minimize weight and maximize safety and reliability. This paper will discuss the present requirements for space based advanced power systems, technological issues associated with the development of these advanced nuclear power systems, and some of the concepts proposed for generating large amounts of power in space. 31 figs.

  17. Chemically deposited tungsten fibre-reinforced tungsten – The way to a mock-up for divertor applications

    Directory of Open Access Journals (Sweden)

    J. Riesch


    Full Text Available The development of advanced materials is essential for sophisticated energy systems like a future fusion reactor. Tungsten fibre-reinforced tungsten composites (Wf/W utilize extrinsic toughening mechanisms and therefore overcome the intrinsic brittleness of tungsten at low temperature and its sensitivity to operational embrittlement. This material has been successfully produced and tested during the last years and the focus is now put on the technological realisation for the use in plasma facing components of fusion devices. In this contribution, we present a way to utilize Wf/W composites for divertor applications by a fabrication route based on the chemical vapour deposition (CVD of tungsten. Mock-ups based on the ITER typical design can be realized by the implementation of Wf/W tiles. A concept based on a layered deposition approach allows the production of such tiles in the required geometry. One fibre layer after the other is positioned and ingrown into the W-matrix until the final sample size is reached. Charpy impact tests on these samples showed an increased fracture energy mainly due to the ductile deformation of the tungsten fibres. The use of Wf/W could broaden the operation temperature window of tungsten significantly and mitigate problems of deep cracking occurring typically in cyclic high heat flux loading. Textile techniques are utilized to optimise the tungsten wire positioning and process speed of preform production. A new device dedicated to the chemical deposition of W enhances significantly, the available machine time for processing and optimisation. Modelling shows that good deposition results are achievable by the use of a convectional flow and a directed temperature profile in an infiltration process.

  18. LMFBR type reactor and power generation system using the same

    Energy Technology Data Exchange (ETDEWEB)

    Otsubo, Akira.


    A reactor core void reactivity of a reactor main body is set to negative or zero. A heat insulation structure is disposed on the inner wall surface of a reactor container. Oxide fuels or nitride fuels are used. A fuel pin cladding tube has a double walled structure having an outer side of stainless steel and an inner side of niobium alloy. Upon imaginary event, boiling is allowed. Even if boiling of coolants should occur by temperature elevation of fuels upon imaginary event, since reactor core fuels comprises oxides or nitrides, they have a heat resistance, further, and since the fuel pin cladding tube has super heat resistance, it has a high temperature strength, so that it is not ruptured and durable to the coolant boiling temperature. Since the reactor core void reactivity is negative or zero, the reactor core is in a subcritical state by the boiling, and the reactor core power is reduced to several % of the rated power. Accordingly, boiling and non-boiling are repeated substantially permanently in the reactor core, during which safety can be kept with no operator's handling. Further, heat generated in the reactor core is gradually removed by an air cooling system for the reactor container. (N.H.).

  19. ITER first mirror mock-ups exposed in Magnum-PSI (United States)

    Marot, L.; De Temmerman, G.; van den Berg, M. A.; Renault, P.-O.; Covarel, G.; Joanny, M.; Travère, J. M.; Steiner, R.; Mathys, D.; Meyer, E.


    The goal of this work was to investigate coated first mirrors under very harsh erosion conditions. Mock-up mirrors were exposed to high-flux hydrogen/argon plasma in the linear plasma facility Magnum-PSI. Rhodium (Rh) and molybdenum (Mo) coated mirrors of different coating thicknesses, with or without water cooling, exhibited different responses to this exposure. Failures of Rh films were demonstrated for 5 micron thick film, 1 micron film revealed 10% decrease in the specular reflectivity only in the exposed area. In comparison, water cooled Mo mock-ups showed a significant diffuse reflectivity on the entire surface leading to more than 50% specular reflectivity losses in the visible range. The losses for non-cooled Mo samples did not exceed 7% in the whole studied wavelength range of 250-2500 nm. Three phenomena were proposed to explain these results. First the mechanical properties of the films as characterized by scratch and hardness measurements as well as residual stress analysis measured by x-ray diffraction. Rh films showed a high compressive stress value of 2.5  ±  0.4 GPa leading to poor adhesion of the thick films deposited on stainless steel substrate due to the high amount of available energy per area stored in the unbuckled film i.e. {{G}0}>30 J m-2. It was confirmed by ANSYS simulation that the von Mises stress for the Rh coating was twice as high as that for the Mo coating due to different mechanical properties. Moreover, the maximum stress for thick Rh film (261 MPa) was higher than the critical buckling stress calculated with a buckle clamped Euler column model demonstrating the failure mode of the film. The second phenomenon was roughening of the mirror surface which was flux and temperature dependent, i.e. at low temperatures the surface would roughen randomly without any oriented surface morphology and at higher temperatures the surface diffusion constants would dominate the process and smoothen the surface. The last phenomenon

  20. Mock-up Test for Isotope Target Transport and Cooling System

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Sang-pil; Kwon, Hyeok-jung; Kim, Han-sung; Cho, Yong-sub; Chung, Bo-hyun; Seol, Kyung-tae; Song, Young-gi; Kim, Dae-il; Min, Yi-sub [KOMAC, Gyeongju (Korea, Republic of)


    In this paper, we described the design and fabrication of the test mock-up of target transport and cooling system for the isotope production by using the 100-MeV proton irradiation. For Sr-82 production, RbCl target and aluminum dummy target was prepared. These targets are contained in the target carrier, which could transported by drive chain and guide rail system. Korea multi-purpose Accelerator Complex (KOMAC) has a plan to construct the new proton beam irradiation facility for the production of radioisotopes. Sr-82 and Cu-67 were selected as the target isotope in this facility, they are promising isotope for the PET imaging and cancer therapy. To produce Sr-82 by 100- MeV proton irradiation, RbCl were chosen as a target material due to their high melting point and easy separation. For the facility construction, we have designed targetry system which consists of target, target transport system and target cooling system. This paper describes the details of targetry system.

  1. The HIE-ISOLDE alignment and monitoring system software and test mock up

    CERN Document Server

    Kautzmann, G; Kadi, Y; Leclercq, Y; Waniorek, S; Williams, L


    For the HIE Isolde project a superconducting linac will be built at CERN in the Isolde facility area. The linac will be based on the creation and installation of 2 high- β and 4 low- β cryomodules containing respectively 5 high-β superconducting cavities and 1 superconducting solenoid for the two first ones, 6 low-β superconducting cavities and 2 superconducting solenoids for the four other ones. An alignment and monitoring system of the RF cavities and solenoids placed inside the cryomodules is needed to reach the optimum linac working conditions. The alignment system is based on opto-electronics, optics and precise mechanical instrumentation. The geometrical frame configuration, the data acquisition and the 3D adjustment will be managed using a dedicated software application. In parallel to the software development, an alignment system test mock-up has been built for software validation and dimensional tests. This paper will present the software concept and the development status, and then will describe...

  2. Characterization of ITER tungsten qualification mock-ups exposed to high cyclic thermal loads

    Energy Technology Data Exchange (ETDEWEB)

    Pintsuk, Gerald, E-mail: [Forschungszentrum Jülich GmbH, D-52425 Jülich (Germany); Bednarek, Maja; Gavila, Pierre [Fusion for Energy, E-08019 Barcelona (Spain); Gerzoskovitz, Stefan [Plansee SE, Innovation Services, 6600 Reutte (Austria); Linke, Jochen [Forschungszentrum Jülich GmbH, D-52425 Jülich (Germany); Lorenzetto, Patrick; Riccardi, Bruno [Fusion for Energy, E-08019 Barcelona (Spain); Escourbiac, Frederic [ITER Organization, Route de Vinon sur Verdon, CS 90 046, 13067 Saint Paul lez Durance (France)


    Highlights: • Mechanical deformation of CuCrZr in case a thermal barrier layer has been formed due to impurity content in the cooling water. • Crack formation at the W/Cu interface starting at the block edge. • Porosity formation in the pure Cu interlayer. • Microstructural changes in tungsten down to the W/Cu interface, which indicates also high temperatures for the pure Cu interlayer. • Macrocrack formation in tungsten which is assumed to be ductile at the initiation point and brittle when proceeding toward the cooling tube. - Abstract: High heat flux tested small-scale tungsten monoblock mock-ups (5000 cycles at 10 MW/m{sup 2} and up to 1000 cycles at 20 MW/m{sup 2}) manufactured by Plansee and Ansaldo were characterized by metallographic means. Therein, the macrocrack formation and propagation in tungsten, its recrystallization behavior and the surface response to different heat load facilities were investigated. Furthermore, debonding at the W/Cu interface, void formation in the soft copper interlayer and microcrack formation at the inner surface of the CuCrZr cooling tube were found.

  3. Structural analysis, design and evaluation of mock-up platform, monorail, and tank plate cut-out

    Energy Technology Data Exchange (ETDEWEB)

    Hundal, T.S.


    Platform - Structural analyses were performed for design seismic, live and dead load combinations for the freestanding platform over the partial DST mock-up section. The platform is to be used for Robotic ultrasonic inspection of the tank wall. It is a free standing structure anchored to floor slab with Hilti Kwik bolts.

  4. Fabrication of W/FMS joint mock-ups using a hot isostatic pressing

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Yang-Il, E-mail: [Korea Atomic Energy Research Institute, 989-111 Daedeokdearo, Yuseong, Daejeon 305-353 (Korea, Republic of); Park, Jeong-Yong; Choi, Byoung-Kwon; Lee, Dong-Won [Korea Atomic Energy Research Institute, 989-111 Daedeokdearo, Yuseong, Daejeon 305-353 (Korea, Republic of); Cho, Seungyon [National Fusion Research Institute, 169-148 Gwahangno, Yuseong, Daejeon 305-806 (Korea, Republic of)


    Highlights: • Tungsten was joined on a ferritic–martensitic steel using a hot isostatic pressing method. • A double-stage HIP was performed to avoid the edge-delamination during a post HIP heat-treatment. • Ti foil was used to minimize the thermal expansion difference between W and ferritic–martensitic steel. • Mo foil was used as a separator not to form a bonding between W and canned materials. • No significant defects or a brittle failure were observed along the joint interface. - Abstract: Blocks of tungsten and ferritic–martensitic steel (FMS) were joined without any interfacial defects or cracks. For the joining, two times of a hot isostatic pressing (HIP) were performed. The first HIP (900 °C, 100 MPa, 1.5 h) facilitates the diffusion bonding between W and FMS. The second HIP (750 °C, 70 MPa, 2 h) corresponds to a tempering process to retain the mechanical properties of the FMS. As an interlayer material, titanium foil that can mitigate the thermal expansion difference between W and FMS was used. In addition, a molybdenum foil was inserted to prevent an unwanted bonding of W to a canning material. The lateral cracks in W plates, which were usually observed in the case of a conventional HIP process, were not observed when the molybdenum separator was used. W/FMS joint mock-ups with a dimension of 50 mm × 50 mm × 32 mm (T) were successfully fabricated. The shear strength of the joints was 89 MPa on average.

  5. Tokamak power reactor ignition and time dependent fractional power operation

    Energy Technology Data Exchange (ETDEWEB)

    Vold, E.L.; Mau, T.K.; Conn, R.W.


    A flexible time-dependent and zero-dimensional plasma burn code with radial profiles was developed and employed to study the fractional power operation and the thermal burn control options for an INTOR-sized tokamak reactor. The code includes alpha thermalization and a time-dependent transport loss which can be represented by any one of several currently popular scaling laws for energy confinement time. Ignition parameters were found to vary widely in density-temperature (n-T) space for the range of scaling laws examined. Critical ignition issues were found to include the extent of confinement time degradation by alpha heating, the ratio of ion to electron transport power loss, and effect of auxiliary heating on confinement. Feedback control of the auxiliary power and ion fuel sources are shown to provide thermal stability near the ignition curve.

  6. UV-Induced Anisotropy In CdBr2-CdBr2: Cu Nanostructures

    Directory of Open Access Journals (Sweden)

    El-Naggar A. M.


    Full Text Available We have found an occurrence of anisotropy in the nanostructure CdBr2-CdBr2: Cu nanocrystalline films. The film thickness was varied from 4 nm up to 80 nm. The films were prepared by successive deposition of the novel layers onto the basic nanocrystals. The detection of anisotropy was performed by occurrence of anisotropy in the polarized light at 633 nm He-Ne laser wavelength. The occurrence of anisotropy was substantially dependent on the film thickness and the photoinduced power density. Possible mechanisms of the observed phenomena are discussed.

  7. MockUp环境下的虚拟装配方法%On the Virtual Assembly in MockUp

    Institute of Scientific and Technical Information of China (English)

    黄延军; 戴春祥; 胡庆夕; 方明伦


    This article introduces the detection of collision during virtual assembly. And that the virtual assembly of some type of industrial sartorius was accomplished in software MockUp is discussed.%介绍了虚拟装配过程中的干涉检测方法.并探讨了运用MockUp实现某工业平缝机的压杆部件的虚拟装配方法.

  8. Experiments of multichannel least-square methods for sound field reproduction inside aircraft mock-up: Objective evaluations (United States)

    Gauthier, P.-A.; Camier, C.; Lebel, F.-A.; Pasco, Y.; Berry, A.; Langlois, J.; Verron, C.; Guastavino, C.


    Sound environment reproduction of various flight conditions in aircraft mock-ups is a valuable tool for the study, prediction, demonstration and jury testing of interior aircraft sound quality and annoyance. To provide a faithful reproduced sound environment, time, frequency and spatial characteristics should be preserved. Physical sound field reproduction methods for spatial sound reproduction are mandatory to immerse the listener's body in the proper sound fields so that localization cues are recreated at the listener's ears. Vehicle mock-ups pose specific problems for sound field reproduction. Confined spaces, needs for invisible sound sources and very specific acoustical environment make the use of open-loop sound field reproduction technologies such as wave field synthesis (based on free-field models of monopole sources) not ideal. In this paper, experiments in an aircraft mock-up with multichannel least-square methods and equalization are reported. The novelty is the actual implementation of sound field reproduction with 3180 transfer paths and trim panel reproduction sources in laboratory conditions with a synthetic target sound field. The paper presents objective evaluations of reproduced sound fields using various metrics as well as sound field extrapolation and sound field characterization.

  9. PWSCC susceptibility evaluation of mill annealed Alloy 600 SG tubings using mock-up specimens in lithium enhanced 360 C water

    Energy Technology Data Exchange (ETDEWEB)

    Tsuruta, T.; Okamoto, S.; Kitera, T. [Mitsubishi Heavy Industries, Ltd., Takasago (Japan); Takamatsu, H.; Matsunaga, T. [Kansai Electric Power Co., Inc., Osaka (Japan)


    Intergranular stress corrosion cracking (IGSCC) behavior of Alloy 600 steam generator (SG) tubing in pressurized water reactor (PWR) primary water environments, or so called PWSCC, has been investigated since the first report by Coriou in 1959. The acceleration of PWSCC by elevating the test temperature up to 365 C, was first reported in 1980 by Bulischeck et al.. Since then, many results of PWSCC phenomena have been reported. It is important for preventive maintenance planning, to predict the life expectation at the operating temperature for a specific structural design. PWSCC susceptibility of mill annealed alloy 600 (MA 600) SG tubing was investigated using mock-up specimens, such as tube sheet expansion specimens and dent shape deformed specimens in lithium enhanced 360 C water (500ppm as B, 20ppm as Li, 45ccH{sub 2}/kgH{sub 2}O). Uniaxial constant load (UCL) tests were carried out for MA 600, thermally treated (TT) Alloy 600, and TT 690 to evaluate the effect of stress, and to evaluate the acceleration factor of the environment compared with the results in normal 360 C reactor coolant system (RCS) water (500ppm as B, 2ppm as Li, 30ccH{sub 2}/kgH{sub 2}O).

  10. Transmutation of actinides in power reactors. (United States)

    Bergelson, B R; Gerasimov, A S; Tikhomirov, G V


    Power reactors can be used for partial short-term transmutation of radwaste. This transmutation is beneficial in terms of subsequent storage conditions for spent fuel in long-term storage facilities. CANDU-type reactors can transmute the main minor actinides from two or three reactors of the VVER-1000 type. A VVER-1000-type reactor can operate in a self-service mode with transmutation of its own actinides.

  11. Power Control Method for Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Baang, Dane; Suh, Yongsuk; Park, Cheol [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)


    Considering safety-oriented design concept and other control environment, we developed a simple controller that provides limiting function of power change- rate as well as fine tracking performance. The design result has been well-proven via simulation and actual application to a TRIGA-II type research reactor. The proposed controller is designed to track the PDM(Power Demand) from operator input as long as maintaining the power change rate lower than a certain value for stable reactor operation. A power control method for a TRIGA-II type research reactor has been designed, simulated, and applied to actual reactor. The control performance during commissioning test shows that the proposed controller provides fine control performance for various changes in reference values (PDM), even though there is large measurement noise from neutron detectors. The overshoot at low power level is acceptable in a sense of reactor operation.

  12. Tritium breeding mock-up experiments containing lithium titanate ceramic pebbles and lead irradiated with DT neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Jakhar, Shrichand; Abhangi, M.; Tiwari, S. [Institute for Plasma Research, Bhat, Gandhinagar 382 428 (India); Makwana, R. [Department of Physics, MS University, Vadodara (India); Chaudhari, V.; Swami, H.L.; Danani, C.; Rao, C.V.S.; Basu, T.K. [Institute for Plasma Research, Bhat, Gandhinagar 382 428 (India); Mandal, D.; Bhade, Sonali; Kolekar, R.V.; Reddy, P.J. [Bhabha Atomic Research Center, Trombay, Mumbai 400 085 (India); Bhattacharyay, R.; Chaudhuri, P. [Institute for Plasma Research, Bhat, Gandhinagar 382 428 (India)


    Highlights: • Breeding benchmark experiment on LLCB TBM in ITER was performed. • Nuclear responses measured are TPR and reaction rate of {sup 115}In(n, n′){sup 115m}In reaction. • Measured responses are compared with calculations by MCNP and FENDL 2.1 library. • TPR measurements agree with calculations in the estimated error bar. • Measured {sup 115}In(n, n′){sup 115m}In reaction rates are underestimated by the calculations. - Abstract: Experiments were conducted with breeding blanket mock-up consisting of two layers of breeder material lithium titanate pebbles and three layers of pure lead as neutron multiplier. The radial dimensions of breeder, neutron multiplier and structural material layers are similar to the current design of the Indian Lead–Lithium cooled Ceramic Breeder (LLCB) blanket. The mock-up assembly was irradiated with 14 MeV neutrons from DT neutron generator. The local tritium production rates (TPR) from {sup 6}Li and {sup 7}Li in breeder layers were measured with the help of two different compositions of Li isotopes (60.69% {sup 6}Li and 7.54% {sup 6}Li) in Li{sub 2}CO{sub 3}. Tritium production in the multiplication layers were also measured with above mentioned two types of pellets to compare the experimental tritium production with calculations. TPR from {sup 6}Li at one location in the breeder layer was also measured by direct online measurement of tritons from {sup 6}Li(n, t){sup 4}He reaction using silicon surface barrier detector and {sup 6}Li to triton converter. Additional verification of neutron spectra (E{sub n} > 0.35 MeV) in the mock-up zones were obtained by measuring {sup 115}In(n, n′){sup 115m}In reaction rate and comparing it with calculated values in all five layers of mock-up. All the measured nuclear responses were compared with transport calculations using code MCNP with FENDL2.1 and FENDL3.0 cross-section libraries. The average C/E ratio for tritium production in enriched Li{sub 2}CO{sub 3} pellets was 1

  13. Diverting activities at Sellafield. [Use of remote technology and computer graphics for mock-up of diversion of raffinate feed pipe

    Energy Technology Data Exchange (ETDEWEB)


    Remote technology is being used to divert a raffinate feed pipe at the Sellafield fuel reprocessing plant in the UK. Extensive use is being made by the contractor of computer graphics and a full-scale mock-up. (author).

  14. ITER校正场线圈试验件VARTM工艺模拟%VARTM Simulation for Mock-up of ITER Correction Coil

    Institute of Scientific and Technical Information of China (English)

    李居平; 吴杰峰; 俞小伍


    The numerical analysis and the experimental investigation for the mock-up of international thermonuclear experimental reactor(ITER) correction coil pieces in the vacuum assisted resin transfer molding(VARTM) process was simulated. The effects of the locations and number of gates on mold-filling process were studied. The results show that the optimized locations of gates are between the two turns. Filling time had exponential decrease with the increase of the number of gates. The locations of vents are based on the flow front of resin, and the mold is designed according to those. The actual filling time (6 600 s) is closed to the calculated time (6 290 s). There are no dry spots in the mock-up. The VARTM simulation is proved useful in strategy design and process optimization for correction coil.%对国际热核实验反应堆(ITER)校正场线圈试验件的真空辅助树脂传递模塑(VARTM)工艺进行了模拟,分析了注射口位置、数量对充模过程的影响。结果表明,最优的注射位置在两匝导体之间;随注射口数量增加,充模时间近似呈指数衰减。对试验件实际浇注过程进行了模拟,通过流动前沿确定了溢料口位置。根据模拟的注射口和溢料口位置设计模具,进行浇注试验。计算充模时间6290 s与实际时间6600 s接近,试验件固化后无干斑缺陷。VARTM工艺模拟可为校正场线圈提供制造方案及工艺优化。

  15. Damage and fatigue crack growth of Eurofer steel first wall mock-up under cyclic heat flux loads. Part 2: Finite element analysis of damage evolution

    Energy Technology Data Exchange (ETDEWEB)

    You, Jeong-Ha, E-mail:


    Highlights: • The surface heat flux load of 3.5 MW/m{sup 2} produced substantial stresses and inelastic strains in the heat-loaded surface region, especially at the notch root. • The notch root exhibited a typical notch effect such as stress concentration and localized inelastic yield leading to a preferred damage development. • The predicted damage evolution feature agrees well with the experimental observation. • The smooth surface also experiences considerable stresses and inelastic strains. However, the stress intensity and the amount of inelastic deformation are not high enough to cause any serious damage. • The level of maximum inelastic strain is higher at the notch root than at the smooth surface. On the other hand, the amplitude of inelastic strain variation is comparable at both positions. • The amount of inelastic deformation is significantly affected by the length of pulse duration time indicating the important role of creep. - Abstract: In the preceding companion article (part 1), the experimental results of the high-heat-flux (3.5 MW/m{sup 2}) fatigue tests of a Eurofer bare steel first wall mock-up was presented. The aim was to investigate the damage evolution and crack initiation feature. The mock-up used there was a simplified model having only basic and generic structural feature of an actively cooled steel FW component for DEMO reactor. In that study, it was found that microscopic damage was formed at the notch root already in the early stage of the fatigue loading. On the contrary, the heat-loaded smooth surface exhibited no damage up to 800 load cycles. In this paper, the high-heat-flux fatigue behavior is investigated with a finite element analysis to provide a theoretical interpretation. The thermal fatigue test was simulated using the coupled damage-viscoplastic constitutive model developed by Aktaa. The stresses, inelastic deformation and damage evolution at the notch groove and at the smooth surface are compared. The different

  16. Experiment of Doppler Heating Starting Point Measurement for Zero Power Physical Test in the Nuclear Power Reactor%核动力反应堆零功率物理试验的多普勒发热点测量试验

    Institute of Scientific and Technical Information of China (English)

    黄礼渊; 付国恩


    In order to determine the upper limit of power,calibrate the range of power and assure testing accura-cy,Doppler heating starting point is measured using digital reactivity meter during zero power physical experi-ment in the reactor.The paper describes the experimental theory,instrument,methods,results and data process-ing methods.The experimental results show that Doppler heating starting point can be determined by modified reactivity values measured by digital reactivity meter.The experience can be used in the following physical ex-periment.%为了定量确定零功率物理试验功率的上限,在反应堆零功率物理试验中,利用数字反应性仪测量多普勒发热点,以确定试验功率的范围、保证试验精度。叙述了本次多普勒发热点测量试验的原理、试验仪器、试验方法、试验结果及数据处理方法等,试验结果表明:利用数字反应性仪测得的反应性经过修正后可以准确地判断多普勒发热点,可为后续物理试验提供参考。

  17. Design and validation of the THMC China-Mock-Up test on buffer material for HLW disposal

    Institute of Scientific and Technical Information of China (English)

    Yuemiao Liu; Like Ma; Dan Ke; Shengfei Cao; Jingli Xie; Xingguang Zhao; Liang Chen; Panpan Zhang


    According to the preliminary concept of the high-level radioactive waste (HLW) repository in China, a large-scale mock-up facility, named China-Mock-Up was constructed in the laboratory of Beijing Research Institute of Uranium Geology (BRIUG). A heater, which simulates a container of radioactive waste, is placed inside the compacted Gaomiaozi (GMZ)-Na-bentonite blocks and pellets. Water inflow through the barrier from its outer surface is used to simulate the intake of groundwater. The numbers of water injection pipes, injection pressure and the insulation layer were determined based on the nu-merical modeling simulations. The current experimental data of the facility are herein analyzed. The experiment is intended to evaluate the thermo-hydro-mechano-chemical (THMC) processes occurring in the compacted bentonite-buffer during the early stage of HLW disposal and to provide a reliable database for numerical modeling and further investigation of engineered barrier system (EBS), and the design of HLW repository.

  18. Design and validation of the THMC China-Mock-Up test on buffer material for HLW disposal

    Directory of Open Access Journals (Sweden)

    Yuemiao Liu


    Full Text Available According to the preliminary concept of the high-level radioactive waste (HLW repository in China, a large-scale mock-up facility, named China-Mock-Up was constructed in the laboratory of Beijing Research Institute of Uranium Geology (BRIUG. A heater, which simulates a container of radioactive waste, is placed inside the compacted Gaomiaozi (GMZ-Na-bentonite blocks and pellets. Water inflow through the barrier from its outer surface is used to simulate the intake of groundwater. The numbers of water injection pipes, injection pressure and the insulation layer were determined based on the numerical modeling simulations. The current experimental data of the facility are herein analyzed. The experiment is intended to evaluate the thermo-hydro-mechano-chemical (THMC processes occurring in the compacted bentonite-buffer during the early stage of HLW disposal and to provide a reliable database for numerical modeling and further investigation of engineered barrier system (EBS, and the design of HLW repository.

  19. Qualification of the numerical simulation of a core disruptive accident on the mars mock-up

    Energy Technology Data Exchange (ETDEWEB)

    Robbe, M.F.; Lepareux, M. [CEA Saclay, Dept. de Mecanique et de Technologie, 91 - Gif-sur-Yvette (France); Cariou, Y. [Novatome, NVPM, 69 - Lyon (France); Treille, E. [Socotec Industrie, 78 - Montigny le Bretonneux (France)


    In case of a Hypothetical Core Disruptive Accident (HCDA) in a Liquid Metal Reactor, the interaction between fuel and liquid sodium creates a high pressure gas bubble in the core. The violent expansion of this bubble loads the vessel and the internal structures, whose deformation is important. A simulation was undertaken using the fluid-structure improvements and the description of the peripheral structures (heat exchangers and pumps) by means of the porosity model. This paper presents the comparison of the results of the third numerical simulation with the experimental results and the numerical results of the previous simulations, as well as a synthesis of all the results of the simulation. (authors)

  20. Thyristor Controlled Reactor for Power Factor Improvement

    Directory of Open Access Journals (Sweden)

    Sheila Mahapatra


    Full Text Available Power factor improvement is the essence of any power sector for reliable operation. This paper provides Thyristor Controlled Reactor regulated by programmed microcontroller which aids in improving power factor and retaining it close to unity under various loading conditions. The implementation is done on 8051 microcontrollerwhich isprogrammed using Keil software. To determine time lag between current and voltage PSpice softwareis used and to display power factor according tothe variation in loadProteus software is used. Whenever a capacitive load is connected to the transmission linea shunt reactor is connected which injects lagging reactive VARs to the power system. As a result the power factor is improved. The results given in this paper provides suitable microcontroller based reactive power compensation and power factor improvement technique using a Thyristor Controlled Reactor module.


    Balent, R.


    This patent describes a method of obtaining a flatter flux and more uniform power generation across the core of a nuclear reactor. The method comprises using moderator elements having differing moderating strength. The elements have an increasing amount of the better moderating material as a function of radial and/or axial distance from the reactor core center. (AEC)

  2. 2D numerical modelling of gas temperature in a nanosecond pulsed longitudinal He-SrBr2 discharge excited in a high temperature gas-discharge tube for the high-power strontium laser (United States)

    Chernogorova, T. P.; Temelkov, K. A.; Koleva, N. K.; Vuchkov, N. K.


    An active volume scaling in bore and length of a Sr atom laser excited in a nanosecond pulse longitudinal He-SrBr2 discharge is carried out. Considering axial symmetry and uniform power input, a 2D model (r, z) is developed by numerical methods for determination of gas temperature in a new large-volume high-temperature discharge tube with additional incompact ZrO2 insulation in the discharge free zone, in order to find out the optimal thermal mode for achievement of maximal output laser parameters. A 2D model (r, z) of gas temperature is developed by numerical methods for axial symmetry and uniform power input. The model determines gas temperature of nanosecond pulsed longitudinal discharge in helium with small additives of strontium and bromine.

  3. Specific power of liquid-metal-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dobranich, D.


    Calculations of the core specific power for conceptual space-based liquid-metal-cooled reactors, based on heat transfer considerations, are presented for three different fuel types: (1) pin-type fuel; (2) cermet fuel; and (3) thermionic fuel. The calculations are based on simple models and are intended to provide preliminary comparative results. The specific power is of interest because it is a measure of the core mass required to produce a given amount of power. Potential problems concerning zero-g critical heat flux and loss-of-coolant accidents are also discussed because these concerns may limit the core specific power. Insufficient experimental data exists to accurately determine the critical heat flux of liquid-metal-cooled reactors in space; however, preliminary calculations indicate that it may be a concern. Results also indicate that the specific power of the pin-type fuels can be increased significantly if the gap between the fuel and the clad is eliminated. Cermet reactors offer the highest specific power because of the excellent thermal conductivity of the core matrix material. However, it may not be possible to take fuel advantage of this characteristic when loss-of-coolant accidents are considered in the final core design. The specific power of the thermionic fuels is dependent mainly on the emitter temperature. The small diameter thermionic fuels have specific powers comparable to those of pin-type fuels. 11 refs., 12 figs, 2 tabs.

  4. Thermionic reactors for space nuclear power (United States)

    Griaznov, Georgii M.; Zhabotinskii, Evgenii E.; Serbin, Victor I.; Zrodnikov, Anatolii V.; Pupko, Victor Ia.; Ponomarev-Stepnoi, Nikolai N.; Usov, V. A.; Nikolaev, Iu. V.

    Compact thermionic nuclear reactor systems with satisfactory mass performance are competitive with space nuclear power systems based on the organic Rankine and closed Brayton cycles. The mass characteristics of the thermionic space nuclear power system are better than that of the solar power system for power levels beyond about 10 kWe. Longlife thermionic fuel element requirements, including their optimal dimensions, and common requirements for the in-core thermionic reactor design are formulated. Thermal and fast in-core thermionic reactors are considered and the ranges of their sensible use are discussed. Some design features of the fast in-core thermionic reactors cores (power range to 1 MWe) including a choice of coolants are discussed. Mass and dimensional performance for thermionic nuclear power reactor system are assessed. It is concluded that thermionic space nuclear power systems are promising power supplies for spacecrafts and that a single basic type of thermionic fuel element may be used for power requirements ranging to several hundred kWe.

  5. Refurbishment of BR2 (Phases 4 and 5)

    Energy Technology Data Exchange (ETDEWEB)

    Gubel, P.; Dekeyser, J.; Van Der Auwera, J


    The BR2 is a materials testing reactor and is SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. In phase 4 of the refurbishment programme, various activities were performed to allow reactor start-up. In phase 5, remaining refurbishment works were carried out as well as the extra studies and upgradings required by the licensing authorities. Major achievements in 1997 are described and discussed.

  6. Topology Zero: Advancing Theory and Experimentation for Power Electronics Education (United States)

    Luchino, Federico

    For decades, power electronics education has been based on the fundamentals of three basic topologies: buck, boost, and buck-boost. This thesis presents the analytical framework for the Topology Zero, a general circuit topology that integrates the basic topologies and provides significant insight into the behaviour of converters. As demonstrated, many topologies are just particular cases of the Topology Zero, an important contribution towards the understanding, integration, and conceptualization of topologies. The investigation includes steady-state, small-signal, and frequency response analysis. The Topology Zero is physically implemented as an educational system. Experimental results are presented to show control applications and power losses analysis using the educational system. The steady-state and dynamic analyses of the Topology Zero provide profuse proof of its suitability as an integrative topology, and of its ability to be indirectly controlled. As well, the implementation of the Topology Zero within an experimentation system is explained and application examples are provided.

  7. Fabrication of full-size mock-up for 10° section of ITER vacuum vessel thermal shield

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Dong Kwon [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Nam, Kwanwoo, E-mail: [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Kang, Kyoung-O; Noh, Chang Hyun; Chung, Wooho [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Lim, Kisuk; Kang, Youngkil [SFA Engineering Corp., Asan-si, Chungcheongnam-do 336-873 (Korea, Republic of); Hamlyn-Harris, Craig; Her, Namil; Robby, Hicks [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France)


    In this paper, a full-scale prototype fabrication for vacuum vessel thermal shield (VVTS) of ITER tokamak is described and test results are reported. All the manufacturing processes except for silver coating were performed in the fabrication of 10° section of VVTS. Pre-qualification test was conducted to compare the vertical and the horizontal welding positions. After shell welding, shell distortion was measured and adjusted. Shell thickness change was also measured after buffing process. Specially, VVTS ports need large bending and complex welding of shell and flange. Bending method for the complex and long cooling tube layout especially for the VVTS ports was developed in detail. Dimensional inspection of the fabricated mock-up was performed with a 3D laser scanner and the scanning data was analyzed.

  8. Introduction to the neutron kinetics of nuclear power reactors

    CERN Document Server

    Tyror, J G; Grant, P J


    An Introduction to the Neutron Kinetics of Nuclear Power Reactors introduces the reader to the neutron kinetics of nuclear power reactors. Topics covered include the neutron physics of reactor kinetics, feedback effects, water-moderated reactors, fast reactors, and methods of plant control. The reactor transients following faults are also discussed, along with the use of computers in the study of power reactor kinetics. This book is comprised of eight chapters and begins with an overview of the reactor physics characteristics of a nuclear power reactor and their influence on system design and

  9. Compact reactor/ORC power source

    Energy Technology Data Exchange (ETDEWEB)

    Meier, K.L.; Kirchner, W.L.; Willcutt, G.J.


    A compact power source that combines an organic Rankine Cycle (ORC) electric generator with a nuclear reactor heat source is being designed and fabricated. Incorporating existing ORC technology with proven reactor technology, the compact reactor/ORC power source offers high reliability while minimizing the need for component development. Thermal power at 125 kWt is removed from the coated particle fueled, graphite moderated reactor by heat pipes operating at 500/sup 0/C. Outside the reactor vessel and connected to the heat pipes are vaporizers in which the toluene ORC working fluid is heated to 370/sup 0/C. In the turbine-alternator-pump (TAP) combined-rotating unit, the thermal energy of the toluene is converted to 25 kWe of electric power. Lumped parameter systems analyses combined with a finite element thermal analysis have aided in the power source design. The analyses have provided assurance of reliable multiyear normal operation as well as full power operation with upset conditions, such as failed heat pipes and inoperative ORC vaporizers. Because of inherent high reliability, long life, and insensitivity to upset conditions, this power source is especially suited for use in remote, inaccessible locations where fuel delivery and maintenance costs are high. 10 refs.

  10. Incorporation of evaporator concentrates mock-up waste in Brazilian bitumen; Incorporacao de rejeitos simulados de concentrado de evaporador em betume nacional

    Energy Technology Data Exchange (ETDEWEB)

    Guzella, Marcia Flavia Righi; Silva, Tania Valeria da [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), Belo Horizonte, MG (Brazil)]. E-mail:;


    The solidification of radioactive liquid waste generated in nuclear power plants is required by the safety standards for the transportation, storage and deposition. CDTN/CNEN has carried out studies and experimental research aiming at the solidification and stabilization of radioactive wastes on different matrixes, such as cement and bitumen, therefore contributing to the improvement of treatment processes of low and intermediate radioactive waste from NPPs in Brazil. Experiments with solidification of waste in national bitumen, using NPP Angra 2 mock-up equipment, were carried out at a pilot-scale at CDTN. The evaporator concentrates were simulated by boric acid solution, salts and corrosion products. Caesium chloride was added to the solution for comparison with former experiments that evaluated the influence of boron and sodium in the matrix resistance to leaching. This word presents the results of the waste form characterization, obtained according to the softening point, flash point, resistance to leaching and penetration tests, plus thermodifferencial analysis. The experiments were performed according to either ABNT or ISO standards. The correct characterization of the waste form is important not only for a safe disposal but also to obtain the necessary safety level during the operation of the radioactive waste bituminization system. (author)

  11. Heat pipe reactors for space power applications (United States)

    Koenig, D. R.; Ranken, W. A.; Salmi, E. W.


    A family of heat pipe reactors design concepts has been developed to provide heat to a variety of electrical conversion systems. Three power plants are described that span the power range 1-500 kWe and operate in the temperature range 1200-1700 K. The reactors are fast, compact, heat-pipe cooled, high-temperature nuclear reactors fueled with fully enriched refractory fuels, UC-ZrC or UO2. Each fuel element is cooled by an axially located molybdenum heat pipe containing either sodium or lithium vapor. Virtues of the reactor designs are the avoidance of single-point failure mechanisms, the relatively high operating temperature, and the expected long lifetimes of the fuel element components.

  12. HHF test with 80x80x1 Be/Cu/SS Mock-ups for verifying the joining technology of the ITER blanket First Wall

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Won; Bae, Young Dug; Kim, Suk Kwon; Hong, Bong Guen; Jeong, Yong Hwan; Park, Jeong Yong; Choi, Byung Kwon; Jung, Hyun Kyu


    Through the fabrication of the Cu/SS and Be/Cu joint specimens, fabrication procedure such as material preparation, canning, degassing, HIP (Hot Isostatic Pressing), PHHT (Post HIP heat treatment) was established. The HIP conditions (1050 .deg. C, 100 MPa 2 hr for Cu/SS, 580 .deg. C 100 MPa 2 hr for Be/Cu) were developed through the investigation on joint specimen fabricated with the various HIP conditions; the destructive tests of joint include the microstructure observation of the interface with the examination of the elemental distribution, tension test, bend test, Charpy impact test and fracture toughness test. However, since the joint should be tested under the High Heat Flux (HHF) conditions like the ITER operation for verifying its joint integrity, several HHF tests were performed like the previous HHF test with the Cu/SS, Be/Cu, Be/Cu/SS Mock-ups. In the present study, the HHF test with Be/Cu/SS Mock-ups, which have 80 mm x 80 mm single Be tile and each material depths were kept to be the same as the ITER blanket FW. The Mock-ups fabricated with three kinds of interlayers such as Cr/Ti/Cu, Ti/Cr/Cu, Ti/Cu, which were different from the developed interlayer (Cr/Cu), total 6 Mock-ups were fabricated. Preliminary analysis were performed to decide the test conditions; they were tested with up to 2.5 MW/m2 of heat fluxes and 20 cycles for each Mock-up in a given heat flux. They were tested with JUDITH-1 at FZJ in Germany. During tests, all Mock-ups showed delamination or full detachment of Be tile and it can be concluded that the joints with these interlayers have a bad joining but it can be used as a good data for developing the Be/Cu joint with HIP.

  13. Topology zero: advancing theory and experimentation for power electronics education


    Luchino, Federico


    For decades, power electronics education has been based on the fundamentals of three basic topologies: buck, boost, and buck-boost. This thesis presents the analytical framework for the Topology Zero, a general circuit topology that integrates the basic topologies and provides significant insight into the behaviour of converters. As demonstrated, many topologies are just particular cases of the Topology Zero, an important contribution towards the understanding, integration, and conceptualizat...

  14. Initial RattleSnake Calculations of the Hot Zero Power BEAVRS

    Energy Technology Data Exchange (ETDEWEB)

    M. Ellis; J. Ortensi; Y. Wang; K. Smith; R.C. Martineau


    The validation of the Idaho National Laboratory's next generation of reactor physics analysis codes is an essential and ongoing task. The validation process requires a large undertaking and includes detailed, realistic models that can accurately predict the behavior of an operational nuclear reactor. Over the past few years the INL has developed the RattleSnake application and supporting tools on the MOOSE framework to perform these reactor physics calculations. RattleSnake solves the linearized Boltzmann transport equation with a variety of solution meth­ ods. Various traditional reactor physics benchmarks have already been performed, but a more realistic light water reactor comparison was needed to solidify the status of the code and deter­ mine its fidelity. The INL team decided to use the Benchmark for Evaluation and Validation of Reactor Simulations, which was made available in early 2013. This benchmark is a one­ of-a-kind document assembled by the Massachusetts Institute of Technology, which includes two cycles of detailed, measured PWR operational data. The results from this initial study of the hot zero power conditions show the current INL analysis procedure with DRAGON4 cross section preparation and using the low order diffusion solver in RattleSnake for the whole core calculations yield very encouraging results for PWR analysis. The radial assembly power distributions, radial detector measurements and control rod worths were computed with good accuracy. The computation of the isothermal temperature coefficients of reactivity require further study.

  15. Liquid Metal Cooled Reactor for Space Power (United States)

    Weitzberg, Abraham


    The conceptual design is for a liquid metal (LM) cooled nuclear reactor that would provide heat to a closed Brayton cycle (CBC) power conversion subsystem to provide electricity for electric propulsion thrusters and spacecraft power. The baseline power level is 100 kWe to the user. For long term power generation, UN pin fuel with Nb1Zr alloy cladding was selected. As part of the SP-100 Program this fuel demonstrated lifetime with greater than six atom percent burnup, at temperatures in the range of 1400-1500 K. The CBC subsystem was selected because of the performance and lifetime database from commercial and aircraft applications and from prior NASA and DOE space programs. The high efficiency of the CBC also allows the reactor to operate at relatively low power levels over its 15-year life, minimizing the long-term power density and temperature of the fuel. The scope of this paper is limited to only the nuclear components that provide heated helium-xenon gas to the CBC subsystem. The principal challenge for the LM reactor concept was to design the reactor core, shield and primary heat transport subsystems to meet mission requirements in a low mass configuration. The LM concept design approach was to assemble components from prior programs and, with minimum change, determine if the system met the objective of the study. All of the components are based on technologies having substantial data bases. Nuclear, thermalhydraulic, stress, and shielding analyses were performed using available computer codes. Neutronics issues included maintaining adequate operating and shutdown reactivities, even under accident conditions. Thermalhydraulic and stress analyses calculated fuel and material temperatures, coolant flows and temperatures, and thermal stresses in the fuel pins, components and structures. Using conservative design assumptions and practices, consistent with the detailed design work performed during the SP-100 Program, the mass of the reactor, shield, primary heat

  16. Magnet powering with zero downtime - a dream?

    CERN Document Server

    Zerlauth, Markus


    Despite a number of improvements already applied in the course of the year, the magnet powering system of the LHC still accounts for around 50% of the premature beam dumps. This number might even further increase when moving to higher beam energies in the next years. With mitigations of radiation effects and the prospects for beam induced magnet quenches being discussed elsewhere, we aim at identifying possible mid- and long-term improvements within the various equipment systems to further reduce the number of equipment failures leading to a loss of the particle beams. Amongst others, this includes the sensitivity of equipment to external causes such as electromagnetic perturbations or perturbations on the electrical network. To conclude, the gain of the identified mitigations will have to be balanced against the potential impact on schedule and cost.

  17. Early Impacts of a Human-in-the-Loop Evaluation in a Space Vehicle Mock-up Facility (United States)

    Byrne, Vicky; Vos, Gordon; Whitmore, Mihriban


    The development of a new space vehicle, the Orion Crew Exploration Vehicle (CEV), provides Human Factors engineers an excellent opportunity to have an impact early in the design process. This case study highlights a Human-in-the-Loop (HITL) evaluation conducted in a Space Vehicle Mock-Up Facility and will describe the human-centered approach and how the findings are impacting design and operational concepts early in space vehicle design. The focus of this HITL evaluation centered on the activities that astronaut crewmembers would be expected to perform within the functional internal volume of the Crew Module (CM) of the space vehicle. The primary objective was to determine if there are aspects of a baseline vehicle configuration that would limit or prevent the performance of dynamically volume-driving activities (e.g. six crewmembers donning their suits in an evacuation scenario). A second objective was to step through concepts of operations for known systems and evaluate them in integrated scenarios. The functional volume for crewmember activities is closely tied to every aspect of system design (e.g. avionics, safety, stowage, seats, suits, and structural support placement). As this evaluation took place before the Preliminary Design Review of the space vehicle with some designs very early in the development, it was not meant to determine definitely that the crewmembers could complete every activity, but rather to provide inputs that could improve developing designs and concepts of operations definition refinement.

  18. Assessment of the thorium fuel cycle in power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Homan, F.J.; Allen, E.J.


    A study was conducted at Oak Ridge National Laboratory to evaluate the role of thorium fuel cycles in power reactors. Three thermal reactor systems were considered: Light Water Reactors (LWRs); High-Temperature Gas-Cooled Reactors (HTGRs); and Heavy Water Reactors (HWRs) of the Canadian Deuterium Uranium Reactor (CANDU) type; most of the effort was on these systems. A summary comparing thorium and uranium fuel cycles in Fast Breeder Reactors (FBRs) was also compiled.

  19. Manufacturing and testing of W/Cu mono-block small scale mock-up for EAST by HIP and HRP technologies

    Energy Technology Data Exchange (ETDEWEB)

    Li, Qiang [Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP), Hefei, Anhui (China); Qin, Sigui [Advanced Technology and Materials Co., Ltd, Beijing (China); Wang, Wanjing; Qi, Pan [Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP), Hefei, Anhui (China); Roccella, Selanna; Visca, Eliseo [Associazione EURATOM-ENEA sulla Fusione, Frascati (Italy); Liu, Guohui [Advanced Technology and Materials Co., Ltd, Beijing (China); Luo, Guang-Nan, E-mail: [Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP), Hefei, Anhui (China)


    ITER-like W/Cu mono-block plasma-facing components (PFCs) will be used in vertical target regions of the experimental advanced superconducting tokamak (EAST) divertor. The first W/Cu mono-block small scale mock-up with five W mono-blocks has been manufactured successfully by technological combination of hot isostatic pressing (HIP) and hot radial pressing (HRP). The joining of a W mono-block and a pure copper interlayer was achieved by means of HIP technology and the bonding strength was over 150 MPa. The good bonding between the pure copper interlayer and a CuCrZr cooling tube was obtained by means of HRP technology. In order to understand deeply the process of HRP, the stress distribution of the mock-up during HRP process was simulated using ANSYS code. Ultrasonic Nondestructive Testing (NDT) of the W/Cu and Cu/CuCrZr interfaces was performed, showing that excellent bonding of the W/Cu and Cu/CuCrZr interfaces. The thermal cycle fatigue testing of the mock-up has been carried out by means of an e-beam device in Southwest Institute of Physics, Chengdu (SWIP) and the mock-up withstood 1000 cycles of heat loads up to 8.4 MW/m{sup 2} with the cooling water of 2 m/s, 20 °C, 0.2 MPa.

  20. Transients in reactors for power systems compensation (United States)

    Abdul Hamid, Haziah

    This thesis describes new models and investigations into switching transient phenomena related to the shunt reactors and the Mechanically Switched Capacitor with Damping Network (MSCDN) operations used for reactive power control in the transmission system. Shunt reactors and MSCDN are similar in that they have reactors. A shunt reactor is connected parallel to the compensated lines to absorb the leading current, whereas the MSCDN is a version of a capacitor bank designed as a C-type filter for use in the harmonic-rich environment. In this work, models have been developed and transient overvoltages due to shunt reactor deenergisation were estimated analytically using MathCad, a mathematical program. Computer simulations used the ATP/EMTP program to reproduce both single-phase and three-phase shunt reactor switching at 275 kV operational substations. The effect of the reactor switching on the circuit breaker grading capacitor was also examined by considering various switching conditions.. The main original achievement of this thesis is the clarification of failure mechanisms occurring in the air-core filter reactor due to MSCDN switching operations. The simulation of the MSCDN energisation was conducted using the ATP/EMTP program in the presence of surge arresters. The outcome of this simulation shows that extremely fast transients were established across the air-core filter reactor. This identified transient event has led to the development of a detailed air-core reactor model, which accounts for the inter-turn RLC parameters as well as the stray capacitances-to-ground. These parameters are incorporated into the transient simulation circuit, from which the current and voltage distribution across the winding were derived using electric field and equivalent circuit modelling. Analysis of the results has revealed that there are substantial dielectric stresses imposed on the winding insulation that can be attributed to a combination of three factors. (i) First, the

  1. A built-in zero valent iron anaerobic reactor to enhance treatment of azo dye wastewater. (United States)

    Zhang, Yaobin; Jing, Yanwen; Quan, Xie; Liu, Yiwen; Onu, Pascal


    Waste scrap iron was packed into an upflow anaerobic sludge blanket (UASB) reactor to form a zero valent iron (ZVI) - UASB reactor system for treatment of azo dye wastewater. The ZVI acted as a reductant to decrease ORP in the reactor by more than 40 mv and functioned as an acid buffer to increase the pH in the reactor from 5.44 to 6.29, both of which improved the performance of the anaerobic reactor. As a result, the removal of color and COD in this reactor was 91.7% and 53%, respectively, which was significantly higher than that of a reference UASB reactor without ZVI. The UV-visible spectrum demonstrated that absorption bands of the azo dye from the ZVI-UASB reactor were substantially reduced. The ZVI promoted methanogenesis, which was confirmed by an increase in CH(4) content in the biogas from 47.9% to 64.8%. The ZVI bed was protected well from rusting, which allowed it to function stably. The effluent could be further purified only by pH adjustment because the Fe(2+) released from ZVI served as a flocculent.


    Institute of Scientific and Technical Information of China (English)

    姜惠强; 林芳华


    Here the authors are interested in the zero set of Sobolev functions and functions of bounded variation with negative power of integrability. The main result is a general Hausdorff dimension estimate on the size of zero set. The research is motivated by the model on van der waal force driven thin film, which is a singular elliptic equation.After obtaining some basic regularity result, the authors get an estimate on the size of singular set; such set corresponds to the thin film rupture set in the thin film model.

  3. 78 FR 73898 - Operator Licensing Examination Standards for Power Reactors (United States)


    ... COMMISSION Operator Licensing Examination Standards for Power Reactors AGENCY: Nuclear Regulatory Commission... available for public comment a draft NUREG, NUREG-1021, Revision 10, ``Operator Licensing Examination Standards for Power Reactors.'' DATES: Submit comments by February 7, 2014. Comments received after...

  4. Investigation on water content in fresco mock-ups in the microwave and near-IR spectral regions (United States)

    Magrini, Donata; Cucci, Costanza; Olmi, Roberto; Picollo, Marcello; Riminesi, Cristiano


    Water diffusion inside masonry is responsible for the majority of the decay phenomena observed in wall paintings and frescos. Thus, the diagnostics of moisture and water content and their monitoring represent a key issue. In order to preserve the integrity of surfaces of artistic interest, investigations by means of non-destructive techniques (NDT) are preferred over others. The aim of this research is to determine methodologies to quantify the moisture content (MC) of frescos by means of the integrated use of two non-invasive techniques, namely fiber optic reflectance spectroscopy (FORS) in the near-IR region and evanescent field dielectrometry (EFD) in the microwave range. The FORS technique has been employed in order to assess the amount of water adsorbed from the surface by means of an analysis of the reflectance spectra in the Vis-NIR (350-2200 nm) range. This technique investigates the electronic and vibrational transitions that are characteristic of each compound and enables their identification. The water content is evaluated on the basis of the 1920 nm and 1450 nm absorption bands. The EFD system consists of a resonant probe connected to a network analyzer. The resonance frequency of the cavity under different moisture-content conditions of frescos is in the 1.0-1.5 GHz range. The device makes it possible to compute, in real time, the MC from a measurement of the transmission coefficient (amplitude versus frequency) through the probe. Fresco mock-ups have been prepared in collaboration with the Opificio delle Pietre Dure in order to recreate most of the possible chromatic shades obtained by mixing iron oxides and hydroxide-based pigments. Measurements were performed by employing both techniques on fresco models after wet-dry cycles obtained by means of poultices with a known water content. The results obtained with these two techniques were compared, and cross relationships between the EFD and FORS data were defined.

  5. Thermal testing of the first mirror unit mock-up for H-alpha and visible spectroscopy in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Orlovskiy, Ilya, E-mail: [National Research Center “Kurchatov Institute”, Moscow (Russian Federation); Alekseev, Andrey; Andreenko, Evgeniy; Vukolov, Konstantin [National Research Center “Kurchatov Institute”, Moscow (Russian Federation); Denisov, Vladimir; Klyatskin, Andrey [OAO Kompozit Company, Korolev (Russian Federation); Lukin, Anatoliy; Melnikov, Andrey; Muslimov, Eduard [NPO GIPO Company, Kazan (Russian Federation)


    Highlights: • Sandwich-like design was proposed for single-crystal molybdenum mirror. • High optical quality was provided for aspheric molybdenum mirror by diamond turning. • First mirror unit made of molybdenum was thermally tested in vacuum chamber. • Observations during the tests showed no degradation of image quality. - Abstract: H-alpha and visible spectroscopy (HA&VS) diagnostic in ITER is aimed at the measurements of the intensity of hydrogen isotopes and of impurity lines in the SOL emission. It is based on the endoscope scheme with the first mirror unit (FMU) containing two mirrors (flat and aspheric) located behind a small entrance pupil and a discharge cleaning system. A primary candidate material for the FMU mirror is a single-crystal molybdenum (SC-Mo). A sandwich-like design is proposed, in which few small-sized SC-Mo plates are bound by a diffusive welding to a large-size polycrystalline Mo substrate. An acceptable welding regime is presented. A recent progress is discussed on the polishing of SC-Mo focusing mirrors up to 150 mm diameter. Thermal tests of the molybdenum FMU mock-up had been performed. The FMU was heated up to 200 °C whereas the mirror temperature was up to 300 °C under the pressure of 8–9 Pa. A technique for the quantitative assessment of the image quality (contrast and resolution) at the elevated temperatures has been suggested. As a result, a sub-mm shift of an image plane along the optical axis had been observed with no significant degradation of the image quality.

