WorldWideScience

Sample records for general electric test reactor

  1. Dynamic Response Testing in an Electrically Heated Reactor Test Facility

    Science.gov (United States)

    Bragg-Sitton, Shannon M.; Morton, T. J.

    2006-01-01

    Non-nuclear testing can be a valuable tool in development of a space nuclear power or propulsion system. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and full nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics, and assess potential design improvements at a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE-100a heat pipe cooled, electrically heated reactor and heat exchanger hardware, utilizing a one-group solution to the point kinetics equations to simulate the expected neutronic response of the system (Bragg-Sitton, 2005). The current paper applies the same testing methodology to a direct drive gas cooled reactor system, demonstrating the applicability of the testing methodology to any reactor type and demonstrating the variation in system response characteristics in different reactor concepts. In each testing application, core power transients were controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. Although both system designs utilize a fast spectrum reactor, the method of cooling the reactor differs significantly, leading to a variable system response that can be demonstrated and assessed in a non-nuclear test facility.

  2. Dynamic Response Testing in an Electrically Heated Reactor Test Facility

    Science.gov (United States)

    Bragg-Sitton, Shannon M.; Morton, T. J.

    2006-01-01

    Non-nuclear testing can be a valuable tool in the development of a space nuclear power or propulsion system. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and fueled nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics, and assess potential design improvements at a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE-100a heat pipe (HP) cooled, electrically heated reactor and heat exchanger hardware, utilizing a one-group solution to the point kinetics equations to simulate the expected neutronic response of the system. Reactivity feedback calculations were then based on a bulk reactivity feedback coefficient and measured average core temperature. This paper presents preliminary results from similar dynamic testing of a direct drive gas cooled reactor system (DDG), demonstrating the applicability of the testing methodology to any reactor type and demonstrating the variation in system response characteristics in different reactor concepts. Although the HP and DDG designs both utilize a fast spectrum reactor, the method of cooling the reactor differs significantly, leading to a variable system response that can be demonstrated and assessed in a non-nuclear test facility. Planned system upgrades to allow implementation of higher fidelity dynamic testing are also discussed. Proposed DDG

  3. General Electric Reactor Protection System Unavailability, 1984-1995

    Energy Technology Data Exchange (ETDEWEB)

    C. D. Gentillon; D. Rasmuson (USNRC); H. Hamzehee; M. B. Calley; S. A. Eide; T. Wierman (INEEL)

    1999-08-01

    An analysis was performed of the safety-related performance of the reactor protection system (RPS) at U.S. General Electric commercial reactors during the period 1984 through 1995. RPS operational data were collected from the Nuclear Plant Reliability Data System and Licensee Event Reports. A risk-based analysis was performed on the data to estimate the observed unavailability of the RPS, based on a fault tree model of the system. Results were compared with existing unavailability estimates from Individual Plant Examinations and other reports.

  4. General Electric Reactor Protection System Unavailability, 1984--1995

    Energy Technology Data Exchange (ETDEWEB)

    Eide, Steven Arvid; Calley, Michael Brennan; Gentillon, Cynthia Ann; Wierman, Thomas Edward; Hamzehee, H.; Rasmuson, D.

    1999-08-01

    An analysis was performed of the safety-related performance of the reactor protection system (RPS) at U. S. General Electric commercial reactors during the period 1984 through 1995. RPS operational data were collected from the Nuclear Plant Reliability Data System and Licensee Event Reports. A risk-based analysis was performed on the data to estimate the observed unavailability of the RPS, based on a fault tree model of the system. Results were compared with existing unavailability estimates from Individual Plant Examinations and other reports.

  5. Simulation of the preliminary General Electric SP-100 space reactor concept using the ATHENA computer code

    Science.gov (United States)

    Fletcher, C. D.

    The capability to perform thermal-hydraulic analyses of a space reactor using the ATHENA computer code is demonstrated. The fast reactor, liquid-lithium coolant loops, and lithium-filled heat pipes of the preliminary General electric SP-100 design were modeled with ATHENA. Two demonstration transient calculations were performed simulating accident conditions. Calculated results are available for display using the Nuclear Plant Analyzer color graphics analysis tool in addition to traditional plots. ATHENA-calculated results appear reasonable, both for steady state full power conditions, and for the two transients. This analysis represents the first known transient thermal-hydraulic simulation using an integral space reactor system model incorporating heat pipes.

  6. REACTOR AND SHIELD PHYSICS. Comprehensive Technical Report, General Electric Direct-Air-Cycle, Aircraft Nuclear Propulsion Program.

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, W.E.; Simpson, J.D.

    1962-01-01

    This volume is one of twenty-one summarizing the Aircraft Nuclear Propulsion Program of the General Electric Company. This volume describes the experimental and theoretical work accomplished in the areas of reactor and shield physics.

  7. Safety of power transformers, power supplies, reactors and similar products - Part 1: General requirements and tests

    CERN Document Server

    International Electrotechnical Commission. Geneva

    1998-01-01

    This International Standard deals with safety aspects of power transformers, power supplies, reactors and similar products such as electrical, thermal and mechanical safety. This standard covers the following types of dry-type transformers, power supplies, including switch mode power supplies, and reactors, the windings of which may be encapsulated or non-encapsulated. It has the status of a group safety publication in accordance with IEC Guide 104.

  8. Radiation-induced electrical degradation experiments in the Japan materials testing reactor

    Science.gov (United States)

    Farnum, Eugene H.; Shikama, Tatsuo; Narui, Minoru; Sagawa, Tsutomu; Scarborough, Kent

    1996-02-01

    An experiment to measure radiation-induced electrical degradation (RIED) in a sapphire sample and in three MgO-insulated cables was conducted at the JMTR light water reactor. The materials were irradiated at about 260°C to a fluence of 3 × 1024 n/m 2 ( E > 1 MeV) with an applied DC electric field between 100 kV/m and 500 kV/m. Even though the results for the sapphire sample are somewhat ambiguous because of an unexplained offset current of about 0.6 μA substantial degradation was not observed in the sapphire: instead, radiation-induced conductivity (RIC) seemed to decrease slightly during the experiment. Substantial increase in leakage current, that increased with applied electric field, occurred in the MgO-insulated cables. This increased conductivity disappeared when the reactor was shut down and sample temperature returned to ambient. However, the physical degradation apparently remained in the material while the reactor was off because restarting the irradiation brought the conductivity back to its previous, degraded, reactor-on value. This effect is different from the RIED effect reported by Hodgson but is similar to previous results reported by Shikama et al. Considerable data were taken to determine the sample temperature and leakage currents during the irradiation.

  9. Radiation-induced electrical degradation experiments in the Japan materials testing reactor

    Energy Technology Data Exchange (ETDEWEB)

    Farnum, E.; Scharborough, K. [Los Alamos National Lab., NM (United States); Shikama, Tatsuo [and others

    1995-04-01

    The objective of this experiment is to determine the extent of degradation during neutron irradiation of electrical and optical properties of candidate dielectric materials. The goals are to identify promising dielectrics for ITER and other fusion machines for diagnostic applications and establish the basis for optimization of candidate materials. An experiment to measure radiation-induced electrical degradation (REID) in sapphire and MgO-insulated cables was conducted at the JMTR light water reactor. The materials were irradiated at about 260 {degree}C to a fluence of 3{times}10{sup 24} n/m{sup 2} (E>1 MeV) with an applied DC electric field between 100 kV/m and 500 kV/m.

  10. 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

    2000-03-01

    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)

  11. FASTER Test Reactor Preconceptual Design Report

    Energy Technology Data Exchange (ETDEWEB)

    Grandy, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Belch, H. [Argonne National Lab. (ANL), Argonne, IL (United States); Brunett, A. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Heidet, F. [Argonne National Lab. (ANL), Argonne, IL (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hoffman, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Jin, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Mohamed, W. [Argonne National Lab. (ANL), Argonne, IL (United States); Moisseytsev, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Passerini, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Sumner, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Vilim, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hayes, S. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-03-31

    The FASTER test reactor plant is a sodium-cooled fast spectrum test reactor that provides high levels of fast and thermal neutron flux for scientific research and development. The 120MWe FASTER reactor plant has a superheated steam power conversion system which provides electrical power to a local grid allowing for recovery of operating costs for the reactor plant.

  12. FASTER test reactor preconceptual design report summary

    Energy Technology Data Exchange (ETDEWEB)

    Grandy, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Belch, H. [Argonne National Lab. (ANL), Argonne, IL (United States); Brunett, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Heidet, F. [Argonne National Lab. (ANL), Argonne, IL (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hoffman, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Jin, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Mohamed, W. [Argonne National Lab. (ANL), Argonne, IL (United States); Moisseytsev, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Passerini, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Sumner, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Vilim, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hayes, Steven [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-02-29

    The FASTER reactor plant is a sodium-cooled fast spectrum test reactor that provides high levels of fast and thermal neutron flux for scientific research and development. The 120MWe FASTER reactor plant has a superheated steam power conversion system which provides electrical power to a local grid allowing for recovery of operating costs for the reactor plant.

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

    Science.gov (United States)

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

    2017-01-01

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

  14. Advanced Carbothermal Electric Reactor Project

    Data.gov (United States)

    National Aeronautics and Space Administration — The overall objective of the Phase 1 effort was to demonstrate the technical feasibility of the Advanced Carbothermal Electric (ACE) Reactor concept. Unlike...

  15. Advanced Carbothermal Electric Reactor Project

    Data.gov (United States)

    National Aeronautics and Space Administration — ORBITEC proposes to develop the Advanced Carbothermal Electric (ACE) reactor to efficiently extract oxygen from lunar regolith. Unlike state-of-the-art carbothermal...

  16. Results from the Operational Testing of the General Electric Smart Grid Capable Electric Vehicle Supply Equipment (EVSE)

    Energy Technology Data Exchange (ETDEWEB)

    Richard Barney Carlson; Don Scoffield; Brion Bennett

    2013-12-01

    The Idaho National Laboratory conducted testing and analysis of the General Electric (GE) smart grid capable electric vehicle supply equipment (EVSE), which was a deliverable from GE for the U.S. Department of Energy FOA-554. The Idaho National Laboratory has extensive knowledge and experience in testing advanced conductive and wireless charging systems though INL’s support of the U.S. Department of Energy’s Advanced Vehicle Testing Activity. This document details the findings from the EVSE operational testing conducted at the Idaho National Laboratory on the GE smart grid capable EVSE. The testing conducted on the EVSE included energy efficiency testing, SAE J1772 functionality testing, abnormal conditions testing, and charging of a plug-in vehicle.

  17. Nozzle for electric dispersion reactor

    Science.gov (United States)

    Sisson, Warren G.; Basaran, Osman A.; Harris, Michael T.

    1995-01-01

    A nozzle for an electric dispersion reactor includes two concentric electrodes, the inner one of the two delivering disperse phase fluid into a continuous phase fluid. A potential difference generated by a voltage source creates a dispersing electric field at the end of the inner electrode.

  18. Development of Electrical Capacitance Sensors for Accident Tolerant Fuel (ATF) Testing at the Transient Reactor Test (TREAT) Facility

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Maolong; Ryals, Matthew; Ali, Amir; Blandford, Edward; Jensen, Colby; Condie, Keith; Svoboda, John; O' Brien, Robert

    2016-08-01

    A variety of instruments are being developed and qualified to support the Accident Tolerant Fuels (ATF) program and future transient irradiations at the Transient Reactor Test (TREAT) facility at Idaho National Laboratory (INL). The University of New Mexico (UNM) is working with INL to develop capacitance-based void sensors for determining the timing of critical boiling phenomena in static capsule fuel testing and the volume-averaged void fraction in flow-boiling in-pile water loop fuel testing. The static capsule sensor developed at INL is a plate-type configuration, while UNM is utilizing a ring-type capacitance sensor. Each sensor design has been theoretically and experimentally investigated at INL and UNM. Experiments are being performed at INL in an autoclave to investigate the performance of these sensors under representative Pressurized Water Reactor (PWR) conditions in a static capsule. Experiments have been performed at UNM using air-water two-phase flow to determine the sensitivity and time response of the capacitance sensor under a flow boiling configuration. Initial measurements from the capacitance sensor have demonstrated the validity of the concept to enable real-time measurement of void fraction. The next steps include designing the cabling interface with the flow loop at UNM for Reactivity Initiated Accident (RIA) ATF testing at TREAT and further characterization of the measurement response for each sensor under varying conditions by experiments and modeling.

  19. Reactors for nuclear electric propulsion

    Energy Technology Data Exchange (ETDEWEB)

    Buden, D.; Angelo, J.A. Jr.

    1981-01-01

    Propulsion is the key to space exploitation and power is the key to propulsion. This paper examines the role of nuclear fission reactors as the primary power source for high specific impulse electric propulsion systems for space missions of the 1980s and 1990s. Particular mission applications include transfer to and a reusable orbital transfer vehicle from low-Earth orbit to geosynchronous orbit, outer planet exploration and reconnaissance missions, and as a versatile space tug supporting lunar resource development. Nuclear electric propulsion is examined as an indispensable component in space activities of the next two decades.

  20. Advanced Demonstration and Test Reactor Options Study

    Energy Technology Data Exchange (ETDEWEB)

    Petti, David Andrew [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Gehin, J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States); Heidet, F. [Argonne National Lab. (ANL), Argonne, IL (United States); Kinsey, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Grandy, Christopher [Argonne National Lab. (ANL), Argonne, IL (United States); Qualls, A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hoffman, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Croson, D. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-01-01

    Global efforts to address climate change will require large-scale decarbonization of energy production in the United States and elsewhere. Nuclear power already provides 20% of electricity production in the United States (U.S.) and is increasing in countries undergoing rapid growth around the world. Because reliable, grid-stabilizing, low emission electricity generation, energy security, and energy resource diversity will be increasingly valued, nuclear power’s share of electricity production has a potential to grow. In addition, there are non electricity applications (e.g., process heat, desalination, hydrogen production) that could be better served by advanced nuclear systems. Thus, the timely development, demonstration, and commercialization of advanced nuclear reactors could diversify the nuclear technologies available and offer attractive technology options to expand the impact of nuclear energy for electricity generation and non-electricity missions. The purpose of this planning study is to provide transparent and defensible technology options for a test and/or demonstration reactor(s) to be built to support public policy, innovation and long term commercialization within the context of the Department of Energy’s (DOE’s) broader commitment to pursuing an “all of the above” clean energy strategy and associated time lines. This planning study includes identification of the key features and timing needed for advanced test or demonstration reactors to support research, development, and technology demonstration leading to the commercialization of power plants built upon these advanced reactor platforms. This planning study is consistent with the Congressional language contained within the fiscal year 2015 appropriation that directed the DOE to conduct a planning study to evaluate “advanced reactor technology options, capabilities, and requirements within the context of national needs and public policy to support innovation in nuclear energy

  1. The NASA Environmentally Responsible Aviation Project/General Electric Open Rotor Test Campaign

    Science.gov (United States)

    Van Zante, Dale

    2013-01-01

    The Open Rotor is a modern version of the UnDucted Fan (UDF) that was flight tested in the late 1980's through a partnership between NASA and General Electric (GE). Tests were conducted in the 9'x15' Low Speed Wind Tunnel and the 8'x6' Supersonic Wind Tunnel starting in late 2009 and completed in early 2012. Aerodynamic and acoustic data were obtained for takeoff, approach and cruise simulations. GE was the primary partner, but other organizations were involved such as Boeing and Airbus who provided additional hardware for fuselage simulations. This test campaign provided the acoustic and performance characteristics for modern open rotor blades designs." NASA and GE conducted joint systems analysis to evaluate how well new blade designs would perform on a B737 class aircraft, and compared the results to an advanced higher bypass ratio turbofan." Acoustic shielding experiments were performed at NASA GRC and Boeing LSAF facilities to provide data for noise estimates of unconventional aircraft configurations with Open Rotor propulsion systems." The work was sponsored by NASA's aeronautics programs, including the Subsonic Fixed Wing (SFW) and the Environmentally Responsible Aviation (ERA) projects."

  2. Reactor Simulator Testing

    Science.gov (United States)

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

    2013-01-01

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

  3. Reactor Simulator Testing

    Science.gov (United States)

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

    2013-01-01

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

  4. Test reactor risk assessment methodology

    Energy Technology Data Exchange (ETDEWEB)

    Jennings, R.H.; Rawlins, J.K.; Stewart, M.E.

    1976-04-01

    A methodology has been developed for the identification of accident initiating events and the fault modeling of systems, including common mode identification, as these methods are applied in overall test reactor risk assessment. The methods are exemplified by a determination of risks to a loss of primary coolant flow in the Engineering Test Reactor.

  5. Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hogerton, John

    1964-01-01

    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.

  6. Severe accident testing of electrical penetration assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Clauss, D.B. (Sandia National Labs., Albuquerque, NM (USA))

    1989-11-01

    This report describes the results of tests conducted on three different designs of full-size electrical penetration assemblies (EPAs) that are used in the containment buildings of nuclear power plants. The objective of the tests was to evaluate the behavior of the EPAs under simulated severe accident conditions using steam at elevated temperature and pressure. Leakage, temperature, and cable insulation resistance were monitored throughout the tests. Nuclear-qualified EPAs were produced from D. G. O'Brien, Westinghouse, and Conax. Severe-accident-sequence analysis was used to generate the severe accident conditions (SAC) for a large dry pressurized-water reactor (PWR), a boiling-water reactor (BWR) Mark I drywell, and a BWR Mark III wetwell. Based on a survey conducted by Sandia, each EPA was matched with the severe accident conditions for a specific reactor type. This included the type of containment that a particular EPA design was used in most frequently. Thus, the D. G. O'Brien EPA was chosen for the PWR SAC test, the Westinghouse was chosen for the Mark III test, and the Conax was chosen for the Mark I test. The EPAs were radiation and thermal aged to simulate the effects of a 40-year service life and loss-of-coolant accident (LOCA) before the SAC tests were conducted. The design, test preparations, conduct of the severe accident test, experimental results, posttest observations, and conclusions about the integrity and electrical performance of each EPA tested in this program are described in this report. In general, the leak integrity of the EPAs tested in this program was not compromised by severe accident loads. However, there was significant degradation in the insulation resistance of the cables, which could affect the electrical performance of equipment and devices inside containment at some point during the progression of a severe accident. 10 refs., 165 figs., 16 tabs.

  7. Reactors

    CERN Document Server

    International Electrotechnical Commission. Geneva

    1988-01-01

    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.

  8. PITR: Princeton Ignition Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    1978-12-01

    The principal objectives of the PITR - Princeton Ignition Test Reactor - are to demonstrate the attainment of thermonuclear ignition in deuterium-tritium, and to develop optimal start-up techniques for plasma heating and current induction, in order to determine the most favorable means of reducing the size and cost of tokamak power reactors. This report describes the status of the plasma and engineering design features of the PITR. The PITR geometry is chosen to provide the highest MHD-stable values of beta in a D-shaped plasma, as well as ease of access for remote handling and neutral-beam injection.

  9. A Preliminary Analysis of Reactor Performance Test (LOEP) for a Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyeonil; Park, Su-Ki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The final phase of commissioning is reactor performance test, which is to prove the integrated performance and safety of the research reactor at full power with fuel loaded such as neutron power calibration, Control Absorber Rod/Second Shutdown Rod drop time, InC function test, Criticality, Rod worth, Core heat removal with natural mechanism, and so forth. The last test will be safety-related one to assure the result of the safety analysis of the research reactor is marginal enough to be sure about the nuclear safety by showing the reactor satisfies the acceptance criteria of the safety functions such as for reactivity control, maintenance of auxiliaries, reactor pool water inventory control, core heat removal, and confinement isolation. After all, the fuel integrity will be ensured by verifying there is no meaningful change in the radiation levels. To confirm the performance of safety equipment, loss of normal electric power (LOEP), possibly categorized as Anticipated Operational Occurrence (AOO), is selected as a key experiment to figure out how safe the research reactor is before turning over the research reactor to the owner. This paper presents a preliminary analysis of the reactor performance test (LOEP) for a research reactor. The results showed how different the transient between conservative estimate and best estimate will look. Preliminary analyses have shown all probable thermal-hydraulic transient behavior of importance as to opening of flap valve, minimum critical heat flux ratio, the change of flow direction, and important values of thermal-hydraulic parameters.

  10. A generalized perturbation program for CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Do Heon; Kim, Jong Kyung [Hanyang University, Seoul (Korea, Republic of); Choi, Hang Bok; Roh, Gyu Hong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Yang, Won Sik [Chosun University, Kwangju (Korea, Republic of)

    1998-12-31

    A generalized perturbation program has been developed for the purpose of estimating zonal power variation of a CANDU reactor upon refueling operation. The forward and adjoint calculation modules of RFSP code were used to construct the generalized perturbation program. The numerical algorithm for the generalized adjoint flux calculation was verified by comparing the zone power estimates upon refueling with those of forward calculation. It was, however, noticed that the truncation error from the iteration process of the generalized adjoint flux is not negligible. 2 refs., 1 figs., 1 tab. (Author)

  11. Reactor Simulator Integration and Testing

    Science.gov (United States)

    Schoenfield, M. P.; Webster, K. L.; Pearson, J. B.

    2013-01-01

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

  12. Construction and operation of parallel electric and magnetic field spectrometers for mass/energy resolved multi-ion charge exchange diagnostics on the Tokamak Fusion Test Reactor

    Science.gov (United States)

    Medley, S. S.; Roquemore, A. L.

    1998-07-01

    A novel charge exchange spectrometer using a dee-shaped region of parallel electric and magnetic fields was developed at the Princeton Plasma Physics Laboratory for neutral particle diagnostics on the Tokamak Fusion Test Reactor (TFTR). The E∥B spectrometer has an energy range of 0.5⩽A (amu)E (keV)⩽600 and provides mass-resolved energy spectra of H+, D+, and T+ (or 3He+) ion species simultaneously during a single discharge. The detector plane exhibits parallel rows of analyzed ions, each row containing the energy dispersed ions of a given mass-to-charge ratio. The detector consists of a large area microchannel plate (MCP) which is provided with three rectangular, semicontinuous active area strips, one coinciding with each of the mass rows for detection of H+, D+, and T+ (or 3He+) and each mass row has 75 energy channels. To suppress spurious signals attending operation of the plate in the magnetic fringe field of the spectrometer, the MCP was housed in a double-walled iron shield with a wire mesh ion entrance window. Using an accelerator neutron generator, the MCP neutron detection efficiency was measured to be 1.7×10-3 and 6.4×10-3 counts/neutron/cm2 for 2.5 MeV-DD and 14 MeV-DT neutrons, respectively. The design and calibration of the spectrometer are described in detail, including the effect of MCP exposure to tritium, and results obtained during high performance D-D operation on TFTR are presented to illustrate the performance of the E∥B spectrometer. The spectrometers were not used during D-T plasma operation due to the cost of providing the required radiation shielding.

  13. Dielectric Heaters for Testing Spacecraft Nuclear Reactors

    Science.gov (United States)

    Sims, William Herbert; Bitteker, Leo; Godfroy, Thomas

    2006-01-01

    A document proposes the development of radio-frequency-(RF)-driven dielectric heaters for non-nuclear thermal testing of the cores of nuclear-fission reactors for spacecraft. Like the electrical-resistance heaters used heretofore for such testing, the dielectric heaters would be inserted in the reactors in place of nuclear fuel rods. A typical heater according to the proposal would consist of a rod of lossy dielectric material sized and shaped like a fuel rod and containing an electrically conductive rod along its center line. Exploiting the dielectric loss mechanism that is usually considered a nuisance in other applications, an RF signal, typically at a frequency .50 MHz and an amplitude between 2 and 5 kV, would be applied to the central conductor to heat the dielectric material. The main advantage of the proposal is that the wiring needed for the RF dielectric heating would be simpler and easier to fabricate than is the wiring needed for resistance heating. In some applications, it might be possible to eliminate all heater wiring and, instead, beam the RF heating power into the dielectric rods from external antennas.

  14. Fast Reactor Subassembly Design Modifications for Increasing Electricity Generation Efficiency

    Energy Technology Data Exchange (ETDEWEB)

    R. Wigeland; K. Hamman

    2009-09-01

    Suggested for Track 7: Advances in Reactor Core Design and In-Core Management _____________________________________________________________________________________ Fast Reactor Subassembly Design Modifications for Increasing Electricity Generation Efficiency R. Wigeland and K. Hamman Idaho National Laboratory Given the ability of fast reactors to effectively transmute the transuranic elements as are present in spent nuclear fuel, fast reactors are being considered as one element of future nuclear power systems to enable continued use and growth of nuclear power by limiting high-level waste generation. However, a key issue for fast reactors is higher electricity cost relative to other forms of nuclear energy generation. The economics of the fast reactor are affected by the amount of electric power that can be produced from a reactor, i.e., the thermal efficiency for electricity generation. The present study is examining the potential for fast reactor subassembly design changes to improve the thermal efficiency by increasing the average coolant outlet temperature without increasing peak temperatures within the subassembly, i.e., to make better use of current technology. Sodium-cooled fast reactors operate at temperatures far below the coolant boiling point, so that the maximum coolant outlet temperature is limited by the acceptable peak temperatures for the reactor fuel and cladding. Fast reactor fuel subassemblies have historically been constructed using a large number of small diameter fuel pins contained within a tube of hexagonal cross-section, or hexcan. Due to this design, there is a larger coolant flow area next to the hexcan wall as compared to flow area in the interior of the subassembly. This results in a higher flow rate near the hexcan wall, overcooling the fuel pins next to the wall, and a non-uniform coolant temperature distribution. It has been recognized for many years that this difference in sodium coolant temperature was detrimental to achieving

  15. REACTOR FUEL ELEMENTS TESTING CONTAINER

    Science.gov (United States)

    Whitham, G.K.; Smith, R.R.

    1963-01-15

    This patent shows a method for detecting leaks in jacketed fuel elements. The element is placed in a sealed tank within a nuclear reactor, and, while the reactor operates, the element is sparged with gas. The gas is then led outside the reactor and monitored for radioactive Xe or Kr. (AEC)

  16. C Reactor overbore test facility review

    Energy Technology Data Exchange (ETDEWEB)

    Carlson, P.A.; Nilson, R.

    1964-04-24

    In 1961, large-size, smooth-bore, Zircaloy process tubes were installed in C-Reactor graphite channels that had been enlarged to 2.275 inches. These tubes were installed to provide a test and demonstration facility for the concept of overboring as a means of securing significant improvement in the production capability of the reactors, After two years of facility operation, it is now appropriate to consider the extent to which original objectives have been achieved, to re-examine the original objectives, and to consider the best future use of this unique facility. This report presents the general results of such a review and re-examination in more detail.

  17. Grey Rod Test in HANARO Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choo, K. N.; Kim, B. G.; Kang, Y. H. (and others)

    2008-08-15

    Westinghouse/KAERI/KNF agreed to perform an irradiation test in the HANARO reactor to obtain irradiation data on the new grey rods that will be part of an AP1000 system. As a preliminary test, two samples containing pure Ag (Reference) and Ag-In-Cd materials provided by Westinghouse Electric Company (WEC) were inserted in a KNF irradiation capsule of 07M-13N. The specimens were irradiated for 95.19days (4 cycles) in the CT test hole of the HANARO of a 30MW thermal output to have a fast neutron fluence of 1.11x10{sup 21}(n/cm{sup 2}) (E>1.0MeV). This report provides all the test conditions and data obtained during the irradiation test of the grey rods in HANARO requested by Westinghouse. The test was prepared according to the meeting minutes (June 26, 2007) and the on-going subject test was stopped midway by the request of Westinghouse.

  18. Testing commercial catalysts in recycle reactors

    Energy Technology Data Exchange (ETDEWEB)

    Berty, J.M.

    1979-01-01

    Recycle reactors for quality control of catalyst production and for testing new catalysts for known or new processes have the following advantages over tubular reactors: they can reproduce the physical and chemical regime which surrounds the catalyst in a commercial reactor; they can achieve high mass and heat transfer; they exhibit uniform coke deposit; and they provide independence of mass velocity and space velocity. Their disadvantage is the unconventional specification of experiments in terms of discharge concentration which derives from the implicit nature of the basic mathematical relationships. Recycle reactor test methods are outlined for quality control and for testing catalysts, e.g., supported nickel from different manufacturers, for processes whose chemistry is well known. Approaches for testing catalysts for new processes are discussed. The standard recycle reactor developed at Union Carbide Corp. and manufactured by Autoclave Engineers, and several of its modifications are described.

  19. General Relativity in Electrical Engineering

    OpenAIRE

    Leonhardt, Ulf; Philbin, Thomas G.

    2006-01-01

    In electrical engineering metamaterials have been developed that offer unprecedented control over electromagnetic fields. Here we show that general relativity lends the theoretical tools for designing devices made of such versatile materials. Given a desired device function, the theory describes the electromagnetic properties that turn this function into fact. We consider media that facilitate space-time transformations and include negative refraction. Our theory unifies the concepts operatin...

  20. Advanced Test Reactor National Scientific User Facility

    Energy Technology Data Exchange (ETDEWEB)

    Frances M. Marshall; Jeff Benson; Mary Catherine Thelen

    2011-08-01

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is a large test reactor for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The ATR is a pressurized, light-water, high flux test reactor with a maximum operating power of 250 MWth. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material irradiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. This paper highlights the ATR NSUF research program and the associated educational initiatives.

  1. Dynamic model of Fast Breeder Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Vaidyanathan, G., E-mail: vaidya@igcar.gov.i [Fast Reactor Technology Group, Indira Gandhi Center for Atomic Research, Kalpakkam (India); Kasinathan, N.; Velusamy, K. [Fast Reactor Technology Group, Indira Gandhi Center for Atomic Research, Kalpakkam (India)

    2010-04-15

    Fast Breeder Test Reactor (FBTR) is a 40 M Wt/13.2 MWe sodium cooled reactor operating since 1985. It is a loop type reactor. As part of the safety analysis the response of the plant to various transients is needed. In this connection a computer code named DYNAM was developed to model the reactor core, the intermediate heat exchanger, steam generator, piping, etc. This paper deals with the mathematical model of the various components of FBTR, the numerical techniques to solve the model, and comparison of the predictions of the code with plant measurements. Also presented is the benign response of the plant to a station blackout condition, which brings out the role of the various reactivity feedback mechanisms combined with a gradual coast down of reactor sodium flow.

  2. Electrical system regulations of the IEA-R1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mello, Jose Roberto de; Madi Filho, Tufic, E-mail: jrmello@ipen.br, E-mail: tmfilho@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2013-07-01

    The IEA-R1 reactor of the Nuclear and Energy Research Institute (IPEN-CNEN/SP), is a research reactor open pool type, designed and built by the U.S. firm Babcock and Wilcox, having, as coolant and moderator, deionized light water and beryllium and graphite, as reflectors. Until about 1988, the reactor safety systems received power from only one source of energy. As an example, it may be cited the control desk that was powered only by the vital electrical system 220V, which, in case the electricity fails, is powered by the generator group: no-break 220V. In the years 1989 and 1990, a reform of the electrical system upgrading to increase the reactor power and, also, to meet the technical standards of the ABNT (Associacao Brasileira de Normas Tecnicas) was carried out. This work has the objective of showing the relationship between the electric power system and the IEA-R1 reactor security. Also, it demonstrates that, should some electrical power interruption occur, during the reactor operation, this occurrence would not start an accident event. (author)

  3. Acrylic acid and electric power cogeneration in an SOFC reactor.

    Science.gov (United States)

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

    2009-04-21

    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.

  4. Irradiation tests of texture controlled cladding for pressurized water reactor in foreign reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ukawa, Kazunori [Hokkaido Electric Power Co., Inc., Sapporo (Japan); Abeta, Sadaaki

    1996-12-01

    PWR electric power companies and makers are promoting a high burn-up program for nuclear fuel aiming at reducing spent fuel. PCI tolerance performance of fuel clad is desirable to promote a high burnup program. It is clear that when we change the direction of a Zircaloy crystal structure by improving a manufacture process (texture control), PCI tolerance performance greatly improved. In this study, an improved clad was burnt in a R2 reactor to 61.5 GWd/t and a power ramp test was carried out in Sweden. Based on irradiation data, power ramp test data and post irradiation examination data, improvement of PCI tolerance performance was confirmed. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-01-01

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

  6. Design requirement for electrical system of an advanced research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Hoan Sung; Kim, H. K.; Kim, Y. K.; Wu, J. S.; Ryu, J. S

    2004-12-01

    An advanced research reactor is being designed since 2002 and the conceptual design has been completed this year for the several types of core. Also the fuel was designed for the potential cores. But the process system, the I and C system, and the electrical system design are under pre-conceptual stage. The conceptual design for those systems will be developed in the next year. Design requirements for the electrical system set up to develop conceptual design. The same goals as reactor design - enhance safety, reliability, economy, were applied for the development of the requirements. Also the experience of HANARO design and operation was based on. The design requirements for the power distribution, standby power supply, and raceway system will be used for the conceptual design of electrical system.

  7. Irradiation rigs in material testing reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rozenblum, F.; Gonnier, C.; Bignan, G. [CEA, Research Centers of Saclay and Cadarache (France)

    2011-07-01

    Osiris is a research reactor with a thermal power of 70 MW. It is a light-water reactor, open-core pool type, the principal aim of which is to carry out tests and irradiate structural materials and fuel elements of nuclear power plants under a high flux of neutrons, and to produce radioisotopes. Osiris operates around 200 days a year, in cycles of varying lengths from 3 to 4 weeks. A shutdown of about 10 days between two cycles allows reloading the core with fuel. Mainly 2 types of irradiation device are present: capsules for materials irradiation (CHOUCA and IRMA devices) and fuels irradiation loops (GRIFFONOS and ISABELLE). Although Osiris is still providing experiments of very good quality, it is facing obsolescence due to its ageing. Osiris is planned to be shut down during next decade. Consequently, it has been decided to launch the construction of the Jules Horowitz Reactor (JHR) in Cadarache. JHR is a water cooled reactor which provides the necessary flexibility and accessibility to manage several highly instrumented experiments, reproducing different reactor environments (water, gas or liquid metal loops), generating transient regimes (key for safety). The JHR facility includes the reactor building, including core, cooling system and the experimental bunkers connected to the core through pool wall penetrations and the auxiliary building, including pools and hot cells necessary for the experimental irradiation process. JHR core is optimised to produce high fast neutron flux to study structural material ageing and high thermal neutrons flux for fuel experiments. The conception of this first fleet of devices integrates the operational experience accumulated by the existing MTR and specifically the Osiris one

  8. Space-reactor electric systems: subsystem technology assessment

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, R.V.; Bost, D.; Determan, W.R.

    1983-03-29

    This report documents the subsystem technology assessment. For the purpose of this report, five subsystems were defined for a space reactor electric system, and the report is organized around these subsystems: reactor; shielding; primary heat transport; power conversion and processing; and heat rejection. The purpose of the assessment was to determine the current technology status and the technology potentials for different types of the five subsystems. The cost and schedule needed to develop these potentials were estimated, and sets of development-compatible subsystems were identified.

  9. Advanced burner test reactor preconceptual design report.

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y. I.; Finck, P. J.; Grandy, C.; Cahalan, J.; Deitrich, L.; Dunn, F.; Fallin, D.; Farmer, M.; Fanning, T.; Kim, T.; Krajtl, L.; Lomperski, S.; Moisseytsev, A.; Momozaki, Y.; Sienicki, J.; Park, Y.; Tang, Y.; Reed, C.; Tzanos, C; Wiedmeyer, S.; Yang, W.; Chikazawa, Y.; JAEA

    2008-12-16

    The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand, to address nuclear waste management concerns and to promote non-proliferation. Implementation of the GNEP requires development and demonstration of three major technologies: (1) Light water reactor (LWR) spent fuel separations technologies that will recover transuranics to be recycled for fuel but not separate plutonium from other transuranics, thereby providing proliferation-resistance; (2) Advanced Burner Reactors (ABRs) based on a fast spectrum that transmute the recycled transuranics to produce energy while also reducing the long term radiotoxicity and decay heat loading in the repository; and (3) Fast reactor fuel recycling technologies to recover and refabricate the transuranics for repeated recycling in the fast reactor system. The primary mission of the ABR Program is to demonstrate the transmutation of transuranics recovered from the LWR spent fuel, and hence the benefits of the fuel cycle closure to nuclear waste management. The transmutation, or burning of the transuranics is accomplished by fissioning and this is most effectively done in a fast spectrum. In the thermal spectrum of commercial LWRs, some transuranics capture neutrons and become even heavier transuranics rather than being fissioned. Even with repeated recycling, only about 30% can be transmuted, which is an intrinsic limitation of all thermal spectrum reactors. Only in a fast spectrum can all transuranics be effectively fissioned to eliminate their long-term radiotoxicity and decay heat. The Advanced Burner Test Reactor (ABTR) is the first step in demonstrating the transmutation technologies. It directly supports development of a prototype full-scale Advanced Burner Reactor, which would be followed by commercial deployment of ABRs. The primary objectives of the ABTR are: (1) To demonstrate reactor-based transmutation of transuranics as part of an

  10. Instrumentation to Enhance Advanced Test Reactor Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    J. L. Rempe; D. L. Knudson; K. G. Condie; J. E. Daw; S. C. Taylor

    2009-09-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR will support basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors has been completed. Based on this review, recommendations are made with respect to what instrumentation is needed at the ATR and a strategy has been developed for obtaining these sensors. Progress toward implementing this strategy is reported in this document. It is anticipated that this report will be updated on an annual basis.

  11. Testing General Relativity

    CERN Document Server

    Tartaglia, Angelo

    2015-01-01

    This lecture will present a review of the past and present tests of the General Relativity theory. The essentials of the theory will be recalled and the measurable effects will be listed and analyzed. The main historical confirmations of General Relativity will be described. Then, the present situation will be reviewed presenting a number of examples. The opportunities given by astrophysical and astrometric observations will be shortly discussed. Coming to terrestrial experiments the attention will be specially focused on ringlasers and a dedicated experiment for the Gran Sasso Laboratories, named by the acronym GINGER, will be presented. Mention will also be made of alternatives to the use of light, such as particle beams and superfluid rings.

  12. Standard Guide for Benchmark Testing of Light Water Reactor Calculations

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This guide covers general approaches for benchmarking neutron transport calculations in light water reactor systems. A companion guide (Guide E2005) covers use of benchmark fields for testing neutron transport calculations and cross sections in well controlled environments. This guide covers experimental benchmarking of neutron fluence calculations (or calculations of other exposure parameters such as dpa) in more complex geometries relevant to reactor surveillance. Particular sections of the guide discuss: the use of well-characterized benchmark neutron fields to provide an indication of the accuracy of the calculational methods and nuclear data when applied to typical cases; and the use of plant specific measurements to indicate bias in individual plant calculations. Use of these two benchmark techniques will serve to limit plant-specific calculational uncertainty, and, when combined with analytical uncertainty estimates for the calculations, will provide uncertainty estimates for reactor fluences with ...

  13. Performance tests for integral reactor nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Dong-Seong; Yim, Jeong-Sik; Lee, Chong-Tak; Kim, Han-Soo; Koo, Yang-Hyun; Lee, Byung-Ho; Cheon, Jin-Sik; Oh, Je-Yong

    2006-02-15

    An integral type reactor SMART plans to utilize metallic Zr-U fuel which is Zr-based alloy with 34{approx}38 wt% U. In order to verify the technologies for the design and manufacturing of the fuel and get a license, performance tests were carried out. Experimental Fuel Assembly (EFA) manufactured in KAERI is being successfully irradiated in the MIR reactor of RIAR from September 4 2004, and it has achieved burnup of 0.21 g/cc as of January 25 2006. Thermal properties of irradiated Zr-U fuel were measured. Up to the phase transformation temperature, thermal diffusivity increased linearly in proportion to temperature. However its dependence on the burnup was not significant. RIA tests with 4 unirradiated Zr-U fuel rods were performed in Kurchatov Institute to establish a safety criterion. In the case of the un-irradiated Zr-U fuel, the energy deposition during the control rod ejection accident should be less than 172 cal/g to prevent the failure accompanying fuel fragmentation and dispersal. Finally the irradiation tests of fuel rods have been performed at HANARO. The HITE-2 test was successfully completed up to a burnup of 0.31 g/cc. The HITE-3 test began in February 2004 and will be continued up to a target burnup of 0.6 g/cc.

  14. Thermal Hydraulic Tests for Reactor Core Safety

    Energy Technology Data Exchange (ETDEWEB)

    Moon, S. K.; Baek, W. P.; Chun, S. Y. (and others)

    2007-06-15

    The main objectives of the present project are to resolve the current issues of reactor core thermal hydraulics, to develop an advanced measurement and analytical techniques, and to perform reactor core safety verification tests. 6x6 reflood experiments, various heat transfer experiments using Freon, and experiments on the spacer grids effects on the post-dryout are carried out using spacer grids developed in Korea in order to resolve the current issues of the reactor core thermal hydraulics. In order to develop a reflood heat transfer model, the detailed reflood phenomena are visualized and measured using round tube and 2x2 rod bundle. A detailed turbulent mixing phenomenon for subchannels is measured using advanced measurement techniques such as LDV and PIV. MARS and MATRA codes developed in Korea are assessed, verified and improved using the obtained experimental data. Finally, a systematic quality assurance program and experimental data generation system has been constructed in order to increase the reliability of the experimental data.

  15. A pilot level decision analysis of thermionic reactor development strategy for nuclear electric propulsion

    Science.gov (United States)

    Menke, M. M.; Judd, B. R.

    1973-01-01

    The development policy for thermionic reactors to provide electric propulsion and power for space exploration was analyzed to develop a logical procedure for selecting development alternatives that reflect the technical feasibility, JPL/NASA project objectives, and the economic environment of the project. The partial evolution of a decision model from the underlying philosophy of decision analysis to a deterministic pilot phase is presented, and the general manner in which this decision model can be employed to examine propulsion development alternatives is illustrated.

  16. 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.

    1991-01-01

    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.

  17. Review of Transient Fuel Test Results at Sandia National Laboratories and the Potential for Future Fast Reactor Fuel Transient Testing in the Annular Core Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wright, Steven A.; Pickard, Paul S.; Parma, Edward J.; Vernon, Milton E.; Kelly, John; Tikare, Veena [Sandia National Laboratories, Org 6872 MS-1146, PO Box 5800 Albuquerque, New Mexico 87185 (United States)

    2009-06-15

    Reactor driven transient tests of fast reactor fuels may be required to support the development and certification of new fuels for Fast Reactors. The results of the transient fuel tests will likely be needed to support licensing and to provide validation data to support the safety case for a variety of proposed fast fuel types and reactors. In general reactor driven transient tests are used to identify basic phenomenology during reactor transients and to determine the fuel performance limits and margins to failure during design basis accidents such as loss of flow, loss of heat sink, and reactivity insertion accidents. This paper provides a summary description of the previous Sandia Fuel Disruption and Transient Axial Relocation tests that were performed in the Annular Core Research Reactor (ACRR) for the U.S. Nuclear Regulatory Commission almost 25 years ago. These tests consisted of a number of capsule tests and flowing gas tests that used fission heating to disrupt fresh and irradiated MOX fuel. The behavior of the fuel disruption, the generation of aerosols and the melting and relocation of fuel and cladding was recorded on high speed cinematography. This paper will present videos of the fuel disruption that was observed in these tests which reveal stark differences in fuel behavior between fresh and irradiated fuel. Even though these tests were performed over 25 years ago, their results are still relevant to today's reactor designs. These types of transient tests are again being considered by the Advanced Fuel Cycle Initiative to support the Global Nuclear Energy Partnership because of the need to perform tests on metal fuels and transuranic fuels. Because the Annular Core Research Reactor is the only transient test facility available within the US, a brief summary of Sandia's continued capability to perform these tests in the ACRR will also be provided. (authors)

  18. Pellet bed reactor concept for nuclear electric propulsion

    Science.gov (United States)

    El-Genk, Mohamed S.; Morley, Nicholas J.; Juhasz, Albert

    1993-01-01

    For Nuclear Electric Propulsion (NEP) applications, gas cooled nuclear reactors with dynamic energy conversion systems offer high specific power and low total mass. This paper describes the Pellet Bed Reactor (PeBR) concept for potential NEP missions to Mars. The helium cooled, 75-80 MWt PeBR, consists of a single annular fuel region filled with a randomly packed bed of spherical fuel pellets, is designed for multiple starts, and offers unique safety and operation features. Each fuel pellet, about 8-10 mm in diameter, is composed of hundreds of TRISO type fuel microspheres embedded in a graphite matrix for a full retention of fission products. To eliminate the likelihood of a single-point failure, the annular core of the PeBR is divided into three 120° sectors. Each sector is self contained and separate and capable of operating and being cooled on its own and in cooperation with either one or two other sectors. Each sector is coupled to a separate, 5 MWe Closed Brayton Cycle (CBC) energy conversion unit and is subcritical for safe handling and launching. In the event of a failure of the cooling system of a core sector, the reactor power level may be reduced, allowing adjacent sectors to convect the heat away using their own cooling system, thus maintaining reactor operation. Also, due to the absence of an internal core structure in the PeBR core, fueling of the reactor can easily be performed either at the launch facility or in orbit, and refueling can be accomplished in orbit as needed to extend the power system lifetime

  19. High Temperature Gas-Cooled Test Reactor Options Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-08-01

    Preliminary scoping calculations are being performed for a 100 MWt gas-cooled test reactor. The initial design uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to identify some reactor design features to investigate further. Current status of the effort is described.

  20. Multi-Purpose Thermal Hydraulic Loop: Advanced Reactor Technology Integral System Test (ARTIST) Facility for Support of Advanced Reactor Technologies

    Energy Technology Data Exchange (ETDEWEB)

    James E. O' Brien; Piyush Sabharwall; SuJong Yoon

    2001-11-01

    Effective and robust high temperature heat transfer systems are fundamental to the successful deployment of advanced reactors for both power generation and non-electric applications. Plant designs often include an intermediate heat transfer loop (IHTL) with heat exchangers at either end to deliver thermal energy to the application while providing isolation of the primary reactor system. In order to address technical feasibility concerns and challenges a new high-temperature multi-fluid, multi-loop test facility “Advanced Reactor Technology Integral System Test facility” (ARTIST) is under development at the Idaho National Laboratory. The facility will include three flow loops: high-temperature helium, molten salt, and steam/water. Details of some of the design aspects and challenges of this facility, which is currently in the conceptual design phase, are discussed

  1. Final Physics Report for the Engineering Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wolfe, I. B. [Savannah River Site (SRS), Aiken, SC (United States)

    1956-06-25

    This report is a summary of the physics design work performed on the Engineering Test Reactor. The ETR presents computational difficulties not found in other reactors because of the large number of experimental holes in the core. The physics of the ETR depends strongly upon the contents of the in-core experimental facilities. In order to properly evaluate the reactor; taking into account the experiments in the core, multi-region, two-dimensional calculations are required. These calculations require .the use of a large computer such as the Remington Rand Univac and are complex and expensive enough to warrant a five-stage program: 1. In the early stages of design, only preliminary two-dimensional calculations were performed .in order to obtain a rough idea of the general behavior of the reactor and its critical mass with tentative experiments in place. 2. A large amount of work was carried out in which the reactor was approximated as one with a uniform homogeneous core. With this model, detailed studies were carried out to investigate the feasibility and to obtain general design data on such points as the design and properties of the gray and black-control rods, the design of the beryllium reflector, gamma and neutron heating, the use of burnable poisons, etc. In performing these calculations, use was made of the IBM 650 PROD code obtained from KAPL. 3. With stages 1 and 2 carried out, two-dimensional calculations of the core at start-up conditions were performed on the Univac computer. 4. Detailed two-dimensional calculations of the properties of the ETR with a proposed first set of experiments in place were carried out. 5. A series of nuclear tests were performed at the reactivity measurements facility at the MTR site in order to confirm the validity of the analytical techniques in physics analysis. In performing the two-dimensional Univac calculations, the MUG code developed by KAPL and the Cuthill code developed at the David Taylor Model Basin were utilized. In

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-12-14

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

  3. TR-EDB: Test Reactor Embrittlement Data Base, Version 1

    Energy Technology Data Exchange (ETDEWEB)

    Stallmann, F.W.; Wang, J.A.; Kam, F.B.K. [Oak Ridge National Lab., TN (United States)

    1994-01-01

    The Test Reactor Embrittlement Data Base (TR-EDB) is a collection of results from irradiation in materials test reactors. It complements the Power Reactor Embrittlement Data Base (PR-EDB), whose data are restricted to the results from the analysis of surveillance capsules in commercial power reactors. The rationale behind their restriction was the assumption that the results of test reactor experiments may not be applicable to power reactors and could, therefore, be challenged if such data were included. For this very reason the embrittlement predictions in the Reg. Guide 1.99, Rev. 2, were based exclusively on power reactor data. However, test reactor experiments are able to cover a much wider range of materials and irradiation conditions that are needed to explore more fully a variety of models for the prediction of irradiation embrittlement. These data are also needed for the study of effects of annealing for life extension of reactor pressure vessels that are difficult to obtain from surveillance capsule results.

  4. Reactor coolant pump testing using motor current signatures analysis

    Energy Technology Data Exchange (ETDEWEB)

    Burstein, N.; Bellamy, J.

    1996-12-01

    This paper describes reactor coolant pump motor testing carried out at Florida Power Corporation`s Crystal River plant using Framatome Technologies` new EMPATH (Electric Motor Performance Analysis and Trending Hardware) system. EMPATH{trademark} uses an improved form of Motor Current Signature Analysis (MCSA), technology, originally developed at Oak Ridge National Laboratories, for detecting deterioration in the rotors of AC induction motors. Motor Current Signature Analysis (MCSA) is a monitoring tool for motor driven equipment that provides a non-intrusive means for detecting the presence of mechanical and electrical abnormalities in the motor and the driven equipment. The base technology was developed at the Oak Ridge National Laboratory as a means for determining the affects of aging and service wear specifically on motor-operated valves used in nuclear power plant safety systems, but it is applicable to a broad range of electric machinery. MCSA is based on the recognition that an electric motor (ac or dc) driving a mechanical load acts as an efficient and permanently available transducer by sensing mechanical load variations, large and small, long-term and rapid, and converting them into variations in the induced current generated in the motor windings. The motor current variations, resulting from changes in load caused by gears, pulleys, friction, bearings, and other conditions that may change over the life of the motor, are carried by the electrical cables powering the motor and are extracted at any convenient location along the motor lead. These variations modulate the 60 Hz carrier frequency and appear as sidebands in the spectral plot.

  5. Generalized and Stability Rational Functions for Dynamic Systems of Reactor Kinetics

    Directory of Open Access Journals (Sweden)

    Ahmed E. Aboanber

    2013-01-01

    Full Text Available The base of reactor kinetics dynamic systems is a set of coupled stiff ordinary differential equations known as the point reactor kinetics equations. These equations which express the time dependence of the neutron density and the decay of the delayed neutron precursors within a reactor are first order nonlinear and essentially describe the change in neutron density within the reactor due to a change in reactivity. Outstanding the particular structure of the point kinetic matrix, a semianalytical inversion is performed and generalized for each elementary step resulting eventually in substantial time saving. Also, the factorization techniques based on using temporarily the complex plane with the analytical inversion is applied. The theory is of general validity and involves no approximations. In addition, the stability of rational function approximations is discussed and applied to the solution of the point kinetics equations of nuclear reactor with different types of reactivity. From the results of various benchmark tests with different types of reactivity insertions, the developed generalized Padé approximation (GPA method shows high accuracy, high efficiency, and stable character of the solution.

  6. Testing and examination of TMI-2 electrical components and discrete devices

    Energy Technology Data Exchange (ETDEWEB)

    Soberano, F.T.

    1982-11-01

    This report discusses the approach and results of the in situ test conducted on TMI-2 reactor building electrical components and discrete devices. Also included are the necessary presumptions and assumptions to correlate observed anomalies to the accident.

  7. Preliminary Options Assessment of Versatile Irradiation Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sen, Ramazan Sonat [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-01-01

    The objective of this report is to summarize the work undertaken at INL from April 2016 to January 2017 and aimed at analyzing some options for designing and building a versatile test reactor; the scope of work was agreed upon with DOE-NE. Section 2 presents some results related to KNK II and PRISM Mod A. Section 3 presents some alternatives to the VCTR presented in [ ] as well as a neutronic parametric study to assess the minimum power requirement needed for a 235U metal fueled fast test reactor capable to generate a fast (>100 keV) flux of 4.0 x 1015 n /cm2-s at the test location. Section 4 presents some results regarding a fundamental characteristic of test reactors, namely displacement per atom (dpa) in test samples. Section 5 presents the INL assessment of the ANL fast test reactor design FASTER. Section 6 presents a summary.

  8. A generalization of Bertrand's test

    Directory of Open Access Journals (Sweden)

    Amirali Tabatabai Adnani

    2014-07-01

    Full Text Available One of the most practical routine tests for convergence of a positive series makes use of the ratio test. If this test fails, we can use Rabbe's test. When Rabbe's test fails the next sharper criteria which may sometimes be used is the Bertrand's test. If this test fails,we can use a generalization of Bertrand's test and such tests can be continued in nitely. For simplicity, we call ratio test, Rabbe's test, Bertrand's test as the Bertrand's test of order 0, 1 and 2, respectively. In this paper, we generalize Bertrand's test in order k for natural k > 2. It is also shown that for any k, there exists a series such that the Bertrand's test of order fails, but such test of order k + 1 is useful, furthermore we show that there exists a series such that for any k, Bertrand's test of order k fails. The only prerequisite for reading this article is a standard knowledge of advanced calculus.

  9. 24 CFR 3280.810 - Electrical testing.