  6. HHF Test with FMS/FMS Mock-ups for Verifying the Joining Technology of the Liquid TBM First Wall in the ITER

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Won; Bae, Young Dug; Kim, Suk Kwon; Song, Woo Sub; Chang, Doo Hee; Kim, Tae Sung; Kim, Bum Ryul; Hong, Bong Guen; Park, Jung Yong; Choi, Byung Kwon; Jeong, Yong Hwan


    Through the fabrication of the FMS joint specimens, fabrication procedure such as material preparation, canning, degassing, HIP (Hot Isostatic Pressing), PHHT (Post HIP heat treatment) was established. The HIP conditions (1050 .deg. C, 100 MPa 2 hr) were developed through the investigation on joint specimen fabricated with the various HIP conditions; the destructive tests of joint include the microstructure observation of the interface with the examination of the elemental distribution, tension test, bend test, Charpy impact test and fracture toughness test. However, since the joint should be tested under the High Heat Flux (HHF) conditions like the ITER operation for verifying its joint integrity, HHF tests were performed like the previous HHF test with the Cu/SS, Be/Cu, Be/Cu/SS mock-ups for the ITER blanket First Wall (FW). In the present study, the HHF test with FMS/FMS mock-ups, which have 26 mm x 30 mm x 50 mm of dimensions with 10 mm x 20 mm of cooling channel, were kept to be the same as the designed FW of the KO HCML TBM. Three mock-ups were fabricated with one, two, and three channels and two channels mock-up among them was used for the microstructure observation after cutting itself. Others were used for the HHF test under 0.5 to 1.5 MW/m{sup 2} of heat fluxes and 20 cycles for each mock-up in a given heat flux. The test conditions were determined through the ANSYS-CFX analysis with the designed KO TBM parameters. They were tested with NBI test stand at KAERI, which were developed and operated for the KSTAR heating. Both mock-ups were survived below 1.5 MW/m{sup 2} of heat flux with 20 cycles but they were broken at 1.5 MW/m{sup 2} of heat flux due to the CHF in the wall, which is around 1.46 MW/m{sup 2} of heat flux in this case. Numbers of the cycle to failure were induced through mechanical analysis with ANSYS code but the expected values were much higher than those of tested cycles.

  7. Modular stellarator reactor: a fusion power plant

    Energy Technology Data Exchange (ETDEWEB)

    Miller, R.L.; Bathke, C.G.; Krakowski, R.A.; Heck, F.M.; Green, L.; Karbowski, J.S.; Murphy, J.H.; Tupper, R.B.; DeLuca, R.A.; Moazed, A.


    A comparative analysis of the modular stellarator and the torsatron concepts is made based upon a steady-state ignited, DT-fueled, reactor embodiment of each concept for use as a central electric-power station. Parametric tradeoff calculations lead to the selection of four design points for an approx. 4-GWt plant based upon Alcator transport scaling in l = 2 systems of moderate aspect ratio. The four design points represent high-aspect ratio. The four design points represent high-(0.08) and low-(0.04) beta versions of the modular stellarator and torsatron concepts. The physics basis of each design point is described together with supporting engineering and economic analyses. The primary intent of this study is the elucidation of key physics and engineering tradeoffs, constraints, and uncertainties with respect to the ultimate power reactor embodiment.

  8. Power reactor noise studies and applications

    Energy Technology Data Exchange (ETDEWEB)

    Arzhanov, V


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

  9. Supplemental Thermal-Hydraulic Transient Analyses of BR2 in Support of Conversion to LEU Fuel

    Energy Technology Data Exchange (ETDEWEB)

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


    Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The RELAP5/Mod 3.3 code has been used to perform transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. A RELAP5 model of BR2 has been validated against select transient BR2 reactor experiments performed in 1963 by showing agreement with measured cladding temperatures. Following the validation, the RELAP5 model was then updated to represent the current use of the reactor; taking into account core configuration, neutronic parameters, trip settings, component changes, etc. Simulations of the 1963 experiments were repeated with this updated model to re-evaluate the boiling risks associated with the currently allowed maximum heat flux limit of 470 W/cm2 and temporary heat flux limit of 600 W/cm2. This document provides analysis of additional transient simulations that are required as part of a modern BR2 safety analysis report (SAR). The additional simulations included in this report are effect of pool temperature, reduced steady-state flow rate, in-pool loss of coolant accidents, and loss of external cooling. The simulations described in this document have been performed for both an HEU- and LEU-fueled core.

  10. Tool coupling for the design and operation of building energy and control systems based on the Functional Mock-up Interface standard

    Energy Technology Data Exchange (ETDEWEB)

    Nouidui, Thierry Stephane; Wetter, Michael


    This paper describes software tools developed at the Lawrence Berkeley National Laboratory (LBNL) that can be coupled through the Functional Mock-up Interface standard in support of the design and operation of building energy and control systems. These tools have been developed to address the gaps and limitations encountered in legacy simulation tools. These tools were originally designed for the analysis of individual domains of buildings, and have been difficult to integrate with other tools for runtime data exchange. The coupling has been realized by use of the Functional Mock-up Interface for co-simulation, which standardizes an application programming interface for simulator interoperability that has been adopted in a variety of industrial domains. As a variety of coupling scenarios are possible, this paper provides users with guidance on what coupling may be best suited for their application. Furthermore, the paper illustrates how tools can be integrated into a building management system to support the operation of buildings. These tools may be a design model that is used for real-time performance monitoring, a fault detection and diagnostics algorithm, or a control sequence, each of which may be exported as a Functional Mock-up Unit and made available in a building management system as an input/output block. We anticipate that this capability can contribute to bridging the observed performance gap between design and operational energy use of buildings.

  11. A SCALE-UP Mock-Up: Comparison of Student Learning Gains in High- and Low-Tech Active-Learning Environments (United States)

    Soneral, Paula A. G.; Wyse, Sara A.


    Student-centered learning environments with upside-down pedagogies (SCALE-UP) are widely implemented at institutions across the country, and learning gains from these classrooms have been well documented. This study investigates the specific design feature(s) of the SCALE-UP classroom most conducive to teaching and learning. Using pilot survey data from instructors and students to prioritize the most salient SCALE-UP classroom features, we created a low-tech “Mock-up” version of this classroom and tested the impact of these features on student learning, attitudes, and satisfaction using a quasi-­experimental setup. The same instructor taught two sections of an introductory biology course in the SCALE-UP and Mock-up rooms. Although students in both sections were equivalent in terms of gender, grade point average, incoming ACT, and drop/fail/withdraw rate, the Mock-up classroom enrolled significantly more freshmen. Controlling for class standing, multiple regression modeling revealed no significant differences in exam, in-class, preclass, and Introduction to Molecular and Cellular Biology Concept Inventory scores between the SCALE-UP and Mock-up classrooms. Thematic analysis of student comments highlighted that collaboration and whiteboards enhanced the learning experience, but technology was not important. Student satisfaction and attitudes were comparable. These results suggest that the benefits of a SCALE-UP experience can be achieved at lower cost without technology features. PMID:28213582

  12. U.S. Nuclear Power Reactor Plant Status (United States)

    Nuclear Regulatory Commission — Demographic data on U.S. commercial nuclear power reactors, including: plant name/unit number, docket number, location, licensee, reactor/containment type, nuclear...

  13. Estudos etnoictiológicos sobre o pirarucu Arapaima gigas na Amazônia Central Ethnoictiology studies on Pirarucu (Arapaima mock-ups in Central Amazon

    Directory of Open Access Journals (Sweden)

    Liane Galvão de Lima


    Full Text Available O presente estudo visou identificar saberes comuns entre o conhecimento científico e o conhecimento local sobre a ecologia e biologia do pirarucu (Arapaima gigas, contribuindo com informações úteis para a implementação e consolidação de projetos de manejo participativo pesqueiro na região. Foram realizadas 57 entrevistas semi-estruturadas, com pescadores profissionais de Manaus e pescadores de subsistência de Manacapuru durante o período de junho a dezembro do ano de 2002. Foi observado que os pescadores profissionais possuem informações igualmente precisas e abrangentes em relação aos saberes dos pescadores ribeirinhos de subsistência nos aspectos de reprodução, predação, migração, crescimento e mortalidade. Os aspectos que não são equivalentes entre os pescadores profissionais comerciais citadinos e ribeirinhos de subsistência são nos aspectos de tipo de alimentação e no tamanho de recrutamento pesqueiro. Concluímos que os pescadores da Amazônia central possuem os conhecimentos necessários que possibilitam o manejo participativo do pirarucu, como um profundo saber nos aspectos comportamentais, biológicos e ecológicos desta espécie, podendo assim contribuir de fato com a participação de gestão nos recursos pesqueiros locais.Present study it aimed at to identify to know common between scientific knowledge and local knowledge on ecology and biology of pirarucu (Arapaima mock-ups, contributing with useful information for implementation and consolidation of projects of participative handling fishing boat in region. 57 half-structuralized interviews had been carried through, with fishing of Manaus and Manacapuru during period of June to December of year 2002. It was observed that professional fishermen also have accurate and comprehensive information in relation to knowledge of subsistence fishermen in coastal aspects of reproduction, predation, migration, growth and mortality. Aspects that are not equivalent

  14. Multisphere measurements in power reactor containments

    Energy Technology Data Exchange (ETDEWEB)

    Scherpelz, R.I.; Tanner, J.E.; Soldat, K.L.; Endres, G.W.R.; Brackenbush, L.W. (Battelle-Pacific Northwest Lab., Richland, WA (United States))


    For more than 12 years, Battelle, Pacific Northwest Laboratories (PNL) has performed neutron measurements in commercial nuclear power plants to determine the response of the plant's personnel dosimeters and survey meters. The multisphere spectrometer system has been used in many of these studies because it covers a wide range of neutron energies and is widely used in the nuclear industry. The system used by PNL employs a 1.27- x 1.27-cm LiI crystal in seven moderated configurations with moderating sphere sizes ranging up to 30.5 cm in diameter. The collected count rates are unfolded using the computer code SPUNIT and a modification of the Sanna response functions. Multisphere measurements have been performed in eight different light water reactors (LWRs), including both boiling water reactor (BWRs) and pressurized water reactors (PWRs). Unfolded energy distributions are typically low with average energies often <100 keV. Spectra in BWRs frequently show higher average energies than those in PWRs. The most useful application of multisphere measurements in LWR containments is for measuring energy distributions for comparisons to other multisphere measurements.

  15. Evaluation of residual stresses in electron-beam welded Zr2.5Nb0.9Hf Zircadyne flange mock-up of a reflector vessel beam tube flange

    Energy Technology Data Exchange (ETDEWEB)

    Muránsky, O., E-mail: [Institute of Material Engineering, ANSTO, Locked Bag 2001, Kirrawee DC, 2234 NSW (Australia); Holden, T.M. [Northern Stress Technologies, Deep River, Ontario, Canada K0J 1P0 (Canada); Kirstein, O. [European Spallation Source, EES AB, Tunavagen 24, SE-211 00 Lund (Sweden); James, J.A. [Open University, Materials Engineering, Milton Keynes MK7 6BJ (United Kingdom); Paradowska, A.M. [Bragg Institute, ANSTO, Locked Bag 2001, Kirrawee DC, 2234 NSW (Australia); Edwards, L. [Institute of Material Engineering, ANSTO, Locked Bag 2001, Kirrawee DC, 2234 NSW (Australia)


    The dual-phase alloy Zr2.5Nb alloy is an important nuclear material, because of its use in current and possible use in future nuclear reactors. It is, however, well-known that Zr2.5Nb weldments can fail through a time-dependent mechanism called delayed hydride cracking which is typically driven by the presence of tensile residual stresses. With a view to understanding the development of residual stresses associated with Zr2.5Nb welds the current study focuses on the evaluation of the residual stresses in a mock-up of a reactor beam tube flange made from Zr2.5Nb0.9Hf. The present results suggests that, like ferritic welds which undergo a solid-state phase transformation upon welding, Zr2.5Nb0.9Hf welds also develop high tensile residual stresses in the heat-affected zone whereas the stresses closer to the weld tip are reduced by the effects of the β → α solid-state phase transformation.

  16. Zero loss magnetic metamaterials using powered active unit cells. (United States)

    Yuan, Yu; Popa, Bogdan-Ioan; Cummer, Steven A


    We report the design and experimental measurement of a powered active magnetic metamaterial with tunable permeability. The unit cell is based on the combination of an embedded radiofrequency amplifier and a tunable phase shifter, which together control the response of the medium. The measurements show that a negative permeability metamaterial with zero loss or even gain can be achieved through an array of such metamaterial cells. This kind of active metamaterial can find use in applications that are performance limited due to material losses.

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

    Energy Technology Data Exchange (ETDEWEB)

    Cammi, Antonio, E-mail: [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)


    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.

  18. Reactor Physics Programme

    Energy Technology Data Exchange (ETDEWEB)

    De Raedt, C


    The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies.

  19. Neutron measurements at nuclear power reactors [55

    CERN Document Server

    Scherpelz, R I


    Staff from the Pacific Northwest National Laboratory (operated by Battelle Memorial Institute), have performed neutron measurements at a number of commercial nuclear power plants in the United States. Neutron radiation fields at light water reactor (LWR) power plants are typically characterized by low-energy distributions due to the presence of large amounts of scattering material such as water and concrete. These low-energy distributions make it difficult to accurately monitor personnel exposures, since most survey meters and dosimeters are calibrated to higher-energy fields such as those produced by bare or D sub 2 O-moderated sup 2 sup 5 sup 2 Cf sources. Commercial plants typically use thermoluminescent dosimeters in an albedo configuration for personnel dosimetry and survey meters based on a thermal-neutron detector inside a cylindrical or spherical moderator for dose rate assessment, so their methods of routine monitoring are highly dependent on the energy of the neutron fields. Battelle has participate...

  20. Zeta function zeros, powers of primes, and quantum chaos. (United States)

    Sakhr, Jamal; Bhaduri, Rajat K; van Zyl, Brandon P


    We present a numerical study of Riemann's formula for the oscillating part of the density of the primes and their integer powers. The formula consists of an infinite series of oscillatory terms, one for each zero of the zeta function on the critical line, and was derived by Riemann in his paper on primes, assuming the Riemann hypothesis. We show that high-resolution spectral lines can be generated by the truncated series at all integer powers of primes and demonstrate explicitly that the relative line intensities are correct. We then derive a Gaussian sum rule for Riemann's formula. This is used to analyze the numerical convergence of the truncated series. The connections to quantum chaos and semiclassical physics are discussed.

  1. Zero valent iron simultaneously enhances methane production and sulfate reduction in anaerobic granular sludge reactors. (United States)

    Liu, Yiwen; Zhang, Yaobin; Ni, Bing-Jie


    Zero valent iron (ZVI) packed anaerobic granular sludge reactors have been developed for improved anaerobic wastewater treatment. In this work, a mathematical model is developed to describe the enhanced methane production and sulfate reduction in anaerobic granular sludge reactors with the addition of ZVI. The model is successfully calibrated and validated using long-term experimental data sets from two independent ZVI-enhanced anaerobic granular sludge reactors with different operational conditions. The model satisfactorily describes the chemical oxygen demand (COD) removal, sulfate reduction and methane production data from both systems. Results show ZVI directly promotes propionate degradation and methanogenesis to enhance methane production. Simultaneously, ZVI alleviates the inhibition of un-dissociated H2S on acetogens, methanogens and sulfate reducing bacteria (SRB) through buffering pH (Fe(0) + 2H(+) = Fe(2+) + H2) and iron sulfide precipitation, which improve the sulfate reduction capacity, especially under deterioration conditions. In addition, the enhancement of ZVI on methane production and sulfate reduction occurs mainly at relatively low COD/ [Formula: see text] ratio (e.g., 2-4.5) rather than high COD/ [Formula: see text] ratio (e.g., 16.7) compared to the reactor without ZVI addition. The model proposed in this work is expected to provide support for further development of a more efficient ZVI-based anaerobic granular system.

  2. An overview of future sustainable nuclear power reactors


    Andreas Poullikkas


    In this paper an overview of the current and future nuclear power reactor technologies is carried out. In particular, the nuclear technology is described and the classification of the current and future nuclear reactors according to their generation is provided. The analysis has shown that generation II reactors currently in operation all around the world lack significantly in safety precautions and are prone to loss of coolant accident (LOCA). In contrast, generation III reactors, which are ...

  3. Analysis of gamma-ray dosimetry experiments in the zero power MINERVE facility

    Energy Technology Data Exchange (ETDEWEB)

    Amharrak, H.; Di Salvo, J.; Lyoussi, A.; Roche, A.; Masson-Fauchier, M.; Bosq, J. C. [CEA, DEN, DER, F-13108 Saint-Paul-lez-Durance (France); Carette, M. [Aix-Marseille Univ., LCP UMR 6264, 13397, Marseille (France)


    The objective of this study is to develop nuclear heating measurement methods in zero power experimental reactors. These developments contribute to the qualification of photonics calculation schemes for the assessment of gamma heating in the future Jules Horowitz Material Testing Reactor. This paper presents the analysis of thermoluminescent detector (TLD) experiments in the UO{sub 2} core of the MINERVE Research Reactor at the French Alternative Energies and Atomic Energy Commission center in Cadarache. The experimental sources of uncertainty in the gamma dose have been reduced by improving the measurement conditions and the repeatability of the calibration step for each individual TLD. The interpretation of these measurements needs to take into account the calculation of cavity correction factors related to calibration and irradiation configurations, as well as neutron correction calculations. These calculations are based on Monte Carlo simulations of neutron-gamma and gamma-electron transport coupled particles. The comparison between calculated and measured integral gamma-ray absorbed doses in the aluminum material surrounding the TLD shows that calculations slightly overestimate the measurement, with a calculated versus experimental ratio equal to 1.04 {+-} 5.7 % (k=2). (authors)

  4. Scaled Facility Design Approach for Pool-Type Lead-Bismuth Eutectic Cooled Small Modular Reactor Utilizing Natural Circulation

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sangrok; Shin, Yong-Hoon; Lee, Jueun; Hwang, Il Soon [Seoul National University, Seoul (Korea, Republic of)


    In low carbon era, nuclear energy is the most prominent energy source of electricity. For steady ecofriendly nuclear energy supply, Generation IV reactors which are future nuclear reactor require safety, sustainability, economics and non-proliferation as four criteria. Lead cooled fast reactor (LFR) is one of these reactor type and Generation IV international forum (GIF) adapted three reference LFR systems which are a small and movable systems with long life without refueling, intermediate size and huge electricity generation system for power grid. NUTRECK (Nuclear Transmutation Energy Center of Korea) has been designed reactor called URANUS (Ubiquitous, Rugged, Accident-forgiving, Non-proliferating, and Ultra-lasting Sustainer) which is small modular reactor and using lead-bismuth eutectic coolant. To prove natural circulation capability of URANUS and analyze design based accidents, scaling mock-up experiment facility will be constructed. In this paper, simple specifications of URANUS will be presented. Then based on this feature, scaling law and scaled facility design results are presented. To validate safety feature and thermodynamics characteristic of URANUS, scaled mockup facility of URANUS is designed based on the scaling law. This mockup adapts two area scale factors, core and lower parts of mock-up are scaled for 3D flow experiment. Upper parts are scaled different size to reduce electricity power and LBE tonnage. This hybrid scaling method could distort some thermal-hydraulic parameters, however, key parameters for experiment will be matched for up-scaling. Detailed design of mock-up will be determined through iteration for design optimization.

  5. Hot Zero and Full Power Validation of PHISICS RELAP-5 Coupling

    Energy Technology Data Exchange (ETDEWEB)

    F. Lodi; C. Rabiti; A. Alfonsi; A. Epiney; M. Sumini


    PHISICS is a reactor analysis toolkit developed in the last 3 years at the Idaho National Laboratory that has been also coupled with the thermo-hydraulic plant simulator RELAP5-3D. PHISICS is aimed to provide an optimal trade off between needed computational resources and accuracy in the range of 10~100 cores. In fact this range has been identified as the next 5 to 10 years average computational capability available to nuclear engineer designing and optimizing nuclear reactor cores. Different publication has been already presented [1] showing test of the single modules composing the PHISICS package. Lately the Idaho National Laboratory had the opportunity to access to plant data for the first cycle of a PWR including Hot Zero Power (HZP) and Hot Full Power (HFP). This data provided the opportunity to validate the transport solver, the interpolation capability for mixed macro and micro cross section and the criticality search option of the PHISICS package. In the following we will firstly recall briefly the structure of the different PHISICS modules and then we will illustrate the modeling process and some preliminary results.

  6. Role of research reactors for nuclear power program in Indonesia

    Energy Technology Data Exchange (ETDEWEB)

    Soentono, S.; Arbie, B. [National Atomic Energy Agency, Batan (Indonesia)


    The main objectives of nuclear development program in Indonesia are to master nuclear science and technology, as well as to utilise peaceful uses of nuclear know-how, aiming at stepwisely socioeconomic development. A Triga Mark II, previously of 250 kW, reactor in Bandung has been in operation since 1965 and its design power has been increased to 1000 kW in 1972. Using core grid of the Triga 250 kW, BATAN designed and constructed the Kartini Reactor in Yogyakarta which started its operation in 1979. Both of these Triga reactors have served a wide spectrum of utilisation, such as training of manpower in nuclear engineering as well as radiochemistry, isotope production and beam research activities in solid state physics. In order to support the nuclear power development program in general and to suffice the reactor experiments further, simultaneously meeting the ever increasing demand for radioisotope, the third reactor, a multipurpose reactor of 30 MW called GA. Siwabessy (RSG-GAS) has been in operation since 1987 at Serpong near Jakarta. Each of these reactors has strong cooperation with Universities, namely the Bandung Institute of Technology at Bandung, the Gadjah Mada University at Yogyakarta, and the Indonesia University at Jakarta and has facilitated the man power development required. The role of these reactors, especially the multipurpose GA. Siwabessy reactor, as essential tools in nuclear power program are described including the experience gained during preproject, construction and commissioning, as well as through their operation, maintenance and utilisation.

  7. Analysis and optimisation of the Zecomix high efficiency zero emission hydrogen and power plant

    Energy Technology Data Exchange (ETDEWEB)

    Antonio Calabro; Paolo Deiana; Paolo Fiorini; Stefano Stendardo; Giuseppe Girardi [ENEA - Italian Agency for New Technologies, Rome (Italy). Energy and Environment


    The paper reports the analysis and the optimization of a high efficiency and zero emissions hydrogen and power production plant. The work is part of a larger project named Zecomix, lead by ENEA, studying an integrated process which produces both hydrogen and electricity from coal. A thermodynamic model of the plant has been realized by adopting a commercial software, ChemCAD. The key element of the proposed plant is the integration of the gasification island characterized by coal hydrogasification, simultaneous steam reforming and carbon dioxide sequestration, with the power island. Crucial characteristics involve the regeneration of the CO{sub 2}-acceptor sorbent, that takes place by a calcination process. The needed heat can be supplied directly, by burning additional coal or a part of the produced syngas in the calciner reactor, or indirectly by mean of an high temperature heat exchanger situated downstream the hydrogen combustor. An advanced steam cycle has been considered for power generation consisting of an internal combustion steam cycle characterized by the presence of innovative components such as the medium pressure steam compressor and medium-pressure high-temperature expander. Possible steps for the optimization of the whole plant have been studied. The solution proposed aims also to identify the most feasible plant configuration, adopting, where possible, current technology solution. The plant could reach an energy conversion efficiency close to 54%, competitive with innovative power cycle proposed in other studies. 6 refs., 3 figs., 2 tabs.

  8. Modular Lead-Bismuth Fast Reactors in Nuclear Power


    Vladimir Petrochenko; Georgy Toshinsky


    On the basis of the unique experience of operating reactors with heavy liquid metal coolant–eutectic lead-bismuth alloy in nuclear submarines, the concept of modular small fast reactors SVBR-100 for civilian nuclear power has been developed and validated. The features of this innovative technology are as follows: a monoblock (integral) design of the reactor with fast neutron spectrum, which can operate using different types of fuel in various fuel cycles including MOX fuel in a self-providing...

  9. Crystal Power: Piezo Coupling to the Quantum Zero Point

    CERN Document Server

    November, Laurence J


    We consider electro-optical constructions in which the Casimir force is modulated in opposition to piezo-crystal elasticity, as in a stack of alternating tunably conductive and piezo layers. Adjacent tunably conducting layers tuned to conduct, attract by the Casimir force compressing the intermediate piezo, but when subsequently detuned to insulate, sandwiched piezo layers expand elastically to restore their original dimension. In each cycle some electrical energy is made available from the quantum zero point (zp). We estimate that the maximum power that could be derived at semiconductor THz modulation rates is megawatts/cm3. Similarly a permittivity wave generated by a THz acoustic wave in a single crystal by the acousto-optic effect produces multiple coherent Casimir wave mode overtones and a bulk mode. We model the Casimir effect in a sinusoidally graded medium finding it to be very enhanced over what is found in a multilayer stack for the equivalent permittivity contrast, and more slowly decreasing with s...

  10. An Innovative Test Platform for Hydrogen Production and Zero Emission Power Generation from Coal

    Energy Technology Data Exchange (ETDEWEB)

    Calabro A; Deiana P; Fiorini P; Stendardo S; Girardi G [ENEA - Italian Agency for New Technologies, Energy and Environment Energy and Plants Division - Via Anguillarese - 301 00060 S Maria di Galeria - Rome (Italy)


    The ZECOMIX project, conceived by ENEA in the framework of Italian National Hydrogen Project, is aimed at studying an integrated process that produces both hydrogen and electricity from coal, with zero emissions and very high efficiency. The key element is the integration of a gasification process, characterized by coal hydro-gasification technology and carbon dioxide sequestration, with the power island, where an oxy-combustion occurs. Many optimization analysis and simulations have been carried out demonstrating the possibility to achieve very high net efficiencies (higher than 50% LHV) and very low (quasi-zero) emissions. The project schedule consists of the design, already started, the construction and the operation of an experimental facility finalized to demonstrate the feasibility of the described reference process. The facility will be realized in the ENEA Research Center of Casaccia, near Rome. It consists of a very flexible plant, in which more components can be tested separately or connected together. The plant is provided with a 50 kg/h coal atmospheric fluid bed gasifier, a fluid bed decarbonator/calcinator reactor filled with calcium oxide pellets, a pressurized hydro-gasifier reactor characterized by a pressure variable from 30 to 100 bar, a 100 kWe micro-turbine test bench, with the combustor chamber modified because of de-carbonized syngas fuelling and finally an oxygen/hydrogen combustor test bench, for experimental activities about the definition of stability limits, operative conditions (dilution, temperature pattern, chemicals) and combustion control. Other auxiliary components are mixing station for hydrogen-based syngas production, and an ordinary steam generator. The first part of the research project is aimed at testing the single component, in particular the main preliminary design criteria adopted for hydro-gasification reactor and carbonator reactor are presented in this paper. The second part of the Project is focused on the integration

  11. Numerical Study of High Heat Flux Performances of Flat-Tile Divertor Mock-ups with Hypervapotron Cooling Concept (United States)

    Chen, Lei; Liu, Xiang; Lian, Youyun; Cai, Laizhong


    The hypervapotron (HV), as an enhanced heat transfer technique, will be used for ITER divertor components in the dome region as well as the enhanced heat flux first wall panels. W-Cu brazing technology has been developed at SWIP (Southwestern Institute of Physics), and one W/CuCrZr/316LN component of 450 mm×52 mm×166 mm with HV cooling channels will be fabricated for high heat flux (HHF) tests. Before that a relevant analysis was carried out to optimize the structure of divertor component elements. ANSYS-CFX was used in CFD analysis and ABAQUS was adopted for thermal-mechanical calculations. Commercial code FE-SAFE was adopted to compute the fatigue life of the component. The tile size, thickness of tungsten tiles and the slit width among tungsten tiles were optimized and its HHF performances under International Thermonuclear Experimental Reactor (ITER) loading conditions were simulated. One brand new tokamak HL-2M with advanced divertor configuration is under construction in SWIP, where ITER-like flat-tile divertor components are adopted. This optimized design is expected to supply valuable data for HL-2M tokamak. supported by the National Magnetic Confinement Fusion Science Program of China (Nos. 2011GB110001 and 2011GB110004)

  12. Validation of a Micrometric remotely controlled pre-alignment system for the CLIC Linear Collider using a test setup (Mock-Up) with 5 degrees of freedom

    CERN Document Server

    Mainaud Durand, H; Griffet, S; Kemppinen, J; Leuxe, R; Sosin, M


    The CLIC main beam quadrupoles need to be prealigned within 17 um rms with respect to a straight reference line along a sliding window of 200 m. A readjustment system based on eccentric cam movers, which will provide stiffness to the support assembly, is being studied. The cam movers were qualified on a 1 degree of freedom (DOF) test setup, where a repeatability of adjustment below 1um was measured along their whole range. This paper presents the 5 DOF mock-up, built for the validation of the eccentric cam movers, as well as the first results of tests carried out: resolution of displacement along the whole range, measurements of the support eigenfrequencies.


    CERN Document Server

    Sosin, Mateusz


    The implementation study of the CLIC (Compact LInear Collider) is under way at CERN with a focus on the challenging issues. The pre-alignment precision and accuracy requirements are part of these technical challenges: the permissible transverse position errors of the linac components are typically 14 micrometers over sliding windows of 200m. To validate the proposed methods and strategies, the Large Scale Metrology section at CERN has performed campaigns of measurements on the CLIC Two Beam Test Modules, focusing inter alia on the alignment performance of the CLIC “snake”- girders configuration and the Main Beam Quadrupoles supporting structures. This paper describes the activities and results of tests which were performed on the test mock-up for the qualification of the CLIC supporting system chain for active pre-alignment. The lessons learnt (“know how”), the issues encountered in the girder position determination as well as the behaviour of the mechanical components are presented.

  14. Issues and Feasibility Demonstration of Positioning Closed Loop Control for the CLIC Supporting System Using a Test Mock-up with Five Degrees of Freedom

    CERN Document Server

    Sosin, M; Chritin, N; Griffet, S; Kemppinen, J; Mainaud Durand, H; Rude, V; Sterbini, G


    Since several years, CERN is studying the feasibility of building a high energy e+ e- linear collider: the CLIC (Compact LInear Collider). One of the challenges of such a collider is the pre-alignment precision and accuracy requirement on the transverse positions of the linac components, which is typically 14 μm over a window of 200 m. To ensure the possibility of positioning within such tight constraints, CERN Beams Department’s Survey team has worked intensively at developing the methods and technology needed to achieve that objective. This paper describes activities which were performed on a test bench (mock-up) with five degrees of freedom (DOF) for the qualification of control algorithms for the CLIC supporting system active-pre-alignment. Present understanding, lessons learned (“know how”), issues of sensors noise and mechanical components nonlinearities are presented.

  15. 间冷回热循环发动机电子样机设计%Digital mock- up design of the intercooled recuperated cycle aero- engine

    Institute of Scientific and Technical Information of China (English)

    曾庆万; 冯松涛; 马健


    Based on the tracking and analysis of the application of intercooled recuperated technology on aero-engine overseas, according to the domestic research status, experiment research of intercooler and re⁃cuperator was carried out. The general structure of the intercooled recuperated aero-engine was primarily studied and built as digital mock-up. The layout rules of components, systems, and special intercooler pipe⁃line and recuperator pipeline were summarized. By building the mock-up, the characteristics of project de⁃sign for the intercooled recuperated aero-engine were grasped, and the layout of components and systems were explored, which provide a basement for the optimization of engineering design.%在追踪和分析国外间冷回热技术在航空发动机领域应用的基础上,根据国内研究现状,对间冷器和回热器进行了模拟实验研究。初步拟定了间冷回热发动机总体结构方案,并将此结构方案建立为电子样机。总结了间冷回热发动机中各部件、系统,以及特有的间冷器管路系统和回热器管路系统的布局规律。通过样机建模,初步掌握了间冷回热发动机总体结构方案设计特点,探索了各部件、系统的布局,为间冷回热发动机的进一步研究积累了经验。

  16. Power distribution control of CANDU reactors based on modal representation of reactor kinetics

    Energy Technology Data Exchange (ETDEWEB)

    Xia, Lingzhi, E-mail: [Department of Electrical and Computer Engineering, The University of Western Ontario, London, Ontario N6A 5B9 (Canada); Jiang, Jin, E-mail: [Department of Electrical and Computer Engineering, The University of Western Ontario, London, Ontario N6A 5B9 (Canada); Luxat, John C., E-mail: [Department of Engineering Physics, McMaster University, Hamilton, Ontario L8S 4L7 (Canada)


    Highlights: • Linearization of the modal synthesis model of neutronic kinetic equations for CANDU reactors. • Validation of the linearized dynamic model through closed-loop simulations by using the reactor regulating system. • Design of a LQR state feedback controller for CANDU core power distribution control. • Comparison of the results of this new controller against those of the conventional reactor regulation system. - Abstract: Modal synthesis representation of a neutronic kinetic model for a CANDU reactor core has been utilized in the analysis and synthesis for reactor control systems. Among all the mode shapes, the fundamental mode of the power distribution, which also coincides with the desired reactor power distribution during operation, is used in the control system design. The nonlinear modal models are linearized around desired operating points. Based on the linearized model, linear quadratic regulator (LQR) control approach is used to synthesize a state feedback controller. The performance of this controller has been evaluated by using the original nonlinear models under load-following conditions. It has been demonstrated that the proposed reactor control system can produce more uniform power distribution than the traditional reactor regulation systems (RRS); in particular, it is more effective in compensating the Xenon induced transients.

  17. Facility for a Low Power Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chalker, R. G.


    Preliminary investigation indicates that a reactor facility with ample research provisions for use by University or other interested groups, featuring safety in design, can be economically constructed in the Los Angeles area. The complete installation, including an underground gas-tight reactor building, with associated storage and experiment assembly building, administration offices, two general laboratory buildings, hot latoratory and lodge, can be constructed for approxinately $1,500,000. This does not include the cost of the reactor itself or of its auxiliary equipment,

  18. Gas core reactor power plants designed for low proliferation potential

    Energy Technology Data Exchange (ETDEWEB)

    Lowry, L.L. (comp.)


    The feasibility of gas core nuclear power plants to provide adequate power while maintaining a low inventory and low divertability of fissile material is studied. Four concepts were examined. Two used a mixture of UF/sub 6/ and helium in the reactor cavities, and two used a uranium-argon plasma, held away from the walls by vortex buffer confinement. Power levels varied from 200 to 2500 MWth. Power plant subsystems were sized to determine their fissile material inventories. All reactors ran, with a breeding ratio of unity, on /sup 233/U born from thorium. Fission product removal was continuous. Newly born /sup 233/U was removed continuously from the breeding blanket and returned to the reactor cavities. The 2500-MWth power plant contained a total of 191 kg of /sup 233/U. Less than 4 kg could be diverted before the reactor shut down. The plasma reactor power plants had smaller inventories. In general, inventories were about a factor of 10 less than those in current U.S. power reactors.

  19. High power density reactors based on direct cooled particle beds (United States)

    Powell, J. R.; Horn, F. L.

    Reactors based on direct cooled High Temperature Gas Cooled Reactor (HTGR) type particle fuel are described. The small diameter particle fuel is packed between concentric porous cylinders to make annular fuel elements, with the inlet coolant gas flowing inwards. Hot exit gas flows out along the central channel of each element. Because of the very large heat transfer area in the packed beds, power densities in particle bed reactors (PBRs) are extremely high resulting in compact, lightweight systems. Coolant exit temperatures are high, because of the ceramic fuel temperature capabilities, and the reactors can be ramped to full power and temperature very rapidly. PBR systems can generate very high burst power levels using open cycle hydrogen coolant, or high continuous powers using closed cycle helium coolant. PBR technology is described and development requirements assessed.

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

    CERN Document Server

    Cammi, Antonio; Chiesa, Davide; Clemenza, Massimiliano; Pozzi, Stefano; Previtali, Ezio; Sisti, Monica; Magrotti, Giovanni; Prata, Michele; Salvini, Andrea


    In this work, the characterization of the full-power steady state of the TRIGA Mark II nuclear reactor of the University of Pavia is performed by coupling Monte Carlo (MC) simulation for neutronics with "Multiphysics" model for thermal-hydraulics. Neutronic analyses have been performed starting from a MC model of the entire reactor system, based on the MCNP5 code, that was already validated in fresh fuel and zero-power configuration (in which thermal effects are negligible) using the available experimental data as benchmark. In order to describe the full-power reactor configuration, the temperature distribution in the core is necessary. To evaluate it, a thermal-hydraulic model has been developed, using the power distribution results from 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 introduced in the MC model and a benchmark analysis is carr...


    Avery, R.


    A nuclear reactor having a region operating predominantly on fast neutrons and another region operating predominantly on slow neutrons is described. The fast region is a plutonium core and the slow region is a natural uranium blanket around the core. Both of these regions are free of moderator. A moderating reflector surrounds the uranium blanket. The moderating material and thickness of the reflector are selected so that fissions in the uranium blanket make a substantial contribution to the reactivity of the reactor.

  2. Electrolysers as a load management mechanism for power systems with wind power and zero-carbon thermal power plant

    Energy Technology Data Exchange (ETDEWEB)

    Troncoso, E. [School of Industrial Engineering, Universidad Las Palmas de Gran Canaria (Spain); Newborough, M. [ITM Power Research Ltd., Mill House, Royston Road, Wendens Ambo, Saffron Walden CB11 4JX (United Kingdom)


    For an isolated power system the deployment of a large stock of electrolysers is investigated as a means for increasing the penetrations of wind power plant and zero-carbon thermal power plant. Consideration is given to the sizing and utilization of an electrolyser stock for three electrolyser implementation cases and three operational strategies, installed capacity ranges of 20-100% for wind power and 10-35% for zero-carbon thermal power plant (as proportions of the power system's maximum electrical demand) were investigated. Relative to wind-hydrogen alone, hydrogen yields are substantially increased especially on low-wind days. The average load placed on fossil-fuelled power plant is substantially decreased (while achieving a virtually flat load profile) and the carbon intensity of electricity can be reduced to values of <0.1 kg CO{sub 2}/kWh{sub e}. The trade-offs between the carbon intensity of the electricity delivered, the carbon intensity of the hydrogen produced and the daily hydrogen yield are explored. For example (on the variable wind day for Strategy C with respective wind power and zero-carbon thermal power penetrations of 100% and 35%), if the carbon intensity of hydrogen is relaxed from 0 to 3 kg CO{sub 2}/kg H{sub 2}, the hydrogen yield can be increased from 435 tonnes to 1115 tonnes (which is the energy equivalent of 120% of consumer demand for electricity on that day). The findings suggest that the deployment of electrolysers on both the supply and demand-side of the power system can contribute nationally-significant amounts of zero or low-carbon hydrogen without exceeding the power system's current maximum system demand. (author)

  3. An overview of future sustainable nuclear power reactors

    Directory of Open Access Journals (Sweden)

    Andreas Poullikkas


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

  4. Investigation of materials for fusion power reactors (United States)

    Bouhaddane, A.; Slugeň, V.; Sojak, S.; Veterníková, J.; Petriska, M.; Bartošová, I.


    The possibility of application of nuclear-physical methods to observe radiation damage to structural materials of nuclear facilities is nowadays a very actual topic. The radiation damage to materials of advanced nuclear facilities, caused by extreme radiation stress, is a process, which significantly limits their operational life as well as their safety. In the centre of our interest is the study of the radiation degradation and activation of the metals and alloys for the new nuclear facilities (Generation IV fission reactors, fusion reactors ITER and DEMO). The observation of the microstructure changes in the reactor steels is based on experimental investigation using the method of positron annihilation spectroscopy (PAS). The experimental part of the work contains measurements focused on model reactor alloys and ODS steels. There were 12 model reactor steels and 3 ODS steels. We were investigating the influence of chemical composition on the production of defects in crystal lattice. With application of the LT 9 program, the spectra of specimen have been evaluated and the most convenient samples have been determined.

  5. Design of megawatt power level heat pipe reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mcclure, Patrick Ray [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Poston, David Irvin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dasari, Venkateswara Rao [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Reid, Robert Stowers [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)


    An important niche for nuclear energy is the need for power at remote locations removed from a reliable electrical grid. Nuclear energy has potential applications at strategic defense locations, theaters of battle, remote communities, and emergency locations. With proper safeguards, a 1 to 10-MWe (megawatt electric) mobile reactor system could provide robust, self-contained, and long-term power in any environment. Heat pipe-cooled fast-spectrum nuclear reactors have been identified as a candidate for these applications. Heat pipe reactors, using alkali metal heat pipes, are perfectly suited for mobile applications because their nature is inherently simpler, smaller, and more reliable than “traditional” reactors. The goal of this project was to develop a scalable conceptual design for a compact reactor and to identify scaling issues for compact heat pipe cooled reactors in general. Toward this goal two detailed concepts were developed, the first concept with more conventional materials and a power of about 2 MWe and a the second concept with less conventional materials and a power level of about 5 MWe. A series of more qualitative advanced designs were developed (with less detail) that show power levels can be pushed to approximately 30 MWe.

  6. Small reactor power systems for manned planetary surface bases (United States)

    Bloomfield, Harvey S.


    A preliminary feasibility study of the potential application of small nuclear reactor space power systems to manned planetary surface base missions was conducted. The purpose of the study was to identify and assess the technology, performance, and safety issues associated with integration of reactor power systems with an evolutionary manned planetary surface exploration scenario. The requirements and characteristics of a variety of human-rated modular reactor power system configurations selected for a range of power levels from 25 kWe to hundreds of kilowatts is described. Trade-off analyses for reactor power systems utilizing both man-made and indigenous shielding materials are provided to examine performance, installation and operational safety feasibility issues. The results of this study have confirmed the preliminary feasibility of a wide variety of small reactor power plant configurations for growth oriented manned planetary surface exploration missions. The capability for power level growth with increasing manned presence, while maintaining safe radiation levels, was favorably assessed for nominal 25 to 100 kWe modular configurations. No feasibility limitations or technical barriers were identified and the use of both distance and indigenous planetary soil material for human rated radiation shielding were shown to be viable and attractive options.

  7. Analysis of Nigeria research reactor-1 thermal power calibration methods

    Energy Technology Data Exchange (ETDEWEB)

    Agbo, Sunday Arome; Ahmed, Yusuf Aminu; Ewa, Ita Okon; Jibrin, Yahaya [Ahmadu Bello University, Zaria (Nigeria)


    This paper analyzes the accuracy of the methods used in calibrating the thermal power of Nigeria Research Reactor-1 (NIRR-1), a low-power miniature neutron source reactor located at the Centre for Energy Research and Training, Ahmadu Bello University, Zaria, Nigeria. The calibration was performed at three different power levels: low power (3.6 kW), half power (15 kW), and full power (30 kW). Two methods were used in the calibration, namely, slope and heat balance methods. The thermal power obtained by the heat balance method at low power, half power, and full power was 3.7 ± 0.2 kW, 15.2 ± 1.2 kW, and 30.7 ± 2.5 kW, respectively. The thermal power obtained by the slope method at half power and full power was 15.8 ± 0.7 kW and 30.2 ± 1.5 kW, respectively. It was observed that the slope method is more accurate with deviations of 4% and 5% for calibrations at half and full power, respectively, although the linear fit (slope method) on average temperature-rising rates during the thermal power calibration procedure at low power (3.6 kW) is not fitting. As such, the slope method of power calibration is not suitable at lower power for NIRR-1.

  8. Nuclear Power from Fission Reactors. An Introduction. (United States)

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

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

  9. Total loss of AC power analysis for EPR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Darnowski, Piotr, E-mail: [Warsaw University of Technology, Institute of Heat Engineering, Nowowiejska 21/25, 00-665 Warsaw (Poland); Skrzypek, Eleonora, E-mail: [Warsaw University of Technology, Institute of Heat Engineering, Nowowiejska 21/25, 00-665 Warsaw (Poland); National Centre for Nuclear Research (NCBJ), A. Sołtana 7, 05-400 Otwock-Świerk (Poland); Mazgaj, Piotr, E-mail: [Warsaw University of Technology, Institute of Heat Engineering, Nowowiejska 21/25, 00-665 Warsaw (Poland); Świrski, Konrad [Warsaw University of Technology, Institute of Heat Engineering, Nowowiejska 21/25, 00-665 Warsaw (Poland); Gandrille, Pascal [AREVA NP SAS, Tour AREVA, 1 place Jean Millier, 92084 Paris La Défense (France)


    Highlights: • Total loss of AC power (Station Blackout) was simulated for the EPR reactor model. • In-vessel phase of the accident is under consideration. • Comparison of MELCOR and MAAP results is presented. • MELCOR and MAAP results are comparable. - Abstract: In this paper the results of severe accident simulations for the EPR reactor in the case of loss of offsite power combined with total failure of all diesel generators (total loss of AC power) are presented. Calculations were performed with MELCOR 2.1 computer code for in-vessel phase of the accident. In this scenario, the unavailability of all offsite and onsite power sources and the lack of cooling leads directly to core degradation, material relocation to the lower plenum and rupture of the reactor pressure vessel. MELCOR results were compared qualitatively and quantitatively with MAAP4 code results and show a good agreement.

  10. Power Reactor Fuel Reprocessing: Mechanical Phase

    Energy Technology Data Exchange (ETDEWEB)

    Klima, B. B.


    The major events in the mechanical phase of the Power Reactor fuels reprocessing program during June were: 1. Feasibility of shearing of fuel elements without disassembly has been demonstrated in tests using porcelain-loaded prototype fuel elements. 2. Further work with the Manco shear was not deemed tb be advisable since permission has been granted to use another shear for cutting UO{sub 2}-loaded fuel elements. 3. Necessity to strip the windows in Building 3048, to sandblast, and repaint them has seriously disrupted occupancy of the cell by July 1. Start of installation probably will not be before August 1. 4. A cold SRE element should be received during July which will permit a direct look a t the problems associated with processing of these irradiated fuel elements. 5. Concurrence with AEC, Atomics International, and ORNL people on the fabrication of a poisoned carrier was obtained and all criteria for the carrier were released and the design was completed. 6. A decision was made to install and use a 24-inch Ty-Sa-Man saw which is on hand and was originally purchased for use in the Segmenting Facility for the SRE reprocessing. This will be used instead of the multipurpose saw to allow more time to refine the design of that saw. The multipurpose saw will be installed for use in subsequent reprocessing programs. This report will chronicle the changes in status which occurred during the calendar month of June. A complete description of each item is not included and may be found in the parent report. The dates indicated on the schedule have slipped since the last report primarily due to increase in scope of the work and postponement on all phases of the work except for the SRE preparations. Twenty-four new items have been added to the schedule. The status of procurement is shown. A total of 93 purchase requests have been turned in to t% Purchasing Department. A total of $199,261.83 has been committed by purchase orders, and a total of 56 purchase orders have been

  11. The effect of biological shielding on fast neutron and photon transport in the VVER-1000 mock-up model placed in the LR-0 reactor. (United States)

    Košťál, Michal; Cvachovec, František; Milčák, Ján; Mravec, Filip


    The paper is intended to show the effect of a biological shielding simulator on fast neutron and photon transport in its vicinity. The fast neutron and photon fluxes were measured by means of scintillation spectroscopy using a 45×45 mm(2) and a 10×10 mm(2) cylindrical stilbene detector. The neutron spectrum was measured in the range of 0.6-10 MeV and the photon spectrum in 0.2-9 MeV. The results of the experiment are compared with calculations. The calculations were performed with various nuclear data libraries.

  12. The state of the art report on the development of manufacturing technology of fusion reactor FW blanket and mock-up in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Park, J. Y.; Jeong, Y. H.; Baek, J. H.; Kim, J. H.; Kim, H. G


    The joining technology of first wall blanket has been developed by JAERI in collaboration with Kawasaki Heavy Industry, Isuau Motors and University of Tsukuba in Japan. A variety of joining technologies including HIP, brazing, casing and friction welding was applied to the manufacturing of SS/SS and Cu/SS joint. In Be/Cu joining, it was emphasized to find the optimal HIP temperature lower than 650 .deg. C in order to avoid excessive SS sensitization because the joining of Be tile to Cu heat sink is a final processing step in the manufacturing of FW blanket. The selected HIP condition were 620 .deg. C, 150MPa and 2hr with Cu interlayer. Sample tests for joints was completed by 1995. The small scale mockup was manufactured and its performance was qualified by end of 2000. From 2001, the manufacturing and the characterization has been carried out for the larger scale mockup.

  13. Performance test results of mock-up model test facility with a full-scale reaction tube for HTTR hydrogen production system. Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Inagaki, Yoshiyuki; Hayashi, Koji; Kato, Michio [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment] [and others


    Research on a hydrogen production system by steam reforming of methane, chemical reaction; CH{sub 4} + H{sub 2}O {yields} 3H{sub 2}O + CO, has been carried out to couple with the HTTR for establishment of high-temperature nuclear heat utilization technology and contribution to hydrogen energy society in future. The mock-up test facility with a full-scale reaction tube test facility, a model simulating one reaction tube of a steam reformer of the HTTR hydrogen production system in full scale, was fabricated to perform tests on controllability, hydrogen production performance etc. under the same pressure and temperature conditions as those of the HTTR hydrogen production system. The design and fabrication of the test facility started from 1997, and the all components were installed until September in 2001. In a performance test conducted from October in 2001 to February in 2002, performance of each component was examined and hydrogen of 120m{sup 3}{sub N}/h was successfully produced with high-temperature helium gas. This report describes the performance test results on components performance, hydrogen production characteristics etc., and main troubles and countermeasures. (author)

  14. Zero-emission fuel-fired power plants with ion transport membrane

    NARCIS (Netherlands)

    Yantovski, E.; Gorski, J.; Smyth, B.; Elshof, ten J.


    Firstly, some points in relation to the history of zero-emissions power cycles are highlighted. Amongst the many schemes, only one which deals with the combustion of a fuel in “artificial air” (i.e. a mixture of oxygen and re-circulated carbon dioxide), is selected. This paper describes the zero em

  15. Technological Transfer from Research Nuclear Reactors to New Generation Nuclear Power Reactors (United States)

    Radulescu, Laura; Pavelescu, Margarit


    The goal of this paper is the analysis of the technological transfer role in the nuclear field, with particular emphasis on nuclear reactors domain. The presentation is sustained by historical arguments. In this frame, it is very important to start with the achievements of the first nuclear systems, for instant those with natural uranium as fuel and heavy water as moderator, following in time through the history until the New Generation Nuclear Power Reactors. Starting with 1940, the accelerated development of the industry has implied the increase of the global demand for energy. In this respect, the nuclear energy could play an important role, being essentially an unlimited source of energy. However, the nuclear option faces the challenges of increasingly demanding safety requirements, economic competitiveness and public acceptance. Worldwide, a significant amount of experience has been accumulated during development, licensing, construction, and operation of nuclear power reactors. The experience gained is a strong basis for further improvements. Actually, the nuclear programs of many countries are addressing the development of advanced reactors, which are intended to have better economics, higher reliability, improved safety, and proliferation-resistant characteristics in order to overcome the current concerns about nuclear power. Advanced reactors, now under development, may help to meet the demand for energy power of both developed and developing countries as well as for district heating, desalination and for process heat. The paper gives historical examples that illustrate the steps pursued from first research nuclear reactors to present advanced power reactors. Emphasis was laid upon the fact that the progress is due to the great discoveries of the nuclear scientists using the technological transfer.

  16. New reactor technology: safety improvements in nuclear power systems. (United States)

    Corradini, M L


    Almost 450 nuclear power plants are currently operating throughout the world and supplying about 17% of the world's electricity. These plants perform safely, reliably, and have no free-release of byproducts to the environment. Given the current rate of growth in electricity demand and the ever growing concerns for the environment, nuclear power can only satisfy the need for electricity and other energy-intensive products if it can demonstrate (1) enhanced safety and system reliability, (2) minimal environmental impact via sustainable system designs, and (3) competitive economics. The U.S. Department of Energy with the international community has begun research on the next generation of nuclear energy systems that can be made available to the market by 2030 or earlier, and that can offer significant advances toward these challenging goals; in particular, six candidate reactor system designs have been identified. These future nuclear power systems will require advances in materials, reactor physics, as well as thermal-hydraulics to realize their full potential. However, all of these designs must demonstrate enhanced safety above and beyond current light water reactor systems if the next generation of nuclear power plants is to grow in number far beyond the current population. This paper reviews the advanced Generation-IV reactor systems and the key safety phenomena that must be considered to guarantee that enhanced safety can be assured in future nuclear reactor systems.

  17. Steady performance of a zero valent iron packed anaerobic reactor for azo dye wastewater treatment under variable influent quality

    Institute of Scientific and Technical Information of China (English)

    Yaobin Zhang; Yiwen Liu; Yanwen Jing; Zhiqiang Zhao; Xie Quan


    Zero valent iron (ZVI) is expected to help create an enhanced anaerobic environment that might improve the performance of anaerobic treatment.Based on this idea,a novel ZVI packed upflow anaerobic sludge blanket (ZVI-UASB) reactor was developed to treat azo dye wastewater with variable influent quality.The results showed that the reactor was less influenced by increases of Reactive Brilliant Red X-3B concentration from 50 to 1000 mg/L and chemical oxygen demand (COD) from 1000 to 7000 mg/L in the feed than a reference UASB reactor without the ZVI.The ZVI decreased oxidation-reduction potential in the reactor by about 80 mV.Iron ion dissolution from the ZVI could buffer acidity in the reactor,the amount of which was related to the COD concentration.Fluorescence in situ hybridization test showed the abundance of methanogens in the sludge of the ZVI-UASB reactor was significantly greater than that of the reference one.Denaturing gradient gel electrophoresis showed that the ZVI increased the diversity of microbial strains responsible for high efficiency.

  18. Steady performance of a zero valent iron packed anaerobic reactor for azo dye wastewater treatment under variable influent quality. (United States)

    Zhang, Yaobin; Liu, Yiwen; Jing, Yanwen; Zhao, Zhiqiang; Quan, Xie


    Zero valent iron (ZVI) is expected to help create an enhanced anaerobic environment that might improve the performance of anaerobic treatment. Based on this idea, a novel ZVI packed upflow anaerobic sludge blanket (ZVI-UASB) reactor was developed to treat azo dye wastewater with variable influent quality. The results showed that the reactor was less influenced by increases of Reactive Brilliant Red X-3B concentration from 50 to 1000 mg/L and chemical oxygen demand (COD) from 1000 to 7000 mg/L in the feed than a reference UASB reactor without the ZVI. The ZVI decreased oxidation-reduction potential in the reactor by about 80 mV. Iron ion dissolution from the ZVI could buffer acidity in the reactor, the amount of which was related to the COD concentration. Fluorescence in situ hybridization test showed the abundance of methanogens in the sludge of the ZVI-UASB reactor was significantly greater than that of the reference one. Denaturing gradient gel electrophoresis showed that the ZVI increased the diversity of microbial strains responsible for high efficiency.

  19. Search for axions at a power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Cavaignac, J.F.; Hoummada, A.; Koang, D.H.; Ost, B.; Vignon, B.; Wilson, R. (Grenoble-1 Univ., 38 (France). Inst. des Sciences Nucleaires); Declais, Y.; Girardi, G.; de Kerret, H.; Pessard, H.


    A search has been conducted for the axion at the Bugey reactor which is owned and operated by Electricite de France. The axion production should be proportional to the magnetic transition of np capture, and be detectable by its decay into 2..gamma.. rays. No signal was observed in this measurement. Also no axion signal was seen from a single proton magnetic transition of /sup 97/Nb. Using those two results, the axion can be excluded with a mass up to 1 MeV in the Peccei-Quinn formalism.

  20. Transportation and storage of foreign spent power reactor fuel

    Energy Technology Data Exchange (ETDEWEB)


    This report describes the generic actions to be taken by the Department of Energy, in cooperation with other US government agencies, foreign governments, and international organizations, in support of the implementation of Administration policies with respect to the following international spent fuel management activities: bilateral cooperation related to expansion of foreign national storage capacities; multilateral and international cooperation related to development of multinational and international spent fuel storage regimes; fee-based transfer of foreign spent power reactor fuel to the US for storage; and emergency transfer of foreign spent power reactor fuel to the US for storage.

  1. Modular Lead-Bismuth Fast Reactors in Nuclear Power

    Directory of Open Access Journals (Sweden)

    Vladimir Petrochenko


    Full Text Available On the basis of the unique experience of operating reactors with heavy liquid metal coolant–eutectic lead-bismuth alloy in nuclear submarines, the concept of modular small fast reactors SVBR-100 for civilian nuclear power has been developed and validated. The features of this innovative technology are as follows: a monoblock (integral design of the reactor with fast neutron spectrum, which can operate using different types of fuel in various fuel cycles including MOX fuel in a self-providing mode. The reactor is distinct in that it has a high level of self-protection and passive safety, it is factory manufactured and the assembled reactor can be transported by railway. Multipurpose application of the reactor is presumed, primarily, it can be used for regional power to produce electricity, heat and for water desalination. The Project is being realized within the framework of state-private partnership with joint venture OJSC “AKME-Engineering” established on a parity basis by the State Atomic Energy Corporation “Rosatom” and the Limited Liability Company “EuroSibEnergo”.

  2. Background radiation measurements at high power research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ashenfelter, J. [Wright Laboratory, Department of Physics, Yale University, New Haven, CT 06520 (United States); Balantekin, B. [Department of Physics, University of Wisconsin, Madison, WI 53706 (United States); Baldenegro, C.X. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Band, H.R. [Wright Laboratory, Department of Physics, Yale University, New Haven, CT 06520 (United States); Barclay, G. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Bass, C.D. [Department of Chemistry and Physics, Le Moyne College, Syracuse, NY 13214 (United States); Berish, D. [Department of Physics, Temple University, Philadelphia, PA 19122 (United States); Bowden, N.S., E-mail: [Nuclear and Chemical Sciences Division, Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Bryan, C.D. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Cherwinka, J.J. [Physical Sciences Laboratory, University of Wisconsin, Madison, WI 53706 (United States); Chu, R. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Department of Physics, University of Tennessee, Knoxville, TN 37996 (United States); Classen, T. [Nuclear and Chemical Sciences Division, Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Davee, D. [Department of Physics, College of William and Mary, Williamsburg, VA 23187 (United States); Dean, D.; Deichert, G. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Dolinski, M.J. [Department of Physics, Drexel University, Philadelphia, PA 19104 (United States); Dolph, J. [Physics Department, Brookhaven National Laboratory, Upton, NY 11973 (United States); Dwyer, D.A. [Physics Division, Lawrence Berkeley National Laboratory, Berkeley, CA 94720 (United States); Fan, S. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Department of Physics, University of Tennessee, Knoxville, TN 37996 (United States); and others


    Research reactors host a wide range of activities that make use of the intense neutron fluxes generated at these facilities. Recent interest in performing measurements with relatively low event rates, e.g. reactor antineutrino detection, at these facilities necessitates a detailed understanding of background radiation fields. Both reactor-correlated and naturally occurring background sources are potentially important, even at levels well below those of importance for typical activities. Here we describe a comprehensive series of background assessments at three high-power research reactors, including γ-ray, neutron, and muon measurements. For each facility we describe the characteristics and identify the sources of the background fields encountered. The general understanding gained of background production mechanisms and their relationship to facility features will prove valuable for the planning of any sensitive measurement conducted therein.

  3. Background Radiation Measurements at High Power Research Reactors

    CERN Document Server

    Ashenfelter, J; Baldenegro, C X; Band, H R; Barclay, G; Bass, C D; Berish, D; Bowden, N S; Bryan, C D; Cherwinka, J J; Chu, R; Classen, T; Davee, D; Dean, D; Deichert, G; Dolinski, M J; Dolph, J; Dwyer, D A; Fan, S; Gaison, J K; Galindo-Uribarri, A; Gilje, K; Glenn, A; Green, M; Han, K; Hans, S; Heeger, K M; Heffron, B; Jaffe, D E; Kettell, S; Langford, T J; Littlejohn, B R; Martinez, D; McKeown, R D; Morrell, S; Mueller, P E; Mumm, H P; Napolitano, J; Norcini, D; Pushin, D; Romero, E; Rosero, R; Saldana, L; Seilhan, B S; Sharma, R; Stemen, N T; Surukuchi, P T; Thompson, S J; Varner, R L; Wang, W; Watson, S M; White, B; White, C; Wilhelmi, J; Williams, C; Wise, T; Yao, H; Yeh, M; Yen, Y -R; Zhang, C; Zhang, X


    Research reactors host a wide range of activities that make use of the intense neutron fluxes generated at these facilities. Recent interest in performing measurements with relatively low event rates, e.g. reactor antineutrino detection, at these facilities necessitates a detailed understanding of background radiation fields. Both reactor-correlated and naturally occurring background sources are potentially important, even at levels well below those of importance for typical activities. Here we describe a comprehensive series of background assessments at three high-power research reactors, including $\\gamma$-ray, neutron, and muon measurements. For each facility we describe the characteristics and identify the sources of the background fields encountered. The general understanding gained of background production mechanisms and their relationship to facility features will prove valuable for the planning of any sensitive measurement conducted therein.

  4. RTC-control of power transients in nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ratemi, Wajdi Mohamed [Alfateh University, PO Box 13040, Tripoli (Libyan Arab Jamahiriya)


    In this paper, the new Reactivity Trace Curve (RTC) method (Ratemi 1993,1994), which is based on the dynamic period studies (Bernard et al.,1984), has been studied for maneuvering of the nuclear reactor power without power shooting. The reactor is modeled with one group of delayed neutrons with temperature feedback effect, as well as, Xenon feedback effect. A precursors concentration model is used to provide for the effective dynamic decay constant (in one group case, it is a static one). The RTC-identifier which is given by a differential equation is then solved at each sampling time (for one group, it has an analytical solution). Its solution is what is called the Reactivity Trace Curve which keeps the power steady at the desired power. An inverse kinetic model which uses the on-line power data for reactivity calculation is used to provide initial condition (initial reactivity) for the RTC- power controller. Also feedback model are needed to evaluate both the temperature and Xenon reactivities which when subtracted from the RTC-value, one then can determine the reactivity required to keep the reactor power steady without power shooting. (authors)

  5. Hybrid zero-voltage switching (ZVS) control for power inverters (United States)

    Amirahmadi, Ahmadreza; Hu, Haibing; Batarseh, Issa


    A power inverter combination includes a half-bridge power inverter including first and second semiconductor power switches receiving input power having an intermediate node therebetween providing an inductor current through an inductor. A controller includes input comparison circuitry receiving the inductor current having outputs coupled to first inputs of pulse width modulation (PWM) generation circuitry, and a predictive control block having an output coupled to second inputs of the PWM generation circuitry. The predictive control block is coupled to receive a measure of Vin and an output voltage at a grid connection point. A memory stores a current control algorithm configured for resetting a PWM period for a switching signal applied to control nodes of the first and second power switch whenever the inductor current reaches a predetermined upper limit or a predetermined lower limit.

  6. Power flow control using distributed saturable reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dimitrovski, Aleksandar D.


    A magnetic amplifier includes a saturable core having a plurality of legs. Control windings wound around separate legs are spaced apart from each other and connected in series in an anti-symmetric relation. The control windings are configured in such a way that a biasing magnetic flux arising from a control current flowing through one of the plurality of control windings is substantially equal to the biasing magnetic flux flowing into a second of the plurality of control windings. The flow of the control current through each of the plurality of control windings changes the reactance of the saturable core reactor by driving those portions of the saturable core that convey the biasing magnetic flux in the saturable core into saturation. The phasing of the control winding limits a voltage induced in the plurality of control windings caused by a magnetic flux passing around a portion of the saturable core.

  7. Innovative oxy-coal combustion process suitable for future and more efficient zero emission power plants

    Energy Technology Data Exchange (ETDEWEB)

    Benelli, G.; Malavasi, M.; Girardi, G. [ENEL Ricerca (Italy)


    The problem with CO{sub 2} capture from a flue gas stream is related to its low concentration, which makes the process of separation very energy-intensive, complex and, as a result, expensive. The CO{sub 2} separation process can be optimized by increasing the concentration of CO{sub 2} and reducing nitrogen concentration in the stream as it happens, in the oxy-fuel combustion process. In such a case, the oxidant flow is typically a mixture of oxygen, steam and carbon dioxide, with a very low concentration of nitrogen. Since the oxy-combustion process leads to very high temperatures, flue gases must be circulating through the chemical reactor to keep the combustion adiabatic temperature below acceptable values, due to the limits imposed by material resistance. This paper focuses on an innovative oxy-coal combustion process named ISOTHERM{reg_sign}, based on a flameless combustion technique which is mentioned in recent literature also as 'Mild' combustion. The combustion process takes place within a pressurized and refractory-lined furnace, approaching temperatures close to 2000 K. The process has been experienced at pressurized conditions up to 4 bar on a 5 MW pilot plant for thousands of hours. In this paper, starting from a detailed description of the process, results obtained by the preliminary experimental tests are presented and discussed. Then, a development and demonstration program to assess the suitability of this technology for zero emission power generation at large scale in one of the units of Brindisi power station is presented. 10 refs., 5 figs., 2 tabs.

  8. Acrylic acid and electric power cogeneration in an SOFC reactor. (United States)

    Ji, Baofeng; Wang, Jibo; Chu, Wenling; Yang, Weishen; Lin, Liwu


    A highly efficient catalyst, MoV(0.3)Te(0.17)Nb(0.12)O, used for acrylic acid (AA) production from propane, was used as an anodic catalyst in an SOFC reactor, from which AA and electric power were cogenerated at 400-450 degrees C.

  9. Fuel element concept for long life high power nuclear reactors (United States)

    Mcdonald, G. E.; Rom, F. E.


    Nuclear reactor fuel elements have burnups that are an order of magnitude higher than can currently be achieved by conventional design practice. Elements have greater time integrated power producing capacity per unit volume. Element design concept capitalizes on known design principles and observed behavior of nuclear fuel.

  10. A gas-cooled reactor surface power system (United States)

    Lipinski, Ronald J.; Wright, Steven A.; Lenard, Roger X.; Harms, Gary A.


    A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life-cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitride clad in Nb1%Zr, which has been extensively tested under the SP-100 program. The fuel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fuel and stabilizing the geometry against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality can not occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars.

  11. Analysis on Zero Power Experiment of High Flux Engineering Test Reactor with Three-Dimensional Continuous Energy Monte Carlo Code%三维堆芯连续能量蒙特卡罗程序用于HFETR零功率物理实验计算分析研究

    Institute of Scientific and Technical Information of China (English)



    采用三维堆芯连续能量蒙特卡罗程序( MCNP)对高通量工程试验堆(HFETR)零功率物理实验进行计算分析.从计算结果可以看出,在零功率反应堆上,径向铍反射层应当考虑金属铍中的杂质和密度修正,同时需要考虑控制棒过渡段的10B含量修正;而HFETR上的铍块则可以认为是纯金属铍,控制棒过渡段中10B含量也需另作考虑.从计算结果来看,多数参数(堆芯中子有效增殖系数keff、中子通量密度相对分布、γ剂量率相对分布以及多数部件反应性价值)均较好满足误差要求,而个别小反应性部件计算误差较大可能与MCNP程序模型有关.%Three-dimensional, continuous energy Monte Carlo code (MCNP) is adopted to carry out the analysis of zero power experiment of HFETR. From the results, the impurity, density of Beryllium block and the B concentration in control rod transition part should be carefully determined in the analysis of zero power experiment. While in the experiment of HFETR, the Beryllium block is considered as pure metal and the 10B concentration in control rod transition part is different from that of zero power experiment. From the calculation results, these parameters (effective neutron multiplication factor Keff relative distribution of neutron flux density, y dose rate distribution and component reactivity) are quite fit with the experiment. The difference of small reactivity between calculation and experiment is quite large, and may be related to the deficiency of MCNP model.

  12. Estimates of power requirements for a manned Mars rover powered by a nuclear reactor (United States)

    Morley, Nicholas J.; El-Genk, Mohamed S.; Cataldo, Robert; Bloomfield, Harvey


    This paper assesses the power requirement for a Manned Mars Rover vehicle. Auxiliary power needs are fulfilled using a hybrid solar photovoltaic/regenerative fuel cell system, while the primary power needs are met using an SP-100 type reactor. The primary electric power needs, which include 30-kWe net user power, depend on the reactor thermal power and the efficiency of the power conversion system. Results show that an SP-100 type reactor coupled to a Free Piston Stirling Engine (FPSE) yields the lowest total vehicle mass and lowest specific mass for the power system. The second lowest mass was for a SP-100 reactor coupled to a Closed Brayton Cycle (CBC) using He/Xe as the working fluid. The specific mass of the nuclear reactor power systrem, including a man-rated radiation shield, ranged from 150-kg/kWe to 190-kg/kWe and the total mass of the Rover vehicle varied depend upon the cruising speed.

  13. High density operation for reactor-relevant power exhaust (United States)

    Wischmeier, M.


    With increasing size of a tokamak device and associated fusion power gain an increasing power flux density towards the divertor needs to be handled. A solution for handling this power flux is crucial for a safe and economic operation. Using purely geometric arguments in an ITER-like divertor this power flux can be reduced by approximately a factor 100. Based on a conservative extrapolation of current technology for an integrated engineering approach to remove power deposited on plasma facing components a further reduction of the power flux density via volumetric processes in the plasma by up to a factor of 50 is required. Our current ability to interpret existing power exhaust scenarios using numerical transport codes is analyzed and an operational scenario as a potential solution for ITER like divertors under high density and highly radiating reactor-relevant conditions is presented. Alternative concepts for risk mitigation as well as strategies for moving forward are outlined.

  14. Light weight space power reactors for nuclear electric propulsion

    Energy Technology Data Exchange (ETDEWEB)

    Ludewig, H.; Mughabghab, S.; Lazareth, O.; Perkins, K.; Schmidt, E.; Powell, J.R.


    A Nuclear Electric Propulsion (NEP) unit capable of propelling a manned vehicle to MARS will be required to have a value of {alpha} (kg/kWe) which is less than five. In order to meet this goal the reactor mass, and thus its contribution to the value of {alpha} will have to be minimized. In this paper a candidate for such a reactor is described. It consists of a gas cooled Particle Bed Reactor (PBR), with specially chosen materials which allow it to operate at an exit temperature of approximately 2000 K. One of the unique features of a PBR is the direct cooling of particulate fuel by the working fluid. This feature allows for high power densities, highest possible gas exit temperatures, for a given fuel temperature and because of the thin particle bed a low pressure drop. The PBR's described in this paper will have a ceramic moderator (Be{sub 2}C), ZrC coated fuel particles and a carbon/carbon hot frit. All the reactors will be designed with sufficient fissile loading to operate at full power for seven years. The burn up possible with particulate fuel is approximately 30%--50%. These rector designs achieve a value of {alpha} less than unity in the power range of interest (5 MWe). 5 refs., 3 figs.

  15. Closed Brayton cycle power conversion systems for nuclear reactors :

    Energy Technology Data Exchange (ETDEWEB)

    Wright, Steven A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Lipinski, Ronald J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Vernon, Milton E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sanchez, Travis [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)


    This report describes the results of a Sandia National Laboratories internally funded research program to study the coupling of nuclear reactors to gas dynamic Brayton power conversion systems. The research focused on developing integrated dynamic system models, fabricating a 10-30 kWe closed loop Brayton cycle, and validating these models by operating the Brayton test-loop. The work tasks were performed in three major areas. First, the system equations and dynamic models for reactors and Closed Brayton Cycle (CBC) systems were developed and implemented in SIMULINKTM. Within this effort, both steady state and dynamic system models for all the components (turbines, compressors, reactors, ducting, alternators, heat exchangers, and space based radiators) were developed and assembled into complete systems for gas cooled reactors, liquid metal reactors, and electrically heated simulators. Various control modules that use proportional-integral-differential (PID) feedback loops for the reactor and the power-conversion shaft speed were also developed and implemented. The simulation code is called RPCSIM (Reactor Power and Control Simulator). In the second task an open cycle commercially available Capstone C30 micro-turbine power generator was modified to provide a small inexpensive closed Brayton cycle test loop called the Sandia Brayton test-Loop (SBL-30). The Capstone gas-turbine unit housing was modified to permit the attachment of an electrical heater and a water cooled chiller to form a closed loop. The Capstone turbine, compressor, and alternator were used without modification. The Capstone systems nominal operating point is 1150 K turbine inlet temperature at 96,000 rpm. The annular recuperator and portions of the Capstone control system (inverter) and starter system also were reused. The rotational speed of the turbo-machinery is controlled by adjusting the alternator load by using the electrical grid as the load bank. The SBL-30 test loop was operated at

  16. Delayed gamma power measurement for sodium-cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Coulon, R., E-mail: romain.coulon@cea.f [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Normand, S., E-mail: stephane.normand@cea.f [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Ban, G., E-mail: ban@lpccaen.in2p3.f [ENSICAEN, 6 Boulevard Marechal Juin, F-14050 Caen Cedex 4 (France); Barat, E.; Montagu, T.; Dautremer, T. [CEA, LIST, Laboratoire Modelisation Simulation et Systemes, F-91191 Gif-sur-Yvette (France); Brau, H.-P. [ICSM, Centre de Marcoule, BP 17171 F-30207 Bagnols sur Ceze (France); Dumarcher, V. [AREVA NP, SET, F-84500 Bollene (France); Michel, M.; Barbot, L.; Domenech, T.; Boudergui, K.; Bourbotte, J.-M. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Jousset, P. [CEA, LIST, Departement des Capteurs, du Signal et de l' Information, F-91191 Gif-sur-Yvette (France); Barouch, G.; Ravaux, S.; Carrel, F. [CEA, LIST, Laboratoire Modelisation Simulation et Systemes, F-91191 Gif-sur-Yvette (France); Saurel, N. [CEA, DAM, Laboratoire Mesure de Dechets et Expertise, F-21120 Is-sur-Tille (France); Frelin-Labalme, A.-M.; Hamrita, H. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France)


    Graphical abstract: Display Omitted Research highlights: {sup 20}F and {sup 23}Ne tagging agents are produced by fast neutron flux. {sup 20}F signal has been measured at the SFR Phenix prototype. A random error of only 3% for an integration time of 2 s could be achieved. {sup 20}F and {sup 23}Ne power measurement has a reduced temperature influence. Burn-up impact could be limited by simultaneous {sup 20}F and {sup 23}Ne measurement. - Abstract: Previous works on pressurized water reactors show that the nitrogen 16 activation product can be used to measure thermal power. Power monitoring using a more stable indicator than ex-core neutron measurements is required for operational sodium-cooled fast reactors, in order to improve their economic efficiency at the nominal operating point. The fluorine 20 and neon 23 produced by (n,{alpha}) and (n,p) capture in the sodium coolant have this type of convenient characteristic, suitable for power measurements with low build-up effects and a potentially limited temperature, flow rate, burn-up and breeding dependence. This method was tested for the first time during the final tests program of the French Phenix sodium-cooled fast reactor at CEA Marcoule, using the ADONIS gamma pulse analyzer. Despite a non-optimal experimental configuration for this application, the delayed gamma power measurement was pre-validated, and found to provide promising results.

  17. 77 FR 3009 - Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors (United States)


    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors..., ``Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors.''...

  18. Radioactive waste management and disposal scenario for fusion power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Tabara, Takashi; Yamano, Naoki [Sumitomo Atomic Energy Industries Ltd., Tokyo (Japan); Seki, Yasushi; Aoki, Isao


    The environmental and economic impact of radioactive waste (radwaste) generated from fusion power reactors using five types of structural materials and a light water reactor (LWR) have been evaluated and compared. At first, the amount and the radioactive level of the radwaste generated in five fusion reactors ware evaluated by an activation calculation code. Next, a possible radwaste disposal scenario applicable to fusion radwaste in Japan is considered and the disposal cost evaluated under certain assumptions. The exposure doses are evaluated for the skyshine of gamma-rays during the disposal operation, groundwater migration scenario during the institutional control period of 300 years and future site use scenario after the institutional period. The radwaste generated from a typical LWR was estimated based on a literature survey and the disposal cost was evaluated using the same assumptions as for the fusion reactors. It is found that the relative cost of disposal is strongly dependent on the cost for interim storage of medium level waste of fusion reactors and the cost of high level waste for the LWR. (author)

  19. Selection of power plant elements for future reactor space electric power systems

    Energy Technology Data Exchange (ETDEWEB)

    Buden, D.; Bennett, G.A.; Copper, K.


    Various types of reactor designs, electric power conversion equipment, and reject-heat systems to be used in nuclear reactor power plants for future space missions were studied. The designs included gas-cooled, liquid-cooled, and heat-pipe reactors. For the power converters, passive types such as thermoelectric and thermionic converters and dynamic types such as Brayton, potassium Rankine, and Stirling cycles were considered. For the radiators, heat pipes for transfer and radiating surface, pumped fluid for heat transfer with fins as the radiating surface, and pumped fluid for heat transfer with heat pipes as the radiating surface were considered. After careful consideration of weights, sizes, reliabilities, safety, and development cost and time, a heat-pipe reactor design, thermoelectric converters, and a heat-pipe radiator for an experimental program were selected.

  20. Diagnostic technology for degradation of feeder pipe and fuel channel; design of nuclear fuel channel mock-up and development of signature analysis technique for its condition monitoring system

    Energy Technology Data Exchange (ETDEWEB)

    Lee, S. K. [Inha University, Incheon (Korea)


    The main purpose of this project is to design the nuclear fuel channel and to develope the signal analysis method for the nonstationary signal. The design of mock-up is necessary for the pre-test of the nuclear plant. The fault signal due to the damage of the nuclear plant is generally nonstationary. It is difficult to analysis this non-stationary signal using traditional method. Thus in this research, time-frequency analysis, and wavelet transform are studied for the analysis of nonstationary signal. Basic program for analysis of nonstationary signal embedded in the background noise is developed. Results from this research can be applied to the early detection of damage of nuclear fuel channel. However, in order to apply to the real nuclear plant, further research project should be processed through the mock-up test. Then that result can be applied to the condition monitoring of a real nuclear plant. 10 refs., 18 figs., 2 tabs. (Author)

  1. Zero Power Levitation Control of Hybrid Electro-Magnetic Levitation System by Load Observer

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Youn Hyun; Lee, Ju [Hanyang University, Seoul (Korea)


    This paper introduces the scheme that improve the control performance of electromagnetic levitation system with zero power controller. Magnetic levitation is used widely, but the electromagnetic force has nonlinear characteristics because it is proportioned to a square of the magnetic flux density and it is in inverse proportion to a square of the air gap. So, it is complicate and difficult to control the electromagnetic force. Besides, it is more difficult to control if the equivalent gap is unknown in case of zero power control. Therefore, this paper proposed the hybrid electro-magnetic levitation control method in which the variable load is estimated by using a load observer and its system is controlled at a new zero power equilibrium air gap position. Also it is confirmed that the proposed control method improve the control performance through simulation and experiment. (author). 10 refs., 17 figs., 1 tab.

  2. Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hogerton, John


    This pamphlet describes how reactors work; discusses reactor design; describes research, teaching, and materials testing reactors; production reactors; reactors for electric power generation; reactors for supply heat; reactors for propulsion; reactors for space; reactor safety; and reactors of tomorrow. The appendix discusses characteristics of U.S. civilian power reactor concepts and lists some of the U.S. reactor power projects, with location, type, capacity, owner, and startup date.

  3. Characteristics of a reactor with power reactivity feedback

    Institute of Scientific and Technical Information of China (English)


    The point-reactor model with power reactivity feedback becomes a nonlinear system. Its dynamic characteristic shows great complexity. According to the mathematic definition of stability in differential equa-tion qualitative theory, the model of a reactor with power reactivity feedback is judged unstable. The equilib-rium point is a saddle-node point. A portion of the trajectory in the neighborhood of the equilibrium point is parabolic fan curve, and the other is hyperbolic fan curve. Based on phase locus near the equilibrium point, it is pointed out that the model is still stable within physical limits. The difference between stabilities in the mathematical sense and in the physical sense is indicated.

  4. The BR2 refurbishment programme: achievements and two years operation feedback

    Energy Technology Data Exchange (ETDEWEB)

    Gubel, P.; Dekeyser, J.; Koonen, E.; Van der Auwera, J. [SCK/CEN, Belgian Nuclear Research Center, Mol (Belgium)


    The BR2 reactor was shutdown end of June 1995 for an extensive refurbishment after more than 30 years utilization. The beryllium matrix needed to be replaced and the aluminium vessel inspected for an envisaged 15 year life extension. Other aspects of the refurbishment programme aimed at the reliability and availability of the installations, safety of operation and compliance with modern safety standards. The reactor was started again in' April '97 and operated only for three cycles in 1997. These first irradiation cycles were intended as a demonstration of the safety and reliability of all components and systems after refurbishment. Also during the extended shutdowns non-critical refurbishment tasks were allowed to be continued and finalized. At the request of the Safety Authorities, some modifications and studies are still in progress without perturbation of the reactor operation. (author)

  5. Reactor Power for Large Displacement Autonomous Underwater Vehicles

    Energy Technology Data Exchange (ETDEWEB)

    McClure, Patrick Ray [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Reid, Robert Stowers [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Poston, David Irvin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dasari, Venkateswara Rao [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)


    This is a PentaChart on reactor power for large displacement autonomous underwater vehicles. Currently AUVs use batteries or combinations of batteries and fuel cells for power. Battery/fuel cell technology is limited by duration. Batteries and cell fuels are a good match for some missions, but other missions could benefit greatly by a longer duration. The goal is the following: to design nuclear systems to power an AUV and meet design constraints including non-proliferation issues, power level, size constraints, and power conversion limitations. The action plan is to continue development of a range of systems for terrestrial systems and focus on a system for Titan Moon as alternative to Pu-238 for NASA.

  6. Summary of space nuclear reactor power systems, 1983--1992

    Energy Technology Data Exchange (ETDEWEB)

    Buden, D.


    This report summarizes major developments in the last ten years which have greatly expanded the space nuclear reactor power systems technology base. In the SP-100 program, after a competition between liquid-metal, gas-cooled, thermionic, and heat pipe reactors integrated with various combinations of thermoelectric thermionic, Brayton, Rankine, and Stirling energy conversion systems, three concepts:were selected for further evaluation. In 1985, the high-temperature (1,350 K), lithium-cooled reactor with thermoelectric conversion was selected for full scale development. Since then, significant progress has been achieved including the demonstration of a 7-y-life uranium nitride fuel pin. Progress on the lithium-cooled reactor with thermoelectrics has progressed from a concept, through a generic flight system design, to the design, development, and testing of specific components. Meanwhile, the USSR in 1987--88 orbited a new generation of nuclear power systems beyond the, thermoelectric plants on the RORSAT satellites. The US has continued to advance its own thermionic fuel element development, concentrating on a multicell fuel element configuration. Experimental work has demonstrated a single cell operating time of about 1 1/2-y. Technology advances have also been made in the Stirling engine; an advanced engine that operates at 1,050 K is ready for testing. Additional concepts have been studied and experiments have been performed on a variety of systems to meet changing needs; such as powers of tens-to-hundreds of megawatts and highly survivable systems of tens-of-kilowatts power.

  7. Sums of the Powers of the Zeros of the Mittag-Leffler Function (United States)

    Hanneken, John W.; Puzio, Raymond S.; Narahari Achar, B. N.


    The Mittag-Leffler function, which is a generalization of the exponential function, occurs naturally in the solution of dynamic problems involving fractional calculus. For example see the book Fractional Differential Equations by Podlubny, or for a specific application see the problem of the fractional oscillator by Achar et. al. Physica A297 (2001) 361 and A309 (2002) 275. The zeros of the Mittag-Leffler function play a significant role in the dynamic solutions. However, very little is known about these zeros. A summary of the available information about the zeros of Mittag-Leffler functions will be given and new results pertaining to the sums of the powers of the zeros will be presented.

  8. Gravity Scaling of a Power Reactor Water Shield (United States)

    Reid, Robert S.; Pearson, J. Boise


    Water based reactor shielding is being considered as an affordable option for use on initial lunar surface power systems. Heat dissipation in the shield from nuclear sources must be rejected by an auxiliary thermal hydraulic cooling system. The mechanism for transferring heat through the shield is natural convection between the core surface and an array of thermosyphon radiator elements. Natural convection in a 100 kWt lunar surface reactor shield design has been previously evaluated at lower power levels (Pearson, 2007). The current baseline assumes that 5.5 kW are dissipated in the water shield, the preponderance on the core surface, but with some volumetric heating in the naturally circulating water as well. This power is rejected by a radiator located above the shield with a surface temperature of 370 K. A similarity analysis on a water-based reactor shield is presented examining the effect of gravity on free convection between a radiation shield inner vessel and a radiation shield outer vessel boundaries. Two approaches established similarity: 1) direct scaling of Rayleigh number equates gravity-surface heat flux products, 2) temperature difference between the wall and thermal boundary layer held constant on Earth and the Moon. Nussult number for natural convection (laminar and turbulent) is assumed of form Nu = CRa(sup n). These combined results estimate similarity conditions under Earth and Lunar gravities. The influence of reduced gravity on the performance of thermosyphon heat pipes is also examined.

  9. Disk magnetohydrodynamic power conversion system for NERVA reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jackson, W.D. (HMJ Corporation. 10400 Connecticut Ave., Kensington, Maryland 20895 (United States)); Bernard, F.E. (Westinghouse Corp., P.O. Box 355, Pittsburgh, Pennsylvania 15230 (United States)); Holman, R.R. (HMJ Corporation, 10400 Connecticut Ave., Kensington, Maryland 20895 (United States)); Maxwell, C.D. (STD Research Corp., P.O. Box C, Arcadia, California 91006 (United States)); Seikel, G.R. (SeiTec, Inc., P.O. Box 81264, Cleveland, Ohio 44181 (United States))


    The combination of a magnetohydrodynamic (MHD) generator of the disk type with a NERVA reactor yields an advanced power system particularly suited to space applications with the capability of producing up to gigawatt pulses and multi-megawatt continuous operation. Several unique features result from the combination of this type of reactor and a disk MHD generator in which hydrogen serves as the plasma working fluid. Cesium seedings is utilized under conditions which enable the generator to operate stably in the non-equilibrium electrical conduction mode. In common with all practical MHD generators, the disk output is DC and voltages in the range 20--100 kV are attainable. This leads to a simplification of the power conditioning system and a major reduction in specific mass. Taken together with the high performance capabilities of the NERVA reactor, the result is an attractively low overall system specific mass. Further, the use of non-equilibrium ionization enables system specific enthalpy extractions in excess of 40% to be attained. This paper reports the results of a study to establish the basis for the design of a cesium seeded hydrogen MHD disk generator. Generator performance results are presented in terms of a stability factor which is related to cesium seeded hydrogen plasma behavior. It is shown that application of the results already obtained with cesium seeded noble gases (argon and helium) to the case of hydrogen as the working fluid in a disk MHD generator enables a high performance power system to be defined.

  10. Reactors

    CERN Document Server

    International Electrotechnical Commission. Geneva


    This standard applies to the following types of reactors: shunt reactors, current-limiting reactors including neutral-earthing reactors, damping reactors, tuning (filter) reactors, earthing transformers (neutral couplers), arc-suppression reactors, smoothing reactors, with the exception of the following reactors: small reactors with a rating generally less than 2 kvar single-phase and 10 kvar three-phase, reactors for special purposes such as high-frequency line traps or reactors mounted on rolling stock.

  11. On the Reaction Mechanism of Br2 with OCS

    Institute of Scientific and Technical Information of China (English)

    Hai Tao YU; Hua ZHONG; Ming Xia LI; Hong Gang FU; Jia Zhong SUN


    The reaction mechanism of photochemical reaction between Br2 ( 1 ∑ ) and OCS ( 1 ∑ ) is predicted by means of theoretical methods. The calculated results indicate that the direct addition of Br2 to the CS bond of OCS molecule is more favorable in energy than the direct addition of Br2to the CO bond. Furthermore, the intermediate isomer syn-BrC(O)SBr is more stable thermodynamically and kinetically than anti-BrC(O)SBr. The original resultant anti-BrC(O)SBr formed in the most favorable reaction channel can easily isomerize into the final product syn-BrC(O)SBr with only 31.72 kJ/mol reaction barrier height. The suggested mechanism is in good agreement with previous experimental study.

  12. Power regulating range broadening of the WWER-type reactor power units

    Energy Technology Data Exchange (ETDEWEB)

    Dement' ev, B.A.; Petrov, V.A.; Proskuryakov, A.G.; Puchkov, V.V. (Moskovskij Ehnergeticheskij Inst. (USSR))


    Calculational studies on the use of sliding pressure (SP) regime to expand the regulating range of the WWER-440 reactor power units are presented. Two operation regimes of a power unit have been considered: according to weekly and daily load swings in electrical grids. The conclusion is made that the use of SP regime in the secondary circuit improves manoeuvable characterstics of the power unit in the second half of operating cycle. T of the reactor (0.6 power regulating range broadening of the reactor. Besides, the use of SP regime during power unit operation with decreased loadings is the more efficient the smaller is the load. The range of operating cycle 0.8 <= T <= 1 makes the greatest contribution to regulating range broadening as a result of SP regime use. Conclusions of the calculational studies can be also applied to WWER reactors of other types as well as to RBMK reactors.

  13. Supercritical Water Reactor Cycle for Medium Power Applications

    Energy Technology Data Exchange (ETDEWEB)

    BD Middleton; J Buongiorno


    Scoping studies for a power conversion system based on a direct-cycle supercritical water reactor have been conducted. The electric power range of interest is 5-30 MWe with a design point of 20 MWe. The overall design objective is to develop a system that has minimized physical size and performs satisfactorily over a broad range of operating conditions. The design constraints are as follows: Net cycle thermal efficiency {ge}20%; Steam turbine outlet quality {ge}90%; and Pumping power {le}2500 kW (at nominal conditions). Three basic cycle configurations were analyzed. Listed in order of increased plant complexity, they are: (1) Simple supercritical Rankine cycle; (2) All-supercritical Brayton cycle; and (3) Supercritical Rankine cycle with feedwater preheating. The sensitivity of these three configurations to various parameters, such as reactor exit temperature, reactor pressure, condenser pressure, etc., was assessed. The Thermoflex software package was used for this task. The results are as follows: (a) The simple supercritical Rankine cycle offers the greatest hardware simplification, but its high reactor temperature rise and reactor outlet temperature may pose serious problems from the viewpoint of thermal stresses, stability and materials in the core. (b) The all-supercritical Brayton cycle is not a contender, due to its poor thermal efficiency. (c) The supercritical Rankine cycle with feedwater preheating affords acceptable thermal efficiency with lower reactor temperature rise and outlet temperature. (d) The use of a moisture separator improves the performance of the supercritical Rankine cycle with feedwater preheating and allows for a further reduction of the reactor outlet temperature, thus it was selected for the next step. Preliminary engineering design of the supercritical Rankine cycle with feedwater preheating and moisture separation was performed. All major components including the turbine, feedwater heater, feedwater pump, condenser, condenser pump

  14. Gas separation membranes for zero-emission fossil power plants: MEM-BRAIN

    NARCIS (Netherlands)

    Czyperek, M.; Zapp, P.; Bouwmeester, H.J.M.; Modigell, M.; Ebert, K.; Voigt, I.; Meulenberg, W.A.; Singheiser, L.; Stöver, D.


    The objective of the “MEM-BRAIN” project is the development and integration of ceramic and polymeric gas separation membranes for zero-emission fossil power plants. This will be achieved using membranes with a high permeability and selectivity for either CO2, O2 or H2, for the three CO2 capture proc

  15. MEM-BRAIN gas separation membranes for zero-emission fossil power plants

    NARCIS (Netherlands)

    Czyperek, M.; Zapp, P.; Bouwmeester, H.J.M.; Modigell, M.; Peinemann, K.-V.; Voigt, I.; Meulenberg, W.A.; Singheiser, L.; Stöver, D.


    The aim of the MEM-BRAIN project is the development and integration of gas separation membranes for zero-emission fossil power plants. This will be achieved by selective membranes with high permeability for CO2, O2 or H2, so that high-purity CO2 is obtained in a readily condensable form. The project

  16. PR-EDB: Power Reactor Embrittlement Database - Version 3

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [ORNL; Subramani, Ranjit [ORNL


    The aging and degradation of light-water reactor pressure vessels is of particular concern because of their relevance to plant integrity and the magnitude of the expected irradiation embrittlement. The radiation embrittlement of reactor pressure vessel materials depends on many factors, such as neutron fluence, flux, and energy spectrum, irradiation temperature, and preirradiation material history and chemical compositions. These factors must be considered to reliably predict pressure vessel embrittlement and to ensure the safe operation of the reactor. Large amounts of data from surveillance capsules are needed to develop a generally applicable damage prediction model that can be used for industry standards and regulatory guides. Furthermore, the investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current codes, Standard Review Plans (SRPs), and Guides for license renewal can be greatly expedited by the use of a well-designed computerized database. The Power Reactor Embrittlement Database (PR-EDB) is such a comprehensive collection of data for U.S. designed commercial nuclear reactors. The current version of the PR-EDB lists the test results of 104 heat-affected-zone (HAZ) materials, 115 weld materials, and 141 base materials, including 103 plates, 35 forgings, and 3 correlation monitor materials that were irradiated in 321 capsules from 106 commercial power reactors. The data files are given in dBASE format and can be accessed with any personal computer using the Windows operating system. "User-friendly" utility programs have been written to investigate radiation embrittlement using this database. Utility programs allow the user to retrieve, select and manipulate specific data, display data to the screen or printer, and fit and plot Charpy impact data. The PR-EDB Version 3.0 upgrades Version 2.0. The package was developed based on the Microsoft .NET framework technology and uses Microsoft Access for

  17. 81Br NQR for Uncoordinated Br ions in trans-[CoBr2(en)2][H5O2]Br2 and trans-[CoBr2(en)2] [D5O2]Br2 (United States)

    Honda, H.; Sasane, A.; Miyagi, K.; Ishikawa, A.; Mori, Y.


    The temperature dependence of the 81Br NQR frequencies (vD) for uncoordinated Br- ions in trans-[CoBr2(en)2] [D5O2]Br2(D) has been determined by a continuous-wave spectrometer. vD amounted to 16.200 MHz at 273 K. This is lower by 418 kHz than the 81Br NQR frequency (vH) for trans-[CoBr2 (en2] [H5O2 ]Br2 (H). The frequency difference (Δv = vH -vD) remained almost constant in the temperature range studied. A shortening of the O-H bond length caused by deuteration could explain the magnitude and the sign of Av on the basis of a point charge model calculation. The compounds D and H yielded 81Br NQR lines in the range 110-320 K and 90-343 K, respectively. As to the 59Co NQR frequencies (7/2 - 5/2), the observed isotope frequency shifts (Δv1 = v1H - v1D) between D and H were smaller than 5 kHz. Below 160 K, 59Co resonances were only available by pulsed experiments. 59Co NQR spin-lattice relaxation times T1Q of 0.54 ms at 194 K and 4.8 s at 77 K for H have been observed.

  18. Small space reactor power systems for unmanned solar system exploration missions

    Energy Technology Data Exchange (ETDEWEB)

    Bloomfield, H.S.


    A preliminary feasibility study of the application of small nuclear reactor space power systems to the Mariner Mark II Cassini spacecraft/mission was conducted. The purpose of the study was to identify and assess the technology and performance issues associated with the reactor power system/spacecraft/mission integration. The Cassini mission was selected because study of the Saturn system was identified as a high priority outer planet exploration objective. Reactor power systems applied to this mission were evaluated for two different uses. First, a very small 1 kWe reactor power system was used as an RTG replacement for the nominal spacecraft mission science payload power requirements while still retaining the spacecraft's usual bipropellant chemical propulsion system. The second use of reactor power involved the additional replacement of the chemical propulsion system with a small reactor power system and an electric propulsion system. The study also provides an examination of potential applications for the additional power available for scientific data collection. The reactor power system characteristics utilized in the study were based on a parametric mass model that was developed specifically for these low power applications. The model was generated following a neutronic safety and operational feasibility assessment of six small reactor concepts solicited from U.S. industry. This assessment provided the validation of reactor safety for all mission phases and generatad the reactor mass and dimensional data needed for the system mass model.


    Energy Technology Data Exchange (ETDEWEB)



    The APPR-I is described and the various hazards are reviewed. Because of the reactor's location near the nation's Capitol, containment is of the utmost importance. The maximum energy release in any possible accident is 7.4 million Btu's which is completely contained within a 7/8 inch thick steel cylindrical shell with hemispherical ends. The vapor container is 60 ft high and 32 ft in diameter and is lined on the inside with 2 ft of reinforced concrete which provides missile protection and is part of the secondary shield. All possible nuclear excursions are reviewed and the energy from any of these is insignificant compared to the stored energy in the water. The maximum credible accident is caused hy the reactor running constantly at its maximum power of 10 Mw and through an extremely unlikely sequence of failures, causing the temperature of the water in the primary and secondary systeras to rise to saturation; whereupon a rupture occurs releasing the stored energy of 7.4 million Btu's into the vapor container. If the reactor core melts during the incident, a maximum of 10/sup 8/ curies of activity is released. While it appears impossible for a rupture of the vapor container to oecur except by sabotage or bombing, the hazards to the surrounding area are discussed in the event of such a rupture occurring simultaneously with the maximum credible accident. (auth)

  20. Utilization of Minor Actinides (Np, Am, Cm) in Nuclear Power Reactor (United States)

    Gerasimov, A.; Bergelson, B.; Tikhomirov, G.


    Calculation research of the utilization process of minor actinides (transmutation with use of power released) is performed for specialized power reactor of the VVER type operating on the level of electric power of 1000 MW. Five subsequent cycles are considered for the reactor with fuel elements containing minor actinides along with enriched uranium. It was shown that one specialized reactor for the one cycle (900 days) can utilize minor actinides from several VVER-1000 reactors without any technological and structural modifications. Power released because of minor actinide fission is about 4% with respect to the total power

  1. Wind power systems for zero net energy housing in the United States

    Energy Technology Data Exchange (ETDEWEB)

    Elkinton, Melissa R.; McGowan, Jon G.; Manwell, James F. [Renewable Energy Research Laboratory, Department of Mechanical and Industrial Engineering, University of Massachusetts, Amherst, MA 01003 (United States)


    This work investigates the feasibility of renewable energy housing development in the U.S. using wind power and solar thermal systems to attain zero net energy consumption. The over all objective was to determine how the wind power and solar thermal system designs and economics differ with various climates, wind and solar resources, energy prices, and state incentives, such as net-metering. Five U.S. cities, one in each of the five climate zones, were selected for this study based on their potential for wind power. A zero net energy housing design tool was developed in order to analyze and compare various system designs. The energy performance and economics of the designs were compared for various sizes of housing development, for seven turbine models, and selected heating systems. The results suggest that while there are some economical options for wind powered zero net energy housing developments, they are generally more expensive (except in the warmest climate zone) than housing with natural gas heating. In all of the cases, the economies of scale for large-scale wind turbines gave more of an economic advantage than net-metering programs gave small- and medium-scale wind turbines. (author)

  2. Robust reactor power control system design by genetic algorithm

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yoon Joon; Cho, Kyung Ho; Kim, Sin [Cheju National University, Cheju (Korea, Republic of)


    The H{sub {infinity}} robust controller for the reactor power control system is designed by use of the mixed weight sensitivity. The system is configured into the typical two-port model with which the weight functions are augmented. Since the solution depends on the weighting functions and the problem is of nonconvex, the genetic algorithm is used to determine the weighting functions. The cost function applied in the genetic algorithm permits the direct control of the power tracking performances. In addition, the actual operating constraints such as rod velocity and acceleration can be treated as design parameters. Compared with the conventional approach, the controller designed by the genetic algorithm results in the better performances with the realistic constraints. Also, it is found that the genetic algorithm could be used as an effective tool in the robust design. 4 refs., 6 figs. (Author)

  3. Numerical study of high-power semiconductor lasers for operation at sub-zero temperatures (United States)

    Hasler, K. H.; Frevert, C.; Crump, P.; Erbert, G.; Wenzel, H.


    We present results on the impact of the Al-content in the waveguide structure on the electro-optical characteristics of 9xx nm, GaAs-based high-power lasers operated at room (300 K) and at sub-zero (200 K) heat sink temperatures. Experimentally a strong improvement of conversion efficiency and output power has been found if the lasers are cooled down. Numerical simulations using a software tool which solves the thermo-dynamic based drift-diffusion equations are able to reproduce the experimental findings. The reasons for the improved performance at lower temperatures are the enhancement of the modal gain and the reduced accumulation of electrons in the p-confinement layers resulting in a reduction of the leakage current. The latter allows the realization of lasers with a reduced Al content having a smaller series resistance and thus further enlarged conversion efficiency at sub-zero temperatures.

  4. Oak Ridge Tokamak experimental power reactor study scoping report

    Energy Technology Data Exchange (ETDEWEB)

    Roberts, M.


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

  5. Review of the status of low power research reactors and considerations for its development

    Energy Technology Data Exchange (ETDEWEB)

    Lim, In Cheol; Wu, Sang Ik; Lee, Byung Chul; Ha, Jae Joo [KAERI, Daejeon (Korea, Republic of)


    At present, 232 research reactors in the world are in operation and two thirds of them have a power less than 1 MW. Many countries have used research reactors as the tools for educating and training students or engineers and for scientific service such as neutron activation analysis. As the introduction of a research reactor is considered a stepping stone for a nuclear power development program, many newcomers are considering having a low power research reactor. The IAEA has continued to provide forums for the exchange of information and experiences regarding low power research reactors. Considering these, the Agency is recently working on the preparation of a guide for the preparation of technical specification possibly for a member state to use when wanting to purchase a low power research reactor. In addition, ANS has stated that special consideration should be given to the continued national support to maintain and expand research and test reactor programs and to the efforts in identifying and addressing the future needs by working toward the development and deployment of next generation nuclear research and training facilities. Thus, more interest will be given to low power research reactors and its role as a facility for education and training. Considering these, the status of low power research reactors was reviewed, and some aspects to be considered in developing a low power research reactor were studied.

  6. Refurbishment of BR2 (Phase 4 and 5)[1997 Scientific Report of the Belgian Nuclear Research Centre

    Energy Technology Data Exchange (ETDEWEB)

    Gubel, P.; Dekeyser, J.; Van der Auwera, J.


    The extensive refurbishment of the BR-2 materials testing reactor should allow another 10 to 15years of continued operation. The refurbishment programme is required in order to comply with modern safety standards, to enhance the reliability of operation, and to compensate for the ageing of the installations of a facility that has reached about 35 years of intensive service. The main objectives and achievements of phase 4 and 5 are described.

  7. Compensation by RGMS for misreading reactor power in case of D2O dilution

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Sang Hoon; Park, Jae Yoon; Choi, Young San; Kim, Young Ki [KAERI, Daejeon (Korea, Republic of)


    In a research reactor Neutron Measurement System (NMS) which uses wide range fission chamber as neutron detector is applied to measure the reactor power. This system has rapid response to power and stable accuracy for wide range. But this has some concerns of relative measured values depending on the installed location of neutron detector and also may cause the loss of accuracy when dilution of heavy water in the D2O tank happens. The NMS is not only used for reactor control and but also used for reactor protection system. Accordingly faulted reactor power with high deviation for second case may lead unexpected increase of the reactor power. In order to prevent this occurrence, Reactor Gamma Measurement System (RGMS) is necessarily applied. Herein the structure, measuring method and application of RGMS will be introduced.

  8. New applications of neutron noise theory in power reactor physics

    Energy Technology Data Exchange (ETDEWEB)

    Arzhanov, Vasiliy


    The present thesis deals with neutron noise theory as applied to three comparatively different topics (or problems) in power reactor physics. Namely they are: theoretical investigation of the possibility to use a newly proposed current-flux (C/F) detector in Pressurized Water Reactors (PWRs) for the localisation of anomalies; both definition and studies on the point kinetic and adiabatic approximations for the relatively recently proposed Accelerator Driven Systems (ADS); development of the general theory of linear reactor kinetics and neutron noise in systems with varying size. One important practical problem is to detect and localise a vibrating control rod pin. The significance comes from the operational experience which indicates that individual pins can execute excessive mechanical vibrations that may lead to damage. Such mechanical vibrations induce neutron noise that can be detected. While the detection is relatively easy, the localisation of a vibrating control rod is much more complicated because only one measuring position is available and one needs to have at least three measured quantities. Therefore it has currently been proposed that the fluctuations of the neutron current vector, called the current noise, can be used in addition to the scalar noise in reactor diagnostic problems. The thesis investigates the possibility of the localization of a vibrating control rod pin in a PWR control assembly by using the scalar neutron noise and the 2-D radial current noise as measured at one central point in the control assembly. An explicit localisation technique is elaborated in which the searched position is determined as the absolute minimum of a minimisation function. The technique is investigated in numerical simulations. The results of the simulation tests show the potential applicability of the method. By design accelerator-driven systems would operate in a subcritical mode with a strong external source. This calls for a revision of many concepts and

  9. 77 FR 38338 - Dairyland Power Cooperative; La Crosse Boiling Water Reactor Exemption From Certain Security... (United States)


    ... COMMISSION Dairyland Power Cooperative; La Crosse Boiling Water Reactor Exemption From Certain Security Requirements 1.0 Background The La Crosse Boiling Water Reactor (LACBWR) is owned and was operated by the Dairyland Power Cooperative (DPC). The LACBWR was a nuclear power plant of nominal 50 Mw electrical...


    Directory of Open Access Journals (Sweden)

    Ashit Chakraborty


    Full Text Available Measurement error is the difference between the true value and the measured value of a quantity that exists in practice and may considerably affect the performance of control charts in some cases. Measurement error variability has uncertainty which can be from several sources. In this paper, we have studied the effect of these sources of variability on the power characteristics of control chart and obtained the values of average run length (ARL for zero-truncated Poisson distribution (ZTPD. Expression of the power of control chart for variable sample size under standardized normal variate for ZTPD is also derived.

  11. Measurements of reaction rate ratios as indexes of breeding performance in mock-up cores of FCA simulating metallic-fueled LMFBR and MOX-fueled LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Sakurai, Takeshi; Nemoto, Tatsuo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kobayashi, Keiji; Unesaki, Hironobu


    Measurements and analysis for reaction rate ratio of {sup 238}U capture to {sup 239}Pu fission (C8/F9) and that of {sup 238}U capture to {sup 235}U fission (C8/F5), which are important as indexes of breeding behavior of a fast breeder reactor, were made at Fast Critical Assembly (FCA) of Japan Atomic Energy Research Institute. The measurements were made by a foil activation technique at two assemblies simulating metallic-fueled liquid metal fast breeder reactor and an assembly simulating a MOX-fueled one. The analysis was made based on JENDL3.2 nuclear data file. Calculation to experiment ratios of C8/F9 and C8/F5 were 0.99-1.02 and 1.0-1.03 respectively. Furthermore, the {sup 238}U capture rate and the {sup 235}U fission rate were measured in a thermal neutron standard field in Heavy Water Facility of Kyoto University Research Reactor to verify accuracy of the present foil activation technique itself and to confirm these C/E values. (author)

  12. High-intensity power-resolved radiation imaging of an operational nuclear reactor (United States)

    Beaumont, Jonathan S.; Mellor, Matthew P.; Villa, Mario; Joyce, Malcolm J.


    Knowledge of the neutron distribution in a nuclear reactor is necessary to ensure the safe and efficient burnup of reactor fuel. Currently these measurements are performed by in-core systems in what are extremely hostile environments and in most reactor accident scenarios it is likely that these systems would be damaged. Here we present a compact and portable radiation imaging system with the ability to image high-intensity fast-neutron and gamma-ray fields simultaneously. This system has been deployed to image radiation fields emitted during the operation of a TRIGA test reactor allowing a spatial visualization of the internal reactor conditions to be obtained. The imaged flux in each case is found to scale linearly with reactor power indicating that this method may be used for power-resolved reactor monitoring and for the assay of ongoing nuclear criticalities in damaged nuclear reactors. PMID:26450669

  13. CARDIOGRAMA: a stochastic, semi-empirical methodology for power-reactor surveillance and diagnostics

    Energy Technology Data Exchange (ETDEWEB)

    March-Leuba, J.A.; deSaussure, G.; Perez, R.B.


    The utilization of stochastic methods (reactor noise) for power reactor diagnostics and surveillance applications is by now a relatively well-established technique. In this technique, the power spectral density (PSD) of the fluctuations of a specified state variable is often used to define the reactor's signature at a given configuration. The purpose of the present work is to address the problem of handling efficiently the substantial amount of information involved in the application of reactor surveillance and diagnostics methods. Specifically, a methodology is described for: (a) representing the PSDs parametrically, and (b) detecting changes from the reactor's baseline PSD (normal signature).

  14. Phase Diagram of the System KCI-CsBr-CaBr2%KCI-CsBr-CaBr2相图的研究

    Institute of Scientific and Technical Information of China (English)

    包新华; 陈华莉; 陆文聪; 阎立诚


    A computer prediction for the regularities of formation of binary bromide system was made. The results indicate that CsBr-CaBr2 molten salt system is binary complex compound-forming system. The phase diagram of CsBr-CaBr2 system has been determined by visual polythermal method in order to prove the prediction. The phase diagrams of the quasibinary system KCl-CsBr, KC1-CaBr2 and the quasiternary system CsBr-CaBr2-KC1 were determined by using visual polythermal method. The phase diagrams of CsBr-CaBr2 and KCl-CaBr2 show that the intermediate compound is formed. The results are in agreement with the computer prediction. Here, the various parts of the phase diagram of the quasiternary system CsBr-CaBr2-KCl were discussed.