    Science.gov (United States)

    2010-04-01

    ... 24 Housing and Urban Development 5 2010-04-01 2010-04-01 false Electrical testing. 3280.810... URBAN DEVELOPMENT MANUFACTURED HOME CONSTRUCTION AND SAFETY STANDARDS Electrical Systems § 3280.810 Electrical testing. (a) Dielectric strength test. The wiring of each manufactured home shall be subjected...

  10. Safety requirements, facility user needs, and reactor concepts for a new Broad Application Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ryskamp, J.M. [ed.; Liebenthal, J.L.; Denison, A.B.; Fletcher, C.D.

    1992-07-01

    This report describes the EG&G Laboratory Directed Research and Development Program (LDRD) Broad Application Test Reactor (BATR) Project that was conducted in fiscal year 1991. The scope of this project was divided into three phases: a project process definition phase, a requirements development phase, and a preconceptual reactor design and evaluation phase. Multidisciplinary teams of experts conducted each phase. This report presents the need for a new test reactor, the project process definition, a set of current and projected regulatory compliance and safety requirements, a set of facility user needs for a broad range of projected testing missions, and descriptions of reactor concepts capable of meeting these requirements. This information can be applied to strategic planning to provide the Department of Energy with management options.

  11. Safety requirements, facility user needs, and reactor concepts for a new Broad Application Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ryskamp, J.M. (ed.); Liebenthal, J.L.; Denison, A.B.; Fletcher, C.D.

    1992-07-01

    This report describes the EG G Laboratory Directed Research and Development Program (LDRD) Broad Application Test Reactor (BATR) Project that was conducted in fiscal year 1991. The scope of this project was divided into three phases: a project process definition phase, a requirements development phase, and a preconceptual reactor design and evaluation phase. Multidisciplinary teams of experts conducted each phase. This report presents the need for a new test reactor, the project process definition, a set of current and projected regulatory compliance and safety requirements, a set of facility user needs for a broad range of projected testing missions, and descriptions of reactor concepts capable of meeting these requirements. This information can be applied to strategic planning to provide the Department of Energy with management options.

  12. Status of the irradiation test vehicle for testing fusion materials in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H.; Gomes, I.C.; Smith, D.L. [Argonne National Lab., IL (United States); Palmer, A.J.; Ingram, F.W. [Lockheed Martin Idaho Technologies Co., Idaho Falls, ID (United States); Wiffen, F.W. [Dept. of Energy, Germantown, MD (United States). Office of Fusion Energy

    1998-09-01

    The design of the irradiation test vehicle (ITV) for the Advanced Test Reactor (ATR) has been completed. The main application for the ITV is irradiation testing of candidate fusion structural materials, including vanadium-base alloys, silicon carbide composites, and low-activation steels. Construction of the vehicle is underway at the Lockheed Martin Idaho Technology Company (LMITCO). Dummy test trains are being built for system checkout and fine-tuning. Reactor insertion of the ITV with the dummy test trains is scheduled for fall 1998. Barring unexpected difficulties, the ITV will be available for experiments in early 1999.

  13. EBR-2 (Experimental Breeder Reactor-2), IFR (Integral Fast Reactor) prototype testing programs

    Energy Technology Data Exchange (ETDEWEB)

    Lehto, W.K.; Sackett, J.I.; Lindsay, R.W. (Argonne National Lab., Idaho Falls, ID (USA). EBR-II Div. Argonne National Lab., IL (USA)); Planchon, H.P.; Lambert, J.D.B. (Argonne National Lab., IL (USA))

    1990-01-01

    The Experimental Breeder Reactor-2 (EBR-2) is a sodium cooled power reactor supplying about 20 MWe to the Idaho National Engineering Laboratory (INEL) grid and, in addition, is the key component in the development of the Integral Fast Reactor (IFR). EBR-2's testing capability is extensive and has seen four major phases: (1) demonstration of LMFBR power plant feasibility, (2) irradiation testing for fuel and material development. (3) testing the off-normal performance of fuel and plant systems and (4) operation as the IFR prototype, developing and demonstrating the IFR technology associated with fuel and plant design. Specific programs being carried out in support of the IFR include advanced fuels and materials development and component testing. This paper discusses EBR-2 as the IFR prototype and the associated testing programs. 29 refs.

  14. Natural radioactive materials at the Arco Reactor Test Site

    Energy Technology Data Exchange (ETDEWEB)

    Singlevich, W; Healy, J W; Paas, H J; Carey, Z E

    1951-05-28

    At the request of the Division of Biology and Medicine of the AEC, the Biophysics Section of the Radiological Sciences Department at Hanford undertook the task of conducting a background survey for naturally occurring radioactive materials in the environs of the Arco Reactor Test Site in Central Idaho. This survey was part of an overall study which included meteorological measurements by the the Air Weather Service, Geological Studies by the USGS, and an ecological survey of plants and animals by members of the Idaho State College at Pocatello. In general, the measurements at Arco followed the pattern established for environmental monitoring at the Hanford Site with some additional measurements made for natural isotopes not normally of concern at Hanford. A number of analysis included materials such as plutonium and I-131 which were carried out for the purpose of establishing analytical backgrounds for the procedures used. 20 refs., 13 figs., 11 tabs.

  15. Large-scale surface dielectric barrier discharge type reactor : effect of the electric wind on the conversion effectiveness

    Energy Technology Data Exchange (ETDEWEB)

    Jolibois, J. [Univ. de Poitiers, Poitiers (France). Centre national de la recherche scientifique, Laboratoire de Catalyse en Chimie Organique; Poitiers Univ., Futuroscope Chasseneuil Cedex (France). Centre national de la recherche scientifique, Inst. Pprime; Zouzou, N.; Moreau, E. [Poitiers Univ., Futuroscope Chasseneuil Cedex (France). Centre national de la recherche scientifique, Inst. Pprime; Tatibouet, J.M. [Univ. de Poitiers, Poitiers (France). Centre national de la recherche scientifique, Laboratoire de Catalyse en Chimie Organique

    2010-07-01

    Non-thermal plasma (NTP) techniques offer an innovative approach for air pollution reduction. Most studies in NTP techniques use volumetric discharge reactors with small dimensions and low flow rates at laboratory scale. The objective of this study was to develop an air pollution control plasma reactor at industrial scale with surface discharge. Propene (C{sub 3}H{sub 6}) was oxidized at high flow rates in a large-scale plasma reactor based on surface dielectric barrier discharge (DBD). Three different configurations of surface discharges were tested with 15 ppm of C{sub 3}H{sub 6} in air at ambient temperature for a flow rate of 50 m{sup 3} per hour. The properties of these different surface discharges were analyzed using chemical measurements and 3 component particle image velocimetry (PIV) measurements. PIV measurements were used characterize the effect of the electric wind on the polluted gas airflow inside the reactor and to explain the differences of effectiveness of the three tested plasma generators. For the three plasma generators, a propene oxidation of up to 45 percent was obtained at one J per liter. The electric wind produced by the surface discharge resulted in the formation of vortices inside the plasma reactor. This electric wind can increase gas mixing inside the plasma reactor and therefore plays a key role in conversion efficiency. It was concluded that the electric wind produced by surface discharges enables the use of this type of discharge for VOC elimination at high flow rate, with the same effectiveness of volumetric discharges. 5 refs., 10 figs.

  16. Furfural Measurement Test Analysis of Transformer and Reactor Oil for Shanxi Electric Power in 2011%山西电力2011年变压器及电抗器油中糠醛普测分析

    Institute of Scientific and Technical Information of China (English)

    俞华; 周安平; 米康民

    2012-01-01

      介绍了糠醛分析的原理、判断方法,并给出了变压器、电抗器油中糠醛的普测结果。此外,还对普测结果进行了分析,最后阐述了糠醛含量分析结果的影响因素。%  The principle and judging method of furfural analysis were described,and furfural test results of oil from the transformer & reactor were made. In addition,the result of furfural measurement test was analyzed,and the factors influencing the analysis result of furfural content were presented.

  17. Modular Electric Propulsion Test Bed Aircraft Project

    Data.gov (United States)

    National Aeronautics and Space Administration — An all electric aircraft test bed is proposed to provide a dedicated development environment for the rigorous study and advancement of electrically powered aircraft....

  18. 压水堆核电站核三级屏蔽电泵试验鉴定程序%Testing of Nuclear Three-stage Shielded Electric Pump for Pressurized Water Reactor Nuclear Power Station

    Institute of Scientific and Technical Information of China (English)

    李作维; 孙悦; 薄大威

    2015-01-01

    This paper analyzes the necessity of obtaining the design/manufacture license of three generation nuclear power plants with two generation of pressurized water reactor nuclear power plant. Describe the function and structure of the three stage pump. Especially introduce the program and test method of the two generation of nuclear power station in three gener-ation pressurized water reactor nuclear power station.%分析了二代加压水堆核电站核三级屏蔽电泵取得民用核安全设备 (屏蔽泵) 设计/制造许可证的必要性, 对核三级屏蔽电泵的功能、 规格及结构进行了详细的阐述, 并重点针对二代加压水堆核电站核三级屏蔽电泵试验鉴定的程序及试验方法进行了介绍, 为核三级屏蔽电泵鉴定提供了依据.

  19. The RES Reactor. A test reactor for the French naval propulsion

    Energy Technology Data Exchange (ETDEWEB)

    Pivet, Sylvestre [CEA, Centre de Cadarache, F-13108 Saint Paul lez Durance (France); Minguet, Jean-Luc [AREVA-Technicatome, BP17, 91192 Gif-sur-Yvette (France)

    2006-07-01

    In the Cadarache nuclear research centre the French Atomic Energy Commission (CEA) operates, with the support of TECHNICATOME as nuclear operator, the experimental facilities which are necessary for the French naval propulsion program. Since the sixties these facilities have brought a large contribution to the development and to the technical support for the nuclear propulsion; they have been used also to train the French Navy operators. The last experimental reactor, the RNG, is now at the end of its life cycle after thirty years of a profitable operation. A replacement reactor is needed to sustain any evolution of the naval propulsion reactors as well as to guarantee a safe operation and a high level of availability of the existing onboard reactors. The aim of the RES program is namely to build such a test facility. Its construction program started in 2003. By the year 2009 the RES reactor will take over the mission of the RNG. We present hereafter: - A brief history of the French experimental reactors built in support to the naval propulsion, - The needs of the naval propulsion and the related objectives of the RES program, - The corresponding architecture and main characteristics of the RES facility, - The current status of the RES construction. The contents of the paper is as follows: 1. Introduction; 2. History of the French nuclear propulsion experimental reactors; 3. Needs of the naval propulsion and related objectives of the RES reactor; 4. RES architecture and main characteristics; 4.1. The pool module; 4.2. The reactor module; 4.3. The RES reactor, an innovative concept; 5. Realisation status; 6. Conclusion. To summarize, from the year 2009 the RES will be an efficient facility available for irradiation and qualification programs. Its large experimental capabilities will allow relevant fuel and core irradiations. This will give access to a real progress in the knowledge of fuel and core physics as well as in the related simulation tools. This reactor

  20. High-Temperature Gas-Cooled Test Reactor Point Design

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Laboratory; Bayless, Paul David [Idaho National Laboratory; Nelson, Lee Orville [Idaho National Laboratory; Gougar, Hans David [Idaho National Laboratory; Kinsey, James Carl [Idaho National Laboratory; Strydom, Gerhard [Idaho National Laboratory; Kumar, Akansha [Idaho National Laboratory

    2016-04-01

    A point design has been developed for a 200 MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technological readiness level, licensing approach and costs.

  1. Testing of an Integrated Reactor Core Simulator and Power Conversion System with Simulated Reactivity Feedback

    Science.gov (United States)

    Bragg-Sitton, Shannon M.; Hervol, David S.; Godfroy, Thomas J.

    2010-01-01

    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.

  2. Acoustic vibration test detects intermittent electrical discontinuities

    Science.gov (United States)

    Grieve, S. M.; Roberts, D. E.

    1970-01-01

    Nondestructive test method detects faulty electrical connections in inaccessible or hidden portions of electronic harness assemblies and connectors. Method employs readily available commercial equipment.

  3. Radioactive waste shipments to Hanford Retrievable Storage from the General Electric Vallecitos Nuclear Center, Pleasanton, California

    Energy Technology Data Exchange (ETDEWEB)

    Vejvoda, E.J.; Pottmeyer, J.A.; DeLorenzo, D.S.; Weyns-Rollosson, M.I. [Los Alamos Technical Associates, Inc., NM (United States); Duncan, D.R. [Westinghouse Hanford Co., Richland, WA (United States)

    1993-10-01

    During the next two decades the transuranic (TRU) wastes now stored in the burial trenches and storage facilities at the Hanford Site are to be retrieved, processed at the Waste Receiving and Processing Facility, and shipped to the Waste Isolation Pilot Plant near Carlsbad, New Mexico for final disposal. Approximately 3.8% of the TRU waste to be retrieved for shipment to WIPP was generated at the General Electric (GE) Vallecitos Nuclear Center (VNC) in Pleasanton, California and shipped to the Hanford Site for storage. The purpose of this report is to characterize these radioactive solid wastes using process knowledge, existing records, and oral history interviews. The waste was generated almost exclusively from the activities, of the Plutonium Fuels Development Laboratory and the Plutonium Analytical Laboratory. Section 2.0 provides further details of the VNC physical plant, facility operations, facility history, and current status. The solid radioactive wastes were associated with two US Atomic Energy Commission/US Department of Energy reactor programs -- the Fast Ceramic Reactor (FCR) program, and the Fast Flux Test Reactor (FFTR) program. These programs involved the fabrication and testing of fuel assemblies that utilized plutonium in an oxide form. The types and estimated quantities of waste resulting from these programs are discussed in detail in Section 3.0. A detailed discussion of the packaging and handling procedures used for the VNC radioactive wastes shipped to the Hanford Site is provided in Section 4.0. Section 5.0 provides an in-depth look at this waste including the following: weight and volume of the waste, container types and numbers, physical description of the waste, radiological components, hazardous constituents, and current storage/disposal locations.

  4. Reduced enrichment for research and test reactors: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1993-07-01

    The 15th annual Reduced Enrichment for Research and Test Reactors (RERTR) international meeting was organized by Ris{o} National Laboratory in cooperation with the International Atomic Energy Agency and Argonne National Laboratory. The topics of the meeting were the following: National Programs, Fuel Fabrication, Licensing Aspects, States of Conversion, Fuel Testing, and Fuel Cycle. Individual papers have been cataloged separately.

  5. Gas cooled fast breeder reactor design for a circulator test facility (modified HTGR circulator test facility)

    Energy Technology Data Exchange (ETDEWEB)

    1979-10-01

    A GCFR helium circulator test facility sized for full design conditions is proposed for meeting the above requirements. The circulator will be mounted in a large vessel containing high pressure helium which will permit testing at the same power, speed, pressure, temperature and flow conditions intended in the demonstration plant. The electric drive motor for the circulator will obtain its power from an electric supply and distribution system in which electric power will be taken from a local utility. The conceptual design decribed in this report is the result of close interaction between the General Atomic Company (GA), designer of the GCFR, and The Ralph M. Parson Company, architect/engineer for the test facility. A realistic estimate of total project cost is presented, together with a schedule for design, procurement, construction, and inspection.

  6. Spacecraft Tests of General Relativity

    Science.gov (United States)

    Anderson, John D.

    1997-01-01

    Current spacecraft tests of general relativity depend on coherent radio tracking referred to atomic frequency standards at the ground stations. This paper addresses the possibility of improved tests using essentially the current system, but with the added possibility of a space-borne atomic clock. Outside of the obvious measurement of the gravitational frequency shift of the spacecraft clock, a successor to the suborbital flight of a Scout D rocket in 1976 (GP-A Project), other metric tests would benefit most directly by a possible improved sensitivity for the reduced coherent data. For purposes of illustration, two possible missions are discussed. The first is a highly eccentric Earth orbiter, and the second a solar-conjunction experiment to measure the Shapiro time delay using coherent Doppler data instead of the conventional ranging modulation.

  7. Spacecraft Tests of General Relativity

    Science.gov (United States)

    Anderson, John D.

    1997-01-01

    Current spacecraft tests of general relativity depend on coherent radio tracking referred to atomic frequency standards at the ground stations. This paper addresses the possibility of improved tests using essentially the current system, but with the added possibility of a space-borne atomic clock. Outside of the obvious measurement of the gravitational frequency shift of the spacecraft clock, a successor to the suborbital flight of a Scout D rocket in 1976 (GP-A Project), other metric tests would benefit most directly by a possible improved sensitivity for the reduced coherent data. For purposes of illustration, two possible missions are discussed. The first is a highly eccentric Earth orbiter, and the second a solar-conjunction experiment to measure the Shapiro time delay using coherent Doppler data instead of the conventional ranging modulation.

  8. Preconceptual design of the new production reactor circulator test facility

    Energy Technology Data Exchange (ETDEWEB)

    Thurston, G.

    1990-06-01

    This report presents the results of a study of a new circulator test facility for the New Production Reactor Modular High-Temperature Gas-Cooled Reactor. The report addresses the preconceptual design of a stand-alone test facility with all the required equipment to test the Main Circulator/shutoff valve and Shutdown Cooling Circulator/shutoff valve. Each type of circulator will be tested in its own full flow, full power helium test loop. Testing will cover the entire operating range of each unit. The loop will include a test vessel, in which the circulator/valve will be mounted, and external piping. The external flow piping will include a throttle valve, flowmeter, and heat exchanger. Subsystems will include helium handling, helium purification, and cooling water. A computer-based data acquisition and control system will be provided. The estimated costs for the design and construction of this facility are included. 2 refs., 15 figs.

  9. Final safeguards analysis, high temperature lattice test reactor. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Hanthorn, H.E.; Brown, W.W.; Clark, R.G.; Heineman, R.E.; Humes, R.M.

    1966-01-01

    The PMACS `reactor-normal` signal signifies that important process variables do not exceed their set points, that various interlocks are properly set, that functional tests of the computer operation are satisfactory, and that the reactor flux level and period derived from two additional, independent, and dissimilar channels are within set limits. This safety circuit combines the features of redundancy, dissimilar components, and frequent testing which are required for best reliability. The experimental equipment auxiliary to the reactor includes two oscillator mechanisms, one to move the test cell or the adjoining cell into and out of position, the other to move small specimens in the test cell or adjoining cells. They have cooling chambers for the removal of specimens from the test cell without the necessity of cooling the reactor. A neutron chopper and time-of-flight spectrometer are provided; the neutron detectors, at the end of a 25-meter flight tube, are in an adjoining small building. Test cores may be assembled on a core dolly have a load capacity of 14,000 lb. Two wire traverse mechanisms are provided for measurements of flux distribution.

  10. Heat Pipe Reactor Dynamic Response Tests: SAFE-100 Reactor Core Prototype

    Science.gov (United States)

    Bragg-Sitton, Shannon M.

    2005-01-01

    The SAFE-I00a test article at the NASA Marshall Space Flight Center was used to simulate a variety of potential reactor transients; the SAFEl00a is a resistively heated, stainless-steel heat-pipe (HP)-reactor core segment, coupled to a gas-flow heat exchanger (HX). For these transients the core power was controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. This type of non-nuclear test is expected to provide reasonable approximation of reactor transient behavior because reactivity feedback is very simple in a compact fast reactor (simple, negative, and relatively monotonic temperature feedback, caused mostly by thermal expansion) and calculations show there are no significant reactivity effects associated with fluid in the HP (the worth of the entire inventory of Na in the core is .tests, the point kinetics model was based on core thermal expansion via deflection measurements. It was found that core deflection was a strung function of how the SAFE-100 modules were fabricated and assembled (in terms of straightness, gaps, and other tolerances). To remove the added variable of how this particular core expands as compared to a different concept, it was decided to use a temperature based feedback model (based on several thermocouples placed throughout the core).

  11. The technology development for surveillance test of reactor vessel materials

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Sun Phil; Park, Day Young; Choi, Kwen Jai

    1997-12-01

    Benchmark test was performed in accordance with the requirement of US NRC Reg. Guide DG-1053 for Kori unit-1 in order to determine best-estimated fast neutron fluence irradiated into reactor vessel. Since the uncertainty of radiation analysis comes from the calculation error due to neutron cross-section data, reactor core geometrical dimension, core source, mesh density, angular expansion and convergence criteria, evaluation of calculational uncertainty due to analytical method was performed in accordance with the regulatory guide and the proof was performed for entire analysis by comparing the measurement value obtained by neutron dosimetry located in surveillance capsule. Best-estimated neutron fluence in reactor vessel was calculated by bias factor, neutron flux measurement value/calculational value, from reanalysis result from previous 1st through 4th surveillance testing and finally fluence prediction was performed for the end of reactor life and the entire period of plant life extension. Pressurized thermal shock analysis was performed in accordance with 10 CFR 50.61 using the result of neutron fluence analysis in order to predict the life of reactor vessel material and the criteria of safe operation for Kori unit 1 was reestablished. (author). 55 refs., 55 figs.

  12. Preliminary conceptual design for electrical and I and C system of a new research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Hoan Sung; Kim, Y. K.; Kim, M. J.; Kim, H. K.; Ryu, J. S

    2004-01-01

    The core type and the process system design will be varied according to the reactor's application and capacity. A New research reactor is being designed by KAERI since 2002 and the process systems are not fixed yet. But control and instrument systems are similar to each other even though the application and the size are not same. So the C and I system that encompasses reactor protection system, reactor control system, and computer system was designed conceptually according to the requirements based on new digital technology and HANARO's proven design. The plant electrical system consists of off-site system that delivers bulk electrical power to the reactor site and on-site system that distributes and controls electrical power at the facility. The electrical system includes building service system that consist of lighting, communication, fire detection, grounding, cathodic protection, etc. also. This report describes the design requirements of on-site and off-site electric power system that set up from the codes and standards and the conceptual design based on the design requirements.

  13. Sequential probability ratio tests for reactor signal validation and sensor surveillance applications

    Energy Technology Data Exchange (ETDEWEB)

    Humenik, K. (Maryland Univ., Baltimore, MD (USA)); Gross, K.C. (Argonne National Lab., IL (USA))

    1989-11-09

    This paper examines the properties of sequential probability ratio tests (SPRT's) and the application of these tests to nuclear power reactor operation. Recently SPRT's have been applied to delayed-neutron (DN) signal data analysis using actual reactor data from the Experimental Breeder Reactor-II, which is operated by Argonne National Laboratory. The implementation of this research as part of an expert system is described. Mathematical properties of the SPRT are investigated, and theoretical results are validated with tests that use DN-signal data taken from the EBR-II in Idaho. Variations of the basic SPRT and applications to general signal validation are also explored. 16 refs., 3 figs.

  14. Moderated heat pipe thermionic reactor (MOHTR) module development and test

    Science.gov (United States)

    Merrigan, Michael A.; Trujillo, Vincent L.

    1992-01-01

    The Moderated Heat Pipe Thermionic Reactor (MOHTR) thermionic space reactor design combines the low risk technology associated with the Thermionic Fuel Element (TFE) Verification Program with the high reliability heat transfer capability of liquid metal heat pipe technology. The resulting design concept, capable of implementation over the power range of 10 to 100 kWe, offers efficiency and reliability with reduced risk of single point failures. The union of TFE and heat pipe technology is achieved by imbedding TFEs and heat pipes in a beryllium matrix to which they are thermally coupled by brazing or by liquid metal (NaK or Na) bonding. The reactor employs an array of TFE modules, each comprising a TFE, a zirconium hydride (ZrH) cylinder for neutron moderation, and heat pipes for transport of heat from the collector surface of the TFE to the waste heat radiator. An advantage of the design is the low temperature drop from the collector surface to the radiating surface. This is a result of the elimination of electrical insulation from the heat transport path through electrical isolation of the modules. The module used in this study consisted of a beryllium core, and electrical cartridge heater simulating the TFE, and three heat pipes to dissipate the waste heat. The investigation was focused on the thermal performance of the assembly, including evaluation of the sodium and braze bonding options for minimizing the thermal resistance between the elements, the temperature distribution in the beryllium matrix, and the heat pipe performance. Continuing subjects of the investigation include performance of the heat pipes through start-up transients, during normal operation, and in a single heat pipe failure mode. Secondary objectives of the investigation include correlation of analytic models for the thermionic element and module including the effects of gap thermal conductances at the modules electrically insulated surfaces.

  15. Application of Simulated Reactivity Feedback in Nonnuclear Testing of a Direct-Drive Gas-Cooled Reactor

    Science.gov (United States)

    Bragg-Sitton, S. M.; Webster, K. L.

    2007-01-01

    Nonnuclear testing can be a valuable tool in the development of an in-space nuclear power or propulsion system. In a nonnuclear test facility, electric heaters are used to simulate heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and full nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response and response characteristics, and assess potential design improvements with a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE 100a heat pipe cooled, electrically heated reactor and heat exchanger hardware. This Technical Memorandum discusses the status of the planned dynamic test methodology for implementation in the direct-drive gas-cooled reactor testing and assesses the additional instrumentation needed to implement high-fidelity dynamic testing.

  16. Materials Test-2 LOCA Simulation in the NRU Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Barner, J. O.; Hesson, G. M.; King, I. L.; Marshall, R. K.; Parchen, L. J.; Pilger, J. P.; Rausch, W. N.; Russcher, G. E.; Webb, B. J.; Wildung, N. J.; Wilson, C. L.; Wismer, M. D.; Mohr, C. L.

    1982-03-01

    A simulated loss-of-coolant accident was performed with a full-length test bundle of pressurized water reactor fuel rods. This third experiment of the program produced fuel cladding temperatures exceeding 1033 K (1400°F) for 155 s and resulted in eight ruptured fuel rods. Experiment data and initial results are presented in the form of photographs and graphical summaries.

  17. Design of an Actinide Burning, Lead-Bismuth Cooled Reactor That Produces Low Cost Electricity

    Energy Technology Data Exchange (ETDEWEB)

    C. Davis; S. Herring; P. MacDonald; K. McCarthy; V. Shah; K. Weaver (INEEL); J. Buongiorno; R. Ballinger; K. Doyoung; M. Driscoll; P. Hejzler; M. Kazimi; N. Todreas (MIT)

    1999-07-01

    The purpose of this project is to investigate the suitability of lead-bismuth cooled fast reactors for producing low-cost electricity as well as for actinide burning. The goal is to identify and analyze the key technical issues in core neutronics, materials, thermal-hydraulics, fuels, and economics associated with the development of this reactor concept. The choice of lead-bismuth for the reactor coolant is an actinide burning fast reactor offers enhanced safety and reliability. The advantages of lead-bismuth over sodium as a coolant are related to the following material characteristics: chemical inertness with air and water; higher atomic number; lower vapor pressure at operating temperatures; and higher boiling temperature. Given the status of the field, it was agreed that the focus of this investigation in the first two years will be on the assessment of approaches to optimize core and plant arrangements in order to provide maximum safety and economic potential in this type of reactor.

  18. Reactor Physics Scoping and Characterization Study on Implementation of TRIGA Fuel in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jennifer Lyons; Wade R. Marcum; Mark D. DeHart; Sean R. Morrell

    2014-01-01

    The Advanced Test Reactor (ATR), under the Reduced Enrichment for Research and Test Reactors (RERTR) Program and the Global Threat Reduction Initiative (GTRI), is conducting feasibility studies for the conversion of its fuel from a highly enriched uranium (HEU) composition to a low enriched uranium (LEU) composition. These studies have considered a wide variety of LEU plate-type fuels to replace the current HEU fuel. Continuing to investigate potential alternatives to the present HEU fuel form, this study presents a preliminary analysis of TRIGA® fuel within the current ATR fuel envelopes and compares it to the functional requirements delineated by the Naval Reactors Program, which includes: greater than 4.8E+14 fissions/s/g of 235U, a fast to thermal neutron flux ratio that is less than 5% deviation of its current value, a constant cycle power within the corner lobes, and an operational cycle length of 56 days at 120 MW. Other parameters outside those put forth by the Naval Reactors Program which are investigated herein include axial and radial power profiles, effective delayed neutron fraction, and mean neutron generation time.

  19. Airvolt Aircraft Electric Propulsion Test Stand

    Science.gov (United States)

    Samuel, Aamod; Lin, Yohan

    2015-01-01

    Development of an electric propulsion test stand that collects high-fidelity data of motor, inverter, and battery system efficiencies; thermal dynamics; and acoustics independent of manufacturer reported values will improve understanding of electric propulsion systems to be used in future aircraft. A buildup approach to this development reveals new areas of research and best practices in testing, and attempts to establish a standard for testing these systems.

  20. Innovative microbial fuel cell for electricity production from anaerobic reactors

    DEFF Research Database (Denmark)

    Min, Booki; Angelidaki, Irini

    2008-01-01

    A submersible microbial fuel cell (SMFC) was developed by immersing an anode electrode and a cathode chamber in an anaerobic reactor. Domestic wastewater was used as the medium and the inoculum in the experiments. The SMFC could successfully generate a stable voltage of 0.428 ± 0.003 V with a fixed...

  1. Reduced enrichment for research and test reactors: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1988-05-01

    The international effort to develop new research reactor fuel materials and designs based on the use of low-enriched uranium, instead of highly-enriched uranium, has made much progress during the eight years since its inception. To foster direct communication and exchange of ideas among the specialist in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at the Argonne National Laboratory, sponsored this meeting as the ninth of a series which began in 1978. All previous meetings of this series are listed on the facing page. The focus of this meeting was on the LEU fuel demonstration which was in progress at the Oak Ridge Research (ORR) reactor, not far from where the meeting was held. The visit to the ORR, where a silicide LEU fuel with 4.8 g A/cm/sup 3/ was by then in routine use, illustrated how far work has progressed.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-12-15

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

  3. Test Results From a Direct Drive Gas Reactor Simulator Coupled to a Brayton Power Conversion Unit

    Science.gov (United States)

    Hervol, David S.; Briggs, Maxwell H.; Owen, Albert K.; Bragg-Sitton, Shannon M.

    2009-01-01

    The Brayton Power Conversion Unit (BPCU) located at NASA Glenn Research Center (GRC) in Cleveland, OH is a closed cycle system incorporating a turboaltemator, recuperator, and gas cooler connected by gas ducts to an external gas heater. For this series of tests, the BPCU was modified by replacing the gas heater with the Direct Drive Gas heater or DOG. The DOG uses electric resistance heaters to simulate a fast spectrum nuclear reactor similar to those proposed for space power applications. The combined system thermal transient behavior was the focus of these tests. The BPCU was operated at various steady state points. At each point it was subjected to transient changes involving shaft rotational speed or DOG electrical input. This paper outlines the changes made to the test unit and describes the testing that took place along with the test results.

  4. Flow through reactors for organic chemistry: directly electrically heated tubular mini reactors as an enabling technology for organic synthesis

    Directory of Open Access Journals (Sweden)

    Ulrich Kunz

    2009-11-01

    Full Text Available Until recently traditional heating in organic chemistry has been done with oil heating baths or using electric heat exchangers. With the advent of microwave equipment, heating by microwaves was rapidly introduced as standard method in organic chemistry laboratories, mainly because of the convenient possibility to operate at high temperature accompanied by accelerated reaction rates. In the present contribution we discuss the method of heating small, continuously operated reactors by passing electric current directly through the reactor wall as an enabling technology in organic chemistry. The benefit of this method is that the heat is generated directly inside the reactor wall. By this means high heating rates comparable to microwave ovens can be reached but at much lower cost for the equipment. A tool for the comparison of microwave heating and traditional heating is provided. As an example kinetic data for the acid catalyzed hydrolysis of methyl formate were measured using this heating concept. The reaction is not only a suitable model but also one of industrial importance since this is the main production process for formic acid.

  5. Flow through reactors for organic chemistry: directly electrically heated tubular mini reactors as an enabling technology for organic synthesis.

    Science.gov (United States)

    Kunz, Ulrich; Turek, Thomas

    2009-11-30

    Until recently traditional heating in organic chemistry has been done with oil heating baths or using electric heat exchangers. With the advent of microwave equipment, heating by microwaves was rapidly introduced as standard method in organic chemistry laboratories, mainly because of the convenient possibility to operate at high temperature accompanied by accelerated reaction rates. In the present contribution we discuss the method of heating small, continuously operated reactors by passing electric current directly through the reactor wall as an enabling technology in organic chemistry. The benefit of this method is that the heat is generated directly inside the reactor wall. By this means high heating rates comparable to microwave ovens can be reached but at much lower cost for the equipment. A tool for the comparison of microwave heating and traditional heating is provided. As an example kinetic data for the acid catalyzed hydrolysis of methyl formate were measured using this heating concept. The reaction is not only a suitable model but also one of industrial importance since this is the main production process for formic acid.

  6. Flow through reactors for organic chemistry: directly electrically heated tubular mini reactors as an enabling technology for organic synthesis

    Science.gov (United States)

    Turek, Thomas

    2009-01-01

    Summary Until recently traditional heating in organic chemistry has been done with oil heating baths or using electric heat exchangers. With the advent of microwave equipment, heating by microwaves was rapidly introduced as standard method in organic chemistry laboratories, mainly because of the convenient possibility to operate at high temperature accompanied by accelerated reaction rates. In the present contribution we discuss the method of heating small, continuously operated reactors by passing electric current directly through the reactor wall as an enabling technology in organic chemistry. The benefit of this method is that the heat is generated directly inside the reactor wall. By this means high heating rates comparable to microwave ovens can be reached but at much lower cost for the equipment. A tool for the comparison of microwave heating and traditional heating is provided. As an example kinetic data for the acid catalyzed hydrolysis of methyl formate were measured using this heating concept. The reaction is not only a suitable model but also one of industrial importance since this is the main production process for formic acid. PMID:20300506

  7. Testing piezoelectric sensors in a nuclear reactor environment

    Science.gov (United States)

    Reinhardt, Brian T.; Suprock, Andy; Tittmann, Bernhard

    2017-02-01

    Several Department of Energy Office of Nuclear Energy (DOE-NE) programs, such as the Fuel Cycle Research and Development (FCRD), Advanced Reactor Concepts (ARC), Light Water Reactor Sustainability, and Next Generation Nuclear Power Plants (NGNP), are investigating new fuels, materials, and inspection paradigms for advanced and existing reactors. A key objective of such programs is to understand the performance of these fuels and materials during irradiation. In DOE-NE's FCRD program, ultrasonic based technology was identified as a key approach that should be pursued to obtain the high-fidelity, high-accuracy data required to characterize the behavior and performance of new candidate fuels and structural materials during irradiation testing. The radiation, high temperatures, and pressure can limit the available tools and characterization methods. In this work piezoelectric transducers capable of making these measurements are developed. Specifically, three piezoelectric sensors (Bismuth Titanate, Aluminum Nitride, and Zinc Oxide) are tested in the Massachusetts Institute of Technology Research reactor to a fast neutron fluence of 8.65×1020 nf/cm2. It is demonstrated that Bismuth Titanate is capable of transduction up to 5 × 1020 nf/cm2, Zinc Oxide is capable of transduction up to at least 6.27 × 1020 nf/cm2, and Aluminum Nitride is capable of transduction up to at least 8.65 × 1020 nf/cm2.

  8. Assessment of the Technical Maturity of Generation IV Concepts for Test or Demonstration Reactor Applications, Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-10-01

    The United States Department of Energy (DOE) commissioned a study the suitability of different advanced reactor concepts to support materials irradiations (i.e. a test reactor) or to demonstrate an advanced power plant/fuel cycle concept (demonstration reactor). As part of the study, an assessment of the technical maturity of the individual concepts was undertaken to see which, if any, can support near-term deployment. A Working Group composed of the authors of this document performed the maturity assessment using the Technical Readiness Levels as defined in DOE’s Technology Readiness Guide . One representative design was selected for assessment from of each of the six Generation-IV reactor types: gas-cooled fast reactor (GFR), lead-cooled fast reactor (LFR), molten salt reactor (MSR), supercritical water-cooled reactor (SCWR), sodium-cooled fast reactor (SFR), and very high temperature reactor (VHTR). Background information was obtained from previous detailed evaluations such as the Generation-IV Roadmap but other technical references were also used including consultations with concept proponents and subject matter experts. Outside of Generation IV activity in which the US is a party, non-U.S. experience or data sources were generally not factored into the evaluations as one cannot assume that this data is easily available or of sufficient quality to be used for licensing a US facility. The Working Group established the scope of the assessment (which systems and subsystems needed to be considered), adapted a specific technology readiness scale, and scored each system through discussions designed to achieve internal consistency across concepts. In general, the Working Group sought to determine which of the reactor options have sufficient maturity to serve either the test or demonstration reactor missions.

  9. Modular Electric Propulsion Test Bed Aircraft Project

    Data.gov (United States)

    National Aeronautics and Space Administration — A hybrid electric aircraft simulation system and test bed is proposed to provide a dedicated development environment for the rigorous study and advancement of hybrid...

  10. Enhanced in-pile instrumentation at the advanced test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rempe, J. L.; Knudson, D. L.; Daw, J. E.; Unruh, T.; Chase, B. M.; Palmer, J.; Condie, K. G.; Davis, K. L. [Idaho National Laboratory, MS 3840, P.O. Box 1625, Idaho Falls, ID 83415 (United States)

    2011-07-01

    Many of the sensors deployed at materials and test reactors cannot withstand the high flux/high temperature test conditions often requested by users at U.S. test reactors, such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory. To address this issue, an instrumentation development effort was initiated as part of the ATR National Scientific User Facility in 2007 to support the development and deployment of enhanced in-pile sensors. This paper reports results from this effort. Specifically, this paper identifies the types of sensors currently available to support in-pile irradiations and those sensors currently available to ATR users. Accomplishments from new sensor technology deployment efforts are highlighted by describing new temperature and thermal conductivity sensors now available to ATR users. Efforts to deploy enhanced in-pile sensors for detecting elongation and realtime flux detectors are also reported, and recently-initiated research to evaluate the viability of advanced technologies to provide enhanced accuracy for measuring key parameters during irradiation testing are noted. (authors)

  11. Electrical Arc Ignition Testing of Spacesuit Materials

    Science.gov (United States)

    Smith, Sarah; Gallus, Tim; Tapia, Susana; Ball, Elizabeth; Beeson, Harold

    2006-01-01

    A viewgraph presentation on electrical arc ignition testing of spacesuit materials is shown. The topics include: 1) Background; 2) Test Objectives; 3) Test Sample Materials; 4) Test Methods; 5) Scratch Test Objectives; 6) Cotton Scratch Test Video; 7) Scratch Test Results; 8) Entire Date Plot; 9) Closeup Data Plot; 10) Scratch Test Problems; 11) Poke Test Objectives; 12) Poke Test Results; 13) Poke Test Problems; 14) Wire-break Test Objectives; 15) Cotton Wire-Break Test Video; 16) High Speed Cotton Wire-break Test Video; 17) Typical Data Plot; 18) Closeup Data Plot; 19) Wire-break Test Results; 20) Wire-break Tests vs. Scratch Tests; 21) Urethane-coated Nylon; and 22) Moleskin.

  12. Development of a general learning algorithm with applications in nuclear reactor systems

    Energy Technology Data Exchange (ETDEWEB)

    Brittain, C.R.; Otaduy, P.J.; Perez, R.B.

    1989-12-01

    The objective of this study was development of a generalized learning algorithm that can learn to predict a particular feature of a process by observation of a set of representative input examples. The algorithm uses pattern matching and statistical analysis techniques to find a functional relationship between descriptive attributes of the input examples and the feature to be predicted. The algorithm was tested by applying it to a set of examples consisting of performance descriptions for 277 fuel cycles of Oak Ridge National Laboratory's High Flux Isotope Reactor (HFIR). The program learned to predict the critical rod position for the HFIR from core configuration data prior to reactor startup. The functional relationship bases its predictions on initial core reactivity, the number of certain targets placed in the center of the reactor, and the total exposure of the control plates. Twelve characteristic fuel cycle clusters were identified. Nine fuel cycles were diagnosed as having noisy data, and one could not be predicted by the functional relationship. 13 refs., 6 figs.

  13. LOCA simulation in the NRU reactor: materials test-1

    Energy Technology Data Exchange (ETDEWEB)

    Russcher, G.E.; Marshall, R.K.; Hesson, G.M.; Wildung, N.J.; Rausch, W.N.

    1981-10-01

    A simulated loss-of-coolant accident was performed with a full-length test bundle of pressurized water reactor fuel rods. This second experiment of the program produced peak fuel cladding temperatures of 1148K (1607/sup 0/F) and resulted in six ruptured fuel rods. Test data and initial results from the experiment are presented here in the form of photographs and graphical summaries. These results are also compared with the preceding prototypic thermal-hydraulic test results and with computer model test predictions.

  14. Design and Test of Advanced Thermal Simulators for an Alkali Metal-Cooled Reactor Simulator

    Science.gov (United States)

    Garber, Anne E.; Dickens, Ricky E.

    2011-01-01

    The Early Flight Fission Test Facility (EFF-TF) at NASA Marshall Space Flight Center (MSFC) has as one of its primary missions the development and testing of fission reactor simulators for space applications. A key component in these simulated reactors is the thermal simulator, designed to closely mimic the form and function of a nuclear fuel pin using electric heating. Continuing effort has been made to design simple, robust, inexpensive thermal simulators that closely match the steady-state and transient performance of a nuclear fuel pin. A series of these simulators have been designed, developed, fabricated and tested individually and in a number of simulated reactor systems at the EFF-TF. The purpose of the thermal simulators developed under the Fission Surface Power (FSP) task is to ensure that non-nuclear testing can be performed at sufficiently high fidelity to allow a cost-effective qualification and acceptance strategy to be used. Prototype thermal simulator design is founded on the baseline Fission Surface Power reactor design. Recent efforts have been focused on the design, fabrication and test of a prototype thermal simulator appropriate for use in the Technology Demonstration Unit (TDU). While designing the thermal simulators described in this paper, effort were made to improve the axial power profile matching of the thermal simulators. Simultaneously, a search was conducted for graphite materials with higher resistivities than had been employed in the past. The combination of these two efforts resulted in the creation of thermal simulators with power capacities of 2300-3300 W per unit. Six of these elements were installed in a simulated core and tested in the alkali metal-cooled Fission Surface Power Primary Test Circuit (FSP-PTC) at a variety of liquid metal flow rates and temperatures. This paper documents the design of the thermal simulators, test program, and test results.

  15. Comprehensive Technical Report, General Electric Direct-Air-Cycle Aircraft Nuclear Propulsion Program; Aircraft Nuclear Propulsion Application Studies

    Energy Technology Data Exchange (ETDEWEB)

    Comassar, S.

    1962-04-30

    This volume is one of twenty-one summarizing the Aircraft Nuclear Propulsion Program of the General Electric Company. This portion describes the studies of advanced applications of nuclear reactors that were performed, including various types of aircraft, missiles, space vehicles, ships, and portable power plants.

  16. Comprehensive Technical Report, General Electric Direct-Air-Cycle Aircraft Nuclear Propulsion Program; Aircraft Nuclear Propulsion Application Studies

    Energy Technology Data Exchange (ETDEWEB)

    Comassar, S.

    1962-04-30

    This volume is one of twenty-one summarizing the Aircraft Nuclear Propulsion Program of the General Electric Company. This portion describes the studies of advanced applications of nuclear reactors that were performed, including various types of aircraft, missiles, space vehicles, ships, and portable power plants.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1979-09-01

    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.

  18. Thermal Hydraulic Integral Effect Tests for Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Baek, W. P.; Song, C. H.; Kim, Y. S. and others

    2005-02-15

    The objectives of the project are to construct a thermal-hydraulic integral effect test facility and to perform various integral effect tests for design, operation, and safety regulation of pressurized water reactors. During the first phase of this project (1997.8{approx}2002.3), the basic technology for thermal-hydraulic integral effect tests was established and the basic design of the test facility was accomplished: a full-height, 1/300-volume-scaled full pressure facility for APR1400, an evolutionary pressurized water reactor that was developed by Korean industry. Main objectives of the present phase (2002.4{approx}2005.2), was to optimize the facility design and to construct the experimental facility. We have performed following researches: 1) Optimization of the basic design of the thermal-hydraulic integral effect test facility for PWRs - ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) - Reduced height design for APR1400 (+ specific design features of KSNP safety injection systems) - Thermal-hydraulic scaling based on three-level scaling methodology by Ishii et al. 2) Construction of the ATLAS facility - Detailed design of the test facility - Manufacturing and procurement of components - Installation of the facility 3) Development of supporting technology for integral effect tests - Development and application of advanced instrumentation technology - Preliminary analysis of test scenarios - Development of experimental procedures - Establishment and implementation of QA system/procedure.

  19. Advanced Gas Reactor (AGR)-5/6/7 Fuel Irradiation Experiments in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    A. Joseph Palmer; David A. Petti; S. Blaine Grover

    2014-04-01

    The United States Department of Energy’s Very High Temperature Reactor (VHTR) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which each consist of at least five separate capsules, are being irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gases also have on-line fission product monitoring the effluent from each capsule to track performance of the fuel during irradiation. The first two experiments (designated AGR-1 and AGR-2), have been completed. The third and fourth experiments have been combined into a single experiment designated AGR-3/4, which started its irradiation in December 2011 and is currently scheduled to be completed in April 2014. The design of the fuel qualification experiment, designated AGR-5/6/7, is well underway and incorporates lessons learned from the three previous experiments. Various design issues will be discussed with particular details related to selection of thermometry.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-07-01

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

  2. High-temperature nuclear reactor power plant cycle for hydrogen and electricity production – numerical analysis

    Directory of Open Access Journals (Sweden)

    Dudek Michał

    2016-01-01

    Full Text Available High temperature gas-cooled nuclear reactor (called HTR or HTGR for both electricity generation and hydrogen production is analysed. The HTR reactor because of the relatively high temperature of coolant could be combined with a steam or gas turbine, as well as with the system for heat delivery for high-temperature hydrogen production. However, the current development of HTR’s allows us to consider achievable working temperature up to 750°C. Due to this fact, industrial-scale hydrogen production using copper-chlorine (Cu-Cl thermochemical cycle is considered and compared with high-temperature electrolysis. Presented calculations show and confirm the potential of HTR’s as a future solution for hydrogen production without CO2 emission. Furthermore, integration of a hightemperature nuclear reactor with a combined cycle for electricity and hydrogen production may reach very high efficiency and could possibly lead to a significant decrease of hydrogen production costs.

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

    Energy Technology Data Exchange (ETDEWEB)

    Nigg, David W.; Nielsen, Joseph W.; Norman, Daren R.

    2017-07-01

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

  4. 77 FR 36014 - Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors

    Science.gov (United States)

    2012-06-15

    ... COMMISSION Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors AGENCY: Nuclear...-1277, ``Initial Test Program of Emergency Core Cooling Systems for Boiling- Water Reactors.'' This... testing features of emergency core cooling systems (ECCSs) for boiling-water reactors (BWRs). DATES...

  5. New results from pulse tests in the CABRI reactor

    Energy Technology Data Exchange (ETDEWEB)

    Schmitz, F.; Papin, J.; Haessler, M. [Institut de Proterction et de Surete Nucleaire, Saint Paul Lez Durance (France)] [and others

    1996-03-01

    At the 21st and 22nd WRSM (1,2), the motivation and objectives of the French program on the behaviour of high burnup PWR fuel under RIA conditions in the CABRI test reactor has been presented. The major results of the three first tests of the test matrix were presented and in particular REP-Na1, which failed at an unexpected low level of fuel enthalpy, was exposed to the community of nuclear safety research. At this time, no final understanding was reached for the origin of the failure. This objective is reached now. Two further tests, REP-Na4 and 5, have been performed in 1995, they demonstrated a satisfactory and safe behaviour by resisting to the early phase of severe loading during the RIA pulse test. Further examination work and analytical testing is in progress and the next tests with MOX fuel are being prepared.

  6. Advanced Test Reactor National Scientific User Facility Progress

    Energy Technology Data Exchange (ETDEWEB)

    Frances M. Marshall; Todd R. Allen; James I. Cole; Jeff B. Benson; Mary Catherine Thelen

    2012-10-01

    The Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) is one of the world’s premier test reactors for studying the effects of intense neutron radiation on reactor materials and fuels. The ATR began operation in 1967, and has operated continuously since then, averaging approximately 250 operating days per year. The combination of high flux, large test volumes, and multiple experiment configuration options provide unique testing opportunities for nuclear fuels and material researchers. The ATR is a pressurized, light-water moderated and cooled, beryllium-reflected highly-enriched uranium fueled, reactor with a maximum operating power of 250 MWth. The ATR peak thermal flux can reach 1.0 x1015 n/cm2-sec, and the core configuration creates five main reactor power lobes (regions) that can be operated at different powers during the same operating cycle. In addition to these nine flux traps there are 68 irradiation positions in the reactor core reflector tank. The test positions range from 0.5” to 5.0” in diameter and are all 48” in length, the active length of the fuel. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material radiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. Goals of the ATR NSUF are to define the cutting edge of nuclear technology research in high temperature and radiation environments, contribute to improved industry performance of current and future light water reactors, and stimulate cooperative research between user groups conducting basic and applied research. The ATR NSUF has developed partnerships with other universities and national laboratories to enable ATR NSUF researchers to perform research at these other facilities, when the research objectives

  7. Reliability and safety of the electrical power supply complex of the Hanford production reactors

    Energy Technology Data Exchange (ETDEWEB)

    Robbins, F.D.

    1960-09-15

    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.

  8. Conceptual design study of a scyllac fusion test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Thomassen, K.I. (comp.)

    1975-07-01

    The report describes a conceptual design study of a fusion test reactor based on the Scyllac toroidal theta-pinch approach to fusion. It is not the first attempt to describe the physics and technology required for demonstrating scientific feasibility of the approach, but it is the most complete design in the sense that the physics necessary to achieve the device goals is extrapolated from experimentally tested MHD theories of toroidal systems,and it uses technological systems whose engineering performance has been carefully calculated to ensure that they meet the machine requirements.

  9. Testing general relativity on accelerators

    CERN Document Server

    Kalaydzhyan, Tigran

    2015-01-01

    Within the general theory of relativity, the curvature of spacetime is related to the energy and momentum of the present matter and radiation. One of the more specific predictions of general relativity is the deflection of light and particle trajectories in the gravitational field of massive objects. Bending angles for electromagnetic waves and light in particular were measured with a high precision. However, the effect of gravity on relativistic massive particles was never studied experimentally. Here we propose and analyse experiments devoted to that purpose. We demonstrate a high sensitivity of the laser Compton scattering at high energy accelerators to the effects of gravity. The main observable -- maximal energy of the scattered photons -- would experience a significant shift in the Earth's gravitational field even for otherwise negligible violation of the equivalence principle. We confirm predictions of general relativity for ultrarelativistic electrons of energy of tens of GeV at a current level of res...

  10. Limitations of eddy current testing in a fast reactor environment

    Science.gov (United States)

    Wu, Tao; Bowler, John R.

    2016-02-01

    The feasibility of using eddy current probes for detecting flaws in fast nuclear reactor structures has been investigated with the aim of detecting defects immersed in electrically conductive coolant including under liquid sodium during standby. For the inspections to be viable, there is a need to use an encapsulated sensor system that can be move into position with the aid of visualization tools. The initial objective being to locate the surface to be investigated using, for example, a combination of electromagnetic sensors and sonar. Here we focus on one feature of the task in which eddy current probe impedance variations due to interaction with the external surface of a tube are evaluated in order to monitor the probe location and orientation during inspection.

  11. Facility Configuration Study of the High Temperature Gas-Cooled Reactor Component Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    S. L. Austad; L. E. Guillen; D. S. Ferguson; B. L. Blakely; D. M. Pace; D. Lopez; J. D. Zolynski; B. L. Cowley; V. J. Balls; E.A. Harvego, P.E.; C.W. McKnight, P.E.; R.S. Stewart; B.D. Christensen

    2008-04-01

    A test facility, referred to as the High Temperature Gas-Cooled Reactor Component Test Facility or CTF, will be sited at Idaho National Laboratory for the purposes of supporting development of high temperature gas thermal-hydraulic technologies (helium, helium-Nitrogen, CO2, etc.) as applied in heat transport and heat transfer applications in High Temperature Gas-Cooled Reactors. Such applications include, but are not limited to: primary coolant; secondary coolant; intermediate, secondary, and tertiary heat transfer; and demonstration of processes requiring high temperatures such as hydrogen production. The facility will initially support completion of the Next Generation Nuclear Plant. It will secondarily be open for use by the full range of suppliers, end-users, facilitators, government laboratories, and others in the domestic and international community supporting the development and application of High Temperature Gas-Cooled Reactor technology. This pre-conceptual facility configuration study, which forms the basis for a cost estimate to support CTF scoping and planning, accomplishes the following objectives: • Identifies pre-conceptual design requirements • Develops test loop equipment schematics and layout • Identifies space allocations for each of the facility functions, as required • Develops a pre-conceptual site layout including transportation, parking and support structures, and railway systems • Identifies pre-conceptual utility and support system needs • Establishes pre-conceptual electrical one-line drawings and schedule for development of power needs.