  15. Phase Diagram of the System KCI-CsBr-CaBr2%KCI-CsBr-CaBr2相图的研究

    Institute of Scientific and Technical Information of China (English)

    包新华; 陈华莉; 陆文聪; 阎立诚


    A computer prediction for the regularities of formation of binary bromide system was made. The results indicate that CsBr-CaBr2 molten salt system is binary complex compound-forming system. The phase diagram of CsBr-CaBr2 system has been determined by visual polythermal method in order to prove the prediction. The phase diagrams of the quasibinary system KCl-CsBr, KC1-CaBr2 and the quasiternary system CsBr-CaBr2-KC1 were determined by using visual polythermal method. The phase diagrams of CsBr-CaBr2 and KCl-CaBr2 show that the intermediate compound is formed. The results are in agreement with the computer prediction. Here, the various parts of the phase diagram of the quasiternary system CsBr-CaBr2-KCl were discussed.

  16. Power conversion systems based on Brayton cycles for fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Linares, J.I., E-mail: [Rafael Marino Chair on New Energy Technologies. Comillas Pontifical University, Alberto Aguilera, 25-28015 Madrid (Spain); Herranz, L.E. [Unit of Nuclear Safety Research. CIEMAT, Madrid (Spain); Moratilla, B.Y.; Serrano, I.P. [Rafael Marino Chair on New Energy Technologies. Comillas Pontifical University, Alberto Aguilera, 25-28015 Madrid (Spain)


    This paper investigates Brayton power cycles for fusion reactors. Two working fluids have been explored: helium in classical configurations and CO{sub 2} in recompression layouts (Feher cycle). Typical recuperator arrangements in both cycles have been strongly constrained by low temperature of some of the energy thermal sources from the reactor. This limitation has been overcome in two ways: with a combined architecture and with dual cycles. Combined architecture couples the Brayton cycle with a Rankine one capable of taking advantage of the thermal energy content of the working fluid after exiting the turbine stage (iso-butane and steam fitted best the conditions of the He and CO{sub 2} cycles, respectively). Dual cycles set a specific Rankine cycle to exploit the lowest quality thermal energy source, allowing usual recuperator arrangements in the Brayton cycle. The results of the analyses indicate that dual cycles could reach thermal efficiencies around 42.8% when using helium, whereas thermal performance might be even better (46.7%), if a combined CO{sub 2}-H{sub 2}O cycle was set.

  17. Preliminary Development of Thermal Power Calculation Code H-Power for a Supercritical Water Reactor

    Directory of Open Access Journals (Sweden)

    Fan Zhang


    Full Text Available SCWR (Supercritical Water Reactor is one of the promising Generation IV nuclear systems, which has higher thermal power efficiency than current pressurized water reactor. It is necessary to perform the thermal equilibrium and thermal power calculation for the conceptual design and further monitoring and calibration of the SCWR. One visual software named H-Power was developed to calculate thermal power and its uncertainty of SCWR, in which the advanced IAPWS-IF97 industrial formulation was used to calculate the thermodynamic properties of water and steam. The ISO-5167-4: 2003 standard was incorporated in the code as the basis of orifice plate to compute the flow rate. New heat balance model and uncertainty estimate have also been included in the code. In order to validate H-Power, an assessment was carried out by using data published by US and Qinshan Phase II. The results showed that H-Power was able to estimate the thermal power of SCWR.

  18. Non-Power Reactor Operator Licensing Examiner Standards. Revision 1

    Energy Technology Data Exchange (ETDEWEB)



    The Non-Power Reactor Operator Licensing Examiner Standards provide policy and guidance to NRC examiners and establish the procedures and practices for examining and licensing of applicants for NRC operator licenses pursuant to Part 55 of Title 10 of the Code of Federal Regulations (10 CFR 55). They are intended to assist NRC examiners and facility licensees to understand the examination process better and to provide for equitable and consistent administration of examinations to all applicants by NRC examiners. These standards are not a substitute for the operator licensing regulations and are subject to revision or other internal operator examination licensing policy changes. As appropriate, these standards will be revised periodically to accommodate comments and reflect new information or experience.

  19. Radioactivity effects of Pb-17Li in fusion power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Casini, G.; Rocco, P. (Commission of the European Communities, Joint Research Centre, Ispra (Italy)); Zucchetti, M. (Dipt. di Energetica, Politecnico Turin (Italy))


    Research on the eutectic Pb-17Li is part of the blanket studies carried out in Europe for fusion power reactors. The use of this breeder makes easier some safety problems as compared to the case of lithium as a consequence of the lower chemical reactivity of Pb-17Li. On the other hand, it increases the radioactivity problems due to the neutron activation of lead and impurities. This paper presents both short-term (accidents) and long-term (waste disposal and recycling) aspects of the Pb-17Li activation products. They include the production, mobilization, release and environmental impact. Concerning accidents, a particular attention is given to Po-210 and Hg-203. Questions related to waste management are also revised. The most attractive solution seems that of recycling the spent Pb-17Li. This will be possible about 20 y after removal from service. As an alternative to recycling, the breeder disposal as radioactive waste is discussed. (orig.).

  20. A Basic LEGO Reactor Design for the Provision of Lunar Surface Power

    Energy Technology Data Exchange (ETDEWEB)

    John Darrell Bess


    A final design has been established for a basic Lunar Evolutionary Growth-Optimized (LEGO) Reactor using current and near-term technologies. The LEGO Reactor is a modular, fast-fission, heatpipe-cooled, clustered-reactor system for lunar-surface power generation. The reactor is divided into subcritical units that can be safely launched with lunar shipments from Earth, and then emplaced directly into holes drilled into the lunar regolith to form a critical reactor assembly. The regolith would not just provide radiation shielding, but serve as neutron-reflector material as well. The reactor subunits are to be manufactured using proven and tested materials for use in radiation environments, such as uranium-dioxide fuel, stainless-steel cladding and structural support, and liquid-sodium heatpipes. The LEGO Reactor system promotes reliability, safety, and ease of manufacture and testing at the cost of an increase in launch mass per overall rated power level and a reduction in neutron economy when compared to a single-reactor system. A single unshielded LEGO Reactor subunit has an estimated mass of approximately 448 kg and provides approximately 5 kWe. The overall envelope for a single subunit with fully extended radiator panels has a height of 8.77 m and a diameter of 0.50 m. Six subunits could provide sufficient power generation throughout the initial stages of establishing a lunar outpost. Portions of the reactor may be neutronically decoupled to allow for reduced power production during unmanned periods of base operations. During later stages of lunar-base development, additional subunits may be emplaced and coupled into the existing LEGO Reactor network, subject to lunar base power demand. Improvements in reactor control methods, fuel form and matrix, shielding, as well as power conversion and heat rejection techniques can help generate an even more competitive LEGO Reactor design. Further modifications in the design could provide power generative opportunities for

  1. Power monitoring in space nuclear reactors using silicon carbide radiation detectors (United States)

    Ruddy, Frank H.; Patel, Jagdish U.; Williams, John G.


    Space reactor power monitors based on silicon carbide (SiC) semiconductor neutron detectors are proposed. Detection of fast leakage neutrons using SiC detectors in ex-core locations could be used to determine reactor power: Neutron fluxes, gamma-ray dose rates and ambient temperatures have been calculated as a function of distance from the reactor core, and the feasibility of power monitoring with SiC detectors has been evaluated at several ex-core locations. Arrays of SiC diodes can be configured to provide the required count rates to monitor reactor power from startup to full power Due to their resistance to temperature and the effects of neutron and gamma-ray exposure, SiC detectors can be expected to provide power monitoring information for the fill mission of a space reactor.

  2. Outline of the safety research results, in the power reactor field, fiscal year 1996

    Energy Technology Data Exchange (ETDEWEB)



    The Power Reactor and Nuclear Fuel Development Corporation (PNC) has promoted the safety research in fiscal year of 1996 according to the Fundamental Research on Safety Research (fiscal year 1996 to 2000) prepared on March, 1996. Here is described on the research results in fiscal year 1996, the first year of the 5 years programme, and whole outline of the fundamental research on safety research, on the power reactor field (whole problems on the new nuclear converter and the fast breeder reactor field and problems relating to the power reactor in the safety for earthquake and probability theoretical safety evaluation field). (G.K.)

  3. Monitoring the Thermal Power of Nuclear Reactors with a Prototype Cubic Meter Antineutrino Detector

    CERN Document Server

    Bernstein, A; Misner, A; Palmer, T


    In this paper, we estimate how quickly and how precisely a reactor's operational status and thermal power can be monitored over hour to month time scales, using the antineutrino rate as measured by a cubic meter scale detector. Our results are obtained from a detector we have deployed and operated at 25 meter standoff from a reactor core. This prototype can detect a prompt reactor shutdown within five hours, and monitor relative thermal power to three percent within seven days. Monitoring of short-term power changes in this way may be useful in the context of International Atomic Energy Agency's (IAEA) Reactor Safeguards Regime, or other cooperative monitoring regimes.

  4. Low-Drift Flow Sensor with Zero-Offset Thermopile-Based Power Feedback

    CERN Document Server

    Dijkstra, M; De Boer, Meint; Berenschot, J W; Wiegerink, Remco; Elwenspoek, M


    A thermal flow sensor has been realised consisting of freely-suspended silicon-rich silicon-nitride microchannels with an integrated Al/poly-Si++ thermopile in combination with up- and downstream Al heater resistors. The inherently zero offset of the thermopile is exploited in a feedback loop controlling the dissipated power in the heater resistors, eliminating inevitable influences of resistance drift and mismatch of the thin-film metal resistors. The control system cancels the flow-induced temperature difference across the thermopile by controlling a power difference between both heater resistors, thereby giving a measure for the flow rate. The flow sensor was characterised for power difference versus water flow rates up to 1.5 ul/min, being in good agreement with a thermal model of the sensor, and the correct low-drift operation of the temperature-balancing control system has been verified.

  5. A Discharge-Excited SrBr2 Vapour Laser

    Institute of Scientific and Technical Information of China (English)

    潘佰良; 姚志欣; 陈钢


    A new-style discharge tube for a metal vapour laser has been designed and built. SrBr2 was successfullyused to replace the metal strontium as a working medium. Multi-line laser oscillations from resonance tometastable transition of strontium atoms (6.45um), ions (1.03um/1.O9um) and from strontium ion recombi-nation (416.2nm/430.5nm) have been obtained through longitudinal pulsed discharge. The problem of an in-compatibility reaction between metallic strontium and the discharge tube in the strontium vapour laser has beensolved. Some proposals are presented for further developments of strontium halide lasers.

  6. Development of a Robust Tri-Carbide Fueled Reactor for Multimegawatt Space Power and Propulsion Applications

    Energy Technology Data Exchange (ETDEWEB)

    Samim Anghaie; Travis W. Knight; Johann Plancher; Reza Gouw


    An innovative reactor core design based on advanced, mixed carbide fuels was analyzed for nuclear space power applications. Solid solution, mixed carbide fuels such as (U,Zr,Nb)c and (U,Zr, Ta)C offer great promise as an advanced high temperature fuel for space power reactors.

  7. 10 CFR 50.44 - Combustible gas control for nuclear power reactors. (United States)


    ... 10 Energy 1 2010-01-01 2010-01-01 false Combustible gas control for nuclear power reactors. 50.44... FACILITIES Standards for Licenses, Certifications, and Regulatory Approvals § 50.44 Combustible gas control for nuclear power reactors. (a) Definitions—(1) Inerted atmosphere means a containment atmosphere...

  8. A zero-voltage switching technique for minimizing the current-source power of implanted stimulators. (United States)

    Çilingiroğlu, Uğur; İpek, Sercan


    The current-source power of an implanted stimulator is reduced almost to the theoretical minimum by driving the electrodes directly from the secondary port of the inductive link with a dedicated zero-voltage switching power supply. A feedback loop confined to the secondary of the inductive link adjusts the timing and conduction angle of switching to provide just the right amount of supply voltage needed for keeping the current-source voltage constant at or slightly above the compliance limit. Since drive is based on current rather than voltage, and supply-voltage update is near real-time, the quality of the current pulses is high regardless of how the electrode impedance evolves during stimulation. By scaling the switching frequency according to power demand, the technique further improves overall power consumption of the stimulator. The technique is implemented with a very simple control circuitry comprising a comparator, a Schmitt trigger and a logic gate of seven devices in addition to an on-chip switch and an off-chip capacitor. The power consumed by the proposed supply circuit itself is no larger than what the linear regulator of a conventional supply typically consumes for the same stimulation current. Still, the sum of supply and current-source power is typically between 20% and 75% of the conventional source power alone. Functionality of the proposed driver is verified experimentally on a proof-of-concept prototype built with 3.3 V devices in a 0.18 μm CMOS technology.

  9. Automatic Meter Reading using Power Line Signaling and Voltage Zero-crossing Detection

    Directory of Open Access Journals (Sweden)

    C.L. Vasu


    Full Text Available In India, the electric power transmission and distribution loss is very high, about 7% in transmission and 26% in distribution. Though deployment of automated meter reading system will reduce losses, particularly in distribution, penetration of automated meter reading is low due to high costs involved. World over, the Two-Way Automatic Communications System (TWACS is the most widely used power line communications technology offering two-way communication between substation and end users. The TWACS introduces disturbance on the power system voltage for short durations near zero-crossing to generate the outbound (from substation to end user signal. To generate the inbound (from end user to substation signal, short duration current pulses are introduced, near voltage zero-crossings. Information is generated as a sequential combination of voltage disturbances for the outbound signal and current pulses for the inbound signal. The current study proposes a low-cost modification of the TWACS to reduce voltage and current harmonics. The proposed system has been modelled and simulated using SIMULINK/SIMPOWER Systems. The simulation results show that there is a reduction in voltage harmonics from 0.84 to 0.17% and in current harmonics from 2.07 to 1.10%.

  10. Issues and future direction of thermal-hydraulics research and development in nuclear power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Saha, P., E-mail: [GE Hitachi Nuclear Energy, Wilmington, NC (United States); Aksan, N. [GRNSPG Group, University of Pisa (Italy); Andersen, J. [GE Hitachi Nuclear Energy, Wilmington, NC (United States); Yan, J. [Westinghouse Electric Co., Columbia, SC (United States); Simoneau, J.P. [AREVA, Lyon (France); Leung, L. [Atomic Energy of Canada Ltd., Chalk River, Ontario (Canada); Bertrand, F. [CEA, DEN, DER, F-13108 Saint-Paul-Lez-Durance (France); Aoto, K.; Kamide, H. [Japan Atomic Energy Agency, Chiyoda-ku, Tokyo (Japan)


    The paper archives the proceedings of an expert panel discussion on the issues and future direction of thermal-hydraulic research and development in nuclear power reactors held at the NURETH-14 conference in Toronto, Canada, in September 2011. Thermal-hydraulic issues related to both operating and advanced reactors are presented. Advances in thermal-hydraulics have significantly improved the performance of operating reactors. Further thermal-hydraulics research and development is continuing in both experimental and computational areas for operating reactors, reactors under construction or ready for near-term deployment, and advanced Generation-IV reactors. As the computing power increases, the fine-scale multi-physics computational models, coupled with the systems analysis code, are expected to provide answers to many challenging problems in both operating and advanced reactor designs.

  11. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A


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

  12. Tokamak experimental power reactor conceptual design. Volume I

    Energy Technology Data Exchange (ETDEWEB)


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

  13. Plasma engineering studies for Tennessee Tokamak (TENTOK) fusion power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yokoyama, K.E.; Lacatski, J.T.; Miller, J.B.; Bryan, W.E.; King, P.W.; Santoro, R.T.; Uckan, N.A.; Shannon, T.E.


    This paper summarizes the results of the plasma engineering and systems analysis studies for the Tennessee Tokamak (TENTOK) fusion power reactor. TENTOK is a 3000-MW(t) central station power plant that uses deuterium-tritium fuel in a D-shaped tokamak plasma configuration with a double-null poloidal divertor. The major parameters are R/sub 0/ = 6.4 m, a = 1.6 m, sigma (elongation) = 1.65, (n) = 1.5 x 10/sup 20/ m/sup -3/, (T) = 15 keV, (..beta..) = 6%, B/sub T/ (on-axis) = 5.6 T, I/sub p/ = 8.5 MA, and wall loading = 3 MW/m/sup 2/. Detailed analyses are performed in the areas of (1) transport simulation using the one-and-one-half-dimensional (1-1/2-D) WHIST transport code, (2) equilibrium/poloidal field coil systems, (3) neutral beam and radiofrequency (rf) heating, and (4) pellet fueling. In addition, impurity control systems, diagnostics and controls, and possible microwave plasma preheating and steady-state current drive options are also considered. Some of the major features of TENTOK include rf heating in the ion cyclotron range of frequencies, superconducting equilibrium field coils outside the superconducting toroidal field coils, a double-null poloidal divertor for impurity control and alpha ash removal, and rf-assisted plasma preheating and current startup.

  14. Development of Improved Burnable Poisons for Commercial Nuclear Power Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Renier, J.A.


    Burnable poisons are used in all modern nuclear reactors to permit higher loading of fuel without the necessity of an overly large control rod system. This not only permits a longer core life but can also be used to level the power distribution. Commercial nuclear reactors commonly use B{sub 4}C in separate non-fueled rods and more recently, zirconium boride coatings on the fuel pellets or gadolinium oxide mixed with the fuel. Although the advantages are great, there are problems with using these materials. Boron, which is an effective neutron absorber, transmutes to lithium and helium upon absorption of a neutron. Helium is insoluble and is eventually released to the interior of the fuel rod, where it produces an internal pressure. When sufficiently high, this pressure stress could cause separation of the cladding from the fuel, causing overly high centerline temperatures. Gadolinium has several very strongly absorbing isotopes, but not all have large cross sections and result in residual burnable poison reactivity worth at the end of the fuel life. Even if the amount of this residual absorber is small and the penalty in operation small, the cost of this penalty, even if only several days, can be very high. The objective of this investigation was to study the performance of single isotopes in order to reduce the residual negative reactivity left over at the end of the fuel cycle. Since the behavior of burnable poisons can be strongly influenced by their configuration, four forms for the absorbers were studied: homogeneously mixed with the fuel, mixed with only the outer one-third of the fuel pellet, coated on the perimeter of the fuel pellets, and alloyed with the cladding. In addition, the numbers of fuel rods containing burnable poison were chosen as 8, 16, 64, and 104. Other configurations were chosen for a few special cases. An enrichment of 4.5 wt% {sup 235}U was chosen for most cases for study in order to achieve a 4-year fuel cycle. A standard pressurized

  15. Building America Case Study: New Town Builders' Power of Zero Energy Center, Denver, Colorado (Brochure)

    Energy Technology Data Exchange (ETDEWEB)


    New Town Builders, a builder of energy efficient homes in Denver, Colorado, offers a zero energy option for all the homes it builds. To attract a wide range of potential homebuyers to its energy efficient homes, New Town Builders created a 'Power of Zero Energy Center' linked to its model home in the Stapleton community of Denver. This case study presents New Town Builders' marketing approach, which is targeted to appeal to homebuyers' emotions rather than overwhelming homebuyers with scientific details about the technology. The exhibits in the Power of Zero Energy Center focus on reduced energy expenses for the homeowner, improved occupant comfort, the reputation of the builder, and the lack of sacrificing the homebuyers' desired design features to achieve zero net energy in the home. The case study also contains customer and realtor testimonials related to the effectiveness of the Center in influencing homebuyers to purchase a zero energy home.

  16. New Whole-House Solutions Case Study: New Town Builders' Power of Zero Energy Center - Denver, Colorado

    Energy Technology Data Exchange (ETDEWEB)



    New Town Builders, a builder of energy efficient homes in Denver, Colorado, offers a zero energy option for all the homes it builds. To attract a wide range of potential homebuyers to its energy efficient homes, New Town Builders created a "Power of Zero Energy Center" linked to its model home in the Stapleton community. This case study presents New Town Builders' marketing approach, which is targeted to appeal to homebuyers' emotions rather than overwhelming homebuyers with scientific details about the technology. The exhibits in the Power of Zero Energy Center focus on reduced energy expenses for the homeowner, improved occupant comfort, the reputation of the builder, and the lack of sacrificing the homebuyers' desired design features to achieve zero net energy in the home. This case study also contains customer and realtor testimonials related to the effectiveness of the Center in influencing homebuyers to purchase a zero energy home.

  17. Preapplication safety evaluation report for the Power Reactor Innovative Small Module (PRISM) liquid-metal reactor. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Donoghue, J.E.; Donohew, J.N.; Golub, G.R.; Kenneally, R.M.; Moore, P.B.; Sands, S.P.; Throm, E.D.; Wetzel, B.A. [Nuclear Regulatory Commission, Washington, DC (United States). Associate Directorate for Advanced Reactors and License Renewal


    This preapplication safety evaluation report (PSER) presents the results of the preapplication desip review for die Power Reactor Innovative Small Module (PRISM) liquid-mew (sodium)-cooled reactor, Nuclear Regulatory Commission (NRC) Project No. 674. The PRISM conceptual desip was submitted by the US Department of Energy in accordance with the NRC`s ``Statement of Policy for the Regulation of Advanced Nuclear Power Plants`` (51 Federal Register 24643). This policy provides for the early Commission review and interaction with designers and licensees. The PRISM reactor desip is a small, modular, pool-type, liquid-mew (sodium)-cooled reactor. The standard plant design consists of dim identical power blocks with a total electrical output rating of 1395 MWe- Each power block comprises three reactor modules, each with a thermal rating of 471 MWt. Each module is located in its own below-grade silo and is co to its own intermediate heat transport system and steam generator system. The reactors utilize a metallic-type fuel, a ternary alloy of U-Pu-Zr. The design includes passive reactor shutdown and passive decay heat removal features. The PSER is the NRC`s preliminary evaluation of the safety features in the PRISM design, including the projected research and development programs required to support the design and the proposed testing needs. Because the NRC review was based on a conceptual design, the PSER did not result in an approval of the design. Instead it identified certain key safety issues, provided some guidance on applicable licensing criteria, assessed the adequacy of the preapplicant`s research and development programs, and concluded that no obvious impediments to licensing the PRISM design had been identified.

  18. Civilian Power Program. Part 1, Summary, Current status of reactor concepts

    Energy Technology Data Exchange (ETDEWEB)

    Author, Not Given


    This study group covered the following: delineation of the specific objectives of the overall US AEC civilian power reactor program, technical objectives of each reactor concept, preparation of a chronological development program for each reactor concept, evaluation of the economic potential of each reactor type, a program to encourage the the development, and yardsticks for measuring the development. Results were used for policy review by AEC, program direction, authorization and appropriation requests, etc. This evaluation encompassed civilian power reactors rated at 25 MW(e) or larger and related experimental facilities and R&D. This Part I summarizes the significant results of the comprehensive effort to determine the current technical and economic status for each reactor concept; it is based on the 8 individual technical status reports (Part III).

  19. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors. (United States)


    ... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... Approvals § 50.46 Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium...

  20. Thermophotovoltaic Energy Conversion in Space Nuclear Reactor Power Systems (United States)


    contrasted with nuclear thermal rockets which use the heat from a nuclear fission reactor to heat propellant to provide rocket thrust and radioisotope...K. Note that the highest temperature (2550 K by the Pewee reactor) was for a nuclear thermal rocket application and has the shortest duration (40 min

  1. Possible optimal configurations for the ZECOMIX high efficiency zero emission hydrogen and power plant

    Energy Technology Data Exchange (ETDEWEB)

    Calabro, A.; Deiana, P.; Florini, P.; Girardi, G.; Stendardo, S. [ENEA, Rome (Italy). Italian Agency for New Technology Energy & Environment


    Coal use for electricity generation will continue growing in importance. In the present work the optimization of a high efficiency and zero emissions coal-fired plant, which produces both hydrogen and electricity, has been developed. The majority of this paper concerns an integration of gasification unit, which is characterized by coal hydrogasification and carbon dioxide (CO{sub 2}) separation, with a power island, where a high-hydrogen content syngas is burnt with pure oxygen stream. Another issue is the high temperature CO{sub 2} desorption. Because of the elevated temperature heat supply, the regeneration process affects the overall performance of ZECOMIX plant. An advanced steam cycle characterized by a medium pressure steam compressor and expander has been considered for power generation. A preliminary study of different components leads to analyze possible routes for optimization of the whole plant. The plant equipped with a CO{sub 2} capture unit could reach efficiency close to 50%. The simulations of a thermodynamic model were carried out using the software ChemCAD. This study is a part of a larger research project, named ZECOMIX, led by ENEA (Italian Research Agency for New technologies, Energy and Environment), other partners being ANSALDO and different Italian Universities. It is aimed at analyzing an integrated hydrogen and power production plant.


    Directory of Open Access Journals (Sweden)

    N. Khaled


    Full Text Available A classification on systems for passive afterheat removal from reactor containment has been developed in the paper.  The classification permits to make a detailed analysis of various concepts pertaining to systems for passive afterheat removal from reactor containment of new generation. The paper considers main classification features of the given systems.

  3. Heat pipe cooled reactors for multi-kilowatt space power supplies

    Energy Technology Data Exchange (ETDEWEB)

    Ranken, W.A.; Houts, M.G.


    Three nuclear reactor space power system designs are described that demonstrate how the use of high temperature heat pipes for reactor heat transport, combined with direct conversion of heat to electricity, can result in eliminating pumped heat transport loops for both primary reactor cooling and heat rejection. The result is a significant reduction in system complexity that leads to very low mass systems with high reliability, especially in the power range of 1 to 20 kWe. In addition to removing heat exchangers, electromagnetic pumps, and coolant expansion chambers, the heat pipe/direct conversion combination provides such capabilities as startup from the frozen state, automatic rejection of reactor decay heat in the event of emergency or accidental reactor shutdown, and the elimination of single point failures in the reactor cooling system. The power system designs described include a thermoelectric system that can produce 1 to 2 kWe, a bimodal modification of this system to increase its power level to 5 kWe and incorporate high temperature hydrogen propulsion capability, and a moderated thermionic reactor concept with 5 to 20 kWe power output that is based on beryllium modules that thermally couple cylindrical thermionic fuel elements (TFEs) to radiator heat pipes.

  4. Heat pipe cooled reactors for multi-kilowatt space power supplies (United States)

    Ranken, W. A.; Houts, M. G.

    Three nuclear reactor space power system designs are described that demonstrate how the use of high temperature heat pipes for reactor heat transport, combined with direct conversion of heat to electricity, can result in eliminating pumped heat transport loops for both primary reactor cooling and heat rejection. The result is a significant reduction in system complexity that leads to very low mass systems with high reliability, especially in the power range of 1 to 20 kWe. In addition to removing heat exchangers, electromagnetic pumps, and coolant expansion chambers, the heat pipe/direct conversion combination provides such capabilities as startup from the frozen state, automatic rejection of reactor decay heat in the event of emergency or accidental reactor shutdown, and the elimination of single point failures in the reactor cooling system. The power system designs described include a thermoelectric system that can produce 1 to 2 kWe, a bimodal modification of this system to increase its power level to 5 kWe and incorporate high temperature hydrogen propulsion capability, and a moderated thermionic reactor concept with 5 to 20 kWe power output that is based on beryllium modules that thermally couple cylindrical thermionic fuel elements (TFE's) to radiator heat pipes.

  5. Capital cost: pressurized water reactor plant. Commercial electric power cost studies

    Energy Technology Data Exchange (ETDEWEB)


    The investment cost study for the 1139 MW(e) pressurized water reactor (PWR) central station power plant consists of two volumes. This volume contains the drawings, equipment list and site description.

  6. Searching for full power control rod patterns in a boiling water reactor using genetic algorithms

    Energy Technology Data Exchange (ETDEWEB)

    Montes, Jose Luis [Departamento Sistemas Nucleares, ININ, Carr. Mexico-Toluca Km. 36.5, Ocoyoacac, Edo. de Mexico (Mexico)]. E-mail:; Ortiz, Juan Jose [Departamento Sistemas Nucleares, ININ, Carr. Mexico-Toluca Km. 36.5, Ocoyoacac, Edo. de Mexico (Mexico)]. E-mail:; Requena, Ignacio [Departamento Ciencias Computacion e I.A. ETSII, Informatica, Universidad de Granada, C. Daniel Saucedo Aranda s/n. 18071 Granada (Spain)]. E-mail:; Perusquia, Raul [Departamento Sistemas Nucleares, ININ, Carr. Mexico-Toluca Km. 36.5, Ocoyoacac, Edo. de Mexico (Mexico)]. E-mail:


    One of the most important questions related to both safety and economic aspects in a nuclear power reactor operation, is without any doubt its reactivity control. During normal operation of a boiling water reactor, the reactivity control of its core is strongly determined by control rods patterns efficiency. In this paper, GACRP system is proposed based on the concepts of genetic algorithms for full power control rod patterns search. This system was carried out using LVNPP transition cycle characteristics, being applied too to an equilibrium cycle. Several operation scenarios, including core water flow variation throughout the cycle and different target axial power distributions, are considered. Genetic algorithm fitness function includes reactor security parameters, such as MLHGR, MCPR, reactor k{sub eff} and axial power density.

  7. Output Feedback Dissipation Control for the Power-Level of Modular High-Temperature Gas-Cooled Reactors

    Directory of Open Access Journals (Sweden)

    Zhe Dong


    Full Text Available Because of its strong inherent safety features and the high outlet temperature, the modular high temperature gas-cooled nuclear reactor (MHTGR is the chosen technology for a new generation of nuclear power plants. Such power plants are being considered for industrial applications with a wide range of power levels, thus power-level regulation is very important for their efficient and stable operation. Exploiting the large scale asymptotic closed-loop stability provided by nonlinear controllers, a nonlinear power-level regulator is presented in this paper that is based upon both the techniques of feedback dissipation and well-established backstepping. The virtue of this control strategy, i.e., the ability of globally asymptotic stabilization, is that it takes advantage of the inherent zero-state detectability property of the MHTGR dynamics. Moreover, this newly built power-level regulator is also robust towards modeling uncertainty in the control rod dynamics. If modeling uncertainty of the control rod dynamics is small enough to be omitted, then this control law can be simplified to a classical proportional feedback controller. The comparison of the control performance between the newly-built power controller and the simplified controller is also given through numerical study and theoretical analysis.

  8. Low-power lead-cooled fast reactor loaded with MOX-fuel (United States)

    Sitdikov, E. R.; Terekhova, A. M.


    Fast reactor for the purpose of implementation of research, education of undergraduate and doctoral students in handling innovative fast reactors and training specialists for atomic research centers and nuclear power plants (BRUTs) was considered. Hard neutron spectrum achieved in the fast reactor with compact core and lead coolant. Possibility of prompt neutron runaway of the reactor is excluded due to the low reactivity margin which is less than the effective fraction of delayed neutrons. The possibility of using MOX fuel in the BRUTs reactor was examined. The effect of Keff growth connected with replacement of natural lead coolant to 208Pb coolant was evaluated. The calculations and reactor core model were performed using the Serpent Monte Carlo code.

  9. 77 FR 38339 - Dairyland Power Cooperative, La Crosse Boiling Water Reactor Exemption From Certain Security... (United States)


    ... COMMISSION Dairyland Power Cooperative, La Crosse Boiling Water Reactor Exemption From Certain Security... Dairyland Power Cooperative (DPC). The LACBWR was a nuclear power plant of nominal 50 Mw electrical output... from the regulations in part 73 as it determines are authorized by law and will not endanger life...

  10. Reactor G1: high power experiments; Experiences a forte puissance

    Energy Technology Data Exchange (ETDEWEB)

    Laage, F. de; Teste du Baillet, A.; Veyssiere, A.; Wanner, G. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Retel, H. [Societe Rateau, D.E.A. (France)


    The experiments carried out in the starting-up programme of the reactor G1 comprised a series of tests at high power, which allowed the following points to be studied: 1- Effect of poisoning by Xenon (absolute value, evolution). 2- Temperature coefficients of the uranium and graphite for a temperature distribution corresponding to heating by fission. 3- Effect of the pressure (due to the coiling system) on the reactivity. 4- Calibration of the security rods as a function of their position in the pile (1). 5- Temperature distribution of the graphite, the sheathing, the uranium and the air leaving the canals, in a pile running normally at high power. 6- Neutron flux distribution in a pile running normally at high power. 7- Determination of the power by nuclear and thermodynamic methods. These experiments have been carried out under two very different pile conditions. From the 1. to the 15. of August 1956, a series of power increases, followed by periods of stabilisation, were induced in a pile containing uranium only, in 457 canals, amounting to about 34 tons of fuel. A knowledge of the efficiency of the control rods in such a pile has made it possible to measure with good accuracy the principal effects at high temperatures, that is, to deal with points 1, 2, 3, 5. Flux charts giving information on the variations of the material Laplacian and extrapolation lengths in the reflector have been drawn up. Finally the thermodynamic power has been measured under good conditions, in spite of some installation difficulties. On September 16, the pile had its final charge of 100 tons. All the canals were loaded, 1,234 with uranium and 53 (i.e. exactly 4 per cent of the total number) with thorium uniformly distributed in a square lattice of 100 cm side. Since technical difficulties prevented the calibration of the control rods, the measurements were limited to the determination of the thermodynamic power and the temperature distributions (points 5 and 7). This report will

  11. A comparison between Zecomix High Efficiency Zero Emission Plant and modern Hydrogen and Power IGCC Plants

    Energy Technology Data Exchange (ETDEWEB)

    Deiana, P.; Calabro, A.; Fiorini, P.; Stendardo, S.; Girardi, G.


    The paper reports the analysis and the comparison of two different plant concepts in the field of high efficiency and zero emissions hydrogen and power production plant. The study has been made as a part of a larger research project, named Zecomix, leaded by ENEA (Italian Research Agency for New Technologies, Energy and Environment), and aimed at studying an integrated process that produces both hydrogen and electricity from coal. A thermodynamic model of the two different plants has been set using the industrial software ChemCAD. The Zecomix plant is based on coal hydrogasification and simultaneous steam reforming and carbon dioxide sequestration. Other crucial characteristics involve high temperature sorbent regeneration. The combustion occurs with pure oxygen and high temperature steam evolves in a nonconventional advanced gas-steam turbine cycle. The considered IGCC plant is capable of producing hydrogen and power adopting current technology solutions. The plant configuration includes a pressurized oxygen blown entrained flow gasifier, syngas cleanup and decarbonization based on high pressure physical absorption, the adoption of class H gas turbine and three pressure level recovery boiler. Moreover a pressure swing adsorption unit has been considered for further hydrogen purification. The comparative analysis, based on the same coal input, underlines the differences between the two plants in terms of efficiency and performance of the single component. Moreover a simple environmental impact analysis has been considered to compare specific CO2 emissions of each alternative. (Author)

  12. Determination of critical assembly absolute power using post-irradiation activation measurement of week-lived fission products. (United States)

    Košťál, Michal; Švadlenková, Marie; Milčák, Ján; Rypar, Vojtěch; Koleška, Michal


    The work presents a detailed comparison of calculated and experimentally determined net peak areas of longer-living fission products after 100 h irradiation on a reactor with power of ~630 W and several days cooling. Specifically the nuclides studied are (140)Ba, (103)Ru, (131)I, (141)Ce, (95)Zr. The good agreement between the calculated and measured net peak areas, which is better than in determination using short lived (92)Sr, is reported. The experiment was conducted on the VVER-1000 mock-up installed on the LR-0 reactor. The Monte Carlo approach has been used for calculations. The influence of different data libraries on results of calculation is discussed as well.


    Energy Technology Data Exchange (ETDEWEB)

    Pickenheim, B.; Hansen, E.; Leishear, R.; Marzolf, A.; Reigel, M.


    A 10-inch READCO mixer is used for mixing the premix (45 (wt%) fly ash, 45 wt% slag, and 10 wt% portland cement) with salt solution in the Saltstone Production Facility (SPF). The Saltstone grout free falls into the grout hopper which feeds the suction line leading to the Watson SPX 100 duplex hose pump. The Watson SPX 100 pumps the grout through approximately 1500 feet of piping prior to being discharged into the Saltstone Disposal Facility (SDF) vaults. The existing grout hopper has been identified by the Saltstone Enhanced Low Activity Waste Disposal (ELAWD) project for re-design. The current nominal working volume of this hopper is 12 gallons and does not permit handling an inadvertent addition of excess dry feeds. Saltstone Engineering has proposed a new hopper tank that will have a nominal working volume of 300 gallons and is agitated with a mechanical agitator. The larger volume hopper is designed to handle variability in the output of the READCO mixer and process upsets without entering set back during processing. The objectives of this task involve scaling the proposed hopper design and testing the scaled hopper for the following processing issues: (1) The effect of agitation on radar measurement. Formation of a vortex may affect the ability to accurately measure the tank level. The agitator was run at varying speeds and with varying grout viscosities to determine what parameters cause vortex formation and whether measurement accuracy is affected. (2) A dry feeds over addition. Engineering Calculating X-ESR-Z-00017 1 showed that an additional 300 pounds of dry premix added to a 300 gallon working volume would lower the water to premix ratio (W/P) from the nominal 0.60 to 0.53 based on a Salt Waste Processing Facility (SWPF) salt simulant. A grout with a W/P of 0.53 represents the upper bound of grout rheology that could be processed at the facility. A scaled amount of dry feeds will be added into the hopper to verify that this is a recoverable situation. (3) The necessity of baffles in the hopper. The preference of the facility is not to have baffles in the hopper; however, if the initial testing indicates inadequate agitation or difficulties with the radar measurement, baffles will be tested.

  14. Power distributions in fresh and depleted LEU and HEU cores of the MITR reactor.

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, E.H.; Horelik, N.E.; Dunn, F.E.; Newton, T.H., Jr.; Hu, L.; Stevens, J.G. (Nuclear Engineering Division); (2MIT Nuclear Reactor Laboratory and Nuclear Science and Engineering Department)


    The Massachusetts Institute of Technology Reactor (MITR-II) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. It delivers a neutron flux comparable to current LWR power reactors in a compact 6 MW core using Highly Enriched Uranium (HEU) fuel. In the framework of its non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context, most research and test reactors both domestic and international have started a program of conversion to the use of Low Enriched Uranium (LEU) fuel. A new type of LEU fuel based on an alloy of uranium and molybdenum (UMo) is expected to allow the conversion of U.S. domestic high performance reactors like the MITR-II reactor. Toward this goal, core geometry and power distributions are presented. Distributions of power are calculated for LEU cores depleted with MCODE using an MCNP5 Monte Carlo model. The MCNP5 HEU and LEU MITR models were previously compared to experimental benchmark data for the MITR-II. This same model was used with a finer spatial depletion in order to generate power distributions for the LEU cores. The objective of this work is to generate and characterize a series of fresh and depleted core peak power distributions, and provide a thermal hydraulic evaluation of the geometry which should be considered for subsequent thermal hydraulic safety analyses.

  15. A novel concentrator with zero-index metamaterial for space solar power station (United States)

    Huang, Jin; Chu, Xue-mei; Fan, Jian-yu; Jin, Qi-bao; Duan, Zhu-zhu


    Space solar power station (SSPS) is a comprehensive system that continuously collects solar energy in space and transmits it to ground with a wireless power transmission (WPT) system. These systems have great potential to provide large-scale energy. To increase the efficiency and reduce the weight and cost of the photovoltaic (PV) components, a huge light-weighted concentrator was introduced in the latest SSPS concepts, such as integrated symmetrical concentrator (ISC) and arbitrarily large phased array (ALPHA). However, for typical SSPS running in Geostationary Earth Orbit (GEO), the sunlight direction varies with time, leading to a great challenge for concentrator design. In ISC, the two-dimensional mast is used to realize sun-tracking. However, a multi-thousand-ton structure is difficult to control precisely in space. For this reason, ALPHA comprises a large number of individually pointed thin-film reflectors to intercept sunlight, mounted on the non-moving structure. However, the real-time adjustment of the thousands of reflectors is still an open problem. Furthermore, the uniformity of the time of the power generation (UTPG) is another factor evaluating the system. Therefore, this paper proposes a novel concentrator based on zero-index metamaterial (ZIM) called Thin-film Energy Terminator (SSPS-TENT). This will aid the control of the massive reflectors while avoiding the rotation of the overall system, the control of the massive reflectors and the influence of the obliquity of the ecliptic. Also, an optimization design method is proposed to increase its solar energy collecting efficiency (ECE) and flux distribution (FD). The ray-tracing simulation results show that the ECE is more than 96% of the day. In terms of the FD, the uniformity varies from 0.3057 to 0.5748. Compared with ALPHA, the UTPG is more stable.

  16. Empirical Risk Analysis of Severe Reactor Accidents in Nuclear Power Plants after Fukushima

    Directory of Open Access Journals (Sweden)

    Jan Christian Kaiser


    Full Text Available Many countries are reexamining the risks connected with nuclear power generation after the Fukushima accidents. To provide updated information for the corresponding discussion a simple empirical approach is applied for risk quantification of severe reactor accidents with International Nuclear and Radiological Event Scale (INES level ≥5. The analysis is based on worldwide data of commercial nuclear facilities. An empirical hazard of 21 (95% confidence intervals (CI 4; 62 severe accidents among the world’s reactors in 100,000 years of operation has been estimated. This result is compatible with the frequency estimate of a probabilistic safety assessment for a typical pressurised power reactor in Germany. It is used in scenario calculations concerning the development in numbers of reactors in the next twenty years. For the base scenario with constant reactor numbers the time to the next accident among the world's 441 reactors, which were connected to the grid in 2010, is estimated to 11 (95% CI 3.7; 52 years. In two other scenarios a moderate increase or decrease in reactor numbers have negligible influence on the results. The time to the next accident can be extended well above the lifetime of reactors by retiring a sizeable number of less secure ones and by safety improvements for the rest.

  17. Multi-reactor configurations for multi-megawatt spacecraft power supplies (United States)

    George, Jeffrey A.


    Several conceptual designs for a multimegawatt space nuclear power supply system were developed on the basis of hardware and technology from the existing multihundred kilowatt-class SP-100 reactor program. It is shown that net power outputs in the multimegawatt range can be achieved by using a modular multireactor configuration of several SP-100-derived nuclear power supplies. A variety of geometries were examined for their applicability to the multireactor configuration, showing that modular multireactor systems have the advantage of an increased redundancy in the power system, as compared with a single-reactor system, allowing higher reliabilities than those achievable with single-reactor systems. Results are presented on the Mars Cargo Mission analysis, showing that modularity allows the option of redeployment of power systems in Mars and facilitates refurbishment and turnaround of NEP transfer vehicles.

  18. Reference reactor module for NASA's lunar surface fission power system

    Energy Technology Data Exchange (ETDEWEB)

    Poston, David I [Los Alamos National Laboratory; Kapernick, Richard J [Los Alamos National Laboratory; Dixon, David D [Los Alamos National Laboratory; Werner, James [INL; Qualls, Louis [ORNL; Radel, Ross [SNL


    Surface fission power systems on the Moon and Mars may provide the first US application of fission reactor technology in space since 1965. The Affordable Fission Surface Power System (AFSPS) study was completed by NASA/DOE to determine the cost of a modest performance, low-technical risk surface power system. The AFSPS concept is now being further developed within the Fission Surface Power (FSP) Project, which is a near-term technology program to demonstrate system-level TRL-6 by 2013. This paper describes the reference FSP reactor module concept, which is designed to provide a net power of 40 kWe for 8 years on the lunar surface; note, the system has been designed with technologies that are fully compatible with a Martian surface application. The reactor concept uses stainless-steel based. UO{sub 2}-fueled, pumped-NaK fission reactor coupled to free-piston Stirling converters. The reactor shielding approach utilizes both in-situ and launched shielding to keep the dose to astronauts much lower than the natural background radiation on the lunar surface. The ultimate goal of this work is to provide a 'workhorse' power system that NASA can utilize in near-term and future Lunar and Martian mission architectures, with the eventual capability to evolve to very high power, low mass systems, for either surface, deep space, and/or orbital missions.


    Directory of Open Access Journals (Sweden)



    Full Text Available The combined effects of reactivity coefficients, along with other core nuclear characteristics, determine reactor core behavior in normal operation and accident conditions. The Power Coefficient of Reactivity (PCR is an aggregate indicator representing the change in reactor core reactivity per unit change in reactor power. It is an integral quantity which captures the contributions of the fuel temperature, coolant void, and coolant temperature reactivity feedbacks. All nuclear reactor designs provide a balance between their inherent nuclear characteristics and the engineered reactivity control features, to ensure that changes in reactivity under all operating conditions are maintained within a safe range. The CANDU® reactor design takes advantage of its inherent nuclear characteristics, namely a small magnitude of reactivity coefficients, minimal excess reactivity, and very long prompt neutron lifetime, to mitigate the demand on the engineered systems for controlling reactivity and responding to accidents. In particular, CANDU reactors have always taken advantage of the small value of the PCR associated with their design characteristics, such that the overall design and safety characteristics of the reactor are not sensitive to the value of the PCR. For other reactor design concepts a PCR which is both large and negative is an important aspect in the design of their engineered systems for controlling reactivity. It will be demonstrated that during Loss of Regulation Control (LORC and Large Break Loss of Coolant Accident (LBLOCA events, the impact of variations in power coefficient, including a hypothesized larger than estimated PCR, has no safety-significance for CANDU reactor design. Since the CANDU 6 PCR is small, variations in the range of values for PCR on the performance or safety of the reactor are not significant.

  20. Studies on advanced water-cooled reactors beyond generation Ⅲ for power generation

    Institute of Scientific and Technical Information of China (English)

    CHENG Xu


    China's ambitious nuclear power program motivates the country's nuclear community to develop advanced reactor concepts beyond generation Ⅲ to ensure a long-term, stable, and sustainable development of nuclear power. The paper discusses some main criteria for the selection of future water-cooled reactors by considering the specific Chinese situation. Based on the suggested selection criteria, two new types of water-cooled reactors are recommended for future Chinese nuclear power generation. The high conversion pressurized water reactor utilizes the present PWR technology to a large extent. With a conversion ratio of about 0.95, the fuel utilization is increased about 5 times. This significantly improves the sustainability of fuel resources. The supercritical water-cooled reactor has favorable features in economics,sustainability and technology availability. It is a logical extension of the generation Ⅲ PWR technology in China.The status of international R&D work is reviewed. A new supercritieal water-cooled reactor (SCWR) core structure (the mixed reactor core) and a new fuel assembly design (two-rows FA) are proposed. The preliminary analysis using a coupled neutron-physics/thermal-hydranlics method is carded out. It shows good feasibility for the new design proposal.

  1. Enhancing purification of an azo dye solution in nanosized zero-valent iron-ZnO photocatalyst system using subsequent semibatch packed-bed reactor


    Khani, Ali; SOHRABI, Mahmoud Reza; KHOSRAVI, Morteza; DAVALLO, Mehran


    In this work, simultaneous synthesis-immobilization of nano zero-valent iron (reductive catalyst) and nano ZnO (photocatalyst) on perlite as a suitable bed (nZVI-P and nZnO-P) was done. Three processes (reduction, oxidation, and reduction-oxidation) were evaluated for removal of azo dye C.I. acid orange 7 (AO7) from aqueous solution in a semibatch packed-bed reactor. Our results show that AO7 was successfully removed, synchronously in terms of its total color and total organic carbon, using...

  2. Zero-crossing angle in the np analyzing power at medium energies and its relation to charge symmetry (United States)

    Bhatia, T. S.; Glass, G.; Hiebert, J. C.; Northcliffe, L. C.; Tippens, W. B.; Bonner, B. E.; Simmons, J. E.; Hollas, C. L.; Newsom, C. R.; Riley, P. J.; Ransome, R. D.


    The angle at which the analyzing power for free np scattering becomes zero, the zero-crossing angle θ0, has been measured simultaneously for the n-->p and np--> scattering processes at 425, 565, and 665 MeV incident neutron energies. A rather strong energy dependence of the zerocrossing angle is found. Knowledge of this energy dependence is important in designing an experiment which tests for charge symmetry breaking forces by comparing high precision measurements of θ0(n-->p) and θ0(np-->). [NUCLEAR REACTIONS p(n,p)n, E=425,565, and 665 MeV; polarized neutron beam, polarized proton target; measured analyzing powers A(E,θ) deduced zero-crossing angles; charge symmetry.

  3. Synthesis of Model Based Robust Stabilizing Reactor Power Controller for Nuclear Power Plant

    Directory of Open Access Journals (Sweden)

    Arshad Habib Malik


    Full Text Available In this paper, a nominal SISO (Single Input Single Output model of PHWR (Pressurized Heavy Water Reactor type nuclear power plant is developed based on normal moderator pump-up rate capturing the moderator level dynamics using system identification technique. As the plant model is not exact, therefore additive and multiplicative uncertainty modeling is required. A robust perturbed plant model is derived based on worst case model capturing slowest moderator pump-up rate dynamics and moderator control valve opening delay. Both nominal and worst case models of PHWR-type nuclear power plant have ARX (An Autoregressive Exogenous structures and the parameters of both models are estimated using recursive LMS (Least Mean Square optimization algorithm. Nominal and worst case discrete plant models are transformed into frequency domain for robust controller design purpose. The closed loop system is configured into two port model form and H? robust controller is synthesized. The H?controller is designed based on singular value loop shaping and desired magnitude of control input. The selection of desired disturbance attenuation factor and size of the largest anticipated multiplicative plant perturbation for loop shaping of H? robust controller form a constrained multi-objective optimization problem. The performance and robustness of the proposed controller is tested under transient condition of a nuclear power plant in Pakistan and found satisfactory.

  4. [Oxidation of mercury by CuBr2 decomposition under controlled-release membrane catalysis condition]. (United States)

    Hu, Lin-Gang; Qu, Zan; Yan, Nai-Qiang; Guo, Yong-Fu; Xie, Jiang-Kun; Jia, Jin-Ping


    CuBr2 in the multi-porous ceramic membrane can release Br2 at high temperature, which was employed as the oxidant for Hg0 oxidation. Hg0 oxidation efficiency was studied by a membrane catalysis device. Meanwhile, a reaction and in situ monitoring device was designed to avoid the impact of Br2 on the downstream pipe. The result showed that the MnO(x)/alpha-Al2O3 catalysis membrane had a considerable "controlled-release" effect on Br2 produced by CuBr2 decomposition. The adsorption and reaction of Hg0 and Br2 on the surface of catalysis membrane obeyed the Langmuir-Hinshelwood mechanism. The removal efficiency of Hg0 increased with the rising of Br2 concentration. However, when Br2 reached a certain concentration, the removal efficiency was limited by adsorption rate and reaction rate of Hg0 and Br2 on the catalysis membrane. From 473 K to 573 K, the variation of Hg0 oxidation efficiency was relatively stable. SO2 in flue gas inhibited the oxidation of Hg0 while NO displayed no obvious effect.

  5. DynMo: Dynamic Simulation Model for Space Reactor Power Systems (United States)

    El-Genk, Mohamed; Tournier, Jean-Michel


    A Dynamic simulation Model (DynMo) for space reactor power systems is developed using the SIMULINK® platform. DynMo is modular and could be applied to power systems with different types of reactors, energy conversion, and heat pipe radiators. This paper presents a general description of DynMo-TE for a space power system powered by a Sectored Compact Reactor (SCoRe) and that employs off-the-shelf SiGe thermoelectric converters. SCoRe is liquid metal cooled and designed for avoidance of a single point failure. The reactor core is divided into six equal sectors that are neutronically, but not thermal-hydraulically, coupled. To avoid a single point failure in the power system, each reactor sector has its own primary and secondary loops, and each loop is equipped with an electromagnetic (EM) pump. A Power Conversion assembly (PCA) and a Thermoelectric Conversion Assembly (TCA) of the primary and secondary EM pumps thermally couple each pair of a primary and a secondary loop. The secondary loop transports the heat rejected by the PCA and the pumps TCA to a rubidium heat pipes radiator panel. The primary loops transport the thermal power from the reactor sector to the PCAs for supplying a total of 145-152 kWe to the load at 441-452 VDC, depending on the selections of the primary and secondary liquid metal coolants. The primary and secondary coolant combinations investigated are lithium (Li)/Li, Li/sodium (Na), Na-Na, Li/NaK-78 and Na/NaK-78, for which the reactor exit temperature is kept below 1250 K. The results of a startup transient of the system from an initial temperature of 500 K are compared and discussed.

  6. Determination of refractive index of a simple negative, positive, or zero power lens using wedged plated interferometer (United States)

    Shukla, R. P.; Perera, G. M.; George, M. C.; Venkateswarlu, P.


    A nondestructive technique for measuring the refractive index of a negative lens using a wedged plate interferometer is described. The method can be also used for measuring the refractive index of convex or zero power lenses. Schematic diagrams are presented for the use of a wedged plate interferometer for measuring the refractive index of a concave lens and of a convex lens.

  7. SVBR-100 module-type fast reactor of the IV generation for regional power industry (United States)

    Zrodnikov, A. V.; Toshinsky, G. I.; Komlev, O. G.; Stepanov, V. S.; Klimov, N. N.


    In the report the following is presented: basic conceptual provisions of the innovative nuclear power technology (NPT) based on modular fast reactors (FR) SVBR-100, summarized results of calculations of the reactor, analysis of the opportunities of multi-purpose application of such reactor facilities (RF) including export potentials with due account of nonproliferation requirements. The most important features of the proposed NPT analyzed in the report are as follows: (1) integral (monoblock) arrangement of the primary circuit equipment with entire elimination of the primary circuit pipelines and valves that considerably reduces the construction and assembly works period and coupling with high boiling point of lead-bismuth coolant (LBC) deterministically eliminates accidents of the LOCA type, (2) option for 100 MWe power and dimensions of the reactor provide: on the one hand, an opportunity to transport the reactor monoblock in factory-readiness by railway as well as other kinds of transport, on the other hand, core breeding ratio (CBR) exceeds 1 while MOX-fuel is used. The preferable area of application of RF SVBR-100 is regional and small power requiring power-units of electric power in a range of (100-600) MW, which could be used for cogeneration-based district heating while locating them nearby cities as well as for generation of electric power in a mode of load tracking in the regions with low network systems.

  8. 10-75-kWe-reactor-powered organic Rankine-cycle electric power systems (ORCEPS) study. Final technical report

    Energy Technology Data Exchange (ETDEWEB)


    This 10-75 kW(e) Reactor-ORCEPS study was concerned with the evaluation of several organic Rankine cycle energy conversion systems which utilized a /sup 235/U-ZrH reactor as a heat source. A liquid metal (NaK) loop employing a thermoelectric converter-powered EM pump was used to transfer the reactor energy to the organic working fluid. At moderate peak cycle temperatures (750/sup 0/F), power conversion unit cycle efficiencies of up to 25% and overall efficiencies of 20% can be obtained. The required operating life of seven years should be readily achievable. The CP-25 (toluene) working fluid cycle was found to provide the highest performance levels at the lowest system weights. Specific weights varies from 100 to 50 lb/kW(e) over the power level range 10 to 75 kW(e). (DLC)

  9. Evaluation of organics removal options. A case study from a zero liquid discharge power plant

    Energy Technology Data Exchange (ETDEWEB)

    Quagraine, Emmanuel K.; Hill, Keith Dean; Omorogbe, Fredrick [Shand Power Station, Estevan (Canada)


    Although the role of organics in power plant cycle chemistry still appears to be controversial, their adverse effects in the course of makeup water treatment are very familiar and include fouling of ion exchange resins. This paper describes the organic/bio-fouling experience in a boiler makeup water treatment train for a zero liquid discharge plant, which draws on treated sewage water and surface water for cooling and utilizes the cooling tower blowdown to make distillate water from an evaporator prior to final treatment with a mixed bed demineralizer. In a case study, which is the focus of this paper, the performance of the pilot plants of two recommended organic removal techniques (i.e. reverse osmosis and organic trap resin) were compared to the existing activated carbon bed for organic removal prior to the mixed beds. Parameters evaluated for these three techniques (before and after each unit) include bacteria plate counts, organic carbons, inorganic nutrients (e.g. NH{sub 3}-N, [NO{sub 3}{sup -}+ NO{sub 2}{sup -}]-N, P, Mn, and Fe), known parameters that could significantly impact on the performance of the mixed beds (i.e. SiO{sub 2}, Cl{sup -}, SO{sub 4}{sup 2-}, Na{sup +}), and various others like pH, conductivity, turbidity, HCO{sub 3}{sup -} etc. The effects of oxidizing (i.e. bleach) and non-oxidizing (glutaraldehyde) biocides on the performance of the activated carbon filter were also evaluated.

  10. A zero power harmonic transponder sensor for ubiquitous wireless μL liquid-volume monitoring. (United States)

    Huang, Haiyu; Chen, Pai-Yen; Hung, Cheng-Hsien; Gharpurey, Ranjit; Akinwande, Deji


    Autonomous liquid-volume monitoring is crucial in ubiquitous healthcare. However, conventional approach is based on either human visual observation or expensive detectors, which are costly for future pervasive monitoring. Here we introduce a novel approach based on passive harmonic transponder antenna sensor and frequency hopping spread spectrum (FHSS) pattern analysis, to provide a very low cost wireless μL-resolution liquid-volume monitoring without battery or digital circuits. In our conceptual demonstration, the harmonic transponder comprises of a passive nonlinear frequency multiplier connected to a metamaterial-inspired 3-D antenna designed to be highly sensitive to the liquid-volume within a confined region. The transponder first receives some FHSS signal from an interrogator, then converts such signal to its harmonic band and re-radiates through the antenna sensor. The harmonic signal is picked up by a sniffer receiver and decoded through pattern analysis of the high dimensional FHSS signal strength data. A robust, zero power, absolute accuracy wireless liquid-volume monitoring is realized in the presence of strong direct coupling, background scatters, distance variance as well as near-field human-body interference. The concepts of passive harmonic transponder sensor, metamaterial-inspired antenna sensor, and FHSS pattern analysis based sensor decoding may help establishing cost-effective, energy-efficient and intelligent wireless pervasive healthcare monitoring platforms.

  11. Estimation of ground reaction force and zero moment point on a powered ankle-foot prosthesis. (United States)

    Martinez-Villalpando, Ernesto C; Herr, Hugh; Farrell, Matthew


    The ground reaction force (GRF) and the zero moment point (ZMP) are important parameters for the advancement of biomimetic control of robotic lower-limb prosthetic devices. In this document a method to estimate GRF and ZMP on a motorized ankle-foot prosthesis (MIT Powered Ankle-Foot Prosthesis) is presented. The method proposed is based on the analysis of data collected from a sensory system embedded in the prosthetic device using a custom designed wearable computing unit. In order to evaluate the performance of the estimation methods described, standing and walking clinical studies were conducted on a transtibial amputee. The results were statistically compared to standard analysis methodologies employed in a gait laboratory. The average RMS error and correlation factor were calculated for all experimental sessions. By using a static analysis procedure, the estimation of the vertical component of GRF had an averaged correlation coefficient higher than 0.94. The estimated ZMP location had a distance error of less than 1 cm, equal to 4% of the anterior-posterior foot length or 12% of the medio-lateral foot width.

  12. Determination of the power density peak factor from out-of-core detector signals in small reactors; Determinacao do fator de pico utilizando sinais de detectores out-of-core em reator de pequeno porte

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Rose Mary Gomes do Prado [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), Belo Horizonte, MG (Brazil); Moreira, Joao Manoel Losada [Centro Tecnologico da Marinha (CTMSP), Sao Paulo, SP (Brazil)


    This paper aims to show that in small reactors the power density peak factor can be determined from the power axial offset, measured from out-of-core detector signals, and from signals indicating the control rod positions in the core. The response of out-of-core detector signals were measured in experiments performed in the IPEN/MB-01 zero-power reactor. Several reactor states with different power density distribution were obtained by positioning the control rods in different configurations. The power distribution and its peak factor were calculated with the CITATION code for each of these reactor states. The obtained results show that for the small reactors there is a correlation between the peak factor, the control rod position and power axial offset. The analysis shows that a complex pattern exist between the peak factor and the power axial offset. In order to improve the accuracy, the results indicate that the peak factor should be determined through a correlation which takes into account the power axial offset and the quadrant power tilt. (author)

  13. Research on Power Ramp Testing Method for PWR Fuel Rod at Research Reactor

    Institute of Scientific and Technical Information of China (English)


    In order to develop high performance fuel assembly for domestic nuclear power plant, it is necessary to master some fundamental test technology. So the research on the power ramp testing methods is proposed. A tentative power ramp test for short PWR fuel rod has been conducted at the heavy water research reactor (HWRR) in China Institute of Atomic Energy (CIAE) in May of 2001. The in-pile test rig was placed into the central channel of the reactor . The test rig consists of pressure pipe assembly, thimble, solid neutron absorbing screen and its driving parts, etc.. The test

  14. A computer program to determine the specific power of prismatic-core reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dobranich, D.


    A computer program has been developed to determine the maximum specific power for prismatic-core reactors as a function of maximum allowable fuel temperature, core pressure drop, and coolant velocity. The prismatic-core reactors consist of hexagonally shaped fuel elements grouped together to form a cylindrically shaped core. A gas coolant flows axially through circular channels within the elements, and the fuel is dispersed within the solid element material either as a composite or in the form of coated pellets. Different coolant, fuel, coating, and element materials can be selected to represent different prismatic-core concepts. The computer program allows the user to divide the core into any arbitrary number of axial levels to account for different axial power shapes. An option in the program allows the automatic determination of the core height that results in the maximum specific power. The results of parametric specific power calculations using this program are presented for various reactor concepts.

  15. Hierarchical Control Strategy of Heat and Power for Zero Energy Buildings including Hybrid Fuel Cell/Photovoltaic Power Sources and Plug-in Electric Vehicle

    DEFF Research Database (Denmark)

    Ghiasi, Mohammad Iman; Aliakbar Golkar, Masoud; Hajizadeh, Amin


    complexities and uncertainties in this kind of hybrid system, a hybrid supervisory control with an adaptive fuzzy sliding power control strategy is proposed to regulate the amount of requested fuel from a fuel cell power source to produce the electrical power and heat. Then, simulation results are used......This paper presents a hierarchical control strategy for heat and electric power control of a building integrating hybrid renewable power sources including photovoltaic, fuel cell and battery energy storage with Plug-in Electric Vehicles (PEV) in smart distribution systems. Because...... of the controllability of fuel cell power, this power sources plays the main role for providing heat and electric power to zero emission buildings. First, the power flow structure between hybrid power resources is described. To do so, all necessary electrical and thermal equations are investigated. Next, due to the many...

  16. Irradiation effects on Zr-2.5Nb in power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Song, C., E-mail: [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)


    Zirconium alloys are widely used as structural materials in nuclear applications because of their attractive properties such as a low absorption cross-section for thermal neutrons, excellent corrosion resistance in water, and good mechanical properties at reactor operating temperatures. Zr-2.5Nb is one of the most commonly used zirconium alloys and has been used for pressure tube materials in CANDU (Canada Deuterium Uranium) and RBMK (Reaktor Bolshoy Moshchnosti Kanalnyy, 'High Power Channel-type Reactor') reactors for over 40 years. In a recent report from the Electric Power Research Institute, Zr-2.5Nb was identified as one of the candidate materials for use in normal structural applications in light-water reactors owing to its increased resistance to irradiation-induced degradation as compared with currently used materials. Historically, the largest program of in-reactor tests on zirconium alloys was performed by Atomic Energy of Canada Limited. Over many years of in-reactor testing and CANDU operating experience with Zr- 2.5Nb, extensive research has been conducted on the irradiation effects on its microstructures, mechanical properties, deformation behaviours, fracture toughness, delayed hydride cracking, and corrosion. Most of the results on Zr-2.5Nb obtained from CANDU experience could be used to predict the material performance under light water reactors. This paper reviews the irradiation effects on Zr-2.5Nb in power reactors (including heavy-water and light-water reactors) and summarizes the current state of knowledge. (author)

  17. Materials technology for an advanced space power nuclear reactor concept: Program summary (United States)

    Gluyas, R. E.; Watson, G. K.


    The results of a materials technology program for a long-life (50,000 hr), high-temperature (950 C coolant outlet), lithium-cooled, nuclear space power reactor concept are reviewed and discussed. Fabrication methods and compatibility and property data were developed for candidate materials for fuel pins and, to a lesser extent, for potential control systems, reflectors, reactor vessel and piping, and other reactor structural materials. The effects of selected materials variables on fuel pin irradiation performance were determined. The most promising materials for fuel pins were found to be 85 percent dense uranium mononitride (UN) fuel clad with tungsten-lined T-111 (Ta-8W-2Hf).

  18. Temperature coefficients in the Dragon low-enriched power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hansen, U.


    The temperature coefficient of the fuel and of the moderator have been evaluated for the Dragon HTR design for different stages in reactor life, initial core, end of no-refuelling period and equilibrium conditions. The investigation has shown the low-enriched HTR to have a strong, positive moderator coefficient. In some cases and for special operating conditions, even leading to a positive total temperature coefficient. This does not imply, however, that the HTR is an unsafe reactor system. By adequate design of the control system, safe and reliable operating characteristics can be achieved. This has already been proved satisfactory through many years of operation of other graphite moderated systems, such as the Magnox stations.

  19. PREMOR: a point reactor exposure model computer code for survey analysis of power plant performance

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.


    The PREMOR computer code was written to exploit a simple, two-group point nuclear reactor power plant model for survey analysis. Up to thirteen actinides, fourteen fission products, and one lumped absorber nuclide density are followed over a reactor history. Successive feed batches are accounted for with provision for from one to twenty batches resident. The effect of exposure of each of the batches to the same neutron flux is determined.

  20. Ultrahigh temperature vapor core reactor-MHD system for space nuclear electric power (United States)

    Maya, Isaac; Anghaie, Samim; Diaz, Nils J.; Dugan, Edward T.


    The conceptual design of a nuclear space power system based on the ultrahigh temperature vapor core reactor with MHD energy conversion is presented. This UF4 fueled gas core cavity reactor operates at 4000 K maximum core temperature and 40 atm. Materials experiments, conducted with UF4 up to 2200 K, demonstrate acceptable compatibility with tungsten-molybdenum-, and carbon-based materials. The supporting nuclear, heat transfer, fluid flow and MHD analysis, and fissioning plasma physics experiments are also discussed.

  1. Analysis of N-16 concentration in primary cooling system of AP1000 power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rohanda, Anis [Center for Reactor Technology and Nuclear Safety – BATAN Kawasan PUSPIPTEK Gd. No. 80 Serpong, Tangerang Selatan 15310 (Indonesia); Waris, Abdul [Physics Department of ITB, Indonesia (Indonesia)


    Nitrogen-16 (N-16) is one of the radiation safety parameter on the primary reactor system. The activation product, N-16, is the predominant contributor to the activity in the reactor coolant system during reactor operation. N-16 is activation product derived from activation of O-16 with fast neutron based on {sup 16}O(n,p){sup 16}N reaction. Thus study is needed and it performs to determine N-16 concentration in reactor coolant (primary coolant) in supporting radiation safety. One of the way is using analytical methode based on activation and redecay princip to obtain N-16 concentration. The analysis was performed on the configuration basis and operational of Westinghouse AP1000 power reactor in several monitoring points at coolant reactor system. The results of the calculation of N-16 concentration at the core outlet, reactor vessel outlet, pressurizer line, inlet and outlet of steam generators, primary pumps, reactor vessels inlet and core inlet are: 281, 257, 255, 250, 145, 142, 129 and 112 µCi/gram respectively. The results of analysis compared with AP1000 design control document as standard values. The verification showed very high accuracy comparation between analytical results and standard values.

  2. Fuel supply of nuclear power industry with the introduction of fast reactors (United States)

    Muraviev, E. V.


    The results of studies conducted for the validation of the updated development strategy for nuclear power industry in Russia in the 21st century are presented. Scenarios with different options for the reprocessing of spent fuel of thermal reactors and large-scale growth of nuclear power industry based on fast reactors of inherent safety with a breeding ratio of ˜1 in a closed nuclear fuel cycle are considered. The possibility of enhanced fuel breeding in fast reactors is also taken into account in the analysis. The potential to establish a large-scale nuclear power industry that covers 100% of the increase in electric power requirements in Russia is demonstrated. This power industry may be built by the end of the century through the introduction of fast reactors (replacing thermal ones) with a gross uranium consumption of up to ˜1 million t and the termination of uranium mining even if the reprocessing of spent fuel of thermal reactors is stopped or suffers a long-term delay.

  3. Neutronic predesign tool for fusion power reactors system assessment

    Energy Technology Data Exchange (ETDEWEB)

    Jaboulay, J.-C., E-mail: [CEA, DEN, Saclay, DM2S, SERMA, F-91191 Gif-sur-Yvette (France); Li Puma, A. [CEA, DEN, Saclay, DM2S, SERMA, F-91191 Gif-sur-Yvette (France); Martínez Arroyo, J. [ETSEIB, Internship in CEA (Spain)


    SYCOMORE, a fusion reactor system code based on a modular approach, is under development at CEA. In this framework, this paper describes a methodology developed to build the neutronic module of SYCOMORE. This neutronic module aims to evaluate main neutronic parameters characterising a fusion reactor (tokamak): tritium breeding ratio, multiplication factor, nuclear heating as a function of the reactor main geometrical parameters (major radius, elongation, etc.), of the radial build, Li enrichment, blanket and shield thickness, etc. It is based on calculations carried out with APOLLO2 and TRIPOLI-4 CEA transport code on simplified 1D and 2D neutronic models. These models are validated versus a more detailed 3D Monte-Carlo model (using TRIPOLI-4). To ease the integration of this neutronic module in SYCOMORE and provide results instantly, a surrogate model that replicates the 1D and 2D neutronic model results was used. Among the different surrogate models types (polynomial interpolation, responses functions, interpolating by Kriging, artificial neural network, etc.) the neural networks were selected for their efficiency and flexibility. The methodology described in this paper to build SYCOMORE neutronic module is devoted to HCLL blanket, but it could be applied to any breeder blanket concept provided that appropriate validation could be carried out.

  4. 78 FR 64029 - Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors (United States)


    ... COMMISSION Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors AGENCY... Systems for Light-Water-Cooled Nuclear Power Reactors,'' in which the NRC made editorial corrections and... analysis for liquid and gaseous radwaste system components for light water nuclear power...

  5. Reactor safety study. An assessment of accident risks in U. S. commercial nuclear power plants. Appendices VII, VIII, IX, and X. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)


    Information is presented concerning the release of radioactivity in reactor accidents; physical processes in reactor meltdown accidents; safety design rationale for nuclear power plants; and design adequacy.

  6. Influence of DC Supply Systems on Unplanned Reactor Trips in Nuclear Power Plants

    Institute of Scientific and Technical Information of China (English)

    李君利; 童节娟; 茆定远


    Operational experience has shown that some components in nuclearpower plants are so important that their failures, which would be a single failure, may cause the entire plant to shutdown. Such shutdowns have often occurred in the past in commercial nuclear power plants. Nuclear power plant authorities try to avoid such unplanned plant shutdowns because of the large economic loss. Unfortunately, it is difficult to identify all the important components from the numerous components in each complex nuclear power plant system. FMEA and FTA methods, which are often applied to probabilistic risk assessments, are used in this paper to identify the key components that may cause unplanned reactor trips. As an example, the 48 V DC power supply system in a typical Chinese nuclear power plant, which is a major cause of many unplanned reactor trips, was analyzed to show how to identify these key components and the causes for nuclear power plant trips.

  7. Key Technology for Digital Mock-up Research and Development of Complex Product%复杂产品数字样机开发的关键技术

    Institute of Scientific and Technical Information of China (English)

    南长峰; 孟祥海; 蔡真


    In order to study the key technology involved in the digital mock-up research and development of complex products, aeroengine is taken as example to analyze the methods of modeling parts, assembly and associated design in the digital prototype. Firstly, 3D models of aeroengine parts are established through the CAD software. Then, for the purpose of studying the lightweight, constrains, interference and other problems occurred in the assembly process, the sub-assemblies and full assembly of the engine prototype are conducted based on those 3D models. Finally, by means of utilizing WAVE technology,the associated parametric design of the blisk is achieved.By replacing the physical prototype with the digital mockup,it promotes the design and manufacturing development of high-quality, high-efficiency and low-cost products.%为研究复杂产品数字样机开发中的关键技术,本文以航空发动机为例,对其数字样机的建模、装配和关联设计方法进行了分析.首先,基于CAD软件,建立了航空发动机零部件三维模型;其次,对上述零部件进行了子装配与整机装配,并对其过程中的轻量化模型、约束关系和干涉检查等问题进行了研究;最后,通过WAVE技术,实现整体叶盘的关联参数设计,使数字样机最优地取代物理样机,推进了高质、高效、低成本的产品设计和制造技术的发展.

  8. Case analysis on pharmacology-toxicology section of mock-up package inserts in registration%申报注册药品说明书样稿[药理毒理]部分问题案例讨论

    Institute of Scientific and Technical Information of China (English)

    李雪梅; 萧惠来


    Pharmacology & Toxicology section is an important element of package inserts. The section aims to describe the essential disciplines of drug interaction with organisms (in particular referring to human body) , including dose-effect relationship, time-effect relationship and mechanism of action, etc. From the several hundreds mock-up package inserts in registration database, specific issues from 17 typical cases in which the section of Pharmacology & Toxicology violated the requirement of related US and Chinese regulatory guidance. The expertise are I-dentified and discussed one by one. The authors wish the review would be a useful reference to the authors, users and governors of package inserts, and also helpful in the improvement for quality of package inserts.%药品说明书中[药理毒理]项是揭示药品对机体(特别是人体)作用的基本规律,包括剂量(浓度)-效应关系、时间—效应关系以及作用机制等,是说明书的重要内容.作者根据我国及美国相关指导原则要求,从当前药品注册申报的数百份说明书样稿中,收集了17个[药理毒理]部分不符合法规要求和违背专业知识的典型案例,进行了逐一讨论,分析存在的问题,供药品说明书的起草者、使用者和监管者借鉴,希望有助于药品说明书质量的提高.

  9. Nanoscale zero-valent iron/persulfate enhanced upflow anaerobic sludge blanket reactor for dye removal: Insight into microbial metabolism and microbial community (United States)

    Pan, Fei; Zhong, Xiaohan; Xia, Dongsheng; Yin, Xianze; Li, Fan; Zhao, Dongye; Ji, Haodong; Liu, Wen


    This study investigated the efficiency of nanoscale zero-valent iron combined with persulfate (NZVI/PS) for enhanced degradation of brilliant red X-3B in an upflow anaerobic sludge blanket (UASB) reactor, and examined the effects of NZVI/PS on anaerobic microbial communities during the treatment process. The addition of NZVI (0.5 g/L) greatly enhanced the decolourization rate of X-3B from 63.8% to 98.4%. The Biolog EcoPlateTM technique was utilized to examine microbial metabolism in the reactor, and the Illumina MiSeq high-throughput sequencing revealed 22 phyla and 88 genera of the bacteria. The largest genera (Lactococcus) decreased from 33.03% to 7.94%, while the Akkermansia genera increased from 1.69% to 20.23% according to the abundance in the presence of 0.2 g/L NZVI during the biological treatment process. Meanwhile, three strains were isolated from the sludge in the UASB reactors and identified by 16 S rRNA analysis. The distribution of three strains was consistent with the results from the Illumina MiSeq high throughput sequencing. The X-ray photoelectron spectroscopy results indicated that Fe(0) was transformed into Fe(II)/Fe(III) during the treatment process, which are beneficial for the microorganism growth, and thus promoting their metabolic processes and microbial community.

  10. A simple dynamic model and transient simulation of the nuclear power reactor on microcomputers

    Energy Technology Data Exchange (ETDEWEB)

    Han, Yang Gee; Park, Cheol [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)


    A simple dynamic model is developed for the transient simulation of the nuclear power reactor. The dynamic model includes the normalized neutron kinetics model with reactivity feedback effects and the core thermal-hydraulics model. The main objective of this paper demonstrates the capability of the developed dynamic model to simulate various important variables of interest for a nuclear power reactor transient. Some representative results of transient simulations show the expected trends in all cases, even though no available data for comparison. In this work transient simulations are performed on a microcomputer using the DESIRE/N96T continuous system simulation language which is applicable to nuclear power reactor transient analysis. 3 refs., 9 figs. (Author)

  11. Multi-reactor power system configurations for multimegawatt nuclear electric propulsion (United States)

    George, Jeffrey A.


    A modular, multi-reactor power system and vehicle configuration for piloted nuclear electric propulsion (NEP) missions to Mars is presented. Such a design could provide enhanced system and mission reliability, allowing a comfortable safety margin for early manned flights, and would allow a range of piloted and cargo missions to be performed with a single power system design. Early use of common power modules for cargo missions would also provide progressive flight experience and validation of standardized systems for use in later piloted applications. System and mission analysis are presented to compare single and multi-reactor configurations for piloted Mars missions. A conceptual design for the Hydra modular multi-reactor NEP vehicle is presented.

  12. Fusion-power-core design of a Compact Reversed-Field Pinch Reactor (CRFPR) (United States)

    Copenhaver, C.; Schnurr, N. M.; Krakowski, R. A.; Hagenson, R. L.; Mynard, R. C.; Cappiello, C.; Lujan, R. E.; Davidson, J. W.; Chaffee, A. D.; Battat, M. E.

    A conceptual design of a fusion power core (FPC, i.e., plasma chamber, first wall, blanket, shield, coils) based on a Reversed-Field Pinch (RFP) has been completed. After a brief statement of rationale and description of the reactor configuraton, the FPC integration is described in terms of power balance, thermal-hydraulics, and mechanical design. The engineering versatility, promise, and problems of this high-power-density approach to fusion are addressed.

  13. Calculated power distribution of a thermionic, beryllium oxide reflected, fast-spectrum reactor (United States)

    Mayo, W.; Lantz, E.


    A procedure is developed and used to calculate the detailed power distribution in the fuel elements next to a beryllium oxide reflector of a fast-spectrum, thermionic reactor. The results of the calculations show that, although the average power density in these outer fuel elements is not far from the core average, the power density at the very edge of the fuel closest to the beryllium oxide is about 1.8 times the core avearge.

  14. Design Study of Modular Nuclear Power Plant with Small Long Life Gas Cooled Fast Reactors Utilizing MOX Fuel (United States)

    Ilham, Muhammad; Su’ud, Zaki


    Growing energy needed due to increasing of the world’s population encourages development of technology and science of nuclear power plant in its safety and security. In this research, it will be explained about design study of modular fast reactor with helium gas cooling (GCFR) small long life reactor, which can be operated over 20 years. It had been conducted about neutronic design GCFR with Mixed Oxide (UO2-PuO2) fuel in range of 100-200 MWth NPPs of power and 50-60% of fuel fraction variation with cylindrical pin cell and cylindrical balance of reactor core geometry. Calculation method used SRAC-CITATION code. The obtained results are the effective multiplication factor and density value of core reactor power (with geometry optimalization) to obtain optimum design core reactor power, whereas the obtained of optimum core reactor power is 200 MWth with 55% of fuel fraction and 9-13% of percentages.

  15. The role of actinide burning and the Integral Fast Reactor in the future of nuclear power

    Energy Technology Data Exchange (ETDEWEB)

    Hollaway, W.R.; Lidsky, L.M.; Miller, M.M.


    A preliminary assessment is made of the potential role of actinide burning and the Integral Fast Reactor (IFR) in the future of nuclear power. The development of a usable actinide burning strategy could be an important factor in the acceptance and implementation of a next generation of nuclear power. First, the need for nuclear generating capacity is established through the analysis of energy and electricity demand forecasting models which cover the spectrum of bias from anti-nuclear to pro-nuclear. The analyses take into account the issues of global warming and the potential for technological advances in energy efficiency. We conclude, as do many others, that there will almost certainly be a need for substantial nuclear power capacity in the 2000--2030 time frame. We point out also that any reprocessing scheme will open up proliferation-related questions which can only be assessed in very specific contexts. The focus of this report is on the fuel cycle impacts of actinide burning. Scenarios are developed for the deployment of future nuclear generating capacity which exploit the advantages of actinide partitioning and actinide burning. Three alternative reactor designs are utilized in these future scenarios: The Light Water Reactor (LWR); the Modular Gas-Cooled Reactor (MGR); and the Integral Fast Reactor (FR). Each of these alternative reactor designs is described in some detail, with specific emphasis on their spent fuel streams and the back-end of the nuclear fuel cycle. Four separation and partitioning processes are utilized in building the future nuclear power scenarios: Thermal reactor spent fuel preprocessing to reduce the ceramic oxide spent fuel to metallic form, the conventional PUREX process, the TRUEX process, and pyrometallurgical reprocessing.

  16. Rotating-bed reactor as a power source for EM gun applications

    Energy Technology Data Exchange (ETDEWEB)

    Powell, J.; Botts, T.; Stickley, C.M.; Meth, S.


    Electromagnetic gun applications of the Rotating Bed Reactor (RBR) are examined. The RBR is a compact (approx. 1 m/sup 3/), (up to several thousand MW(th)), high-power reactor concept, capable of producing a high-temperature (up to approx. 300/sup 0/K) gas stream with a MHD generator coupled to it, the RBR can generate electric power (up to approx. 1000 MW(e)) in the pulsed or cw modes. Three EM gun applications are investigated: a rail gun thruster for orbit transfer, a rapid-fire EM gun for point defense, and a direct ground-to-space launch. The RBR appears suitable for all applications.

  17. The design and simulation of TCR(thyristor control reactor) reactive power compensation system based on Arene

    Institute of Scientific and Technical Information of China (English)

    WANG Shu-fang; ZHANG Li; JIANG Jian-guo; WANG Ru-lin


    Inevitably, the question of reactive power compensation was aroused by applied of power electronics. Based on the study of the instantaneous reactive power theory, the designs of TCR(thyristor control reactor) thyristor control reactor reactive power compensation system and TCR single closed loop strategy was proposed. In addition, as digital simulation software, Arene was applied to simulate the Jining coal mine No.2 system. The simulation results validate that the design is effective to improve power factor and stabilization of the system.

  18. Gas Cooled, Natural Uranium, D20 Moderated Power Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dahlberg, R.C.; Beasley, E.G.; DeBoer, T.K.; Evans, T.C.; Molino, D.F.; Rothwell, W.S.; Slivka, W.R.


    The attractiveness of a helium cooled, heavy water moderated, natural uranium central station power plant has been investigated. A fuel element has been devised which allows the D20 to be kept at a low pressure while the exit gas temperature is high. A preliminary cost analysis indicates that, using currently available materials, competitive nuclear power in foreign countries is possible.

  19. High Efficiency Nuclear Power Plants Using Liquid Fluoride Thorium Reactor Technology (United States)

    Juhasz, Albert J.; Rarick, Richard A.; Rangarajan, Rajmohan


    An overall system analysis approach is used to propose potential conceptual designs of advanced terrestrial nuclear power plants based on Oak Ridge National Laboratory (ORNL) Molten Salt Reactor (MSR) experience and utilizing Closed Cycle Gas Turbine (CCGT) thermal-to-electric energy conversion technology. In particular conceptual designs for an advanced 1 GWe power plant with turbine reheat and compressor intercooling at a 950 K turbine inlet temperature (TIT), as well as near term 100 MWe demonstration plants with TITs of 950 and 1200 K are presented. Power plant performance data were obtained for TITs ranging from 650 to 1300 K by use of a Closed Brayton Cycle (CBC) systems code which considered the interaction between major sub-systems, including the Liquid Fluoride Thorium Reactor (LFTR), heat source and heat sink heat exchangers, turbo-generator machinery, and an electric power generation and transmission system. Optional off-shore submarine installation of the power plant is a major consideration.

  20. Power level effects on thorium-based fuels in pressure-tube heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bromley, B.P.; Edwards, G.W.R., E-mail: [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Sambavalingam, P. [Univ. of Ontario Inst. of Technology, Oshawa, Ontario (Canada)


    Lattice and core physics modeling and calculations have been performed to quantify the impact of power/flux levels on the reactivity and achievable burnup for 35-element fuel bundles made with Pu/Th or U-233/Th. The fissile content in these bundles has been adjusted to produce on the order of 20 MWd/kg burnup in homogeneous cores in a 700 MWe-class pressure-tube heavy water reactor, operating on a once-through thorium cycle. Results demonstrate that the impact of the power/flux level is modest for Pu/Th fuels but significant for U-233/Th fuels. In particular, high power/flux reduces the breeding and burnup potential of U-233/Th fuels. Thus, there may be an incentive to operate reactors with U-233/Th fuels at a lower power density or to develop alternative refueling schemes that will lower the time-average specific power, thereby increasing burnup.(author)


    Directory of Open Access Journals (Sweden)

    ED Iftitah


    -EELS. The catalysts spesific surface area and porosity determined by adsorption-desorption of dinitrogen at 77 K. Pore distribution and volume were determined by the desorption isotherm at P/Po ≥ 0.3. The result showed that there was correlation between the catalyst characteristics and catalytic activity to (R-(+-Citronellal isomerisation and hydrogenation product. The activity test were performed in a mini fixed bed reactor with 0.5 g of catalyst and 3 mL of (R-(+-Citronellal using N2 and/or H2 gas atmosphere in 5 and 24 hours at each temperature 90 and 120 oC. The catalyst composition of the choice of gas atmosphere and temperature greatly influenced the activity as well as the selectivity of isomerisation and hydrogenation product formation. The highest conversion was achieved for A3=Ni/ZnBr2/γ-Al2O3 (2:3 with complete conversion of (R-(+-Citronellal were obtained when it was running in 5 hours (4 hours N2 + 1 hour H2 at 90 oC and 24 hours (4 hours N2 + 20 hour H2 at 120 oC.

  2. Power Beaming to Space Using a Nuclear Reactor-Pumped Laser (United States)

    Lipinski, Ronald J.; Monroe, David K.; Pickard, Paul S.


    The present political and environmental climate may slow the inevitable direct utilization of nuclear power in space. In the meantime, there is another approach for using nuclear energy for space power. That approach is to let nuclear energy generate a laser beam in a ground-based nuclear reactor-pumped laser (RPL), and then beam the optical energy into space. Potential space applications for a ground-based RPL include (1) illuminating geosynchronous communication satellites in the earth's shadow to extend their lives, (2) beaming power to orbital transfer vehicles, (3) providing power (from earth) to a lunar base during the long lunar night, and (4) removing space debris. FALCON is a high-power, steady-state, nuclear reactor-pumped laser (RPL) concept which is being developed by the Department of Energy with Sandia National Laboratories as the lead laboratory. The FALCON program has experimentally demonstrated reactor-pumped lasing in various mixtures of xenon, argon, neon, and helium at wavelengths of 585, 703, 725, 1271, 1733, 1792, 2032, 2630, 2650, and 3370 nm with intrinsic efficiency as high as 2.5%. Frequency-doubling the 1733-nm line would yield a good match for photovoltaic arrays at 867 nm. Preliminary designs of an RPL suitable for power beaming have been completed. The MW- class laser is fairly simple in construction, self-powered, closed-cycle (no exhaust gases), and modular. This paper describes the FALCON program accomplishments and power-beaming applications.

  3. Critical experiments supporting close proximity water storage of power reactor fuel. Technical progress report, October 1, 1977-December 31, 1977

    Energy Technology Data Exchange (ETDEWEB)

    Baldwin, M.N.; Hoovler, G.S.


    Experiments are being conducted on critical configurations of clusters of fuel rods, mocking up LWR-type fuel elements in close proximity water storage. Spacings between fuel clusters and the intervening material are being varied to provide a variety of benchmark loadings. (DLC)

  4. Random processes in nuclear reactors

    CERN Document Server

    Williams, M M R


    Random Processes in Nuclear Reactors describes the problems that a nuclear engineer may meet which involve random fluctuations and sets out in detail how they may be interpreted in terms of various models of the reactor system. Chapters set out to discuss topics on the origins of random processes and sources; the general technique to zero-power problems and bring out the basic effect of fission, and fluctuations in the lifetime of neutrons, on the measured response; the interpretation of power reactor noise; and associated problems connected with mechanical, hydraulic and thermal noise sources

  5. Delayed Neutrons Effect on Power Reactor with Variation of Fluid Fuel Velocity at MSR Fuji-12 (United States)

    Kuncoro Aji, Indarta; Pramuditya, Syeilendra; Novitrian; Irwanto, Dwi; Waris, Abdul


    As the nuclear reactor operate with liquid fuel, controlling velocity of the fuel flow on the Molten salt reactor very influence on the neutron kinetics in that reactor system. The effect of the pace fuel changes to the populations number of neutrons and power density on vertical direction (1 dimension) from the first until fifth year reactor operating had been analyzed on this research. This research had been conducted on MSR Fuji-12 with a two meters core high, and LiF-BeF2-ThF4-233UF4 as fuel composition respectively 71.78%-16%-11.86%-0.36%. Data of reactivity, neutron flux, and the macroscopic fission cross section obtained from ouput of SRAC (neutronic calculation code has been developed by JAEA, with JENDL-4.0 as data library on the SRAC calculation) was being used for the calculation process of this research. The calculation process of this research had been performed numerically by SOR (successive over relaxation) and finite difference methode, as well as using C programing language. From the calculation, regarding to the value of power density resulting from delayed neutrons, concluded that 20 m/s is the optimum fuel flow velocity in all the years reactor had operated. Where the increases number of power are inversely proportional with the fuel flow speed.

  6. Novel, Integrated Reactor / Power Conversion System (LMR-AMTEC)

    Energy Technology Data Exchange (ETDEWEB)

    Pablo Rubiolo, Principal Investigator


    The main features of this project were the development of a long life (up to 10 years) Liquid Metal Reactor (LMR) and a static conversion subsystem comprising an Alkali Metal Thermal-to-Electric (AMTEC) topping cycle and a ThermoElectric (TE) Bottom cycle. Various coupling options of the LMR with the energy conversion subsystem were explored and, base in the performances found in this analysis, an Indirect Coupling (IC) between the LMR and the AMTEC/TE converters with Alkali Metal Boilers (AMB) was chosen as the reference design. The performance model of the fully integrated sodium-and potassium-AMTEC/TE converters shows that a combined conversion efficiency in excess of 30% could be achieved by the plant. (B204)

  7. Safety Re-evaluation of Kyoto University Research Reactor by reflecting the Accident of Fukushima Daiichi Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Nakajima, K.; Yamamoto, T. [Kyoto Univ., Kyoto (Japan)


    Kyoto University Research Reactor (KUR) is a light-water moderated tank-type reactor operated at rated thermal power of 5MW. After the accident of Fukushima Daiichi nuclear power plant, we have settled a 40-ton water tank near the reactor room, and prepared a mobile fire pump and a mobile power generator as additional safety measures for beyond design basis accidents (BDBAs). We also have conducted the safety re-evaluation of KUR, and confirmed that the integrity of KUR fuels could be kept against the BDBA with the use of the additional safety measures when the several restrictions were imposed on the reactor operation.

  8. Space and Terrestrial Power System Integration Optimization Code BRMAPS for Gas Turbine Space Power Plants With Nuclear Reactor Heat Sources (United States)

    Juhasz, Albert J.


    In view of the difficult times the US and global economies are experiencing today, funds for the development of advanced fission reactors nuclear power systems for space propulsion and planetary surface applications are currently not available. However, according to the Energy Policy Act of 2005 the U.S. needs to invest in developing fission reactor technology for ground based terrestrial power plants. Such plants would make a significant contribution toward drastic reduction of worldwide greenhouse gas emissions and associated global warming. To accomplish this goal the Next Generation Nuclear Plant Project (NGNP) has been established by DOE under the Generation IV Nuclear Systems Initiative. Idaho National Laboratory (INL) was designated as the lead in the development of VHTR (Very High Temperature Reactor) and HTGR (High Temperature Gas Reactor) technology to be integrated with MMW (multi-megawatt) helium gas turbine driven electric power AC generators. However, the advantages of transmitting power in high voltage DC form over large distances are also explored in the seminar lecture series. As an attractive alternate heat source the Liquid Fluoride Reactor (LFR), pioneered at ORNL (Oak Ridge National Laboratory) in the mid 1960's, would offer much higher energy yields than current nuclear plants by using an inherently safe energy conversion scheme based on the Thorium --> U233 fuel cycle and a fission process with a negative temperature coefficient of reactivity. The power plants are to be sized to meet electric power demand during peak periods and also for providing thermal energy for hydrogen (H2) production during "off peak" periods. This approach will both supply electric power by using environmentally clean nuclear heat which does not generate green house gases, and also provide a clean fuel H2 for the future, when, due to increased global demand and the decline in discovering new deposits, our supply of liquid fossil fuels will have been used up. This is

  9. Nuclear resurrection: Must Ontario fire up more reactors to power its future?

    Energy Technology Data Exchange (ETDEWEB)

    Dewar, E.


    An extensive historical review of Canada's nuclear reactor program is provided. The author also examines the role of nuclear power generation in Ontario's energy future, concluding that given the limited capacity for additional hydro power, and the uncertainty of natural gas supply, nuclear power will likely remain a significant source of energy for Ontario for the foreseeable future. Nevertheless, the challenge to bring nuclear power generation under control remains, considering that despite the best efforts of generations of nuclear engineers, politicians and regulators the industry appears close to being unmanageable, and Ontario taxpayers are likely to be paying its old debt far into the future. The current contingent of reactors is rapidly aging and the disposal of used nuclear fuel still defies a satisfactory solution. These formidable challenges notwithstanding, best estimates are that Ontario has few viable alternatives, and will have to embark on a new cycle of nuclear construction before the end of this decade.

  10. Power flattening on modified CANDLE small long life gas-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Monado, Fiber [Nuclear Physics and Biophysics Research Group, Dept. of Physics, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Bandung, Indonesia and Dept. of Physics, Faculty of Mathematics and Natural Sciences, Sriwijaya University (Indonesia); Su' ud, Zaki; Waris, Abdul; Basar, Khairul [Nuclear Physics and Biophysics Research Group, Dept. of Physics, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Bandung (Indonesia); Ariani, Menik [Dept. of Physics, Faculty of Mathematics and Natural Sciences, Sriwijaya University (Indonesia); Sekimoto, Hiroshi [CRINES, Tokyo Institute of Technology, O-okoyama, Meguro-ku, Tokyo 152-8550 (Japan)


    Gas-cooled Fast Reactor (GFR) is one of the candidates of next generation Nuclear Power Plants (NPPs) that expected to be operated commercially after 2030. In this research conceptual design study of long life 350 MWt GFR with natural uranium metallic fuel as fuel cycle input has been performed. Modified CANDLE burn-up strategy with first and second regions located near the last region (type B) has been applied. This reactor can be operated for 10 years without refuelling and fuel shuffling. Power peaking reduction is conducted by arranging the core radial direction into three regions with respectively uses fuel volume fraction 62.5%, 64% and 67.5%. The average power density in the modified core is about 82 Watt/cc and the power peaking factor decreased from 4.03 to 3.43.

  11. Oxygen transport membrane reactor based method and system for generating electric power

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, Sean M.; Chakravarti, Shrikar; Li, Juan


    A carbon capture enabled system and method for generating electric power and/or fuel from methane containing sources using oxygen transport membranes by first converting the methane containing feed gas into a high pressure synthesis gas. Then, in one configuration the synthesis gas is combusted in oxy-combustion mode in oxygen transport membranes based boiler reactor operating at a pressure at least twice that of ambient pressure and the heat generated heats steam in thermally coupled steam generation tubes within the boiler reactor; the steam is expanded in steam turbine to generate power; and the carbon dioxide rich effluent leaving the boiler reactor is processed to isolate carbon. In another configuration the synthesis gas is further treated in a gas conditioning system configured for carbon capture in a pre-combustion mode using water gas shift reactors and acid gas removal units to produce hydrogen or hydrogen-rich fuel gas that fuels an integrated gas turbine and steam turbine system to generate power. The disclosed method and system can also be adapted to integrate with coal gasification systems to produce power from both coal and methane containing sources with greater than 90% carbon isolation.

  12. Introduction to Nuclear Fusion Power and the Design of Fusion Reactors. An Issue-Oriented Module. (United States)

    Fillo, J. A.

    This three-part module focuses on the principles of nuclear fusion and on the likely nature and components of a controlled-fusion power reactor. The physical conditions for a net energy release from fusion and two approaches (magnetic and inertial confinement) which are being developed to achieve this goal are described. Safety issues associated…

  13. Controlling the power output of a nuclear reactor with fuzzy logic

    NARCIS (Netherlands)

    Ruan, D.; Wal, A.J. van der


    The application of fuzzy logic control (FLC) in the domain of nuclear industry presents a tremendous challenge. The main reason for this is the public awareness of the risks of nuclear reactors and the very strict safety regulations in force for nuclear power plants. The very same regulations preven

  14. Controlling the Power Output of a Nuclear Reactor with Fuzzy Logic

    NARCIS (Netherlands)

    Ruan, D.; Wal, A.J. van der


    The application of fuzzy logic control (FLC) in the domain of nuclear industry presents a tremendous challenge. The main reason for this is the public awareness of the risks of nuclear reactors and the very strict safety regulations in force for nuclear power plants. The very same regulations preven

  15. Thermal-hydraulics and safety analysis of sectored compact reactor for lunar surface power

    Energy Technology Data Exchange (ETDEWEB)

    Schriener, T. M. [Inst. for Space and Nuclear Power Studies, Univ. of New Mexico, Albuquerque, NM (United States); Chemical and Nuclear Engineering Dept., Univ. of New Mexico, Albuquerque, NM (United States); El-Genk, M. S. [Inst. for Space and Nuclear Power Studies, Univ. of New Mexico, Albuquerque, NM (United States); Chemical and Nuclear Engineering Dept., Univ. of New Mexico, Albuquerque, NM (United States); Mechanical Engineering Dept., Univ. of New Mexico, Albuquerque, NM (United States)


    The liquid NaK-cooled, fast-neutron spectrum, Sectored Compact Reactor (SCoRe-N 5) concept has been developed at the Univ. of New Mexico for lunar surface power applications. It is loaded with highly enriched UN fuel pins in a triangular lattice, and nominally operates at exit and inlet coolant temperatures of 850 K and 900 K. This long-life reactor generates up to 1 MWth continuously for {>=} 20 years. To avoid a single point failure in reactor cooling, the core is divided into 6 sectors that are neutronically and thermally coupled, but hydraulically independent. This paper performs a 3-D the thermal-hydraulic analysis of SCoRe--N 5 at nominal operation temperatures and a power level of 1 MWth. In addition, the paper investigates the potential of continuing reactor operation at a lower power in the unlikely event that one sector in the core experiences a loss of coolant (LOC). Redesigning the core with a contiguous steel matrix enhances the cooling of the sector experiencing a LOC. Results show that with a core sector experiencing a LOC, SCORE-N 5 could continue operating safely at a reduced power of 166.6 kWth. (authors)

  16. Gaseous-fuel nuclear reactor research for multimegawatt power in space (United States)

    Thom, K.; Schneider, R. T.; Helmick, H. H.


    In the gaseous-fuel reactor concept, the fissile material is contained in a moderator-reflector cavity and exists in the form of a flowing gas or plasma separated from the cavity walls by means of fluid mechanical forces. Temperatures in excess of structural limitations are possible for low-specific-mass power and high-specific-impulse propulsion in space. Experiments have been conducted with a canister filled with enriched UF6 inserted into a beryllium-reflected cavity. A theoretically predicted critical mass of 6 kg was measured. The UF6 was also circulated through this cavity, demonstrating stable reactor operation with the fuel in motion. Because the flowing gaseous fuel can be continuously processed, the radioactive waste in this type of reactor can be kept small. Another potential of fissioning gases is the possibility of converting the kinetic energy of fission fragments directly into coherent electromagnetic radiation, the nuclear pumping of lasers. Numerous nuclear laser experiments indicate the possibility of transmitting power in space directly from fission energy. The estimated specific mass of a multimegawatt gaseous-fuel reactor power system is from 1 to 5 kg/kW while the companion laser-power receiver station would be much lower in specific mass.

  17. Experimental detailed power distribution in a fast spectrum thermionic reactor fuel element at the core/BeO reflector interface region (United States)

    Klann, P. G.; Lantz, E.


    A zero-power critical assembly was designed, constructed, and operated for the prupose of conducting a series of benchmark experiments dealing with the physics characteristics of a UN-fueled, Li-7-cooled, Mo-reflected, drum-controlled compact fast reactor for use with a space-power conversion system. The critical assembly was modified to simulate a fast spectrum advanced thermionics reactor by: (1) using BeO as a reflector in place of some of the existing molybdenum, (2) substituting Nb-1Zr tubing for some of the existing Ta tubing, and (3) inserting four full-scale mockups of thermionic type fuel elements near the core and BeO reflector boundary. These mockups were surrounded with a buffer zone having the equivalent thermionic core composition. In addition to measuring the critical mass of this thermionic configuration, a detailed power distribution in one of the thermionic element stages in the mixed spectrum region was measured. A power peak to average ratio of two was observed for this fuel stage at the midplane of the core and adjacent to the reflector. Also, the power on the outer surface adjacent to the BeO was slightly more than a factor of two larger than the power on the inside surface of a 5.08 cm (2.0 in.) high annular fuel segment with a 2.52 cm (0.993 in. ) o.d. and a 1.86 cm (0.731 in.) i.d.

  18. Dynamic neutronic and stability analysis of a burst mode, single cavity gas core reactor Brayton cycle space power system (United States)

    Dugan, Edward T.; Kutikkad, Kiratadas

    The conceptual, burst-mode gaseous-core reactor (GCR) space nuclear power system presently subjected to reactor-dynamics and system stability studies operates on a closed Brayton cycle, via disk MHD generator for energy conversion. While the gaseous fuel density power coefficient of reactivity is found to be capable of rapidly stabilizing the GCR system, the power of this feedback renders standard external reactivity insertions inadequate for significant power-level changes during normal operation.

  19. A modular gas-cooled cermet reactor system for planetary base power (United States)

    Jahshan, Salim N.; Borkowski, Jeffrey A.


    Fission nuclear power is foreseen as the source for electricity in planetary colonization and exploration. A six module gas-cooled, cermet-fueled reactor is proposed that can meet the design objectives. The highly enriched core is compact and can operate at high temperature for a long life. The helium coolant powers six modular Brayton cycles that compare favorably with the SP-100-based Brayton cycle.

  20. A neutron tomography facility at a low power research reactor

    CERN Document Server

    Körner, S; Von Tobel, P; Rauch, H


    Neutron radiography (NR) provides a very efficient tool in the field of non-destructive testing as well as for many applications in fundamental research. A neutron beam penetrating a specimen is attenuated by the sample material and detected by a two-dimensional (2D) imaging device. The image contains information about materials and structure inside the sample because neutrons are attenuated according to the basic law of radiation attenuation. Contrary to X-rays, neutrons can be attenuated by some light materials, as for example, hydrogen and boron, but penetrate many heavy materials. Therefore, NR can yield important information not obtainable by more traditional methods. Nevertheless, there are many aspects of structure, both quantitative and qualitative, that are not accessible from 2D transmission images. Hence, there is an interest in three-dimensional neutron imaging. At the 250 kW TRIGA Mark II reactor of the Atominstitut in Austria a neutron tomography facility has been installed. The neutron flux at ...

  1. Optimization of the self-sufficient thorium fuel cycle for CANDU power reactors

    Directory of Open Access Journals (Sweden)

    Bergelson Boris R.


    Full Text Available The results of optimization calculations for CANDU reactors operating in the thorium cycle are presented in this paper. Calculations were performed to validate the feasibility of operating a heavy-water thermal neutron power reactor in a self-sufficient thorium cycle. Two modes of operation were considered in the paper: the mode of preliminary accumulation of 233U in the reactor itself and the mode of operation in a self-sufficient cycle. For the mode of accumulation of 233U, it was assumed that enriched uranium or plutonium was used as additional fissile material to provide neutrons for 233U production. In the self-sufficient mode of operation, the mass and isotopic composition of heavy nuclei unloaded from the reactor should provide (after the removal of fission products the value of the multiplication factor of the cell in the following cycle K>1. Additionally, the task was to determine the geometry and composition of the cell for an acceptable burn up of 233U. The results obtained demonstrate that the realization of a self-sufficient thorium mode for a CANDU reactor is possible without using new technologies. The main features of the reactor ensuring a self-sufficient mode of operation are a good neutron balance and moving of fuel through the active core.

  2. Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production

    Energy Technology Data Exchange (ETDEWEB)

    Mac Donald, Philip Elsworth; Buongiorno, Jacopo; Davis, Cliff Bybee; Weaver, Kevan Dean


    The use of supercritical temperature and pressure light water as the coolant in a direct-cycle nuclear reactor offers potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies of the power conversion cycle are possible (i.e. up to 46%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type recirculation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed. If a tight fuel rod lattice is adopted, it is possible to significantly reduce the neutron moderation and attain fast neutron energy spectrum conditions. In this project a supercritical water reactor concept with a simple, blanket-free, pancake-shaped core will be developed. This type of core can make use of either fertile or fertile-free fuel and retain the hard spectrum to effectively burn plutonium and minor actinides from LWR spent fuel while efficiently generating electricity.

  3. Space Molten Salt Reactor Concept for Nuclear Electric Propulsion and Surface Power (United States)

    Eades, M.; Flanders, J.; McMurray, N.; Denning, R.; Sun, X.; Windl, W.; Blue, T.

    Students at The Ohio State University working under the NASA Steckler Grant sought to investigate how molten salt reactors with fissile material dissolved in a liquid fuel medium can be applied to space applications. Molten salt reactors of this kind, built for non-space applications, have demonstrated high power densities, high temperature operation without pressurization, high fuel burn up and other characteristics that are ideal for space fission systems. However, little research has been published on the application of molten salt reactor technology to space fission systems. This paper presents a conceptual design of the Space Molten Salt Reactor (SMSR), which utilizes molten salt reactor technology for Nuclear Electric Propulsion (NEP) and surface power at the 100 kWe to 15 MWe level. Central to the SMSR design is a liquid mixture of LiF, BeF2 and highly enriched U235F4 that acts as both fuel and core coolant. In brief, some of the positive characteristics of the SMSR are compact size, simplified core design, high fuel burn up percentages, proliferation resistant features, passive safety mechanisms, a considerable body of previous research, and the possibility for flexible mission architecture.

  4. Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels

    CERN Document Server

    American Society for Testing and Materials. Philadelphia


    1.1 This practice covers procedures for designing a surveillance program for monitoring the radiation-induced changes in the mechanical properties of ferritic materials in light-water moderated nuclear power reactor vessels. This practice includes the minimum requirements for the design of a surveillance program, selection of vessel material to be included, and the initial schedule for evaluation of materials. 1.2 This practice was developed for all light-water moderated nuclear power reactor vessels for which the predicted maximum fast neutron fluence (E > 1 MeV) at the end of license (EOL) exceeds 1 × 1021 neutrons/m2 (1 × 1017 n/cm2) at the inside surface of the reactor vessel. 1.3 This practice applies only to the planning and design of surveillance programs for reactor vessels designed and built after the effective date of this practice. Previous versions of Practice E185 apply to earlier reactor vessels. 1.4 This practice does not provide specific procedures for monitoring the radiation induced cha...

  5. Oak Ridge Tokamak experimental power reactor study. Part 5. Engineering

    Energy Technology Data Exchange (ETDEWEB)

    Shannon, T.E.; Reid, R.L. (eds.)


    The results of the past year's effort in the engineering areas are documented in this report. The discussion covers system identification and overall description, power conversion, plant layout, structural design, remote maintenance and assembly, program planning, and schedules. Five companion reports being issued as parts of this composite report present information in the other discipline areas.

  6. Development of TSC SAMG for Nuclear Power Reactor

    Institute of Scientific and Technical Information of China (English)

    HUANGDong-xing; PUSheng-di; ZHAOShou-zhi; LIJi-gen


    Severe Accident Management Guidance (SAMG) is developed to prevent the severe accident progress and mitigate the consequences of severe accident in nuclear power plant. Technical Support Center (TSC) SAMG is one of important components of the strategies. It contains Severe Accident Guidelines (SAGs) and Severe Challenge Guidelines (SCGs).

  7. Continuous preparation of nanoscale zero-valent iron using impinging stream-rotating packed bed reactor and their application in reduction of nitrobenzene (United States)

    Jiao, Weizhou; Qin, Yuejiao; Luo, Shuai; Feng, Zhirong; Liu, Youzhi


    Nanoscale zero-valent iron (nZVI) was continuously prepared by high-gravity reaction precipitation through a novel impinging stream-rotating packed bed (IS-RPB). Reactant solutions of FeSO4 and NaBH4 were conducted into the IS-RPB with flow rates of 60 L/h and rotating speed of 1000 r/min for the preparation of nZVI. As-prepared nZVI obtained by IS-RPB were quasi-spherical morphology and almost uniformly distributed with a particle size of 10-20 nm. The reactivity of nZVI was estimated by the degradation of 100 ml nitrobenzene (NB) with initial concentration of 250 mg/L. The optimum dosage of nZVI obtained by IS-RPB was 4.0 g/L as the NB could be completely removed within 10 min, which reduced 20% compared with nZVI obtained by stirred tank reactor (STR). The reduction of NB and production of aniline (AN) followed pseudo-first-order kinetics, and the pseudo-first-order rate constants were 0.0147 and 0.0034 s-1, respectively. Furthermore, the as-prepared nZVI using IS-RPB reactor in this work can be used within a relatively wide range pH of 1-9.

  8. Review on Application of Control Algorithms to Power Regulations of Reactor Cores



    This research is to solve the stability analysis issue of nonlinear pressurized water reactor cores. On the basis of modeling a nonlinear pressurized water reactor core using the lumped parameter method, its linearized model is achieved via the small perturbation linearization way. Linearized models of the nonlinear core at six power levels are selected as local models of this core. The T-S fuzzy idea for the core is exploited to construct the T-S fuzzy model of the nonlinear core based on th...