  12. Non destructive testing of irradiated fuel assemblies at the IEA-R1 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Jose Eduardo Rosa da; Terremoto, Luis Antonio Albiac; Castanheira, Myrthes; Teodoro, Celso Antonio; Silva, Antonio Teixeira e; Damy, Margaret de Almeida; Lucki, Georgi [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)]. E-mails: jersilva@ipen.br; laaterre@ipen.br; myrthes@ipen.br; cteodoro@ipen.br; teixeira@ipen.br; madamy@ipen.br; glucki@ipen.br

    2007-07-01

    Fuel performance and nuclear fuel qualification require a post-irradiation analysis. Non-destructive methods are utilised both in irradiated fuel storage pools and in hot-cells laboratories. As Brazil does not have hot-cells facilities for post-irradiation analysis, a qualification program for the Material Testing Reactor (MTR) fuel elements made at IPEN/CNEN-SP was adopted, based on non-destructive tests. The IPEN Fuel Engineering Group - CENC developed basic facilities for fuels post-irradiated analysis inside the reactor pool, which gives indications of: general state, by visual inspection; the integrity of the irradiated fuel cladding, by sipping tests; thickness measurements of the fuel miniplates during the irradiation time, for swelling evaluation; and, local burn-up evaluation by gamma spectrometry along the active area of the fuel element. This work describes that facilities, equipment and examples of some irradiated fuels analysis performed. (author)

  13. Generalization of Selection Test Validity.

    Science.gov (United States)

    Colbert, G. A.; Taylor, L. R.

    1978-01-01

    This is part three of a three-part series concerned with the empirical development of homogeneous families of insurance company jobs based on data from the Position Analysis Questionnaire (PAQ). This part involves validity generalizations within the job families which resulted from the previous research. (Editor/RK)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-06-30

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

  15. General Electric : Immelt surve all / Jena McGregor

    Index Scriptorium Estoniae

    McGregor, Jena

    2008-01-01

    Ettevõtte General Electric juht Jeffrey R. Immelt võitleb selle nimel, et taastada usaldus firma vastu, kuna laenukriis ja kehvad aktsiahinnad on tekitanud olukorra, kus investorid avaldavad suurt survet ettevõtte juhile. Lisa: GE kindralid

  16. General Electric tõrjus Microsofti liidrikohalt / Kaja Koovit

    Index Scriptorium Estoniae

    Koovit, Kaja, 1968-

    2001-01-01

    General Electric võttis Microsofti ees liidrikoha maailma suurima turuväärtusega ettevõtte edetabelis FT500. Tabelid: Maailma suurima turukapitalisatsiooniga ettevõtete TOP25, Skandinaavia firmade turuväärtuse TOP10

  17. General Electric tõrjus Microsofti liidrikohalt / Kaja Koovit

    Index Scriptorium Estoniae

    Koovit, Kaja, 1968-

    2001-01-01

    General Electric võttis Microsofti ees liidrikoha maailma suurima turuväärtusega ettevõtte edetabelis FT500. Tabelid: Maailma suurima turukapitalisatsiooniga ettevõtete TOP25, Skandinaavia firmade turuväärtuse TOP10

  18. General Electric : Immelt surve all / Jena McGregor

    Index Scriptorium Estoniae

    McGregor, Jena

    2008-01-01

    Ettevõtte General Electric juht Jeffrey R. Immelt võitleb selle nimel, et taastada usaldus firma vastu, kuna laenukriis ja kehvad aktsiahinnad on tekitanud olukorra, kus investorid avaldavad suurt survet ettevõtte juhile. Lisa: GE kindralid

  19. General Electric presents $325,000 to College of Engineering

    OpenAIRE

    Nystrom, Lynn A.

    2007-01-01

    The General Electric Company has presented a $300,000 check to Virginia Tech's College of Engineering to create three endowed fellowships in memory of its three slain engineering professors killed on April 16.

  20. General Electric presents $325,000 to College of Engineering

    OpenAIRE

    Nystrom, Lynn A.

    2007-01-01

    The General Electric Company has presented a $300,000 check to Virginia Tech's College of Engineering to create three endowed fellowships in memory of its three slain engineering professors killed on April 16.

  1. Present status and future perspectives of research and test reactor in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kaneko, Yoshihiko [Atomic Energy Research Laboratory, Musashi Institute of Technology, Kawasaki, Kanagawa (Japan); Kaieda, Keisuke [Department of Research Reactor, Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)

    2000-10-01

    Since 1957, Japan Atomic Energy Research Institute (JAERI) has constructed several research and test reactors to fulfill a major role in the study of nuclear energy and fundamental research. At present four reactors, the Japan Research Reactor No. 3 and No. 4 (JRR-3M and JRR-4 respectively), the Japan Materials Testing Reactor (JMTR) and the Nuclear Safety Research Reactor (NSRR) are in operation, and a new High Temperature Engineering Test Reactor (HTTR) has recently reached first criticality and now in the power up test. In 1966, the Kyoto University built the Kyoto University Reactor (KUR) and started its operation for joint use program of the Japanese universities. This paper introduces these reactors and describes their present operational status and also efforts for aging management. The recent tendency of utilization and future perspectives is also reported. (author)

  2. Reactor Testing and Qualification: Prioritized High-level Criticality Testing Needs

    Energy Technology Data Exchange (ETDEWEB)

    S. Bragg-Sitton; J. Bess; J. Werner; G. Harms; S. Bailey

    2011-09-01

    Researchers at the Idaho National Laboratory (INL) were tasked with reviewing possible criticality testing needs to support development of the fission surface power system reactor design. Reactor physics testing can provide significant information to aid in development of technologies associated with small, fast spectrum reactors that could be applied for non-terrestrial power systems, leading to eventual system qualification. Several studies have been conducted in recent years to assess the data and analyses required to design and build a space fission power system with high confidence that the system will perform as designed [Marcille, 2004a, 2004b; Weaver, 2007; Parry et al., 2008]. This report will provide a summary of previous critical tests and physics measurements that are potentially applicable to the current reactor design (both those that have been benchmarked and those not yet benchmarked), summarize recent studies of potential nuclear testing needs for space reactor development and their applicability to the current baseline fission surface power (FSP) system design, and provide an overview of a suite of tests (separate effects, sub-critical or critical) that could fill in the information database to improve the accuracy of physics modeling efforts as the FSP design is refined. Some recommendations for tasks that could be completed in the near term are also included. Specific recommendations on critical test configurations will be reserved until after the sensitivity analyses being conducted by Los Alamos National Laboratory (LANL) are completed (due August 2011).

  3. Local stability tests in Dresden 2 boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    March-Leuba, J.; Fry, D.N.; Buchanan, M.E.; McNew, C.O.

    1984-04-01

    This report presents the results of a local stability test performed at Dresden Unit 2 in May 1983 to determine the effect of a new fuel element design on local channel stability. This test was performed because the diameter of the new fuel rods increases the heat transfer coefficient, making the reactor more responsive and, thus, more susceptible to instabilities. After four of the new fuel elements with a 9 x 9 array of fuel rods were loaded into Dresden 2, the test was performed by inserting an adjacent control rod all the way in and then withdrawing it to its original position at maximum speed. At the moment of the test, reactor conditions were 52.7% power and 38.9% flow. Both the new 9 x 9 fuel elements and the standard 8 x 8 ones proved to be locally stable when operating at minimum pump speed at the beginning of cycle in Dresden 2, and no significant difference was found between the behavior of the two fuel types. Finally, Dresden 2 showed a high degree of stability during control rod and normal noise type perturbations.

  4. Electrical condutivity test in Jacaranda micacrantha seeds

    Directory of Open Access Journals (Sweden)

    Gabriela Fernanda Souza

    2016-03-01

    Full Text Available The objective of this study was to evaluate the physiological quality of Jacaranda micranta Cham. seeds, by electrical conductivity test. We used three lots of seeds, stored for different periods. Electrical conductivity tests were conducted at different amounts of seeds and deionised water for soaking, and then they were associated with germination and vigour data. It was observed that the longer the seeds were storage, the greater their electrical conductivity, and therefore the lower were their germination vigour. It was also observed that different quantities of seeds and soaking water did not present significant difference, so it is recommended to use 50 seeds in 75 mL of deionised water.

  5. Testing general relativity on accelerators

    Directory of Open Access Journals (Sweden)

    Tigran Kalaydzhyan

    2015-11-01

    Full Text Available Within the general theory of relativity, the curvature of spacetime is related to the energy and momentum of the present matter and radiation. One of the more specific predictions of general relativity is the deflection of light and particle trajectories in the gravitational field of massive objects. Bending angles for electromagnetic waves and light in particular were measured with a high precision. However, the effect of gravity on relativistic massive particles was never studied experimentally. Here we propose and analyze experiments devoted to that purpose. We demonstrate a high sensitivity of the laser Compton scattering at high energy accelerators to the effects of gravity. The main observable – maximal energy of the scattered photons – would experience a significant shift in the ambient gravitational field even for otherwise negligible violation of the equivalence principle. We confirm predictions of general relativity for ultrarelativistic electrons of energy of tens of GeV at a current level of resolution and expect our work to be a starting point of further high-precision studies on current and future accelerators, such as PETRA, European XFEL and ILC.

  6. Safety Analyses at the Idaho National Engineering and Environmental Laboratory Test Reactor Area - Past to Present

    Energy Technology Data Exchange (ETDEWEB)

    Ambrosek, Richard Garry; Ingram, Frederick William

    1999-11-01

    Test reactors are unique in that the core configuration may change with each operating interval. The process of safety analyses for test reactors at the Idaho National Engineering and Environmental Test Reactor Area has evolved as the computing capabilities, software, and regulatory requirements have changed. The evaluations for experiments and the reactor have moved from measurements in a set configuration and then application to other configurations with a relatively large error to modeling in three-dimensions and explicit analyses for each experiment and operating interval. This evolution is briefly discussed for the Test Reactor Area.

  7. Action Memorandum for the Engineering Test Reactor under the Idaho Cleanup Project

    Energy Technology Data Exchange (ETDEWEB)

    A. B. Culp

    2007-01-26

    This Action Memorandum documents the selected alternative for decommissioning of the Engineering Test Reactor at the Idaho National Laboratory under the Idaho Cleanup Project. Since the missions of the Engineering Test Reactor Complex have been completed, an engineering evaluation/cost analysis that evaluated alternatives to accomplish the decommissioning of the Engineering Test Reactor Complex was prepared adn released for public comment. The scope of this Action Memorandum is to encompass the final end state of the Complex and disposal of the Engineering Test Reactor vessol. The selected removal action includes removing and disposing of the vessel at the Idaho CERCLA Disposal Facility and demolishing the reactor building to ground surface.

  8. High temperature indentation tests on fusion reactor candidate materials

    Energy Technology Data Exchange (ETDEWEB)

    Montanari, R. [Dipartimento di Ingegneria Meccanica, Universita di Roma-Tor Vergata, Via del Politecnico 1, I-00133 Rome (Italy)]. E-mail: roberto.montanari@uniroma2.it; Filacchioni, G. [ENEA CR Casaccia, Via Anguillarese 301, I-00060 S.M. di Galeria, Rome (Italy); Iacovone, B. [Dipartimento di Ingegneria Meccanica, Universita di Roma-Tor Vergata, Via del Politecnico 1, I-00133 Rome (Italy); Plini, P. [Dipartimento di Ingegneria Meccanica, Universita di Roma-Tor Vergata, Via del Politecnico 1, I-00133 Rome (Italy); Riccardi, B. [Associazione EURATOM-ENEA sulla Fusione, P.O. Box 65, I-00044 Frascati, Rome (Italy)

    2007-08-01

    Flat-top cylinder indenter for mechanical characterization (FIMEC) is an indentation technique employing cylindrical punches with diameters ranging from 0.5 to 2 mm. The test gives pressure-penetration curves from which the yield stress can be determined. The FIMEC apparatus was developed to test materials in the temperature range from -180 to +200 {sup o}C. Recently, the heating system of FIMEC apparatus has been modified to operate up to 500 {sup o}C. So, in addition to providing yield stress over a more extended temperature range, it is possible to perform stress-relaxation tests at temperatures of great interest for several nuclear fusion reactor (NFR) alloys. Data on MANET-II, F82H mod., Eurofer-97, EM-10, AISI 316 L, Ti6Al4V and CuCrZr are presented and compared with those obtained by mechanical tests with standard methods.

  9. Data on test results of vessel cooling system of high temperature engineering test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Saikusa, Akio [Secretariat of Nuclear Safety Commission, Tokyo (Japan); Nakagawa, Shigeaki; Fujimoto, Nozomu; Tachibana, Yukio; Iyoku, Tatsuo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2003-02-01

    High Temperature Engineering Test Reactor (HTTR) is the first graphite-moderated helium gas cooled reactor in Japan. The rise-to-power test of the HTTR started on September 28, 1999 and thermal power of the HTTR reached its full power of 30 MW on December 7, 2001. Vessel Cooling System (VCS) of the HTTR is the first Reactor Cavity Cooling System (RCCS) applied for High Temperature Gas Cooled Reactors. The VCS cools the core indirectly through the reactor pressure vessel to keep core integrity during the loss of core flow accidents such as depressurization accident. Minimum heat removal of the VCS to satisfy its safety requirement is 0.3MW at 30 MW power operation. Through the performance test of the VCS in the rise-to-power test of the HTTR, it was confirmed that the VCS heat removal at 30 MW power operation was higher than 0.3 MW. This paper shows outline of the VCS and test results on the VCS performance. (author)

  10. Multi-reactor power system configurations for multimegawatt nuclear electric propulsion

    Science.gov (United States)

    George, Jeffrey A.

    1991-01-01

    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.

  11. The General Aptitude Test Battery.

    Science.gov (United States)

    Goguen, Lucille

    The development and use of the GATB in the United States is presented by a Canadian author. The history of establishing the norms for the GATB is also discussed. The use of the GATB as a counseling and selection tool is outlined while another section of the article points out the advantages and disadvantages of the test battery. There are also…

  12. 30 CFR 75.512 - Electric equipment; examination, testing and maintenance.

    Science.gov (United States)

    2010-07-01

    ...-General § 75.512 Electric equipment; examination, testing and maintenance. All electric equipment shall be... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Electric equipment; examination, testing and maintenance. 75.512 Section 75.512 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT...

  13. Probe for testing electrical properties of a test sample

    DEFF Research Database (Denmark)

    2012-01-01

    A probe for testing electrical properties of test samples includes a body having a probe arm defining proximal and distal ends, the probe arm extending from the body at the proximal end of the probe arm, whereby a first axis is defined by the proximal and the distal ends. The probe arm defines...

  14. Entrained Flow Reactor Test of Potassium Capture by Kaolin

    DEFF Research Database (Denmark)

    Wang, Guoliang; Jensen, Peter Arendt; Wu, Hao

    2015-01-01

    In the present study a method to simulate the reaction between gaseous KCl and kaolin at suspension fired condition was developed using a pilot-scale entrained flow reactor (EFR). Kaolin was injected into the EFR for primary test of this method. By adding kaolin, KCl can effectively be captured......, forming water-insoluble K-aluminosilicate. The amount of K captured by 1 g kaolin rose when increasing the molar ratio of K/Si in the reactant. Changing of reaction temperature from 1100 °C to 1300 °C did not influence the extent of reaction, which is different from the results observed in previous fixed...

  15. Reduced enrichment for research and test reactors: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1993-08-01

    November 9--10, 1978, marked the first of what has become an annual event--the International Meeting on Reduced Enrichment for Research and Test Reactors (RERTR). The meeting brought together for the first time many people who became major program participants in later years. This first meeting emphasized fuel development, and it established the basis for all later meetings. Believing that the proceedings of this first meeting are important as a historical record of the beginning of the international RERTR effort. This report provides presentations and discussions of this original meeting. Individual papers have been cataloged separately.

  16. EPR/PTFE dosimetry for test reactor environments

    Energy Technology Data Exchange (ETDEWEB)

    Vehar, D.W.; Griffin, P.J.; Quirk, T.J. [Sandia National Laboratories, Albuquerque, NM 87185-1146 (United States)

    2011-07-01

    The use of Electron Paramagnetic Resonance (EPR) spectroscopy with materials such as alanine is well established as a technique for measurement of ionizing radiation absorbed dose in photon and electron fields such as Co-60, high-energy bremsstrahlung and electron-beam fields [1]. In fact, EPR/Alanine dosimetry has become a routine transfer standard for national standards bodies such as NIST and NPL. In 1992 the Radiation Metrology Laboratory (RML) at Sandia National Laboratories implemented EPR/Alanine capabilities for use in routine and calibration activities at its Co-60 and pulsed-power facilities. At that time it also investigated the usefulness of the system for measurement of absorbed dose in the mixed neutron/photon environments of reactors such as the Sandia Pulsed Reactor and the Annular Core Research Reactor used for hardness testing of electronics. The RML concluded that the neutron response of alanine was a sufficiently high fraction of the overall dosimeter response that the resulting uncertainties in the photon dose would be unacceptably large for silicon-device testing. However, it also suggested that non-hydrogenous materials such as polytetrafluoroethylene (PTFE) would exhibit smaller neutron response and might be useful in mixed environments. Preliminary research with PTFE in photon environments indicated considerable promise, but further development was not pursued at that time. Because of renewed interest in absorbed dose measurements that could better define the individual contributions of photon and neutron components to the overall dose delivered to a test object, the RML has re-initiated the development of an EPR/PTFE dosimetry system. This effort consists of three stages: 1) Identification of PTFE materials that may be suitable for dosimetry applications. It was speculated that the inconsistency of EPR signatures in the earlier samples may have been due to variability in PTFE manufacturing processes. 2) Characterization of dosimetry in

  17. LIGHT WATER REACTOR ACCIDENT TOLERANT FUELS IRRADIATION TESTING

    Energy Technology Data Exchange (ETDEWEB)

    Carmack, William Jonathan [Idaho National Laboratory; Barrett, Kristine Eloise [Idaho National Laboratory; Chichester, Heather Jean MacLean [Idaho National Laboratory

    2015-09-01

    The purpose of Accident Tolerant Fuels (ATF) experiments is to test novel fuel and cladding concepts designed to replace the current zirconium alloy uranium dioxide (UO2) fuel system. The objective of this Research and Development (R&D) is to develop novel ATF concepts that will be able to withstand loss of active cooling in the reactor core for a considerably longer time period than the current fuel system while maintaining or improving the fuel performance during normal operations, operational transients, design basis, and beyond design basis events. It was necessary to design, analyze, and fabricate drop-in capsules to meet the requirements for testing under prototypic LWR temperatures in Idaho National Laboratory's Advanced Test Reactor (ATR). Three industry led teams and one DOE team from Oak Ridge National Laboratory provided fuel rodlet samples for their new concepts for ATR insertion in 2015. As-built projected temperature calculations were performed on the ATF capsules using the BISON fuel performance code. BISON is an application of INL’s Multi-physics Object Oriented Simulation Environment (MOOSE), which is a massively parallel finite element based framework used to solve systems of fully coupled nonlinear partial differential equations. Both 2D and 3D models were set up to examine cladding and fuel performance.

  18. 14 CFR 29.1363 - Electrical system tests.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Electrical system tests. 29.1363 Section 29... AIRWORTHINESS STANDARDS: TRANSPORT CATEGORY ROTORCRAFT Equipment Electrical Systems and Equipment § 29.1363 Electrical system tests. (a) When laboratory tests of the electrical system are conducted— (1) The tests...

  19. 14 CFR 25.1363 - Electrical system tests.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Electrical system tests. 25.1363 Section 25... AIRWORTHINESS STANDARDS: TRANSPORT CATEGORY AIRPLANES Equipment Electrical Systems and Equipment § 25.1363 Electrical system tests. (a) When laboratory tests of the electrical system are conducted— (1) The tests...

  20. Meso-scale modeling of irradiated concrete in test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Giorla, A. [Oak Ridge National Laboratory, One Bethel Valley Road, Oak Ridge, TN 37831 (United States); Vaitová, M. [Czech Technical University, Thakurova 7, 166 29 Praha 6 (Czech Republic); Le Pape, Y., E-mail: lepapeym@ornl.gov [Oak Ridge National Laboratory, One Bethel Valley Road, Oak Ridge, TN 37831 (United States); Štemberk, P. [Czech Technical University, Thakurova 7, 166 29 Praha 6 (Czech Republic)

    2015-12-15

    Highlights: • A meso-scale finite element model for irradiated concrete is developed. • Neutron radiation-induced volumetric expansion is a predominant degradation mode. • Confrontation with expansion and damage obtained from experiments is successful. • Effects of paste shrinkage, creep and ductility are discussed. - Abstract: A numerical model accounting for the effects of neutron irradiation on concrete at the mesoscale is detailed in this paper. Irradiation experiments in test reactor (Elleuch et al., 1972), i.e., in accelerated conditions, are simulated. Concrete is considered as a two-phase material made of elastic inclusions (aggregate) subjected to thermal and irradiation-induced swelling and embedded in a cementitious matrix subjected to shrinkage and thermal expansion. The role of the hardened cement paste in the post-peak regime (brittle-ductile transition with decreasing loading rate), and creep effects are investigated. Radiation-induced volumetric expansion (RIVE) of the aggregate cause the development and propagation of damage around the aggregate which further develops in bridging cracks across the hardened cement paste between the individual aggregate particles. The development of damage is aggravated when shrinkage occurs simultaneously with RIVE during the irradiation experiment. The post-irradiation expansion derived from the simulation is well correlated with the experimental data and, the obtained damage levels are fully consistent with previous estimations based on a micromechanical interpretation of the experimental post-irradiation elastic properties (Le Pape et al., 2015). The proposed modeling opens new perspectives for the interpretation of test reactor experiments in regards to the actual operation of light water reactors.

  1. 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

    2002-01-01

    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.

  2. Space Molten Salt Reactor Concept for Nuclear Electric Propulsion and Surface Power

    Science.gov (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.

  3. Generalized sensitivity evaluation of electrical power systems

    Energy Technology Data Exchange (ETDEWEB)

    Grewal, H.K.

    1987-01-01

    The material presented in this thesis is a logical extension of and addition to previous work on network sensitivities as applied to power-system analysis and planning. The continuing tendency of supplementing the existing extra-high voltage a.c. transmission systems with high-voltage d.c. (HVDC) lines was taken into consideration, and various relevant component models were investigated using a new hybrid network formulation based on the methodology developed by Bandler and El-Kady. The load buses, frequently modeled as PQ buses at which both the real and reactive injected powers are known, and the generator buses characterized by a constant voltage magnitude and constant real injected power, have been dealt with by exploiting a special complex conjugate notation. In addition, the current, voltage and/or power relationships associated with the transmission network branches were investigated. A hybrid formulation for generalized power system component models were developed. This novel formulation not only encompasses the work established on the basis of one-port theory, but it is also capable of manipulating multiport, nonreciprocal, a.c. as well as integrated a.c.-d.c. bulk transmission networks.

  4. 10 CFR 110.26 - General license for the export of nuclear reactor components.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false General license for the export of nuclear reactor components. 110.26 Section 110.26 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) EXPORT AND IMPORT OF NUCLEAR EQUIPMENT AND MATERIAL Licenses § 110.26 General license for the export of nuclear...

  5. Hydraulic Shuttle Irradiation System (HSIS) Recently Installed in the Advanced Test Reactor (ATR)

    Energy Technology Data Exchange (ETDEWEB)

    A. Joseph Palmer; Gerry L. McCormick; Shannon J. Corrigan

    2010-06-01

    2010 International Congress on Advances in Nuclear Power Plants (ICAPP’10) ANS Annual Meeting Imbedded Topical San Diego, CA June 13 – 17, 2010 Hydraulic Shuttle Irradiation System (HSIS) Recently Installed in the Advanced Test Reactor (ATR) Author: A. Joseph Palmer, Mechanical Engineer, Irradiation Test Programs, 208-526-8700, Joe.Palmer@INL.gov Affiliation: Idaho National Laboratory P.O. Box 1625, MS-3840 Idaho Falls, ID 83415 INL/CON-10-17680 ABSTRACT Most test reactors are equipped with shuttle facilities (sometimes called rabbit tubes) whereby small capsules can be inserted into the reactor and retrieved during power operations. With the installation of Hydraulic Shuttle Irradiation System (HSIS) this capability has been restored to the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). The general design and operating principles of this system were patterned after the hydraulic rabbit at Oak Ridge National Laboratory’s (ORNL) High Flux Isotope Reactor (HFIR), which has operated successfully for many years. Using primary coolant as the motive medium the HSIS system is designed to simultaneously transport fourteen shuttle capsules, each 16 mm OD x 57 mm long, to and from the B-7 position of the reactor. The B-7 position is one of the higher flux positions in the reactor with typical thermal and fast (>1 Mev) fluxes of 2.8E+14 n/cm2/sec and 1.9E+14 n/cm2/sec respectively. The available space inside each shuttle is approximately 14 mm diameter x 50 mm long. The shuttle containers are made from titanium which was selected for its low neutron activation properties and durability. Shuttles can be irradiated for time periods ranging from a few minutes to several months. The Send and Receive Station (SRS) for the HSIS is located 2.5 m deep in the ATR canal which allows irradiated shuttles to be easily moved from the SRS to a wet loaded cask, or transport pig. The HSIS system first irradiated (empty) shuttles in September 2009 and has since completed

  6. Topical report: Natural convection shutdown heat removal test facility (NSTF) evaluation for generating additional reactor cavity cooling system (RCCS) data.

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, M. T.; Kilsdonk, D. J.; Tzanos, C.P.; Lomperski, S.; Aeschlimann, R.W.; Pointer, D.; Nuclear Engineering Division

    2005-09-01

    As part of the Department of Energy (DOE) Generation IV roadmapping activity, the Very High Temperature gas cooled Reactor (VHTR) has been selected as the principal concept for hydrogen production and other process-heat applications such as district heating and potable water production. On this basis, the DOE has selected the VHTR for additional R&D with the ultimate goal of demonstrating emission-free electricity and hydrogen production with this advanced reactor concept. One of the key passive safety features of the VHTR is the potential for decay heat removal by natural circulation of air in a Reactor Cavity Cooling System (RCCS). The air-cooled RCCS concept is notably similar to the Reactor Vessel Auxiliary Cooling System (RVACS) that was developed for the General Electric PRISM sodium-cooled fast reactor. As part of the DOE R&D program that supported the development of this fast reactor concept, the Natural Convection Shutdown Heat Removal Test Facility (NSTF) was developed at ANL to provide proof-of-concept data for the RVACS under prototypic natural convection flow, temperature, and heat flux conditions. Due to the similarity between RVACS and the RCCS, current VHTR R&D plans call for the utilization of the NSTF to provide RCCS model development and validation data, in addition to supporting design validation and optimization activities. Both air-cooled and water-cooled RCCS designs are to be included. In support of this effort, ANL has been tasked with the development of an engineering plan for mechanical and instrumentation modifications to NSTF to ensure that sufficiently detailed temperature, heat flux, velocity and turbulence profiles are obtained to adequately qualify the codes under the expected range of air-cooled RCCS flow conditions. Next year, similar work will be carried out for the alternative option of a water-cooled RCCS design. Analysis activities carried out in support of this experiment planning task have shown that: (a) in the RCCS, strong

  7. Study for improvement of performance of the test and research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sakurai, Fumio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-03-01

    Current utilization needs for the test and research reactors become more advanced and diversified along with the advance of nuclear science and technology. Besides, the requested safety for the research and test reactors grows strictly every year as well as a case of the power reactors. Under this circumstance, every effort to improve reactor performance including its safety is necessary to be sustained for allowing more effective utilization of the test and research reactors as experimental apparatus for advanced researches. In this study, the following three themes i.e., JMTR high-performance fuel element, evaluation method of fast neutron irradiation dose in the JMTR, evaluation method of performance of siphon break valve as core covering system for water-cooled test and research reactors, were investigated respectively from the views of improvement of core performance as a neutron source, utilization performance as an experimental apparatus, and safety as a reactor plant. (author)

  8. Irradiation Tests Supporting LEU Conversion of Very High Power Research Reactors in the US

    Energy Technology Data Exchange (ETDEWEB)

    Woolstenhulme, N. E.; Cole, J. I.; Glagolenko, I.; Holdaway, K. K.; Housley, G. K.; Rabin, B. H.

    2016-10-01

    The US fuel development team is developing a high density uranium-molybdenum alloy monolithic fuel to enable conversion of five high-power research reactors. Previous irradiation tests have demonstrated promising behavior for this fuel design. A series of future irradiation tests will enable selection of final fuel fabrication process and provide data to qualify the fuel at moderately-high power conditions for use in three of these five reactors. The remaining two reactors, namely the Advanced Test Reactor and High Flux Isotope Reactor, require additional irradiation tests to develop and demonstrate the fuel’s performance with even higher power conditions, complex design features, and other unique conditions. This paper reviews the program’s current irradiation testing plans for these moderately-high irradiation conditions and presents conceptual testing strategies to illustrate how subsequent irradiation tests will build upon this initial data package to enable conversion of these two very-high power research reactors.

  9. Method of boundary testing of the electric circuits and its application for calculating electric tolerances. [electric equipment tests

    Science.gov (United States)

    Redkina, N. P.

    1974-01-01

    Boundary testing of electric circuits includes preliminary and limiting tests. Preliminary tests permit determination of the critical parameters causing the greatest deviation of the output parameter of the system. The boundary tests offer the possibility of determining the limits of the fitness of the system with simultaneous variation of its critical parameters.

  10. VISTA : thermal-hydraulic integral test facility for SMART reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choi, K. Y.; Park, H. S.; Cho, S.; Park, C. K.; Lee, S. J.; Song, C. H.; Chung, M. K. [KAERI, Taejon (Korea, Republic of)

    2003-07-01

    Preliminary performance tests were carried out using the thermal-hydraulic integral test facility, VISTA (Experimental Verification by Integral Simulation of Transients and Accidents), which has been constructed to simulate the SMART-P. The VISTA facility is an integral test facility including the primary and secondary systems as well as safety-related Passive Residual Heat Removal (PRHR) systems. Its scaled ratio with respect to the SMART-P is 1/1 in height and 1/96 in volume and heater power. Several steady states and power changing tests have been carried out to verify the overall thermal hydraulic primary and secondary characteristics in the range of 10% to 100% power operation. As for the preliminary results, the steady state conditions were found to coincide with the expected design values of the SMART-P. But the major thermal hydraulic parameters are greatly affected by the initial water level and the nitrogen pressure in the reactor's upper annular cavity. The power step/ramp changing tests are successfully carried out and the system responses are observed. The primary natural circulation operation is achieved, but advanced control logics need to be developed to reach the natural circulation mode without pressure excursion. In the PRHR transient tests, the natural circulation flow rate through the PRHR system was found to be about 10 percent in the early phases of PRHR operation.

  11. 77 FR 26607 - Energy Conservation Program: Test Procedures for Electric Motors and Small Electric Motors

    Science.gov (United States)

    2012-05-04

    ... II.2--Summary of SNOPR Commenters Company Abbreviation Interested party type Baldor Electric Co Baldor Manufacturer. WEG Electric WEG Manufacturer. Advanced Energy Advanced Energy Independent Test...

  12. BENCHMARK EVALUATION OF THE START-UP CORE REACTOR PHYSICS MEASUREMENTS OF THE HIGH TEMPERATURE ENGINEERING TEST REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    John Darrell Bess

    2010-05-01

    The benchmark evaluation of the start-up core reactor physics measurements performed with Japan’s High Temperature Engineering Test Reactor, in support of the Next Generation Nuclear Plant Project and Very High Temperature Reactor Program activities at the Idaho National Laboratory, has been completed. The evaluation was performed using MCNP5 with ENDF/B-VII.0 nuclear data libraries and according to guidelines provided for inclusion in the International Reactor Physics Experiment Evaluation Project Handbook. Results provided include updated evaluation of the initial six critical core configurations (five annular and one fully-loaded). The calculated keff eigenvalues agree within 1s of the benchmark values. Reactor physics measurements that were evaluated include reactivity effects measurements such as excess reactivity during the core loading process and shutdown margins for the fully-loaded core, four isothermal temperature reactivity coefficient measurements for the fully-loaded core, and axial reaction rate measurements in the instrumentation columns of three core configurations. The calculated values agree well with the benchmark experiment measurements. Fully subcritical and warm critical configurations of the fully-loaded core were also assessed. The calculated keff eigenvalues for these two configurations also agree within 1s of the benchmark values. The reactor physics measurement data can be used in the validation and design development of future High Temperature Gas-cooled Reactor systems.

  13. Four-terminal electrical testing device

    Science.gov (United States)

    Robinson, Robert L.; Graves, Thomas J.; Hoffman, William C., III

    1987-04-01

    The invention relates to a four-terminal electrical connector device for testing and measuring unknown resistances of initiators used for starting pyrotechnic events aboard the space shuttle. The testing device minimizes contact resistance degradation effects and so improves the reliability of resistance measurements taken with the device. Separate and independent voltage sensing and current supply circuits each include a pair of socket contacts for mating engagement with the pins of the initiator. The unknown resistance that is measured by the device is the resistance of the bridgewire of the initiator which is required to be between 0.95 and 1.15 ohms.

  14. Efficiency Test Method for Electric Vehicle Chargers

    DEFF Research Database (Denmark)

    Kieldsen, Andreas; Thingvad, Andreas; Martinenas, Sergejus

    2016-01-01

    This paper investigates different methods for measuring the charger efficiency of mass produced electric vehicles (EVs), in order to compare the different models. The consumers have low attention to the loss in the charger though the impact on the driving cost is high. It is not a high priority...... different vehicles. A unified method for testing the efficiency of the charger in EVs, without direct access to the component, is presented. The method is validated through extensive tests of the models Renault Zoe, Nissan LEAF and Peugeot iOn. The results show a loss between 15 % and 40 %, which is far...

  15. Standard Test Method for Application and Analysis of Helium Accumulation Fluence Monitors for Reactor Vessel Surveillance, E706 (IIIC)

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2007-01-01

    1.1 This test method describes the concept and use of helium accumulation for neutron fluence dosimetry for reactor vessel surveillance. Although this test method is directed toward applications in vessel surveillance, the concepts and techniques are equally applicable to the general field of neutron dosimetry. The various applications of this test method for reactor vessel surveillance are as follows: 1.1.1 Helium accumulation fluence monitor (HAFM) capsules, 1.1.2 Unencapsulated, or cadmium or gadolinium covered, radiometric monitors (RM) and HAFM wires for helium analysis, 1.1.3 Charpy test block samples for helium accumulation, and 1.1.4 Reactor vessel (RV) wall samples for helium accumulation. This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-02-07

    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.

  17. Development and verification test of integral reactor major components

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. I.; Kim, Y. W.; Kim, J. H. and others

    1999-03-01

    The conceptual designs for SG, MCP, CEDM to be installed in the integral reactor SMART were developed. Three-dimensional CAD models for the major components were developed to visualize the design concepts. Once-through helical steam generator was conceptually designed for SMART. Canned motor pump was adopted in the conceptual design of MCP. Linear pulse motor type and ballscrew type CEDM, which have fine control capabilities were studied for adoption in SMART. In parallel with the structural design, the electro-magnetic design was performed for the sizing motors and electro-magnet. Prototypes for the CEDM and MCP sub-assemblies were developed and tested to verify the performance. The impeller design procedure and the computer program to analyze the dynamic characteristics of MCP rotor shaft were developed. The design concepts of SG, MCP, CEDM were also invetigated for the fabricability.

  18. DOE Hybrid and Electric Vehicle Test Platform

    Energy Technology Data Exchange (ETDEWEB)

    Gao, Yimin [Advanced Vehicle Research Center, Danville, VA (United States)

    2012-03-31

    Based on the contract NT-42790 to the Department of Energy, “Plug-in Hybrid Ethanol Research Platform”, Advanced Vehicle Research Center (AVRC) Virginia has successfully developed the phase I electric drive train research platform which has been named as Laboratory Rapid Application Testbed (LabRAT). In phase II, LabRAT is to be upgraded into plug-in hybrid research platform, which will be capable of testing power systems for electric vehicles, and plug-in hybrid electric vehicles running on conventional as well as alternative fuels. LabRAT is configured as a rolling testbed with plentiful space for installing various component configurations. Component connections are modularized for flexibility and are easily replaced for testing various mechanisms. LabRAT is designed and built as a full functional vehicle chassis with a steering system, brake system and four wheel suspension. The rear drive axle offers maximum flexibility with a quickly changeable gear ratio final drive to accommodate different motor speed requirements. The electric drive system includes an electric motor which is mechanically connected to the rear axle through an integrated speed/torque sensor. Initially, a 100 kW UQM motor and corresponding UQM motor controller is used which can be easily replaced with another motor/controller combination. A lithium iron phosphate (LiFePO4) battery pack is installed, which consists of 108 cells of 100 AH capacity, giving the total energy capacity of 32.5 kWh. Correspondingly, a fully functional battery management system (BMS) is installed to perform battery cell operation monitoring, cell voltage balancing, and reporting battery real time operating parameters to vehicle controller. An advanced vehicle controller ECU is installed for controlling the drive train. The vehicle controller ECU receives traction or braking torque command from driver through accelerator and brake pedal position sensors and battery operating signals from the BMS through CAN BUS

  19. Use of small reactors as an alternative to supply electricity to Baja California Sur; Uso de reactores pequenos como alternativa de suministro de electricidad para Baja California Sur

    Energy Technology Data Exchange (ETDEWEB)

    Alonso, G.; Portes, E.; Ramirez, J. R. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Ortega, G., E-mail: gustavo.alonso@inin.gob.mx [Comision Federal de Electricidad, Rio Rodano No. 14, 06500 Ciudad de Mexico (Mexico)

    2016-09-15

    The state of Baja California Sur (Mexico) does not form part of the national interconnected electrical system of the country, reason why is local its electrical power supply; one of the alternatives to cover future demands is the use of gas-based combined cycles, which presents the additional problem of including a high price for gas transportation in its costs. In order to reduce total costs, including investment, fuels and operation and maintenance in the operation of the Baja California Sur state electricity system in the coming years, mainly due to the estimated natural gas cost order of $11.50 dollars per million BTU, a proposal is presented to reduce the costs of the electrical system by replacing the necessary combined cycles with the new Small Modular Reactor type nuclear reactors, this alternative is economically competitive. (Author)

  20. Recent results on the RIA test in IGR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Asmolov, V.; Yegorova, L. [Nuclear Safety Institute, Moscow (Russian Federation)

    1997-01-01

    At the 23d WRSM meeting the data base characterizing results of VVER high burnup fuel rods tests under reactivity-initiated accident (RIA) conditions was presented. Comparison of PWR and VVER failure thresholds was given also. Additional analysis of the obtained results was being carried out during 1996. The results of analysis show that the two different failure mechanisms were observed for PWR and VVER fuel rods. Some factors which can be as the possible reasons of these differences are presented. First of them is the state of preirradiated cladding. Published test data for PWR high burnup fuel rods demonstrated that the PWR high burnup fuel rods failed at the RIA test are characterized by very high level of oxidation and hydriding for the claddings. Corresponding researches were performed at Institute of Atomic Reactors (RLAR, Dimitrovgrad, Russia) for large set of VVER high burnup fuel rods. Results of these investigations show that preirradiated commercial Zr-1%Nb claddings practically keep their initial levels of oxidation and H{sub 2} concentration. Consequently the VVER preirradiated cladding must keep the high level of mechanical properties. The second reason leading to differences between failure mechanisms for two types of high burnup fuel rods can be the test conditions. Now such kind of analysis have been performed by two methods.

  1. Electrically Driven General Systems for UAV’s

    Science.gov (United States)

    2007-11-01

    systems are discussed in this paper. First the Barracuda M-05 UAV Demonstrator and second the 270 VDC More Electric Aircraft project launched by the...German Federal Office of Defense Technology and Procurement (BWB). 3.1 Barracuda M-05 The Barracuda M-05 is a company founded unmanned...test flight. Figure 4 showed a picture of the first flight. Figure 4: Barracuda M-05 First Flight With respect to electrically driven

  2. 9 CFR 113.5 - General testing.

    Science.gov (United States)

    2010-01-01

    ..., and the test may be repeated. (e) When new test methods are developed and approved by Animal and Plant... 9 Animals and Animal Products 1 2010-01-01 2010-01-01 false General testing. 113.5 Section 113.5 Animals and Animal Products ANIMAL AND PLANT HEALTH INSPECTION SERVICE, DEPARTMENT OF AGRICULTURE VIRUSES...

  3. NOTICE OF ELECTRICAL CUT - TEST OF THE SECURED NETWORK

    CERN Document Server

    Electrical Service ST/EL

    2001-01-01

    The electrical service ST/EL will test the switching sequence between the secured network and the diesel generators on January 8, 2002. The normal network, general services of the sites Meyrin, Prevessin, SPS, Zone Nord, LHC1 and LHC18 will be cut between 6:00am and 6:10am. The secured network will be resupplied by the diesel generators after approximately 1 minute. The UPS network will not be affected. To facilitate the restart of the electrical network and to minimize the impact of the tests on critical equipment, we would like to ask you to stop any equipment that might suffer major inconveniences during the tests (e.g. computers). For any further information, please do not hesitate to contact the Technical Control Room TCR (72201) or G. Cumer (160592).

  4. Deterministic Modeling of the High Temperature Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ortensi, J.; Cogliati, J. J.; Pope, M. A.; Ferrer, R. M.; Ougouag, A. M.

    2010-06-01

    Idaho National Laboratory (INL) is tasked with the development of reactor physics analysis capability of the Next Generation Nuclear Power (NGNP) project. In order to examine INL’s current prismatic reactor deterministic analysis tools, the project is conducting a benchmark exercise based on modeling the High Temperature Test Reactor (HTTR). This exercise entails the development of a model for the initial criticality, a 19 column thin annular core, and the fully loaded core critical condition with 30 columns. Special emphasis is devoted to the annular core modeling, which shares more characteristics with the NGNP base design. The DRAGON code is used in this study because it offers significant ease and versatility in modeling prismatic designs. Despite some geometric limitations, the code performs quite well compared to other lattice physics codes. DRAGON can generate transport solutions via collision probability (CP), method of characteristics (MOC), and discrete ordinates (Sn). A fine group cross section library based on the SHEM 281 energy structure is used in the DRAGON calculations. HEXPEDITE is the hexagonal z full core solver used in this study and is based on the Green’s Function solution of the transverse integrated equations. In addition, two Monte Carlo (MC) based codes, MCNP5 and PSG2/SERPENT, provide benchmarking capability for the DRAGON and the nodal diffusion solver codes. The results from this study show a consistent bias of 2–3% for the core multiplication factor. This systematic error has also been observed in other HTTR benchmark efforts and is well documented in the literature. The ENDF/B VII graphite and U235 cross sections appear to be the main source of the error. The isothermal temperature coefficients calculated with the fully loaded core configuration agree well with other benchmark participants but are 40% higher than the experimental values. This discrepancy with the measurement stems from the fact that during the experiments the

  5. 78 FR 72552 - Airworthiness Directives; General Electric Company Turbofan Engines

    Science.gov (United States)

    2013-12-03

    ...: We are adopting a new airworthiness directive (AD) for certain General Electric Company model GEnx... consumed more cyclic life than they would have in revenue flight cycles. These parts were then installed into engines and introduced into revenue service without adjustment to remaining cyclic life. This...

  6. Miniaturized Charpy test for reactor pressure vessel embrittlement characterization

    Energy Technology Data Exchange (ETDEWEB)

    Manahan, M.P. Sr. [MPM Research and Consulting, Lemont, PA (United States)

    1999-10-01

    Modifications were made to a conventional Charpy machine to accommodate the miniaturized Charpy V-Notch (MCVN) specimens which were fabricated from an archived reactor pressure vessel (RPV) steel. Over 100 dynamic MCVN tests were performed and compared to the results from conventional Charpy V-Notch (CVN) tests to demonstrate the efficacy of the miniature specimen test. The optimized sidegrooved MCVN specimens exhibit transitional fracture behavior over essentially the same temperature range as the CVN specimens which indicates that the stress fields in the MCVN specimens reasonably simulate those of the CVN specimens and this fact has been observed in finite element calculations. This result demonstrates a significant breakthrough since it is now possible to measure the ductile-brittle transition temperature (DBTT) using miniature specimens with only small correction factors, and for some materials as in the present study, without the need for any correction factor at all. This development simplifies data interpretation and will facilitate future regulatory acceptance. The non-sidegrooved specimens yield energy-temperature data which is significantly shifted downward in temperature (non-conservative) as a result of the loss of constraint which accompanies size reduction.

  7. Testing of an advanced thermochemical conversion reactor system

    Science.gov (United States)

    1990-01-01

    This report presents the results of work conducted by MTCI to verify and confirm experimentally the ability of the MTCI gasification process to effectively generate a high-quality, medium-Btu gas from a wider variety of feedstock and waste than that attainable in air-blown, direct gasification systems. The system's overall simplicity, due to the compact nature of the pulse combustor, and the high heat transfer rates attainable within the pulsating flow resonance tubes, provide a decided and near-term potential economic advantage for the MTCI indirect gasification system. The primary objective was the design, construction, and testing of a Process Design Verification System for an indirectly heated, thermochemical fluid-bed reactor and a pulse combustor an an integrated system that can process alternative renewable sources of energy such as biomass, black liquor, municipal solid waste and waste hydrocarbons, including heavy oils into a useful product gas. The test objectives for the biomass portion of this program were to establish definitive performance data on biomass feedstocks covering a wide range of feedstock qualities and characteristics. The test objectives for the black liquor portion of this program were to verify the operation of the indirect gasifier on commercial black liquor containing 65 percent solids at several temperature levels and to characterize the bed carbon content, bed solids particle size and sulfur distribution as a function of gasification conditions.

  8. Testing of an advanced thermochemical conversion reactor system

    Energy Technology Data Exchange (ETDEWEB)

    1990-01-01

    This report presents the results of work conducted by MTCI to verify and confirm experimentally the ability of the MTCI gasification process to effectively generate a high-quality, medium-Btu gas from a wider variety of feedstock and waste than that attainable in air-blown, direct gasification systems. The system's overall simplicity, due to the compact nature of the pulse combustor, and the high heat transfer rates attainable within the pulsating flow resonance tubes, provide a decided and near-term potential economic advantage for the MTCI indirect gasification system. The primary objective of this project was the design, construction, and testing of a Process Design Verification System for an indirectly heated, thermochemical fluid-bed reactor and a pulse combustor an an integrated system that can process alternative renewable sources of energy such as biomass, black liquor, municipal solid waste and waste hydrocarbons, including heavy oils into a useful product gas. The test objectives for the biomass portion of this program were to establish definitive performance data on biomass feedstocks covering a wide range of feedstock qualities and characteristics. The test objectives for the black liquor portion of this program were to verify the operation of the indirect gasifier on commercial black liquor containing 65 percent solids at several temperature levels and to characterize the bed carbon content, bed solids particle size and sulfur distribution as a function of gasification conditions. 6 refs., 59 figs., 29 tabs.

  9. High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nelson, Lee Orville [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kinsey, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-03-01

    A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

  10. High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nelson, Lee Orville [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-01-01

    A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

  11. Propagation of negative electrical discharges through 2-dimensional packed bed reactors

    Science.gov (United States)

    Kruszelnicki, Juliusz; Engeling, Kenneth W.; Foster, John E.; Xiong, Zhongmin; Kushner, Mark J.

    2017-01-01

    Plasma-based pollutant remediation and value-added gas production have recently gained increased attention as possible alternatives to the currently-deployed chemical reactor systems. Electrical discharges in packed bed reactors (PBRs) are of interest, due to their ability to synergistically combine catalytic and plasma chemical processes. In principle, these systems could be tuned to produce specific products, based on their application by combinations of power formats, materials, geometries and working gases. Negative voltage, atmospheric-pressure plasma discharges sustained in humid air in a PBR-like geometry were experimentally characterized using ICCD imaging and simulated in 2-dimensions (2D) to provide insights into possible routes to this tunability. Surface ionization waves (SIWs) and positive restrikes through the lattice of dielectric rods were shown to be the principal means of producing reactive species. The number and intensity of SIWs and restrikes are sensitive functions of the alignment of the lattice of dielectric beads (or rods in 2D) with respect to the applied electric field. Decreased spacing between the dielectric elements leads to an increased electric field enhancement in the gas, and therefore locally higher plasma densities, but does not necessarily impact the types of discharges that occur through the lattice.

  12. THE COMPONENT TEST FACILITY – A NATIONAL USER FACILITY FOR TESTING OF HIGH TEMPERATURE GAS-COOLED REACTOR (HTGR) COMPONENTS AND SYSTEMS

    Energy Technology Data Exchange (ETDEWEB)

    David S. Duncan; Vondell J. Balls; Stephanie L. Austad

    2008-09-01

    The Next Generation Nuclear Plant (NGNP) and other High-Temperature Gas-cooled Reactor (HTGR) Projects require research, development, design, construction, and operation of a nuclear plant intended for both high-efficiency electricity production and high-temperature industrial applications, including hydrogen production. During the life cycle stages of an HTGR, plant systems, structures and components (SSCs) will be developed to support this reactor technology. To mitigate technical, schedule, and project risk associated with development of these SSCs, a large-scale test facility is required to support design verification and qualification prior to operational implementation. As a full-scale helium test facility, the Component Test facility (CTF) will provide prototype testing and qualification of heat transfer system components (e.g., Intermediate Heat Exchanger, valves, hot gas ducts), reactor internals, and hydrogen generation processing. It will perform confirmation tests for large-scale effects, validate component performance requirements, perform transient effects tests, and provide production demonstration of hydrogen and other high-temperature applications. Sponsored wholly or in part by the U.S. Department of Energy, the CTF will support NGNP and will also act as a National User Facility to support worldwide development of High-Temperature Gas-cooled Reactor technologies.

  13. Design of an Actinide-Burning, Lead or Lead-Bismuth Cooled Reactor that Produces Low-Cost Electricity

    Energy Technology Data Exchange (ETDEWEB)

    Mac Donald, Philip Elsworth; Weaver, Kevan Dean; Davis, Cliff Bybee; MIT folks

    2000-07-01

    The purpose of this Idaho National Engineering and Environmental Laboratory (INEEL) and Massachusetts Institute of Technology (MIT) University Research Consortium (URC) project is to investigate the suitability of lead or lead-bismuth cooled fast reactors for producing low-cost electricity as well as for actinide burning. The goal is to identify and analyze the key technical issues in core neutronics, materials, thermal-hydraulics, fuels, and economics associated with the development of this reactor concept. Work has been accomplished in four major areas of research: core neutronic design, material compatibility, plant engineering, and coolant activation. In the area of core neutronic design, the reactivity vs. burnup and discharge isotopics of both non-fertile and fertile fuels were evaluated. An innovative core for pure actinide burning that uses streaming, fertile-free fuel assemblies was studied in depth. This particular core exhibits excellent reactivity performance upon coolant voiding, even for voids that occur in the core center, and has a transuranic (TRU) destruction rate that is comparable to the proposed accelerator transmutation of waste (ATW) facility. These studies suggest that a core can be designed to achieve a long life while maintaining safety and minimizing waste. In the area of material compatibility studies, an experimental apparatus for the investigation of the flow-assisted dissolution and precipitation (corrosion) of potential fuel cladding and structural materials has been designed and built at the INEEL. The INEEL forced-convection corrosion cell consists of a small heated vessel with a shroud and gas flow system. The corrosion cell is being used to test steel that is commercially available in the United States to temperatures above 650°C. Progress in plant engineering was made for two reactor concepts, one utilizing an indirect cycle with heat exchangers and the other utilizing a direct-contact steam cycle. The evaluation of the

  14. Black Hole Based Tests of General Relativity

    CERN Document Server

    Yagi, Kent

    2016-01-01

    General relativity has passed all solar system experiments and neutron star based tests, such as binary pulsar observations, with flying colors. A more exotic arena for testing general relativity is in systems that contain one or more black holes. Black holes are the most compact objects in the universe, providing probes of the strongest-possible gravitational fields. We are motivated to study strong-field gravity since many theories give large deviations from general relativity only at large field strengths, while recovering the weak-field behavior. In this article, we review how one can probe general relativity and various alternative theories of gravity by using electromagnetic waves from a black hole with an accretion disk, and gravitational waves from black hole binaries. We first review model-independent ways of testing gravity with electromagnetic/gravitational waves from a black hole system. We then focus on selected examples of theories that extend general relativity in rather simple ways. Some impor...

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

    Energy Technology Data Exchange (ETDEWEB)

    Douglas M. Gerstner

    2009-05-01

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

  16. Electric Power quality Analysis in research reactor: Impacts on nuclear safety assessment and electrical distribution reliability

    Energy Technology Data Exchange (ETDEWEB)

    Touati, Said; Chennai, Salim; Souli, Aissa [Nuclear Research Centre of Birine, Ain Oussera, Djelfa Province (Algeria)

    2015-07-01

    The increased requirements on supervision, control, and performance in modern power systems make power quality monitoring a common practise for utilities. Large databases are created and automatic processing of the data is required for fast and effective use of the available information. Aim of the work presented in this paper is the development of tools for analysis of monitoring power quality data and in particular measurements of voltage and currents in various level of electrical power distribution. The study is extended to evaluate the reliability of the electrical system in nuclear plant. Power Quality is a measure of how well a system supports reliable operation of its loads. A power disturbance or event can involve voltage, current, or frequency. Power disturbances can originate in consumer power systems, consumer loads, or the utility. The effect of power quality problems is the loss power supply leading to severe damage to equipments. So, we try to track and improve system reliability. The assessment can be focused on the study of impact of short circuits on the system, harmonics distortion, power factor improvement and effects of transient disturbances on the Electrical System during motor starting and power system fault conditions. We focus also on the review of the Electrical System design against the Nuclear Directorate Safety Assessment principles, including those extended during the last Fukushima nuclear accident. The simplified configuration of the required system can be extended from this simple scheme. To achieve these studies, we have used a demo ETAP power station software for several simulations. (authors)

  17. CERCA LEU fuel assemblies testing in Maria Reactor - safety analysis summary and testing program scope.

    Energy Technology Data Exchange (ETDEWEB)

    Pytel, K.; Mieleszczenko, W.; Lechniak, J.; Moldysz, A.; Andrzejewski, K.; Kulikowska, T.; Marcinkowska, A.; Garner, P. L.; Hanan, N. A.; Nuclear Engineering Division; Institute of Atomic Energy (Poland)

    2010-03-01

    The presented paper contains neutronic and thermal-hydraulic (for steady and unsteady states) calculation results prepared to support annex to Safety Analysis Report for MARIA reactor in order to obtain approval for program of testing low-enriched uranium (LEU) lead test fuel assemblies (LTFA) manufactured by CERCA. This includes presentation of the limits and operational constraints to be in effect during the fuel testing investigations. Also, the scope of testing program (which began in August 2009), including additional measurements and monitoring procedures, is described.