  9. Electronic imaging system for neutron radiography at a low power research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ferreira, F.J.O., E-mail: [Instituto de Engenharia Nuclear, Comissao Nacional de Energia Nuclear, Caixa Postal 68550, CEP 21945-970, Rio de Janeiro (Brazil); Silva, A.X.; Crispim, V.R. [PEN/COPPE-DNC/POLI CT, Universidade Federal do Rio de Janeiro, Ilha do Fundao, Caixa Postal 68509, 21945-970 Rio de Janeiro (Brazil)


    This paper describes an electronic imaging system for producing real time neutron radiography from a low power research reactor, which will allow inspections of samples with high efficiency, in terms of measuring time and result analysis. This system has been implanted because of its potential use in various scientific and industrial areas where neutron radiography with photographic film could not be applied. This real time system is installed in neutron radiography facility of Argonauta nuclear research reactor, at the Instituto de Engenharia Nuclear of the Comissao Nacional de Energia Nuclear, in Brazil. It is adequate to perform real time neutron radiography of static and dynamic events of samples.

  10. Utilization of stable isotopes in power reactor; Utilisation des isotopes stables dans les reacteurs de puissance

    Energy Technology Data Exchange (ETDEWEB)

    Desmoulins, P. [Electricite de France (EDF), 75 - Paris (France)


    The stable isotopes, besides uranium, used in EDF power nuclear reactors are mainly the boron 10 and the lithium 7. Boron is used in reactors as a neutrophagous agent for core reactivity control, and lithium, and more especially lithium 7, is extensively used as a solution in PWR moderators for primary fluid pH control. Boron and lithium ore reserves and producers are presented; industrial isotopic separation techniques are described: for the boron 10, they include dissociative distillation (Sulzer process) and separation on anionic resins, and for lithium 7, ion exchange columns (Cogema). 1 tab.

  11. Nonlinear Adaptive Dynamic Output-Feedback Power-Level Control of Nuclear Heating Reactors

    Directory of Open Access Journals (Sweden)

    Zhe Dong


    Full Text Available Due to the high safety performance of small nuclear reactors, there is a promising future for small reactors. Nuclear heating reactor (NHR is a small reactor that has many advanced safety features such as the integrated arrangement, natural circulation at any power levels, self-pressurization, hydraulic control rod driving, and passive residual heating removing and can be applied to the fields of district heating, seawater desalination, and electricity production. Since the NHR dynamics has strong nonlinearity and uncertainty, it is meaningful to develop the nonlinear adaptive power-level control technique. From the idea of physically based control design method, a novel nonlinear adaptive power-level control is given for the NHR in this paper. It is theoretically proved that this newly built controller does not only provide globally asymptotic closed-loop stability but is also adaptive to the system uncertainty. Numerical simulation results show the feasibility of this controller and the relationship between the performance and controller parameters.

  12. Setting Limits On The Power Of A Geo-reactor With Kamland Detector

    CERN Document Server

    Maricic, J


    The Earth's magnetic field has existed for at least 3 billion years with high and on average stable intensity, though with many fluctuations and reversals. One of the models, albeit rather controversial, proposed as the energy source of the Earth's magnetic field is a natural nuclear reactor inside the Earth's core [1] and [2]. This author maintains that this is the only model that generates sufficient power to energize the geo-magnetic field for 3 billion years. Even more, the reactor's ability to produce variable power levels including stops and restarts in its operations, provides a viable explanation, according to [2], for the random reversals of the geo-magnetic field that have been recorded numerous times during the Earth's history. In this study, Kamioka Liquid scintillator Anti-Neutrino Detector (KamLAND) is used to set limits on the power of the putative geo-reactor. KamLAND is designed to detect anti-neutrinos from reactors around Japan, and thus can make a direct measurement of the anti-neutrino ra...

  13. Possibility of fusion power reactor to transmute minor actinides of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Serikov, A. E-mail:; Shatalov, G.; Sheludjakov, S.; Shpansky, Yu.; Vasiliev, N


    A possibility to use fusion power reactor (FPR) is considered for burning long-life elements of spent nuclear fuel in parallel with energy production. In this study a principal design of FPR blanket was examined for transmutation of long-life minor actinides (Np, Am, Cm). A production of minor actinide isotopes is equal to 20-30 kg/1 GW{sub (e)} year for now operating fission reactors, and their amounts will rise with the expected growth of fission reactor power. These isotopes have long-life time and can be dangerous in big amounts in future. Plutonium isotopes are not included in an assumption that they will be used in fission reactors. The major goals of the study were to determine FPR blanket composition corresponding to fast transmutation rate of actinides and tritium self-supply simultaneously. Tritium breeding ratio (TBR) was obtained at level 1.11 for water cooling and reached up 1.56 in variant with helium-cooled assemblies with Np nitride. It was concluded that rows with actinides from processed waste fuel should be arranged near the plasma first wall. Advantages of helium above water cooling are observed in the twice-increased loading of waste fissionable materials and essential increase of achievable TBR. Burnout of Np, Am, Cm would remain at a level {approx}40-50% after 4 full power years.

  14. NASA's Kilopower Reactor Development and the Path to Higher Power Missions (United States)

    Gibson, Marc A.; Oleson, Steven R.; Poston, David I.; McClure, Patrick


    The development of NASAs Kilopower fission reactor is taking large strides toward flight development with several successful tests completed during its technology demonstration trials. The Kilopower reactors are designed to provide 1-10 kW of electrical power to a spacecraft which could be used for additional science instruments as well as the ability to power electric propulsion systems. Power rich nuclear missions have been excluded from NASA proposals because of the lack of radioisotope fuel and the absence of a flight qualified fission system. NASA has partnered with the Department of Energy's National Nuclear Security Administration to develop the Kilopower reactor using existing facilities and infrastructure to determine if the design is ready for flight development. The 3-year Kilopower project started in 2015 with a challenging goal of building and testing a full-scale flight prototypic nuclear reactor by the end of 2017. As the date approaches, the engineering team shares information on the progress of the technology as well as the enabling capabilities it provides for science and human exploration.

  15. Nuclear design of the burst power ultrahigh temperature UF4 vapor core reactor system (United States)

    Kahook, Samer D.; Dugan, Edward T.


    Static and dynamic neutronic analyses are being performed, as part of an integrated series of studies, on an innovative burst power UF4 Ultrahigh Temperature Vapor Core Reactor (UTVR)/Disk Magnetohydrodynamic (MHD) generator for space nuclear power applications. This novel reactor concept operates on a direct, closed Rankine cycle in the burst power mode (hundreds of MWe for thousands of seconds). The fuel/working fluid is a mixture of UF4 and metal fluoride. Preliminary calculations indicate high overall system efficiencies (≊20%), small radiator size (≊5 m2/MWe), and high specific power (≊5 kWe/kg). Neutronic analysis has revealed a number of attractive features for this novel reactor concept. These include some unique and very effective inherent negative reactivity control mechanisms such as the vapor-fuel density power coefficient of reactivity, the direct neutronic coupling among the multiple fissioning core regions (the central vapor core and the surrounding boiler columns), and the mass flow coupling feedback between the fissioning cores.

  16. Vital area identification for U.S. Nuclear Regulatory Commission nuclear power reactor licensees and new reactor applicants.

    Energy Technology Data Exchange (ETDEWEB)

    Whitehead, Donnie Wayne; Varnado, G. Bruce


    U.S. Nuclear Regulatory Commission nuclear power plant licensees and new reactor applicants are required to provide protection of their plants against radiological sabotage, including the placement of vital equipment in vital areas. This document describes a systematic process for the identification of the minimum set of areas that must be designated as vital areas in order to ensure that all radiological sabotage scenarios are prevented. Vital area identification involves the use of logic models to systematically identify all of the malicious acts or combinations of malicious acts that could lead to radiological sabotage. The models available in the plant probabilistic risk assessment and other safety analyses provide a great deal of the information and basic model structure needed for the sabotage logic model. Once the sabotage logic model is developed, the events (or malicious acts) in the model are replaced with the areas in which the events can be accomplished. This sabotage area logic model is then analyzed to identify the target sets (combinations of areas the adversary must visit to cause radiological sabotage) and the candidate vital area sets (combinations of areas that must be protected against adversary access to prevent radiological sabotage). Any one of the candidate vital area sets can be selected for protection. Appropriate selection criteria will allow the licensee or new reactor applicant to minimize the impacts of vital area protection measures on plant safety, cost, operations, or other factors of concern.

  17. Adaptive fuzzy control of neutron power of the TRIGA Mark III reactor; Control difuso adaptable de la potencia neutronica del reactor Triga Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Rojas R, E.


    The design and implementation of an identification and control scheme of the TRIGA Mark III research nuclear reactor of the Instituto Nacional de Investigaciones Nucleares (ININ) of Mexico is presented in this thesis work. The identification of the reactor dynamics is carried out using fuzzy logic based systems, in which a learning process permits the adjustment of the membership function parameters by means of techniques based on neural networks and bio-inspired algorithms. The resulting identification system is a useful tool that allows the emulation of the reactor power behavior when different types of insertions of reactivity are applied into the core. The identification of the power can also be used for the tuning of the parameters of a control system. On the other hand, the regulation of the reactor power is carried out by means of an adaptive and stable fuzzy control scheme. The control law is derived using the input-output linearization technique, which permits the introduction of a desired power profile for the plant to follow asymptotically. This characteristic is suitable for managing the ascent of power from an initial level n{sub o} up to a predetermined final level n{sub f}. During the increase of power, a constraint related to the rate of change in power is considered by the control scheme, thus minimizing the occurrence of a safety reactor shutdown due to a low reactor period value. Furthermore, the theory of stability in the sense of Lyapunov is used to obtain a supervisory control law which maintains the power error within a tolerance region, thus guaranteeing the stability of the power of the closed loop system. (Author)

  18. Decommissioning strategy and schedule for a multiple reactor nuclear power plant site

    Energy Technology Data Exchange (ETDEWEB)

    Monteiro, Deiglys Borges; Moreira, Joao M.L.; Maiorino, Jose Rubens, E-mail:, E-mail:, E-mail: [Universidade Federal do ABC (CECS/UFABC), Santo Andre, SP (Brazil). Centro de Engenharia, Modelagem e Ciencias Aplicadas


    The decommissioning is an important part of every Nuclear Power Plant life cycle gaining importance when there are more than one plant at the same site due to interactions that can arise from the operational ones and a decommissioning plant. In order to prevent undesirable problems, a suitable strategy and a very rigorous schedule should implemented and carried. In this way, decommissioning tasks such as fully decontamination and dismantling of activated and contaminated systems, rooms and structures could be delayed, posing as an interesting option to multiple reactor sites. The present work aims to purpose a strategy and a schedule for the decommissioning of a multiple reactor site highlighting the benefits of delay operational tasks and constructs some auxiliary services in the site during the stand by period of the shutdown plants. As a case study, will be presented a three-reactor site which the decommissioning process actually is in planning stage and that should start in the next decade. (author)

  19. Testing of Passive Safety System Performance for Higher Power Advanced Reactors

    Energy Technology Data Exchange (ETDEWEB)

    brian G. Woods; Jose Reyes, Jr.; John Woods; John Groome; Richard Wright


    This report describes the results of NERI research on the testing of advanced passive safety performance for the Westinghouse AP1000 design. The objectives of this research were: (a) to assess the AP1000 passive safety system core cooling performance under high decay power conditions for a spectrum of breaks located at a variety of locations, (b) to compare advanced thermal hydraulic computer code predictions to the APEX high decay power test data and (c) to develop new passive safety system concepts that could be used for Generation IV higher power reactors.

  20. Axial power distribution calculation using a neural network in the nuclear reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Y. H.; Cha, K. H.; Lee, S. H. [Korea Electric Power Research Institute, Taejon (Korea, Republic of)


    This paper is concerned with an algorithm based on neural networks to calculate the axial power distribution using excore detector signals in the nuclear reactor core. The fundamental basis of the algorithm is that the detector response can be fairly accurately estimated using computational codes. In other words, the training set, which represents relationship between detector signals and axial power distributions, for the neural network can be obtained through calculations instead of measurements. Application of the new method to the Yonggwang nuclear power plant unit 3 (YGN-3) shows that it is superior to the current algorithm in place. 7 refs., 4 figs. (Author)

  1. Thermal analysis of heat and power plant with high temperature reactor and intermediate steam cycle

    Directory of Open Access Journals (Sweden)

    Fic Adam


    Full Text Available Thermal analysis of a heat and power plant with a high temperature gas cooled nuclear reactor is presented. The main aim of the considered system is to supply a technological process with the heat at suitably high temperature level. The considered unit is also used to produce electricity. The high temperature helium cooled nuclear reactor is the primary heat source in the system, which consists of: the reactor cooling cycle, the steam cycle and the gas heat pump cycle. Helium used as a carrier in the first cycle (classic Brayton cycle, which includes the reactor, delivers heat in a steam generator to produce superheated steam with required parameters of the intermediate cycle. The intermediate cycle is provided to transport energy from the reactor installation to the process installation requiring a high temperature heat. The distance between reactor and the process installation is assumed short and negligable, or alternatively equal to 1 km in the analysis. The system is also equipped with a high temperature argon heat pump to obtain the temperature level of a heat carrier required by a high temperature process. Thus, the steam of the intermediate cycle supplies a lower heat exchanger of the heat pump, a process heat exchanger at the medium temperature level and a classical steam turbine system (Rankine cycle. The main purpose of the research was to evaluate the effectiveness of the system considered and to assess whether such a three cycle cogeneration system is reasonable. Multivariant calculations have been carried out employing the developed mathematical model. The results have been presented in a form of the energy efficiency and exergy efficiency of the system as a function of the temperature drop in the high temperature process heat exchanger and the reactor pressure.


    Energy Technology Data Exchange (ETDEWEB)

    M. G. McKellar; E. A. Harvego; A. M. Gandrik


    An updated reference design for a commercial-scale high-temperature electrolysis (HTE) plant for hydrogen production has been developed. The HTE plant is powered by a high-temperature gas-cooled reactor (HTGR) whose configuration and operating conditions are based on the latest design parameters planned for the Next Generation Nuclear Plant (NGNP). The current HTGR reference design specifies a reactor power of 600 MWt, with a primary system pressure of 7.0 MPa, and reactor inlet and outlet fluid temperatures of 322°C and 750°C, respectively. The reactor heat is used to produce heat and electric power to the HTE plant. A Rankine steam cycle with a power conversion efficiency of 44.4% was used to provide the electric power. The electrolysis unit used to produce hydrogen includes 1.1 million cells with a per-cell active area of 225 cm2. The reference hydrogen production plant operates at a system pressure of 5.0 MPa, and utilizes a steam-sweep system to remove the excess oxygen that is evolved on the anode (oxygen) side of the electrolyzer. The overall system thermal-to-hydrogen production efficiency (based on the higher heating value of the produced hydrogen) is 42.8% at a hydrogen production rate of 1.85 kg/s (66 million SCFD) and an oxygen production rate of 14.6 kg/s (33 million SCFD). An economic analysis of this plant was performed with realistic financial and cost estimating The results of the economic analysis demonstrated that the HTE hydrogen production plant driven by a high-temperature helium-cooled nuclear power plant can deliver hydrogen at a competitive cost. A cost of $3.03/kg of hydrogen was calculated assuming an internal rate of return of 10% and a debt to equity ratio of 80%/20% for a reactor cost of $2000/kWt and $2.41/kg of hydrogen for a reactor cost of $1400/kWt.

  3. Reliability and safety of the electrical power supply complex of the Hanford production reactors

    Energy Technology Data Exchange (ETDEWEB)

    Robbins, F.D.


    Safety has been and must continue to be the inviolable modulus by which the operation of a nuclear reactor must be judged. A malfunction in any reactor may well result in a release of fission products which may dissipate over a wide geographical area. Such dissipation may place the health, happiness and even the lives of the people in the region in serious jeopardy. As a result, the property damage and liability cost may reach astronomical values in the order of magnitude of billions of dollars. Reliability of the electrical network is an indispensable factor in attaining a high order of safety assurance. Progress in the peaceful use of atomic energy may take the form of electrical power generation using the nuclear reactor as a source of thermal energy. In view of these factors it seems appropriate and profitable that a critical engineering study be made of the safety and reliability of the Hanford reactors without regard to cost economics. This individual and independent technical engineering analysis was made without regard to Hanford traditional engineering and administration assignments. The main objective has been to focus attention on areas which seem to merit further detailed study on conditions which seem to need adjustment but most of all on those changes which will improve reactor safety. This report is the result of such a study.

  4. Study on the selection of nuclear fuel type for a hybrid power extraction reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Dong Han; Park, Won Suk [Korea Atomic Energy Research Institute, Taejon (Korea)


    The development of a subcritical transmutation reactor concept is emerging for reducing the amounts of actinides and long-lived nuclides in the spent fuel from nuclear power plants. This technology may make contribution to reduce the human risks associated with constructing radio-waste disposal facilities. One of the important issues for the design of the reactor is the selection of a suitable nuclear fuel type. Choosing the best nuclear fuel type for the reactor may not be easy since there exist several criteria associated with neutronic aspects, thermal performance, safety problem, cost problem, radiation damage in the reactor, etc. The best option should be chosen based on the maximization of our needs in this situation. This study presents a logical decision model for this issue using an analytic hierarchy process (AHP). Hierarchy is a representation of a system to study the functional relations of its components and its impact on the entire system. The study shows first how to construct hierarchy representing their relations and then measure the individual element's impact to the entire system for a quantitative decision making. Current four fuel types; metal, oxide, molten salt, and nitride, were selected and analyzed based on several characteristics with respect to overall comparison. Based on the decision model developed, the study concludes that the metal fuel type is the best choice for the transmutation reactor. The proposed approach is intended to help people be rational and logical in making decisions such complex task. 13 refs., 16 figs., 16 tabs. (Author)

  5. Wireless power transmission for biomedical implants: The role of near-zero threshold CMOS rectifiers. (United States)

    Mohammadi, Ali; Redoute, Jean-Michel; Yuce, Mehmet R


    Biomedical implants require an electronic power conditioning circuitry to provide a stable electrical power supply. The efficiency of wireless power transmission is strongly dependent on the power conditioning circuitry specifically the rectifier. A cross-connected CMOS bridge rectifier is implemented to demonstrate the impact of thresholds of rectifiers on wireless power transfer. The performance of the proposed rectifier is experimentally compared with a conventional Schottky diode full wave rectifier over 9 cm distance of air and tissue medium between the transmitter and receiver. The output voltage generated by the CMOS rectifier across a 1 KΩ resistive load is around twice as much as the Schottky rectifier.

  6. Model of punctual kinetic for studies on fast reactor stability; Modelo de cinetica pontual para estudos de estabilidade de reatores rapidos

    Energy Technology Data Exchange (ETDEWEB)

    Rocamora, Francisco Dias Jr.; Rosa, Mauricio A. Pinheiro; Braz Filho, Francisco A.; Borges, Eduardo M.; Guimaraes, Lamartine


    The neutron kinetics equations are used to obtain the Zero Power Transfer Function which establishes a relationship between a reactor core reactivity perturbation and the corresponding reactor power response. This transfer function should be coupled with those obtained from the fuel element and coolant thermal-hydraulics models in order to study fast reactor stability 'in the small'. (author)

  7. Nonlinear control for core power of pressurized water nuclear reactors using constant axial offset strategy

    Energy Technology Data Exchange (ETDEWEB)

    Ansarifar, Gholam Reza; Saadatzi, Saeed [Dept. of Nuclear Engineering, Faculty of Advanced Sciences and Technology, University of Isfahan, Isfahan (Iran, Islamic Republic of)


    One of the most important operations in nuclear power plants is load following, in which an imbalance of axial power distribution induces xenon oscillations. These oscillations must be maintained within acceptable limits otherwise the nuclear power plant could become unstable. Therefore, bounded xenon oscillation is considered to be a constraint for the load following operation. In this paper, the design of a sliding mode control (SMC), which is a robust nonlinear controller, is presented. SMC is a means to control pressurized water nuclear reactor (PWR) power for the load following operation problem in a way that ensures xenon oscillations are kept bounded within acceptable limits. The proposed controller uses constant axial offset (AO) strategy to ensure xenon oscillations remain bounded. The constant AO is a robust state constraint for the load following problem. The reactor core is simulated based on the two-point nuclear reactor model with a three delayed neutron groups. The stability analysis is given by means of the Lyapunov approach, thus the control system is guaranteed to be stable within a large range. The employed method is easy to implement in practical applications and moreover, the SMC exhibits the desired dynamic properties during the entire output-tracking process independent of perturbations. Simulation results are presented to demonstrate the effectiveness of the proposed controller in terms of performance, robustness, and stability. Results show that the proposed controller for the load following operation is so effective that the xenon oscillations are kept bounded in the given region.

  8. An evaluation of the ecological consequences of partial-power operation of the K Reactor, SRS

    Energy Technology Data Exchange (ETDEWEB)

    Gladden, J.B.; Mackey, H.E.; Paller, M.H.; Specht, W.L.; Wike, L.D.; Wilde, E.W.


    The K Reactor at the Savannah River Site (SRS) shut-down in spring 1988 for maintenance and safety upgrades. Since that time the receiving stream for thermal effluent, Indian Grave Branch and Pen Branch, have undergone a pattern of post-thermal recovery that is typical of other SRS streams following removal of thermal stress. Divesity of fish and aquatic macroinvertebrate communities has increased and available habitats have been colonized by numerous species of herbaceous and woody plants. K Reactor is scheduled to resume operation in 1991 and operate through 1992 without a cooling tower to cool the discharge. It is likely that the reactor will operate at approximately one-third to one-half of full power (800--1200 MW thermal) during this period and effluent temperatures will be substantially lower than earlier operation at full power. Monthly average discharge temperatures at half-power operation will range from approximately 42{degrees}C in winter to 49{degrees}C in summer. The volume of water discharged will not be affected by altered power levels and will average approximately 10--11 m{sup 3}/s. The ecological consequences of this mode of operation on the Indian Grave/Pen Branch stream system have been evaluated.

  9. Fiscal impacts associated with power reactor siting: a paired case study

    Energy Technology Data Exchange (ETDEWEB)

    Bjornstad, D.J.


    The paper examines the fiscal impacts associated with siting nuclear-powered electrical stations. First, a framework for examining fiscal impacts is constructed. This framework consists of four elements: the ability of a local community to raise revenues, the degree to which this ability is used, the uses to which tax revenues are applied, and the effect of tax/expenditure decisions on the local economy. Changes in these four elements caused by the siting are termed fiscal impacts. Second, this framework is applied to two communities, Waterford, Connecticut and Plymouth, Massachusetts, which host operating reactors. In each community the ability to raise revenues through the property tax--the prime local revenue source--approximately doubled. As a result both communities chose ultimately to reduce tax rates. Moreover, it appears that the annual revenues raised through the public sector as a result of the reactor siting exceeded income changes that resulted from increased local employment associated with each reactor's operation. It therefore appears that for these two towns, the primary economic impact occurred through the public sector. The report concludes with suggestions for further research into local fiscal and economic effects associated with power reactor siting.

  10. Testing of an Integrated Reactor Core Simulator and Power Conversion System with Simulated Reactivity Feedback (United States)

    Bragg-Sitton, Shannon M.; Hervol, David S.; Godfroy, Thomas J.


    A Direct Drive Gas-Cooled (DDG) reactor core simulator has been coupled to a Brayton Power Conversion Unit (BPCU) for integrated system testing at NASA Glenn Research Center (GRC) in Cleveland, Ohio. This is a closed-cycle system that incorporates an electrically heated reactor core module, turboalternator, recuperator, and gas cooler. Nuclear fuel elements in the gas-cooled reactor design are replaced with electric resistance heaters to simulate the heat from nuclear fuel in the corresponding fast spectrum nuclear reactor. The thermodynamic transient behavior of the integrated system was the focus of this test series. In order to better mimic the integrated response of the nuclear-fueled system, a simulated reactivity feedback control loop was implemented. Core power was controlled by a point kinetics model in which the reactivity feedback was based on core temperature measurements; the neutron generation time and the temperature feedback coefficient are provided as model inputs. These dynamic system response tests demonstrate the overall capability of a non-nuclear test facility in assessing system integration issues and characterizing integrated system response times and response characteristics.

  11. Advanced Fusion Reactors for Space Propulsion and Power Systems (United States)

    Chapman, John J.


    In recent years the methodology proposed for conversion of light elements into energy via fusion has made steady progress. Scientific studies and engineering efforts in advanced fusion systems designs have introduced some new concepts with unique aspects including consideration of Aneutronic fuels. The plant parameters for harnessing aneutronic fusion appear more exigent than those required for the conventional fusion fuel cycle. However aneutronic fusion propulsion plants for Space deployment will ultimately offer the possibility of enhanced performance from nuclear gain as compared to existing ionic engines as well as providing a clean solution to Planetary Protection considerations and requirements. Proton triggered 11Boron fuel (p- 11B) will produce abundant ion kinetic energy for In-Space vectored thrust. Thus energetic alpha particles "exhaust" momentum can be used directly to produce high ISP thrust and also offer possibility of power conversion into electricity. p- 11B is an advanced fusion plant fuel with well understood reaction kinematics but will require some new conceptual thinking as to the most effective implementation.

  12. High power millimeter wave experiment of ITER relevant electron cyclotron heating and current drive system. (United States)

    Takahashi, K; Kajiwara, K; Oda, Y; Kasugai, A; Kobayashi, N; Sakamoto, K; Doane, J; Olstad, R; Henderson, M


    High power, long pulse millimeter (mm) wave experiments of the RF test stand (RFTS) of Japan Atomic Energy Agency (JAEA) were performed. The system consists of a 1 MW/170 GHz gyrotron, a long and short distance transmission line (TL), and an equatorial launcher (EL) mock-up. The RFTS has an ITER-relevant configuration, i.e., consisted by a 1 MW-170 GHz gyrotron, a mm wave TL, and an EL mock-up. The TL is composed of a matching optics unit, evacuated circular corrugated waveguides, 6-miter bends, an in-line waveguide switch, and an isolation valve. The EL-mock-up is fabricated according to the current design of the ITER launcher. The Gaussian-like beam radiation with the steering capability of 20°-40° from the EL mock-up was also successfully proved. The high power, long pulse power transmission test was conducted with the metallic load replaced by the EL mock-up, and the transmission of 1 MW/800 s and 0.5 MW/1000 s was successfully demonstrated with no arcing and no damages. The transmission efficiency of the TL was 96%. The results prove the feasibility of the ITER electron cyclotron heating and current drive system.

  13. Bromine-79 NQR for uncoordinated Br- ions in trans-[CoBr2(en)2][H5O2]Br2 (United States)

    Sasane, A.; Matsuda, T.; Honda, H.; Mori, Y.


    A single 79Br NQR line showing a frequency of 19.594 MHz at room temperature has been observed in the crystals of trans-[CoBr2(en)2] [ H5O2 ] Br2 and assigned to the Br - ions which are not coordinated to the central Co(III) atom. The electric field gradient (EFG) at the Br - nuclei arises from O-H • • • Br - hydrogen bond formation between the Br - ions and the terminal O - H hydrogen atoms in [ H5O2 ] + ions. The induced EFG is greater for the present bromine complex than that for the isostructural chlorine complex. A point charge model calculation explains well the relative magnitude of the EFG in the two crystals by introducing Sternheimer's antishielding factors for the halogen ions.

  14. Analysis of Possible Application of High-Temperature Nuclear Reactors to Contemporary Large-Output Steam Power Plants on Ships

    Directory of Open Access Journals (Sweden)

    Kowalczyk T.


    Full Text Available This paper is aimed at analysis of possible application of helium to cooling high-temperature nuclear reactor to be used for generating steam in contemporary ship steam-turbine power plants of a large output with taking into account in particular variable operational parameters. In the first part of the paper types of contemporary ship power plants are presented. Features of today applied PWR reactors and proposed HTR reactors are discussed. Next, issues of load variability of the ship nuclear power plants, features of the proposed thermal cycles and results of their thermodynamic calculations in variable operational conditions, are presented.

  15. First in-core simultaneous measurements of nuclear heating and thermal neutron flux obtained with the innovative mobile calorimeter CALMOS inside the OSIRIS reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lepeltier, Valerie; Bubendorff, Jacques; Carcreff, Hubert [Nuclear studies and reactor irradiation Service, CEA Saclay 91191 Gif sur Yvette (France); Salmon, Laurent [Thermalhydraulics and Fluid Mechanics Section, CEA Saclay 91191 Gif sur Yvette, (France)


    Nuclear heating inside a MTR reactor has to be known in order to design and to run irradiation experiments which have to fulfill target temperature constraints. This measurement is usually carried out by calorimetry. The innovative calorimetric system, CALMOS, has been studied and built in 2011 for the 70 MWth OSIRIS reactor operated by CEA. Thanks to a new type of calorimetric probe, associated to a specific displacement system, it provides measurements along the fissile height and above the core. This development required preliminary modelling and irradiation of mock-ups of the calorimetric probe in the ex-core area, where nuclear heating rate does not exceed 2 W.g{sup -1}. The calorimeter working modes, the different measurement procedures allowed with such a new probe, the main modeling and experimental results and expected advantages of this new technique have been already presented. However, these first in-core measurements were not performed beyond 6 W.g{sup -1}, due to an inside temperature limitation imposed by a safety authority requirement. In this paper, we present the first in-core simultaneous measurements of nuclear heating and conventional thermal neutron flux obtained by the CALMOS device at the 70 MW nominal reactor power. For the first time, this experimental system was operated in nominal in-core conditions, with nominal neutron flux up to 2.7 10{sup 14}{sup -2}.s{sup -1} and nuclear heating up to 12 W.g{sup -1}. A comprehensive measurement campaign carried out from 2013 to 2015 inside all accessible irradiation locations of the core, allowed to qualify definitively this new device, not only in terms of measurement ability but also in terms of reliability. After a brief reminder of the calorimetric cell configuration and displacement system specificities, first nuclear heating distributions at nominal power are presented and discussed. In order to reinforce the heating evaluation, a systematic comparison is made between results obtained by

  16. The nuclear battery: a very small reactor power supply for remote locations

    Energy Technology Data Exchange (ETDEWEB)

    Kozier, K.S. (Atomic Energy of Canada Ltd., Pinawa, MB (Canada). Whiteshell Nuclear Research Establishment)

    The Nuclear Battery is a small reactor power supply being developed by Atomic Energy of Canada Limited for use in locations that are remote from utility grids and natural gas pipelines. Key technical features of the Nuclear Battery reactor core include a heat-pipe primary heat transport system, graphite neutron moderator, low-enriched uranium TRISO coated-particle fuel, and the use of burnable poisons for long-term reactivity control. An external secondary heat transport system extracts useful heat energy that may be converted into electricity in an organic Rankine cycle engine, or used to produce a high-pressure steam. The reference design is capable of producing about 2400 kW(t) (about 600 kW(e) net) for 15 full-power years without refueling. (author).

  17. The nuclear battery: a very small reactor power supply for remote locations; Technical note

    Energy Technology Data Exchange (ETDEWEB)

    Kozier, K.S. (Atomic Energy of Canada Ltd., Pinawa, MB (Canada). Whiteshell Labs.)


    The Nuclear Battery is a small reactor power supply being developed by Atomic Energy of Canada Limited for use in locations that are remote from utility grids and natural gas pipelines. Key technical features of the Nuclear Battery reactor core include a heat-pipe primary heat transport system, graphite neutron moderator, low enriched uranium TRISO coated-particle fuel and the use of burnable poisons for long-term reactivity control. An external secondary heat transport system extracts useful heat energy that may be converted into electricity in an organic Rankine cycle engine, or used to produce high-pressure steam. The reference design is capable of producing about 2400 kW(t) (about 500 kW(e) net) for 15 full-power years without refuelling. (orig.).

  18. Status of R&D Activities on Materials for Fusion Power Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Baluc, N.; Abe, K.; Boutard, J. L.; Chernov, V. M.; Diegele, E.; Jitsukawa, S.; Kimura, Akihiko; Klueh, R. L.; Kohyama, Akira; Kurtz, Richard J.; Lasser, R.; Matsui, H.; Moslang, A.; Muroga, T.; Odette, George R.; Tran, M. Q.; van der Schaaf, B.; Wu, Y.; Yu, J.; Zinkle, Steven J.


    Current R&D activities on materials for fusion power reactors are mainly focused on plasma facing, structural and tritium breeding materials for plasma facing (first wall, divertor) and breeding blanket components. Most of these activities are being performed in Europe, Japan, P.R. China, Russia and the USA. They relate to development of new high temperature, radiation resistant materials, development of coatings that shall act as erosion, corrosion, permeation or electrical/MHD barriers, characterization of the whole candidate materials in terms of mechanical and physical properties, assessment of irradiation effects, compatibility experiments, development of reliable joints, and development and/or validation of design rules. Priorities defined worldwide in the field of materials for fusion power reactors are summarized, as well as the main achievements obtained during the last few years and the near-term perspectives in the different investigation areas.

  19. Power up-grading study for the first Egyptian research reactor

    Energy Technology Data Exchange (ETDEWEB)

    El-Sawy Temraz, H.; Ashoub, N. E-mail:; Fathallah, A


    In the present work, power up-grading study is performed, for the first Egyptian Research Reactor (ET-RR-1), using the present fuel basket with 4x4 fuel rods, (17.5 mm pitch), and a proposed fuel basket with 5x5 fuel rods, (14.0 mm pitch), without violating the thermal hydraulic safety criteria. These safety criteria are; fuel centerline temperature (fuel melting), clad surface temperature (surface boiling), outlet coolant temperature, and maximum heat flux (critical heat flux ratio). Different thermal reactor powers (2-10 MW) and different core coolant flow rates (450, 900, 1350 m{sup 3} h{sup -1}) are considered. The thermal hydraulic analysis was performed using the subchannel code COBRA-IIIC for the estimation of temperatures, coolant velocities and critical heat flux. The neutronic calculations were performed using WIMS-D4 code with 5-group neutron cross section library. These cross sections were adapted to use in the two-dimensional (2-D) diffusion code DIXY for core calculations. The study concluded that ET-RR-1 power can be upgraded safely up to 4 MW with the present 4x4-fuel basket and with the proposed 5x5-fuel basket up to 5 MW with the present coolant flow rate (900 m{sup 3} h{sup -1}). With the two fuel arrays, the reactor power can be upgraded to 6 MW with coolant flow rate of 1350 m{sup 3} h{sup -1} without violating the safety criterion. It is also concluded that, loading the ET-RR-1 core with the proposed fuel basket (5x5) increases the excess reactivity of the reactor core than the present 4x4 fuel matrix with equal U-235 mass load and gave better fuel economy of fuel utilization.

  20. Reactor (United States)

    Evans, Robert M.


    1. A neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch.

  1. Thermal and neutron-physical features of the nuclear reactor for a power pulsation plant for space applications (United States)

    Gordeev, É. G.; Kaminskii, A. S.; Konyukhov, G. V.; Pavshuk, V. A.; Turbina, T. A.


    We have explored the possibility of creating small-size reactors with a high power output with the provision of thermal stability and nuclear safety under standard operating conditions and in emergency situations. The neutron-physical features of such a reactor have been considered and variants of its designs preserving the main principles and approaches of nuclear rocket engine technology are presented.

  2. Boiling water reactor in a prestressed reinforced concrete vessel for an atomic central heating-and-power plant

    Energy Technology Data Exchange (ETDEWEB)

    Tokarev, Yu.I.; Sokolov, I.N.; Skvortsov, S.A.; Sidorov, A.M.; Krauze, L.V.


    The possibility of using a boiling water reactor in a prestressed reinforced concrete vessel for an atomic central heating-and-power plant (CHPP) was considered, with design features of the reactor intended for a two-purpose plant. A prestressed reinforced concrete vessel and integral arrangement of the primary circuit ensured reliability of the atomic CHPP using various CHPP flowsheets.

  3. Preliminary Evaluation of Convective Heat Transfer in a Water Shield for a Surface Power Reactor (United States)

    Pearson J. Boise; Reid, Robert S.


    As part of the Vision for Space Exploration, the end of the next decade will bring man back to the surface of the moon. A crucial issue for the establishment of human presence on the moon will be the availability of compact power sources. This presence could require greater than 10's of kWt's in follow on years. Nuclear reactors are well suited to meet the needs for power generation on the lunar or Martian surface. Radiation shielding is a key component of any surface power reactor system. Several competing concepts exist for lightweight, safe, robust shielding systems such as a water shield, lithium hydride (LiH), and boron carbide. Water offers several potential advantages, including reduced cost, reduced technical risk, and reduced mass. Water has not typically been considered for space reactor applications because of the need for gravity to fix the location of any vapor that could form radiation streaming paths. The water shield concept relies on the predictions of passive circulation of the shield water by natural convection to adequately cool the shield. This prediction needs to be experimentally evaluated, especially for shields with complex geometries. NASA Marshall Space Flight Center has developed the experience and facilities necessary to do this evaluation in its Early Flight Fission - Test Facility (EFF-TF).

  4. The outlook for application of powerful nuclear thermionic reactor - powered space electric jet propulsion engines

    Energy Technology Data Exchange (ETDEWEB)

    Semyonov, Y.P.; Bakanov, Y.A.; Synyavsky, V.V.; Yuditsky, V.D. [Rocket-Space Corp. `Energia`, Moscow (Russian Federation)


    This paper summarizes main study results for application of powerful space electric jet propulsion unit (EJPUs) which is powered by Nuclear Thermionic Power Unit (NTPU). They are combined in Nuclear Power/Propulsion Unit (NPPU) which serves as means of spacecraft equipment power supply and spacecraft movement. Problems the paper deals with are the following: information satellites delivery and their on-orbit power supply during 10-15 years, removal of especially hazardous nuclear wastes, mining of asteroid resources and others. Evaluations on power/time/mass relationship for this type of mission are given. EJPU parameters are compatible with Russian existent or being under development launch vehicle. (author)

  5. Nickel Base Alloy Strip Electroslag Cladding of Nuclear Power Equipment%核电设备中的镍基合金带极电渣堆焊

    Institute of Scientific and Technical Information of China (English)



    以Inconel 690镍基合金堆焊为例,简述了堆焊工艺评定、模拟件试验和产品应用,结果表明:带极电渣堆焊可用于压水堆核电站核岛主设备蒸汽发生器管板大面积镍基合金堆焊.%Taking Inconel 690 nickel base alloy strip cladding for example,the cladding procedure qualification test, mock ?up test and product application were briefly described. Test results indicate that strip electroslag cladding technology can apply to large area nickel base alloy cladding on tube - sheet of steam generators for pressurized water reactor nuclear power plant nuclear island main equipment.

  6. Development of Liquid-Vapor Core Reactors with MHD Generator for Space Power and Propulsion Applications

    Energy Technology Data Exchange (ETDEWEB)

    Samim Anghaie


    Any reactor that utilizes fuel consisting of a fissile material in a gaseous state may be referred to as a gaseous core reactor (GCR). Studies on GCRs have primarily been limited to the conceptual phase, mostly due to budget cuts and program cancellations in the early 1970's. A few scientific experiments have been conducted on candidate concepts, primarily of static pressure fissile gas filling a cylindrical or spherical cavity surrounded by a moderating shell, such as beryllium, heavy water, or graphite. The main interest in this area of nuclear power generation is for space applications. The interest in space applications has developed due to the promise of significant enhancement in fuel utilization, safety, plant efficiency, special high-performance features, load-following capabilities, power conversion optimization, and other key aspects of nuclear power generation. The design of a successful GCR adapted for use in space is complicated. The fissile material studied in the pa st has been in a fluorine compound, either a tetrafluoride or a hexafluoride. Both of these molecules have an impact on the structural material used in the making of a GCR. Uranium hexafluoride as a fuel allows for a lower operating temperature, but at temperatures greater than 900K becomes essentially impossible to contain. This difficulty with the use of UF6 has caused engineers and scientists to use uranium tetrafluoride, which is a more stable molecule but has the disadvantage of requiring significantly higher operating temperatures. Gas core reactors have traditionally been studied in a steady state configuration. In this manner a fissile gas and working fluid are introduced into the core, called a cavity, that is surrounded by a reflector constructed of materials such as Be or BeO. These reactors have often been described as cavity reactors because the density of the fissile gas is low and criticality is achieved only by means of the reflector to reduce neutron leakage from the

  7. Management and construction of a major zero discharge water supply and treatment scheme at Mount Piper Power Station, Australia

    Energy Technology Data Exchange (ETDEWEB)

    Wong, P.; Docker, P.; Gabbrielli, E.; Wade, P. [Pacific Power, Sydney, NSW (Australia)


    Excellence in the organisation and management of a large multi-disciplinary design and construct project such as the zero discharge water supply and treatment scheme for Mount Piper coal-fired power station was essential to its successful execution. This paper discusses the management philosophies and strategies of Pacific Power and Transfield-PWT Asia/Pacific that resulted in this project being completed to the satisfaction of the client in the minimum time without delays and cost increases caused by interfaces between the parties. It highlights the advantages of contract packaging and placing the total responsibility for the performance and integration of a complex project system on one organisation. 2 refs., 6 figs.

  8. Three-Phase High-Power and Zero-Current-Switching OBC for Plug-In Electric Vehicles

    Directory of Open Access Journals (Sweden)

    Cheng-Shan Wang


    Full Text Available In this paper, an interleaved high-power zero-current-switching (ZCS onboard charger (OBC based on the three-phase single-switch buck rectifier is proposed for application to plug-in electric vehicles (EVs. The multi-resonant structure is used to achieve high efficiency and high power density, which are necessary to reduce the volume and weight of the OBC. This study focuses on the border conditions of ZCS converting with a battery load, which means the variation ranges of the output voltage and current are very large. Furthermore, a novel hybrid control method combining pulse frequency modulation (PFM and pulse width modulation (PWM together is presented to ensure a driving frequency higher than 10 kHz, and this will reduce the unexpected inner resonant power flow and decrease the total harmonic distortion (THD of the input current under a light load at the end of the charging process. Finally, a prototype is established, and experiments are carried out. According to the experimental results, the conversion efficiency is higher than 93.5%, the THD about 4.3% and power factor (PF 0.98 under the maximum power output condition. Besides, a three-stage charging process is also carried out the experimental platform.

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

    Energy Technology Data Exchange (ETDEWEB)

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


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

  10. Cutting Technology for Decommissioning of the Reactor Pressure Vessels in Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Kwan Seong; Kim, Geun Ho; Moon, Jei Kwon; Choi, Byung Seon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)


    Lots of nuclear power plants have been decommissioned during the last 2 decades. An essential part of this work is the dismantling of the Reactor Pressure Vessel and its Internals. For this purpose a wide variety of different cutting technologies have been developed, adapted and applied. A detailed introduction to Plasma Arc cutting, Contact Arc Metal cutting and Abrasive Water Suspension Jet cutting is given, as it turned out that these cutting technologies are particularly suitable for these type of segmentation work. A comparison of these technologies including gaseous emissions, cutting power, manipulator requirements as well as selected design approaches are given. Process limits as well as actual limits of application are presented

  11. Harmonics and voltage stability analysis in power systems including thyristor-controlled reactor

    Indian Academy of Sciences (India)

    M Uzunoglu


    In this study, non-sinusoidal quantities and voltage stability, both known as power quality criteria, are examined together in detail. The widespread use of power electronics elements cause the existence of significant non-sinusoidal quantities in the system. These non-sinusoidal quantities can create serious harmonic distortions in transmission and distribution systems. In this paper, harmonic generation of a static VAR compensator with thyristor-controlled reactor and effects of the harmonics on steady-state voltage stability are examined for various operational conditions.

  12. Parameter Identification with the Random Perturbation Particle Swarm Optimization Method and Sensitivity Analysis of an Advanced Pressurized Water Reactor Nuclear Power Plant Model for Power Systems

    Directory of Open Access Journals (Sweden)

    Li Wang


    Full Text Available The ability to obtain appropriate parameters for an advanced pressurized water reactor (PWR unit model is of great significance for power system analysis. The attributes of that ability include the following: nonlinear relationships, long transition time, intercoupled parameters and difficult obtainment from practical test, posed complexity and difficult parameter identification. In this paper, a model and a parameter identification method for the PWR primary loop system were investigated. A parameter identification process was proposed, using a particle swarm optimization (PSO algorithm that is based on random perturbation (RP-PSO. The identification process included model variable initialization based on the differential equations of each sub-module and program setting method, parameter obtainment through sub-module identification in the Matlab/Simulink Software (Math Works Inc., Natick, MA, USA as well as adaptation analysis for an integrated model. A lot of parameter identification work was carried out, the results of which verified the effectiveness of the method. It was found that the change of some parameters, like the fuel temperature and coolant temperature feedback coefficients, changed the model gain, of which the trajectory sensitivities were not zero. Thus, obtaining their appropriate values had significant effects on the simulation results. The trajectory sensitivities of some parameters in the core neutron dynamic module were interrelated, causing the parameters to be difficult to identify. The model parameter sensitivity could be different, which would be influenced by the model input conditions, reflecting the parameter identifiability difficulty degree for various input conditions.

  13. Impact of membrane characteristics on the performance and cycling of the Br-2-H-2 redox flow cell

    Energy Technology Data Exchange (ETDEWEB)

    Tucker, MC; Cho, KT; Spingler, FB; Weber, AZ; Lin, GY


    The Br-2/H-2 redox flow cell shows promise as a high-power, low-cost energy storage device. In this paper, the effect of various aspects of material selection and processing of proton exchange membranes on the operation of the Br-2/H-2 redox flow cell is determined. Membrane properties have a significant impact on the performance and efficiency of the system. In particular, there is a tradeoff between conductivity and crossover, where conductivity limits system efficiency at high current density and crossover limits efficiency at low current density. The impact of thickness, pretreatment procedure, swelling state during cell assembly, equivalent weight, membrane reinforcement, and addition of a microporous separator layer on this tradeoff is assessed. NR212 (50 mu m) pretreated by soaking in 70 degrees C water is found to be optimal for the studied operating conditions. For this case, an energy efficiency of greater than 75% is achieved for current density up to 400 mA cm(-2), with a maximum obtainable energy efficiency of 88%. A cell with this membrane was cycled continuously for 3164 h. Membrane transport properties, including conductivity and bromine and water crossover, were found to decrease moderately upon cycling but remained higher than those for the as-received membrane. (C) 2015 Elsevier B.V. All rights reserved.

  14. ORNL R and D on advanced small and medium power reactors: Selected topics

    Energy Technology Data Exchange (ETDEWEB)

    White, J.D.; Trauger, D.B.


    From 1984-1985, ORNL studied several innovative small and medium power nuclear concepts with respect to viability. Criteria for assessment of market attractiveness were developed and are described here. Using these criteria and descriptions of selected advanced reactor concepts, and assessment of their projected market viability in the time period 2000-2010 was made. All of these selected concepts could be considered as having the potential for meeting the criteria but, in most cases, considerable RandD would be required to reduce uncertainties. This work and later studies of safety and licensing of advanced, passively safe reactor concepts by ORNL are described. The results of these studies are taken into account in most of the current (FY 1989) work at ORNL on advanced reactors. A brief outline of this current work is given. One of the current RandD efforts at ORNL which addresses the operability and safety of advanced reactors is the Advanced Controls Program. Selected topics from this Program are described. 13 refs., 1 fig.

  15. Nuclear renaissance in the reactor training of Areva

    Energy Technology Data Exchange (ETDEWEB)

    De Braquilanges, Bertrand [Reactor Training Center/France Manager, La Tour Areva - 1, place Jean Millier - 92084 Paris - La Defense (France); Napior, Amy [Reactor Training Center/USA Manager, 1300 Old Graves Mill Road - Lynchburg VA, 2450 (United States); Schoenfelder, Christian [Reactor Training Center/Germany Manager, Kaiserleistrasse 29 - 63067 Offenbach (Germany)


    Because of the perspectives of new builds, a significant increase in the number of design, construction and management personnel working in AREVA, their clients and sub-contractors has been estimated for the next future. In order to cope with the challenge to integrate newly hired people quickly and effectively into the AREVA workforce, a project - 'Training Task Force (TTF)' - was launched in 2008. The objective was to develop introductory and advanced courses and related tools harmonized between AREVA Training Centers in France, Germany and USA. First, a Global Plants Introductory Session (GPIS) was developed for newly hired employees. GPIS is a two weeks training course introducing in a modular way AREVA and specifically the activities and the reactors technical basics. As an example, design and operation of a nuclear power plant is illustrated on EPRTM. Since January 2009, these GPIS are held regularly in France, Germany and the US with a mixing of employees from these 3 regions. Next, advanced courses for more experienced employees were developed: - Advanced EPR{sup TM}, giving a detailed presentation of the EPR{sup TM} reactor design; - Codes and Standards; - Technical Nuclear Safety. Finally, feasibility studies on a Training Material Management (TMM) system, able to manage the training documentation, and on a worldwide training administration tool, were performed. The TTF project was completed mid of 2009; it transferred their recurrent activities to a new AREVA training department. This unit now consists of the French, German and US Reactors Training Centers. In particular, all courses developed by the TTF are now implemented worldwide with an opening to external trainees. The current worldwide course catalogue includes training courses for operation and maintenance personnel as well as for managers, engineers and non technical personnel of nuclear operators, suppliers, safety authorities and expert organizations. Training delivery is supported

  16. Fuel rod behavior under normal operating conditions in Super Fast Reactor with high power density

    Energy Technology Data Exchange (ETDEWEB)

    Ju, Haitao, E-mail: [Science and Technology on Reactor System Design Technology Laboratory, Chengdu, Sichuan 610041 (China); Ishiwatari, Yuki [Department of Nuclear Engineering and Management, The University of Tokyo, Hongo, Bunkyo, Tokyo 113-8656 (Japan); Oka, Yoshiaki [Joint Department of Nuclear Energy, Waseda University, Totsukamachi, Shinjuku, Tokyo 169-8050 (Japan)


    Highlights: • The improved core of Super Fast Reactor with high power density is analyzed. • We analyzed four types of the limiting fuel rods. • The influence of Pu enrichment and compressive stress to yield strength ratio are analyzed. • The improved fuel rod design of the new core is suggested. - Abstract: A Super Fast Reactor is a pressure-vessel type, fast spectrum SuperCritical Water Reactor (SCWR) which is presently researched in a Japanese project. A preliminary core has an average power density of 158.8 W/cc. However one of the most important advantages of the Super Fast Reactor is the higher power density compared to the thermal spectrum SCWR, which reduces the capital cost. After the sensitivity analyses on the fuel rod configurations, the fuel assembly configurations and the core configurations, an improved core with an average power density of 294.8 W/cc is designed by 3-D neutronic/thermal-hydraulic coupled calculations. In order to ensure the fuel rod integrity of new core design with high power density, the fuel rod behaviors under normal operating condition are analyzed using fuel performance code FEMAXI-6. The power histories of each fuel rod are taken from the neutronics calculation results in the core design. The cladding surface temperature histories are generated from the thermal-hydraulic calculation results in the core design. Four types of the limiting fuel rods, individually with the Maximum Cladding Surface Temperature (MCST), Maximum Power Peak (MPP), Maximum Discharge Burnup (MDB) and Different Coolant Flow Pattern (DCFP), are chosen to cover all the fuel rods in the core. The available design range of the fuel rod design parameters, such as initial gas plenum pressure, gas plenum position, gas plenum length, grain size and gap size, are found out in order to satisfy the following design criteria: (1) Maximum fuel centerline temperature should be less than 1900 °C. (2) Maximum cladding stress in circumferential direction should

  17. 76 FR 17160 - Office of New Reactors; Final Interim Staff Guidance on the Review of Nuclear Power Plant Designs... (United States)


    ... COMMISSION Office of New Reactors; Final Interim Staff Guidance on the Review of Nuclear Power Plant Designs... Guidance (ISG) DC/COL-ISG-021 titled ``Interim Staff Guidance on the Review of Nuclear Power Plant Designs... Nuclear Power Plants,'' March 2007, Standard Review Plan (SRP), Section 8.3.1 and Sections 9.5.4 through...

  18. 75 FR 5632 - Office of New Reactors; Interim Staff Guidance on the Review of Nuclear Power Plant Designs Using... (United States)


    ... COMMISSION Office of New Reactors; Interim Staff Guidance on the Review of Nuclear Power Plant Designs Using... Review of Nuclear Power Plant Designs Using a Gas Turbine Driven Standby Emergency Alternating Current... for Nuclear Power Plants (LWR Edition),'' June 2007. Background: Emergency diesel generators...

  19. Inquiry into the radiological consequences of power uprates at light-water reactors worldwide

    Energy Technology Data Exchange (ETDEWEB)

    Bilic Zabric, Tea; Tomic, Bojan; Lundgren, Klas; Sjoeberg, Mats


    In Sweden, most of the nuclear power plants are planning power uprates within the next few years. The Dept. of Occupational and Medical Exposures at the Swedish Radiation Protection Agency, SSI, has initiated a research project to investigate the radiological implications of power uprates on light-water reactors throughout the world. The project was divided into three tasks: 1. A compilation of power uprates of light-water reactors worldwide. The compilation contains a technical description in brief of how the power uprates were carried out. 2. An analysis of the radiological consequences at four selected Nuclear Power Plants, which was the main objective of the inquiry. Affects on the radiological and chemical situation due to the changed situation were discussed. 3. Review of technical and organisational factors to be considered in uprate projects to keep exposures ALARA. The project was carried out, starting with the collecting of information on the implemented and planned uprates on reactors internationally. The information was catalogued in accordance with criteria focusing on radiological impact. A detailed analysis followed of four plants selected for uprates chosen according to established criteria, in line with the project requirements. The selected plants were Olkiluoto 1 and 2, Cofrentes, Asco and Tihange. The plants were selected with design and operation conditions close to the Swedish plants. All information was compiled to identify good and bad practices that are impacting on the occupational exposure. Important factors were discussed concerning BWRs and PWRs which affect radiation levels and occupational exposures in general, and especially at power uprates. Conclusions related to each task are in detail presented in a particular chapter of the report. Taking into account the whole project and its main objective the following conclusions are considered to be emphasized: Optimisation of the work processes to limit the duration of the time spent in

  20. Design considerations regarding slug ruptures in the intermediate power level reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pearl, W.L.; Pursel, C.A.


    The minimum shutdown time, to permit accessibility, for the Intermediate Power Reactor is estimated to be 38 hours. In case the reactor were shutdown following each rupture this long shutdown period would have serious disadvantages. The desirability of being able to make firm power commitments (independent of slug ruptures) has led to a study of the possibility of continuous operation following a rupture. There is evidence to indicate that, at the proposed water temperature, the rate of corrosion of uranium may be so high that at least a major portion of the rupture products may have entered the system before the reactor can be shutdown. A pushout of the affected column would then be a pushout of only those slugs which are still intact and the problem would still remain of removing the rupture products from the system. The first portion of this report is concerned with the rate of corrosion of a slug following rupture and the possible limitations to the principle of non-shutdown operation. These limitations include a flow stoppage by the ruptured can, undue increase in gamma activity, increased corrosion by the rupture products, and adherence of rupture products to the piping. The latter portion of the document is concerned with design considerations of the shielding and water plant so as to eliminate or minimize the effects of the introduction of rupture products into the cooling system. 7 refs., 2 figs.


    Energy Technology Data Exchange (ETDEWEB)

    REID, ROBERT S. [Los Alamos National Laboratory; PEARSON, J. BOSIE [Los Alamos National Laboratory; STEWART, ERIC T. [Los Alamos National Laboratory


    Water based reactor shielding is being investigated for use on initial lunar surface power systems. A water shield may lower overall cost (as compared to development cost for other materials) and simplify operations in the setup and handling. The thermal hydraulic performance of the shield is of significant interest. The mechanism for transferring heat through the shield is natural convection. Natural convection in a 100 kWt lunar surface reactor shield design is evaluated with 2 kW power input to the water in the Water Shield Testbed (WST) at the NASA Marshall Space Flight Center. The experimental data from the WST is used to validate a CFD model. Performance of the water shield on the lunar surface is then predicted with a CFD model anchored to test data. The experiment had a maximum water temperature of 75 C. The CFD model with 1/6-g predicts a maximum water temperature of 88 C with the same heat load and external boundary conditions. This difference in maximum temperature does not greatly affect the structural design of the shield, and demonstrates that it may be possible to use water for a lunar reactor shield.

  2. Minimization of the external heating power by long fusion power rise-up time for self-ignition access in the helical reactor FFHR2m (United States)

    Mitarai, O.; Sagara, A.; Chikaraishi, H.; Imagawa, S.; Watanabe, K.; Shishkin, A. A.; Motojima, O.


    Minimization of the external heating power to access self-ignition is advantageous to increase the reactor design flexibility and to reduce the capital and operating costs of the plasma heating device in a helical reactor. In this work we have discovered that a larger density limit leads to a smaller value of the required confinement enhancement factor, a lower density limit margin reduces the external heating power and over 300 s of the fusion power rise-up time makes it possible to reach a minimized heating power. While the fusion power rise-up time in a tokamak is limited by the OH transformer flux or the current drive capability, any fusion power rise-up time can be employed in a helical reactor for reducing the thermal stresses of the blanket and shields, because the confinement field is generated by the external helical coils.

  3. Development of technology-neutral safety requirements for the regulation of future nuclear power reactors: Back to basics

    Energy Technology Data Exchange (ETDEWEB)

    Tronea, Madalina, E-mail: [Faculty of Physics, University of Bucharest (Romania)


    This paper explores the current trends as regards the development of technology-neutral safety requirements to be used in the regulation of future nuclear power reactors and the role of the quantitative safety goals in the design of reactor safety systems. The use of the recommendations of the International Commission on Radiological Protection (ICRP) on protection against potential exposure could form the basis of a technology-neutral framework for safety requirements on new reactor designs and could contribute to international harmonisation of nuclear safety assessment practices as part of the licensing processes for future nuclear power plants.

  4. 239Pu Prompt Fission Neutron Spectra Impact on a Set of Criticality and Experimental Reactor Benchmarks (United States)

    Peneliau, Y.; Litaize, O.; Archier, P.; De Saint Jean, C.


    A large set of nuclear data are investigated to improve the calculation predictions of the new neutron transport simulation codes. With the next generation of nuclear power plants (GEN IV projects), one expects to reduce the calculated uncertainties which are mainly coming from nuclear data and are still very important, before taking into account integral information in the adjustment process. In France, future nuclear power plant concepts will probably use MOX fuel, either in Sodium Fast Reactors or in Gas Cooled Fast Reactors. Consequently, the knowledge of 239Pu cross sections and other nuclear data is crucial issue in order to reduce these sources of uncertainty. The Prompt Fission Neutron Spectra (PFNS) for 239Pu are part of these relevant data (an IAEA working group is even dedicated to PFNS) and the work presented here deals with this particular topic. The main international data files (i.e. JEFF-3.1.1, ENDF/B-VII.0, JENDL-4.0, BRC-2009) have been considered and compared with two different spectra, coming from the works of Maslov and Kornilov respectively. The spectra are first compared by calculating their mathematical moments in order to characterize them. Then, a reference calculation using the whole JEFF-3.1.1 evaluation file is performed and compared with another calculation performed with a new evaluation file, in which the data block containing the fission spectra (MF=5, MT=18) is replaced by the investigated spectra (one for each evaluation). A set of benchmarks is used to analyze the effects of PFNS, covering criticality cases and mock-up cases in various neutron flux spectra (thermal, intermediate, and fast flux spectra). Data coming from many ICSBEP experiments are used (PU-SOL-THERM, PU-MET-FAST, PU-MET-INTER and PU-MET-MIXED) and French mock-up experiments are also investigated (EOLE for thermal neutron flux spectrum and MASURCA for fast neutron flux spectrum). This study shows that many experiments and neutron parameters are very sensitive to

  5. Nuclear reactor power for a space-based radar. SP-100 project (United States)

    Bloomfield, Harvey; Heller, Jack; Jaffe, Leonard; Beatty, Richard; Bhandari, Pradeep; Chow, Edwin; Deininger, William; Ewell, Richard; Fujita, Toshio; Grossman, Merlin


    A space-based radar mission and spacecraft, using a 300 kWe nuclear reactor power system, has been examined, with emphasis on aspects affecting the power system. The radar antenna is a horizontal planar array, 32 X 64 m. The orbit is at 61 deg, 1088 km. The mass of the antenna with support structure is 42,000 kg; of the nuclear reactor power system, 8,300 kg; of the whole spacecraft about 51,000 kg, necessitating multiple launches and orbital assembly. The assembly orbit is at 57 deg, 400 km, high enough to provide the orbital lifetime needed for orbital assembly. The selected scenario uses six Shuttle launches to bring the spacecraft and a Centaur G upper-stage vehicle to assembly orbit. After assembly, the Centaur places the spacecraft in operational orbit, where it is deployed on radio command, the power system started, and the spacecraft becomes operational. Electric propulsion is an alternative and allows deployment in assembly orbit, but introduces a question of nuclear safety.

  6. Research on pressure control of pressurizer in pressurized water reactor nuclear power plant (United States)

    Dai, Ling; Yang, Xuhong; Liu, Gang; Ye, Jianhua; Qian, Hong; Xue, Yang


    Pressurizer is one of the most important components in the nuclear reactor system. Its function is to keep the pressure of the primary circuit. It can prevent shutdown of the system from the reactor accident under the normal transient state while keeping the setting value in the normal run-time. This paper is mainly research on the pressure system which is running in the Daya Bay Nuclear Power Plant. A conventional PID controller and a fuzzy controller are designed through analyzing the dynamic characteristics and calculating the transfer function. Then a fuzzy PID controller is designed by analyzing the results of two controllers. The fuzzy PID controller achieves the optimal control system finally.

  7. Design of a Low Power, Fast-Spectrum, Liquid-Metal Cooled Surface Reactor System (United States)

    Marcille, T. F.; Dixon, D. D.; Fischer, G. A.; Doherty, S. P.; Poston, D. I.; Kapernick, R. J.


    In the current 2005 US budget environment, competition for fiscal resources make funding for comprehensive space reactor development programs difficult to justify and accommodate. Simultaneously, the need to develop these systems to provide planetary and deep space-enabling power systems is increasing. Given that environment, designs intended to satisfy reasonable near-term surface missions, using affordable technology-ready materials and processes warrant serious consideration. An initial lunar application design incorporating a stainless structure, 880 K pumped NaK coolant system and a stainless/UO2 fuel system can be designed, fabricated and tested for a fraction of the cost of recent high-profile reactor programs (JIMO, SP-100). Along with the cost reductions associated with the use of qualified materials and processes, this design offers a low-risk, high-reliability implementation associated with mission specific low temperature, low burnup, five year operating lifetime requirements.

  8. CALMOS: Innovative device for the measurement of nuclear heating in material testing reactors

    Energy Technology Data Exchange (ETDEWEB)

    Carcreff, H. [Alternative Energies and Atomic Energy Commission CEA, Saclay Center, DEN/DANS/DRSN/SIREN, Gif Sur Yvette, 91191 (France)


    An R and D program has been carried out since 2002 in order to improve gamma heating measurements in the 70 MWth OSIRIS Material Testing Reactor operated by CEA's Nuclear Energy Div. at the Saclay research center. Throughout this program an innovative calorimetric probe associated to a specific handling system has been designed in order to make measurements both along the fissile height and on the upper part of the core, where nuclear heating rates still remain high. Two mock-ups of the probe were manufactured and tested in 2005 and 2009 in ex-core area of OSIRIS reactor for the process validation, while a displacement system has been especially designed to move the probe axially. A final probe has been designed thanks to modeling results and to preliminary measurements obtained with mock-ups irradiated to a heating level of 2W/g, This paper gives an overview of the development, describes the calorimetric probe, and expected advantages such as the possibility to use complementary methods to get the nuclear heating measurement. Results obtained with mock-ups irradiated in ex-core area of the reactor are presented and discussed. (authors)

  9. Zero Power Buoyancy Control for Distributed Autonomous Sensor Networks Conducted During Crimson Viper 2010 Field Experiment (United States)


    STATEMENT 13. SUPPLEMENTARY NOTES 14. ABSTRACT 15. SUBJECT TERMS 16. SECURITY CLASSIFICATION OF: a. REPORT 19a. NAME OF RESPONSIBLE PERSON 19b. TELEPHONE...producing simple, small, power-efficient data harvesting nodes with variable buoyancy, enabling unsupervised underwater sensing with subsequent surfacing...aerosol samples for subsequent analysis at NRL. At regular intervals, a hand-held hyperspectral sensor as to be used to collect near-water optical

  10. Optimization of power-cycle arrangements for Supercritical Water cooled Reactors (SCWRs) (United States)

    Lizon-A-Lugrin, Laure

    The world energy demand is continuously rising due to the increase of both the world population and the standard of life quality. Further, to assure both a healthy world economy as well as adequate social standards, in a relatively short term, new energy-conversion technologies are mandatory. Within this framework, a Generation IV International Forum (GIF) was established by the participation of 10 countries to collaborate for developing nuclear power reactors that will replace the present technology by 2030. The main goals of these nuclear-power reactors are: economic competitiveness, sustainability, safety, reliability and resistance to proliferation. As a member of the GIF, Canada has decided to orient its efforts towards the design of a CANDU-type Super Critical Water-cooled Reactor (SCWR). Such a system must run at a coolant outlet temperature of about 625°C and at a pressure of 25 MPa. It is obvious that at such conditions the overall efficiency of this kind of Nuclear Power Plant (NPP) will compete with actual supercritical water-power boilers. In addition, from a heat-transfer viewpoint, the use of a supercritical fluid allows the limitation imposed by Critical Heat Flux (CHF) conditions, which characterize actual technologies, to be removed. Furthermore, it will be also possible to use direct thermodynamic cycles where the supercritical fluid expands right away in a turbine without the necessity of using intermediate steam generators and/or separators. This work presents several thermodynamic cycles that could be appropriate to run SCWR power plants. Improving both thermal efficiency and mechanical power constitutes a multi-objective optimization problem and requires specific tools. To this aim, an efficient and robust evolutionary algorithm, based on genetic algorithm, is used and coupled to an appropriate power plant thermodynamic simulation model. The results provide numerous combinations to achieve a thermal efficiency higher than 50% with a

  11. Thermal noise informatics: Totally secure communication via a wire; Zero-power communication; and Thermal noise driven computing

    CERN Document Server

    Kish, Laszlo B; gingl, Zoltan


    Very recently, it has been shown that Gaussian thermal noise and its artificial versions (Johnson-like noises) can be utilized as an information carrier with peculiar properties therefore it may be proper to call this topic Thermal Noise Informatics. Zero Power (Stealth) Communication, Thermal Noise Driven Computing, and Totally Secure Classical Communication are relevant examples. In this paper, while we will briefly describe the first and the second subjects, we shall focus on the third subject, the secure classical communication via wire. This way of secure telecommunication utilizes the properties of Johnson(-like) noise and those of a simple Kirchhoff's loop. The communicator is unconditionally secure at the conceptual (circuit theoretical) level and this property is (so far) unique in communication systems based on classical physics. The communicator is superior to quantum alternatives in all known aspects, except the need of using a wire.

  12. Xenon-induced power oscillations in a generic small modular reactor (United States)

    Kitcher, Evans Damenortey

    As world demand for energy continues to grow at unprecedented rates, the world energy portfolio of the future will inevitably include a nuclear energy contribution. It has been suggested that the Small Modular Reactor (SMR) could play a significant role in the spread of civilian nuclear technology to nations previously without nuclear energy. As part of the design process, the SMR design must be assessed for the threat to operations posed by xenon-induced power oscillations. In this research, a generic SMR design was analyzed with respect to just such a threat. In order to do so, a multi-physics coupling routine was developed with MCNP/MCNPX as the neutronics solver. Thermal hydraulic assessments were performed using a single channel analysis tool developed in Python. Fuel and coolant temperature profiles were implemented in the form of temperature dependent fuel cross sections generated using the SIGACE code and reactor core coolant densities. The Power Axial Offset (PAO) and Xenon Axial Offset (XAO) parameters were chosen to quantify any oscillatory behavior observed. The methodology was benchmarked against results from literature of startup tests performed at a four-loop PWR in Korea. The developed benchmark model replicated the pertinent features of the reactor within ten percent of the literature values. The results of the benchmark demonstrated that the developed methodology captured the desired phenomena accurately. Subsequently, a high fidelity SMR core model was developed and assessed. Results of the analysis revealed an inherently stable SMR design at beginning of core life and end of core life under full-power and half-power conditions. The effect of axial discretization, stochastic noise and convergence of the Monte Carlo tallies in the calculations of the PAO and XAO parameters was investigated. All were found to be quite small and the inherently stable nature of the core design with respect to xenon-induced power oscillations was confirmed. Finally, a

  13. High temporal resolution Br2, BrCl and BrO observations in coastal Antarctica

    Directory of Open Access Journals (Sweden)

    R. von Glasow


    Full Text Available There are few observations of speciated inorganic bromine in polar regions against which to test current theory. Here we report the first high temporal resolution measurements of Br2, BrCl and BrO in coastal Antarctica, made at Halley during spring 2007 using a Chemical Ionisation Mass Spectrometer (CIMS. We find indications for an artefact in daytime BrCl measurements arising from conversion of HOBr, similar to that already identified for observations of Br2 made using a similar CIMS method. Using the MISTRA model, we estimate that the artefact represents a conversion of HOBr to Br2 of the order of several tens of percent, while that for HOBr to BrCl is less but non-negligible. If the artefact is indeed due to HOBr conversion, then nighttime observations were unaffected. It also appears that all daytime BrO observations were artefact-free. Mixing ratios of BrO, Br2 and BrCl ranged from instrumental detection limits to 13 pptv (daytime, 45 pptv (nighttime, and 6 pptv (nighttime, respectively. We see considerable variability in the Br2 and BrCl observations over the measurement period which is strongly linked to the prevailing meteorology, and thus air mass origin. Higher mixing ratios of these species were generally observed when air had passed over the sea-ice zone prior to arrival at Halley, than from over the continent. Variation in the diurnal structure of BrO is linked to previous model work where differences in the photolysis spectra of Br2 and O3 is suggested to lead to a BrO maximum at sunrise and sunset, rather than a noon-time maxima. This suite of Antarctic data provides the first analogue to similar measurements made in the Arctic, and of note is that our maximum measured BrCl (nighttime is less than half of the maximum measured during a similar period (spring-time in the Arctic (also nighttime. This difference in maximum measured BrCl may also be the cause of a difference in the Br2 : BrCl ratio between the Arctic and Antarctic. An

  14. Post 9-11 Security Issues for Non-Power Reactor Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Zaffuts, P. J.


    This paper addresses the legal and practical issues arising out of the design and implementation of a security-enhancement program for non power reactor nuclear facilities. The security enhancements discussed are derived from the commercial nuclear power industry's approach to security. The nuclear power industry's long and successful experience with protecting highly sensitive assets provides a wealth of information and lessons that should be examined by other industries contemplating security improvements, including, but not limited to facilities using or disposing of nuclear materials. This paper describes the nuclear industry's approach to security, the advantages and disadvantages of its constituent elements, and the legal issues that facilities will need to address when adopting some or all of these elements in the absence of statutory or regulatory requirements to do so.

  15. Excitation Method of Linear-Motor-Type Rail Brake without Using Power Sources by Dynamic Braking with Zero Electrical Output (United States)

    Sakamoto, Yasuaki; Kashiwagi, Takayuki; Hasegawa, Hitoshi; Sasakawa, Takashi; Fujii, Nobuo

    The eddy current rail brake is a type of braking system used in railway vehicles. Because of problems such as rail heating and problems associated with ensuring that power is supplied when the feeder malfunctions, this braking system has not been used for practical applications in Japan. Therefore, we proposed the use of linear induction motor (LIM) technology in eddy current rail brake systems. The LIM rail brake driven by dynamic braking can reduce rail heating and generate the energy required for self-excitation. In this paper, we present an excitation system and control method for the LIM rail brake driven by “dynamic braking with zero electrical output”. The proposed system is based on the concept that the LIM rail brake can be energized without using excitation power sources such as a feeder circuit and that high reliability can be realized by providing an independent excitation system. We have studied this system and conducted verification tests using a prototype LIM rail brake on a roller rig. The results show that the system performance is adequate for commercializing the proposed system, in which the LIM rail brake is driven without using any excitation power source.

  16. Structural materials challenges for advanced reactor systems (United States)

    Yvon, P.; Carré, F.


    mechanical properties and corrosion resistance, as well as component mock-up tests on technology loops to validate potential applications while accounting for mechanical design rules and manufacturing processes. The selection, assessment and validation of materials necessitate a large number of experiments, involving rare and expensive facilities such as research reactors, hot laboratories or corrosion loops. The modelling and the codification of the behaviour of materials will always involve the use of such technological experiments, but it is of utmost importance to develop also a predictive material science. Finally, the paper stresses the benefit of prospects of multilateral collaboration to join skills and share efforts of R&D to achieve in the nuclear field breakthroughs on materials that have already been achieved over the past decades in other industry sectors (aeronautics, metallurgy, chemistry, etc.).

  17. Concept of a nuclear powered submersible research vessel and a compact reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kusunoki, Tsuyoshi; Odano, Naoteru; Yoritsune, Tsutomu; Ishida, Toshihisa [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Takahashi, Teruo [Energis, Co., Kobe, Hyogo (Japan); Nishimura, Hajime [Japan Marine Science and Technology Center, Yokosuka, Kanagawa (Japan); Tokunaga, Sango [Japan Deep Sea Technology Association, Tokyo (Japan)


    A conceptual design study of a submersible research vessel navigating in 600 m depth and a compact nuclear reactor were carried out for the expansion of the nuclear power utilization. The mission of the vessel is the research of mechanism of the climate change to predict the global environment. Through conditions of the Arctic Ocean and the sea at high latitude have significant impacts on the global environmental change, it is difficult to investigate those areas by ordinary ships because of thick ice or storm. Therefore the research vessel is mainly utilized in the Arctic Ocean and the sea at high latitude. By taking account of the research mission, the basic specifications of the vessel are decided; the total weight is 500 t, the submersible depth is 600 m, the maximum speed is 12 knots (22.2 km/h), and the number of crews is 16. Nuclear power has an advantage in supplying large power of electricity in the sea for long period. Based on the requirements, it has been decided that two sets of submersible compact reactor, SCR, which is light-weighted and of enhanced safety characteristics of supply the total electricity of 500 kW. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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


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

  19. Collaborative investigations of in-service irradiated material from the Japan Power Demonstration Reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Corwin, W.R.; Broadhead, B.L. [Oak Ridge National Lab., TN (United States); Suzuki, M.; Kohsaka, A. [Japan Atomic Energy Research Institute, Tokai (Japan)


    There is a need to validate the results of irradiation effects research by the examination of material taken directly from the wall of a pressure vessel that has been irradiated during normal service. Just such an evaluation is currently being conducted on material from the wall of the pressure vessel from the Japan Power Demonstration Reactor (JPDR). The research is being jointly performed at the Tokai Research Establishment of the Japan Atomic Energy Research Institute (JAERI) and by the Nuclear Regulatory Commission (NRC)-funded Heavy-Section Steel Irradiation Program at the Oak Ridge National Laboratory (ORNL).

  20. Vibration behavior of fuel-element vibration suppressors for the advanced power reactor (United States)

    Adams, D. W.; Fiero, I. B.


    Preliminary shock and vibration tests were performed on vibration suppressors for the advanced power reactor for space application. These suppressors position the fuel pellets in a pin type fuel element. The test determined the effect of varying axial clearance on the behavior of the suppressors when subjected to shock and vibratory loading. The full-size suppressor was tested in a mockup model of fuel and clad which required scaling of test conditions. The test data were correlated with theoretical predictions for suppressor failure. Good agreement was obtained. The maximum difference with damping neglected was about 30 percent. Neglecting damping would result in a conservative design.

  1. Interactive Virtual Reactor and Control Room for Education and Training at Universities and Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Satoh, Yoshinori; Li, Ye; Zhu, Xuefeng; Rizwan, Uddin [University of Illinois, Urbana (United States)


    Efficient and effective education and training of nuclear engineering students and nuclear workers are critical for the safe operation and maintenance of nuclear power plants. With an eye toward this need, we have focused on the development of 3D models of virtual labs for education, training as well as to conduct virtual experiments. These virtual labs, that are expected to supplement currently available resources, and have the potential to reduce the cost of education and training, are most easily developed on game-engine platforms. We report some recent extensions to the virtual model of the University of Illinois TRIGA reactor.

  2. The mode of operation of CANDU power reactor in thorium self-sufficient fuel cycle

    Directory of Open Access Journals (Sweden)

    Bergelson Boris R.


    Full Text Available This paper presents the results of calculations for CANDU reactor operation in the thorium fuel cycle. The calculations were performed to estimate feasibility of operation of a heavy-water thermal neutron power reactor in the self-sufficient thorium cycle. The parameters of the active core and the scheme of fuel reloading were considered to be the same as for the standard operation in the uranium cycle. Two modes of operation are discussed in the paper: the mode of preliminary accumulation of 233U and the mode of operation in the self-sufficient cycle. For calculations for the mode of accumulation of 233U, it was assumed that plutonium was used as the additional fissile material to provide neutrons for 233U production. Plutonium was placed in fuel channels, while 232Th was located in target channels. The maximum content of 233U in the target channels was about 13 kg/t of ThO2. This was achieved by six year irradiation. The start of reactor operation in the self-sufficient mode requires content of 233U not less than 12 kg/t. For the mode of operation in the self-sufficient cycle, it was assumed that all the channels were loaded with the identical fuel assemblies containing ThO2 and a certain amount of 233U. It was shown that the non-uniform distribution of 233U in a fuel assembly is preferable.

  3. Knowledge and abilities catalog for nuclear power plant operators: Boiling water reactors, Revision 1

    Energy Technology Data Exchange (ETDEWEB)



    The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling-Water Reactors (BWRs) (NUREG-1123, Revision 1) provides the basis for the development of content-valid licensing examinations for reactor operators (ROs) and senior reactor operators (SROs). The examinations developed using the BWR Catalog along with the Operator Licensing Examiner Standards (NUREG-1021) and the Examiner`s Handbook for Developing Operator Licensing Written Examinations (NUREG/BR-0122), will cover the topics listed under Title 10, Code of Federal Regulations, Part 55 (10 CFR 55). The BWR Catalog contains approximately 7,000 knowledge and ability (K/A) statements for ROs and SROs at BWRs. The catalog is organized into six major sections: Organization of the Catalog, Generic Knowledge and Ability Statements, Plant Systems grouped by Safety Functions, Emergency and Abnormal Plant Evolutions, Components, and Theory. Revision 1 to the BWR Catalog represents a modification in form and content of the original catalog. The K/As were linked to their applicable 10 CFR 55 item numbers. SRO level K/As were identified by 10 CFR 55.43 item numbers. The plant-wide generic and system generic K/As were combined in one section with approximately one hundred new K/As. Component Cooling Water and Instrument Air Systems were added to the Systems Section. Finally, High Containment Hydrogen Concentration and Plant Fire On Site evolutions added to the Emergency and Abnormal Plant Evolutions section.

  4. Operation of CANDU power reactor in thorium self-sufficient fuel cycle

    Indian Academy of Sciences (India)

    B R Bergelson; A S Gerasimov; G V Tikhomirov


    This paper presents the results of calculations for CANDU reactor operation in thorium fuel cycle. Calculations are performed to estimate the feasibility of operation of heavy-water thermal neutron power reactor in self-sufficient thorium cycle. Parameters of active core and scheme of fuel reloading were considered to be the same as for standard operation in uranium cycle. Two modes of operations are discussed in the paper: mode of preliminary accumulation of 233U and mode of operation in self-sufficient cycle. For the mode of accumulation of 233U it was assumed for calculations that plutonium can be used as additional fissile material to provide neutrons for 233U production. Plutonium was placed in fuel channels, while 232Th was located in target channels. Maximum content of 233U in target channels was estimated to be ∼ 13 kg/t of ThO2. This was achieved by irradiation for six years. The start of the reactor operation in the self-sufficient mode requires 233U content to be not less than 12 kg/t. For the mode of operation in self-sufficient cycle, it was assumed that all channels were loaded with identical fuel assemblies containing ThO2 and certain amount of 233U. It is shown that nonuniform distribution of 233U in fuel assembly is preferable.

  5. R&D on high-power dc reactor prototype for ITER poloidal field converter

    Energy Technology Data Exchange (ETDEWEB)

    Li, Chuan [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); School of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Song, Zhiquan; Fu, Peng [Institute of Plasma Physics, Chinese Academy of Science, Hefei 230031 (China); Zhang, Ming, E-mail: [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); School of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Yu, Kexun [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); School of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Qin, Xiuqi [School of Electrical Engineering and Automation, Hefei University of Technology, Hefei 230009 (China)


    Highlights: • A new prototype design structure of dry-type air-core water-cooling reactor with epoxy resin casting technique is presented. • Theoretical analysis, finite-element simulation and prototype test verification are applied on the design. • The results of temperature rise and transient fault current test of prototypes are introduced and analyzed. • The success of tests demonstrates that the proposed structure is of high reliability and availability. - Abstract: This paper mainly introduces the research and development (R&D) of the high-power dc reactor prototype, whose functions are to limit the circulating current and ripple current in the ITER poloidal field (PF) converter. It needs to operate at rated large direct current 27.5 kA and withstand peak fault current up to 175 kA. Therefore, in order to meet the special requirements of the dynamic and thermal stability, a new prototype design structure of dry-type air-core water-cooling reactor with epoxy resin casting technique is presented, which is based on the theoretical analysis, finite-element simulation calculation and small prototype test verification. Now the full prototype has been fabricated by China industry, and the dynamic and thermal stability tests of the prototype have also been accomplished successfully. The test results are in compliance with the design and it shows the availability and feasibility of the proposed design, which may be a reference for relevant applications.

  6. PR-EDB: Power Reactor Embrittlement Data Base, Version 2. Revision 2, Program description

    Energy Technology Data Exchange (ETDEWEB)

    Stallmann, F.W.; Wang, J.A.; Kam, F.B.K.; Taylor, B.J. [Oak Ridge National Lab., TN (United States)


    Investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current codes Standard Review Plans (SRP`s) and Guides for license renewal can be greatly expedited by the use of a well-designed computerized data base. Also, such a data base is essential for the validation of embrittlement prediction models by researchers. The Power Reactor Embrittlement Data Base (PR-EDB) is such a comprehensive collection of data for US commercial nuclear reactors. The current version of the PR-EDB contains the Charpy test data that were irradiated in 252 capsules of 96 reactors and consists of 207 data points for heat-affected-zone (HAZ) materials (98 different HAZ), 227 data points for weld materials (105 different welds), 524 data points for base materials (136 different base materials), including 297 plate data points (85 different plates), 119 forging data points (31) different forging), and 108 correlation monitor materials data points (3 different plates). The data files are given in dBASE format and can be accessed with any computer using the DOS operating system. ``User-friendly`` utility programs are used to retrieve and select specific data, manipulate data, display data to the screen or printer, and to fit and plot Charpy impact data. The results of several studies investigated are presented in Appendix D.

  7. Technology, safety, and costs of decommissioning a reference pressurized water reactor power station

    Energy Technology Data Exchange (ETDEWEB)

    Smith, R.I.; Konzek, G.J.; Kennedy, W.E. Jr.


    Safety and cost information was developed for the conceptual decommissioning of a large (1175 MW(e)) pressurized water reactor (PWR) power station. Two approaches to decommissioning, Immediate Dismantlement and Safe Storage with Deferred Dismantlement, were studied to obtain comparisons between costs, occupational radiation doses, potential radiation dose to the public, and other safety impacts. Immediate Dismantlement was estimated to require about six years to complete, including two years of planning and preparation prior to final reactor shutdown, at a cost of $42 million, and accumulated occupational radiation dose, excluding transport operations, of about 1200 man-rem. Preparations for Safe Storage were estimated to require about three years to complete, including 1/sup 1///sub 2/ years for planning and preparation prior to final reactor shutdown, at a cost of $13 million and an accumulated occupational radiation dose of about 420 man-rem. The cost of continuing care during the Safe Storage period was estimated to be about $80 thousand annually. Accumulated occupational radiation dose during the Safe Storage period was estimated to range from about 10 man-rem for the first 10 years to about 14 man-rem after 30 years or more. The cost of decommissioning by Safe Storage with Deferred Dismantlement was estimated to be slightly higher than Immediate Dismantlement. Cost reductions resulting from reduced volumes of radioactive material for disposal, due to the decay of the radioactive containments during the deferment period, are offset by the accumulated costs of surveillance and maintenance during the Safe Storage period.

  8. Russian design studies of the DEMO-S demonstration fusion power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kolbasov, B.; Belyakov, V.; Borisov, A.; Kirillov, I.; Shatalov, G.; Sokolov, Yu.; Strebkov, Yu.; Vasiliev, N. [Kurchatov Institut (Russian Federation)


    Different concepts for a fusion power plant have been studied in Russia since 1975. Researchers have considered power facilities using tokamaks, stellarators and inertial fusion devices. Tokamak reactors appear the most promising at this stage of science development. Application of fusion reactors for generation of electricity, production of domestic and industrial heat, hydrogen production, transmutation of non-fissionable isotopes into fissionable ones, water desalination, and burning out of minor actinides was considered. Conceptual design studies of a tokamak-based demonstration fusion reactor have been carried out since 1991. The preferred concept was selected, which was a steady-state operating tokamak with superconducting magnets, one-null divertor configuration and a high contribution of bootstrap current into plasma current drive. The general reactor layout was determined. Plasma characteristics were optimized. Two most attractive blanket concepts were analyzed: (1) a He-cooled ceramic (Li{sub 4}SiO{sub 4}) design for tritium breeding, using ferritic steel as structural material, and (2) a blanket using liquid Li as tritium breeding material and coolant and a V-Cr-Ti alloy as structural material. The studies were supported by neutronic, heat-hydraulic and mechanical calculations. A conventional type of water or Li cooled divertor targets with maximum heat load of {proportional_to}10 MW/m{sup 2} was chosen. Blankets of both types require Be as a neutron multiplier and have to be replaced after the integral fusion neutron load on the first wall reaches 10 MW/m{sup 2}. Heat to electricity conversion schemes enable operation with net efficiency of 34% for the He-coolant design and 40% for the liquid Li one. Aspects of radioactive waste management and scarce materials refabrication are considered. In particular, a radiochemical extraction technology for separation of V alloy components and their purification from activation products after reactor

  9. Effect of non-graphitized carbon electrodes on the electrochemical characteristics of a thermocell with a Br 2/Br - redox couple (United States)

    Shindo, Kazuhiko; Arakawa, Masayasu; Hirai, Toshiro

    Electrodes made from a non-graphitized carbon (Maxsorb) sheet with a large specific surface area greatly improved the electrochemical characteristics of thermocells with a bromine/bromide-ion redox couple and a KBr aqueous electrolyte saturated with Br 2. The Maxsorb electrodes provide a higher open-circuit voltage than graphite (Perma Foil) electrodes and the maximum discharge current density is 2.4 mA/cm 2 until the voltage reaches 0 V. The Maxsorb electrodes generate 5.68 mV/°C of thermoelectric power at a temperature difference greater than 40°C, which is 2.5 times the power reported previously. The Maxsorb electrodes also improve the cathodic polarization in the presence of a gas-liquid-solid interface in Br 2-free electrolyte, while the graphite electrodes do not. Nevertheless, this improvement is insufficient to observe when the KBr electrolyte is saturated with Br 2. An experimental cell with Maxsorb electrodes maintains a stable voltage up to 2.0 mA/cm 2 for over 450 h, even though the cell configuration has still to be optimized.

  10. The scalability of OTR (out-of-core thermionic reactor) space nuclear power systems

    Energy Technology Data Exchange (ETDEWEB)

    Gallup, D.R.


    In this document, masses of the STAR-C power system and an optimized out-of-core thermionic reactor (OTR) power system versus power level are investigated. The impacts of key system parameters on system performance are also addressed. The STAR-C is mass competitive below about 15 kWe, but at higher power levels the scalability is relatively poor. An optimized OR is the least massive space nuclear power system below 25 kWe, and scales well to 50 kWe. The system parameters that have a significant impact on the scalability of the STAR-C are core thermal flux, thermionic converter efficiency, and core length to diameter ratio. The emissivity of the core surface is shown to be a relatively unimportant parameter. For an optimized OR power system, the most significant system parameter is the maximum allowable fuel temperature. It is also shown that if advanced radiation-hardened electronics are used in the satellite payload, a very large mass savings is realized. 10 refs., 23 figs., 7 tabs.

  11. EOIL power scaling in a 1-5 kW supersonic discharge-flow reactor (United States)

    Davis, Steven J.; Lee, Seonkyung; Oakes, David B.; Haney, Julie; Magill, John C.; Paulsen, Dwane A.; Cataldi, Paul; Galbally-Kinney, Kristin L.; Vu, Danthu; Polex, Jan; Kessler, William J.; Rawlins, Wilson T.


    Scaling of EOIL systems to higher powers requires extension of electric discharge powers into the kW range and beyond with high efficiency and singlet oxygen yield. We have previously demonstrated a high-power microwave discharge approach capable of generating singlet oxygen yields of ~25% at ~50 torr pressure and 1 kW power. This paper describes the implementation of this method in a supersonic flow reactor designed for systematic investigations of the scaling of gain and lasing with power and flow conditions. The 2450 MHz microwave discharge, 1 to 5 kW, is confined near the flow axis by a swirl flow. The discharge effluent, containing active species including O II(a1Δ g, b1Σ g +), O( 3P), and O 3, passes through a 2-D flow duct equipped with a supersonic nozzle and cavity. I2 is injected upstream of the supersonic nozzle. The apparatus is water-cooled, and is modular to permit a variety of inlet, nozzle, and optical configurations. A comprehensive suite of optical emission and absorption diagnostics is used to monitor the absolute concentrations of O II(a), O II(b), O( 3P), O 3, I II, I(2P 3/2), I(2P 1/2), small-signal gain, and temperature in both the subsonic and supersonic flow streams. We discuss initial measurements of singlet oxygen and I* excitation kinetics at 1 kW power.

  12. Dynamic parameters test of Haiyang Nuclear Power Engineering in reactor areas, China (United States)

    Zhou, N.; Zhao, S.; Sun, L.


    Haiyang Nuclear Power Project is located in Haiyang city, China. It consists of 6×1000MW AP1000 Nuclear Power generator sets. The dynamic parameters of the rockmass are essential for the design of the nuclear power plant. No.1 and No.2 reactor area are taken as research target in this paper. Sonic logging, single hole and cross-hole wave velocity are carried out respectively on the site. There are four types of rock lithology within the measured depth. They are siltstone, fine sandstone, shale and allgovite. The total depth of sonic logging is 409.8m and 2049 test points. The sound wave velocity of the rocks are respectively 5521 m/s, 5576m/s, 5318 m/s and 5576 m/s. Accroding to the statistic data, among medium weathered fine sandstone, fairly broken is majority, broken and relatively integrity are second, part of integrity. Medium weathered siltstone, relatively integrity is mojority, fairly broken is second. Medium weathered shale, fairly broken is majority, broken and relatively integrity for the next and part of integrity. Slight weathered fine sandstone, siltstone, shale and allgovite, integrity is the mojority, relatively integrity for the next, part of fairly broken.The single hole wave velocity tests are set in two boreholesin No.1 reactor area and No.2 reactor area respectively. The test depths of two holes are 2-24m, and the others are 2-40m. The wave velocity data are calculated at different depth in each holes and dynamic parameters. According to the test statistic data, the wave velocity and the dynamic parameter values of rockmass are distinctly influenced by the weathering degree. The test results are list in table 1. 3 groups of cross hole wave velocity tests are set for No.1 and 2 reactor area, No.1 reactor area: B16, B16-1, B20(Direction:175°, depth: 100m); B10, B10-1, B11(269°, 40m); B21, B21-1, B17(154°, 40m); with HB16, HB10, HB21 as trigger holes; No.2 reactor area: B47, B47-1, HB51(176°, 100m); B40, B40-1, B41(272°, 40m); B42, B42-1, B

  13. Experimental Setup for the Measurement of the Thermoelectric Power in Zero and Applied Magnetic Field

    CERN Document Server

    Mun, Eundeok; Torikachvili, Milton S; Canfield, Paul C


    An experimental setup was developed for the measurement of the thermoelectric power (TEP, Seebeck coefficient) in the temperature range from 2 to 350 K and magnetic fields up to 140 kOe. The system was built to fit in a commercial cryostat and is versatile, accurate and automated; using two heaters and two thermometers increases the accuracy of the TEP measurement. High density data of temperature sweeps from 2 to 350 K can be acquired in under 16 hours and high density data of isothermal field sweeps from 0 to 140 kOe can be obtained in under 2 hours. Calibrations for the system have been performed on a platinum wire and Bi$_{2}$Sr$_{2}$CaCu$_{2}$O$_{8+\\delta}$ high $T_{c}$ superconductors. The measured TEP of phosphor-bronze (voltage lead wire) turns to be very small, where the absolute TEP value of phosphor-bronze wire is much less than 0.5 $\\mu$V/K below 80 K. For copper and platinum wires measured against to the phosphor-bronze wire, the agreement between measured results and the literature data is good....

  14. The Zero Emission Fossil Fuel Power Plant - from vision to reality.

    Energy Technology Data Exchange (ETDEWEB)

    Stroemberg, L.; Sauthoff, M.


    Sufficient supply of energy without fossil fuels is not possible the next fifty years. Thus, we must find a solution to use coal, without endangering the environment. Carbon Capture and Storage, CCS, might be the answer. At a cost of about 20 Euro/ton CO{sub 2}, there exist technologies, which can be ready for commercial application in 2020. After that, even more cost effective technologies will be developed. To reduce emissions by more than half until 2050, cannot be reached without CCS. However, CCS is very powerful, but not the only tool. All ways to reduce emissions, including renewables and nuclear must be used. To put emphasis behind the words, Vattenfall has started an R and D program to develop technology for CCS in a ten year program. As part of that, Vattenfall is building a Pilot Plant including all process steps from coal input to liquid CO{sub 2}. It will be ready in 2008. In parallel, preparations for a demonstration plant are ongoing. It will be a coal fired full size plant with storage on shore. That will be ready for operation in 2015. (auth)

  15. Radiotoxicity and decay heat power of spent nuclear fuel of VVER type reactors at long-term storage. (United States)

    Bergelson, B R; Gerasimov, A S; Tikhomirov, G V


    Radiotoxicity and decay heat power of the spent nuclear fuel of VVER-1000 type reactors are calculated during storage time up to 300,000 y. Decay heat power of radioactive waste (radwaste) determines parameters of the heat removal system for the safe storage of spent nuclear fuel. Radiotoxicity determines the radiological hazard of radwaste after its leakage and penetration into the environment.

  16. Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production, 3rd Quarterly Report

    Energy Technology Data Exchange (ETDEWEB)

    Mac Donald, Philip Elsworth


    The use of light water at supercritical pressures as the coolant in a nuclear reactor offers the potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies of the power conversion cycle are possible (i.e. up to about 45%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type re-circulation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel and smaller containment building than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed.

  17. A design study of reactor core optimization for direct nuclear heat-to-electricity conversion in a space power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yoshikawa, Hidekazu; Takahashi, Makoto; Shimoda, Hiroshi; Takeoka, Satoshi [Kyoto Univ. (Japan); Nakagawa, Masayuki; Kugo, Teruhiko


    To propose a new design concept of a nuclear reactor used in the space, research has been conducted on the conceptual design of a new nuclear reactor on the basis of the following three main concepts: (1) Thermionic generation by thermionic fuel elements (TFE), (2) reactivity control by rotary reflector, and (3) reactor cooling by liquid metal. The outcomes of the research are: (1) A calculation algorithm was derived for obtaining convergent conditions by repeating nuclear characteristic calculation and thermal flow characteristic calculation for the space nuclear reactor. (2) Use of this algorithm and the parametric study established that a space nuclear reactor using 97% enriched uranium nitride as the fuel and lithium as the coolant and having a core with a radius of about 25 cm, a height of about 50 cm and a generation efficiency of about 7% can probably be operated continuously for at least more than ten years at 100 kW only by reactivity control by rotary reflector. (3) A new CAD/CAE system was developed to assist design work to optimize the core characteristics of the space nuclear reactor comprehensively. It is composed of the integrated design support system VINDS using virtual reality and the distributed system WINDS to collaboratively support design work using Internet. (N.H.)

  18. A Burst Mode, Ultrahigh Temperature UF4 Vapor Core Reactor Rankine Cycle Space Power System Concept (United States)

    Dugan, E. T.; Kahook, S. D.; Diaz, N. J.


    Static and dynamic neutronic analyses have been performed on an innovative burst mode (100's of MW output for a few thousand seconds) Ulvahigh Temperature Vapor Core Reactor (UTVR) space nuclear power system. The NVTR employs multiple, neutronically-coupled fissioning cores and operates on a direct, closed Rankine cycle using a disk Magnetohydrodynamic (MHD) generater for energy conversion. The UTVR includes two types of fissioning core regions: (1) the central Ultrahigh Temperature Vapor Core (UTVC) which contains a vapor mixture of highly enriched UF4 fuel and a metal fluoride working fluid and (2) the UF4 boiler column cores located in the BeO moderator/reflector region. The gaseous nature of the fuel the fact that the fuel is circulating, the multiple coupled fissioning cores, and the use of a two phase fissioning fuel lead to unique static and dynamic neutronic characteristics. Static neutronic analysis was conducted using two-dimensional S sub n, transport theory calculations and three-dimensional Monte Carlo transport theory calculations. Circulating-fuel, coupled-core point reactor kinetics equations were used for analyzing the dynamic behavior of the UTVR. In addition to including reactivity feedback phenomena associated with the individual fissioning cores, the effects of core-to-core neutronic and mass flow coupling between the UTVC and the surrounding boiler cores were also included in the dynamic model The dynamic analysis of the UTVR reveals the existence of some very effectlve inherent reactivity feedback effects that are capable of quickly stabilizing this system, within a few seconds, even when large positive reactivity insertions are imposed. If the UTVC vapor fuel density feedback is suppressed, the UTVR is still inherently stable because of the boiler core liquid-fuel volume feedback; in contrast, suppression of the vapor fuel density feedback in 'conventional" gas core cavity reactors causes them to become inherently unstable. Due to the

  19. Zero-Power-Consumption Solar-Blind Photodetector Based on β-Ga2O3/NSTO Heterojunction. (United States)

    Guo, Daoyou; Liu, Han; Li, Peigang; Wu, Zhenping; Wang, Shunli; Cui, Can; Li, Chaorong; Tang, Weihua


    A solar-blind photodetector based on β-Ga2O3/NSTO (NSTO = Nb:SrTiO3) heterojunctions were fabricated for the first time, and its photoelectric properties were investigated. The device presents a typical positive rectification in the dark, while under 254 nm UV light illumination, it shows a negative rectification, which might be caused by the generation of photoinduced electron-hole pairs in the β-Ga2O3 film layer. With zero bias, that is, zero power consumption, the photodetector shows a fast photoresponse time (decay time τd = 0.07 s) and the ratio Iphoto/Idark ≈ 20 under 254 nm light illumination with a light intensity of 45 μW/cm(2). Such behaviors are attributed to the separation of photogenerated electron-hole pairs driven by the built-in electric field in the depletion region of β-Ga2O3 and the NSTO interface, and the subsequent transport toward corresponding electrodes. The photocurrent increases linearly with increasing the light intensity and applied bias, while the response time decreases with the increase of the light intensity. Under -10 V bias and 45 μW/cm(2) of 254 nm light illumination, the photodetector exhibits a responsivity Rλ of 43.31 A/W and an external quantum efficiency of 2.1 × 10(4) %. The photo-to-electric conversion mechanism in the β-Ga2O3/NSTO heterojunction photodetector is explained in detail by energy band diagrams. The results strongly suggest that a photodetector based on β-Ga2O3 thin-film heterojunction structure can be practically used to detect weak solar-blind signals because of its high photoconductive gain.

  20. Development of the breeding blanket and shield model for the fusion power reactors system SYCOMORE

    Energy Technology Data Exchange (ETDEWEB)

    Li-Puma, Antonella, E-mail: [CEA, DEN, Saclay, DM2S, SERMA, F-91191 Gif-sur-Yvette (France); Jaboulay, Jean-Charles, E-mail: [CEA, DEN, Saclay, DM2S, SERMA, F-91191 Gif-sur-Yvette (France); Martin, Brunella, E-mail: [Incka, 19-21 Rue du 8 mai 1945, F-94110 Arcueil (France)


    SYCOMORE, a fusion reactor system code based on a modular approach is under development at CEA. Within this framework, this paper describes the relevant sub-modules which have been implemented to model the main outputs of the breeding blanket and shield block of the system code: tritium breeding ratio, peak energy deposition in toroidal field coils, reactor layout and power deposition, blanket pressure drops and materials inventory. Blanket and shield requirements are calculated by several sub-modules: the blanket assembly and layout sub-module, the neutronic sub-module, the blanket design sub-module (thermal hydraulic and thermo-mechanic pre-design tool). A power flow module has also been developed which is directly linked to the blanket thermo-dynamic performances, which is not described in this paper. For the blanket assembly and layout and the blanket module design sub-modules, explicit analytic models have been developed and implemented; for the neutronic sub-module neural networks that replicate the results of appropriate simplified 1D and 2D neutronic simulations have been built. Presently, relevant model for the Helium Cooled Lithium Lead is available. Sub-modules have been built in a way that they can run separately or coupled into the breeding blanket and shield module in order to be integrated in SYCOMORE. In the paper, the objective and main input/output parameters of each sub-module are reported and relevant models discussed. The application to previous studied reactor models (PPCS model AB, DEMO-HCLL 2006–2007 studies) is also presented.

  1. Office of Analysis and Evaluation of Operational Data 1989 annual report, Power reactors

    Energy Technology Data Exchange (ETDEWEB)



    The annual report of the US Nuclear Regulatory Commission's Office for Analysis and Evaluation of Operational Data (AEOD) is devoted to the activities performed during 1989. The report is published in two separate parts. This document, NUREG-1272, Vol. 4, No. 1, covers power reactors and presents an overview of the operating experience of the nuclear power industry from the NRC perspective, including comments about the trends of some key performance measures. The report also includes the principal findings and issues identified in AEOD studies over the past year and summarizes information from such sources as licensee event reports, diagnostic evaluations, and reports to the NRC's Operations Center. This report also compiles the status of staff actions resulting from previous Incident Investigation Team (IIT) reports. 16 figs., 9 tabs.

  2. Proving test on the seismic reliability of nuclear power plant: PWR reactor containment vessel

    Energy Technology Data Exchange (ETDEWEB)

    Akiyama, Hiroshi; Yoshikawa, Teiichi; Ohno, Tokue; Yoshikawa, Eiji.


    Seismic reliability proving tests of nuclear power plant facilities are carried out by the Nuclear Power Engineering Test Center, using the large-scale, high-performance vibration table of Tadotsu Engineering Laboratory, and sponsored by the Ministry of International Trade and Industry. In 1982, the seismic reliability proving test of a PWR containment vessel was conducted using a test component of reduced scale 1/3.7. As a result of this test, the test component proved to have structural soundness against earthquakes, and at the same time its stable function was proved by leak tests which were carried out before and after the vibration test. In 1983, the detailed analysis and evaluation of these test results were carried out, and the analysis methods for evaluating strength against earthquakes were established. The seismic analysis and evaluation on the actual containment vessel were then performed using these analysis methods, and the safety and reliability of the PWR reactor containment vessel were confirmed.

  3. Experimental Evaluation of a Water Shield for a Surface Power Reactor (United States)

    Pearson, J. B.; Reid, R.; Sadasivan, P.; Stewart, E.


    A water based shielding system is being investigated for use on initial lunar surface power systems. The use of water may lower overall cost (as compared to development cost for other materials) and simplify operations in the setup and handling. The thermal hydraulic performance of the shield is of significant interest. The mechanism for transferring heat through the shield is natural convection. A representative lunar surface reactor design is evaluated at various power levels in the Water Shield Testbed (WST) at the NASA Marshall Space Flight Center. The evaluation compares the experimental data from the WST to CFD models. Performance of a water shield on the lunar surface is predicted by CFD models anchored to test data, and by matching relevant dimensionless parameters.

  4. Experimental determination of creep properties of Beryllium irradiated to relevant fusion power reactor doses (United States)

    Scibetta, M.; Pellettieri, A.; Sannen, L.


    A dead weight machine has been developed to measure creep in irradiated beryllium relevant to fusion power reactors. Due to the external compressive load, the material will creep and the specimen will shrink. However, the specimen also swells due to the combined effect of internal pressure in helium bubbles and creep. One of the major challenges is to unmask swelling and derive intrinsic creep properties. This has been achieved through appropriate pre-annealing experiments. Creep has been measured on irradiated and unirradiated specimens. The temperature and stress dependence is characterized and modeled using the product of an Arrhenius' law for the temperature dependence and a power law for the stress dependence. Irradiation increases the sensitivity to creep but the irradiation effects can be rationalized by taking into account the irradiation-induced porosity. Experimental evidence supports dislocation climb by vacancy absorption to be the most plausible intrinsic creep mechanism.

  5. High-power-density approaches to magnetic fusion energy: Problems and promise of compact reversed-field pinch reactors (CRFPR) (United States)

    Hagenson, Randy L.; Krakowski, Robert A.; Dreicer, Harry


    If the costing assumptions upon which the positive assessment of conventional large superconducting fusion reactors are based proves overly optimistic, approaches that promise considerably increased system power density and reduced mass utilization will be required. These more compact reactor embodiments generally must operate with reduced shield thickness and resistive magnets. Because of the unique magnetic topology associated with the Reversed-Field Pinch (RFP), the compact reactor embodiment for this approach is particularly attractive from the viewpoint of low-field resistive coils operating with ohmic losses that can be made small relative to the fusion power. The RFP, therefore, is used as one example of a high-power-density (HPD) approach to magnetic fusion energy. A comprehensive system model is described and applied to select a unique, cost-optimized design point that will be used for a subsequent conceptual engineering design of the compact RFP Reactor (CRFPR). This cost-optimized CRFPR design serves as an example of a HPD fusion reactor that would operate with system power densities and mass utilizations that are comparable to fission power plants, these measures of system performance being an order of magnitude more favorable than the conventional approaches to magnetic fusion energy (MFE).

  6. A study on neural network representation of reactor power control procedures 2

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Byung Soo; Park, Jea Chang; Kim, Young Taek; Lee, Hee Cho; Yang, Sung Uoon; Hwang, Hee Sun; Hwang, In Ah


    The major results of this study are as follows; the first is the algorithm developed through this study for computing the spline interpolation coefficients without solving the matrix equation involved. This is expected to be used in various numerical analysis problems. If this algorithm can be extended to functions of two independent variables in the future, then it could be a big help for the finite element method used in solving various boundary value problems. The second is the method developed to reduce systematically the number of output fuzzy sets for fuzzy systems representing functions of two variables. this may be considered as an indication that the neural network representation of functions has advantages over other conventional methods. The third result is an artificial neural network system developed for automating the manual procedures being used to change the reactor power level by adding boric acid or water to the reactor coolant. This along with the neural networks developed earlier can be used in nuclear power plants as an operator aid after a verification process. (author). 8 refs., 13 tabs., 5 figs.

  7. 14C release from a Soviet-designed pressurized water reactor nuclear power plant. (United States)

    Uchrin, G; Csaba, E; Hertelendi, E; Ormai, P; Barnabas, I


    The Paks Nuclear Power Plant in Hungary runs with four pressurized water reactors, each of 440-MWe capacity. Sampling systems have been developed and used to determine the 14C of various chemical forms (14CO2, 14CO, 14CnHm) in the airborne releases. The average normalized yearly discharge rates for the time period 1988-1991 are equal to 0.77 TBq GWe-1 y-1 for hydrocarbons and 0.05 TBq GWe-1 y-1 for CO2. The contribution of 14CO was less than 0.5% of the total emission. The 14C discharge rate is estimated to be four times higher than the corresponding mean data of Western European pressurized water reactors. The calculated effective dose equivalent to individuals living in the vicinity of the power plant, due to 14C release, was 0.64 microSv in 1991 while the effective dose equivalent due to the natural 14C level was 15 microSv y-1. The long-term global impact of the 14C release in the operational period of the plant (1982-1991) was 1,270 man-Sv. The 14C excess in the environmental air has been measured since 1989 by taking biweekly samples at a distance of 1.7 km from the nuclear power plant. The long-term average of radiocarbon excess coming from the power plant was 2 mBq m-3. The local 14C deposition was followed by tree ring analysis, too. No 14C increase higher than the uncertainty of the measurement (four per thousand = 0.17 mBq m-3) was observed.