  18. Design considerations of the irradiation test vehicle for the advanced test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H.; Gomes, I.C.; Smith, D.L. [Argonne National Lab., IL (United States)] [and others

    1997-08-01

    An irradiation test vehicle (ITV) for the Advanced Test Reactor (ATR) is being jointly developed by the Lockheed Martin Idaho Technologies Company (LMIT) and the U.S. Fusion Program. The vehicle is intended for neutron irradiation testing of candidate structural materials, including vanadium-based alloys, silicon carbide composites, and low activation steels. It could possibly be used for U.S./Japanese collaboration in the Jupiter Program. The first test train is scheduled to be completed by September 1998. In this report, we present the functional requirements for the vehicle and a preliminary design that satisfies these requirements.

  19. A cycle timer for testing electric vehicles

    Science.gov (United States)

    Soltis, R. F.

    1978-01-01

    A cycle timer was developed to assist the driver of an electric vehicle in more accurately following and repeating SAE driving schedules. These schedules require operating an electric vehicle in a selected stop-and-go driving cycle and repeating this cycle pattern until the vehicle ceases to meet the requirements of the cycle. The heart of the system is a programmable read-only memory (PROM) that has the required test profiles permanently recorded on plug-in cards, one card for each different driving schedule. The PROM generates a direct current analog signal that drives a speedometer displayed on one scale of a dual movement meter. The second scale of the dual movement meter displays the actual speed of the vehicle as recorded by the fifth wheel. The vehicle operator controls vehicle speed to match the desired profile speed. The PROM controls the recycle start time as well as the buzzer activation. The cycle programmer is powered by the test vehicle's 12-volt accessory battery, through a 5-volt regulator and a 12-volt dc-to-dc converter.

  20. Parametric Thermal Models of the Transient Reactor Test Facility (TREAT)

    Energy Technology Data Exchange (ETDEWEB)

    Bradley K. Heath

    2014-03-01

    This work supports the restart of transient testing in the United States using the Department of Energy’s Transient Reactor Test Facility at the Idaho National Laboratory. It also supports the Global Threat Reduction Initiative by reducing proliferation risk of high enriched uranium fuel. The work involves the creation of a nuclear fuel assembly model using the fuel performance code known as BISON. The model simulates the thermal behavior of a nuclear fuel assembly during steady state and transient operational modes. Additional models of the same geometry but differing material properties are created to perform parametric studies. The results show that fuel and cladding thermal conductivity have the greatest effect on fuel temperature under the steady state operational mode. Fuel density and fuel specific heat have the greatest effect for transient operational model. When considering a new fuel type it is recommended to use materials that decrease the specific heat of the fuel and the thermal conductivity of the fuel’s cladding in order to deal with higher density fuels that accompany the LEU conversion process. Data on the latest operating conditions of TREAT need to be attained in order to validate BISON’s results. BISON’s models for TREAT (material models, boundary convection models) are modest and need additional work to ensure accuracy and confidence in results.

  1. Investigation of TASS/SMR Capability to Predict a Natural Circulation in the Test Facility for an Integral Reactor

    Directory of Open Access Journals (Sweden)

    Young-Jong Chung

    2014-01-01

    Full Text Available System-integrated modular advanced reactor (SMART is a small-sized advanced integral type pressurized water reactor (PWR with a rated thermal power of 330 MW. It can produce 100 MW of electricity or 90 MW of electricity and 40,000 ton of desalinated water concurrently, which is sufficient for 100,000 residents. The design features contributing to safety enhancement are basically inherent safety improvement and passive safety features. TASS/SMR code was developed for an analysis of design based events and accidents in an integral type reactor reflecting the characteristics of the SMART design. The main purpose of the code is to analyze all relevant phenomena and processes. The code should be validated using experimental data in order to confirm prediction capability. TASS/SMR predicts well the overall thermal-hydraulic behavior under various natural circulation conditions at the experimental test facility for an integral reactor. A pressure loss should be provided a function of Reynolds number at low velocity conditions in order to simulate the mass flow rate well under natural circulations.

  2. Energy Multiplier Module (EM{sup 2}) - advanced small modular reactor for electricity generation

    Energy Technology Data Exchange (ETDEWEB)

    Bertch, T.; Schleicher, R.; Choi, H.; Rawls, J., E-mail: timothy.bertch@ga.com [General Atomics, San Diego, California (United States)

    2013-07-01

    In order to provide cost effective nuclear energy in other than large reactor, large grid applications, fission technology needs to make further advances. 'Convert and burn' fast reactors offer long life cores, improved fuel utilization, reduced waste and other benefits while achieving cost effective energy production in a smaller reactor. General Atomics' Energy Multiplier Module (EM{sup 2}), a helium-cooled compact fast reactor that augments its fissile fuel load with either depleted uranium (DU) or used nuclear fuel (UNF). The convert and burn in-situ provides 250 MWe with a 30 year core life. High temperature provides a simple, high efficiency direct cycle gas turbine which along with modular construction, fewer systems, road shipment and minimum on site construction support cost effectiveness. Additional advantages in fuel cycle, non-proliferation and siting flexibility and its ability to meet all safety requirements make for an attractive power source, especially in remote and small grid regions. (author)

  3. SPECIFICATIONS FOR HIGH TEMPERATURE LATTICE TEST REACTOR BUILDING 318 PROJECT CAH-100

    Energy Technology Data Exchange (ETDEWEB)

    Vitro Engineering Company

    1964-07-15

    This is the specifications for the High Temperature Lattice Test Reactor Building 318 and it is divided into the following 21 divisions or chapters: (1) Excavating, Backfill & Grading; (2) Reinforced Concrete; (3) Masonry; (4) Structural Steel & Miscellaneous Metal Items, Contents - Division V; (5) Plumbing, Process & Service Piping; (6) Welding; (7) Insulated Metal Siding; (8) Roof Decks & Roofing; (9) Plaster Partitions & Ceiling; (10) Standard Doors, Windows & Hardware; (11) Shielding Doors; (12) Sprinkler System & Fire Extinguishers, Contents - Division XIII; (13) Heating, Ventilating & Air Conditioning; (14) Painting, Protective Coating & Floor Covering, Contents - Division XV; (15) Electrical; (16) Communications & Alarm Systems; (17) Special Equipment & Furnishings; (18) Overhead Bridge Crane; (19) Prefabricated Steel Building; (20) Paved Drive; and (21) Landscaping & Irrigation Sprinklers.

  4. Sipping test update device for fuel elements cladding inspections in IPR-r1 TRIGA reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rodrigues, R.R.; Mesquita, A.Z.; Andrade, E.P.D.; Gual, Maritza R., E-mail: rrr@cdtn.br, E-mail: amir@cdtn.br, E-mail: edson@cdtn.br, E-mail: maritzargual@gmail.com [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    It is in progress at the Centro de Desenvolvimento da Tecnologia Nuclear - CDTN (Nuclear Technology Development Center), a research project that aims to investigate possible leaks in the fuel elements of the TRIGA reactor, located in this research center. This paper presents the final form of sipping test device for TRIGA reactor, and results of the first experiments setup. Mechanical support strength tests were made by knotting device on the crane, charged with water from the conventional water supply, and tests outside the reactor pool with the use of new non-irradiated fuel elements encapsulated in stainless steel, and available safe stored in this unit. It is expected that tests with graphite elements from reactor pool are done soon after and also the test experiment with the first fuel elements in service positioned in the B ring (central ring) of the reactor core in the coming months. (author)

  5. 78 FR 47235 - Airworthiness Directives; General Electric Company Turbofan Engines

    Science.gov (United States)

    2013-08-05

    ... Electric Company (GE) model GEnx-2B67 and GEnx-2B67B turbofan engines. This proposed AD was prompted by the... in Boeing 747-8 flight tests had consumed more cyclic life than they would have in revenue flight cycles. These parts were then installed into engines and introduced into revenue service...

  6. FreedomCAR :electrical energy storage system abuse test manual for electric and hybrid electric vehicle applications.

    Energy Technology Data Exchange (ETDEWEB)

    Doughty, Daniel Harvey; Crafts, Chris C.

    2006-08-01

    This manual defines a complete body of abuse tests intended to simulate actual use and abuse conditions that may be beyond the normal safe operating limits experienced by electrical energy storage systems used in electric and hybrid electric vehicles. The tests are designed to provide a common framework for abuse testing various electrical energy storage systems used in both electric and hybrid electric vehicle applications. The manual incorporates improvements and refinements to test descriptions presented in the Society of Automotive Engineers Recommended Practice SAE J2464 ''Electric Vehicle Battery Abuse Testing'' including adaptations to abuse tests to address hybrid electric vehicle applications and other energy storage technologies (i.e., capacitors). These (possibly destructive) tests may be used as needed to determine the response of a given electrical energy storage system design under specifically defined abuse conditions. This manual does not provide acceptance criteria as a result of the testing, but rather provides results that are accurate and fair and, consequently, comparable to results from abuse tests on other similar systems. The tests described are intended for abuse testing any electrical energy storage system designed for use in electric or hybrid electric vehicle applications whether it is composed of batteries, capacitors, or a combination of the two.

  7. Development of automatic Ultrasonic testing equipment for reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kor R.; Kim, Jae H.; Lee, Jae C.

    1996-06-01

    The selected weld areas of a reactor pressure vessel and adjacent piping are examined by the remote mechanized ultrasonic testing (MUT) equipment. Since the MUT equipment was purchased from southwest Research Institute (SwRI) in April 1985, 15 inservice inspections and 5 preservice inspections are performed with this MUT equipment. However due to the old age of the equipment and frequent movements to plant sites, the reliability of examination was recently decreased rapidly and it is very difficult to keep spare parts. In order to resolve these problems and to meet the strong request from plant sites, we intend to develop a new 3-axis control system including hardware and software. With this control system, we expect more efficient and reliable examination of the nozzle to shell weld areas, which is specified in ASME Code Section XI. The new 3-axis control system hardware and software were designed and development of our own control system, the advanced technologies of computer control mechanism were established and examination reliability of the nozzle to shell weld area was improved. With the development of our 3-axis control system for PaR ISI-2 computer control system, the reliability of nozzle to shell weld area examination has been improved. The established technologies from the development and detailed analysis of existing control system, are expected to be applied to the similar control systems in nuclear power plants. (author). 12 refs., 4 tabs., 33 figs.

  8. ELECTRICAL RESISTANCE HEATING OF SOILS AT C-REACTOR AT THE SAVANNAH RIVER SITE

    Energy Technology Data Exchange (ETDEWEB)

    Blundy, R; Michael Morgenstern, M; Joseph Amari, J; Annamarie MacMurray, A; Mark Farrar, M; Terry Killeen, T

    2007-09-10

    Chlorinated solvent contamination of soils and groundwater is an endemic problem at the Savannah River Site (SRS), and originated as by-products from the nuclear materials manufacturing process. Five nuclear reactors at the SRS produced special nuclear materials for the nation's defense program throughout the cold war era. An important step in the process was thorough degreasing of the fuel and target assemblies prior to irradiation. Discharges from this degreasing process resulted in significant groundwater contamination that would continue well into the future unless a soil remediation action was performed. The largest reactor contamination plume originated from C-Reactor and an interim action was selected in 2004 to remove the residual trichloroethylene (TCE) source material by electrical resistance heating (ERH) technology. This would be followed by monitoring to determine the rate of decrease in concentration in the contaminant plume. Because of the existence of numerous chlorinated solvent sources around SRS, it was elected to generate in-house expertise in the design and operation of ERH, together with the construction of a portable ERH/SVE system that could be deployed at multiple locations around the site. This paper describes the waste unit characteristics, the ERH system design and operation, together with extensive data accumulated from the first deployment adjacent to the C-Reactor building. The installation heated the vadose zone down to 62 feet bgs over a 60 day period during the summer of 2006 and raised soil temperatures to over 200 F. A total of 730 lbs of trichloroethylene (TCE) were removed over this period, and subsequent sampling indicated a removal efficiency of 99.4%.

  9. Design, Test and Demonstration of Saturable Reactor High-Temperature Superconductor Fault Current Limiters

    Energy Technology Data Exchange (ETDEWEB)

    Darmann, Frank [Zenergy Power, Inc., Burlingame, CA (United States); Lombaerde, Robert [Zenergy Power, Inc., Burlingame, CA (United States); Moriconi, Franco [Zenergy Power, Inc., Burlingame, CA (United States); Nelson, Albert [Zenergy Power, Inc., Burlingame, CA (United States)

    2012-03-01

    Zenergy Power has successfully designed, built, tested, and installed in the US electrical grid a saturable reactor Fault Current Limiter. Beginning in 2007, first as SC Power Systems and from 2008 as Zenergy Power, Inc., ZP used DOE matching grant and ARRA funds to help refine the design of the saturated reactor fault current limiter. ZP ultimately perfected the design of the saturated reactor FCL to the point that ZP could reliably design a suitable FCL for most utility applications. Beginning with a very basic FCL design using 1G HTS for a coil housed in a LN2 cryostat for the DC bias magnet, the technology progressed to a commercial system that was offered for sale internationally. Substantial progress was made in two areas. First, the cryogenics cooling system progressed from a sub-cooled liquid nitrogen container housing the HTS coils to cryostats utilizing dry conduction cooling and reaching temperatures down to less than 20 degrees K. Large, round cryostats with warm bore diameters of 1.7 meters enabled the design of large tanks to hold the AC components. Second, the design of the AC part of the FCL was refined from a six legged spider design to a more compact and lighter design with better fault current limiting capability. Further refinement of the flux path and core shape led to an efficient saturated reactor design requiring less Ampere-turns to saturate the core. In conclusion, the development of the saturable reactor FCL led to a more efficient design not requiring HTS magnets and their associated peripheral equipment, which yielded a more economical product in line with the electric utility industry expectations. The original goal for the DOE funding of the ZP project Design, Test and Demonstration of Saturable Reactor High-Temperature Superconductor Fault Current Limiters was to stimulate the HTS wire industry with, first 1G, then 2G, HTS wire applications. Over the approximately 5 years of ZP's product development program, the amount of HTS

  10. A general theory of phase noise in electrical oscillators

    OpenAIRE

    Hajimiri, Ali; Lee, Thomas H.

    1998-01-01

    A general model is introduced which is capable of making accurate, quantitative predictions about the phase noise of different types of electrical oscillators by acknowledging the true periodically time-varying nature of all oscillators. This new approach also elucidates several previously unknown design criteria for reducing close-in phase noise by identifying the mechanisms by which intrinsic device noise and external noise sources contribute to the total phase noise. In particular, it expl...

  11. STATUS OF TRISO FUEL IRRADIATIONS IN THE ADVANCED TEST REACTOR SUPPORTING HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGNS

    Energy Technology Data Exchange (ETDEWEB)

    Davenport, Michael; Petti, D. A.; Palmer, Joe

    2016-11-01

    The United States Department of Energy’s Advanced Reactor Technologies (ART) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The experiments will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of several independent capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and completed in October 2013. The third and fourth experiments have been combined into a single experiment designated (AGR-3/4), which started its irradiation in December 2011 and completed in April 2014. Since the purpose of this experiment was to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment was significantly different from the first two experiments, though the control

  12. Black hole based tests of general relativity

    Science.gov (United States)

    Yagi, Kent; Stein, Leo C.

    2016-03-01

    General relativity has passed all solar system experiments and neutron star based tests, such as binary pulsar observations, with flying colors. A more exotic arena for testing general relativity is in systems that contain one or more black holes. Black holes are the most compact objects in the Universe, providing probes of the strongest-possible gravitational fields. We are motivated to study strong-field gravity since many theories give large deviations from general relativity only at large field strengths, while recovering the weak-field behavior. In this article, we review how one can probe general relativity and various alternative theories of gravity by using electromagnetic waves from a black hole with an accretion disk, and gravitational waves from black hole binaries. We first review model-independent ways of testing gravity with electromagnetic/gravitational waves from a black hole system. We then focus on selected examples of theories that extend general relativity in rather simple ways. Some important characteristics of general relativity include (but are not limited to) (i) only tensor gravitational degrees of freedom, (ii) the graviton is massless, (iii) no quadratic or higher curvatures in the action, and (iv) the theory is four-dimensional. Altering a characteristic leads to a different extension of general relativity: (i) scalar-tensor theories, (ii) massive gravity theories, (iii) quadratic gravity, and (iv) theories with large extra dimensions. Within each theory, we describe black hole solutions, their properties, and current and projected constraints on each theory using black hole based tests of gravity. We close this review by listing some of the open problems in model-independent tests and within each specific theory.

  13. Appendix to HDC 2118 design criteria 100-X reactor water plant, general description - section II

    Energy Technology Data Exchange (ETDEWEB)

    1952-03-29

    The factors responsible for the advances of 100-X compared with the older areas are: Simplification of the process, such as elimination of separate process water clearwells, by having the filtered water reservoirs perform that function. Combination of separate buildings into one building, such as combining filter pump house and process pump house. Use of electric standby. Use of higher capacity pumps and filter basins, and so fewer number of units. Centralization of control and operation. More compact arrangement of plant components. Use of waste heat for space heating, recovered from reactor effluent, backed up by steam plant.

  14. Evaluation of zinc addition on fuel cladding corrosion at the Halden test reactor. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Kolstad, E.; Symons, W.J.; Bryhn-Integrigtsen, K.; Oberlaender, B.C.

    1996-08-01

    Experimental studies have shown that addition of zinc to a PWR environment reduces the general corrosion rates of materials in the primary system and delays the initiation of primary water stress corrosion cracking (PWSCC) in Alloy 600. In order to provide an early warning of any potential adverse effects on the fuel cladding, corrosion studies were initiated at the Halden test reactor. These tests were carried out in a PWR rig inserted in the Halden reactor core. The rig simulated thermal hydraulic and coolant conditions typical of a MR. It had two flow channels where the fuel rod segments were exposed to the coolant under irradiation flux. Selected pre-characterized rodlets with fresh and pre-irradiated standard and low-tin Zircaloy-4 material were irradiated for three cycles. First cycle lasted for 110 effective full power days (EFPDs), the second for 95 EFPDs and the last 62 EFPDs. The cladding corrosion behavior was monitored by initial, interim and final oxide thickness measurements by eddy current lift-off probe. Crud sampling was performed in both channels after cycle 1 and 2. Destructive post-irradiation examinations (PIE) of two rodlets, irradiated during cycle 1 and 2, have also been completed at the conclusion of the in-pile testing. This report presents the results on oxide thickness measurements, irradiation history and water chemistry data, and the PIE.

  15. Mechanical properties test data for structural materials. Quarterly progress report for period ending October 31, 1976

    Energy Technology Data Exchange (ETDEWEB)

    Hill, M R [comp.

    1976-12-01

    Test data on heat resisting reactor materials are presented. These data were obtained in research at EG and G Idaho, Argonne National Laboratory, Oak Ridge National Laboratory, Naval Research Laboratory, Hanford Engineering Development Laboratory, Westinghouse Advanced Reactors Division, General Electric Company, University of Cincinnati, and University of California at Los Angeles. (JRD)

  16. Differential Prediction Generalization in College Admissions Testing

    Science.gov (United States)

    Aguinis, Herman; Culpepper, Steven A.; Pierce, Charles A.

    2016-01-01

    We introduce the concept of "differential prediction generalization" in the context of college admissions testing. Specifically, we assess the extent to which predicted first-year college grade point average (GPA) based on high-school grade point average (HSGPA) and SAT scores depends on a student's ethnicity and gender and whether this…

  17. Differential Prediction Generalization in College Admissions Testing

    Science.gov (United States)

    Aguinis, Herman; Culpepper, Steven A.; Pierce, Charles A.

    2016-01-01

    We introduce the concept of "differential prediction generalization" in the context of college admissions testing. Specifically, we assess the extent to which predicted first-year college grade point average (GPA) based on high-school grade point average (HSGPA) and SAT scores depends on a student's ethnicity and gender and whether this…

  18. Adaptation of Crack Growth Detection Techniques to US Material Test Reactors

    Energy Technology Data Exchange (ETDEWEB)

    A. Joseph Palmer; Sebastien P. Teysseyre; Kurt L. Davis; Joy L. Rempe; Gordon Kohse; Yakov Ostrovsky; David M. Carpenter

    2014-04-01

    A key component in evaluating the ability of Light Water Reactors to operate beyond 60 years is characterizing the degradation of materials exposed to radiation and various water chemistries. Of particular concern is the response of reactor materials to Irradiation Assisted Stress Corrosion Cracking (IASCC). Some materials testing reactors (MTRs) outside the U.S., such as the Halden Boiling Water Reactor (HBWR), have deployed a technique to measure crack growth propagation during irradiation. This technique incorporates a compact loading mechanism to stress the specimen during irradiation. A crack in the specimen is monitored using the Direct Current Potential Drop (DCPD) method. A project is underway to develop and demonstrate the performance of a similar type of test rig for use in U.S. MTRs. The first year of this three year project was devoted to designing, analyzing, fabricating, and bench top testing a mechanism capable of applying a controlled stress to specimens while they are irradiated in a pressurized water loop (simulating PWR reactor conditions). During the second year, the mechanism will be tested in autoclaves containing high pressure, high temperature water with representative water chemistries. In addition, necessary documentation and safety reviews for testing in a reactor environment will be completed. In the third year, the assembly will be tested in the Massachusetts Institute of Technology Reactor (MITR) and Post Irradiation Examinations (PIE) will be performed.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-07-01

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

  20. Tensile and electrical properties of copper alloys irradiated in a fission reactor

    Energy Technology Data Exchange (ETDEWEB)

    Fabritsiev, S.A. [D.V. Efremov Inst., St. Petersburg (Russian Federation); Pokrovsky, A.S. [Scientific Research Inst. of Atomic Reactors, Dimitrovgrad (Russian Federation); Zinkle, S.J.; Rowcliffe, A.F. [Oak Ridge National Laboratory, TN (United States)] [and others

    1996-04-01

    Postirradiation electrical sensitivity and tensile measurements have been completed on pure copper and copper alloy sheet tensile specimens irradiated in the SM-2 reactor to doses of {approx}0.5 to 5 dpa and temperatures between {approx}80 and 400{degrees}C. Considerable radiation hardening and accompanying embrittlement was observed in all of the specimens at irradiation temperature below 200{degrees}C. The radiation-induced electrical conductivity degradation consisted of two main components: solid transmutation effects and radiation damage (defect cluster and particle dissolution) effects. The radiation damage component was nearly constant for the doses in this study, with a value of {approx}1.2n{Omega}m for pure copper and {approx}1.6n{Omega}m for dispersion strengthened copper irradiated at {approx}100{degrees}C. The solid transmutation component was proportional to the thermal neutron flux, and became larger than the radiation damage component for fluences larger than {approx}5 10{sup 24} n.m{sup 2}. The radiation hardening and electrical conductivity degradation decreased with increasing irradiation temperature, and became negligible for temperatures above {approx}300{degrees}C.

  1. Tests of Electric Polarizability on the Lattice

    CERN Document Server

    Guerrero, V X; Christensen, J

    2008-01-01

    Using clover fermions on CP-PACS dynamical configurations, we consider a number of ways of measuring hadronic electric polarizability, an $|\\mathbf{E}|^{2}$ effect in hadron masses, using lattice techniques. We consider the effects of periodic and Dirichlet boundary conditions, the field linearization postulate as well as a quantized electric field. We also consider two ways of formulating the classical vector potential which describes a uniform electric field in combination with the other possibilities.

  2. Electric field gradient, generalized Sternheimer shieldings and electric field gradient polarizabilities by multiconfigurational SCF response

    Science.gov (United States)

    Rizzo, Antonio; Ruud, Kenneth; Helgaker, Trygve; Jaszuński, Michał

    1998-08-01

    The electric field gradient (EFG) at the nuclei, the generalized Sternheimer shielding constants and the EFG electric dipole polarizabilities are computed for eight small molecules employing multiconfigurational self-consistent field wave functions and the corresponding linear and quadratic response functions. The molecules studied are H2, N2, CO, HF, C2H2, HCl, HCN, and HNC, all of which are linear. For the hydrogen molecule, full configuration-interaction results for the properties are also reported. The dependence of the computed quantities on the basis set and the electron-correlation treatment is analyzed.

  3. The Effect of Current-Limiting Reactors on the Tripping of Short Circuits in High-Voltage Electrical Equipment

    Energy Technology Data Exchange (ETDEWEB)

    Volkov, M. S.; Gusev, Yu. P., E-mail: GusevYP@mpei.ru; Monakov, Yu. V.; Cho, Gvan Chun [National Research University “Moscow Power Engineering Institute,” (Russian Federation)

    2016-01-15

    The insertion of current-limiting reactors into electrical equipment operating at a voltage of 110 and 220 kV produces a change in the parameters of the transient recovery voltages at the contacts of the circuit breakers for disconnecting short circuits, which could be the reason for the increase in the duration of the short circuit, damage to the electrical equipment and losses in the power system. The results of mathematical modeling of the transients, caused by tripping of the short circuit in a reactive electric power transmission line are presented, and data are given on the negative effect of a current-limiting resistor on the rate of increase and peak value of the transient recovery voltages. Methods of ensuring the standard requirements imposed on the parameters of the transient recovery voltages when using current-limiting reactors in the high-voltage electrical equipment of power plants and substations are proposed and analyzed.

  4. 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

    1998-01-01

    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.)

  5. Test Plan for the Boiling Water Reactor Dry Cask Simulator

    Energy Technology Data Exchange (ETDEWEB)

    Durbin, Samuel [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Lindgren, Eric R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-11-01

    The thermal performance of commercial nuclear spent fuel dry storage casks are evaluated through detailed numerical analysis . These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and by increasing the internal convection through greater canister helium pressure. These same vertical, canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both above and below-ground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of above-ground and below-ground canistered dry cask systems. The purpose of the investigation described in this report is to produce a data set that can be used to test the validity of the assumptions associated with the calculations presently used to determine steady-state cladding temperatures in modern vertical, canistered dry cask systems. The BWR cask simulator (BCS) has been designed in detail for both the above-ground and below-ground venting configurations. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure (MAWP) rating of 24 bar at 400 deg C. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly is being deployed inside of a representative storage basket and cylindrical pressure vessel that represents the

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

    Science.gov (United States)

    Bragg-Sitton, Shannon M.; Forsbacka, Matthew

    2004-01-01

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

  7. Preliminary Study on the High Efficiency Supercritical Pressure Water-Cooled Reactor for Electricity Generation

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Yoon Yeong; Park, Jong Kyun; Cho, Bong Hyun and others

    2006-01-15

    This research has been performed to introduce a concept of supercritical pressure water cooled reactor(SCWR) in Korea The area of research includes core conceptual design, evaluation of candidate fuel, fluid systems conceptual design with mechanical consideration, preparation of safety analysis code, and construction of supercritical pressure heat transfer test facility, SPHINX, and preliminary test. As a result of the research, a set of tools for the reactor core design has been developed and the conceptual core design with solid moderator was proposed. The direct thermodynamic cycle has been studied to find a optimum design. The safety analysis code has also been adapted to supercritical pressure condition. A supercritical pressure CO2 heat transfer test facility has been constructed and preliminary test proved the facility works as expected. The result of this project will be good basis for the participation in the international collaboration under GIF GEN-IV program and next 5-year mid and long term nuclear research program of MOST. The heat transfer test loop, SPHINX, completed as a result of this project may be used for the power cycle study as well as further heat transfer study for the various geometries.

  8. Multimegawatt nuclear electric propulsion with gaseous and vapor core reactors with MHD

    Science.gov (United States)

    Knight, Travis; Anghaie, Samim; Smith, Blair; Houts, Michael

    2001-02-01

    This study investigated the development of a system concept for space power generation and nuclear electric propulsion based on a fissioning plasma core reactor (FPCR) with magnetohydrodynamic (MHD) power conversion system, coupled to a magnetoplasmadynamic (MPD) thruster. The FPCR is a liquid-vapor core reactor concept operating with metallic uranium or uranium tetrafluoride (UF4) vapor as the fissioning fuel and alkali metals or their fluorides as working fluid in a closed Rankine cycle with MHD energy conversion. Candidate working fluids include K, Li, Na, KF, LiF, NaF, etc. The system features core outlet temperatures of 3000 to 4000 K at pressures of about 1 to 10 MPa, MHD temperatures of 2000 to 3000 K, and radiator temperatures of 1200 to 2000 K. This combination of parameters offers the potential for low total system specific mass in the range of 0.4 to 0.6 kg/kWe. The MHD output could be coupled with minimal power conditioning to the variable specific impulse magnetoplasma rocket (VASIMR), MPD thrusters or other types of thruster for producing thrust at very high specific impulse (Isp=1500 to 10,000 s). .

  9. Reactor

    Science.gov (United States)

    Evans, Robert M.

    1976-10-05

    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.

  10. ENERGY EFFICIENCY OF ELECTRIC HEATING OF REACTORS IN THE MANUFACTURE OF VARNISHES AND PAINTS

    Directory of Open Access Journals (Sweden)

    Tovajniansky L.L.

    2014-08-01

    Full Text Available The drawbacks of the traditional design of the heaters, which make known imperfections in manufacturing processes, realized with the use of electric heating. This determines the need for a radically new design of the heating devices. Created by high-temperature ceramics, characterized by abnormally high thermal stability and clarified the parameters that allow a certain degree change its thermal conductivity. On this basis the contact type ceramic heaters that provide thermal flow direction using different materials in the body of the heater - of high thermal conductivity, the surface facing the heat transfer and low which differs opposite sides of the heater are designed to eliminate the dissipation of heat into the surrounding space. This made it possible to equip the modern production paint industry energy efficient heating equipment with explosion and fire heating system reactors.

  11. CArbon-free Electricity by SEWGS. Advanced materials, reactor-, and process design CAESAR

    Energy Technology Data Exchange (ETDEWEB)

    Jansen, D. [ECN Hydrogen and Clean Fossil Fuels, Petten (Netherlands)

    2009-11-15

    In this presentation attention is paid to the activities within the CAESAR project. CAESAR is the acronym for 'CArbon-free Electricity by SEWGS: Advanced materials, Reactor-, and process design'. SEWGS stands for 'Sorption Enhanced Water Gas Shift'. The CAESAR project builds on CACHET SEWGS. CACHET is the acronym of the project with the overall goal to develop innovative technologies which will substantially reduce the cost of CO2 capture whilst simultaneously producing H2 from natural gas fuel. The first 12 month of CAESAR focused on sorbent development and process feasibility for NGCC (Natural Gas Combined Cycle) and IGCC (Integrated Gasification Combined Cycle). Improved sorbent capacity and improved mechanical strength are demonstrated. Proper integration of the SEWGS process in NGCC will improve power generation efficiencies.

  12. 76 FR 70166 - Electrical Standards for Construction and General Industry; Extension of the Office of Management...

    Science.gov (United States)

    2011-11-10

    ... installation and maintenance of electric utilization equipment that prevent death and serious injuries among... Occupational Safety and Health Administration Electrical Standards for Construction and General Industry... collection requirements contained in the Electrical Standards for Construction (29 CFR part 1926, Subpart...

  13. Study of fast reactor safety test facilities. Preliminary report

    Energy Technology Data Exchange (ETDEWEB)

    Bell, G.I.; Boudreau, J.E.; McLaughlin, T.; Palmer, R.G.; Starkovich, V.; Stein, W.E.; Stevenson, M.G.; Yarnell, Y.L.

    1975-05-01

    Included are sections dealing with the following topics: (1) perspective and philosophy of fast reactor safety analysis; (2) status of accident analysis and experimental needs; (3) experiment and facility definitions; (4) existing in-pile facilities; (5) new facility options; and (6) data acquisition methods. (DG)

  14. Emission and transmission tomography systems to be developed for the future needs of Jules Horowitz material testing reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kotiluoto, Petri [VTT Technical Research Centre of Finland, P.O.Box 1000, FI-02044 VTT (Finland)], E-mail: petri.kotiluoto@vtt.fi; Wasastjerna, Frej; Kekki, Tommi [VTT Technical Research Centre of Finland, P.O.Box 1000, FI-02044 VTT (Finland); Sipilae, Heikki; Banzuzi, Kukka [Oxford Instruments Analytical Oy, Nihtisillankuja 5, P.O.Box 85, FI-02631 Espoo (Finland); Kinnunen, Petri; Heikinheimo, Liisa [VTT Technical Research Centre of Finland, P.O.Box 1000, FI-02044 VTT (Finland)

    2009-08-01

    The new 100 MW Jules Horowitz material testing reactor will be built in Cadarache, France. It will support, for instance, research on new types of innovative nuclear fuel. As a Finnish in-kind contribution, 3D emission and transmission tomography equipment will be delivered for both the reactor and the active component storage pool. The image reconstruction of activities inside the used nuclear fuel will be based on gamma spectrometry measurements. A new type of underwater digital X-ray linear detector array is under development for transmission imaging, based on GaAs and direct conversion of X-rays into an electrical signal. A shared collimator will be used for both emission and transmission measurements. Some preliminary design has been performed. For the current design, the expected gamma spectrometric response of a typical high-purity germanium detector has been simulated with MCNP for minimum and maximum source activities (specified by CEA) to be measured in future.

  15. Aging assessment of Westinghouse PWR and General Electric BWR containment isolation functions

    Energy Technology Data Exchange (ETDEWEB)

    Lee, B.S.; Travis, R.; Grove, E.; DiBiasio, A.

    1996-03-01

    A study was performed to assess the effects of aging on the Containment Isolation (CI) functions of Westinghouse Pressurized Water Reactors and General Electric Boiling Water Reactors. This study is part of the Nuclear Plant Aging Research (NPAR) program, sponsored by the U.S. Nuclear Regulatory Commission. The objectives of this program are to provide an understanding of the aging process and how it affects plant safety so that it can be properly managed. This is one of a number of studies performed under the NPAR program which provide a technical basis for the identification and evaluation of degradation caused by age. Failure data from two national databases, Nuclear Plant Reliability Data System (NPRDS) and Licensee Event Reports (LERs), as well as plant specific data were reviewed and analyzed to understand the effects of aging on the CI functions. This study provided information on the effects of aging on component failure frequency, failure modes, and failure causes. Current inspection, surveillance, and monitoring practices were also reviewed.

  16. Electrical Emissions from Concrete under Three-point Bending Tests

    Institute of Scientific and Technical Information of China (English)

    LIU Qingping; SUN Mingqing; LI Zhuoqiu; LI Guoqiang

    2005-01-01

    Electrical emission(EM) signals, which are generated from the concrete specimens under three-point bending tests, were conducted. It is shown that electrical emission phenomena are related to cracking of the specimens, cohesive failure, contact-separation etc. The simultaneous appearance of electric emission signals and visible cracks during the flexure loading of beams was also observed.

  17. IRRADIATION TESTING OF THE RERTR FUEL MINIPLATES WITH BURNABLE ABSORBERS IN THE ADVANCED TEST REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    I. Glagolenko; D. Wachs; N. Woolstenhulme; G. Chang; B. Rabin; C. Clark; T. Wiencek

    2010-10-01

    Based on the results of the reactor physics assessment, conversion of the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) can be potentially accomplished in two ways, by either using U-10Mo monolithic or U-7Mo dispersion type plates in the ATR fuel element. Both designs, however, would require incorporation of the burnable absorber in several plates of the fuel element to compensate for the excess reactivity and to flatten the radial power profile. Several different types of burnable absorbers were considered initially, but only borated compounds, such as B4C, ZrB2 and Al-B alloys, were selected for testing primarily due to the length of the ATR fuel cycle and fuel manufacturing constraints. To assess and compare irradiation performance of the U-Mo fuels with different burnable absorbers we have designed and manufactured 28 RERTR miniplates (20 fueled and 8 non-fueled) containing fore-mentioned borated compounds. These miniplates will be tested in the ATR as part of the RERTR-13 experiment, which is described in this paper. Detailed plate design, compositions and irradiations conditions are discussed.

  18. 76 FR 31381 - Office Of New Reactors; Proposed Revision 4 to Standard Review Plan; Section 8.1 on Electric...

    Science.gov (United States)

    2011-05-31

    ... Analysis Reports for Nuclear Power Plants,'' on a proposed Revision 4 to Standard Review Plan (SRP... Position (BTP) 8-8 on ``Onsite (Emergency Diesel Generators) and Offsite Power Sources Allowed Outage Time... COMMISSION Office Of New Reactors; Proposed Revision 4 to Standard Review Plan; Section 8.1 on Electric...

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

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    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

  20. An Analysis of Testing Requirements for Fluoride Salt Cooled High Temperature Reactor Components

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, David Eugene [ORNL; Cetiner, Sacit M [ORNL; Flanagan, George F [ORNL; Peretz, Fred J [ORNL; Yoder Jr, Graydon L [ORNL

    2009-11-01

    This report provides guidance on the component testing necessary during the next phase of fluoride salt-cooled high temperature reactor (FHR) development. In particular, the report identifies and describes the reactor component performance and reliability requirements, provides an overview of what information is necessary to provide assurance that components will adequately achieve the requirements, and then provides guidance on how the required performance information can efficiently be obtained. The report includes a system description of a representative test scale FHR reactor. The reactor parameters presented in this report should only be considered as placeholder values until an FHR test scale reactor design is completed. The report focus is bounded at the interface between and the reactor primary coolant salt and the fuel and the gas supply and return to the Brayton cycle power conversion system. The analysis is limited to component level testing and does not address system level testing issues. Further, the report is oriented as a bottom-up testing requirements analysis as opposed to a having a top-down facility description focus.

  1. Fatigue Test of Domestic Manufactured Reactor Pressure Vessel Steel

    Institute of Scientific and Technical Information of China (English)

    ZHONG; Wei-hua; TONG; Zhen-feng; NING; Guang-sheng; YU; Bin-tao

    2013-01-01

    The CAP1400 will be built by our country,after the self-dependent innovation work on the imported technology of AP1000,which is a 3rd generation NPP.Now,the design of CAP1400 key equipment is ongoing,and the fatigue design of the domestic manufactured key equipment,such as reactor pressure vessel(RPV),is found to be a main problem in the design work,as the fatigue data is lacked.Thus the

  2. 10 CFR 830 Major Modification Determination for the Advanced Test Reactor Remote Monitoring and Management Capability

    Energy Technology Data Exchange (ETDEWEB)

    Bohachek, Randolph Charles [Idaho National Laboratory (INL), Idaho Falls, ID (United States)

    2015-09-01

    The Advanced Test Reactor (ATR; TRA-670), which is located in the ATR Complex at Idaho National Laboratory, was constructed in the 1960s for the purpose of irradiating reactor fuels and materials. Other irradiation services, such as radioisotope production, are also performed at ATR. While ATR is safely fulfilling current mission requirements, assessments are continuing. These assessments intend to identify areas to provide defense–in-depth and improve safety for ATR. One of the assessments performed by an independent group of nuclear industry experts recommended that a remote accident management capability be provided. The report stated that: “contemporary practice in commercial power reactors is to provide a remote shutdown station or stations to allow shutdown of the reactor and management of long-term cooling of the reactor (i.e., management of reactivity, inventory, and cooling) should the main control room be disabled (e.g., due to a fire in the control room or affecting the control room).” This project will install remote reactor monitoring and management capabilities for ATR. Remote capabilities will allow for post scram reactor management and monitoring in the event the main Reactor Control Room (RCR) must be evacuated.

  3. Towards a general theory of optimal testing

    Science.gov (United States)

    Pericchi, Luis R. G.; Pereira, Carlos A. B.

    2012-10-01

    In Pericchi and Pereira [1] it is argued against the traditional way on which testing is based on fixed significance level, either using p-values (with fixed levels of evidence, like the 5% rule) or α values. We instead, follow an approach put forward by [2], on which an optimal test is chosen by minimizing type I and type II errors. Morris DeGroot in his authoritative book [2], Probability and Statistics 2nd Edition, stated that it is more reasonable to minimize a weighted sum of Type I and Type II error than to specify a value of type I error and then minimize Type II error. He showed it beyond reasonable doubt, but only in the very restrictive scenario of simple VS simple hypothesis, and it is not clear how to generalize it. We propose here a very natural generalization for composite hypothesis, by using general weight functions in the parameter space. This was also the position taken by [3, 4, 5]. We show, in a parallel manner to DeGroot's proof and Pereira's discussion, that the optimal test statistics are Bayes Factors, when the weighting functions are priors with mass on the whole parameter space. On the other hand when the weight functions are point masses in specific parameter values of practical significance, then a procedure is designed for which the sum of Type I error and Type II error in the specified points of practical significance is minimized. This can be seen as bridge between Bayesian Statistics and a new version of Hypothesis testing, more in line with statistical consistency and scientific insight.

  4. A superconducting gyroscope to test Einstein's general theory of relativity

    Science.gov (United States)

    Everitt, C. W. F.

    1978-01-01

    Schiff (1960) proposed a new test of general relativity based on measuring the precessions of the spin axes of gyroscopes in earth orbit. Since 1963 a Stanford research team has been developing an experiment to measure the two effects calculated by Schiff. The gyroscope consists of a uniform sphere of fused quartz 38 mm in diameter, coated with superconductor, electrically suspended and spinning at about 170 Hz in vacuum. The paper describes the proposed flight apparatus and the current state of development of the gyroscope, including techniques for manufacturing and measuring the gyro rotor and housing, generating ultralow magnetic fields, and mechanizing the readout.

  5. Tests of General Relativity with GW150914.

    Science.gov (United States)

    Abbott, B P; Abbott, R; Abbott, T D; Abernathy, M R; Acernese, F; Ackley, K; Adams, C; Adams, T; Addesso, P; Adhikari, R X; Adya, V B; Affeldt, C; Agathos, M; Agatsuma, K; Aggarwal, N; Aguiar, O D; Aiello, L; Ain, A; Ajith, P; Allen, B; Allocca, A; Altin, P A; Anderson, S B; Anderson, W G; Arai, K; Araya, M C; Arceneaux, C C; Areeda, J S; Arnaud, N; Arun, K G; Ascenzi, S; Ashton, G; Ast, M; Aston, S M; Astone, P; Aufmuth, P; Aulbert, C; Babak, S; Bacon, P; Bader, M K M; Baker, P T; Baldaccini, F; Ballardin, G; Ballmer, S W; Barayoga, J C; Barclay, S E; Barish, B C; Barker, D; Barone, F; Barr, B; Barsotti, L; Barsuglia, M; Barta, D; Bartlett, J; Bartos, I; Bassiri, R; Basti, A; Batch, J C; Baune, C; Bavigadda, V; Bazzan, M; Behnke, B; Bejger, M; Bell, A S; Bell, C J; Berger, B K; Bergman, J; Bergmann, G; Berry, C P L; Bersanetti, D; Bertolini, A; Betzwieser, J; Bhagwat, S; Bhandare, R; Bilenko, I A; Billingsley, G; Birch, J; Birney, R; Birnholtz, O; Biscans, S; Bisht, A; Bitossi, M; 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Gopakumar, A; Gordon, N A; Gorodetsky, M L; Gossan, S E; Gosselin, M; Gouaty, R; Graef, C; Graff, P B; Granata, M; Grant, A; Gras, S; Gray, C; Greco, G; Green, A C; Groot, P; Grote, H; Grunewald, S; Guidi, G M; Guo, X; Gupta, A; Gupta, M K; Gushwa, K E; Gustafson, E K; Gustafson, R; Hacker, J J; Hall, B R; Hall, E D; Hammond, G; Haney, M; Hanke, M M; Hanks, J; Hanna, C; Hannam, M D; Hanson, J; Hardwick, T; Harms, J; Harry, G M; Harry, I W; Hart, M J; Hartman, M T; Haster, C-J; Haughian, K; Healy, J; Heidmann, A; Heintze, M C; Heitmann, H; Hello, P; Hemming, G; Hendry, M; Heng, I S; Hennig, J; Heptonstall, A W; Heurs, M; Hild, S; Hoak, D; Hodge, K A; Hofman, D; Hollitt, S E; Holt, K; Holz, D E; Hopkins, P; Hosken, D J; Hough, J; Houston, E A; Howell, E J; Hu, Y M; Huang, S; Huerta, E A; Huet, D; Hughey, B; Husa, S; Huttner, S H; Huynh-Dinh, T; Idrisy, A; Indik, N; Ingram, D R; Inta, R; Isa, H N; Isac, J-M; Isi, M; Islas, G; Isogai, T; Iyer, B R; Izumi, K; Jacqmin, T; Jang, H; Jani, K; Jaranowski, P; Jawahar, S; Jiménez-Forteza, F; Johnson, W W; Johnson-McDaniel, N K; Jones, D I; Jones, R; Jonker, R J G; Ju, L; Haris, M K; Kalaghatgi, C V; Kalogera, V; Kandhasamy, S; Kang, G; Kanner, J B; Karki, S; Kasprzack, M; Katsavounidis, E; Katzman, W; Kaufer, S; Kaur, T; Kawabe, K; Kawazoe, F; Kéfélian, F; Kehl, M S; Keitel, D; Kelley, D B; Kells, W; Kennedy, R; Key, J S; Khalaidovski, A; Khalili, F Y; Khan, I; Khan, S; Khan, Z; Khazanov, E A; Kijbunchoo, N; Kim, C; Kim, J; Kim, K; Kim, Nam-Gyu; Kim, Namjun; Kim, Y-M; King, E J; King, P J; Kinzel, D L; Kissel, J S; Kleybolte, L; Klimenko, S; Koehlenbeck, S M; Kokeyama, K; Koley, S; Kondrashov, V; Kontos, A; Korobko, M; Korth, W Z; Kowalska, I; Kozak, D B; Kringel, V; Krishnan, B; Królak, A; Krueger, C; Kuehn, G; Kumar, P; Kuo, L; Kutynia, A; Lackey, B D; Landry, M; Lange, J; Lantz, B; Lasky, P D; Lazzarini, A; Lazzaro, C; Leaci, P; Leavey, S; Lebigot, E O; Lee, C H; Lee, H K; Lee, H M; Lee, K; Lenon, A; Leonardi, M; Leong, J R; Leroy, N; Letendre, N; 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Verkindt, D; Vetrano, F; Viceré, A; Vinciguerra, S; Vine, D J; Vinet, J-Y; Vitale, S; Vo, T; Vocca, H; Vorvick, C; Voss, D; Vousden, W D; Vyatchanin, S P; Wade, A R; Wade, L E; Wade, M; Walker, M; Wallace, L; Walsh, S; Wang, G; Wang, H; Wang, M; Wang, X; Wang, Y; Ward, R L; Warner, J; Was, M; Weaver, B; Wei, L-W; Weinert, M; Weinstein, A J; Weiss, R; Welborn, T; Wen, L; Weßels, P; Westphal, T; Wette, K; Whelan, J T; White, D J; Whiting, B F; Williams, D; Williams, R D; Williamson, A R; Willis, J L; Willke, B; Wimmer, M H; Winkler, W; Wipf, C C; Wittel, H; Woan, G; Worden, J; Wright, J L; Wu, G; Yablon, J; Yam, W; Yamamoto, H; Yancey, C C; Yap, M J; Yu, H; Yvert, M; Zadrożny, A; Zangrando, L; Zanolin, M; Zendri, J-P; Zevin, M; Zhang, F; Zhang, L; Zhang, M; Zhang, Y; Zhao, C; Zhou, M; Zhou, Z; Zhu, X J; Zucker, M E; Zuraw, S E; Zweizig, J; Boyle, M; Campanelli, M; Hemberger, D A; Kidder, L E; Ossokine, S; Scheel, M A; Szilagyi, B; Teukolsky, S; Zlochower, Y

    2016-06-03

    The LIGO detection of GW150914 provides an unprecedented opportunity to study the two-body motion of a compact-object binary in the large-velocity, highly nonlinear regime, and to witness the final merger of the binary and the excitation of uniquely relativistic modes of the gravitational field. We carry out several investigations to determine whether GW150914 is consistent with a binary black-hole merger in general relativity. We find that the final remnant's mass and spin, as determined from the low-frequency (inspiral) and high-frequency (postinspiral) phases of the signal, are mutually consistent with the binary black-hole solution in general relativity. Furthermore, the data following the peak of GW150914 are consistent with the least-damped quasinormal mode inferred from the mass and spin of the remnant black hole. By using waveform models that allow for parametrized general-relativity violations during the inspiral and merger phases, we perform quantitative tests on the gravitational-wave phase in the dynamical regime and we determine the first empirical bounds on several high-order post-Newtonian coefficients. We constrain the graviton Compton wavelength, assuming that gravitons are dispersed in vacuum in the same way as particles with mass, obtaining a 90%-confidence lower bound of 10^{13}  km. In conclusion, within our statistical uncertainties, we find no evidence for violations of general relativity in the genuinely strong-field regime of gravity.

  6. Tests of General Relativity with GW150914

    Science.gov (United States)

    Abbott, B. P.; Abbott, R.; Abbott, T. D.; Abernathy, M. R.; Acernese, F.; Ackley, K.; Adams, C.; Adams, T.; Addesso, P.; Adhikari, R. X.; Adya, V. B.; Affeldt, C.; Agathos, M.; Agatsuma, K.; Aggarwal, N.; Aguiar, O. D.; Aiello, L.; Ain, A.; Ajith, P.; Allen, B.; Allocca, A.; Altin, P. A.; Anderson, S. B.; Anderson, W. G.; Arai, K.; Araya, M. C.; Arceneaux, C. C.; Areeda, J. S.; Arnaud, N.; Arun, K. G.; Ascenzi, S.; Ashton, G.; Ast, M.; Aston, S. M.; Astone, P.; Aufmuth, P.; Aulbert, C.; Babak, S.; Bacon, P.; Bader, M. K. M.; Baker, P. T.; Baldaccini, F.; Ballardin, G.; Ballmer, S. W.; Barayoga, J. C.; Barclay, S. E.; Barish, B. C.; Barker, D.; Barone, F.; Barr, B.; Barsotti, L.; Barsuglia, M.; Barta, D.; Bartlett, J.; Bartos, I.; Bassiri, R.; Basti, A.; Batch, J. C.; Baune, C.; Bavigadda, V.; Bazzan, M.; Behnke, B.; Bejger, M.; Bell, A. S.; Bell, C. J.; Berger, B. K.; Bergman, J.; Bergmann, G.; Berry, C. P. L.; Bersanetti, D.; Bertolini, A.; Betzwieser, J.; Bhagwat, S.; Bhandare, R.; Bilenko, I. A.; Billingsley, G.; Birch, J.; Birney, R.; Birnholtz, O.; Biscans, S.; Bisht, A.; Bitossi, M.; Biwer, C.; Bizouard, M. A.; Blackburn, J. K.; Blair, C. D.; Blair, D. G.; Blair, R. M.; Bloemen, S.; Bock, O.; Bodiya, T. P.; Boer, M.; Bogaert, G.; Bogan, C.; Bohe, A.; Bojtos, P.; Bond, C.; Bondu, F.; Bonnand, R.; Boom, B. A.; Bork, R.; Boschi, V.; Bose, S.; Bouffanais, Y.; Bozzi, A.; Bradaschia, C.; Brady, P. R.; Braginsky, V. B.; Branchesi, M.; Brau, J. E.; Briant, T.; Brillet, A.; Brinkmann, M.; Brisson, V.; Brockill, P.; Brooks, A. F.; Brown, D. A.; Brown, D. D.; Brown, N. M.; Buchanan, C. C.; Buikema, A.; Bulik, T.; Bulten, H. J.; Buonanno, A.; Buskulic, D.; Buy, C.; Byer, R. L.; Cadonati, L.; Cagnoli, G.; Cahillane, C.; Calderón Bustillo, J.; Callister, T.; Calloni, E.; Camp, J. B.; Cannon, K. C.; Cao, J.; Capano, C. D.; Capocasa, E.; Carbognani, F.; Caride, S.; Casanueva Diaz, J.; Casentini, C.; Caudill, S.; Cavaglià, M.; Cavalier, F.; Cavalieri, R.; Cella, G.; Cepeda, C. B.; Cerboni Baiardi, L.; Cerretani, G.; Cesarini, E.; Chakraborty, R.; Chalermsongsak, T.; Chamberlin, S. J.; Chan, M.; Chao, S.; Charlton, P.; Chassande-Mottin, E.; Chen, H. Y.; Chen, Y.; Cheng, C.; Chincarini, A.; Chiummo, A.; Cho, H. S.; Cho, M.; Chow, J. H.; Christensen, N.; Chu, Q.; Chua, S.; Chung, S.; Ciani, G.; Clara, F.; Clark, J. A.; Cleva, F.; Coccia, E.; Cohadon, P.-F.; Colla, A.; Collette, C. G.; Cominsky, L.; Constancio, M.; Conte, A.; Conti, L.; Cook, D.; Corbitt, T. R.; Cornish, N.; Corsi, A.; Cortese, S.; Costa, C. A.; Coughlin, M. W.; Coughlin, S. B.; Coulon, J.-P.; Countryman, S. T.; Couvares, P.; Cowan, E. E.; Coward, D. M.; Cowart, M. J.; Coyne, D. C.; Coyne, R.; Craig, K.; Creighton, J. D. E.; Cripe, J.; Crowder, S. G.; Cumming, A.; Cunningham, L.; Cuoco, E.; Dal Canton, T.; Danilishin, S. L.; D'Antonio, S.; Danzmann, K.; Darman, N. S.; Dattilo, V.; Dave, I.; Daveloza, H. P.; Davier, M.; Davies, G. S.; Daw, E. J.; Day, R.; DeBra, D.; Debreczeni, G.; Degallaix, J.; De Laurentis, M.; Deléglise, S.; Del Pozzo, W.; Denker, T.; Dent, T.; Dereli, H.; Dergachev, V.; De Rosa, R.; DeRosa, R. T.; DeSalvo, R.; Dhurandhar, S.; Díaz, M. C.; Di Fiore, L.; Di Giovanni, M.; Di Lieto, A.; Di Pace, S.; Di Palma, I.; Di Virgilio, A.; Dojcinoski, G.; Dolique, V.; Donovan, F.; Dooley, K. L.; Doravari, S.; Douglas, R.; Downes, T. P.; Drago, M.; Drever, R. W. P.; Driggers, J. C.; Du, Z.; Ducrot, M.; Dwyer, S. E.; Edo, T. B.; Edwards, M. C.; Effler, A.; Eggenstein, H.-B.; Ehrens, P.; Eichholz, J.; Eikenberry, S. S.; Engels, W.; Essick, R. C.; Etzel, T.; Evans, M.; Evans, T. M.; Everett, R.; Factourovich, M.; Fafone, V.; Fair, H.; Fairhurst, S.; Fan, X.; Fang, Q.; Farinon, S.; Farr, B.; Farr, W. M.; Favata, M.; Fays, M.; Fehrmann, H.; Fejer, M. M.; Ferrante, I.; Ferreira, E. C.; Ferrini, F.; Fidecaro, F.; Fiori, I.; Fiorucci, D.; Fisher, R. P.; Flaminio, R.; Fletcher, M.; Fournier, J.-D.; Franco, S.; Frasca, S.; Frasconi, F.; Frei, Z.; Freise, A.; Frey, R.; Frey, V.; Fricke, T. T.; Fritschel, P.; Frolov, V. V.; Fulda, P.; Fyffe, M.; Gabbard, H. A. G.; Gair, J. R.; Gammaitoni, L.; Gaonkar, S. G.; Garufi, F.; Gatto, A.; Gaur, G.; Gehrels, N.; Gemme, G.; Gendre, B.; Genin, E.; Gennai, A.; George, J.; Gergely, L.; Germain, V.; Ghosh, Abhirup; Ghosh, Archisman; Ghosh, S.; Giaime, J. A.; Giardina, K. D.; Giazotto, A.; Gill, K.; Glaefke, A.; Goetz, E.; Goetz, R.; Gondan, L.; González, G.; Gonzalez Castro, J. M.; Gopakumar, A.; Gordon, N. A.; Gorodetsky, M. L.; Gossan, S. E.; Gosselin, M.; Gouaty, R.; Graef, C.; Graff, P. B.; Granata, M.; Grant, A.; Gras, S.; Gray, C.; Greco, G.; Green, A. C.; Groot, P.; Grote, H.; Grunewald, S.; Guidi, G. M.; Guo, X.; Gupta, A.; Gupta, M. K.; Gushwa, K. E.; Gustafson, E. K.; Gustafson, R.; Hacker, J. J.; Hall, B. R.