  8. Development and evaluation of a novel low power, high frequency piezoelectric-based ultrasonic reactor for intensifying the transesterification reaction

    Directory of Open Access Journals (Sweden)

    Mortaza Aghbashlo


    Full Text Available In this study, a novel low power, high frequency piezoelectric-based ultrasonic reactor was developed and evaluated for intensifying the transesterification process. The reactor was equipped with an automatic temperature control system, a heating element, a precise temperature sensor, and a piezoelectric-based ultrasonic module. The conversion efficiency and specific energy consumption of the reactor were examined under different operational conditions, i.e., reactor temperature (40‒60 °C, ultrasonication time (6‒10 min, and alcohol/oil molar ratio (4:1‒8:1. Transesterification of waste cooking oil (WCO was performed in the presence of a base-catalyst (potassium hydroxide using methanol. According to the obtained results, alcohol/oil molar ratio of 6:1, ultrasonication time of 10 min, and reactor temperature of 60 °C were found as the best operational conditions. Under these conditions, the reactor converted WCO to biodiesel with a conversion efficiency of 97.12%, meeting the ASTM standard satisfactorily, while the lowest specific energy consumption of 378 kJ/kg was also recorded. It should be noted that the highest conversion efficiency of 99.3 %, achieved at reactor temperature of 60 °C, ultrasonication time of 10 min, and alcohol/oil molar ratio of 8:1, was not favorable as the associated specific energy consumption was higher at 395 kJ/kg. Overall, the low power, high frequency piezoelectric-based ultrasonic module could be regarded as an efficient and reliable technology for intensifying the transesterification process in terms of energy consumption, conversion efficiency, and processing time, in comparison with high power, low frequency ultrasonic system reported previously. Finally, this technology could also be considered for designing, developing, and retrofitting chemical reactors being employed for non-biofuel applications as well.

  9. Zero phase sequence impedance and tank heating model for three phase three leg core type power transformers coupling magnetic field and electric circuit equations in finite element software

    Energy Technology Data Exchange (ETDEWEB)

    Ngnegueu, T.; Mailhot, M.; Munar, A. [Jeumont Schneider Transformateurs, Lyon (France); Sacotte, M. [France-Transfo. Voie romaine, Mezieres-Les-Metz (France)


    In this paper, the authors present a finite element model for the calculation of zero phase sequence reactance for three phase three leg core type power transformers. An axisymmetrical approximation is assumed. A simplified model is used to assess the tank`s hottest spot temperature.

  10. A physical description of fission product behavior fuels for advanced power reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Kaganas, G.; Rest, J.; Nuclear Engineering Division; Florida International Univ.


    The Global Nuclear Energy Partnership (GNEP) is considering a list of reactors and nuclear fuels as part of its chartered initiative. Because many of the candidate materials have not been explored experimentally under the conditions of interest, and in order to economize on program costs, analytical support in the form of combined first principle and mechanistic modeling is highly desirable. The present work is a compilation of mechanistic models developed in order to describe the fission product behavior of irradiated nuclear fuel. The mechanistic nature of the model development allows for the possibility of describing a range of nuclear fuels under varying operating conditions. Key sources include the FASTGRASS code with an application to UO{sub 2} power reactor fuel and the Dispersion Analysis Research Tool (DART ) with an application to uranium-silicide and uranium-molybdenum research reactor fuel. Described behavior mechanisms are divided into subdivisions treating fundamental materials processes under normal operation as well as the effect of transient heating conditions on these processes. Model topics discussed include intra- and intergranular gas-atom and bubble diffusion, bubble nucleation and growth, gas-atom re-solution, fuel swelling and ?scion gas release. In addition, the effect of an evolving microstructure on these processes (e.g., irradiation-induced recrystallization) is considered. The uranium-alloy fuel, U-xPu-Zr, is investigated and behavior mechanisms are proposed for swelling in the {alpha}-, intermediate- and {gamma}-uranium zones of this fuel. The work reviews the FASTGRASS kinetic/mechanistic description of volatile ?scion products and, separately, the basis for the DART calculation of bubble behavior in amorphous fuels. Development areas and applications for physical nuclear fuel models are identified.

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

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byung Cheol; Chang, Ki Oak


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

  12. Innovative self-powered submersible microbial electrolysis cell (SMEC) for biohydrogen production from anaerobic reactors. (United States)

    Zhang, Yifeng; Angelidaki, Irini


    A self-powered submersible microbial electrolysis cell (SMEC), in which a specially designed anode chamber and external electricity supply were not needed, was developed for in situ biohydrogen production from anaerobic reactors. In batch experiments, the hydrogen production rate reached 17.8 mL/L/d at the initial acetate concentration of 410 mg/L (5 mM), while the cathodic hydrogen recovery ( [Formula: see text] ) and overall systemic coulombic efficiency (CE(os)) were 93% and 28%, respectively, and the systemic hydrogen yield ( [Formula: see text] ) peaked at 1.27 mol-H(2)/mol-acetate. The hydrogen production increased along with acetate and buffer concentration. The highest hydrogen production rate of 32.2 mL/L/d and [Formula: see text] of 1.43 mol-H(2)/mol-acetate were achieved at 1640 mg/L (20 mM) acetate and 100 mM phosphate buffer. Further evaluation of the reactor under single electricity-generating or hydrogen-producing mode indicated that further improvement of voltage output and reduction of electron losses were essential for efficient hydrogen generation. In addition, alternate exchanging the electricity-assisting and hydrogen-producing function between the two cell units of the SMEC was found to be an effective approach to inhibit methanogens. Furthermore, 16S rRNA genes analysis showed that this special operation strategy resulted same microbial community structures in the anodic biofilms of the two cell units. The simple, compact and in situ applicable SMEC offers new opportunities for reactor design for a microbial electricity-assisted biohydrogen production system.

  13. Production of Fission Product 99Mo using High-Enriched Uranium Plates in Polish Nuclear Research Reactor MARIA: Technology and Neutronic Analysis

    Directory of Open Access Journals (Sweden)

    Jaroszewicz Janusz


    Full Text Available The main objective of 235U irradiation is to obtain the 99mTc isotope, which is widely used in the domain of medical diagnostics. The decisive factor determining its availability, despite its short lifetime, is a reaction of radioactive decay of 99Mo into 99mTc. One of the possible sources of molybdenum can be achieved in course of the 235U fission reaction. The paper presents activities and the calculation results obtained upon the feasibility study on irradiation of 235U targets for production of 99Mo in the MARIA research reactor. Neutronic calculations and analyses were performed to estimate the fission products activity for uranium plates irradiated in the reactor. Results of dummy targets irradiation as well as irradiation uranium plates have been presented. The new technology obtaining 99Mo is based on irradiation of high-enriched uranium plates in standard reactor fuel channel and calculation of the current fission power generation. Measurements of temperatures and the coolant flow in the molybdenum installation carried out in reactor SAREMA system give online information about the current fission power generated in uranium targets. The corrective factors were taken into account as the heat generation from gamma radiation from neighbouring fuel elements as well as heat exchange between channels and the reactor pool. The factors were determined by calibration measurements conducted with aluminium mock-up of uranium plates. Calculations of fuel channel by means of REBUS code with fine mesh structure and libraries calculated by means of WIMS-ANL code were performed.

  14. Zero-power receiver

    Energy Technology Data Exchange (ETDEWEB)

    Brocato, Robert W.


    An unpowered signal receiver and a method for signal reception detects and responds to very weak signals using pyroelectric devices as impedance transformers and/or demodulators. In some embodiments, surface acoustic wave devices (SAW) are also used. Illustrative embodiments include satellite and long distance terrestrial communications applications.

  15. Passive residual energy utilization system in thermal cycles on water-cooled power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Placco, Guilherme M.; Guimaraes, Lamartine N.F., E-mail:, E-mail: [Instituto de Estudos Avancados (IEAV/DCTA) Sao Jose dos Campos, SP (Brazil); Santos, Rubens S. dos, E-mail: [Instituto de Engenharia Nuclear (IEN/CNEN -RJ), Rio de Janeiro, RJ (Brazil)


    This work presents a concept of a residual energy utilization in nuclear plants thermal cycles. After taking notice of the causes of the Fukushima nuclear plant accident, an idea arose to adapt a passive thermal circuit as part of the ECCS (Emergency Core Cooling System). One of the research topics of IEAv (Institute for Advanced Studies), as part of the heat conversion of a space nuclear power system is a passive multi fluid turbine. One of the main characteristics of this device is its passive capability of staying inert and be brought to power at moments notice. During the first experiments and testing of this passive device, it became clear that any small amount of gas flow would generate power. Given that in the first stages of the Fukushima accident and even during the whole event there was plenty availability of steam flow that would be the proper condition to make the proposed system to work. This system starts in case of failure of the ECCS, including loss of site power, loss of diesel generators and loss of the battery power. This system does not requires electricity to run and will work with bleed steam. It will generate enough power to supply the plant safety system avoiding overheating of the reactor core produced by the decay heat. This passive system uses a modified Tesla type turbine. With the tests conducted until now, it is possible to ensure that the operation of this new turbine in a thermal cycle is very satisfactory and it performs as expected. (author)

  16. Process Model of A Fusion Fuel Recovery System for a Direct Drive IFE Power Reactor (United States)

    Natta, Saswathi; Aristova, Maria; Gentile, Charles


    A task has been initiated to develop a detailed representative model for the fuel recovery system (FRS) in the prospective direct drive inertial fusion energy (IFE) reactor. As part of the conceptual design phase of the project, a chemical process model is developed in order to observe the interaction of system components. This process model is developed using FEMLAB Multiphysics software with the corresponding chemical engineering module (CEM). Initially, the reactants, system structure, and processes are defined using known chemical species of the target chamber exhaust. Each step within the Fuel recovery system is modeled compartmentally and then merged to form the closed loop fuel recovery system. The output, which includes physical properties and chemical content of the products, is analyzed after each step of the system to determine the most efficient and productive system parameters. This will serve to attenuate possible bottlenecks in the system. This modeling evaluation is instrumental in optimizing and closing the fusion fuel cycle in a direct drive IFE power reactor. The results of the modeling are presented in this paper.

  17. Two-Dimensional Neutronic and Fuel Cycle Analysis of the Transatomic Power Molten Salt Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Betzler, Benjamin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Worrall, Andrew [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Robertson, Sean [Transatomic Power Corporation, Cambridge, MA (United States); Dewan, Leslie [Transatomic Power Corporation, Cambridge, MA (United States); Massie, Mark [Transatomic Power Corporation, Cambridge, MA (United States)


    This status report presents the results from the first phase of the collaboration between Transatomic Power Corporation (TAP) and Oak Ridge National Laboratory (ORNL) to provide neutronic and fuel cycle analysis of the TAP core design through the Department of Energy Gateway for Accelerated Innovation in Nuclear, Nuclear Energy Voucher program. The TAP design is a molten salt reactor using movable moderator rods to shift the neutron spectrum in the core from mostly epithermal at beginning of life to thermal at end of life. Additional developments in the ChemTriton modeling and simulation tool provide the critical moderator-to-fuel ratio searches and time-dependent parameters necessary to simulate the continuously changing physics in this complex system. Results from simulations with these tools show agreement with TAP-calculated performance metrics for core lifetime, discharge burnup, and salt volume fraction, verifying the viability of reducing actinide waste production with this design. Additional analyses of time step sizes, mass feed rates and enrichments, and isotopic removals provide additional information to make informed design decisions. This work further demonstrates capabilities of ORNL modeling and simulation tools for analysis of molten salt reactor designs and strongly positions this effort for the upcoming three-dimensional core analysis.

  18. Monte Carlo simulation of a research reactor with nominal power of 7 MW to design new control safety rods (United States)

    Shoushtari, M. K.; Kakavand, T.; Sadat Kiai, S. M.; Ghaforian, H.


    The Monte Carlo simulation has been established for a research reactor with nominal power of 7 MW. A detailed model of the reactor core was employed including standard and control fuel elements, reflectors, irradiation channels, control rods, reactor pool and thermal column. The following physical parameters of reactor core were calculated for the present LEU core: core reactivity ( ρ), control rod (CR) worth, thermal and epithermal neutron flux distributions, shutdown margin and delayed neutron fraction. Reduction of unfavorable effects of blockage probability of control safety rod (CSR)s in their interiors because of not enough space in their sites, and lack of suitable capabilities to fabricate very thin plates for CSR cladding, is the main aim of the present study. Making the absorber rod thinner and CSR cladding thicker by introducing a better blackness absorbing material and a new stainless steel alloy, respectively, are two studied ways to reduce the effects of mentioned problems.

  19. Novel Co(III)/Co(II) mixed valence compound [Co(bapen)Br2]2[CoBr4] (bapen = N,N‧-bis(3-aminopropyl)ethane-1,2-diamine): Synthesis, crystal structure and magnetic properties (United States)

    Smolko, Lukáš; Černák, Juraj; Kuchár, Juraj; Miklovič, Jozef; Boča, Roman


    Green crystals of Co(III)/Co(II) mixed valence compound [Co(bapen)Br2]2[CoBr4] (bapen = N,N‧-bis(3-aminopropyl)ethane-1,2-diamine) were isolated from the aqueous system CoBr2 - bapen - HBr, crystallographically studied and characterized by elemental analysis and IR spectroscopy. Its ionic crystal structure is built up of [Co(bapen)Br2]+ cations and [CoBr4]2- anions. The Co(III) central atoms within the complex cations are hexacoordinated (donor set trans-N4Br2) with bromido ligands placed in the axial positions. The Co(II) atoms exhibit distorted tetrahedral coordination. Beside ionic forces weak Nsbnd H⋯Br intermolecular hydrogen bonding interactions contribute to the stability of the structure. Temperature variable magnetic measurements confirm the S = 3/2 behavior with the zero-field splitting of an intermediate strength: D/hc = 8.7 cm-1.

  20. An impact source localization technique for a nuclear power plant by using sensors of different types. (United States)

    Choi, Young-Chul; Park, Jin-Ho; Choi, Kyoung-Sik


    In a nuclear power plant, a loose part monitoring system (LPMS) provides information on the location and the mass of a loosened or detached metal impacted onto the inner surface of the primary pressure boundary. Typically, accelerometers are mounted on the surface of a reactor vessel to localize the impact location caused by the impact of metallic substances on the reactor system. However, in some cases, the number of accelerometers is not sufficient to estimate the impact location precisely. In such a case, one of useful methods is to utilize other types of sensor that can measure the vibration of the reactor structure. For example, acoustic emission (AE) sensors are installed on the reactor structure to detect leakage or cracks on the primary pressure boundary. However, accelerometers and AE sensors have a different frequency range. The frequency of interest of AE sensors is higher than that of accelerometers. In this paper, we propose a method of impact source localization by using both accelerometer signals and AE signals, simultaneously. The main concept of impact location estimation is based on the arrival time difference of the impact stress wave between different sensor locations. However, it is difficult to find the arrival time difference between sensors, because the primary frequency ranges of accelerometers and AE sensors are different. To overcome the problem, we used phase delays of an envelope of impact signals. This is because the impact signals from the accelerometer and the AE sensor are similar in the whole shape (envelope). To verify the proposed method, we have performed experiments for a reactor mock-up model and a real nuclear power plant. The experimental results demonstrate that we can enhance the reliability and precision of the impact source localization. Therefore, if the proposed method is applied to a nuclear power plant, we can obtain the effect of additional installed sensors.

  1. TRIP-Br2 promotes oncogenesis in nude mice and is frequently overexpressed in multiple human tumors

    Directory of Open Access Journals (Sweden)

    Peh Bee


    Full Text Available Abstract Background Members of the TRIP-Br/SERTAD family of mammalian transcriptional coregulators have recently been implicated in E2F-mediated cell cycle progression and tumorigenesis. We, herein, focus on the detailed functional characterization of the least understood member of the TRIP-Br/SERTAD protein family, TRIP-Br2 (SERTAD2. Methods Oncogenic potential of TRIP-Br2 was demonstrated by (1 inoculation of NIH3T3 fibroblasts, which were engineered to stably overexpress ectopic TRIP-Br2, into athymic nude mice for tumor induction and (2 comprehensive immunohistochemical high-throughput screening of TRIP-Br2 protein expression in multiple human tumor cell lines and human tumor tissue microarrays (TMAs. Clinicopathologic analysis was conducted to assess the potential of TRIP-Br2 as a novel prognostic marker of human cancer. RNA interference of TRIP-Br2 expression in HCT-116 colorectal carcinoma cells was performed to determine the potential of TRIP-Br2 as a novel chemotherapeutic drug target. Results Overexpression of TRIP-Br2 is sufficient to transform murine fibroblasts and promotes tumorigenesis in nude mice. The transformed phenotype is characterized by deregulation of the E2F/DP-transcriptional pathway through upregulation of the key E2F-responsive genes CYCLIN E, CYCLIN A2, CDC6 and DHFR. TRIP-Br2 is frequently overexpressed in both cancer cell lines and multiple human tumors. Clinicopathologic correlation indicates that overexpression of TRIP-Br2 in hepatocellular carcinoma is associated with a worse clinical outcome by Kaplan-Meier survival analysis. Small interfering RNA-mediated (siRNA knockdown of TRIP-Br2 was sufficient to inhibit cell-autonomous growth of HCT-116 cells in vitro. Conclusion This study identifies TRIP-Br2 as a bona-fide protooncogene and supports the potential for TRIP-Br2 as a novel prognostic marker and a chemotherapeutic drug target in human cancer.

  2. Neutronic design studies of a conceptual DCLL fusion reactor for a DEMO and a commercial power plant (United States)

    Palermo, I.; Veredas, G.; Gómez-Ros, J. M.; Sanz, J.; Ibarra, A.


    Neutronic analyses or, more widely, nuclear analyses have been performed for the development of a dual-coolant He/LiPb (DCLL) conceptual design reactor. A detailed three-dimensional (3D) model has been examined and optimized. The design is based on the plasma parameters and functional materials of the power plant conceptual studies (PPCS) model C. The initial radial-build for the detailed model has been determined according to the dimensions established in a previous work on an equivalent simplified homogenized reactor model. For optimization purposes, the initial specifications established over the simplified model have been refined on the detailed 3D design, modifying material and dimension of breeding blanket, shield and vacuum vessel in order to fulfil the priority requirements of a fusion reactor in terms of the fundamental neutronic responses. Tritium breeding ratio, energy multiplication factor, radiation limits in the TF coils, helium production and displacements per atom (dpa) have been calculated in order to demonstrate the functionality and viability of the reactor design in guaranteeing tritium self-sufficiency, power efficiency, plasma confinement, and re-weldability and structural integrity of the components. The paper describes the neutronic design improvements of the DCLL reactor, obtaining results for both DEMO and power plant operational scenarios.

  3. Worldwide assessment of steam-generator problems in pressurized-water-reactor nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Woo, H.H.; Lu, S.C.


    Objective is to assess the reliability of steam generators of pressurized water reactor (PWR) power plants in the United States and abroad. The assessment is based on operation experience of both domestic and foreign PWR plants. The approach taken is to collect and review papers and reports available from the literature as well as information obtained by contacting research institutes both here and abroad. This report presents the results of the assessment. It contains a general background of PWR plant operations, plant types, and materials used in PWR plants. A review of the worldwide distribution of PWR plants is also given. The report describes in detail the degradation problems discovered in PWR steam generators: their causes, their impacts on the performance of steam generators, and the actions to mitigate and avoid them. One chapter is devoted to operating experience of PWR steam generators in foreign countries. Another discusses the improvements in future steam generator design.

  4. Photochemical/Microchannel Plasma Reactors Driven By High Power Vacuum Ultraviolet Lamps (United States)

    Shin, Chul; Park, Sung-Jin; Eden, Gary


    Experiments are being conducted in which molecular dissociation or other chemical reactions in microchannel plasmas are accelerated by the introduction of vacuum ultraviolet photons. Initial emphasis is being placed on recently-developed Xe2 lamps that are efficient sources of 172 nm (h ν 7.2 eV) photons. Thin, flat lamps, fabricated from fused silica and having microcavity arrays internal to the lamp, have been developed by the University of Illinois and Eden Park Illumination and produce intensities above 200 mW/cm2. Integrating such lamps into a microcavity plasma reactor yields a hybrid photochemical/plasma system in which product yield and power consumption can be optimized. The selectivity of photodissociation in generating radicals and atomic fragments offers new synergies in plasma processing. Data concerning CO2 dissociation in arrays of microchannel plasmas, and the modification of this process by external 172 nm radiation, will be presented.

  5. Survey of electronics capability for SP-100 space reactor power system applications (United States)

    Manvi, Ram; Fujita, Tosh


    Because of reports indicating improvements in the radiation tolerance of some electronic parts, a survey was recently performed by SP-100 project personnel to determine the advisability of revising SP-100 SRPS (space reactor power systems) allowable neutron and gamma dose rates in order to reduce the size and mass of the radiation shield and thereby achieve system mass reductions. The survey results indicate that recent developments to increase the radiation tolerance of a limited set of electronics parts do not justify increasing the allowable SP-100 dose rates for electronic components. Specifically, the recent improvements on a limited set of parts do not justify increasing the current SP-100 allowable specifications of 5 x 10 exp 5 rads gamma dose and 1 x 10 exp 13 neutrons/sq cm fluence. However, if the improvements of 108 rads for gammas and 10 exp 15 neutrons/sq cm can be extended to a wide range of parts, significant mass savings would result.

  6. Big ambitions for small reactors as investors size up power options

    Energy Technology Data Exchange (ETDEWEB)

    Shepherd, John [nuclear24, Redditch (United Kingdom)


    Earlier this year, US nuclear developer NuScale Power completed a study for the UK's National Nuclear Laboratory (NNL) that supported the suitability of NuScale's small modular reactor (SMR) technology for the effective disposition of plutonium. The UK is a frontrunner to compete in the SMR marketplace, both in terms of technological capabilities, trade and political commitment. Industry observers are openly speculating whether SMR design and construction could start to move ahead faster than 'big and conventional' nuclear construction projects - not just in the UK but worldwide. Economies of scale could increase the attraction of SMRs to investors and the general public.

  7. Evaluation of a Business Case for Safeguards by Design in Nuclear Power Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Wood, Thomas W.; Seward, Amy M.; Lewis, Valerie A.; Gitau, Ernest TN; Zentner, Michael D.


    Safeguards by Design (SbD) is a well-known paradigm for consideration and incorporation of safeguards approaches and associated design features early in the nuclear facility development process. This paradigm has been developed as part of the Next Generation Safeguards Initiative (NGSI), and has been accepted as beneficial in many discussions and papers on NGSI or specific technologies under development within NGSI. The Office of Nuclear Safeguards and Security funded the Pacific Northwest National Laboratory to examine the business case justification of SbD for nuclear power reactors. Ultimately, the implementation of SbD will rely on the designers of nuclear facilities. Therefore, it is important to assess the incentives which will lead designers to adopt SbD as a standard practice for nuclear facility design. This report details the extent to which designers will have compelling economic incentives to adopt SbD.

  8. Synthetic fuel production via carbon neutral cycles with high temperature nuclear reactors as a power source

    Energy Technology Data Exchange (ETDEWEB)

    Konarek, E.; Coulas, B.; Sarvinis, J. [Hatch Ltd., Mississauga, Ontario (Canada)


    This paper analyzes a number of carbon neutral cycles, which could be used to produce synthetic hydrocarbon fuels. Synthetic hydrocarbons are produced via the synthesis of Carbon Monoxide and Hydrogen. The . cycles considered will either utilize Gasification processes, or carbon capture as a source of feed material. In addition the cycles will be coupled to a small modular Nuclear Reactor (SMR) as a power and heat source. The goal of this analysis is to reduce or eliminate the need to transport diesel and other fossil fuels to remote regions and to provide a carbon neutral, locally produced hydrocarbon fuel for remote communities. The technical advantages as well as the economic case are discussed for each of the cycles presented. (author)

  9. Conceptual design of the Fast-Liner Reactor (FLR) for fusion power

    Energy Technology Data Exchange (ETDEWEB)

    Moses, R.W.; Krakowski, R.A.; Miller, R.L.


    The generation of fusion power from the Fast-Liner Reactor (FLR) concept envisages the implosion of a thin (3-mm) metallic cylinder (0.2-m radius by 0.2-m length) onto a preinjected plasma. This plasma would be heated to thermonuclear temperatures by adiabatic compression, pressure confinement would be provided by the liner inertia, and thermal insulation of the wall-confined plasma would be established by an embedded azimuthal magnetic field. A 2- to 3-mu s burn would follow the approx. 10/sup 4/ m/s radial implosion and would result in a thermonuclear yield equal to 10 to 15 times the energy initially invested into the liner kinetic energy. For implosions occurring once every 10 s a gross thermal power of 430 MWt would be generated. The results of a comprehensive systems study of both physics and technology (economics) optima are presented. Despite unresolved problems associated with both the physics and technology of the FLR, a conceptual power plant design is presented.

  10. Annual report of Power Reactor and Nuclear Fuel Development Corporation, fiscal year 1996

    Energy Technology Data Exchange (ETDEWEB)



    The experimental FBR `Joyo` has continued the irradiation operation at 100 MWt. After the 11th periodic inspection, the 30th cycle operation was carried out. The cumulative operation time as of the end of the fiscal year was 51,630 hours, and the cumulative heat output was about 4.2 billion kWh. The prototype FBR `Monju` has succeeded in electric power generation in August, 1995, but the sodium leak accident occurred in December, 1995. The elucidation of the cause of the sodium leak accident and the total inspection for the safety have been carried out. As for FBRs, the research and development of the reactor physics, the design of a large FBR, the equipment systems, the fuel and materials, the structures and the safety have been advanced. The ATR `Fugen` Power Station has continued the operation smoothly, and as of the end of the fiscal year, the total generated electric power was about 17.3 billion kWh, and the capacity factor was 66.3%. It boasts about the result of using MOX fuel. The exploration of uranium resources, the development of uranium conversion, uranium enrichment and plutonium fuel, the reprocessing of spent fuel, the development of environmental technology for radioactive waste, creative and innovative research and development, safety control and safety research and others are reported. (K.I.)

  11. Investigation on the mechanism and applications of the reaction Cl2+2HBr=2HCl+Br2

    Institute of Scientific and Technical Information of China (English)


    The mechanism of reaction Cl2+2HBr=2HCl+Br2 has been carefully investigated with density functional theory (DFT) at B3LYP/6-311G** level. A series of three-centred and four-cen- tred transition states have been obtained. The activation energy (138.96 and 147.24 kJ/mol, re-spectively) of two bimolecular elementary reactions Cl2+HBr→HCl+BrCl and BrCl+HBr→HCl+Br2 is smaller than the dissociation energy of Cl2, HBr and BrCl, indicating that it is favorable for the title reaction occurring in the bimolecular form. The reaction has been applied to the chemical engi-neering process of recycling Br2 from HBr. Gaseous Cl2 directly reacts with HBr gas, which pro-duces gaseous mixtures containing Br2, and liquid Br2 and HCl are obtained by cooling the mixtures and further separated by absorption with CCl4. The recovery percentage of Br2 is more than 96%, and the Cl2 remaining in liquid Br2 is less than 3.0%. The paper provides a good example of solving the difficult problem in chemical engineering with basic theory.

  12. Investigation on the mechanism and applications of the reaction Cl2+2HBr=2HCl+Br2

    Institute of Scientific and Technical Information of China (English)

    肖鹤鸣; 王遵尧; 肖继军; 陈兆旭


    The mechanism of reaction CI2+2HBr=2HCI+Br2 has been carefully investigated with density functional theory (DFT) at B3LYP/6-311G** level. A series of three-centred and four-centred transition states have been obtained. The activation energy (138.96 and 147.24 kJ/mol, respectively) of two bimolecular elementary reactions CI2+HBr→HCI+BrCI and BrCI+HBr→HCI+Br2 is smaller than the dissociation energy of CI2, HBr and BrCI, indicating that it is favorable for the title reaction occurring in the bimolecular form. The reaction has been applied to the chemical engineering process of recycling Br2 from HBr. Gaseous CI2 directly reacts with HBr gas, which produces gaseous mixtures containing Br2, and liquid Br2 and HCI are obtained by cooling the mixtures and further separated by absorption with CCI4. The recovery percentage of Br2 is more than 96%, and the CI2 remaining in liquid Br2 is less than 3.0%. The paper provides a good example of solving the difficult problem in chemical engineering with basic theory.

  13. Impact of the use of low or medium enriched uranium on the masses of space nuclear reactor power systems

    Energy Technology Data Exchange (ETDEWEB)



    The design process for determining the mass increase for the substitution of low-enriched uranium (LEU) for high-enriched uranium (HEU) in space nuclear reactor systems is an optimization process which must simultaneously consider several variables. This process becomes more complex whenever the reactor core operates on an in-core thermionic power conversion, in which the fissioning of the nuclear fuel is used to directly heat thermionic emitters, with the subsequent elimination of external power conversion equipment. The increased complexity of the optimization process for this type of system is reflected in the work reported herein, where considerably more information has been developed for the moderated in-core thermionic reactors.

  14. Closed Brayton Cycle power system with a high temperature pellet bed reactor heat source for NEP applications (United States)

    Juhasz, Albert J.; El-Genk, Mohamed S.; Harper, William B., Jr.


    Capitalizing on past and future development of high temperature gas reactor (HTGR) technology, a low mass 15 MWe closed gas turbine cycle power system using a pellet bed reactor heating helium working fluid is proposed for Nuclear Electric Propulsion (NEP) applications. Although the design of this directly coupled system architecture, comprising the reactor/power system/space radiator subsystems, is presented in conceptual form, sufficient detail is included to permit an assessment of overall system performance and mass. Furthermore, an attempt is made to show how tailoring of the main subsystem design characteristics can be utilized to achieve synergistic system level advantages that can lead to improved reliability and enhanced system life while reducing the number of parasitic load driven peripheral subsystems.

  15. Complex risk analysis for loss of electric power in liquid metal nuclear reactor by system dynamics (SD) method

    Energy Technology Data Exchange (ETDEWEB)

    Woo, Tae Ho [Seoul National Univ. (Korea, Republic of). Dept. of Nuclear Engineering


    The power stabilization of the nuclear power plants (NPPs) is investigated in the aspect of the liquid metal coolant. The quantification of the risk analysis is performed by the system dynamics (SD) method which is processed by the feedback and accumulation complex algorithms. The Vensim software package is used for the simulations, which is supported by the Monte-Carlo method. There are 2 kinds of considerations as the economic and safety properties. The result shows the stability of the operations when the power can be decided. This shows the higher efficiency of the reactor. The failure frequency is 16/60 = 27%. In the event of Power Stabilized, the failure event is in the quite lower frequency rate. The commercial use of the reactor is important in the operations. (orig.)

  16. Synthesis and crystal structure determination of Br2SeIBr polyhalogen–chalcogen

    Indian Academy of Sciences (India)

    A A Alemi; E Solaimani


    In this paper polyhalogen–chalcogen Br2SeIBr was synthesized and the crystal structure was determined by single crystal X-ray diffraction method. This compound was prepared in the temperature range 150–50°C which was brownish-red in colour and crystallized in monoclinic crystal system and space group 21/c with four molecules per unit cell. Lattice parameters were: = 6.3711(1), = 6.7522(2), = 16.8850(5) Å, = = 90°, = 95·96°, = 722·45 Å3.

  17. Phase diagram of the CsBr-CaBr2 system

    Institute of Scientific and Technical Information of China (English)

    BAO Xinhua; CHEN Nianyi; LU Wencong; CHENG Zhixuan; LUO Yunyun; LU Weiying; XIA Yiben


    The phase diagram of the CsBr-CaBr2 system was re-determined by using differential thermal analysis and high temperature and room temperature X-ray diffraction analysis. It is concluded that there are three intermediate compounds in this system: a congruently melting compound, CsCaBr3, with a melting point of 823℃ and two incongruently melting compounds, Cs2CaBr4 and Cs3Ca2Br7, whose peritectic points being 597℃ and 635℃, respectively. X-ray diffraction analysis indicated that compound CsCaBr3 is of slightly distorted perovskite structure.

  18. 10 CFR 50.60 - Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for... (United States)


    ... 10 Energy 1 2010-01-01 2010-01-01 false Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for normal operation. 50.60 Section 50.60 Energy NUCLEAR REGULATORY... of Licenses and Construction Permits § 50.60 Acceptance criteria for fracture prevention measures...

  19. Zero effluent; Efluente zero

    Energy Technology Data Exchange (ETDEWEB)

    Monteiro, Silvio Rogerio; Santos, Angelo Francisco dos [Liquigas Distribuidora S.A., Sao Paulo, SP (Brazil)


    A scenery of water shortage and the search for profitability improvement obligate the companies to exercise their creativity and to adopt alternative methods to the conventional ones to preserve the environmental resources. The 'Effluent Zero' project comes from a paradigms changing that the environmental preservation is a necessary cost. It brings a new analysis approach of this problem with the purpose to adapt the investments and operational costs with the effluents treatment to the demands of the productive processes. In Liquigas, the project brought significant results; made a potential reduction of nearly 90% in the investments of the effluents treatment systems. That means nearly 13% in reduction in the total investments in modernization and upgrade of the existents companies installations and of 1,6% in the total operational costs of the Company. Further more, it has contributed for a reduction of until 43% of the water consumption in the bottling process of the Liquefied Petroleum Gas (LPG). This way, the project resulted in effective actions of environmental protection with relevant economic benefits. (author)

  20. Technology, safety and costs of decommissioning a Reference Boiling Water Reactor Power Station. Main report. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Oak, H.D.; Holter, G.M.; Kennedy, W.E. Jr.; Konzek, G.J.


    Technology, safety and cost information is given for the conceptual decommissioning of a large (1100MWe) boiling water reactor (BWR) power station. Three approaches to decommissioning, immediate dismantlement, safe storage with deferred dismantlement and entombment, were studied to obtain comparisons between costs, occupational radiation doses, potential dose to the public and other safety impacts. It also shows the sensitivity of decommissioning safety and costs to the power rating of a BWR in the range of 200 to 1100 MWe.

  1. An intrinsically safe facility for forefront research and training on nuclear technologies — A zero-power experiment (United States)

    Alloni, D.; Borio di Tigliole, A.; Cagnazzo, M.; Magrotti, G.; Manera, S.; Panza, F.; Prata, M.; Salvini, A.; Cammi, A.; Chiesa, D.; Clemenza, M.; Previtali, E.; Sisti, M.


    The study of subcritical reactors needs some test to correctly evaluate many parameters that must be introduced in a reactor model. The Sub-critical Multiplication installation at the University of Pavia was used to measure neutron fluxes and multiplication factors that will be modelled with a MCNP Monte Carlo numerical simulation. The comparison between the measurements and the simulations were very promising demonstrating that also in a sub-critical system the Monte Carlo model could be very helpful in describing the reactor characteristics. Using all the collected data from measurements and simulations, it was possible to determine the k eff of the reactor plant with a reasonable accuracy. The ratio between the fast and the thermal component of the neutron fluxes was also determined.

  2. Structural and Physical Properties of τ--(EDO-S,S-DMEDT-TTF) 2(AuBr2)1+Y and τ-(P-S,S-DMEDT-TTF)2(AuBr2)1+Y (United States)

    Papavassiliou, G. C.; Mousdis, G. A.; Anyfantis, G. C.; Murata, K.; Li, L.; Yoshino, H.; Tajima, H.; Konoike, T.; Brooks, J. S.; Graf, D.; Choi, E. S.


    Some new findings concerning the structural, optical, transport and magnetotransport properties of τ -(EDO-S,S-DMEDT-TTF)2(AuBr2)1+y and τ -(P-S,S-DMEDT-TTF)2(AuBr2)1+y (y ≈ 0.75) are reported. Some similarities and dissimilarities in the properties are discussed. The dissimilarities are more pronounced at low temperatures and high magnetic fields. The results are compared to those obtained of other conducting multilayered systems.

  3. Ionothermal synthesis and structural characterization of [Cu(C4H6N2)4]Br2 and [Ni(C4H6N2)4]Br2

    Indian Academy of Sciences (India)

    Hong Wang; Bin Lu; Jingxiang Zhao; Qinghai Cai


    The ionothermal synthesis and spectroscopic, thermal and structural characterization of two new compounds [Cu(C4H6N2)4]Br2 (1) and [Ni(C4H6N2)4]Br2 (2) [(C4H6N2) = N-methylimidazole] are reported. In both 1 and 2, the central metal Cu (or Ni) ion adopts a square planar geometry and is bonded to the N-atoms of four terminal N-methylimidazole ligands.

  4. Analytical evaluation on loss of off-side electric power simulation of the High Temperature Engineering Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Takeshi; Nakagawa, Shigeaki; Tachibana, Yukio; Takada, Eiji; Kunitomi, Kazuhiko [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment


    A rise-to-power test of the high temperature engineering test reactor (HTTR) started on September 28 in 1999 for establishing and upgrading the technological basis for the high temperature gas-cooled reactor (HTGR). A loss of off-site electric power test of the HTTR from the normal operation under 15 and 30 MW thermal power will be carried out in the rise-to-power test. Analytical evaluations on transient behaviors of the reactor and plant during the loss of off-site electric power were conducted. These estimations are proposed as benchmark problems for the IAEA coordinated research program on 'Evaluation of HTGR Performance'. This report describes an event scenario of transient during the loss of off-site electric power, the outline of major components and system, detailed thermal and nuclear data set for these problems and pre-estimation results of the benchmark problems by an analytical code 'ACCORD' for incore and plant dynamics of the HTGR. (author)

  5. Simulation of power excursions - Osiris reactor; Simulation des excursions de puissance - pile Osiris

    Energy Technology Data Exchange (ETDEWEB)

    Pascouet, A. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires


    Following the experimental work accomplished in the U.S.A. on Borax 1 and SPERT 1 and the accident of SL 1, the 'Commissariat a l'Energie Atomique' started a research program about the safety of its own swimming Pool reactors, with regard to power excursions. The first research work led to the design of programmed explosive charges, adapted to the simulation of a power excursion. This report describes the application of these methods to the investigation of Osiris safety. (author) [French] A la suite des essais effectues aux U.S.A. sur BORAX 1 et SPERT 1 et de l'accident survenu a SL 1, le Commissariat a l'Energie Atomique a lance un programme d'etudes sur la surete de ses reacteurs piscines vis-a-vis des excursions de puissance. Les premieres etudes ont abouti A la mise au point de charges programmees capables de simuler une excursion de puissance. On trouvera dans le present rapport l'application de ces methodes a l'etude de la surete d'OSIRIS. (auteur)

  6. Experimental Evaluation of the Thermal Performance of a Water Shield for a Surface Power Reactor (United States)

    Pearson, J. Boise; Stewart, Eric T.; Reid, Robert S.


    A water based shielding system is being investigated for use on initial lunar surface power systems. The use of water may lower overall cost (as compared to development cost for other materials) and simplify operations in the setup and handling. The thermal hydraulic performance of the shield is of significant interest. The mechanism for transferring heat through the shield is natural convection. Natural convection in a representative lunar surface reactor shield design is evaluated at various power levels in the Water Shield Testbed (WST) at the NASA Marshall Space Flight Center. The experimental data from the WST is used to anchor a CFD model. Performance of a water shield on the lunar surface is then predicted by CFD models anchored to test data. The accompanying viewgraph presentation includes the following topics: 1) Testbed Configuration; 2) Core Heater Placement and Instrumentation; 3) Thermocouple Placement; 4) Core Thermocouple Placement; 5) Outer Tank Thermocouple Placement; 6) Integrated Testbed; 7) Methodology; 8) Experimental Results: Core Temperatures; 9) Experimental Results; Outer Tank Temperatures; 10) CFD Modeling; 11) CFD Model: Anchored to Experimental Results (1-g); 12) CFD MOdel: Prediction for 1/6-g; and 13) CFD Model: Comparison of 1-g to 1/6-g.

  7. Comparative study of survived displacement damage defects in iron irradiated in IFMIF and fusion power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Simakov, S.P. [Forschungszentrum Karlsruhe, Institute for Reactor Safety, D-76021 Karlsruhe (Germany)], E-mail:; Konobeyev, A.Yu.; Fischer, U.; Heinzel, V. [Forschungszentrum Karlsruhe, Institute for Reactor Safety, D-76021 Karlsruhe (Germany)


    The assessment of the primary survived defects rates in iron such as vacancies-interstitials pairs and simplest clusters have been performed for the IFMIF, fusion power plant and research reactor. This was achieved by a modified version of the NJOY code, when processing evaluated nuclear cross section file. The modifications accounted for the reduction of the available damage energy predicted by the standard NRT model by the surviving defects fractions. These fractions were picked-up from the molecular dynamics and binary collisions simulation results available in the literature. The number of primary survived and clustered defects in the {alpha}-iron irradiated in the high flux test module of IFMIF was estimated as 10 and 6 dpa/fpy or several times less than standard NRT estimates at the level of 30 dpa/fpy. The comparison with damages in iron calculated for irradiation in the first wall of fusion power plant gave however the same reduction factors, that supports the qualification of IFMIF as a fusion material irradiation facility.

  8. Defining the "proven technology" technical criterion in the reactor technology assessment for Malaysia's nuclear power program (United States)

    Anuar, Nuraslinda; Kahar, Wan Shakirah Wan Abdul; Manan, Jamal Abdul Nasir Abd


    Developing countries that are considering the deployment of nuclear power plants (NPPs) in the near future need to perform reactor technology assessment (RTA) in order to select the most suitable reactor design. The International Atomic Energy Agency (IAEA) reported in the Common User Considerations (CUC) document that "proven technology" is one of the most important technical criteria for newcomer countries in performing the RTA. The qualitative description of five desired features for "proven technology" is relatively broad and only provides a general guideline to its characterization. This paper proposes a methodology to define the "proven technology" term according to a specific country's requirements using a three-stage evaluation process. The first evaluation stage screens the available technologies in the market against a predefined minimum Technology Readiness Level (TRL) derived as a condition based on national needs and policy objectives. The result is a list of technology options, which are then assessed in the second evaluation stage against quantitative definitions of CUC desired features for proven technology. The potential technology candidates produced from this evaluation is further narrowed down to obtain a list of proven technology candidates by assessing them against selected risk criteria and the established maximum allowable total score using a scoring matrix. The outcome of this methodology is the proven technology candidates selected using an accurate definition of "proven technology" that fulfills the policy objectives, national needs and risk, and country-specific CUC desired features of the country that performs this assessment. A simplified assessment for Malaysia is carried out to demonstrate and suggest the use of the proposed methodology. In this exercise, ABWR, AP1000, APR1400 and EPR designs assumed the top-ranks of proven technology candidates according to Malaysia's definition of "proven technology".

  9. Thermal properties for the thermal-hydraulics analyses of the BR2 maximum nominal heat flux.

    Energy Technology Data Exchange (ETDEWEB)

    Dionne, B.; Kim, Y. S.; Hofman, G. L. (Nuclear Engineering Division)


    This memo describes the assumptions and references used in determining the thermal properties for the various materials used in the BR2 HEU (93% enriched in {sup 235}U) to LEU (19.75% enriched in {sup 235}U) conversion feasibility analysis. More specifically, this memo focuses on the materials contained within the pressure vessel (PV), i.e., the materials that are most relevant to the study of impact of the change of fuel from HEU to LEU. This section is regrouping all of the thermal property tables. Section 2 provides a summary of the thermal properties in form of tables while the following sections present the justification of these values. Section 3 presents a brief background on the approach used to evaluate the thermal properties of the dispersion fuel meat and specific heat capacity. Sections 4 to 7 discuss the material properties for the following materials: (i) aluminum, (ii) dispersion fuel meat (UAlx-Al and U-7Mo-Al), (iii) beryllium, and (iv) stainless steel. Section 8 discusses the impact of irradiation on material properties. Section 9 summarizes the material properties for typical operating temperatures. Appendix A elaborates on how to calculate dispersed phase's volume fraction. Appendix B shows the evolution of the BR2 maximum heat flux with burnup.

  10. Flow tests of a single fuel element coolant channel for a compact fast reactor for space power (United States)

    Springborn, R. H.


    Water flow tests were conducted on a single-fuel-element cooling channel for a nuclear concept to be used for space power. The tests established a method for measuring coolant flow rate which is applicable to water flow testing of a complete mockup of the reference reactor. The inlet plenum-to-outlet plenum pressure drop, which approximates the overall core pressure drop, was measured and correlated with flow rate. This information can be used for reactor coolant flow and heat transfer calculations. An analytical study of the flow characteristics was also conducted.

  11. Pebble Bed Reactor Power Systems for Lunar Outposts: Long Operation Life and End-of Life Storage (United States)

    El-Genk, Mohamed S.; Schriener, Timothy M.


    The Pellet Bed Reactor(PeBR) and power system for supporting future lunar outposts offer many desirable design, operation and safety features and address post operation storage of spent nuclear fuel. In addition to its long, full power operation life of 66 year, the PeBR is launched without fuel and loaded after placement below grade on the lunar surface with spherical fuel pellets, designed to fully contain fission products. The fuel pellets(~1.0 cm dia.) are launched separately in subcritical canisters. The post-operation PeBR is stored below grade for > 300 year to allow the radioactivity in the spent fuel to decay to a sufficiently low level. The PeBR power system, designed for avoidance of single point failures in reactor cooling and energy conversion, nominally generates ~100 kWe at a thermal efficiency of ~ 21%. In addition to the sectored reactor core, it uses three Closed Brayton Cycle loops with centrifugal flow turbo-machines for energy conversion and He-Xe(40 g/mol) binary gas mixture working fluid and reactor coolant.

  12. Formation of a nuclear reactor's molten core bath in a crucible-type corium catcher for a nuclear power station equipped with VVER reactors (United States)

    Beshta, S. V.; Vitol', S. A.; Granovskii, V. S.; Kalyago, E. K.; Kovtunova, S. V.; Krushinov, E. V.; Sulatskaya, M. B.; Sulatskii, A. A.; Khabenskii, V. B.; Al'Myashev, V. I.; Gusarov, V. V.


    Results from a calculation study on analyzing the formation of a melt bath in a crucible-type catcher for the conditions of a severe accident at a nuclear power station equipped with VVER-1000 reactors are presented. It is shown that the heat loads exerted on the water-cooled walls of the corium catcher shell are limited to a permissible level at which the necessary margins to nucleate boiling crisis and to destruction are ensured under the conditions of thermal and mechanical loading of the shell. An important role of sacrificial material in the efficient operation of the corium catcher is pointed out.

  13. Comparative CFD analyses of liquid metal cooled reactor for lunar surface power

    Energy Technology Data Exchange (ETDEWEB)

    Schriener, Timothy M. [Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM (United States); Chemical and Nuclear Engineering Department, University of New Mexico, Albuquerque, NM (United States); El-Genk, Mohamed S., E-mail: [Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM (United States); Chemical and Nuclear Engineering Department, University of New Mexico, Albuquerque, NM (United States); Mechanical Engineering Department, University of New Mexico, Albuquerque, NM (United States)


    Highlights: • Performed 3-D thermal-hydraulics analyses of a NaK-56 cooled, sector core reactor for lunar surface power. • Investigated effects mesh refinements and choice of two-layer k–ε and SST k–ω models on results. • Choice of turbulence model slightly affects calculated temperatures and velocities. • Estimates of pressure losses require a finer mesh for convergence than temperature and velocities fields. • SST k–ω model provides more details of mixing eddies, but requires more computation time. - Abstract: This paper presents the results of comparative CFD and thermal-hydraulics analyses of the Solid Core–Sectored Compact Reactor (SC-SCoRe) for lunar surface power. This fast-neutron spectrum, liquid NaK-56 cooled reactor is loaded with highly enriched UN fuel. It nominally generates 1.0 MW{sub th} for ∼21 full-power years at NaK-56 inlet and exit temperatures of 850 K and 900 K. The analyses examine the realizable two-layer k–ε and the Shear Stress Transport (SST) k–ω turbulence models, with different numerical mesh refinements, for simulating the performance of the SC-SCoRe core during nominal operation as well as investigate the effect on the computation time and calculated parameters. In addition, the results calculated for a single tri-lobe flow channel are compared to those obtained using the Detached Eddy Simulation (DES), a hybrid LES and RANS turbulence model. The numerical mesh refinement beyond 2 × 10{sup 7} cells in the flow channels and the choice of turbulence model slightly affect the calculated fuel, core structure, and liquid metal temperatures. They more strongly affect the pressure losses and the intensity of flow mixing and the formation of turbulence eddies in the inlet and exit plenums and exit duct. With the same numerical mesh refinements, the total computation time with the k–ω model is 40–120% longer than with the k-ε model, while the calculated operation parameters are almost identical. The flow

  14. Power Law or Logarithmic Law?—A data Analysis for Zero Pressure Gradient Turbulent Boundary Layers with Low Reδs

    Institute of Scientific and Technical Information of China (English)



    The paper presents an analysis of two-dimensional zero pressure gradient(ZPG) turbulent boundary layers(TBL) with regard to the application of power laws,only TBL with low Reynolds number 300power law of the form u+=Cpow*yα,This power law region is not a priori identical with the overlap region.An algorithm for the determination of the wall skin friction using the power law is proposed.The method was applied with good result to ZPG TBL and to adverse pressure gradient(APG) TBL.To brdge the gap between the wall and the power law region an approach for the turbulent viscosity is suggested.

  15. Solid State Reactor Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Mays, G.T.


    The Solid State Reactor (SSR) is an advanced reactor concept designed to take advantage of Oak Ridge National Laboratory's (ORNL's) recently developed graphite foam that has enhanced heat transfer characteristics and excellent high-temperature mechanical properties, to provide an inherently safe, self-regulated, source of heat for power and other potential applications. This work was funded by the U.S. Department of Energy's Nuclear Energy Research Initiative (NERI) program (Project No. 99-064) from August 1999 through September 30, 2002. The initial concept of utilizing the graphite foam as a basis for developing an advanced reactor concept envisioned that a suite of reactor configurations and power levels could be developed for several different applications. The initial focus was looking at the reactor as a heat source that was scalable, independent of any heat removal/power conversion process. These applications might include conventional power generation, isotope production and destruction (actinides), and hydrogen production. Having conducted the initial research on the graphite foam and having performed the scoping parametric analyses from neutronics and thermal-hydraulic perspectives, it was necessary to focus on a particular application that would (1) demonstrate the viability of the overall concept and (2) require a reasonably structured design analysis process that would synthesize those important parameters that influence the concept the most as part of a feasible, working reactor system. Thus, the application targeted for this concept was supplying power for remote/harsh environments and a design that was easily deployable, simplistic from an operational standpoint, and utilized the new graphite foam. Specifically, a 500-kW(t) reactor concept was pursued that is naturally load following, inherently safe, optimized via neutronic studies to achieve near-zero reactivity change with burnup, and proliferation resistant. These four major areas

  16. Questions to the reactors power upgrade of the Nuclear Power Plant of Laguna Verde; Cuestionamientos al aumento de potencia de los reactores de la Central Nuclear de Laguna Verde

    Energy Technology Data Exchange (ETDEWEB)

    Salas M, B., E-mail: [UNAM, Facultad de Ciencias, Departamento de Fisica, Circuito exterior s/n, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)


    The two reactors of the Nuclear Power Plant of Laguna Verde (NPP-L V) were subjected to power upgrade labors with the purpose of achieving 20% upgrade on the original power; these labors concluded in August 24, 2010 for the Reactor 1 and in January 16, 2011 for the Reactor 2, however in January of 2014, the NNP-L V has not received by part of the Comision Nacional de Seguridad Nuclear y Salvaguardias (CNSNS) the new Operation License to be able to work with the new power, because it does not fulfill all the necessary requirements of safety. In this work is presented and analyzed the information obtained in this respect, with data provided by the Instituto Federal de Acceso a la Informacion Publica y Proteccion de Datos (IFAI) and the Comision Federal de Electricidad (CFE) in Mexico, as well as the opinion of some workers of the NPP-L V. The Governing Board of the CFE announcement that will give special continuation to the behavior on the operation and reliability of the NPP-L V, because the frequency of not announced interruptions was increased 7 times more in the last three years. (Author)


    Energy Technology Data Exchange (ETDEWEB)

    GAVRON, VICTOR I. [Los Alamos National Laboratory; HILL, TONY S. [Los Alamos National Laboratory; PITCHER, ERIC J. [Los Alamos National Laboratory; TOVESSON, FREDERIK K. [Los Alamos National Laboratory


    measurements in progress include {sup 240}Pu and {sup 242}Pu. The United States recently announced the Global Nuclear Energy Partnership (GNEP), with the goal of closing the commercial nuclear fuel cycle while minimizing proliferation risk. GNEP achieves these goals using fast-spectrum nuclear reactors powered by new transmutation fuels that contain significant quantities of minor actinides. The proposed Materials Test Station (MTS) will provide the GNEP with a cost-effective means of obtaining domestic fast-spectrum irradiations of advanced transmutation fuel forms and structural materials, which is an important step in the fuels qualification process. The MTS will be located at the LANSCE, and will be driven by a 1.08-MW proton beam. Th epeak neutron flux in the irradiation region is 1.67 x 10{sup 15} n/cm{sup 2}/s, and the energy spectrum is similar to that of a fast reactor, with the addition of a high-energy tail. The facility is expected to operate at least 4,400 hours per year. Fuel burnup rates will exceed 4% per year, and the radiation damage rate in iron will be 18 dpa (displacements per atom) per year. The construction cost is estimated to be $73M (including 25% contingency), with annual operating costs in the range of $6M to $10M. Appropriately funded, the MTS could begin operation in 2010.

  18. Count-to-count time interval distribution analysis in a fast reactor; Estudio de la distribucion de intervalos de tiempo entre detecciones consecutivas de neutrones en un reactor rapido

    Energy Technology Data Exchange (ETDEWEB)

    Perez-Navarro Gomez, A.


    The most important kinetic parameters have been measured at the zero power fast reactor CORAL-I by means of the reactor noise analysis in the time domain, using measurements of the count-to-count time intervals. (Author) 69 refs.

  19. The Impact of Climate Changes on the Thermal Performance of a Proposed Pressurized Water Reactor: Nuclear-Power Plant

    Directory of Open Access Journals (Sweden)

    Said M. A. Ibrahim


    Full Text Available This paper presents a methodology for studying the impact of the cooling water temperature on the thermal performance of a proposed pressurized water reactor nuclear power plant (PWR NPP through the thermodynamic analysis based on the thermodynamic laws to gain some new aspects into the plant performance. The main findings of this study are that an increase of one degree Celsius in temperature of the coolant extracted from environment is forecasted to decrease by 0.39293 and 0.16% in the power output and the thermal efficiency of the nuclear-power plant considered, respectively.

  20. Power Challenges of Large Scale Research Infrastructures: the Square Kilometer Array and Solar Energy Integration; Towards a zero-carbon footprint next generation telescope

    CERN Document Server

    Barbosa, Domingos; Ruiz, Valeriano; Silva, Manuel; Verdes-Montenegro, Lourdes; Santander-Vela, Juande; Maia, Dalmiro; Antón, Sonia; van Ardenne, Arnold; Vetter, Matthias; Kramer, Michael; Keller, Reinhard; Pereira, Nuno; Silva, Vitor


    The Square Kilometer Array (SKA) will be the largest Global science project of the next two decades. It will encompass a sensor network dedicated to radioastronomy, covering two continents. It will be constructed in remote areas of South Africa and Australia, spreading over 3000Km, in high solar irradiance latitudes. Solar Power supply is therefore an option to power supply the SKA and contribute to a zero carbon footprint next generation telescope. Here we outline the major characteristics of the SKA and some innovation approaches on thermal solar energy Integration with SKA prototypes.