    2016-06-01

    The LIGO detection of GW150914 provides an unprecedented opportunity to study the two-body motion of a compact-object binary in the large-velocity, highly nonlinear regime, and to witness the final merger of the binary and the excitation of uniquely relativistic modes of the gravitational field. We carry out several investigations to determine whether GW150914 is consistent with a binary black-hole merger in general relativity. We find that the final remnant's mass and spin, as determined from the low-frequency (inspiral) and high-frequency (postinspiral) phases of the signal, are mutually consistent with the binary black-hole solution in general relativity. Furthermore, the data following the peak of GW150914 are consistent with the least-damped quasinormal mode inferred from the mass and spin of the remnant black hole. By using waveform models that allow for parametrized general-relativity violations during the inspiral and merger phases, we perform quantitative tests on the gravitational-wave phase in the dynamical regime and we determine the first empirical bounds on several high-order post-Newtonian coefficients. We constrain the graviton Compton wavelength, assuming that gravitons are dispersed in vacuum in the same way as particles with mass, obtaining a 90%-confidence lower bound of 1013 km . In conclusion, within our statistical uncertainties, we find no evidence for violations of general relativity in the genuinely strong-field regime of gravity.

  7. Electrical conductivity as a state indicator for the start-up period of anaerobic fixed-bed reactors.

    Science.gov (United States)

    Robles, A; Latrille, E; Ribes, J; Bernet, N; Steyer, J P

    2016-01-01

    The aim of this work was to analyse the applicability of electrical conductivity sensors for on-line monitoring the start-up period of an anaerobic fixed-bed reactor. The evolution of bicarbonate concentration and methane production rate was analysed. Strong linear relationships between electrical conductivity and both bicarbonate concentration and methane production rate were observed. On-line estimations of the studied parameters were carried out in a new start-up period by applying simple linear regression models, which resulted in a good concordance between both observed and predicted values. Electrical conductivity sensors were therefore identified as an interesting method for monitoring the start-up period of anaerobic fixed-bed reactors due to its reliability, robustness, easy operation, low cost, and minimum maintenance compared with the currently used sensors.

  8. Mixed oxide fuels testing in the advanced test reactor to support plutonium disposition

    Energy Technology Data Exchange (ETDEWEB)

    Ryskamp, J.M.; Sterbentz, J.W.; Chang, G.S. [and others

    1995-09-01

    An intense worldwide effort is now under way to find means of reducing the stockpile of weapons-grade plutonium. One of the most attractive solutions would be to use WGPu as fuel in existing light water reactors (LWRs) in the form of mixed oxide (MOX) fuel - i.e., plutonia (PUO{sub 2}) mixed with urania (UO{sub 2}). Before U.S. reactors could be used for this purpose, their operating licenses would have to be amended. Numerous technical issues must be resolved before LWR operating licenses can be amended to allow the use of MOX fuel. These issues include the following: (1) MOX fuel fabrication process verification, (2) Whether and how to use burnable poisons to depress MOX fuel initial reactivity, which is higher than that of urania, (3) The effects of WGPu isotopic composition, (4) The feasibility of loading MOX fuel with plutonia content up to 7% by weight, (5) The effects of americium and gallium in WGPu, (6) Fission gas release from MOX fuel pellets made from WGPu, (7) Fuel/cladding gap closure, (8) The effects of power cycling and off-normal events on fuel integrity, (9) Development of radial distributions of burnup and fission products, (10) Power spiking near the interfaces of MOX and urania fuel assemblies, and (11) Fuel performance code validation. We have performed calculations to show that the use of hafnium shrouds can produce spectrum adjustments that will bring the flux spectrum in ATR test loops into a good approximation to the spectrum anticipated in a commercial LWR containing MOX fuel while allowing operation of the test fuel assemblies near their optimum values of linear heat generation rate. The ATR would be a nearly ideal test bed for developing data needed to support applications to license LWRs for operation with MOX fuel made from weapons-grade plutonium. The requirements for planning and implementing a test program in the ATR have been identified.

  9. Testing for Spatial Isotropy Under General Designs.

    Science.gov (United States)

    Maity, Arnab; Sherman, Michael

    2012-05-01

    Spatial modeling is typically composed of a specification of a mean function and a model for the correlation structure. A common assumption on the spatial correlation is that it is isotropic. This means that the correlation between any two observations depends only on the distance between those sites and not on their relative orientation. The assumption of isotropy is often made due to a simpler interpretation of correlation behavior and to an easier estimation problem under an assumed isotropy. The assumption of isotropy, however, can have serious deleterious effects when not appropriate. In this paper we formulate a test of isotropy for spatial observations located according to a general class of stochastic designs. Distribution theory of our test statistic is derived and we carry out extensive simulations which verify the efficacy of our approach. We apply our methodology to a data set on longleaf pine trees from an oldgrowth forest in the southern United States.

  10. The Jules Horowitz Reactor - A new High Performance European Material Testing Reactor open to International Users Present Status and Objectives

    Energy Technology Data Exchange (ETDEWEB)

    Iracane, Daniel; Bignan, Gilles [CEA Atomic Energy Commission Saclay Batiment 121- 91191 Gif Sur Yvette (France); Lindbaeck, Jan-Erik; Blomgren, Jan [VATTENFALL AB Nuclear Power Jaemtlandsgatan 99 SE-16287 Stockholm (Sweden)

    2010-07-01

    The development of sustainable nuclear energy requires R and D on fuel and material behaviour under irradiation with a high level of performance in order to meet the needs and challenges for the benefit of industry, research and public bodies. These stakes require a sustainable and secured access to an up-to-date high performance Material Testing Reactor. Following a broad survey within the European Research Area, the international community agreed that the need for Material Test Reactors in support of nuclear power plant safety and operation will continue in the context of sustainable nuclear energy. The Jules Horowitz Reactor project (JHR) copes with this context. JHR is designed as a user facility addressing the needs of the international community. This means: - flexibility with irradiation loops able to reproduce a large variation in operation conditions of different power reactor technologies, - high flux capacity to address Generations II, III, and IV needs. JHR is designed, built and operated as an international user facility because: - Given the maturity and globalization of the industry, domestic tools have no more the required level of economic and technical efficiency. Meanwhile, countries with nuclear energy need an access to high performance irradiation experimental capabilities to support technical skill and guarantee the competitiveness and safety of nuclear energy. - Many research items related to safety or public policy (waste management, etc.) require international cooperation to share costs and benefits of resulting consensus. JHR design is optimised for offering high performance material and fuel irradiation capability for the coming decades. This project is driven and funded by an international consortium gathering vendors, utilities and public stakeholders. This consortium has been set up in March 2007 when the construction began. The construction is in progress and the start of operation is scheduled for 2014. The JHR is a research

  11. A generalized exponential time series regression model for electricity prices

    DEFF Research Database (Denmark)

    Haldrup, Niels; Knapik, Oskar; Proietti, Tomasso

    We consider the issue of modeling and forecasting daily electricity spot prices on the Nord Pool Elspot power market. We propose a method that can handle seasonal and non-seasonal persistence by modelling the price series as a generalized exponential process. As the presence of spikes can distort...... the estimation of the dynamic structure of the series we consider an iterative estimation strategy which, conditional on a set of parameter estimates, clears the spikes using a data cleaning algorithm, and reestimates the parameters using the cleaned data so as to robustify the estimates. Conditional...... on the estimated model, the best linear predictor is constructed. Our modeling approach provides good fit within sample and outperforms competing benchmark predictors in terms of forecasting accuracy. We also find that building separate models for each hour of the day and averaging the forecasts is a better...

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

    Science.gov (United States)

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

    2013-01-01

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

  13. Efficiency Test Method for Electric Vehicle Chargers

    DEFF Research Database (Denmark)

    Kieldsen, Andreas; Thingvad, Andreas; Martinenas, Sergejus

    2016-01-01

    This paper investigates different methods for measuring the charger efficiency of mass produced electric vehicles (EVs), in order to compare the different models. The consumers have low attention to the loss in the charger though the impact on the driving cost is high. It is not a high priority a...

  14. Common-Cause Failure Analysis for Reactor Protection System Reliability Studies

    Energy Technology Data Exchange (ETDEWEB)

    Gentillon, C.; Rasmuson, D.; Eide, S.; Wierman, T.

    1999-08-01

    Analyses were performed of the safety-related performance of the reactor protection system (RPS) at U.S. Westinghouse and General Electric commercial reactors during the period 1984 through 1995. RPS operational data from these reactors were collected from the Nuclear Plant Reliability Data System (NPRDS) and Licensee Event Reports (LER). The common-cause failure (CCF) modeling in the fault trees developed for these studies and the analysis and use of common-cause failure data were sophisticated, state-of-the-art efforts. The overall CCF effort helped to test and expand the limits of the U.S. Nuclear Regulatory Commission's CCF methodology.

  15. Evaluation of the Initial Isothermal Physics Measurements at the Fast Flux Test Facility, a Prototypic Liquid Metal Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess

    2010-03-01

    The Fast Flux Test Facility (FFTF) was a 400-MWt, sodium-cooled, low-pressure, high-temperature, fast-neutron flux, nuclear fission reactor plant designed for the irradiation testing of nuclear reactor fuels and materials for the development of liquid metal fast breeder reactors (LMFBRs). The FFTF was fueled with plutonium-uranium mixed oxide (MOX) and reflected by Inconel-600. Westinghouse Hanford Company operated the FFTF as part of the Hanford Engineering Development Laboratory (HEDL) for the U.S. Department of Energy on the Hanford Site near Richland, Washington. Although the FFTF was a testing facility not specifically designed to breed fuel or produce electricity, it did provide valuable information for LMFBR projects and base technology programs in the areas of plant system and component design, component fabrication, prototype testing, and site construction. The major objectives of the FFTF were to provide a strong, disciplined engineering base for the LMFBR program, provide fast flux testing for other U.S. programs, and contribute to the development of a viable self-sustaining competitive U.S. LMFBR industry. During its ten years of operation, the FFTF acted as a national research facility to test advanced nuclear fuels, materials, components, systems, nuclear power plant operating and maintenance procedures, and active and passive reactor safety technologies; it also produced a large number of isotopes for medical and industrial users, generated tritium for the U.S. fusion research program, and participated in cooperative, international research work. Prior to the implementation of the reactor characterization program, a series of isothermal physics measurements were performed; this acceptance testing program consisted of a series of control rod worths, critical rod positions, subcriticality measurements, maximum reactivity addition rates, shutdown margins, excess reactivity, and isothermal temperature coefficient reactivity. The results of these

  16. Baseline tests of the Volkswagen transporter electric delivery van

    Science.gov (United States)

    Soltis, R. F.; Mcbrien, E. F.; Bozek, J. M.; Gourash, F.

    1978-01-01

    The Volkswagen Transporter, an electric delivery van, was tested as part of an Energy Research and Development Administration (ERDA) project to characterize the state of the art of electric vehicles. The Volkswagen Transporter is a standard Volkswagen van that has been converted to an electric vehicle. It is powered by a 144-volt traction battery. A direct current (dc) chopper controller, actuated by a conventional accelerator pedal, regulates the voltage or power applied to the 16-kilowatt (21-hp) motor. The braking system uses conventional hydraulic braking in combination with an electric regenerative braking system. The Volkswagen vehicle performance test results are presented.

  17. ADAPTATION OF CRACK GROWTH DETECTION TECHNIQUES TO US MATERIAL TEST REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    A. Joseph Palmer; Sebastien P. Teysseyre; Kurt L. Davis; Gordon Kohse; Yakov Ostrovsky; David M. Carpenter; Joy L. Rempe

    2015-04-01

    A key component in evaluating the ability of Light Water Reactors to operate beyond 60 years is characterizing the degradation of materials exposed to radiation and various water chemistries. Of particular concern is the response of reactor materials to Irradiation Assisted Stress Corrosion Cracking (IASCC). Some test reactors outside the United States, such as the Halden Boiling Water Reactor (HBWR), have developed techniques to measure crack growth propagation during irradiation. The basic approach is to use a custom-designed compact loading mechanism to stress the specimen during irradiation, while the crack in the specimen is monitored in-situ using the Direct Current Potential Drop (DCPD) method. In 2012 the US Department of Energy commissioned the Idaho National Laboratory and the MIT Nuclear Reactor Laboratory (MIT NRL) to take the basic concepts developed at the HBWR and adapt them to a test rig capable of conducting in-pile IASCC tests in US Material Test Reactors. The first two and half years of the project consisted of designing and testing the loader mechanism, testing individual components of the in-pile rig and electronic support equipment, and autoclave testing of the rig design prior to insertion in the MIT Reactor. The load was applied to the specimen by means of a scissor like mechanism, actuated by a miniature metal bellows driven by pneumatic pressure and sized to fit within the small in-core irradiation volume. In addition to the loader design, technical challenges included developing robust connections to the specimen for the applied current and voltage measurements, appropriate ceramic insulating materials that can endure the LWR environment, dealing with the high electromagnetic noise environment of a reactor core at full power, and accommodating material property changes in the specimen, due primarily to fast neutron damage, which change the specimen resistance without additional crack growth. The project culminated with an in

  18. The General Configuration of CEV1 Electric Vehicle's Electrical System and the Design of Its Control Sequence

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    The general configuration of CEV1 electric vehicle's electrical system and the design scheme of its control sequence are presented, which are modularized by using VMU as master control unit, PMU as power management unit, BMU as battery management unit. It is a rather advanced and practical general design scheme of electric vehicle, because the division of its module function is definite, which is advantage for research, manufacture and maintenance.

  19. Generalized F test and generalized deviance test in two-way ANOVA models for randomized trials.

    Science.gov (United States)

    Shen, Juan; He, Xuming

    2014-01-01

    We consider the problem of detecting treatment effects in a randomized trial in the presence of an additional covariate. By reexpressing a two-way analysis of variance (ANOVA) model in a logistic regression framework, we derive generalized F tests and generalized deviance tests, which provide better power in detecting common location-scale changes of treatment outcomes than the classical F test. The null distributions of the test statistics are independent of the nuisance parameters in the models, so the critical values can be easily determined by Monte Carlo methods. We use simulation studies to demonstrate how the proposed tests perform compared with the classical F test. We also use data from a clinical study to illustrate possible savings in sample sizes.

  20. Upgrading the electrical system of the IEA-R1 reactor to avoid triggering event of accidents

    Energy Technology Data Exchange (ETDEWEB)

    Mello, Jose Roberto de; Madi Filho, Tufic, E-mail: jrmello@ipen.br, E-mail: tmfilho@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    The IEA-R1 research reactor at the Institute of Energy and Nuclear Research (IPEN) is a research reactor open pool type, built and designed by the American firm 'Babcox and Wilcox', having as coolant and moderator demineralized light water and Beryllium and graphite, as reflectors. The power supply system is designed to meet the electricity demand required by the loads of the reactor (Security systems and systems not related to security) in different situations the plant can meet, such as during startup, normal operation at power, shutdown, maintenance, exchange of fuel elements and accident situations. Studies have been done on possible accident initiating events and deterministic techniques were applied to assess the consequences of such incidents. Thus, the methods used to identify and select the accident initiating events, the methods of analysis of accidents, including sequence of events, transient analysis and radiological consequences, have been described. Finally, acceptance criteria of radiological doses are described. Only a brief summary of the item concerning loss of electrical power will be presented. The loss of normal electrical power at the IEA-R1 reactor is very common. In the case of Electric External Power Loss, at the IEA-R1 reactor building, there may be different sequences of events, as described below. When the supply of external energy in the IEA-R1 facility fails, the Electrical Distribution Vital System, consisting of 4 (four) generators type 'UPS', starts operation, immediately and it will continue supplying power to the reactor control table, core cooling system and other security systems. To contribute to security, in the electric power failure, starts to operate the Emergency Cooling System (SRE). SRE has the function of removing residual heat from the core to prevent the melting of fuel elements in the event of loss of refrigerant to the core. Adding to the generators with batteries group system, new auxiliary

  1. Final report on the in situ testing of electrical components and devices at TMI-2

    Energy Technology Data Exchange (ETDEWEB)

    Soberano, F T

    1984-06-01

    A total of 88 electrical components and devices were in situ tested. Of these, 11 totally failed and 21 suffered degradation that varied from mild to severe. The equipment that failed or incurred severe degradation was located in areas of known environmental extremes. Several motor operated valves in the Reactor Building basement failed because of submersion in water. Others severely degraded from contamination tracking, resulting in the alteration of their circuit electrical characteristics - a circumstance that could compromise their designed function. One backup oil lift pump motor for a reactor coolant pump motor, although located well above the Reactor Building basement high water mark, failed because of a break in its armature and field circuits; this failure was surmised to be a result of corrosion. The limit switch of a Class 1E solenoid valve likewise failed due to moisture intrusion. Components that noticeably degraded exhibited anomalies, likely due to the incursion of moisture, that varied from high capacitance to increased circuit resistance. The effect of the other degenerating conditions that existed during the accident, such as high temperature, high radiation levels, and the hydrogen burn, could not be evaluated individually or synergistically.

  2. Electrical impedance potential mammography for visualization of objects (electrochemical tests)

    Science.gov (United States)

    Karpov, A.; Korotkova, M.; Tsofin, Yu; Tsyplyonkov, V.; Machin, M.

    2010-04-01

    The article presents the result of the electrochemical testing of the potential electric impedance mammograph "MEIK" (version 5.6) which reconstructs the image using the reciprocal projection method. The testing was carried out on several categories: biological objects visualization, definition size, location, depth and shape of the biological objects, concentration and electrical conductivity, resolution capability. The research has been conducted in the original tank filled with homogeneous and heterogeneous medium. The research of the visualization of abiological and bialogical objects, objects dimensions, objects location, objects structure. According to the data of the electrochemical testing, the resolution capability of the objects with high electrical conductivity ranges from 1 to 3 mm and of the objects with low electrical conductivity - from 3 to 5 mm. Due to the given electrochemical testing results potencial electric impedance mammograph "MEIK" (version 5.6) can be rated as the medical visualization device.

  3. Production test PTA-002, increased graphite temperature limit -- B, C and D Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Russell, A.

    1965-12-17

    The fundamental objective of the graphite temperature limit is to prevent excessive oxidation of the graphite moderator blocks with carbon dioxide and water vapor in the reactor atmosphere. Laboratory tests have shown that 10% uniform oxidation of graphite results in a loss in strength of approximately 50%. Production Test IP-725 was conducted at F Reactor for a period of six months at graphite temperatures approximately 50 and 100 C higher than the present graphite temperature limit of 650 C. The results from the F Reactor test suggest that an increase in the graphite temperature limit from 650 C to 700 C is technically feasible from the standpoint of oxidation of the graphite moderator with CO{sub 2}. Any significant additional increase was shown to lead to excessively high oxidation rates and is therefore not considered feasible. The objective of this test, therefore, is to extend the higher temperature investigations to B, C, and D Reactors. For the duration of this test, the graphite temperature limit will be increased from 650 C and 700 C, corresponding to an increase in the graphite stringer temperature limit from 735 C to 790 C. The test is expected to last for approximately six months but may be terminated early on any or all the reactors.

  4. 76 FR 647 - Energy Conservation Program: Test Procedures for Electric Motors and Small Electric Motors

    Science.gov (United States)

    2011-01-05

    ... 2008 NOPR, NEMA and Baldor commented that Congress created confusion when it struck the EPAct 1992.... According to NEMA and Baldor, the DOE definition of general purpose electric motor (subtype I) does not... and subtype II motors. Further, Baldor suggested that DOE restore the original definition of electric...

  5. Safety Issues at the DOE Test and Research Reactors. A Report to the U.S. Department of Energy.

    Science.gov (United States)

    National Academy of Sciences - National Research Council, Washington, DC. Commission on Physical Sciences, Mathematics, and Resources.

    This report provides an assessment of safety issues at the Department of Energy (DOE) test and research reactors. Part A identifies six safety issues of the reactors. These issues include the safety design philosophy, the conduct of safety reviews, the performance of probabilistic risk assessments, the reliance on reactor operators, the fragmented…

  6. Thermal analysis for a spent reactor fuel storage test in granite

    Energy Technology Data Exchange (ETDEWEB)

    Montan, D.N.

    1980-09-01

    A test is conducted in which spent fuel assemblies from an operating commercial nuclear power reactor are emplaced in the Climax granite at the US Department of Energy`s Nevada Test Site. In this generic test, 11 canisters of spent PWR fuel are emplaced vertically along with 6 electrical simulator canisters on 3 m centers, 4 m below the floor of a storage drift which is 420 m below the surface. Two adjacent parallel drifts contain electrical heaters, operated to simulate (in the vicinity of the storage drift) the temperature fields of a large repository. This test, planned for up to five years duration, uses fairly young fuel (2.5 years out of core) so that the thermal peak will occur during the time frame of the test and will not exceed the peak that would not occur until about 40 years of storage had older fuel (5 to 15 years out of core) been used. This paper describes the calculational techniques and summarizes the results of a large number of thermal calculations used in the concept, basic design and final design of the spent fuel test. The results of the preliminary calculations show the effects of spacing and spent fuel age. Either radiation or convection is sufficient to make the drifts much better thermal conductors than the rock that was removed to create them. The combination of radiation and convection causes the drift surfaces to be nearly isothermal even though the heat source is below the floor. With a nominal ventilation rate of 2 m{sup 3}/s and an ambient rock temperature of 23{sup 0}C, the maximum calculated rock temperature (near the center of the heat source) is about 100{sup 0}C while the maximum air temperature in the drift is around 40{sup 0}C. This ventilation (1 m{sup 3}/s through the main drift and 1/2 m{sup 3}/s through each of the side drifts) will remove about 1/3 of the heat generated during the first five years of storage.

  7. Halbach array motor/generators: A novel generalized electric machine

    Energy Technology Data Exchange (ETDEWEB)

    Merritt, B.T.; Post, R.F.; Dreifuerst, G.R.; Bender, D.A. [Lawrence Livermore National Lab., CA (United States)

    1995-02-01

    For many years Klaus Halbach has been investigating novel designs for permanent magnet arrays, using advanced analytical approaches and employing a keen insight into such systems. One of his motivations for this research was to find more efficient means for the utilization of permanent magnets for use in particle accelerators and in the control of particle beams. As a result of his pioneering work, high power free-electron laser systems, such as the ones built at the Lawrence Livermore Laboratory, became feasible, and his arrays have been incorporated into other particle-focusing systems of various types. This paper reports another, quite different, application of Klaus` work, in the design of high power, high efficiency, electric generators and motors. When tested, these motor/generator systems display some rather remarkable properties. Their success derives from the special properties which these arrays, which the authors choose to call {open_quotes}Halbach arrays,{close_quotes} possess.

  8. Subsize specimen testing of nuclear reactor pressure vessel material

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, A.S. [Missouri Univ., Rolla, MO (United States). Materials Research Center; Rosinski, S.T. [Sandia National Labs., Albuquerque, NM (United States); Cannon, N.S. [Westinghouse Hanford Co., Richland, WA (United States); Hamilton, M.L. [Pacific Northwest Lab., Richland, WA (United States)

    1991-12-31

    A new methodology is proposed to correlate the upper shelf energy (USE) of full size and subsize Charpy specimens of a nuclear reactor pressure vessel plate material, A533B. The methodology appears to be more satisfactory than the methodologies proposed earlier. USE of a notched-only specimen is partitioned into macro-crack initiation and crack propagation energies. USE of a notched and precracked specimen provides the crack propagation energy. {Delta}USE, the difference between the USE`s of notched-only and precracked specimens, is an estimate of the crack initiation energy. {Delta}USE was normalized by a factor involving the dimensions of the Charpy specimen and the stress concentration factor at the notch root. The normalized values of the {Delta}USE were found to be invariant with specimen size.

  9. Subsize specimen testing of nuclear reactor pressure vessel material

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, A.S. (Missouri Univ., Rolla, MO (United States). Materials Research Center); Rosinski, S.T. (Sandia National Labs., Albuquerque, NM (United States)); Cannon, N.S. (Westinghouse Hanford Co., Richland, WA (United States)); Hamilton, M.L. (Pacific Northwest Lab., Richland, WA (United States))

    1991-01-01

    A new methodology is proposed to correlate the upper shelf energy (USE) of full size and subsize Charpy specimens of a nuclear reactor pressure vessel plate material, A533B. The methodology appears to be more satisfactory than the methodologies proposed earlier. USE of a notched-only specimen is partitioned into macro-crack initiation and crack propagation energies. USE of a notched and precracked specimen provides the crack propagation energy. [Delta]USE, the difference between the USE's of notched-only and precracked specimens, is an estimate of the crack initiation energy. [Delta]USE was normalized by a factor involving the dimensions of the Charpy specimen and the stress concentration factor at the notch root. The normalized values of the [Delta]USE were found to be invariant with specimen size.

  10. 76 FR 11437 - Application To Export Electric Energy; Societe Generale Energy Corp.

    Science.gov (United States)

    2011-03-02

    ... Application To Export Electric Energy; Societe Generale Energy Corp. AGENCY: Office of Electricity Delivery.... (SGEC) has applied for authority to transmit electric energy from the United States to Canada pursuant... application from the SGEC for authority to transmit electric energy from the United States to Canada as...

  11. Electrochemical Recovery of Gold from Waste Electric and Electronic Equipment Using Circulating Particulate Bed Reactor (CPBE)

    Science.gov (United States)

    Ravinder, T.; Ali, U. F. M.; Ridwan, F. M.; Ibrahim, N.; Azmi, N. H.

    2017-06-01

    The utilization of electrochemical process recovery involving low reactant concentrations of metal requires electrodes with high mass transport rates and specific surface areas. This is essential to increase cross-sectional current densities whilst optimizing the capital and operating costs. Experimental results demonstrated that Circulating Particulate Bed Reactor (CPBE) is suitable for the recovery of low concentrations of gold from aqueous chloride solution containing {{AuCl}}4- and {{AuCl}}2- of less than 0.5 mol m-3(mass transport controlled at an applied potential of +0.20 V (SCE), specific electrical energy consumption (SEEC) of ca. 800-1300 kWh h (tonne Au)-1 for cell voltages (U) of 2.0-3.0 V, and fractional current efficiencies of ca. 0.95. However, atomic absorption and UV spectrophotometry established that as the ([{{AuCl}}4-+[{{AuCl}}2-]) concentration decayed, the [{{AuCl}}4-]:[{{AuCl}}2-] molar ratio changed. A multi-step mechanism for reduction of {{AuCl}}4- ions explained this behavior in terms of changing overpotentials for {{AuCl}}4- and {{AuCl}}2- reduction as total dissolved gold concentration decreased. In addition, SEM images confirmed that adherent and coherent Au deposits were achieved with CPBE for Au deposition under mass transport control at 0.20 V (SCE).

  12. An numerical analysis of high-temperature helium reactor power plant for co-production of hydrogen and electricity

    Science.gov (United States)

    Dudek, M.; Podsadna, J.; Jaszczur, M.

    2016-09-01

    In the present work, the feasibility of using a high temperature gas cooled nuclear reactor (HTR) for electricity generation and hydrogen production are analysed. The HTR is combined with a steam and a gas turbine, as well as with the system for heat delivery for medium temperature hydrogen production. Industrial-scale hydrogen production using copper-chlorine (Cu-Cl) thermochemical cycle is considered and compared with high temperature electrolysis. Presented cycle shows a very promising route for continuous, efficient, large-scale and environmentally benign hydrogen production without CO2 emissions. The results show that the integration of a high temperature helium reactor, with a combined cycle for electric power generation and hydrogen production, may reach very high efficiency and could possibly lead to a significant decrease of hydrogen production costs.

  13. Reference architecture for interoperability testing of Electric Vehicle charging

    NARCIS (Netherlands)

    Lehfuss, F.; Nohrer, M.; Werkmany, E.; Lopezz, J.A.; Zabalaz, E.

    2015-01-01

    This paper presents a reference architecture for interoperability testing of electric vehicles as well as their support equipment with the smart grid and the e-Mobility environment. Pan-European Electric Vehicle (EV)-charging is currently problematic as there are compliance and interoperability

  14. 30 CFR 75.831 - Electrical work; troubleshooting and testing.

    Science.gov (United States)

    2010-07-01

    ... and continuous mining machine electrical work procedures. Prior to performing electrical work, other than troubleshooting and testing, on the high-voltage trailing cable or the continuous mining machine... determine voltages and currents; (3) By qualified persons; and (4) When using protective gloves...

  15. Recirculation pump discharge line break tests at ROSA-III for a boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, M.; Anoda, Y.; Kumamaru, H.; Nakamura, H.; Shiba, M.; Tasaka, K.

    1985-08-01

    Three loss-of-coolant accident (LOCA) tests were conducted at the Rig of Safety Assessment (ROSA)-III test facility, which simulates boiling water reactor (BWR)/6-251 with a volumetric scaling factor of 1/424. The fundamental features of the recirculation pump discharge line break LOCA and the effects of break areas on the features are investigated. It has been confirmed experimentally that the LOCA phenomena in the discharge line break are analogous to those in the suction line break with the same effective choking flow area, which is a sum of the least choking flow areas along the break flow paths and controls the system pressure responses. In general, the maximum effective choking flow area is (A /SUB j/ + A /SUB p/ ) for discharge line breaks and (A /SUB j/ + A /SUB o/ ) for suction line breaks, where A /SUB j/ , A /SUB p/ , and A /SUB o/ are the flow areas of the jet pump drive nozzles, the main recirculation pump discharge nozzle, and the break, respectively. The similarity between the ROSA-III test and a BWR LOCA has been confirmed in the key phenomena by the analyses using the RELAP5/MOD1 code. An atypical behavior is observed in the fuel rod surface temperature transient in the early phase of blowdown due to the limitation of the ROSA-III initial core power.

  16. Opportunities for mixed oxide fuel testing in the advanced test reactor to support plutonium disposition

    Energy Technology Data Exchange (ETDEWEB)

    Terry, W.K.; Ryskamp, J.M.; Sterbentz, J.W. [and others

    1995-08-01

    Numerous technical issues must be resolved before LWR operating licenses can be amended to allow the use of MOX fuel. These issues include the following: (1) MOX fuel fabrication process verification; (2) Whether and how to use burnable poisons to depress MOX fuel initial reactivity, which is higher than that of urania; (3) The effects of WGPu isotopic composition; (4) The feasibility of loading MOX fuel with plutonia content up to 7% by weight; (5) The effects of americium and gallium in WGPu; (6) Fission gas release from MOX fuel pellets made from WGPu; (7) Fuel/cladding gap closure; (8) The effects of power cycling and off-normal events on fuel integrity; (9) Development of radial distributions of burnup and fission products; (10) Power spiking near the interfaces of MOX and urania fuel assemblies; and (11) Fuel performance code validation. The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory possesses many advantages for performing tests to resolve most of the issues identified above. We have performed calculations to show that the use of hafnium shrouds can produce spectrum adjustments that will bring the flux spectrum in ATR test loops into a good approximation to the spectrum anticipated in a commercial LWR containing MOX fuel while allowing operation of the test fuel assemblies near their optimum values of linear heat generation rate. The ATR would be a nearly ideal test bed for developing data needed to support applications to license LWRs for operation with MOX fuel made from weapons-grade plutonium. The requirements for planning and implementing a test program in the ATR have been identified. The facilities at Argonne National Laboratory-West can meet all potential needs for pre- and post-irradiation examination that might arise in a MOX fuel qualification program.

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

    Energy Technology Data Exchange (ETDEWEB)

    M. Chen; CM Regan; D. Noe

    2006-01-09

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

  18. Flow-induced vibration for light water reactors. Final progress report, July 1981-September 1981

    Energy Technology Data Exchange (ETDEWEB)

    Torres, M.R.

    1981-11-01

    Flow-Induced Vibration for Light Water Reactors (FIV for LWRs) is a program designed to improve the FIV performance of light water reactors through the development of design criteria, analytical models for predicting behavior of components, and general scaling laws to improve the accuracy of reduced-scale tests, and through the identification of high FIV risk areas. The program is managed by the General Electric Nuclear Power Systems Engineering Department and has three major contributors: General Electric Nuclear Power Systems Engineering Department (NPSED), General Electric Corporate Research and Development (CR and D) and Argonne National Laboratory (ANL). The program commenced December 1, 1976. This progress report summarizes the accomplishments achieved during the final period from July 1981 to September 1981. This is the last quarterly progress report to be issued for this program.

  19. A cycle timer for testing electric vehicles

    Science.gov (United States)

    Soltis, R. F.

    1978-01-01

    The paper presents a cycle timer which enables the accurate following and repetition of SAE driving schedules of stop and go cycles, for electric vehicles, by reducing the human factor. The system which consists of a programmable read-only memory (PROM) stores each of these cycles, which are detailed, on its own plug-in card. The actual vehicle speed, and the PROM indicated desired speed are displayed on a dual scale meter allowing the driver to match them. A speed change is preceded by a half second buzzer warning and a new cycle by a one second warning. The PROM controls the recycle start time as well as the buzzer activation. A 5 volt regulator providing logic power, and a 12 volt dc-dc converter providing analog and memory power are described.

  20. Testing and qualification of Control and Safety Rod and its drive mechanism of Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rajan Babu, V., E-mail: vrb@igcar.gov.i [Indira Gandhi Centre for Atomic Research, Department of Atomic Energy, Kalpakkam 603 102 (India); Veerasamy, R.; Patri, Sudheer; Ignatius Sundar Raj, S.; Kumar Krovvidi, S.C.S.P.; Dash, S.K.; Meikandamurthy, C.; Rajan, K.K.; Puthiyavinayagam, P.; Chellapandi, P.; Vaidyanathan, G.; Chetal, S.C. [Indira Gandhi Centre for Atomic Research, Department of Atomic Energy, Kalpakkam 603 102 (India)

    2010-07-15

    Prototype Fast Breeder Reactor (PFBR) has two independent fast acting diverse shutdown systems. The absorber rod of the first system is called Control and Safety Rod (CSR). CSR and its Drive Mechanism (CSRDM) are used for reactor control and for safe shutdown of the reactor by scram action. In view of the safety role, the qualification of CSRDM is one of the important requirements. CSR and CSRDM were qualified in two stages by extensive testing. In the first stage, the critical subassemblies of the mechanism, such as scram release electromagnet, hydraulic dashpot and dynamic seals and CSR subassembly, were tested and qualified individually simulating the operating conditions of the reactor. Experiments were also carried out on sodium vapour deposition in the annular gaps between the stationary and mobile parts of the mechanism. In the second stage, full-scale CSRDM and CSR were subjected to all the integrated functional tests in air, hot argon and subsequently in sodium simulating the operating conditions of the reactor and finally subjected to endurance tests. Since the damage occurring in CSRDM and CSR is mainly due to fatigue cycles during scram actions, the number of test cycles was decided based on the guidelines given in ASME, Section III, Div. 1. The results show that the performance of CSRDM and CSR is satisfactory. Subsequent to the testing in sodium, the assemblies having contact with liquid sodium/sodium vapour were cleaned using CO{sub 2} process and the total cleaning process has been established, so that the mechanism can be reused in sodium. The various stages of qualification programmes have raised the confidence level on the performance of the system as a whole for the intended and reliable operation in the reactor.

  1. Design of Seismic Test Rig for Control Rod Drive Mechanism of Jordan Research and Training Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sun, Jongoh; Kim, Gyeongho; Yoo, Yeonsik; Cho, Yeonggarp; Kim, Jong In [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The reactor assembly is submerged in a reactor pool filled with water and its reactivity is controlled by locations of four control absorber rods(CARs) inside the reactor assembly. Each CAR is driven by a stepping motor installed at the top of the reactor pool and they are connected to each other by a tie rod and an electromagnet. The CARs scram the reactor by de-energizing the electromagnet in the event of a safe shutdown earthquake(SSE). Therefore, the safety function of the control rod drive mechanism(CRDM) which consists of a drive assembly, tie rod and CARs is to drop the CAR into the core within an appropriate time in case of the SSE. As well known, the operability for complex equipment such as the CRDM during an earthquake is very hard to be demonstrated by analysis and should be verified through tests. One of them simulates the reactor assembly and the guide tube of the CAR, and the other one does the pool wall where the drive assembly is installed. In this paper, design of the latter test rig and how the test is performed are presented. Initial design of the seismic test rig and excitation table had its first natural frequency at 16.3Hz and could not represent the environment where the CRDM was installed. Therefore, experimental modal analyses were performed and an FE model for the test rig and table was obtained and tuned based on the experimental results. Using the FE model, the design of the test rig and table was modified in order to have higher natural frequency than the cutoff frequency. The goal was achieved by changing its center of gravity and the stiffness of its sliding bearings.

  2. Strategies of General.Electric company's Development

    Institute of Scientific and Technical Information of China (English)

    杨娜

    2016-01-01

    As the biggest multinational corporate providing technology and services, the success of General Electric (GE) is always used for reference by other companies. This report is to evaluate the suitability of GE's strategic position from 2001 to early 2014. After a brief company overview, the report first points out current strategies which GE applies and analyses GE's activities with strategy concepts such as Porter's generic strategy and' Ansoff matrix. Strategic position analysis consists pf environmental analysis, capability analysis and stakeholders' expectations and purposes. GE's environmental analysis is drawn from macro-environment with PEST analysis and micro-environment with Porter's five forces analysis, and opportunities and threats are evaluated based on that. After that, the report introduces resources deployed by GE and therefore outlines the strengths and weakness of GE's strategy. Following GE's capability analysis, the stakeholders' power distribution and purposes part describes corporate governance, stakeholders' expectation and CSR to illustrate how GE's strategies are affected by stakeholder's exnectations.

  3. Risk factors for visible lesions or positive laboratory tests in bovine tuberculosis reactor cattle in Northern Ireland.

    Science.gov (United States)

    O'Hagan, M J H; Courcier, E A; Drewe, J A; Gordon, A W; McNair, J; Abernethy, D A

    2015-07-01

    An observational case-control study was conducted to investigate risk factors for confirmed bovine tuberculosis (bTB) infection in cattle reacting positively to the single intradermal comparative cervical test (SICCT) in Northern Ireland in the years 1998, 2002 and 2006. Macroscopic lesions were detected at slaughter (positive visible lesion (VL) status) in 43.0% of reactor cattle, whilst 45.3% of those sampled were confirmed as bTB positive due to the presence of lesions or positive histopathology/mycobacterial culture (positive bTB status). In 97.5% of the reactors, the VL status and bTB status were either both negative or both positive. Generalized linear mixed model analyses were conducted on data of 24,923 reactor cattle with the variables herd identifier, local veterinary office (DVO) and abattoir being used as random effects within all the models generated at univariable and multivariable level. The other variables within the dataset were used as fixed effects. Significant risk factors associated with VL status and bTB status at multivariable level (pbovine tuberculin injection site, epidemiological status of skin test, total number of reactors at the disclosure test, mean herd size and prior response to the skin test. These risk factors are likely related to the time since infection, the strength of the challenge of infection and the susceptibility of the animal. These findings are important as the detection of visible lesions and the confirmation of bTB are an integral part of the overall bTB control programme in Northern Ireland and the veterinary meat inspection and hygiene programme. The visible lesion status and bTB status of an animal can affect the way in which bTB breakdowns are managed, since failure to detect visible lesions and recovery of Mycobacterium bovis can lead to a less stringent follow-up after other risk factors have been taken into account.

  4. Liquid Metal Thermal Electric Converter bench test module

    Energy Technology Data Exchange (ETDEWEB)

    Lukens, L.L.; Andraka, C.E.; Moreno, J.B.

    1988-04-01

    This report describes the design, fabrication, and test of a Liquid Metal Thermal Electric Converter Bench Test Module. The work presented in this document was conducted as a part of Heat Engine Task of the US Department of Energy's (DOE) Solar Thermal Technology Program. The objective of this task is the development and evaluation of heat engine technologies applicable to distributed receiver systems, in particular, dish electric systems.

  5. Cold Model Study and Commercial Test on Novel Vapor-Liquid Distributor of Hydroprocessing Reactor

    Institute of Scientific and Technical Information of China (English)

    Wang Shaobing; Zhang Zhanzhu; Wu Defei; Guo Qingming

    2007-01-01

    A novel vapor-liquid distributor was developed on the basis of sufficient study on the existing distributors applied in hydroprocessing reactors.The cold model test data showed that the fluid distribution performance of the novel vapor-liquid distributor was evidently better than the traditional one.Commercial tests of the new distributor were carried out in the 300 kt/a gas oil hydrotreating reactor at SINOPEC Changling Branch Company,showing that the new vapor-liquid distributor could improve the fluid distribution,promote the hydrotreating efficiency and lead to better performance than the traditional one.

  6. Postirradiation examination of recycle test elements from the Peach Bottom Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tiegs, T.N.; Long, E.L. Jr.

    1978-12-01

    The Recycle Test Elements were a series of tests of High-Temperature Gas-Cooled Reactor fuels irradiated in Core 2 of the Peach Bottom Unit 1 Reactor. They tested a wide variety of fissile and fertile fuel types of prime interest when the tests were designed. The fuel types included UO/sub 2/, UC/sub 2/, (2Th,U)O/sub 2/, (4Th,U)O/sub 2/, ThC/sub 2/, and ThO/sub 2/. The mixed thorium--uranium oxides and the pure thorium oxide were tested as Biso-coated particles only, while the others were tested as both Biso- and Triso-coated particles. The Biso coatings on the fissile kernels contained the fission products inadequately but on the fertile kernels they did so acceptably. The results from accelerated and real-time tests on the particle types agreed well.

  7. A generalization information management system applied to electrical distribution

    Energy Technology Data Exchange (ETDEWEB)

    Geisler, K.I.; Neumann, S.A.; Nielsen, T.D.; Bower, P.K. (Empros Systems International (US)); Hughes, B.A.

    1990-07-01

    This article presents a system solution approach that meets the requirements being imposed by industry trends and the electric utility customer. Specifically, the solution addresses electric distribution management systems. Electrical distribution management is a particularly well suited area of application because it involves a high diversity of tasks, which are currently supported by a proliferation of automated islands. Islands of automation which currently exist include (among others) distribution operations, load management, automated mapping, facility management, work order processing, and planning.

  8. Fusion Reactor and Fusion Reactor Materials:Concept Design of the ITER Test Blanket Modules

    Institute of Scientific and Technical Information of China (English)

    HUANGJinhua; LIZaixing; ZHUYukun; HUGang

    2003-01-01

    Performances required: prospect to be adopted in DEMO. Shielding for V.V. and TFC in ITER. Design principles: the peak temperature and stress should not exceed technical limits. The structure of test blanket modules (TBM) should be simple for easy fabrication, and TBM should be robust for reliability.

  9. Energy Systems Test Area (ESTA) Electrical Power Systems Test Operations: User Test Planning Guide

    Science.gov (United States)

    Salinas, Michael J.

    2012-01-01

    Test process, milestones and inputs are unknowns to first-time users of the ESTA Electrical Power Systems Test Laboratory. The User Test Planning Guide aids in establishing expectations for both NASA and non-NASA facility customers. The potential audience for this guide includes both internal and commercial spaceflight hardware/software developers. It is intended to assist their test engineering personnel in test planning and execution. Material covered includes a roadmap of the test process, roles and responsibilities of facility and user, major milestones, facility capabilities, and inputs required by the facility. Samples of deliverables, test article interfaces, and inputs necessary to define test scope, cost, and schedule are included as an appendix to the guide.

  10. ROSA-III base test series for a large break loss-of-coolant accident in a boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tasaka, K.; Abe, N.; Anoda, Y.; Koizumi, Y.; Shiba, M.

    1982-05-01

    The rig of safety assessment (ROSA)-III facility is a volumetrically scaled (1/424) boiling water reactor (BWR) system with an electrically heated core designed for integral loss-of-coolant accident (LOCA) and emergency core cooling system (ECCS) tests. It is confirmed from the experimental results obtained so far that the ROSA-III test facility can simulate major aspects of a BWR LOCA, such as boiling transition by lowering of the mixture level in the core, rewetting by the lower plenum flashing, and final quenching by the ECCS. The overall agreement between the calculated results by the RELAP5/ MOD0 code and the experimental results is good; however, the calculated lower plenum flashing rewetted the whole core and the calculated cladding temperature considerably underpredicts the measured value at the upper part of the core.

  11. Electric and Weak Electric Dipole Form Factors for Heavy Fermions in a General Two Higgs Doublet Model

    CERN Document Server

    Gómez-Dumm, D; Gomez-Dumm, Daniel

    1999-01-01

    The electric and weak electric dipole form factors for heavy fermions are calculated in the context of the most general two-Higgs-doublet model (2HDM). We find that the large top mass can produce a significant enhancement of the electric dipole form factor in the case of the b and c quarks. This effect can be used to distinguish between different 2HDM scenarios.

  12. TEMPERATURE MONITORING OPTIONS AVAILABLE AT THE IDAHO NATIONAL LABORATORY ADVANCED TEST REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    J.E. Daw; J.L. Rempe; D.L. Knudson; T. Unruh; B.M. Chase; K.L Davis

    2012-03-01

    As part of the Advanced Test Reactor National Scientific User Facility (ATR NSUF) program, the Idaho National Laboratory (INL) has developed in-house capabilities to fabricate, test, and qualify new and enhanced sensors for irradiation testing. To meet recent customer requests, an array of temperature monitoring options is now available to ATR users. The method selected is determined by test requirements and budget. Melt wires are the simplest and least expensive option for monitoring temperature. INL has recently verified the melting temperature of a collection of materials with melt temperatures ranging from 100 to 1000 C with a differential scanning calorimeter installed at INL’s High Temperature Test Laboratory (HTTL). INL encapsulates these melt wires in quartz or metal tubes. In the case of quartz tubes, multiple wires can be encapsulated in a single 1.6 mm diameter tube. The second option available to ATR users is a silicon carbide temperature monitor. The benefit of this option is that a single small monitor (typically 1 mm x 1 mm x 10 mm or 1 mm diameter x 10 mm length) can be used to detect peak irradiation temperatures ranging from 200 to 800 C. Equipment has been installed at INL’s HTTL to complete post-irradiation resistivity measurements on SiC monitors, a technique that has been found to yield the most accurate temperatures from these monitors. For instrumented tests, thermocouples may be used. In addition to Type-K and Type-N thermocouples, a High Temperature Irradiation Resistant ThermoCouple (HTIR-TC) was developed at the HTTL that contains commercially-available doped molybdenum paired with a niobium alloy thermoelements. Long duration high temperature tests, in furnaces and in the ATR and other MTRs, demonstrate that the HTIR-TC is accurate up to 1800 C and insensitive to thermal neutron interactions. Thus, degradation observed at temperatures above 1100 C with Type K and N thermocouples and decalibration due to transmutation with tungsten

  13. Advanced high-pressure bench-scale reactor for testing with hot corrosive gases

    Energy Technology Data Exchange (ETDEWEB)

    Abbasian, J.; Bachta, R.P.; Wangerow, J.R. (Inst. of Gas Technology, Chicago, IL (United States)); Mojtahedi, W.; Salo, K. (Enviropower Inc., Espoo (Finland))

    1994-01-01

    A bench-scale, high-pressure/high-temperature fluidized-bed reactor (HPTR) system is described that is capable of operating at a maximum temperature and pressure of 1,000 C and 30 bar in a corrosive atmosphere. The design of the unit is based on a double-shell balanced-pressure system. All the hot parts of the reactor that are wetted by the corrosive (and/or reactive) gases and the entire sampling line are constructed of inert material to prevent corrosion and loss of the reactant gases. The unit has been used for over 200 high-pressure hot coal gas desulfurization tests at 20 bars and up to 750 C without any experimental problem and with excellent sulfur balance, indicating that this reactor system is ideal for testing with reactive and corrosive gases at elevated pressures and temperatures.

  14. Use of research and test reactors for SPD development and calibration

    Energy Technology Data Exchange (ETDEWEB)

    LaFontaine, M.W.R. [Physics Solutions Inc., Kitchener, ON (Canada)

    2011-07-01

    Prior to using a research or test reactor for performance studies or calibration of self powered detectors, it is first necessary to fully characterize the reactor environment in the region to be utilized. This presentation details Characterization Experiments performed to quantify research/test reactor core/site parameters as they would apply for use with SPD applications. Methods will be described to: Determine the Westcott parameter, r (T {sub n}/T {sub o}) , for the region of interest; Characterize the neutron energy spectrum in terms of the cadmium absorption cut-off, i.e., consider neutrons of energy 5kT < 0.13 eV to be thermal neutrons, and neutrons of energy 5kT > 0.13 eV to be epithermal neutrons; Determine T {sub n}, the effective neutron temperature, in the region of interest; Determine the gamma flux in the region of interest; and, Establish SPD calibration standard detectors.

  15. 2011 Hyundai Sonata 3539 - Hybrid Electric Vehicle Battery Test Results

    Energy Technology Data Exchange (ETDEWEB)

    Matthew Shirk; Tyler Gray; Jeffrey Wishart

    2014-09-01

    The U.S. Department of Energy’s Advanced Vehicle Testing Activity Program consists of vehicle, battery, and infrastructure testing on advanced technology related to transportation. The activity includes tests on hybrid electric vehicles, including testing hybrid electric vehicle batteries when both the vehicles and batteries are new and at the conclusion of 160,000 miles of on-road fleet testing. This report documents battery testing performed for the 2011 Hyundai Sonata Hybrid (VIN KMHEC4A47BA003539). Battery testing was performed by Intertek Testing Services NA. The Idaho National Laboratory and Intertek collaborate on the Advanced Vehicle Testing Activity for the Vehicle Technologies Program of the U.S. Department of Energy.

  16. High Temperature Stress Analysis on 61-pin Test Assembly for Reactor Core Sub-channel Flow Test

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dongwon; Kim, Hyungmo; Lee, Hyeongyeon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    In this study, a high temperature heat transfer and stress analysis of a 61-pin test fuel assembly scaled down from the full scale 217-pin sub-assembly was conducted. The reactor core subchannel flow characteristic test will be conducted to evaluate uncertainties in computer codes used for reactor core thermal hydraulic design. Stress analysis for a 61-pin fuel assembly scaled down from Prototype Generation IV Sodium-cooled Fast Reactor was conducted and structural integrity in terms of load controlled stress limits was conducted. In this study, The evaluations on load-controlled stress limits for a 61-pin test fuel assembly to be used for reactor core subchannel flow distribution tests were conducted assuming that the test assembly is installed in a Prototype Generation IV Sodium-cooled fast reactor core. The 61-pin test assembly has the geometric similarity on P/D and H/D with PGSFR and material of fuel assembly is austenitic stainless steel 316L. The stress analysis results showed that 4.05MPa under primary load occurred at mid part of the test assembly and it was shown that the value of 4.05Mpa was far smaller than the code allowable of 127MPa. , it was shown that the stress intensity due to due to primary load is very small. The stress analysis results under primary and secondary loads showed that maximum stress intensity of 84.08MPa occurred at upper flange tangent to outer casing and the value was well within the code allowable of 268.8MPa. Integrity evaluations based on strain limits and creep-fatigue damage are underway according to the elevated design codes.

  17. Evaluating and planning the radioactive waste options for dismantling the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rule, K.; Scott, J.; Larson, S. [Princeton Plasma Physics Lab., NJ (United States)] [and others

    1995-12-31

    The Tokamak Fusion Test Reactor (TFTR) is a one-of-a kind tritium fusion research reactor, and is planned to be decommissioned within the next several years. This is the largest fusion reactor in the world and as a result of deuterium-tritum reactions is tritium contaminated and activated from 14 Mev neutrons. This presents many unusual challenges when dismantling, packaging and disposing its components and ancillary systems. Special containers are being designed to accommodate the vacuum vessel, neutral beams, and tritium delivery and processing systems. A team of experienced professionals performed a detailed field study to evaluate the requirements and appropriate methods for packaging the radioactive materials. This team focused on several current and innovative methods for waste minimization that provides the oppurtunmost cost effective manner to package and dispose of the waste. This study also produces a functional time-phased schedule which conjoins the waste volume, weight, costs and container requirements with the detailed project activity schedule for the entire project scope. This study and project will be the first demonstration of the decommissioning of a tritium fusion test reactor. The radioactive waste disposal aspects of this project are instrumental in demonstrating the viability of a fusion power reactor with regard to its environmental impact and ultimate success.

  18. Performance tests of a small hydrogen reactor based on Mg-Al pellets

    Energy Technology Data Exchange (ETDEWEB)

    Capurso, Giovanni, E-mail: giovanni.capurso@studenti.unipd.it [Dipartimento di Ingegneria Meccanica, Settore Materiali, Universita di Padova, via Marzolo 9, 35131 Padova (Italy); Agresti, Filippo [Dipartimento di Ingegneria Meccanica, Settore Materiali, Universita di Padova, via Marzolo 9, 35131 Padova (Italy); Russo, Sergio Lo [Dipartimento di Fisica and CNISM, Universita di Padova, via Marzolo 8, 35131 Padova (Italy); Maddalena, Amedeo; Principi, Giovanni [Dipartimento di Ingegneria Meccanica, Settore Materiali, Universita di Padova, via Marzolo 9, 35131 Padova (Italy); Cavallari, Andrea; Guardamagna, Cristina [ERSE s.p.a., via Rubattino 54, 20134 Milano (Italy)

    2011-09-15

    On the basis of a previously acquired experience on scaling up issues concerning the use of magnesium hydride as a base material for solid-state hydrogen storage, a small reactor was designed and tested in different operating conditions. It contains about 10 g of catalyzed magnesium hydride powder mixed with 5 wt.% aluminium powder and pressed in the form of cylindrical pellets and the heat flow is managed by means of an oil circulation system. Carbon paper is used to ensure good heat conductivity between the pellets and the inner wall of the reactor and between one pellet and another. A number of hydrogen absorption and desorption cycles at different temperatures and pressures was carried out to compare the behaviour of the small reactor with the laboratory data obtained on small amounts (fractions of grams) of powdered and pelletized samples. Data acquisition for gas flow, pressure and temperature in different positions of the reactor allow a good understanding of internal dynamics. The results in terms of hydrogen absorption/desorption kinetics and of stability to ongoing cycles are stimulating, so that the tested small reactor can be considered as a basic element for further studies and improvements.

  19. The influence of biofilm formation on electricity production from tempe wastewater using tubular membraneless microbial fuel cell reactor

    Science.gov (United States)

    Siagian, Nathania Dwi Karina; Arbianti, Rita; Utami, Tania Surya

    2017-05-01

    Microbial fuel cell (MFC) technology can be potentially developed as an alternative energy source since it can convert various substrates from renewable sources into electricity using bacteria as biocatalyst. Tempe wastewater as MFC substrate gives advantages in tempe wastewater treatment and reducing the purchasing cost of bacteria. Currently, the applications of MFCs are still limited due to the relatively low electricity production, so many studies have been conducted to improve the electricity production by MFC. This study focused on investigating the influence of biofilm formation time and the use of macromolecule as additional substrate towards electricity production from MFC system with tubular membranless reactor and tempe wastewater as substrate. This study suggested that biofilm formation on anode could improve the electricity production up to 10-folds while the use of glucose as substrate addition reduce the electricity production up to 60%. The biggest electricity output was obtained from the experiment of biofilm formation for 14 days with EPS content in biofilm 0,13 mg/cm2 where the maximum voltage and power density produced was respectively 34,81 mV and 0,26 mW/m2.

  20. Standard technical specifications General Electric plants, BWR/6. Volume 1, Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-04-01

    This report documents the results of the combined effort of the NRC and the industry to produce improved Standard Technical Specifications (STS), Revision 1 for General Electric BWR/6 Plants. The changes reflected in Revision 1 resulted from the experience gained from license amendment applications to convert to these improved STS or to adopt partial improvements to existing technical specifications. This NUREG is the result of extensive public technical meetings and discussions between the Nuclear Regulatory Commission (NRC) staff and various nuclear power plant licensees, Nuclear Steam Supply System (NSSS) Owners Groups, NSSS vendors, and the Nuclear Energy Institute (NEI). The improved STS were developed based on the criteria in the Final Commission Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, 1993. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.10 of the improved STS.

  1. The Next Generation Nuclear Plant Graphite Creep Experiment Irradiation in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Blaine Grover

    2010-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating six gas reactor graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These graphite irradiations are being accomplished to support development of the next generation reactors in the United States. The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the Next Generation Nuclear Plant (NGNP) Very High Temperature Gas Reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain six stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six stacks will have differing compressive loads applied to the top half of each pair of specimen stacks, while a seventh stack will not have a compressive load. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be the capability of sampling the sweep gas effluent to determine if any oxidation or off-gassing of the specimens occurs during initial start-up of

  2. Operation, test, research and development of the high temperature engineering test reactor (HTTR). FY1999-2001

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-05-01

    The HTTR (High Temperature Engineering Test Reactor) with the thermal power of 30 MW and the reactor outlet coolant temperature of 850/950 degC is the first high temperature gas-cooled reactor (HTGR) in Japan, which uses coated fuel particle, graphite for core components, and helium gas for primary coolant. The HTTR, which locates at the south-west area of 50,000 m{sup 2} in the Oarai Research Establishment, had been constructed since 1991 before accomplishing the first criticality on November 10, 1998. Rise to power tests of the HTTR started in September, 1999 and the rated thermal power of 30 MW and the reactor outlet coolant temperature of 850 degC was attained in December 2001. JAERI received the certificate of pre-operation test, that is, the commissioning license for the HTTR in March 2002. This report summarizes operation, tests, maintenance, radiation control, and construction of components and facilities for the HTTR as well as R and Ds on HTGRs from FY1999 to 2001. (author)

  3. Abbreviated sampling and analysis plan for planning decontamination and decommissioning at Test Reactor Area (TRA) facilities

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-10-01

    The objective is to sample and analyze for the presence of gamma emitting isotopes and hazardous constituents within certain areas of the Test Reactor Area (TRA), prior to D and D activities. The TRA is composed of three major reactor facilities and three smaller reactors built in support of programs studying the performance of reactor materials and components under high neutron flux conditions. The Materials Testing Reactor (MTR) and Engineering Test Reactor (ETR) facilities are currently pending D/D. Work consists of pre-D and D sampling of designated TRA (primarily ETR) process areas. This report addresses only a limited subset of the samples which will eventually be required to characterize MTR and ETR and plan their D and D. Sampling which is addressed in this document is intended to support planned D and D work which is funded at the present time. Biased samples, based on process knowledge and plant configuration, are to be performed. The multiple process areas which may be potentially sampled will be initially characterized by obtaining data for upstream source areas which, based on facility configuration, would affect downstream and as yet unsampled, process areas. Sampling and analysis will be conducted to determine the level of gamma emitting isotopes and hazardous constituents present in designated areas within buildings TRA-612, 642, 643, 644, 645, 647, 648, 663; and in the soils surrounding Facility TRA-611. These data will be used to plan the D and D and help determine disposition of material by D and D personnel. Both MTR and ETR facilities will eventually be decommissioned by total dismantlement so that the area can be restored to its original condition.

  4. Electric vehicle test report, Cutler-Hammer Corvette

    Energy Technology Data Exchange (ETDEWEB)

    1981-01-01

    The work described was part of the effort to characterize vehicles for the state-of-the-art assessment of electric vehicles. The vehicle evaluated was a Chevrolet Corvette converted to electric operation. The vehicle was based on a standard production 1967 chassis and body. The original internal combustion engine was replaced by an electric traction motor. Eighteen batteries supplied the electrical energy. A controller, an onboard battery charger, and several dashboard instruments completed the conversion. The remainder of the vehicle, and in particular the remainder of the drive-train (clutch, driveshaft, and differential), was stock, except for the transmission. The overall objective of the tests was to develop performance data at the system and subsystem level. The emphasis was on the electrical portion of the drive train, although some analysis and discussion of the mechanical elements are included. There was no evaluation of other aspects of the vehicle such as braking, ride, handling, passenger accomodations, etc. Included are a description of the vehicle, the tests performed and a discussion of the results. Tests were conducted both on the road (actually a mile long runway) and in a chassis dynamometer equipped laboratory. The majority of the tests performed were according to SAE Procedure J227a and included maximum effort accelerations, constant-speed range, and cyclic range. Some tests that are not a part of the SAE Procedure J227a are described and the analysis of the data from all tests is discussed. (LCL)

  5. Warm Water Oxidation Verification - Scoping and Stirred Reactor Tests

    Energy Technology Data Exchange (ETDEWEB)

    Braley, Jenifer C.; Sinkov, Sergey I.; Delegard, Calvin H.; Schmidt, Andrew J.

    2011-06-15

    Scoping tests to evaluate the effects of agitation and pH adjustment on simulant sludge agglomeration and uranium metal oxidation at {approx}95 C were performed under Test Instructions(a,b) and as per sections 5.1 and 5.2 of this Test Plan prepared by AREVA. (c) The thermal testing occurred during the week of October 4-9, 2010. The results are reported here. For this testing, two uranium-containing simulant sludge types were evaluated: (1) a full uranium-containing K West (KW) container sludge simulant consisting of nine predominant sludge components; (2) a 50:50 uranium-mole basis mixture of uraninite [U(IV)] and metaschoepite [U(VI)]. This scoping study was conducted in support of the Sludge Treatment Project (STP) Phase 2 technology evaluation for the treatment and packaging of K-Basin sludge. The STP is managed by CH2M Hill Plateau Remediation Company (CHPRC) for the U.S. Department of Energy. Warm water ({approx}95 C) oxidation of sludge, followed by immobilization, has been proposed by AREVA and is one of the alternative flowsheets being considered to convert uranium metal to UO{sub 2} and eliminate H{sub 2} generation during final sludge disposition. Preliminary assessments of warm water oxidation have been conducted, and several issues have been identified that can best be evaluated through laboratory testing. The scoping evaluation documented here was specifically focused on the issue of the potential formation of high strength sludge agglomerates at the proposed 95 C process operating temperature. Prior hydrothermal tests conducted at 185 C produced significant physiochemical changes to genuine sludge, including the formation of monolithic concretions/agglomerates that exhibited shear strengths in excess of 100 kPa (Delegard et al. 2007).

  6. Analytical Study of High Concentration PCB Paint at the Heavy Water Components Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lowry, N.J.

    1998-10-21

    This report provides results of an analytical study of high concentration PCB paint in a shutdown nuclear test reactor located at the US Department of Energy's Savannah River Site (SRS). The study was designed to obtain data relevant for an evaluation of potential hazards associated with the use of and exposure to such paints.

  7. Analysis of Removal Alternatives for the Heavy Water Components Test Reactor at the Savannah River Site

    Energy Technology Data Exchange (ETDEWEB)

    Owen, M.B. [Westinghouse Savannah River Co., Aiken, SC (United States)

    1996-08-01

    This engineering study was developed to evaluate different options for decommissioning of the Heavy Water Components Test Reactor (HWCTR) at the Savannah River Site. This document will be placed in the DOE-SRS Area reading rooms for a period of 30 days in order to obtain public input to plans for the demolition of HWCTR.

  8. Hawaiian Electric Advanced Inverter Test Plan - Result Summary

    Energy Technology Data Exchange (ETDEWEB)

    Hoke, Anderson; Nelson, Austin; Prabakar, Kumaraguru; Nagarajan, Adarsh

    2016-10-14

    This presentation is intended to share the results of lab testing of five PV inverters with the Hawaiian Electric Companies and other stakeholders and interested parties. The tests included baseline testing of advanced inverter grid support functions, as well as distribution circuit-level tests to examine the impact of the PV inverters on simulated distribution feeders using power hardware-in-the-loop (PHIL) techniques. hardware-in-the-loop (PHIL) techniques.

  9. Continuous-flow stirred-tank reactor 20-L demonstration test: Final report

    Energy Technology Data Exchange (ETDEWEB)

    Lee, D.D.; Collins, J.L.

    2000-02-01

    One of the proposed methods of removing the cesium, strontium, and transuranics from the radioactive waste storage tanks at Savannah River is the small-tank tetraphenylborate (TPB) precipitation process. A two-reactor-in-series (15-L working volume each) continuous-flow stirred-tank reactor (CSTR) system was designed, constructed, and installed in a hot cell to test the Savannah River process. The system also includes two cross-flow filtration systems to concentrate and wash the slurry produced in the process, which contains the bulk of radioactivity from the supernatant processed through the system. Installation, operational readiness reviews, and system preparation and testing were completed. The first test using the filtration systems, two CSTRs, and the slurry concentration system was conducted over a 61-h period with design removal of Cs, Sr, and U achieved. With the successful completion of Test 1a, the following tests, 1b and 1c, were not required.

  10. Digital System Reliability Test for the Evaluation of safety Critical Software of Digital Reactor Protection System

    Directory of Open Access Journals (Sweden)

    Hyun-Kook Shin

    2006-08-01

    Full Text Available A new Digital Reactor Protection System (DRPS based on VME bus Single Board Computer has been developed by KOPEC to prevent software Common Mode Failure(CMF inside digital system. The new DRPS has been proved to be an effective digital safety system to prevent CMF by Defense-in-Depth and Diversity (DID&D analysis. However, for practical use in Nuclear Power Plants, the performance test and the reliability test are essential for the digital system qualification. In this study, a single channel of DRPS prototype has been manufactured for the evaluation of DRPS capabilities. The integrated functional tests are performed and the system reliability is analyzed and tested. The results of reliability test show that the application software of DRPS has a very high reliability compared with the analog reactor protection systems.

  11. Mechanical properties of reactor pressure vessel steels studied by static and dynamic torsion tests

    Science.gov (United States)

    Munier, A.; Maamouri, M.; Schaller, R.; Mercier, O.

    1993-06-01

    Internal friction measurements and torsional plastic deformation tests have been performed in reactor pressure vessel steels (unirradiated, irradiated and irradiated/annealed specimens). The results of these experiments have been interpreted with help of transmission electron microscopy observations (conventional and in situ). It is shown how the interactions between screw dislocations and obstacles (Peierls valleys, impurities and precipitates) could explain the low temperature hardening and the irradiation embrittlement of ferritic steels. In addition, it appears that the nondestructive internal friction technique could be used advantageously to follow the evolution of the material properties under irradiation, as for instance the irradiation embrittlement of the reactor pressure vessel steels.

  12. Acoustic emission monitoring of hot functional testing: Watts Bar Unit 1 Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hutton, P.H.; Dawson, J.F.; Friesel, M.A.; Harris, J.C.; Pappas, R.A.

    1984-06-01

    Acoustic emission (AE) monitoring of selected pressure boundary areas at TVA's Watts Bar, Unit 1 Nuclear Power Plant during hot functional preservice testing is described in this report. The report deals with background, methodology, and results. The work discussed here is a major milestone in a program supported by NRC to develop and demonstrate application of AE monitoring for continuous surveillance of reactor pressure boundaries to detect and evaluate growing flaws. The subject work demonstrated that anticipated problem areas can be overcome. Work is continuing toward AE monitoring during reactor operation.

  13. Development and Testing of a High Capacity Plasma Chemical Reactor in the Ukraine

    Energy Technology Data Exchange (ETDEWEB)

    Reilly, Raymond W.

    2012-07-30

    This project, Development and Testing of a High Capacity Plasma Chemical Reactor in the Ukraine was established at the Kharkiv Institute of Physics and Technology (KIPT). The associated CRADA was established with Campbell Applied Physics (CAP) located in El Dorado Hills, California. This project extends an earlier project involving both CAP and KIPT conducted under a separate CRADA. The initial project developed the basic Plasma Chemical Reactor (PCR) for generation of ozone gas. This project built upon the technology developed in the first project, greatly enhancing the output of the PCR while also improving reliability and system control.

  14. Pilot-plant testing of magnetic filters for the N-Reactor primary cooling circuit

    Energy Technology Data Exchange (ETDEWEB)

    Emory, B.B.

    1982-01-01

    Data obtained during the laboratory loop test program using the high power HGMF indicates that removal efficiency for /sup 60/Co and subsequently the bulk of the crud, will be greater than 90% at field strength above .1 Tesla for the expanded metal mesh matrix. However, since /sup 54/Mn seems to exhibit paramagnetic behavior and the possibility of quantities of alpha iron forming during reactor shut down from oxygen inleakage, a field strength of .5 to 1 Tesla may be more appropriate for a full scale on-reactor installation. Crud loading of 50 gm per kg of matrix weight are readily obtainable and up to twice that amount has been reached.

  15. Proceedings of the 1988 International Meeting on Reduced Enrichment for Research and Test Reactors

    Energy Technology Data Exchange (ETDEWEB)

    1993-07-01

    The international effort to develop and implement new research reactor fuels utilizing low-enriched uranium, instead of highly- enriched uranium, continues to make solid progress. This effort is the cornerstone of a widely shared policy aimed at reducing, and possibly eliminating, international traffic in highly-enriched uranium and the nuclear weapon proliferation concerns associated with this traffic. To foster direct communication and exchange of ideas among the specialists in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Argonne National Laboratory, sponsored this meeting as the eleventh of a series which began 1978. Individual papers presented at the meeting have been cataloged separately.

  16. Tests of Lorentz and CPT Violation in the Medium Baseline Reactor Antineutrino Experiment

    CERN Document Server

    Li, Yu-Feng

    2014-01-01

    Tests of Lorentz and CPT violation in the medium baseline reactor antineutrino experiment are presented in the framework of the Standard Model Extension (SME). Both the spectral distortion and sidereal variation are employed to derive the limits of Lorentz violation (LV) coefficients. We do the numerical analysis of the sensitivity of LV coefficients by taking the Jiangmen Underground Neutrino Observatory (JUNO) as an illustration, which can improve the sensitivity by more than two orders of magnitude compared with the current limits from reactor antineutrino experiments.

  17. Calculation of the electric field gradients, generalized Sternheimer shielding constants, and electric-field-gradient polarizabilities for ten small molecules

    Science.gov (United States)

    Bishop, David M.; Cybulski, sławomir M.

    1994-05-01

    Electric field gradients, generalized Sternheimer shielding constants, and electric-field-gradient polarizabilities are calculated for H2, N2, F2, HF, HCl, CO, HCN, HNC, H2O, and NH3. The calculations are performed at both the Hartree-Fock and second order Møller-Plesset levels of approximation using large basis sets. For most of these molecules this is the first time that the shielding constants and electric field gradient polarizabilities have been determined. Electron correlation is generally found to be a significant factor.

  18. Measuring Test Measurement Error: A General Approach

    Science.gov (United States)

    Boyd, Donald; Lankford, Hamilton; Loeb, Susanna; Wyckoff, James

    2013-01-01

    Test-based accountability as well as value-added asessments and much experimental and quasi-experimental research in education rely on achievement tests to measure student skills and knowledge. Yet, we know little regarding fundamental properties of these tests, an important example being the extent of measurement error and its implications for…

  19. Hybrid Test Bed of Wind Electric Generator with Photovoltaic Panels

    Directory of Open Access Journals (Sweden)

    G.D.Anbarasi Jebaselvi

    2014-03-01

    Full Text Available Driven by the increasing costs of power production and decreasing fossil fuel reserves with the addition of global environmental concerns, renewable energy is now becoming significant fraction of total electricity production in the world. Advancements in the field of wind electric generator technology and power electronics help to achieve rapid progress in hybrid power system which mainly involves wind, solar and diesel energy with a good battery back-up. Here the discussion brings about the installation of real time test bed with a small electric generator and dynamic solar panels with battery backups.

  20. Reactor operation

    CERN Document Server

    Shaw, J

    2013-01-01

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

  1. Vibration test on KMRR reactor structure and primary cooling system piping

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Seung Hoh; Kim, Tae Ryong; Park, Jin Hoh; Park, Jin Suk; Ryoo, Jung Soo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-10-01

    Most equipments, piping systems and reactor structures in nuclear power plants are subjected to flow induced vibration due to high temperature and high pressure coolant flowing inside or outside of the equipments, systems and structures. Because the flow induced vibration sometimes causes significant damage to reactor structures and piping systems, it is important and necessary to evaluate the vibration effect on them and to prove their structural integrity. Korea Multipurpose Research Reactor (KMRR) being constructed by KAERI is 30 MWt pool type research reactor. Since its main structures and piping systems were designed and manufactured in accordance with the standards and guidelines for commercial nuclear power plant, it was decided to evaluate their vibratory response in accordance with the standards and guidelines for commercial NPP. The objective of this vibration test is the assessment of vibration levels of KMRR reactor structure and primary cooling piping system for their structural integrity under the steady-state or transient operating condition. 38 figs, 14 tabs, 2 refs. (Author).

  2. Proceedings of the 1990 International Meeting on Reduced Enrichment for Research and Test Reactors

    Energy Technology Data Exchange (ETDEWEB)

    1993-07-01

    The global effort to reduce, and possibly, eliminate the international traffic in highly-enriched uranium caused by its use in research reactors requires extensive cooperation and free exchange of information among all participants. To foster this free exchange of information, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Argonne National Laboratory, sponsored this meeting as the thirteenth of a series which began in 1978. The common effort brought together, past, a large number of specialists from many countries. On hundred twenty-three participants from 26 countries, including scientists, reactor operators, and personnel from commercial fuel suppliers, research centers, and government organizations, convened in Newport, Rhode Island to discuss their results, their activities, and their plans relative to converting research reactors to low-enriched fuels. As more and more reactors convert to the use of low-enriched uranium, the emphasis of our effort has begun to shift from research and development to tasks more directly related to implementation of the new fuels and technologies that have been developed, and to refinements of those fuels and technologies. It is appropriate, for this reason, that the emphasis of this meeting was placed on safety and on conversion experiences. This individual papers in this report have been cataloged separately.

  3. Evaluation of the Shielding Characteristics Test around the Reactor Core in the Prototype Fbr Monju

    Science.gov (United States)

    Usami, Shin; Suzuoki, Zenro; Deshimaru, Takehide; Nakashima, Fumiaki; Hikichi, Takuo

    2003-06-01

    In Monju, shielding measurements were made around the reactor core as a part of the system start-up tests in order to evaluate the design margins of the shielding performance, to demonstrate the validity of the shielding analysis method, and to acquire basic data for use in future FBR design. The measured reaction rates have been obtained radially from the core to the in-vessel storage rack and axially to the reactor vessel upper plenum. The measured values (E) were compared with the calculated values (C) obtained with the FBR shielding analysis system on the basis of the nuclear data library JENDL-3.2. Based upon these results, the design margins around the reactor core have been examined.

  4. Two-Compartment Photoelectrochemical Reactors Tested under various solar Light Concentration ratios

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez-Ibanez, P.; Malato, S. [Plataforma Solar de Almeria. CIEMAT (Spain)

    1999-07-01

    A new type of photo reactor, made of a cylindrical photo anode placed around an inner comportment, has been adapted to the compound parabolic (CPC's) and to the parabolic trough (PTC, Helioman) solar collectors. The photoelectrochemical performances of such two-compartment photo reactors are noticeably improved with respect to those previously obtained with photo reactors having flat photoanodes and a single compartment. The abatement of model pollutants shows up to threefold higher organic oxidation rates compared to Ti O{sub 2} slurries tested in the same experimental conditions. Clearly, charge separation is much better when an external electrochemical bias is applied to Ti/Ti O{sub 2} photoanodes under irradiation. (Author) 8 refs.

  5. Monitoring and Analysis of In-Pile Phenomena in Advanced Test Reactor using Acoustic Telemetry

    Energy Technology Data Exchange (ETDEWEB)

    Agarwal, Vivek [Idaho National Lab. (INL), Idaho Falls, ID (United States). Dept. of Human Factors, Controls, and Statistics; Smith, James A. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Dept. of Fuel Performance and Design; Jewell, James Keith [Idaho National Lab. (INL), Idaho Falls, ID (United States). Dept. of Fuel Performance and Design

    2015-02-01

    The interior of a nuclear reactor presents a particularly harsh and challenging environment for both sensors and telemetry due to high temperatures and high fluxes of energetic and ionizing particles among the radioactive decay products. A number of research programs are developing acoustic-based sensing approach to take advantage of the acoustic transmission properties of reactor cores. Idaho National Laboratory has installed vibroacoustic receivers on and around the Advanced Test Reactor (ATR) containment vessel to take advantage of acoustically telemetered sensors such as thermoacoustic (TAC) transducers. The installation represents the first step in developing an acoustic telemetry infrastructure. This paper presents the theory of TAC, application of installed vibroacoustic receivers in monitoring the in-pile phenomena inside the ATR, and preliminary data processing results.

  6. 2011 Hyundai Sonata 4932 - Hybrid Electric Vehicle Battery Test Results

    Energy Technology Data Exchange (ETDEWEB)

    Tyler Gray; Matthew Shirk; Jeffrey Wishart

    2013-07-01

    The U.S. Department of Energy Advanced Vehicle Testing Activity Program consists of vehicle, battery, and infrastructure testing on advanced technology related to transportation. The activity includes tests on hybrid electric vehicles (HEVs), including testing the HEV batteries when both the vehicles and batteries are new and at the conclusion of 160,000 miles of on-road fleet testing. This report documents battery testing performed for the 2011 Hyundai Sonata Hybrid HEV (VIN KMHEC4A43BA004932). Battery testing was performed by the Electric Transportation Engineering Corporation dba ECOtality North America. The Idaho National Laboratory and ECOtality North America collaborate on the AVTA for the Vehicle Technologies Program of the DOE.

  7. 78 FR 63516 - Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors

    Science.gov (United States)

    2013-10-24

    ... COMMISSION Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors AGENCY... Cooling Systems for New Boiling-Water Reactors.'' This RG describes testing methods the NRC staff considers acceptable for demonstrating the operability of emergency core cooling systems (ECCSs) for boiling...

  8. Submersible microbial desalination cell for simultaneous ammonia recovery and electricity production from anaerobic reactors containing high levels of ammonia

    DEFF Research Database (Denmark)

    Zhang, Yifeng; Angelidaki, Irini

    2015-01-01

    High ammonia concentration in anaerobic reactors can seriously inhibit the anaerobic digestion process. In this study, a submersible microbial desalination cell (SMDC) was developed as an innovative method to lower the ammonia level in a continuous stirred tank reactor (CSTR) by in situ ammonia...... recovery and electricity production. In batch experiment, the ammonia concentration in the CSTR decreased from 6 to 0.7g-N/L during 30days, resulting in an average recovery rate of 80g-N/m2/d. Meanwhile, a maximum power density of 0.71±0.5W/m2 was generated at 2.85A/m2. Both current driven NH4+ migration...

  9. Improvement of Efficiency and Electrical Energy Consumption of AB74 Degradation Process using A Novel Cylindrical Batch Photochemical Reactor.

    Directory of Open Access Journals (Sweden)

    Mohammad Ebrahim Olya

    2016-06-01

    Full Text Available Optimization of electric energy consumption in degradation of Indigo Carmine based on a new cylindrical batch suspension photocatalytic reactor designing concerning the main engineering parameters such as mixing, reaction kinetic, mass transfer, catalyst illumination, and inherent potential of contaminant was done. The most degradation efficiency was observed within in air flow rate 1.2L.min-1, initial hydrogen peroxide concentration 0.0294mol.L-1, pH between 2-3, initial dye concentration 10mg.L-1, and initial TiO2 dose 2g.L-1. Applying optimum condition eventuated 76.09% TOC reduction after 60 min reaction time. The results show that designed reactor is extremely affective in reduction of operational cost.

  10. Continuous preparation of carbon-nanotube-supported platinum catalysts in a flow reactor directly heated by electric current

    Directory of Open Access Journals (Sweden)

    Alicja Schlange

    2011-10-01

    Full Text Available In this contribution we present for the first time a continuous process for the production of highly active Pt catalysts supported by carbon nanotubes by use of an electrically heated tubular reactor. The synthesized catalysts show a high degree of dispersion and narrow distributions of cluster sizes. In comparison to catalysts synthesized by the conventional oil-bath method a significantly higher electrocatalytic activity was reached, which can be attributed to the higher metal loading and smaller and more uniformly distributed Pt particles on the carbon support. Our approach introduces a simple, time-saving and cost-efficient method for fuel cell catalyst preparation in a flow reactor which could be used at a large scale.

  11. FeCrO Nanoparticles as Anode Catalyst for Ethane Proton Conducting Fuel Cell Reactors to Coproduce Ethylene and Electricity

    Directory of Open Access Journals (Sweden)

    Jian-Hui Li

    2011-01-01

    Full Text Available Ethylene and electrical power are cogenerated in fuel cell reactors with FeCr2O4 nanoparticles as anode catalyst, La0.7Sr0.3FeO3- (LSF as cathode material, and BaCe0.7Zr0.1Y0.2O3- (BCZY perovskite oxide as proton-conducting ceramic electrolyte. FeCr2O4, BCZY and LSF are synthesized by a sol-gel combustion method. The power density increases from 70 to 240 mW cm−2, and the ethylene yield increases from about 14.1% to 39.7% when the operating temperature of the proton-conducting fuel cell reactor increases from 650∘C to 750∘C. The FeCr2O4 anode catalyst exhibits better catalytic performance than nanosized Cr2O3 anode catalyst.

  12. Stand for testing the electrical race car engine

    Science.gov (United States)

    Baier, M.; Franiasz, J.; Mierzwa, P.; Wylenzek, D.

    2015-11-01

    An engine test stand created especially for research of electrical race car is described in the paper. The car is an aim of Silesian Greenpower project whose participants build and test electrical vehicles to take part in international races in Great Britain. The engine test stand is used to test and measure the characteristics of vehicles and their engines. It has been designed particularly to test the electric cars engineered by students of Silesian Greenpower project. The article contains a description how the test stand works and shows its versatility in many areas. The paper presents both construction of the test stand, control system and sample results of conducted research. The engine test stand was designed and modified using PLM Siemens NX 8.5. The construction of the test stand is highly modular, which means it can be used both for testing the vehicle itself or for tests without the vehicle. The test stand has its own wheel, motor, powertrain and braking system with second engine. Such solution enables verifying various concepts without changing the construction of the vehicle. The control system and measurement system are realized by enabling National Instruments product myRIO (RIO - Reconfigurable Input/Output). This controller in combination with powerful LabVIEW environment performs as an advanced tool to control torque and speed simultaneously. It is crucial as far as the test stand is equipped in two motors - the one being tested and the braking one. The feedback loop is realized by an optical encoder cooperating with the rotor mounted on the wheel. The results of tests are shown live on the screen both as a chart and as single values. After performing several tests there is a report generated. The engine test stand is widely used during process of the Silesian Greenpower vehicle design. Its versatility enables powertrain testing, wheels and tires tests, thermal analysis and more.

  13. State Waste Discharge Permit Application: Electric resistance tomography testing

    Energy Technology Data Exchange (ETDEWEB)

    1994-04-01

    This permit application documentation is for a State Waste Discharge Permit issued in accordance with requirements of Washington Administrative Code 173-216. The activity being permitted is a technology test using electrical resistance tomography. The electrical resistance tomography technology was developed at Lawrence Livermore National Laboratory and has been used at other waste sites to track underground contamination plumes. The electrical resistance tomography technology measures soil electrical resistance between two electrodes. If a fluid contaminated with electrolytes is introduced into the soil, the soil resistance is expected to drop. By using an array of measurement electrodes in several boreholes, the areal extent of contamination can be estimated. At the Hanford Site, the purpose of the testing is to determine if the electrical resistance tomography technology can be used in the vicinity of large underground metal tanks without the metal tank interfering with the test. It is anticipated that the electrical resistance tomography technology will provide a method for accurately detecting leaks from the bottom of underground tanks, such as the Hanford Site single-shell tanks.

  14. The RES reactor, a test reactor for naval propulsion; Le reacteur d'essais RES, reacteur d'essais de la propulsion navale

    Energy Technology Data Exchange (ETDEWEB)

    Pivet, S. [CEA Bruyeres-le-Chatel, 91 (France); Minguet, J.L. [AREVA-Technicatome, 13 - Aix en Provence (France)

    2005-07-01

    The RES, the new test reactor for naval propulsion, will replace the RNG that nears the end of its operating life after 30 years in service. The main asset of a land-based installation is to provide an in-core instrumented reactor while the on-board system must stay as simple as possible for robustness reasons. The objective of the RES is fivefold: 1) to foresee and help solving problems likely to happen on on-board reactor, 2) to validate nuclear fuels and reactor systems for naval propulsion, 3) to validate reactor system and equipment for the Barracuda submarine program, 4) to upgrade the on-ground facility located at Cadarache, and 5) to provide the Cea with a new capacity for the storing of spent fuels from naval propulsion systems and from Cea research reactors. The RES facility is made of 2 parts: one that houses the reactor and the other that is dedicated to the handling on spent fuels, their examination through a gamma spectrometry bench and their storing in a pool. The RES facility is scheduled to open in 2009. (A.C.)

  15. Preliminary Benchmark Evaluation of Japan’s High Temperature Engineering Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    John Darrell Bess

    2009-05-01

    A benchmark model of the initial fully-loaded start-up core critical of Japan’s High Temperature Engineering Test Reactor (HTTR) was developed to provide data in support of ongoing validation efforts of the Very High Temperature Reactor Program using publicly available resources. The HTTR is a 30 MWt test reactor utilizing graphite moderation, helium coolant, and prismatic TRISO fuel. The benchmark was modeled using MCNP5 with various neutron cross-section libraries. An uncertainty evaluation was performed by perturbing the benchmark model and comparing the resultant eigenvalues. The calculated eigenvalues are approximately 2-3% greater than expected with an uncertainty of ±0.70%. The primary sources of uncertainty are the impurities in the core and reflector graphite. The release of additional HTTR data could effectively reduce the benchmark model uncertainties and bias. Sensitivity of the results to the graphite impurity content might imply that further evaluation of the graphite content could significantly improve calculated results. Proper characterization of graphite for future Next Generation Nuclear Power reactor designs will improve computational modeling capabilities. Current benchmarking activities include evaluation of the annular HTTR cores and assessment of the remaining start-up core physics experiments, including reactivity effects, reactivity coefficient, and reaction-rate distribution measurements. Long term benchmarking goals might include analyses of the hot zero-power critical, rise-to-power tests, and other irradiation, safety, and technical evaluations performed with the HTTR.

  16. Advanced gas cooled nuclear reactor materials evaluation and development program. Progress report, July 1--September 30, 1978

    Energy Technology Data Exchange (ETDEWEB)

    1978-11-24

    Results of work performed from July 1, 1978 through September 30, 1978 on the Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program are presented. Candidate alloys were evaluated for Very High Temperature Reactor Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the affect of simulated reactor primary coolant (Helium containing small amounts of various other gases), the high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. The activities associated with the characterization of the materials for the screening test program are reported, i.e., test specimen preparation, information from the materials characterization tests performed by General Electric, and the status of the simulated reactor helium supply system, testing equipment, and gas chemistry analysis instrumentation and equipment. The status of the data management system is presented.

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

    Energy Technology Data Exchange (ETDEWEB)

    1977-06-01

    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.

  18. The Chlorella killed by pulsed electrical discharge in liquid with two different reactors

    Science.gov (United States)

    Gao, Z. Y.; Sun, B.; Yan, Z. Y.; Zhu, X. M.; Liu, H.; Song, Y. J.; Sato, M.

    2013-03-01

    The application of pulsed high-voltage discharge in liquid has attracted wide attention as an effective water treatment. In this paper, two different liquid high-voltage discharge systems were constructed with plate-hole-plate and needle-plate electrode structures, and the inactivation behaviors of Chlorella were studied in the two reactors. The results show that the killing rates of algae in both reactors all increased significantly with increasing discharge voltage and the killing rates were intensely related to discharge power, instantaneous power and single pulse input energy. Furthermore, the inactivation effect in needle-plate reactor was superior to that in plate-hole-plate reactor under the same experimental conditions.

  19. Effects of an electric field on white sharks: in situ testing of an electric deterrent.

    Directory of Open Access Journals (Sweden)

    Charlie Huveneers

    Full Text Available Elasmobranchs can detect minute electromagnetic fields, <1 nV cm(-1, using their ampullae of Lorenzini. Behavioural responses to electric fields have been investigated in various species, sometimes with the aim to develop shark deterrents to improve human safety. The present study tested the effects of the Shark Shield Freedom7™ electric deterrent on (1 the behaviour of 18 white sharks (Carcharodon carcharias near a static bait, and (2 the rates of attacks on a towed seal decoy. In the first experiment, 116 trials using a static bait were performed at the Neptune Islands, South Australia. The proportion of baits taken during static bait trials was not affected by the electric field. The electric field, however, increased the time it took them to consume the bait, the number of interactions per approach, and decreased the proportion of interactions within two metres of the field source. The effect of the electric field was not uniform across all sharks. In the second experiment, 189 tows using a seal decoy were conducted near Seal Island, South Africa. No breaches and only two surface interactions were observed during the tows when the electric field was activated, compared with 16 breaches and 27 surface interactions without the electric field. The present study suggests that the behavioural response of white sharks and the level of risk reduction resulting from the electric field is contextually specific, and depends on the motivational state of sharks.

  20. Ultrahigh temperature vapor core reactor-MHD system for space nuclear electric power

    Science.gov (United States)

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

    1991-01-01

    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. Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production

    Energy Technology Data Exchange (ETDEWEB)

    Philip MacDonald; Jacopo Buongiorno; James Sterbentz; Cliff Davis; Robert Witt; Gary Was; J. McKinley; S. Teysseyre; Luca Oriani; Vefa Kucukboyaci; Lawrence Conway; N. Jonsson: Bin Liu

    2005-02-13

    The supercritical water reactor (SCWR) has been the object of interest throughout the nuclear Generation IV community because of its high potential: a simple, direct cycle, compact configuration; elimination of many traditional LWR components, operation at coolant temperatures much higher than traditional LWRs and thus high thermal efficiency. It could be said that the SWR was viewed as the water counterpart to the high temperature gas reactor.

  2. [A literature analysis of power frequency electric field testing data].

    Science.gov (United States)

    Zhang, Suli; Guo, Zehua; Yu, Xintian; Ding, Yan; Zhu, Zhiliang

    2015-06-01

    To analyze the literature on power frequency electric field testing data and to propose views and suggestions for current testing. The literature on power frequency electric field testing data published in the previous years was searched to identify 306 articles involving 193 valid testing data. Mann-Whitney test and Wilcoxon W test were used for analyzing the testing data. The classification of data was carried out according to one quarter of occupational exposure limit (1.25 kV/m), one half of the exposure limit (2.5 kV/m), and the exposure limit (5 kV/m). The structure of testing data showed a significant difference between the non-power facility group and the power facility group (Pelectric field is extensive. However, the power frequency electric field testing data in actual workplaces except high-voltage power facilities are far less than the occupational exposure limit with little harmfulness. There is a phenomenon of excessive testing at present.

  3. The 5th surveillance testing for Kori unit 2 reactor vessel materials

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwon Jae; Gong, Un Sik; Chang, Jong Hwa; Joo, Yong Sun; Ahn, Sang Bok; Hong, Joon Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2001-03-01

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 5th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Kori site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Kori unit 2 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules V, R, P, T and N are 2.837E+18, 1.105E+19, 2.110E+19, 3.705E+19 and 4.831E+19n/cm{sup 2}, respectively. The bias factor, the ratio of measurement/calculation, was 0.918 for the 1st through 5th testing and the calculational uncertainty, 11.6% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.898E+19n/cm{sup 2} based on the end of 15th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 4.203E+19, 5.232E+19, 6.262E+19 and 7.291E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Kori unit 2 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 49 refs., 35 figs., 48 tabs. (Author)

  4. The 4th surveillance testing for Kori unit 3 reactor vessel materials

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwun Jae; Gong, Un Sik; Chang, Jong Hwa; Joo, Yong Sun; Ahn, Sang Bok; Hong, Joon Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-10-01

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 4th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Kori site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Kori unit 3 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X and W are 4.983E+18, 1.641E+19, 3.158E+19, and 4.469E+19n/cm{sup 2}, respectively. The bias factor, the ratio of calculation/measurement, was 0.840 for the 1st through 4th testing and the calculational uncertainty, 12% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.362E+19n/cm{sup 2} based on the end of 12th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 3.481E+19, 4.209E+19, 5.144E+19 and 5.974E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Kori unit 3 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 48 refs., 35 figs., 41 tabs. (Author)

  5. Gravitational tests of the generalized uncertainty principle

    Energy Technology Data Exchange (ETDEWEB)

    Scardigli, Fabio [American University of the Middle East, Department of Mathematics, College of Engineering, P.O. Box 220, Dasman (Kuwait); Politecnico di Milano, Dipartimento di Matematica, Milan (Italy); Casadio, Roberto [Alma Mater Universita di Bologna, Dipartimento di Fisica e Astronomia, Bologna (Italy); INFN, Sezione di Bologna, Bologna (Italy)

    2015-09-15

    We compute the corrections to the Schwarzschild metric necessary to reproduce the Hawking temperature derived from a generalized uncertainty principle (GUP), so that the GUP deformation parameter is directly linked to the deformation of the metric. Using this modified Schwarzschild metric, we compute corrections to the standard general relativistic predictions for the light deflection and perihelion precession, both for planets in the solar system and for binary pulsars. This analysis allows us to set bounds for the GUP deformation parameter from well-known astronomical measurements. (orig.)

  6. 77 FR 4203 - Energy Conservation Program: Test Procedures for General Service Fluorescent Lamps, General...

    Science.gov (United States)

    2012-01-27

    ... Fluorescent Lamps, General Service Incandescent Lamps, and Incandescent Reflector Lamps AGENCY: Office of... the test procedures for general service fluorescent lamps (GSFLs), general service incandescent lamps (GSILs), and incandescent reflector lamps (IRLs). That proposed rulemaking serves as the basis for...

  7. Development and testing of a recorder and controller for a microalgae culture reactor; Desarrollo y prueba de un registrador y controlador para un reactor de cultivo de microalgas

    Energy Technology Data Exchange (ETDEWEB)

    Esquivel, Wilson; Reyes, Jose Fernando; Bruijn, Johannes; Hernandez, Alejandro [Universidad de Concepcion, Chilan (Chile). Facultad de Ingenieria Agricola. Dept. de Mecanizacion y Energia], Emails: wesquive@udec.cl., jreyes@udec.cl., jdebruij@udec.cl., alehernandez@udec.cl

    2010-07-01

    An electronic system to monitor and control operational variables in a Raceway type of reactor for the culture of the Scenedesmus spinosus microalgae and later production for biodiesel and mitigating CO{sub 2} was developed and tested. The electronic system is constituted by a micro controller, a card reader SD, a card SD, a real-time clock, a power supply, a screen GLCD, a keyboard and a card for data acquisition, all implemented for 4-20 mA and 0-5 V output sensors. Temperature, pH, electrical conductivity, dissolved oxygen and solar radiation were measured digitalized and saved every 10 minutes. These variables were digitalized and kept in the SD memory every 10 minutes. It was determined that the most favorable conditions for the proliferation of the culture are near pH neutral and a temperature of 30 deg C, existing a strong correlation between pH and the dissolved CO{sub 2} level. Using the digital outputs of temperature and pH of the microcontroller, the CO{sub 2} injection and the elimination of O{sub 2} were controlled to maintain an adequate environment for the development of the culture. (author)

  8. Thick Films acoustic sensors devoted to MTR environment measurements. Thick Films acoustic sensors devoted to Material Testing Reactor environment measurements

    Energy Technology Data Exchange (ETDEWEB)

    Very, F.; Rosenkrantz, E.; Combette, P.; Ferrandis, J.Y. [University Montpellier, IES, UMR 5214, F-34000, Montpellier (France); CNRS, IES, UMR 5214, F-34000, Montpellier (France); Fourmentel, D.; Destouches, C.; Villard, J.F. [CEA, DEN, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St Paul lez Durance (France)

    2015-07-01

    The development of advanced instrumentation for in-pile experiments in Material Testing Reactor constitutes a main goal for the improvement of the nuclear fuel behavior knowledge. An acoustic method for fission gas release detection was tested with success during a first experiment called REMORA 3 in 2010 and 2011, and the results were used to differentiate helium and fission gas release kinetics under transient operating conditions. This experiment was lead at OSIRIS reactor (CEA Saclay, France). The maximal temperature on the sensor during the irradiation was about 150 deg. C. In this paper we present a thick film transducer produce by screen printing process. The screen printing of piezoelectric offers a wide range of possible applications for the development of acoustic sensors and piezoelectric structure for measurements in high temperature environment. We firstly produced a Lead Zirconate Titanate (PZT) based paste composed of Pz27 powder from Ferroperm, CF7575 glass, and organic solvent ESL 400. Likewise a Bismuth Titanate based paste synthesized in our laboratory was produced. With these inks we produced thick film up to 130 μm by screen printing process. Material properties characterizations of these thick-film resonators are essential for device design and applications. The piezoelectric coefficients d33 and pyro-electric P(T) coefficient are investigated. The highest P(T) and d33 are respectively 80 μC.m{sup -2}.K{sup -1} and 130 μC.N{sup -1} for the PZT transducer -which validates the fabrication process-. In view of the development of this transducer oriented for high temperature and irradiation environment, we investigated the electrical properties of the transducers for different ranges of frequencies and temperature - from 20 Hz up to 40 MHz between 30 and 400 deg. C. We highlight the evolution of the impedance response and piezoelectric parameters of screen printed piezoelectric structures on alumina. Shortly an irradiation will be realized in

  9. Electrical test prediction using hybrid metrology and machine learning

    Science.gov (United States)

    Breton, Mary; Chao, Robin; Muthinti, Gangadhara Raja; de la Peña, Abraham A.; Simon, Jacques; Cepler, Aron J.; Sendelbach, Matthew; Gaudiello, John; Emans, Susan; Shifrin, Michael; Etzioni, Yoav; Urenski, Ronen; Lee, Wei Ti

    2017-03-01

    Electrical test measurement in the back-end of line (BEOL) is crucial for wafer and die sorting as well as comparing intended process splits. Any in-line, nondestructive technique in the process flow to accurately predict these measurements can significantly improve mean-time-to-detect (MTTD) of defects and improve cycle times for yield and process learning. Measuring after BEOL metallization is commonly done for process control and learning, particularly with scatterometry (also called OCD (Optical Critical Dimension)), which can solve for multiple profile parameters such as metal line height or sidewall angle and does so within patterned regions. This gives scatterometry an advantage over inline microscopy-based techniques, which provide top-down information, since such techniques can be insensitive to sidewall variations hidden under the metal fill of the trench. But when faced with correlation to electrical test measurements that are specific to the BEOL processing, both techniques face the additional challenge of sampling. Microscopy-based techniques are sampling-limited by their small probe size, while scatterometry is traditionally limited (for microprocessors) to scribe targets that mimic device ground rules but are not necessarily designed to be electrically testable. A solution to this sampling challenge lies in a fast reference-based machine learning capability that allows for OCD measurement directly of the electrically-testable structures, even when they are not OCD-compatible. By incorporating such direct OCD measurements, correlation to, and therefore prediction of, resistance of BEOL electrical test structures is significantly improved. Improvements in prediction capability for multiple types of in-die electrically-testable device structures is demonstrated. To further improve the quality of the prediction of the electrical resistance measurements, hybrid metrology using the OCD measurements as well as X-ray metrology (XRF) is used. Hybrid metrology

  10. Evaluation program for secondary spacecraft cells: Initial evaluation tests of General Electric Company 40.0 ampere hour nickel cadmium spacecraft cells for the tracking data relay satellite system

    Science.gov (United States)

    Hall, S. W.

    1980-01-01

    Average end of charge voltages and pressures, and capacity output in ampere hours are presented. Test limits specify those values at which a cell is to be terminated from charge or discharge. Requirements are based on past cell performance data. The requirement does not constitute a limit for discontinuance from testing. The nickel cadmium batteries were screened for internal shorts, low capacity, electrolyte leakage, or inability of any cell to recover its open circuit voltage above 1.150 volts during the internal short test.

  11. External fuel thermionic reactor system.

    Science.gov (United States)

    Mondt, J. F.; Peelgren, M. L.

    1971-01-01

    Thermionic reactors are prime candidates for nuclear electric propulsion. The national thermionic reactor effort is concentrated on the flashlight concept with the external-fuel concept as the backup. The external-fuel concept is very adaptable to a completely modular power subsystem which is attractive for highly reliable long-life applications. The 20- to 25-cm long, externally-fueled converters have been designed, fabricated, and successfully tested with many thermal cycles by electrical heating. However, difficulties have been encountered during encapsulation for nuclear heated tests and none have been started to date. These nuclear tests are required to demonstrate the concept feasibility.

  12. Nuclear safety considerations in the conceptual design of a fast reactor for space electric power and propulsion

    Science.gov (United States)

    Hsieh, T.-M.; Koenig, D. R.

    1977-01-01

    Some nuclear safety aspects of a 3.2 mWt heat pipe cooled fast reactor with out-of-core thermionic converters are discussed. Safety related characteristics of the design including a thin layer of B4C surrounding the core, the use of heat pipes and BeO reflector assembly, the elimination of fuel element bowing, etc., are highlighted. Potential supercriticality hazards and countermeasures are considered. Impacts of some safety guidelines of space transportation system are also briefly discussed, since the currently developing space shuttle would be used as the primary launch vehicle for the nuclear electric propulsion spacecraft.

  13. Generalization of the Analytical Exponential Model for Homogeneous Reactor Kinetics Equations

    Directory of Open Access Journals (Sweden)

    Abdallah A. Nahla

    2012-01-01

    Full Text Available Mathematical form for two energy groups of three-dimensional homogeneous reactor kinetics equations and average one group of the precursor concentration of delayed neutrons is presented. This mathematical form is called “two energy groups of the point kinetics equations.” We rewrite two energy groups of the point kinetics equations in the matrix form. Generalization of the analytical exponential model (GAEM is developed for solving two energy groups of the point kinetics equations. The GAEM is based on the eigenvalues and the corresponding eigenvectors of the coefficient matrix. The eigenvalues of the coefficient matrix are calculated numerically using visual FORTRAN code, based on Laguerre’s method, to calculate the roots of an algebraic equation with real coefficients. The eigenvectors of the coefficient matrix are calculated analytically. The results of the GAEM are compared with the traditional methods. These comparisons substantiate the accuracy of the results of the GAEM. In addition, the GAEM is faster than the traditional methods.

  14. Review of Byzantine General Problems in the Reactor Protection System of Korean Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Eungse; Kim, Yungoo [Korea Hydro and Nuclear Power Co. Ltd. Central Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    This kind of complex error scenario is known as a Byzantine General Problem and the error is called a Byzantine Error (BE). The BE is now considered as one of the plausible common-cause failure in the nuclear power plant's (NPP) computer systems. The Reactor Protection System (RPS) in the Korean NPP consists of multiple redundant digital computer systems to increase system availability and redundancy. This system architecture is inherited form the well proved analog system's architecture. Failure modes and effects on the RPS system functions are reviewed when a BE assumed in the system's decision making logic path. For this review, one channel and two channels of BE problems in the RPS trip function are considered. If a BE occurs in any one channel of the RPS, the systems trip function has no harm affects from the BE. If two BEs occur in any channels of the RPS, the systems trip function may or may not work properly. This BE review method can be applied to other decision-making parts of the protection system in NPP.

  15. Validation of RELAP5 model of experimental test rig simulating the natural convection in MTR research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Khedr, A.; Abdel-Latif, Salwa H. [Nuclear and Radiological Regulatory Authority, Cairo (Egypt); Abdel-Hadi, Eed A. [Benha Univ., Cairo (Egypt). Shobra Faculty of Engineering; D' Auria, F. [Pisa Univ. (Italy)

    2016-03-15

    In an attempt to understand the built-up of natural circulation in MTR pool type upward flow research reactors after loss of power, an experimental test rig was built to simulate the loop of natural circulation in MTR reactors. The test rig consisting of two vertically oriented branches, in one of them the core is simulated by two rectangular, electrically heated, parallel channels. The other branch simulates the part of the return pipe that participates in the development of core natural circulation. In the first phase of the work, many experimental runs at different conditions of channel's power and branch's initial temperatures are performed. The channel's coolant and surface temperatures were measured. The measurements and their interpretation were published by the first three authors. In the present work the thermal hydraulic behavior of the test rig is complemented by theoretical analysis using RELAP5 Mod 3.3 system code. The analysis consisting of two parts; in the first part RELAP5 model is validated against the measured values and in the second part some of the other not measured hydraulic parameters are predicted and analyzed. The test rig is typically nodalized and an input dick is prepared. In spite of the low pressure of the test rig, the results show that RELAP5 qualitatively predicts the thermal hydraulic behaviour and the accompanied phenomenon of flow inversion of such facilities. Quantitatively, there is a difference between the predicted and measured values especially the channel's surface temperature. This difference may be return to the uncertainties in initial conditions of experimental runs, the position of the thermocouples which buried inside the heat structure, and the heat transfer package in RELAP5.

  16. Light Water Reactor Sustainability Program Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor Demonstration Case Study

    Energy Technology Data Exchange (ETDEWEB)

    Curtis Smith; David Schwieder; Cherie Phelan; Anh Bui; Paul Bayless

    2012-08-01

    Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about LWR design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the RISMC Pathway R&D is to support plant decisions for risk-informed margins management with the aim to improve economics, reliability, and sustain safety of current NPPs. Goals of the RISMC Pathway are twofold: (1) Develop and demonstrate a risk-assessment method coupled to safety margin quantification that can be used by NPP decision makers as part of their margin recovery strategies. (2) Create an advanced “RISMC toolkit” that enables more accurate representation of NPP safety margin. This report describes the RISMC methodology demonstration where the Advanced Test Reactor (ATR) was used as a test-bed for purposes of determining safety margins. As part of the demonstration, we describe how both the thermal-hydraulics and probabilistic safety calculations are integrated and used to quantify margin management strategies.

  17. Design, Testing and Modeling of the Direct Reactor Auxiliary Cooling System for AHTRs

    Energy Technology Data Exchange (ETDEWEB)

    Lv, Quiping [The Ohio State Univ., Columbus, OH (United States); Sun, Xiaodong [The Ohio State Univ., Columbus, OH (United States); Chtistensen, Richard [The Ohio State Univ., Columbus, OH (United States); Blue, Thomas [The Ohio State Univ., Columbus, OH (United States); Yoder, Graydon [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wilson, Dane [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-05-08

    The principal objective of this research is to test and model the heat transfer performance and reliability of the Direct Reactor Auxiliary Cooling System (DRACS) for AHTRs. In addition, component testing of fluidic diodes is to be performed to examine the performance and viability of several existing fluidic diode designs. An extensive database related to the thermal performance of the heat exchangers involved will be obtained, which will be used to benchmark a computer code for the DRACS design and to evaluate and improve, if needed, existing heat transfer models of interest. The database will also be valuable for assessing the viability of the DRACS concept and benchmarking any related computer codes in the future. The experience of making a liquid fluoride salt test facility available, with lessons learned, will greatly benefit the development of the Fluoride Salt-cooled High-temperature Reactor (FHR) and eventually the AHTR programs.

  18. Battery Test Manual For Plug-In Hybrid Electric Vehicles

    Energy Technology Data Exchange (ETDEWEB)

    Jeffrey R. Belt

    2010-09-01

    This battery test procedure manual was prepared for the United States Department of Energy (DOE), Office of Energy Efficiency and Renewable Energy (EERE), Vehicle Technologies Program. It is based on technical targets established for energy storage development projects aimed at meeting system level DOE goals for Plug-in Hybrid Electric Vehicles (PHEV). The specific procedures defined in this manual support the performance and life characterization of advanced battery devices under development for PHEV’s. However, it does share some methods described in the previously published battery test manual for power-assist hybrid electric vehicles. Due to the complexity of some of the procedures and supporting analysis, a revision including some modifications and clarifications of these procedures is expected. As in previous battery and capacitor test manuals, this version of the manual defines testing methods for full-size battery systems, along with provisions for scaling these tests for modules, cells or other subscale level devices.

  19. Battery Test Manual For Plug-In Hybrid Electric Vehicles

    Energy Technology Data Exchange (ETDEWEB)

    Jeffrey R. Belt

    2010-12-01

    This battery test procedure manual was prepared for the United States Department of Energy (DOE), Office of Energy Efficiency and Renewable Energy (EERE), Vehicle Technologies Program. It is based on technical targets established for energy storage development projects aimed at meeting system level DOE goals for Plug-in Hybrid Electric Vehicles (PHEV). The specific procedures defined in this manual support the performance and life characterization of advanced battery devices under development for PHEV’s. However, it does share some methods described in the previously published battery test manual for power-assist hybrid electric vehicles. Due to the complexity of some of the procedures and supporting analysis, a revision including some modifications and clarifications of these procedures is expected. As in previous battery and capacitor test manuals, this version of the manual defines testing methods for full-size battery systems, along with provisions for scaling these tests for modules, cells or other subscale level devices.

  20. Radiochemical problems of fusion reactors. 1. Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Crespi, M.B.A.

    1984-02-01

    A list of fusion reactor candidate materials is given, for use in connection with blanket structure, breeding, moderation, neutron multiplication, cooling, magnetic field generation, electrical insulation and radiation shielding. The phenomena being studied for each group of materials are indicated. Suitable irradiation test facilities are discussed under the headings (1) accelerator-based neutron sources, (2) fission reactors, and (3) ion accelerators.

  1. The 5th surveillance testing for Kori unit 1 reactor vessel materials

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwun Jae; Gong, Un Sik; Chang, Jong Hwa; Joo, Yong Sun; Ahn, Sang Bok; Hong, Joon Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-08-01

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 5th surveillance testing was performed primarily by Korea Atomic Energy Research Institute and Westinhouse corporation partially involved in testing and calculation data evaluation in order to obtain reliable test result. Fast neutron fluences for capsule V, T, S, R and P were 5.087E+18, 1.115E+19, 1.228E+19, 2.988E+19, and 3.938E+19n/cm2, respectively. The bias factor, the ratio of calculation/measurement, was 0.940 for the 1st through 5th testing and the calculational uncertainty, 7% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.9846E+19n/cm{sup 2} based on the end of 17th fuel cycle and it was predicted that the fluences of vessel inside surface at 24, 32, 40 and 48EFPY would reach 3.0593E+19, 4.0695E+19, 5.0797E+19 and 6.0900E+19n/cm{sup 2} based on the current calculation. PTS analysis for Kori unit 1 showed that 27.93EFPY was the threshold value for 300 deg F requirement. 71 refs., 33 figs., 52 tabs. (Author)

  2. In-reactor tests of the nuclear light bulb rocket concept

    Science.gov (United States)

    Gauntt, R. O.; Slutz, S. A.; Latham, T. S.; Roman, W. C.; Rogers, R. J.

    1992-07-01

    An overview is given of the closed-cycle Gas Core Nuclear Rocket outlining scenarios for its use in short-duration Mars missions and results of Nuclear Light Bulb (NLB) tests. Isothermal and nonnuclear tests are described which confirmed the fundamental concepts behind the NLB. NLB reference-engine performance characteristics are given for hypothetical engines that could be used for manned Mars missions. Vehicle/propulsion sizing is based on a Mars mission with three trans-Mars impulse burns, capture and escape burns, and a total mission duration of 600 days. The engine would have a specific impulse of 1870 seconds, a 412-kN thrust, and a thrust/weight ratio of 1.3. Reactor tests including small-scale in-reactor tests are shown to be prerequisites for studying: (1) fluid mechanical confinement of the gaseous nuclear fuel; (2) buffer gas separation and circulation; and (3) the minimization of transparent wall-heat loading. The reactor tests are shown to be critical for establishing the feasibility of the NLB concept.

  3. Battery Test Manual For 48 Volt Mild Hybrid Electric Vehicles

    Energy Technology Data Exchange (ETDEWEB)

    Walker, Lee Kenneth [Idaho National Laboratory

    2017-03-01

    This manual details the U.S. Advanced Battery Consortium and U.S. Department of Energy Vehicle Technologies Program goals, test methods, and analysis techniques for a 48 Volt Mild Hybrid Electric Vehicle system. The test methods are outlined stating with characterization tests, followed by life tests. The final section details standardized analysis techniques for 48 V systems that allow for the comparison of different programs that use this manual. An example test plan is included, along with guidance to filling in gap table numbers.

  4. Pin-wise Reactor Analysis Based on the Generalized Equivalence Theory

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Hwan Yeal; Heo, Woong; Kim, Yong Hee [KAIST, Daejeon (Korea, Republic of)

    2016-05-15

    In this paper, a pin-wise reactor analysis is performed based on the generalized equivalence theory. From the conventional fuel assembly lattice calculations, pin-wise 2-group cross sections and pin DFs are generated. Based on the numerical results on a small PWR benchmark, it is observed that the pin-wise core analysis provide quite accurate prediction on the effective multiplication factor and the peak pin power error is bounded by about 3% in peripheral fuel assemblies facing the baffle-reflector. Also, it was found that relatively large pin power errors occur along the interface between clearly different fuel assemblies. It is expected that the GET-based pin-by-pin core calculation can be further developed as an advanced method for reactor analysis via improving the group constants and discontinuity factors. Recently, high-fidelity multi-dimensional analysis tools are gaining more attention because of their accurate prediction of local parameters for core design and safety assessment. In terms of accuracy, direct whole-core transport is quite promising. However, it is clear that it is still very costly in terms of the computing time and memory requirements. Another possible solution is the pin-by-pin core analysis in which only small fuel pins are homogenized and the 3-D core analysis is still performed using a low-order operator such as the diffusion theory. In this paper, a pin-by-pin core analysis is performed using the hybrid CMFD (HCMFD) method. Hybrid CMFD is a new global-local iteration method that has been developed for efficient parallel calculation of pinby-pin heterogeneous core analysis. For the HCMFD method, the one-node CMFD scheme is combined with a local two-node CMFD method in a non-linear way. Since the SPH method is iterative and SPH factors are not direction dependent, it is clear that SPH method takes more computing cost and cannot take into account the different heterogeneity and transport effects at each pin interface. Unlike the SPH

  5. Comprehensive Technical Report, General Electric Direct-Air-Cycle Aircraft Nuclear Propulsion Program, Program Summary and References

    Energy Technology Data Exchange (ETDEWEB)

    Thornton, G.; Rothstein, A.J.

    1962-06-28

    This is one of twenty-one volumes sumarizing the Aircraft Nuclear Propulsion Program of the General Electric Company. This volume discusses the background to the General Electric program, and summarizes the various direct-air-cycle nuclear test assemblies and power plants that were developed. Because of the requirements of high performance, low weight, and small size, vast improvements in existing technology were required to meet the flight objectives. The technological progress achieved during the program is also summarized. The last appendix contains a compilation of the abstracts, tables of contents, and reference lists of the other twenty volumes.

  6. On the road performance tests of electric test vehicle for correlation with road load simulator

    Science.gov (United States)

    Dustin, M. O.; Slavik, R. J.

    1982-08-01

    A dynamometer (road load simulator) is used to test and evaluate electric vehicle propulsion systems. To improve correlation between system tests on the road load simulator and on the road, similar performance tests are conducted using the same vehicle. The results of track tests on the electric propulsion system test vehicle are described. The tests include range at constant speeds and over SAE J227a driving cycles, maximum accelerations, maximum gradability, and tire rolling resistance determination. Road power requirements and energy consumption were also determined from coast down tests.

  7. Effects of an electric field on white sharks: in situ testing of an electric deterrent.

    Science.gov (United States)

    Huveneers, Charlie; Rogers, Paul J; Semmens, Jayson M; Beckmann, Crystal; Kock, Alison A; Page, Brad; Goldsworthy, Simon D

    2013-01-01

    Elasmobranchs can detect minute electromagnetic fields, shark deterrents to improve human safety. The present study tested the effects of the Shark Shield Freedom7™ electric deterrent on (1) the behaviour of 18 white sharks (Carcharodon carcharias) near a static bait, and (2) the rates of attacks on a towed seal decoy. In the first experiment, 116 trials using a static bait were performed at the Neptune Islands, South Australia. The proportion of baits taken during static bait trials was not affected by the electric field. The electric field, however, increased the time it took them to consume the bait, the number of interactions per approach, and decreased the proportion of interactions within two metres of the field source. The effect of the electric field was not uniform across all sharks. In the second experiment, 189 tows using a seal decoy were conducted near Seal Island, South Africa. No breaches and only two surface interactions were observed during the tows when the electric field was activated, compared with 16 breaches and 27 surface interactions without the electric field. The present study suggests that the behavioural response of white sharks and the level of risk reduction resulting from the electric field is contextually specific, and depends on the motivational state of sharks.

  8. Evaluation program for secondary spacecraft cells: Initial evaluation tests of General Electric Company 40.0 ampere-hour nickel-cadmium spacecraft cells for the tracking data relay satellite system

    Science.gov (United States)

    Harkness, J. D.

    1978-01-01

    Five cells provided by NASA's Goddard Space Flight Center were evaluated at room temperature and pressure (25 C plus or minus 2 C) with discharges at the 2 hour rate. Measurements of the cell containers following test, indicated an average increase of .006 inches at the plate thickness. Average end of charge voltages and pressures, and capacity output in ampere hours were determined. Three cells exceeded the voltage requirements of 1.52 volts during both c/10 charges at 20 C. All cells exceeded the voltage requirement of 1.52 volts during the 0 C overcharge test, although their end charges were below 1.50 volts. The pressure requirement of 65 psia was exceeded by both pressure transducer cells during c/10 charges at 25 C and 20 C and also during the 0 C overcharge test. The cells with pressure transducers reached a pressure of 20 psia before reaching the voltage limit of 1.550 volts during the pressure versus capacity test, and exhibited a pressure decay of 2 psia during the last 30 minutes of the 1 hour open circuit stand. Average capacity was 51.3 ampere hours.

  9. 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.

    1989-01-01

    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.

  10. 77 FR 70484 - Preoperational Testing of Onsite Electric Power Systems To Verify Proper Load Group Assignments...

    Science.gov (United States)

    2012-11-26

    ... COMMISSION Preoperational Testing of Onsite Electric Power Systems To Verify Proper Load Group Assignments... Power Systems to Verify Proper Load Group Assignments, Electrical Separation, and Redundancy.'' DG-1294... encompass preoperational testing of electrical power systems used to meet current Station...

  11. Assessment of Startup Fuel Options for a Test or Demonstration Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Carmack, Jon [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hayes, Steven [Idaho National Lab. (INL), Idaho Falls, ID (United States); Walters, L. C. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

    This document explores startup fuel options for a proposed test/demonstration fast reactor. The fuel options considered are the metallic fuels U-Zr and U-Pu-Zr and the ceramic fuels UO2 and UO2-PuO2 (MOX). Attributes of the candidate fuel choices considered were feedstock availability, fabrication feasibility, rough order of magnitude cost and schedule, and the existing irradiation performance database. The reactor-grade plutonium bearing fuels (U-Pu-Zr and MOX) were eliminated from consideration as the initial startup fuels because the availability and isotopics of domestic plutonium feedstock is uncertain. There are international sources of reactor grade plutonium feedstock but isotopics and availability are also uncertain. Weapons grade plutonium is the only possible source of Pu feedstock in sufficient quantities needed to fuel a startup core. Currently, the available U.S. source of (excess) weapons-grade plutonium is designated for irradiation in commercial light water reactors (LWR) to a level that would preclude diversion. Weapons-grade plutonium also contains a significant concentration of gallium. Gallium presents a potential issue for both the fabrication of MOX fuel as well as possible performance issues for metallic fuel. Also, the construction of a fuel fabrication line for plutonium fuels, with or without a line to remove gallium, is expected to be considerably more expensive than for uranium fuels. In the case of U-Pu-Zr, a relatively small number of fuel pins have been irradiated to high burnup, and in no case has a full assembly been irradiated to high burnup without disassembly and re-constitution. For MOX fuel, the irradiation database from the Fast Flux Test Facility (FFTF) is extensive. If a significant source of either weapons-grade or reactor-grade Pu became available (i.e., from an international source), a startup core based on Pu could be reconsidered.

  12. A High-power Electric Propulsion Test Platform in Space

    Science.gov (United States)

    Petro, Andrew J.; Reed, Brian; Chavers, D. Greg; Sarmiento, Charles; Cenci, Susanna; Lemmons, Neil

    2005-01-01

    This paper will describe the results of the preliminary phase of a NASA design study for a facility to test high-power electric propulsion systems in space. The results of this design study are intended to provide a firm foundation for subsequent detailed design and development activities leading to the deployment of a valuable space facility. The NASA Exploration Systems Mission Directorate is sponsoring this design project. A team from the NASA Johnson Space Center, Glenn Research Center, the Marshall Space Flight Center and the International Space Station Program Office is conducting the project. The test facility is intended for a broad range of users including government, industry and universities. International participation is encouraged. The objectives for human and robotic exploration of space can be accomplished affordably, safely and effectively with high-power electric propulsion systems. But, as thruster power levels rise to the hundreds of kilowatts and up to megawatts, their testing will pose stringent and expensive demands on existing Earth-based vacuum facilities. These considerations and the human access to near-Earth space provided by the International Space Station (ISS) have led to a renewed interest in space testing. The ISS could provide an excellent platform for a space-based test facility with the continuous vacuum conditions of the natural space environment and no chamber walls to modify the open boundary conditions of the propulsion system exhaust. The test platform could take advantage of the continuous vacuum conditions of the natural space environment. Space testing would provide open boundary conditions without walls, micro-gravity and a realistic thermal environment. Testing on the ISS would allow for direct observation of the test unit, exhaust plume and space-plasma interactions. When necessary, intervention by on-board personnel and post-test inspection would be possible. The ISS can provide electrical power, a location for

  13. Criticality Safety Evaluation for the Advanced Test Reactor U-Mo Demonstration Elements

    Energy Technology Data Exchange (ETDEWEB)

    Leland M. Montierth

    2010-12-01

    The Reduced Enrichment Research Test Reactors (RERTR) fuel development program is developing a high uranium density fuel based on a (LEU) uranium-molybdenum alloy. Testing of prototypic RERTR fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. Two RERTR-Full Size Demonstration fuel elements based on the ATR-Reduced YA elements (all but one plate fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). The two fuel elements will be irradiated in alternating cycles such that only one element is loaded in the reactor at a time. Existing criticality analyses have analyzed Standard (HEU) ATR elements (all plates fueled) from which controls have been derived. This criticality safety evaluation (CSE) documents analysis that determines the reactivity of the Demonstration fuel elements relative to HEU ATR elements and shows that the Demonstration elements are bound by the Standard HEU ATR elements and existing HEU ATR element controls are applicable to the Demonstration elements.

  14. On0Line Fuel Failure Monitor for Fuel Testing and Monitoring of Gas Cooled Very High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ayman I. Hawari; Mohamed A. Bourham

    2010-04-22

    IVery High Temperature Reactors (VHTR) utilize the TRISO microsphere as the fundamental fuel unit in the core. The TRISO microsphere (~ 1- mm diameter) is composed of a UO2 kernel surrounded by a porous pyrolytic graphite buffer, an inner pyrolytic graphite layer, a silicon carbide (SiC) coating, and an outer pyrolytic graphite layer. The U-235 enrichment of the fuel is expected to range from 4% – 10% (higher enrichments are also being considered). The layer/coating system that surrounds the UO2 kernel acts as the containment and main barrier against the environmental release of radioactivity. To understand better the behavior of this fuel under in-core conditions (e.g., high temperature, intense fast neutron flux, etc.), the US Department of Energy (DOE) is launching a fuel testing program that will take place at the Advanced Test Reactor (ATR) located at Idaho National Laboratory (INL). During this project North Carolina State University (NCSU) researchers will collaborate with INL staff for establishing an optimized system for fuel monitoring for the ATR tests. In addition, it is expected that the developed system and methods will be of general use for fuel failure monitoring in gas cooled VHTRs.

  15. Development of in-vessel type control rod drive mechanism for a innovative small reactor (Contract research)

    Energy Technology Data Exchange (ETDEWEB)

    Yoritsune, Tsutomu; Ishida, Toshihisa [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    Although the control rod drive mechanism of an existing large scale light water reactor is generally installed outside the reactor vessel, an in-vessel type control rod drive mechanism (INV-CRDM) is installed inside the reactor vessel. The INV-CRDM contributes to compactness and simplicity of the reactor system, and it can eliminate the possibility of a rod ejection accident. Therefore, INV-CRDM is an important technology adopted in an innovative small reactor. Japan Atomic Energy Research Institute (JAERI) has developed this type of CRDM driven by an electric motor, which can work under high temperature and high pressure water for the advanced marine reactor. On the basis of this research result, a driving motor coil and a bearing were developed to be used under the high temperature steam, severe condition for an innovative small reactor. About the driving motor, we manufactured the driving motor available for high temperature steam and carried out performance test under room temperature atmosphere to confirm the electric characteristic and coolability of the driving coil. With these test results and the past test results under high temperature water, we analyzed and evaluated the electric performance and coolability of the driving coil under high temperature steam. Concerning bearing, we manufactured the test pieces using some candidate material for material characteristic test and carried out the rolling wear test under high temperature steam to select the material. Consequently, we confirmed that performance of the driving coil for the advanced type driving motor, is enough to be used under high temperature steam. And, we evaluated the performance of the bearing and selected the material of the bearing, which can be used under high temperature steam. From these results, we have obtained the prospect that the INV-CRDM can be used for an innovative small reactor under steam atmosphere could be developed. (author)

  16. Development of in-vessel type control rod drive mechanism for a innovative small reactor (Contract research)

    Energy Technology Data Exchange (ETDEWEB)

    Yoritsune, Tsutomu; Ishida, Toshihisa [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    Although the control rod drive mechanism of an existing large scale light water reactor is generally installed outside the reactor vessel, an in-vessel type control rod drive mechanism (INV-CRDM) is installed inside the reactor vessel. The INV-CRDM contributes to compactness and simplicity of the reactor system, and it can eliminate the possibility of a rod ejection accident. Therefore, INV-CRDM is an important technology adopted in an innovative small reactor. Japan Atomic Energy Research Institute (JAERI) has developed this type of CRDM driven by an electric motor, which can work under high temperature and high pressure water for the advanced marine reactor. On the basis of this research result, a driving motor coil and a bearing were developed to be used under the high temperature steam, severe condition for an innovative small reactor. About the driving motor, we manufactured the driving motor available for high temperature steam and carried out performance test under room temperature atmosphere to confirm the electric characteristic and coolability of the driving coil. With these test results and the past test results under high temperature water, we analyzed and evaluated the electric performance and coolability of the driving coil under high temperature steam. Concerning bearing, we manufactured the test pieces using some candidate material for material characteristic test and carried out the rolling wear test under high temperature steam to select the material. Consequently, we confirmed that performance of the driving coil for the advanced type driving motor, is enough to be used under high temperature steam. And, we evaluated the performance of the bearing and selected the material of the bearing, which can be used under high temperature steam. From these results, we have obtained the prospect that the INV-CRDM can be used for an innovative small reactor under steam atmosphere could be developed. (author)

  17. Electric vehicles pick up speed as oil prices climb. [General Motors planning EVs in 1985

    Energy Technology Data Exchange (ETDEWEB)

    Bowe, J.

    1980-02-01

    General Motors recently announced that in 1985, its Chevrolet division will begin manufacturing either two-passenger electric automobiles, or small electric vans, or both. Key features of the electric-vehicles will be a separately excited DC motor, regenerative braking, electronic controls, and coil-type suspension. Newly developed zinc-nickel batteries promise higher speeds (up to about 55 mph) and a longer cruising range (about 100 mi before needing a recharge and 30,000 mi before needing replacement). General Motors' commitment to the electric-vehicle market signals the beginning of a radical departure from a traditional petroleum-based economy. (1 diagram, 1 drawing, 2 photos)

  18. Vortex Diode Analysis and Testing for Fluoride Salt-Cooled High-Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Yoder Jr, Graydon L [ORNL; Elkassabgi, Yousri M. [Texas A& M University, Kingsville; De Leon, Gerardo I. [Texas A& M University, Kingsville; Fetterly, Caitlin N. [Texas A& M University, Kingsville; Ramos, Jorge A. [Texas A& M University, Kingsville; Cunningham, Richard Burns [University of Tennessee, Knoxville (UTK)

    2012-02-01

    Fluidic diodes are presently being considered for use in several fluoride salt-cooled high-temperature reactor designs. A fluidic diode is a passive device that acts as a leaky check valve. These devices are installed in emergency heat removal systems that are designed to passively remove reactor decay heat using natural circulation. The direct reactor auxiliary cooling system (DRACS) uses DRACS salt-to-salt heat exchangers (DHXs) that operate in a path parallel to the core flow. Because of this geometry, under normal operating conditions some flow bypasses the core and flows through the DHX. A flow diode, operating in reverse direction, is-used to minimize this flow when the primary coolant pumps are in operation, while allowing forward flow through the DHX under natural circulation conditions. The DRACSs reject the core decay heat to the environment under loss-of-flow accident conditions and as such are a reactor safety feature. Fluidic diodes have not previously been used in an operating reactor system, and therefore their characteristics must be quantified to ensure successful operation. This report parametrically examines multiple design parameters of a vortex-type fluidic diode to determine the size of diode needed to reject a particular amount of decay heat. Additional calculations were performed to size a scaled diode that could be tested in the Oak Ridge National Laboratory Liquid Salt Flow Loop. These parametric studies have shown that a 152.4 mm diode could be used as a test article in that facility. A design for this diode is developed, and changes to the loop that will be necessary to test the diode are discussed. Initial testing of a scaled flow diode has been carried out in a water loop. The 150 mm diode design discussed above was modified to improve performance, and the final design tested was a 171.45 mm diameter vortex diode. The results of this testing indicate that diodicities of about 20 can be obtained for diodes of this size. Experimental

  19. 40 CFR 86.1773-99 - Test sequence; general requirements.

    Science.gov (United States)

    2010-07-01

    ... required for fuel-flexible and dual-fuel vehicles when operating on gasoline. Natural gas, hybrid electric and diesel-fueled vehicles shall also be exempt from 50 °F testing. (3) The following schedule... vehicle shall be approximately level during all phases of the test sequence to prevent abnormal fuel...

  20. Fluidized-bed reactor model with generalized particle balances. Part 1. Formulation and solution

    Energy Technology Data Exchange (ETDEWEB)

    Overturf, B.W.; Reklaitis, G.V.

    1983-09-01

    In this first part, a particle balance model is developed for a fluidized-bed gas-solid reactor which accommodates particle distributions dependent on both size and density, as well as populations consisting of multiple solids.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-10-01

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

  2. Field test and evaluation of the passive neutron coincidence collar for prototype fast reactor fuel subassemblies

    Energy Technology Data Exchange (ETDEWEB)

    Menlove, H.O.; Keddar, A.

    1982-08-01

    The passive neutron Coincidence Collar, which was developed for the verification of plutonium content in fast reactor fuel subassemblies, has been field tested using Prototype Fast Reactor fuel. For passive applications, the system measures the /sup 240/Pu-effective mass from the spontaneous fission rate, and in addition, a self-interrogation technique is used to determine the fissile content in the subassembly. Both the passive and active modes were evaluated at the Windscale Works in the United Kingdom. The results of the tests gave a standard deviation 0.75% for the passive count and 3 to 7% for the active measurement for a 1000-s counting time. The unit will be used in the future for the verification of plutonium in fresh fuel assemblies.

  3. A preliminary neutronic evaluation of high temperature engineering test reactor using the SCALE6 code

    Science.gov (United States)

    Tanure, L. P. A. R.; Sousa, R. V.; Costa, D. F.; Cardoso, F.; Veloso, M. A. F.; Pereira, C.

    2014-02-01

    Neutronic parameters of some fourth generation nuclear reactors have been investigated at the Departamento de Engenharia Nuclear/UFMG. Previous studies show the possibility to increase the transmutation capabilities of these fourth generation systems to achieve significant reduction concerning transuranic elements in spent fuel. To validate the studies, a benchmark on core physics analysis, related to initial testing of the High Temperature Engineering Test Reactor and provided by International Atomic Energy Agency (IAEA) was simulated using the Standardized Computer Analysis for Licensing Evaluation (SCALE). The CSAS6/KENO-VI control sequence and the 44-group ENDF/B-V 0 cross-section neutron library were used to evaluate the keff (effective multiplication factor) and the result presents good agreement with experimental value.

  4. An Evaluation of a General Venting Strategy in CANDU 6 Reactor Building

    Energy Technology Data Exchange (ETDEWEB)

    Kim, See Darl; Kim, Dong Ha; Park, Soo Yong; Song, Yong Man; Jin, Young Ho

    2006-03-15

    If the reactor building sprays or local air coolers are not available, depressurization by reactor building venting is considered as a useful mitigation strategy for a severe accident management of the Wolsong plants. As the CFVS is not established in the Wolsong Units, the reactor building isolation system can be a substitute for reactor building venting. The D{sub 2}O Vapour recovery system which has a 30' diameter penetration is expected to meet the NRC requirements. To investigate the effectiveness of the Reactor Building Venting Strategy, three kinds of accidents are analyzed: a SBO, a SLOCA and a Large LOCA. The reactor building pressure behavior was analyzed with ISAAC 2.0.2 for four different cases: without venting, 55psig/50psig, 50psig/40psig and 50psig/30psig valve open/close pressures. It was found that applying venting for a SBO reduces the mass fraction of the CsI released to the environment by 67.8% (valve open/close pressure of 55psig/50psig), by 64.4% (valve open/close pressure of 50psig/40psig) and by 63.5% (valve open/close pressure of 50psig/30psig). For a SLOCA, venting strategy reduces the mass fraction of the CsI by 58.3% (valve open/close pressure of 55psig/50psig), by 55.0% (valve open/close pressure 50psig/40psig) and by 48.3% (valve open/close pressure 50psig/30psig). For a LLOCA, reactor building venting reduces the mass fraction of the CsI released to the environment by less than 10% when compared to that without reactor building venting. When the reactor building spray or local air coolers can not be operated, a depressurization strategy by using the D{sub 2}O Vapour Recovery System could prevent a reactor building failure and reduce the amount of CsI released to the environment. The present study shows that the operation of valves at a pressure of 55psig/50psig is safe and effective. Based on the current study, the strategy of reactor building venting is involved SAMG-5.

  5. Optimal DO Setpoint Decision and Electric Cost Saving in Aerobic Reactor Using Respirometer and Air Blower Control

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kwang Su; Yoo, Changkyoo [Kyung Hee University, Yongin (Korea, Republic of); Kim, Minhan [Pangaea21 Ltd., Seongnam (Korea, Republic of); Kim, Jongrack [UnUsoft Ltd., Seoul (Korea, Republic of)

    2014-10-15

    Main objects for wastewater treatment operation are to maintain effluent water quality and minimize operation cost. However, the optimal operation is difficult because of the change of influent flow rate and concentrations, the nonlinear dynamics of microbiology growth rate and other environmental factors. Therefore, many wastewater treatment plants are operated for much more redundant oxygen or chemical dosing than the necessary. In this study, the optimal control scheme for dissolved oxygen (DO) is suggested to prevent over-aeration and the reduction of the electric cost in plant operation while maintaining the dissolved oxygen (DO) concentration for the metabolism of microorganisms in oxic reactor. The oxygen uptake rate (OUR) is real-time measured for the identification of influent characterization and the identification of microorganisms' oxygen requirement in oxic reactor. Optimal DO set-point needed for the micro-organism is suggested based on real-time measurement of oxygen uptake of micro-organism and the control of air blower. Therefore, both stable effluent quality and minimization of electric cost are satisfied with a suggested optimal set-point decision system by providing the necessary oxygen supply requirement to the micro-organisms coping with the variations of influent loading.

  6. ACCELERATED AGING AND ELECTRICAL CONDUCTIVITY TESTS IN CRAMBE SEEDS

    Directory of Open Access Journals (Sweden)

    Juliana Joice Pereira Lima

    2015-02-01

    Full Text Available The aim of this study was to adapt the methodology of the accelerated aging and electrical conductivity tests for determination of physiological potential in crambe seeds. Six seed lots of crambe (cv. FMS Brilhante were subjected to determination of moisture content, germination test, first count germination, emergence, and emergence speed index. For the accelerated aging test, the traditional methodology was used with water, and with a saturated potassium chloride and sodium chloride solution in three periods of exposure (24, 48, and 72 hours at 41°C; the electrical conductivity test was performed with four pre-soaking treatments (0, 2, 4, and 8 hours and four soaking periods (4, 8, 16, and 24 hours at 25°C. The accelerated aging test with water for 72 hours and the electrical conductivity test with 2 hours of pre-soaking and assessment after 16 hours were effective for classification of the crambe seed lots in regard to physiological quality.

  7. Technology, safety, and costs of decommissioning reference nuclear research and test reactors. Main report

    Energy Technology Data Exchange (ETDEWEB)

    Konzek, G.J.; Ludwick, J.D.; Kennedy, W.E. Jr.; Smith, R.I.

    1982-03-01

    Safety and Cost Information is developed for the conceptual decommissioning of two representative licensed nuclear research and test reactors. Three decommissioning alternatives are studied to obtain comparisons between costs (in 1981 dollars), occupational radiation doses, potential radiation dose to the public, and other safety impacts. The alternatives considered are: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). The study results are presented in two volumes. Volume 1 (Main Report) contains the results in summary form.

  8. After Action Report: Advanced Test Reactor Complex 2015 Evaluated Drill October 6, 2015

    Energy Technology Data Exchange (ETDEWEB)

    Holmes, Forest Howard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-11-01

    The Advanced Test Reactor (ATR) Complex, operated by Battelle Energy Alliance, LLC, at the Idaho National Laboratory (INL) conducted an evaluated drill on October 6, 2015, to allow the ATR Complex emergency response organization (ERO) to demonstrate the ability to respond to and mitigate an emergency by implementing the requirements of DOE O 151.1C, “Comprehensive Emergency Management System.”

  9. Dynamic parameters test of Haiyang Nuclear Power Engineering in reactor areas, China

    Science.gov (United States)

    Zhou, N.; Zhao, S.; Sun, L.

    2012-12-01

    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

  10. Highly Perturbed Operational Test Configurations at the WSMR Fast Burst Reactor

    Directory of Open Access Journals (Sweden)

    Flanders T. Michael

    2016-01-01

    Full Text Available The White Sands Missile Range (WSMR MoLLY-G reactor has a long history of producing a well characterized environment for testing electronic systems/devices in fission environments. As an unmoderated, unreflected, bare critical assembly, it provides a slightly degraded fission spectrum with a 1/E tail. For radiation hardness testing of electronics, the neutron fluence is usually reported as the 1-MeV Equivalent Neutron Fluence for Silicon. In this paper, we examine additional neutron environments and characterizations ranging from low intensity neutron fields to more extreme modifications of our normal test environment.

  11. General Electric PETtrace cyclotron as a neutron source for boron neutron capture therapy

    Science.gov (United States)

    Bosko, Andrey

    This research investigates the use of a PETtrace cyclotron produced by General Electric (GE) as a neutron source for boron neutron capture therapy (BNCT). The GE PETtrace was chosen for this investigation because this type of cyclotron is popular among nuclear pharmacies and clinics in many countries; it is compact and reliable; it produces protons with energies high enough to produce neutrons with appropriate energy and fluence rate for BNCT and it does not require significant changes in design to provide neutrons. In particular, the standard PETtrace 18O target is considered. The cyclotron efficiency may be significantly increased if unused neutrons produced during radioisotopes production could be utilized for other medical modalities such as BNCT at the same time. The resulting dose from the radiation emitted from the target is evaluated using the Monte Carlo radiation transport code MCNP at several depths in a brain phantom for different scattering geometries. Four different moderating materials of various thicknesses were considered: light water, carbon, heavy water, arid Fluental(TM). The fluence rate tally was used to calculate photon and neutron dose, by applying fluence rate-to-dose conversion factors. Fifteen different geometries were considered and a 30-cm thick heavy water moderator was chosen as the most suitable for BNCT with the GE PETtrace cyclotron. According to the Brookhaven Medical Research Reactor (BMRR) protocol, the maximum dose to the normal brain is set to 12.5 RBEGy, which for the conditions of using a heavy water moderator, assuming a 60 muA beam current, would be reached with a treatment time of 258 min. Results showed that using a PETtrace cyclotron in this configuration provides a therapeutic ratio of about 2.4 for depths up to 4 cm inside a brain phantom. Further increase of beam current proposed by GE should significantly improve the beam quality or the treatment time and allow treating tumors at greater depths.

  12. Initial Testing of the Microscopic Depletion Implementation in the MAMMOTH Reactor Physics Application

    Energy Technology Data Exchange (ETDEWEB)

    J. Ortensi; Y. Wang; S. Schunert; B.D. Ganapol; F.N. Gleicher; B. Baker; M.D. DeHart

    2016-09-01

    Present and new nuclear fuels that will be tested at the Transient Reactor Test (TREAT) facility will be analyzed with the MAMMOTH reactor physics application, currently under development, at Idaho National Laboratory. MAMMOTH natively couples the BISON, RELAP-7, and Rattlesnake applications within the MOOSE framework. This system allows the irradiation of fuel from beginning of life in a nuclear reactor until it is placed in TREAT for fuel testing within the same analysis mesh and, thus, retaining a very high level of resolution and fidelity. The calculation of the isotopic distribution in fuel requires the solution to the decay and transmutation equations coupled to the neutron transport equation. The Chebyshev Rational Approximation Method (CRAM) is the current state-of-the-art in the field, as was chosen to be the solver for the decay and transmutation equations. This report shows that the implementation of the CRAM solver within MAMMOTH is correct with various analytic benchmarks for decay and transmutation of nuclides. The results indicate that the solutions with CRAM order 16 achieve the level of precision of the benchmark. The CRAM solutions show little sensitivity to the time step size and consistently produce a high level of accuracy for isotopic decay for time steps of 1x10^11 years. Comparisons to DRAGON5 with 297 isotopes yield comparable results, but some differences need to be further analyzed.

  13. Long term out-of-pile thermocouple tests in conditions representative for nuclear gas-cooled high temperature reactors

    OpenAIRE

    LAURIE Mathias; FOURREZ Stéphane; FUETTERER Michael; LAPETITE Jean-Marc; SADLI M.; MORICE Ronan; FAILLEAU G

    2013-01-01

    During irradiation tests at high temperature failure of commercial Inconel 600 sheathed thermocouples is commonly encountered. As instrumentation, in particular thermocouples are considered safety-relevant both for irradiation tests and for commercial reactors, JRC and THERMOCOAX joined forces to solve this issue by performing out-of-pile tests with thermocouples mimicking the environment encountered by high temperature reactor (HTR) in-core instrumentation. The objective was to screen innova...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

  16. NUMERICAL SIMULATION FOR MECHANICAL BEHAVIOR OF U10MO MONOLITHIC MINIPLATES FOR RESEARCH AND TEST REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Hakan Ozaltun & Herman Shen

    2011-11-01

    This article presents assessment of the mechanical behavior of U-10wt% Mo (U10Mo) alloy based monolithic fuel plates subject to irradiation. Monolithic, plate-type fuel is a new fuel form being developed for research and test reactors to achieve higher uranium densities within the reactor core to allow the use of low-enriched uranium fuel in high-performance reactors. Identification of the stress/strain characteristics is important for understanding the in-reactor performance of these plate-type fuels. For this work, three distinct cases were considered: (1) fabrication induced residual stresses (2) thermal cycling of fabricated plates; and finally (3) transient mechanical behavior under actual operating conditions. Because the temperatures approach the melting temperature of the cladding during the fabrication and thermal cycling, high temperature material properties were incorporated to improve the accuracy. Once residual stress fields due to fabrication process were identified, solution was used as initial state for the subsequent simulations. For thermal cycling simulation, elasto-plastic material model with thermal creep was constructed and residual stresses caused by the fabrication process were included. For in-service simulation, coupled fluid-thermal-structural interaction was considered. First, temperature field on the plates was calculated and this field was used to compute the thermal stresses. For time dependent mechanical behavior, thermal creep of cladding, volumetric swelling and fission induced creep of the fuel foil were considered. The analysis showed that the stresses evolve very rapidly in the reactor. While swelling of the foil increases the stress of the foil, irradiation induced creep causes stress relaxation.

  17. Method of determining an electrical property of a test sample

    DEFF Research Database (Denmark)

    2010-01-01

    A method of obtaining an electrical property of a test sample, comprising a non-conductive area and a conductive or semi-conductive test area, byperforming multiple measurements using a multi-point probe. The method comprising the steps of providing a magnetic field having field lines passing...... each tip, selecting one tip to be a current source positioned between conductive tips being used for determining a voltage in the test sample, performing a first measurement, moving the probe and performing a second measurement, calculating on the basis of the first and second measurement...... perpendicularly through the test area, bringing the probe into a first position on the test area, the conductive tips of the probe being in contact with the test area, determining a position for each tip relative to the boundary between the non- conductive area and the test area, determining distances between...

  18. An Experimental Test Facility to Support Development of the Fluoride Salt Cooled High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yoder Jr, Graydon L [ORNL; Aaron, Adam M [ORNL; Cunningham, Richard Burns [University of Tennessee, Knoxville (UTK); Fugate, David L [ORNL; Holcomb, David Eugene [ORNL; Kisner, Roger A [ORNL; Peretz, Fred J [ORNL; Robb, Kevin R [ORNL; Wilgen, John B [ORNL; Wilson, Dane F [ORNL

    2014-01-01

    The need for high-temperature (greater than 600 C) energy exchange and delivery systems is significantly increasing as the world strives to improve energy efficiency and develop alternatives to petroleum-based fuels. Liquid fluoride salts are one of the few energy transport fluids that have the capability of operating at high temperatures in combination with low system pressures. The Fluoride Salt-Cooled High-Temperature Reactor design uses fluoride salt to remove core heat and interface with a power conversion system. Although a significant amount of experimentation has been performed with these salts, specific aspects of this reactor concept will require experimental confirmation during the development process. The experimental facility described here has been constructed to support the development of the Fluoride Salt Cooled High Temperature Reactor concept. The facility is capable of operating at up to 700 C and incorporates a centrifugal pump to circulate FLiNaK salt through a removable test section. A unique inductive heating technique is used to apply heat to the test section, allowing heat transfer testing to be performed. An air-cooled heat exchanger removes added heat. Supporting loop infrastructure includes a pressure control system; trace heating system; and a complement of instrumentation to measure salt flow, temperatures, and pressures around the loop. The initial experiment is aimed at measuring fluoride salt heat transfer inside a heated pebble bed similar to that used for the core of the pebble bed advanced high-temperature reactor. This document describes the details of the loop design, auxiliary systems used to support the facility, the inductive heating system, and facility capabilities.

  19. 国家标准GB9706.1-2007医用电气设备安全通用要求所需试验及仪器分析%Analysis of The Needed Test-instrument in Medical Electrical Equipment General Requirements for Safety GB9706.1-2007 Standard

    Institute of Scientific and Technical Information of China (English)

    魏景锋

    2013-01-01

    研究了G B9706.1-2007医用电气安全通用要求标准。针对其标准条款按顺序归纳出具体型式试验项目,并对其中需要通过仪器进行测量的试验给出具体仪器名称和主要技术参数。解决了将抽象的标准条款与具体检测试验及所用仪器设备相互结合转化的实际问题。为研究学习GB9706.1标准提供了一个新的角度。%Analysis of the medical electrical equipment general requirements for safety GB9706.1-2007 standard. Generalization of test items according to the requirements of content. According to the experiment concludes instrument and the instrument main parameters. To solve the mutual transformation between the standard terms and speciifc tests and the instruments. Provides a new perspective for the study of GB9706.1 standard.

  20. Reactor Physics Methods and Preconceptual Core Design Analyses for Conversion of the Advanced Test Reactor to Low-Enriched Uranium Fuel Annual Report for Fiscal Year 2012

    Energy Technology Data Exchange (ETDEWEB)

    David W. Nigg; Sean R. Morrell

    2012-09-01

    Under the current long-term DOE policy and planning scenario, both the ATR and the ATRC will be reconfigured at an appropriate time within the next several years to operate with low-enriched uranium (LEU) fuel. This will be accomplished under the auspices of the Reduced Enrichment Research and Test Reactor (RERTR) Program, administered by the DOE National Nuclear Security Administration (NNSA). At a minimum, the internal design and composition of the fuel element plates and support structure will change, to accommodate the need for low enrichment in a manner that maintains total core excess reactivity at a suitable level for anticipated operational needs throughout each cycle while respecting all control and shutdown margin requirements and power distribution limits. The complete engineering design and optimization of LEU cores for the ATR and the ATRC will require significant multi-year efforts in the areas of fuel design, development and testing, as well as a complete re-analysis of the relevant reactor physics parameters for a core composed of LEU fuel, with possible control system modifications. Ultimately, revalidation of the computational physics parameters per applicable national and international standards against data from experimental measurements for prototypes of the new ATR and ATRC core designs will also be required for Safety Analysis Report (SAR) changes to support routine operations with LEU. This report is focused on reactor physics analyses conducted during Fiscal Year (FY) 2012 to support the initial development of several potential preconceptual fuel element designs that are suitable candidates for further study and refinement during FY-2013 and beyond. In a separate, but related, effort in the general area of computational support for ATR operations, the Idaho National Laboratory (INL) is conducting a focused multiyear effort to introduce modern high-fidelity computational reactor physics software and associated validation protocols to replace

  1. Progress of General Test Stand for Intensive Beam Cyclotron

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    <正>The general test stand for intensive beam cyclotron is one of the preliminary tasks of BRIF project at CIAE. The test stand, which actually is a small compact cyclotron with designed energy of 10 MeV,

  2. Test case specifications for coupled neutronics-thermal hydraulics calculation of Gas-cooled Fast Reactor

    Science.gov (United States)

    Osuský, F.; Bahdanovich, R.; Farkas, G.; Haščík, J.; Tikhomirov, G. V.

    2017-01-01

    The paper is focused on development of the coupled neutronics-thermal hydraulics model for the Gas-cooled Fast Reactor. It is necessary to carefully investigate coupled calculations of new concepts to avoid recriticality scenarios, as it is not possible to ensure sub-critical state for a fast reactor core under core disruptive accident conditions. Above mentioned calculations are also very suitable for development of new passive or inherent safety systems that can mitigate the occurrence of the recriticality scenarios. In the paper, the most promising fuel material compositions together with a geometry model are described for the Gas-cooled fast reactor. Seven fuel pin and fuel assembly geometry is proposed as a test case for coupled calculation with three different enrichments of fissile material in the form of Pu-UC. The reflective boundary condition is used in radial directions of the test case and vacuum boundary condition is used in axial directions. During these condition, the nuclear system is in super-critical state and to achieve a stable state (which is numerical representation of operational conditions) it is necessary to decrease the reactivity of the system. The iteration scheme is proposed, where SCALE code system is used for collapsing of a macroscopic cross-section into few group representation as input for coupled code NESTLE.

  3. Ultracold neutron source at the PULSTAR reactor: Engineering design and cryogenic testing

    Energy Technology Data Exchange (ETDEWEB)

    Korobkina, E., E-mail: ekorobk@ncsu.edu [Department of Nuclear Engineering, North Carolina State University, 2500 Stinson Drive, Box 7909, Raleigh, NC 27695 (United States); Medlin, G. [Department of Physics, North Carolina State University, 2401 Stinson Drive, Box 8202, Raleigh, NC 27695 (United States); Triangle Universities Nuclear Laboratory, 116 Science Drive, Box 90308, Durham, NC 27708 (United States); Wehring, B.; Hawari, A.I. [Department of Nuclear Engineering, North Carolina State University, 2500 Stinson Drive, Box 7909, Raleigh, NC 27695 (United States); Huffman, P.R.; Young, A.R. [Department of Physics, North Carolina State University, 2401 Stinson Drive, Box 8202, Raleigh, NC 27695 (United States); Triangle Universities Nuclear Laboratory, 116 Science Drive, Box 90308, Durham, NC 27708 (United States); Beaumont, B. [Department of Physics, North Carolina State University, 2401 Stinson Drive, Box 8202, Raleigh, NC 27695 (United States); Palmquist, G. [Department of Physics, North Carolina State University, 2401 Stinson Drive, Box 8202, Raleigh, NC 27695 (United States); Triangle Universities Nuclear Laboratory, 116 Science Drive, Box 90308, Durham, NC 27708 (United States)

    2014-12-11

    Construction is completed and commissioning is in progress for an ultracold neutron (UCN) source at the PULSTAR reactor on the campus of North Carolina State University. The source utilizes two stages of neutron moderation, one in heavy water at room temperature and the other in solid methane at ∼40K, followed by a converter stage, solid deuterium at 5 K, that allows a single down scattering of cold neutrons to provide UCN. The UCN source rolls into the thermal column enclosure of the PULSTAR reactor, where neutrons will be delivered from a bare face of the reactor core by streaming through a graphite-lined assembly. The source infrastructure, i.e., graphite-lined assembly, heavy-water system, gas handling system, and helium liquefier cooling system, has been tested and all systems operate as predicted. The research program being considered for the PULSTAR UCN source includes the physics of UCN production, fundamental particle physics, and material surface studies of nanolayers containing hydrogen. In the present paper we report details of the engineering and cryogenic design of the facility as well as results of critical commissioning tests without neutrons.

  4. Ultracold neutron source at the PULSTAR reactor: Engineering design and cryogenic testing

    Science.gov (United States)

    Korobkina, E.; Medlin, G.; Wehring, B.; Hawari, A. I.; Huffman, P. R.; Young, A. R.; Beaumont, B.; Palmquist, G.

    2014-12-01

    Construction is completed and commissioning is in progress for an ultracold neutron (UCN) source at the PULSTAR reactor on the campus of North Carolina State University. The source utilizes two stages of neutron moderation, one in heavy water at room temperature and the other in solid methane at ~ 40 K, followed by a converter stage, solid deuterium at 5 K, that allows a single down scattering of cold neutrons to provide UCN. The UCN source rolls into the thermal column enclosure of the PULSTAR reactor, where neutrons will be delivered from a bare face of the reactor core by streaming through a graphite-lined assembly. The source infrastructure, i.e., graphite-lined assembly, heavy-water system, gas handling system, and helium liquefier cooling system, has been tested and all systems operate as predicted. The research program being considered for the PULSTAR UCN source includes the physics of UCN production, fundamental particle physics, and material surface studies of nanolayers containing hydrogen. In the present paper we report details of the engineering and cryogenic design of the facility as well as results of critical commissioning tests without neutrons.

  5. 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)

    2011-07-01

    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)

  6. Assessment of Feasibility of the Beneficial Use of Waste Heat from the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Donna P. Guillen

    2012-07-01

    This report investigates the feasibility of using waste heat from the Advanced Test Reactor (ATR). A proposed glycol waste heat recovery system was assessed for technical and economic feasibility. The system under consideration would use waste heat from the ATR secondary coolant system to preheat air for space heating of TRA-670. A tertiary coolant stream would be extracted from the secondary coolant system loop and pumped to a new plate and frame heat exchanger, where heat would be transferred to a glycol loop for preheating outdoor air in the heating and ventilation system. Historical data from Advanced Test Reactor operations over the past 10 years indicates that heat from the reactor coolant was available (when needed for heating) for 43.5% of the year on average. Potential energy cost savings by using the waste heat to preheat intake air is $242K/yr. Technical, safety, and logistics considerations of the glycol waste heat recovery system are outlined. Other opportunities for using waste heat and reducing water usage at ATR are considered.

  7. NRC review of Electric Power Research Institute's Advanced Light Reactor Utility Requirements Document - Program summary, Project No. 669

    Energy Technology Data Exchange (ETDEWEB)

    1992-08-01

    The staff of the US Nuclear Regulatory Commission has prepared Volume 1 of a safety evaluation report (SER), NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document -- Program Summary,'' to document the results of its review of the Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document.'' This SER provides a discussion of the overall purpose and scope of the Requirements Document, the background of the staff's review, the review approach used by the staff, and a summary of the policy and technical issues raised by the staff during its review.

  8. The development and the tests of the electrostatic probe for dust particle collection in thermonuclear reactors

    Science.gov (United States)

    Begrambekov, L. B.; Voityuk, A. N.; Zakharov, A. M.

    2016-09-01

    Formation of dust particles in thermonuclear reactors can greatly affect the plasma parameters and lead to accumulation of tritium. The rates of formation and deposition of dust need to be measured, and the parameters of formation of dust particles and clusters need to be studied. A model of a device for collection of fine conductive particles capable of removing them from the reactor chamber for future research is proposed in this paper. The dust collector's operation is based on a principle of applied electrostatic field. The model was tested in different operating conditions: in vacuum, at the atmospheric pressure in the atmosphere of air and dry nitrogen. The experiments were conducted with a stationary system and with the dust collector in motion relative to the dusty surface. It is shown that, during the probe moving relative to the surface, it can remove up to 95% of fine tungsten particles with sizes ranging from 1 to 10 μm.

  9. Operational Philosophy for the Advanced Test Reactor National Scientific User Facility

    Energy Technology Data Exchange (ETDEWEB)

    J. Benson; J. Cole; J. Jackson; F. Marshall; D. Ogden; J. Rempe; M. C. Thelen

    2013-02-01

    In 2007, the Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF). At its core, the ATR NSUF Program combines access to a portion of the available ATR radiation capability, the associated required examination and analysis facilities at the Idaho National Laboratory (INL), and INL staff expertise with novel ideas provided by external contributors (universities, laboratories, and industry). These collaborations define the cutting edge of nuclear technology research in high-temperature and radiation environments, contribute to improved industry performance of current and future light-water reactors (LWRs), and stimulate cooperative research between user groups conducting basic and applied research. To make possible the broadest access to key national capability, the ATR NSUF formed a partnership program that also makes available access to critical facilities outside of the INL. Finally, the ATR NSUF has established a sample library that allows access to pre-irradiated samples as needed by national research teams.

  10. Electric vehicle chassis dynamometer test methods at JPL and their correlation to track tests

    Science.gov (United States)

    Marte, J.; Bryant, J.

    1983-01-01

    Early in its electric vehicle (EV) test program, JPL recognized that EV test procedures were too vague and too loosely defined to permit much meaningful data to be obtained from the testing. Therefore, JPL adopted more stringent test procedures and chose the chassis dynamometer rather than the track as its principal test technique. Through the years, test procedures continued to evolve towards a methodology based on chassis dynamometers which would exhibit good correlation with track testing. Based on comparative dynamometer and track test results on the ETV-1 vehicle, the test methods discussed in this report demonstrate a means by which excellent track-to-dynamometer correlation can be obtained.

  11. A Generalized Model for Transport of Contaminants in Soil by Electric Fields

    DEFF Research Database (Denmark)

    Paz-Garcia, Juan Manuel; Baek, Kitae; Alshawabkeh, Iyad D.

    2012-01-01

    A generalized model applicable to soils contaminated with multiple species under enhanced boundary conditions during treatment by electric fields is presented. The partial differential equations describing species transport are developed by applying the law of mass conservation to their fluxes...

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

    Energy Technology Data Exchange (ETDEWEB)

    Blaine Grover

    2012-10-01

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

  13. Industrial Hygiene Concerns during the Decontamination and Decommissioning of the Tokamak Fusion Test Reactor

    CERN Document Server

    Lumia, M E

    2002-01-01

    A significant industrial hygiene concern during the Decontamination and Decommissioning (D and D) of the Tokamak Fusion Test Reactor (TFTR) was the oxidation of the lead bricks' surface, which were utilized for radiation shielding. This presented both airborne exposure and surface contamination issues for the workers in the field removing this material. This paper will detail the various protection and control methods tested and implemented to protect the workers, including those technologies deployed to decontaminate the work surfaces. In addition, those techniques employed to recycle the lead for additional use at the site will be discussed.

  14. 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

    2004-12-31

    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.

  15. Bayesian Zero-Failure (BAZE) reliability demonstration testing procedure for components of nuclear reactor safety systems

    Energy Technology Data Exchange (ETDEWEB)

    Waller, R.A.

    1977-06-01

    A Bayesian-Zero-Failure (BAZE) reliability demonstration testing procedure is presented. The method is developed for an exponential failure-time model and a gamma prior distribution on the failure-rate. A simple graphical approach using percentiles is used to fit the prior distribution. The procedure is given in an easily applied step-by-step form which does not require the use of a computer for its implementation. The BAZE approach is used to obtain sample test plans for selected components of nuclear reactor safety systems.

  16. Supported Pd-Au Membrane Reactor for Hydrogen Production: Membrane Preparation, Characterization and Testing

    OpenAIRE

    Adolfo Iulianelli; Marjan Alavi; Giuseppe Bagnato; Simona Liguori; Jennifer Wilcox; Mohammad Reza Rahimpour; Reza Eslamlouyan; Bryce Anzelmo; Angelo Basile

    2016-01-01

    A supported Pd-Au (Au 7wt%) membrane was produced by electroless plating deposition. Permeation tests were performed with pure gas (H2, H2, N2, CO2, CH4) for long time operation. After around 400 h under testing, the composite Pd-Au membrane achieved steady state condition, with an H2/N2 ideal selectivity of around 500 at 420 °C and 50 kPa as transmembrane pressure, remaining stable up to 1100 h under operation. Afterwards, the membrane was allocated in a membrane reactor module for methane s...

  17. Observed Changes in As-Fabricated U-10Mo Monolithic Fuel Microstructures After Irradiation in the Advanced Test Reactor

    Science.gov (United States)

    Keiser, Dennis; Jue, Jan-Fong; Miller, Brandon; Gan, Jian; Robinson, Adam; Madden, James

    2017-08-01

    A low-enriched uranium U-10Mo monolithic nuclear fuel is being developed by the Material Management and Minimization Program, earlier known as the Reduced Enrichment for Research and Test Reactors Program, for utilization in research and test reactors around the world that currently use high-enriched uranium fuels. As part of this program, reactor experiments are being performed in the Advanced Test Reactor. It must be demonstrated that this fuel type exhibits mechanical integrity, geometric stability, and predictable behavior to high powers and high fission densities in order for it to be a viable fuel for qualification. This paper provides an overview of the microstructures observed at different regions of interest in fuel plates before and after irradiation for fuel samples that have been tested. These fuel plates were fabricated using laboratory-scale fabrication methods. Observations regarding how microstructural changes during irradiation may impact fuel performance are discussed.

  18. Economic aspects of electricity and industrial heat generating reactors; Aspect economique des reacteurs produisant de l'electricite et de la chaleur industrielle

    Energy Technology Data Exchange (ETDEWEB)

    Gaussens, J.; Moulle, N.; Dutheil, F. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Aldebert, J. [Institut National des Sciences et Techniques Nucleaires (INSTN), CEA Saclay, 91 - Gif sur Yvette (France)

    1964-07-01

    The economic advantage of electricity-generating nuclear stations decreases when their size decreases. However, when a counter-pressure turbine is joined on to a reactor and the residual heat can be properly used, it can be shown that fairly low capacity nuclear equipment may compete with conventional equipment under certain realistic enough conditions. The aim of this paper is to define these special conditions under which nuclear energy can be profitable. They are connected with the location and the general economic environment of the station, the pattern of the electricity and heat demands it must meet, the level of fuel and specific capital costs, nuclear and conventional. These conditions entail certain technical and economic specifications for the reactors used in this way otherwise they are unlikely to be competitive. In addition, these results are referred to the potential steam and electricity market, which leads us to examine certain uses for the heat generated by double purpose power stations; for example, to supply combined industrial plants, various types of town heating and for removal of salt from sea water. (authors) [French] L'interet economique de centrales nucleaires productrices d'electricite decroit lorsque la puissance decroit. Cependant, lorsqu'on associe une turbine a contrepression a un reacteur et qu'il est possible d'utiliser dans de bonnes conditions la chaleur residuelle, on peut montrer que dans certaines conditions assez realistes, des equipements nucleaires d'une puissance unitaire peu elevee peuvent etre competitifs avec des equipements conventionnels. Cette communication a donc pour but de mettre en evidence quelles sont ces conditions particulieres de rentabilite de l'energie nucleaire. Elles sont liees a la localisation de la centrale et a son contexte economique general, a la structure de la demande d'energie electrique et thermique a laquelle elle doit satisfaire, au niveau des couts des

  19. Conformal invariance and generalized Hojman conserved quantities of mechanico-electrical systems

    Institute of Scientific and Technical Information of China (English)

    Li Yuan-Cheng; Xia Li-Li; Wang Xiao-Ming

    2009-01-01

    This paper studies conformal invariance and generalized Hojman conserved quantities of mechanico-electrical systems. The definition and the determining equation of conformal invariance for mechanico-electrical systems are provided. The conformal factor expression is deduced from conformal invariance and Lie symmetry under the infinitesimal singleparameter transformation group. The generalized Hojman conserved quantities from the conformal invariance of the system are given. An example is given to illustrate the application of the result.

  20. General theories for the electrical transport properties of carbon nanotubes

    Energy Technology Data Exchange (ETDEWEB)

    Singh, L T; Nanda, K K, E-mail: nanda@mrc.iisc.ernet.in [Materials Research Centre, Indian Institute of Science, Bangalore-560012 (India)

    2011-08-05

    We have shown that the general theories of metals and semiconductors can be employed to understand the diameter and voltage dependency of current through metallic and semiconducting carbon nanotubes, respectively. The current through a semiconducting multiwalled carbon nanotube (MWCNT) is associated with the energy gap that is different for different shells. The contribution of the outermost shell is larger as compared to the inner shells. The general theories can also explain the diameter dependency of maximum current through nanotubes. We have also compared the current carrying ability of a MWCNT and an array of the same diameter of single wall carbon nanotubes (SWCNTs) and found that MWCNTs are better suited and deserve further investigation for possible applications as interconnects.

  1. Fusion Reactor Materials

    Energy Technology Data Exchange (ETDEWEB)

    Decreton, M

    2000-07-01

    SCK-CEN's research and development programme on fusion reactor materials includes: (1) the study of the mechanical behaviour of structural materials under neutron irradiation (including steels, inconel, molybdenum, chromium); (2) the determination and modelling of the characteristics of irradiated first wall materials such as beryllium; (3) the detection of abrupt electrical degradation of insulating ceramics under high temperature and neutron irradiation; (4) the study of the dismantling and waste disposal strategy for fusion reactors.; (5) a feasibility study for the testing of blanket modules under neutron radiation. Main achievements in these topical areas in the year 1999 are summarised.

  2. Improved Generalized Predictive Control Algorithm with Offline and Online Identification and Its Application to Fixed Bed Reactor

    Institute of Scientific and Technical Information of China (English)

    余世明; 王海清

    2003-01-01

    An improved generalized predictive control algorithm is presented in this paper by incorporating offline identification into onlie identification.Unlike the existing generalized predictive control algorithms.the proposed approach divides parameters of a predictive model into the time invariant and time-varying ones,which are treated respectively by offline and onlie identification algorithms.Therefore,both the reliability and accuracy of the predictive model are improved,Two simulation examples of control of a fixed bed reactor show that this new algorithm is not only reliable and stable in the case of uncertainties and abnormal distrubances,but also adaptable to slow time varying processes.

  3. Report on emergency electrical power supply systems for nuclear fuel cycle and reactor facilities security systems

    Energy Technology Data Exchange (ETDEWEB)

    1977-01-01

    The report includes information that will be useful to those responsible for the planning, design and implementation of emergency electric power systems for physical security and special nuclear materials accountability systems. Basic considerations for establishing the system requirements for emergency electric power for security and accountability operations are presented. Methods of supplying emergency power that are available at present and methods predicted to be available in the future are discussed. The characteristics of capacity, cost, safety, reliability and environmental and physical facility considerations of emergency electric power techniques are presented. The report includes basic considerations for the development of a system concept and the preparation of a detailed system design.

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

    Energy Technology Data Exchange (ETDEWEB)

    Donna Post Guillen

    2012-11-01

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

  5. IMPROVED COMPUTATIONAL NEUTRONICS METHODS AND VALIDATION PROTOCOLS FOR THE ADVANCED TEST REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    David W. Nigg; Joseph W. Nielsen; Benjamin M. Chase; Ronnie K. Murray; Kevin A. Steuhm

    2012-04-01

    The Idaho National Laboratory (INL) is in the process of modernizing the various reactor physics modeling and simulation tools used to support operation and safety assurance of the Advanced Test Reactor (ATR). Key accomplishments so far have encompassed both computational as well as experimental work. A new suite of stochastic and deterministic transport theory based reactor physics codes and their supporting nuclear data libraries (HELIOS, KENO6/SCALE, NEWT/SCALE, ATTILA, and an extended implementation of MCNP5) has been installed at the INL. Corresponding models of the ATR and ATRC are now operational with all five codes, demonstrating the basic feasibility of the new code packages for their intended purpose. Of particular importance, a set of as-run core depletion HELIOS calculations for all ATR cycles since August 2009 was successfully completed during 2011. This demonstration supported a decision late in the year to proceed with the phased incorporation of the HELIOS methodology into the ATR fuel cycle management process beginning in 2012. On the experimental side of the project, new hardware was fabricated, measurement protocols were finalized, and the first four of six planned physics code validation experiments based on neutron activation spectrometry were conducted at the ATRC facility. Data analysis for the first three experiments, focused on characterization of the neutron spectrum in one of the ATR flux traps, has been completed. The six experiments will ultimately form the basis for a flexible, easily-repeatable ATR physics code validation protocol that is consistent with applicable ASTM standards.

  6. Measurement of basic thermal-hydraulic characteristics under the test facility and reactor conditions

    Energy Technology Data Exchange (ETDEWEB)

    Eduard A Boltenko; Victor P Sharov [Elektrogorsk Research and Engineering Center, EREC, Bezimyannaja Street, 6, Elektrogorsk, Moscow Region, 142530 (Russian Federation); Dmitriy E Boltenko [State Scientific Center of Russian Federation IPPE, Bondarenko Square, Obhinsk, Kaluga Region, 249020 (Russian Federation)

    2005-07-01

    Full text of publication follows: The nuclear power of Russia is based on the reactors of two types: water-water - WWER and uranium - graphite channel RBMK. The nuclear power development is possible with performance of the basic condition - level of nuclear power plants (NPP) safety should satisfy the rigid requirements. The calculated proof of NPPs safety made by means of thermal-hydraulic codes of improved estimation, verified on experimental data is the characteristic of this level. The data for code verification can be obtained at the integral facilities simulating a circulation circuit of NPP with the basic units and intended for investigation of circuit behaviour in transient and accident conditions. For verification of mathematical models in transient and accident conditions, development of physically reasonable methods for definition of the various characteristics of two-phase flow the experimental data, as the integrated characteristics of a flow, and data on the local characteristics and structure of a flow is necessary. For safety assurance of NPP it is necessary to monitor and determine the basic thermalhydraulic characteristics of reactor facility (RF). It is possible to refer coolant flow-rate, core input and output water temperature, heat-power. The description of the EREC works in the field completion and adaptation of certain methods with reference to measurements in dynamic modes of test facility conditions and development of methods for measurements of basic thermal-hydraulic characteristics of reactor facilities is presented in the paper. (authors)

  7. Bench-scale reactor tests of low-temperature, catalytic gasification of wet, industrial wastes

    Energy Technology Data Exchange (ETDEWEB)

    Elliott, D.C.; Neuenschwander, G.G.; Baker, E.G.; Butner, R.S.; Sealock, L.J.

    1990-04-01

    Bench-scale reactor tests are under way at Pacific Northwest Laboratory to develop a low-temperature, catalytic gasification system. The system, licensed under the trade name Thermochemical Environmental Energy System (TEES{reg sign}), is designed for to a wide variety of feedstocks ranging from dilute organics in water to waste sludges from food processing. The current research program is focused on the use of a continuous-feed, tubular reactor. The catalyst is nickel metal on an inert support. Typical results show that feedstocks such as solutions of 2% para-cresol or 5% and 10% lactose in water or cheese whey can be processed to >99% reduction of chemical oxygen demand (COD) at a rate of up to 2 L/hr. The estimated residence time is less than 5 min at 360{degree}C and 3000 psig, not including 1 to 2 min required in the preheating zone of the reactor. The liquid hourly space velocity has been varied from 1.8 to 2.9 L feedstock/L catalyst/hr depending on the feedstock. The product fuel gas contains 40% to 55% methane, 35% to 50% carbon dioxide, and 5% to 10% hydrogen with as much as 2% ethane, but less than 0.1% ethylene or carbon monoxide, and small amounts of higher hydrocarbons. The byproduct water stream carries residual organics amounting to less than 500 mg/L COD. 9 refs., 1 fig., 4 tabs.

  8. Dissolution of Material and Test reactor Fuel in an H-Canyon Dissolver

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-01-26

    In an amended record of decision for the management of spent nuclear fuel (SNF) at the Savannah River Site, the US Department of Energy has authorized the dissolution and recovery of U from 1000 bundles of Al-clad SNF. The SNF is fuel from domestic and foreign research reactors and is typically referred to as Material Test Reactor (MTR) fuel. Bundles of MTR fuel containing assemblies fabricated from U-Al alloys (or other U compounds) are currently dissolved using a Hg-catalyzed HNO3 flowsheet. Since the development of the existing flowsheet, improved experimental methods have been developed to more accurately characterize the offgas composition and generation rate during laboratory dissolutions. Recently, these new techniques were successfully used to develop a flowsheet for the dissolution of High Flux Isotope Reactor (HFIR) fuel. Using the data from the HFIR dissolution flowsheet development and necessary laboratory experiments, the Savannah River National Laboratory (SRNL) was requested to define flowsheet conditions for the dissolution of MTR fuels. With improved offgas characterization techniques, SRNL will be able define the number of bundles of fuel which can be charged to an H-Canyon dissolver with much less conservatism.

  9. Four-terminal electrical testing device. [initiator bridgewire resistance

    Science.gov (United States)

    Robinson, Robert L. (Inventor); Graves, Thomas J. (Inventor); Hoffman, William C., III (Inventor)

    1987-01-01

    The invention relates to a four-terminal electrical connector device for testing and measuring unknown resistances of initiators used for starting pyrotechnic events aboard the space shuttle. The testing device minimizes contact resistance degradation effects and so improves the reliability of resistance measurements taken with the device. Separate and independent voltage sensing and current supply circuits each include a pair of socket contacts for mating engagement with the pins of the initiator. The unknown resistance that is measured by the device is the resistance of the bridgewire of the initiator which is required to be between 0.95 and 1.15 ohms.

  10. Baseline tests of the EPC Hummingbird electric passenger vehicle

    Science.gov (United States)

    Slavik, R. J.; Maslowski, E. A.; Sargent, N. B.; Birchenough, A. G.

    1977-01-01

    The rear-mounted internal combustion engine in a four-passenger Volkswagen Thing was replaced with an electric motor made by modifying an aircraft generator and powered by 12 heavy-duty, lead-acid battery modules. Vehicle performance tests were conducted to measure vehicle maximum speed, range at constant speed, range over stop-and-go driving schedules, maximum acceleration, gradeability limit, road energy consumption, road power, indicated energy consumption, braking capability, battery charger efficiency, and battery characteristics. Test results are presented in tables and charts.

  11. Generalized M-Fluctuation Tests for Parameter Instability

    OpenAIRE

    Zeileis, Achim; Hornik, Kurt

    2003-01-01

    A general class of fluctuation tests for parameter instability in an M-estimation framework is suggested. The tests are based on partial sum processes of M-estimation scores for which functional central limit theorems are derived under the null hypothesis of parameter stability and local alternatives. Special emphasis is given to parameter instability in (generalized) linear regression models and it is shown that the introduced M-fluctuation tests contain a large number of parameter instabili...

  12. Submersible microbial desalination cell for simultaneous ammonia recovery and electricity production from anaerobic reactors containing high levels of ammonia.

    Science.gov (United States)

    Zhang, Yifeng; Angelidaki, Irini

    2015-02-01

    High ammonia concentration in anaerobic reactors can seriously inhibit the anaerobic digestion process. In this study, a submersible microbial desalination cell (SMDC) was developed as an innovative method to lower the ammonia level in a continuous stirred tank reactor (CSTR) by in situ ammonia recovery and electricity production. In batch experiment, the ammonia concentration in the CSTR decreased from 6 to 0.7 g-N/L during 30 days, resulting in an average recovery rate of 80 g-N/m(2)/d. Meanwhile, a maximum power density of 0.71±0.5 W/m(2) was generated at 2.85 A/m(2). Both current driven NH4(+) migration and free NH3 diffusion were identified as the mechanisms responsible for the ammonia transportation. With an increase in initial ammonia concentration and a decrease in external resistance, the SMDC performance was enhanced. In addition, the coexistence of other cations in CSTR or cathode had no negative effect on the ammonia transportation. Copyright © 2014 Elsevier Ltd. All rights reserved.

  13. Reference site selection report for the advanced liquid metal reactor at the Idaho National Engineering Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Sivill, R.L.

    1990-03-01

    This Reference Site Selection Report was prepared by EG G, Idaho Inc., for General Electric (GE) to provide information for use by the Department of Energy (DOE) in selecting a Safety Test Site for an Advanced Liquid Metal Reactor. Similar Evaluation studies are planned to be conducted at other potential DOE sites. The Power Reactor Innovative Small Module (PRISM) Concept was developed for ALMR by GE. A ALMR Safety Test is planned to be performed on a DOE site to demonstrate features and meet Nuclear Regulatory Commission Requirements. This study considered possible locations at the Idaho National Engineering Laboratory that met the ALMR Prototype Site Selection Methodology and Criteria. Four sites were identified, after further evaluation one site was eliminated. Each of the remaining three sites satisfied the criteria and was graded. The results were relatively close. Thus concluding that the Idaho National Engineering Laboratory is a suitable location for an Advanced Liquid Metal Reactor Safety Test. 23 refs., 13 figs., 9 tabs.

  14. 3D thermal hydraulic simulation of the hot channel of a typical material testing reactor under normal operation conditions

    Energy Technology Data Exchange (ETDEWEB)

    El-Morshedy, Salah El-Din; Salama, Amgad [Atomic Energy Authority, Cairo (Egypt). Reactors Dept.

    2010-09-15

    The hot channel in a typical Material Testing Reactor (MTR) is subjected to 3D simulation. Because of the existence of similarity planes, only a quarter of the hot channel including meat thickness, clad, and coolant channel is considered for CFD analysis using the FLUENT code. For the simulation, steady state normal operation regime at the reactor nominal power is assumed. In order to build confidence in our modeling approach, the results obtained in this work are compared with those obtained from the one-dimensional simulation code, MTRTHA. That is, modified variables were generated in order to match those obtained by MTRTHA and to allow comparisons. Quite good agreement is generally observed, however, the maximum clad surface temperature predicted by the 3D calculations, located at the clad mid-width, is higher than the 1D prediction by about 8 C but still below the onset of subcooled boiling by adequate safety margin. The results show quite interesting 3D patterns in both the flow field and the heat transfer. Temperature profiles, velocity profiles and contours are all presented to highlight the essential 3D features of this system. (orig.)

  15. Criticality safety evaluation for the Advanced Test Reactor enhanced low enriched uranium fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Montierth, Leland M. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-07-19

    The Global Threat Reduction Initiative (GTRI) convert program is developing a high uranium density fuel based on a low enriched uranium (LEU) uranium-molybdenum alloy. Testing of prototypic GTRI fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. GTRI Enhanced LEU Fuel (ELF) elements based on the ATR-Standard Size elements (all plates fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). While a specific ELF element design will eventually be provided for detailed analyses and in-core testing, this criticality safety evaluation (CSE) is intended to evaluate a hypothetical ELF element design for criticality safety purposes. Existing criticality analyses have analyzed Standard (HEU) ATR elements from which controls have been derived. This CSE documents analysis that determines the reactivity of the hypothetical ELF fuel elements relative to HEU ATR elements and whether the existing HEU ATR element controls bound the ELF element. The initial calculations presented in this CSE analyzed the original ELF design, now referred to as Mod 0.1. In addition as part of a fuel meat thickness optimization effort for reactor performance other designs have been evaluated. As of early 2014 the most current conceptual designs are Mk1A and Mk1B that were previously referred to as conceptual designs Mod 0.10 and Mod 0.11, respectively. Revision 1 evaluates the reactivity of the ATR HEU Mark IV elements for a comparison with the Mark VII elements.

  16. A review on the utilization of the Japan materials testing reactor (JMTR)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D. H.; Kang, Y. H.; Kim, B. G.; Choo, K. N.; Oh, J. M.; Park, S. J.; Shin, Y. T

    1999-04-01

    The HANARO has possessed the potential capability for the testing of materials and fuels since the beginning of its operation in 1995. Recently, this reactor has contributed to various activities in nuclear power research in Korea. We need the recent technical data of developed countries to support these activities in nuclear power. Most of the developed countries in nuclear power have more than thirty years' experience in the irradiation test of nuclear fuel and material for performing their complicated in-core measurements of the change of material properties. They also have developed various types of sensors, equipment and techniques. This report describes the status of utilization of the irradiation facilities of the Japan Materials Testing Reactor(JMTR). It also describes the recent efforts of the JMTR in order to develop new irradiation test techniques. It will be our great pleasure for this report to help a broad range of people understand the generic contents (JMTR utilization, new techniques) of the JMTR. (author)

  17. Criticality safety evaluation for the Advanced Test Reactor enhanced low enriched uranium fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Montierth, Leland M. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-07-19

    The Global Threat Reduction Initiative (GTRI) convert program is developing a high uranium density fuel based on a low enriched uranium (LEU) uranium-molybdenum alloy. Testing of prototypic GTRI fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. GTRI Enhanced LEU Fuel (ELF) elements based on the ATR-Standard Size elements (all plates fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). While a specific ELF element design will eventually be provided for detailed analyses and in-core testing, this criticality safety evaluation (CSE) is intended to evaluate a hypothetical ELF element design for criticality safety purposes. Existing criticality analyses have analyzed Standard (HEU) ATR elements from which controls have been derived. This CSE documents analysis that determines the reactivity of the hypothetical ELF fuel elements relative to HEU ATR elements and whether the existing HEU ATR element controls bound the ELF element. The initial calculations presented in this CSE analyzed the original ELF design, now referred to as Mod 0.1. In addition, as part of a fuel meat thickness optimization effort for reactor performance, other designs have been evaluated. As of early 2014 the most current conceptual designs are Mk1A and Mk1B, that were previously referred to as conceptual designs Mod 0.10 and Mod 0.11, respectively. Revision 1 evaluates the reactivity of the ATR HEU Mark IV elements for a comparison with the Mark VII elements.

  18. Battery Test Manual For Electric Vehicles, Revision 3

    Energy Technology Data Exchange (ETDEWEB)

    Christophersen, Jon P. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-06-01

    This battery test procedure manual was prepared for the United States Department of Energy (DOE), Office of Energy Efficiency and Renewable Energy (EERE), Vehicle Technologies Office. It is based on technical targets for commercial viability established for energy storage development projects aimed at meeting system level DOE goals for Electric Vehicles (EV). The specific procedures defined in this manual support the performance and life characterization of advanced battery devices under development for EVs. However, it does share some methods described in the previously published battery test manual for plug-in hybrid electric vehicles. Due to the complexity of some of the procedures and supporting analysis, future revisions including some modifications and clarifications of these procedures are expected. As in previous battery and capacitor test manuals, this version of the manual defines testing methods for full-size battery systems, along with provisions for scaling these tests for modules, cells or other subscale level devices. The DOE-United States Advanced Battery Consortium (USABC), Technical Advisory Committee (TAC) supported the development of the manual. Technical Team points of contact responsible for its development and revision are Chul Bae of Ford Motor Company and Jon P. Christophersen of the Idaho National Laboratory. The development of this manual was funded by the Unites States Department of Energy, Office of Energy Efficiency and Renewable Energy, Vehicle Technologies Office. Technical direction from DOE was provided by David Howell, Energy Storage R&D Manager and Hybrid Electric Systems Team Leader. Comments and questions regarding the manual should be directed to Jon P. Christophersen at the Idaho National Laboratory (jon.christophersen@inl.gov).

  19. Status of the Combined Third and Fourth NGNP Fuel Irradiations In the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover; David A. Petti; Michael E. Davenport

    2013-07-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The experiments will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of several independent capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and is currently scheduled to be completed in September 2013. The third and fourth experiments have been combined into a single experiment designated (AGR-3/4), which started its irradiation in December 2011 and is currently scheduled to be completed in April 2014. Since the purpose of this combined experiment is to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment is

  20. DESIGN CHARACTERISTICS OF THE IDAHO NATIONAL LABORATORY HIGH-[TEMPERATURE GAS-COOLED TEST REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James; Bayless, Paul; Strydom, Gerhard; Kumar, Akansha; Gougar, Hans

    2016-11-01

    A point design for a graphite-moderated, high-temperature, gas-cooled test reactor (HTG TR) has been developed by Idaho National Laboratory (INL) as part of a United States (U.S.) Department of Energy (DOE) initiative to explore and potentially expand the existing U.S. test reactor capabilities. This paper provides a summary of the design and its main attributes. The 200 MW HTG TR is a thermal-neutron spectrum reactor composed of hexagonal prismatic fuel and graphite reflector blocks. Twelve fuel columns (96 fuel blocks total and 6.34 m active core height) are arranged in two hexagonal rings to form a relatively compact, high-power density, annular core sandwiched between inner, outer, top, and bottom graphite reflectors. The HTG-TR is designed to operate at 7 MPa with a coolant inlet/outlet temperature of 325°C/650°C, and utilizes TRISO particle fuel from the DOE AGR Program with 425 ?m uranium oxycarbide (UCO) kernels and an enrichment of 15.5 wt% 235U. The primary mission of the HTG TR is material irradiation and therefore the core has been specifically designed and optimized to provide the highest possible thermal and fast neutron fluxes. The highest thermal neutron flux (3.90E+14 n/cm2s) occurs in the outer reflector, and the maximum fast flux levels (1.17E+14 n/cm2s) are produced in the central reflector column where most of the graphite has been removed. Due to high core temperatures under accident conditions, all the irradiation test facilities have been located in the inner and outer reflectors where fast flux levels decline. The core features a large number of irradiation positions with large test volumes and long test lengths, ideal for thermal neutron irradiation of large test articles. The total available test volume is more than 1100 liters. Up to four test loop facilities can be accommodated with pressure tube boundaries to isolate test articles and test fluids (e.g., liquid metal, liquid salt, light water) from the helium primary coolant system.

  1. Classical tests of general relativity in brane world models

    CERN Document Server

    Boehmer, Christian G; Harko, Tiberiu; Lobo, Francisco S N

    2009-01-01

    The classical tests of general relativity (perihelion precession, deflection of light, and the radar echo delay) are considered for several spherically symmetric static vacuum solutions in brane world models. Generally, the spherically symmetric vacuum solutions of the brane gravitational field equations have properties quite distinct as compared to the standard black hole solutions of general relativity. As a first step a general formalism that facilitates the analysis of general relativistic Solar System tests for any given spherically symmetric metric is developed. It is shown that the existing observational Solar System data on the perihelion shift of Mercury, on the light bending around the Sun (obtained using long-baseline radio interferometry), and ranging to Mars using the Viking lander, constrain the numerical values of the parameters of the specific models. Hence Solar System tests represent very convenient and efficient tools to test the viability of the different black hole solutions in brane worl...

  2. Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production, Progress Report for Work Through September 2003, 2nd Annual/8th Quarterly Report

    Energy Technology Data Exchange (ETDEWEB)

    Philip E. MacDonald

    2003-09-01

    The supercritical water-cooled reactor (SCWR) is one of the six reactor technologies selected for research and development under the Generation-IV program. SCWRs are promising advanced nuclear systems because of their high thermal efficiency (i.e., about 45% vs. about 33% efficiency for current Light Water Reactors, LWRs) and considerable plant simplification. SCWRs are basically LWRs operating at higher pressure and temperatures with a direct once-through cycle. Operation above the critical pressure eliminates coolant boiling, so the coolant remains single-phase throughout the system. Thus the need for recirculation and jet pumps, a pressurizer, steam generators, steam separators and dryers is eliminated. The main mission of the SCWR is generation of low-cost electricity. It is built upon two proven technologies, LWRs, which are the most commonly deployed power generating reactors in the world, and supercritical fossil-fired boilers, a large number of which is also in use around the world.

  3. Creep-Fatigue Damage Evaluation of a Model Reactor Vessel and Reactor Internals of Sodium Test Facility according to ASME-NH and RCC-MRx Codes

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Dong-Won; Lee, Hyeong-Yeon; Eoh, Jae-Hyuk; Son, Seok-Kwon; Kim, Jong-Bum; Jeong, Ji-Young [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The objective of the STELLA-2 is to support the specific design approval for PGSFR by synthetic reviews of key safety issues and code validations through the integral effect tests. Due to its high temperature operation in SFRs (and in a testing facility) up to 550 °C, thermally induced creep-fatigue damage is very likely in components including a reactor vessel, reactor internals (interior structures), heat exchangers, pipelines, etc. In this study, structural integrity of the components such as reactor vessel and internals in STELLA-2 has been evaluated against creep-fatigue failures at a concept-design step. As 2D analysis yields far conservative results, a realistic 3D simulation is performed by a commercial software. A design integrity guarding against a creep-fatigue damage failure operating at high temperature was evaluated for the reactor vessel with its internal structure of the STELLA-2. Both the high temperature design codes were used for the evaluation, and results were compared. All the results showed the vessel as a whole is safely designed at the given operating conditions, while the ASME-NH gives a conservative evaluation.

  4. Application of the small punch test to reactor pressure vessel integrity

    Energy Technology Data Exchange (ETDEWEB)

    Rosinski, S.T. [EPRI Nondestructive Evaluation Center, Charlotte, NC (United States); Viswanathan, R. [EPRI, Palo Alto, CA (United States); Foulds, J.R. [Failure Analysis Associates, Inc., Menlo Park, CA (United States)

    1998-07-01

    Based on prior success with fossil plant steels, EPRI is investigating the feasibility of applying the Small Punch test to determine the fracture toughness (K{sub ic}) of irradiated reactor pressure vessel (RPV) materials. The small punch test specimen is sufficiently small to alleviate future surveillance material availability concerns, as well as provide a means of direct vessel material interrogation by non-disruptive miniature sample removal and testing. A limited series of small punch tests on unirradiated and irradiated RPV steel materials has shown that the method can be used to estimate ductile-to-brittle transition temperatures and to determine the material fracture toughness (K{sub lc}, J{sub lc}). The results to date are described and the experimental difficulties that need to be resolved in achieving valid results are identified. (authors)

  5. Command control of reactors and factories: general architecture; Controle-commande des reacteurs et des usines: architecture generale

    Energy Technology Data Exchange (ETDEWEB)

    Appell, B.; Guesnier, G. [Electricite de France, 75 - Paris (France). Service Etudes et Projets Thermiques et Nucleaires; Chabert, J. [Cogema, 78 - Velizy-Villacoublay (France)

    1998-07-01

    As any industrial installation, the nuclear power plants and the fuel reprocessing plants require means to survey and to command the physical process and the associated equipment. These means are grouped under the designation of `command-control`. The command control is constituted by captors allowing to transform the physical quantities in electric signals, by automates allowing to treat these signals, by surveillance and control means at operators disposal and finally activators allowing to transform the command electric signals in mechanical actions on the process. The equipment has to answer to specifications imposed by nuclear safety. (N.C.)

  6. 60th Anniversary of electricity production from light water reactors: Historical review of the contribution of materials science to the safety of the pressure vessel

    Science.gov (United States)

    van Duysen, J. C.; Meric de Bellefon, G.

    2017-02-01

    The first light water nuclear reactor dedicated to electricity production was commissioned in Shippingport, Pennsylvania in the United States in 1957. Sixty years after the event, it is clear that this type of reactor will be a major source of electricity and one of the key solutions to limit climate change in the 21st century. This article pays homage to the teams that contributed to this achievement by their involvement in research and development and their determination to push back the frontiers of knowledge. Via a few examples of scientific or technological milestones, it describes the evolution of ideas, models, and techniques during the last 60 years, and gives the current state-of-the-art in areas related to the safety of the reactor pressure vessel. Among other topics, it focuses on vessel manufacturing, steel fracture mechanics analysis, and understanding of irradiation-induced damage.

  7. Design of the VISTA-ITL Test Facility for an Integral Type Reactor of SMART and a Post-Test Simulation of a SBLOCA Test

    OpenAIRE

    2014-01-01

    To validate the performance and safety of an integral type reactor of SMART, a thermal-hydraulic integral effect test facility, VISTA-ITL, is introduced with a discussion of its scientific design characteristics. The VISTA-ITL was used extensively to assess the safety and performance of the SMART design, especially for its passive safety system such as a passive residual heat removal system, and to validate various thermal-hydraulic analysis codes. The VISTA-ITL program includes several tests...

  8. 40 CFR 86.230-11 - Test sequence: general requirements.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 18 2010-07-01 2010-07-01 false Test sequence: general requirements. 86.230-11 Section 86.230-11 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR... New Medium-Duty Passenger Vehicles; Cold Temperature Test Procedures § 86.230-11 Test...

  9. 40 CFR 86.230-94 - Test sequence: general requirements.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 18 2010-07-01 2010-07-01 false Test sequence: general requirements. 86.230-94 Section 86.230-94 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR... New Medium-Duty Passenger Vehicles; Cold Temperature Test Procedures § 86.230-94 Test...

  10. 14 CFR 33.57 - General conduct of block tests.

    Science.gov (United States)

    2010-01-01

    ... Section 33.57 Aeronautics and Space FEDERAL AVIATION ADMINISTRATION, DEPARTMENT OF TRANSPORTATION AIRCRAFT AIRWORTHINESS STANDARDS: AIRCRAFT ENGINES Block Tests; Reciprocating Aircraft Engines § 33.57 General conduct of block tests. (a) The applicant may, in conducting the block tests, use separate engines of...

  11. 14 CFR 33.99 - General conduct of block tests.

    Science.gov (United States)

    2010-01-01

    ... Section 33.99 Aeronautics and Space FEDERAL AVIATION ADMINISTRATION, DEPARTMENT OF TRANSPORTATION AIRCRAFT AIRWORTHINESS STANDARDS: AIRCRAFT ENGINES Block Tests; Turbine Aircraft Engines § 33.99 General conduct of block tests. (a) Each applicant may, in making a block test, use separate engines of identical design...

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

    CERN Document Server

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

    2012-01-01

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

  13. Technology, safety, and costs of decommissioning reference nuclear research and test reactors. Appendices

    Energy Technology Data Exchange (ETDEWEB)

    Konzek, G.J.; Ludwick, J.D.; Kennedy, W.E. Jr.; Smith, R.I.

    1982-03-01

    Safety and Cost Information is developed for the conceptual decommissioning of two representative licensed nuclear research and test reactors. Three decommissioning alternatives are studied to obtain comparisons between costs (in 1981 dollars), occupational radiation doses, potential radiation dose to the public, and other safety impacts. The alternatives considered are: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and EMTOMB (entombment). The study results are presented in two volumes. Volume 2 (Appendices) contains the detailed data that support the results given in Volume 1, including unit-component data.

  14. Evaluation of dynamic fracture toughness for Yong Gwang unit 5 reactor pressure vessel materials (Baseline Tests)

    Energy Technology Data Exchange (ETDEWEB)

    Chi Se Hwan; Kim, Joo Hag; Hong, Jun Hwa; Kwon, Sun Chil; Lee, Bong Sang [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-10-01

    The dynamic fracture toughness (K{sub d}) of intermediate shell and its weld in SA 508 CI. 3 Yong Gwang 5 reactor pressure vessel was determined and evaluated. Precracked thirty six Charpy specimens were tested by using an instrumented impact tester. The purpose of present work is to evaluate and confirm the un-irradiated dynamic fracture toughness and to provide pre-irradiation baseline data for future evaluation on dynamic fracture toughness change during operation. 18 refs., 5 figs., 5 tabs. (Author)

  15. EVA Metro Sedan electric-propulsion system: test and evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Reimers, E.

    1979-09-01

    The procedure and results of the performance evaluation of the EVA Metro Sedan (car No. 1) variable speed dc chopper motor drive and its three speed automatic transmission are presented. The propulsion system for a battery powered vehicle manufactured by Electric Vehicle Associates, Valley View, Ohio, was removed from the vehicle, mounted on the programmable electric dynamometer test facility and evaluated with the aid of a hp 3052A Data Acquisition System. Performance data for the automatic transmission, the solid state dc motor speed controller, and the dc motor in the continuous and pulsating dc power mode, as derived on the dynamometer test facility, as well as the entire propulsion system are given. This concept and the system's components were evaluated in terms of commercial applicability, maintainability, and energy utility to establish a design base for the further development of this system or similar propulsion drives. The propulsion system of the EVA Metro Sedan is powered by sixteen 6-volt traction batteries, Type EV 106 (Exide Battery Mfg. Co.). A thyristor controlled cable form Pulsomatic Mark 10 controller, actuated by a foot throttle, controls the voltage applied to a dc series field motor, rated at 10 hp at 3800 rpm (Baldor Electric Co.). Gear speed reduction to the wheel is accomplished by the original equipment three speed automatic transmission with torque converter (Renault 12 Sedan). The brake consists of a power-assisted, hydraulic braking system with front wheel disk and rear drum. An ability to recuperate electric energy with subsequent storage in the battery power supply is not provided.

  16. Development of Mini-Compact Tension Test Method for Determining Fracture Toughness Master Curves for Reactor Pressure Vessel Steels

    Energy Technology Data Exchange (ETDEWEB)

    Sokolov, Mikhail A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-05-01

    Small specimens are playing the key role in evaluating properties of irradiated materials. The use of small specimens provides several advantages. Typically, only a small volume of material can be irradiated in a reactor at desirable conditions in terms of temperature, neutron flux, and neutron dose. A small volume of irradiated material may also allow for easier handling of specimens. Smaller specimens reduce the amount of radioactive material, minimizing personnel exposures and waste disposal. However, use of small specimens imposes a variety of challenges as well. These challenges are associated with proper accounting for size effects and transferability of small specimen data to the real structures of interest. Any fracture toughness specimen that can be made out of the broken halves of standard Charpy specimens may have exceptional utility for evaluation of reactor pressure vessels (RPVs) since it would allow one to determine and monitor directly actual fracture toughness instead of requiring indirect predictions using correlations established with impact data. The Charpy V-notch specimen is the most commonly used specimen geometry in surveillance programs. Validation of the mini compact tension specimen (mini-CT) geometry has been performed on previously well characterized Midland beltline Linde 80 (WF-70) weld in the unirradiated condition. It was shown that the fracture toughness transition temperature, To, measured by these Mini-CT specimens is almost the same as To value that was derived from various larger fracture toughness specimens. Moreover, an International collaborative program has been established to extend the assessment and validation efforts to irradiated Linde 80 weld metal. The program is underway and involves the Oak Ridge National Laboratory (ORNL), Central Research Institute for Electrical Power Industry (CRIEPI), and Electric Power Research Institute (EPRI). The irradiated Mini-CT specimens from broken halves of previously tested Charpy

  17. A generalized finite difference method for modeling cardiac electrical activation on arbitrary, irregular computational meshes.

    Science.gov (United States)

    Trew, Mark L; Smaill, Bruce H; Bullivant, David P; Hunter, Peter J; Pullan, Andrew J

    2005-12-01

    A generalized finite difference (GFD) method is presented that can be used to solve the bi-domain equations modeling cardiac electrical activity. Classical finite difference methods have been applied by many researchers to the bi-domain equations. However, these methods suffer from the limitation of requiring computational meshes that are structured and orthogonal. Finite element or finite volume methods enable the bi-domain equations to be solved on unstructured meshes, although implementations of such methods do not always cater for meshes with varying element topology. The GFD method solves the bi-domain equations on arbitrary and irregular computational meshes without any need to specify element basis functions. The method is useful as it can be easily applied to activation problems using existing meshes that have originally been created for use by finite element or finite difference methods. In addition, the GFD method employs an innovative approach to enforcing nodal and non-nodal boundary conditions. The GFD method performs effectively for a range of two and three-dimensional test problems and when computing bi-domain electrical activation moving through a fully anisotropic three-dimensional model of canine ventricles.

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

    Energy Technology Data Exchange (ETDEWEB)

    MH Lane

    2006-02-15

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

  19. Completion summary for borehole USGS 136 near the Advanced Test Reactor Complex, Idaho National Laboratory, Idaho

    Science.gov (United States)

    Twining, Brian V.; Bartholomay, Roy C.; Hodges, Mary K.V.

    2012-01-01

    In 2011, the U.S. Geological Survey, in cooperation with the U.S. Department of Energy, cored and completed borehole USGS 136 for stratigraphic framework analyses and long-term groundwater monitoring of the eastern Snake River Plain aquifer at the Idaho National Laboratory. The borehole was initially cored to a depth of 1,048 feet (ft) below land surface (BLS) to collect core, open-borehole water samples, and geophysical data. After these data were collected, borehole USGS 136 was cemented and backfilled between 560 and 1,048 ft BLS. The final construction of borehole USGS 136 required that the borehole be reamed to allow for installation of 6-inch (in.) diameter carbon-steel casing and 5-in. diameter stainless-steel screen; the screened monitoring interval was completed between 500 and 551 ft BLS. A dedicated pump and water-level access line were placed to allow for aquifer testing, for collecting periodic water samples, and for measuring water levels. Geophysical and borehole video logs were collected after coring and after the completion of the monitor well. Geophysical logs were examined in conjunction with the borehole core to describe borehole lithology and to identify primary flow paths for groundwater, which occur in intervals of fractured and vesicular basalt. A single-well aquifer test was used to define hydraulic characteristics for borehole USGS 136 in the eastern Snake River Plain aquifer. Specific-capacity, transmissivity, and hydraulic conductivity from the aquifer test were at least 975 gallons per minute per foot, 1.4 × 105 feet squared per day (ft2/d), and 254 feet per day, respectively. The amount of measureable drawdown during the aquifer test was about 0.02 ft. The transmissivity for borehole USGS 136 was in the range of values determined from previous aquifer tests conducted in other wells near the Advanced Test Reactor Complex: 9.5 × 103 to 1.9 × 105 ft2/d. Water samples were analyzed for cations, anions, metals, nutrients, total organic

  20. Increase of energy efficiency of testing of traction electric machines of direct and pulsating current

    Directory of Open Access Journals (Sweden)

    A.M. Afanasov

    2015-03-01

    Full Text Available The results of the analysis of the effect of the load current of traction electric machines when tested for heating on the total electricity consumption for the test are presented. It is shown that increase of load current at the heating test permits to significantly reduce the consumption of electrical energy, and reduce the testing time without reducing its quality.

  1. General Electric ATS program technical review: Phase 2 activities

    Energy Technology Data Exchange (ETDEWEB)

    Chance, T. [GE Power Generation, Schenectady, NY (United States); Smith, D. [GE Corporate Research & Development Center, Schenectady, NY (United States)

    1995-10-01

    In response to the industrial and utility objectives specified for the ATS, the GE Power Generation ATS Phase 2 Program consisted of a dual approach. These were (1) development of an Industrial ATS (aircraft engine based) led by GE Aircraft Engines, and (2) development of a Utility ATS which was already underway at GEPG. Both programs required the identification and resolution of critical technical issues. Both systems were studied in Tasks 3-7, and both have resulted in designs that meet all ATS goals. The Industrial ATS as defined (130 MW) did not meet projected market power size requirements, and emphasis has remained on the Utility ATS development. The design and testing effort has been focused on the MS7001H combined cycle gas turbine, as the next product evolution in GE Power Generation`s product line. Common technology derived from the ATS Program is also being incorporated into the 50 Hz version of the ATS utility machine designated as the MS9001H.

  2. Advanced Test Reactor National Scientific User Facility: Addressing advanced nuclear materials research

    Energy Technology Data Exchange (ETDEWEB)

    John Jackson; Todd Allen; Frances Marshall; Jim Cole

    2013-03-01

    The Advanced Test Reactor National Scientific User Facility (ATR NSUF), based at the Idaho National Laboratory in the United States, is supporting Department of Energy and industry research efforts to ensure the properties of materials in light water reactors are well understood. The ATR NSUF is providing this support through three main efforts: establishing unique infrastructure necessary to conduct research on highly radioactive materials, conducting research in conjunction with industry partners on life extension relevant topics, and providing training courses to encourage more U.S. researchers to understand and address LWR materials issues. In 2010 and 2011, several advanced instruments with capability focused on resolving nuclear material performance issues through analysis on the micro (10-6 m) to atomic (10-10 m) scales were installed primarily at the Center for Advanced Energy Studies (CAES) in Idaho Falls, Idaho. These instruments included a local electrode atom probe (LEAP), a field-emission gun scanning transmission electron microscope (FEG-STEM), a focused ion beam (FIB) system, a Raman spectrometer, and an nanoindentor/atomic force microscope. Ongoing capability enhancements intended to support industry efforts include completion of two shielded, irradiation assisted stress corrosion cracking (IASCC) test loops, the first of which will come online in early calendar year 2013, a pressurized and controlled chemistry water loop for the ATR center flux trap, and a dedicated facility intended to house post irradiation examination equipment. In addition to capability enhancements at the main site in Idaho, the ATR NSUF also welcomed two new partner facilities in 2011 and two new partner facilities in 2012; the Oak Ridge National Laboratory, High Flux Isotope Reactor (HFIR) and associated hot cells and the University California Berkeley capabilities in irradiated materials analysis were added in 2011. In 2012, Purdue University’s Interaction of Materials

  3. Supported Pd-Au Membrane Reactor for Hydrogen Production: Membrane Preparation, Characterization and Testing.

    Science.gov (United States)

    Iulianelli, Adolfo; Alavi, Marjan; Bagnato, Giuseppe; Liguori, Simona; Wilcox, Jennifer; Rahimpour, Mohammad Reza; Eslamlouyan, Reza; Anzelmo, Bryce; Basile, Angelo

    2016-05-09

    A supported Pd-Au (Au 7wt%) membrane was produced by electroless plating deposition. Permeation tests were performed with pure gas (H₂, H₂, N₂, CO₂, CH₄) for long time operation. After around 400 h under testing, the composite Pd-Au membrane achieved steady state condition, with an H₂/N₂ ideal selectivity of around 500 at 420 °C and 50 kPa as transmembrane pressure, remaining stable up to 1100 h under operation. Afterwards, the membrane was allocated in a membrane reactor module for methane steam reforming reaction tests. As a preliminary application, at 420 °C, 300 kPa of reaction pressure, space velocity of 4100 h(-1), 40% methane conversion and 35% hydrogen recovery were reached using a commercial Ni/Al₂O₃ catalyst. Unfortunately, a severe coke deposition affected irreversibly the composite membrane, determining the loss of the hydrogen permeation characteristics of the supported Pd-Au membrane.

  4. Optimization of enhanced bioelectrical reactor with electricity from microbial fuel cells for groundwater nitrate removal.

    Science.gov (United States)

    Liu, Ye; Zhang, Baogang; Tian, Caixing; Feng, Chuanping; Wang, Zhijun; Cheng, Ming; Hu, Weiwu

    2016-01-01

    Factors influencing the performance of a continual-flow bioelectrical reactor (BER) intensified by microbial fuel cells for groundwater nitrate removal, including nitrate load, carbon source and hydraulic retention time (HRT), were investigated and optimized by response surface methodology (RSM). With the target of maximum nitrate removal and minimum intermediates accumulation, nitrate load (for nitrogen) of 60.70 mg/L, chemical oxygen demand (COD) of 849.55 mg/L and HRT of 3.92 h for the BER were performed. COD was the dominant factor influencing performance of the system. Experimental results indicated the undistorted simulation and reliable optimized values. These demonstrate that RSM is an effective method to evaluate and optimize the nitrate-reducing performance of the present system and can guide mathematical models development to further promote its practical applications.

  5. Computer Analysis of 400 HZ Aircraft Electrical Generator Test Data.

    Science.gov (United States)

    1980-06-01

    7 D-AIO0 705 AIR FORCE INST OF TECH WRIGHT-PATTERSON AlS OH SCHOO-CTC F/ /2 AOMPUTER ANALYSIS OF 400 HZ AIRCRAFT ELECTRICAL GEERATOR TES JUN 80 P G...software system to analyze the data available from the Generator-Test Facility. I - -- - --- - - - - -- I, 0 Co 4.3 CO COm- I : W z I 40 g(2) • 10 4 E- c...triangles. Figure 4 presents the basis for the following linear inter- polation used to estimate time of zero crossover. Z T = ___- (T 2-T 1 ) + T1(4

  6. Electricity price forecasting using generalized regression neural network based on principal components analysis

    Institute of Scientific and Technical Information of China (English)

    牛东晓; 刘达; 邢棉

    2008-01-01

    A combined model based on principal components analysis (PCA) and generalized regression neural network (GRNN) was adopted to forecast electricity price in day-ahead electricity market. PCA was applied to mine the main influence on day-ahead price, avoiding the strong correlation between the input factors that might influence electricity price, such as the load of the forecasting hour, other history loads and prices, weather and temperature; then GRNN was employed to forecast electricity price according to the main information extracted by PCA. To prove the efficiency of the combined model, a case from PJM (Pennsylvania-New Jersey-Maryland) day-ahead electricity market was evaluated. Compared to back-propagation (BP) neural network and standard GRNN, the combined method reduces the mean absolute percentage error about 3%.

  7. Determination of students’ general test anxiety and their test anxiety concerning English classes

    OpenAIRE

    Çiğdem Hürsen; İbrahim Tekman

    2014-01-01

    This study was conducted to determine the students’ general test anxiety and their test anxiety concerning English classes. The research was performed with 871 students. The results of the study reveal that students’ general test anxiety is in the boundaries of “rarely”. However, according to the results about students’ levels of test anxiety concerning English classes, their test anxiety level is in the boundaries of “sometimes”. Female students have a higher level of test anxiety and 15-18 ...

  8. Advanced gas cooled nuclear reactor materials evaluation and development program. Progress report, April 1--June 30, 1978

    Energy Technology Data Exchange (ETDEWEB)

    1978-08-31

    The objectives of the program are to evaluate candidate alloys for Very High Temperature Reactor Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the affect of simulated reactor primary coolant (Helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in the report includes the activities associated with the procurement of the materials for the screening test program, information from vendor certification for the materials receiver, and preliminary information from the materials characterization tests performed by General Electric. The construction status of the simulated reactor helium supply system, testing equipment, and gas chemistry analysis instrumentation and equipment are discussed. The status of the data management system is also reviewed.

  9. Status Report on Efforts to Enhance Instrumentation to Support Advanced Test Reactor Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    J. Rempe; D. Knudson; J. Daw; T. Unruh; B. Chase; R. Schley; J. Palmer; K. Condie

    2014-01-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support the growth of nuclear science and technology in the United States (US). By attracting new research users - universities, laboratories, and industry - the ATR NSUF facilitates basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort at the Idaho National Laboratory (INL) is to design, develop, and deploy new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors was completed. Based on this initial review, recommendations were made with respect to what instrumentation is needed at the ATR, and a strategy was developed for obtaining these sensors. In 2009, a report was issued documenting this program’s strategy and initial progress toward accomplishing program objectives. Since 2009, annual reports have been issued to provide updates on the program strategy and the progress made on implementing the strategy. This report provides an update reflecting progress as of January 2014.

  10. Status Report on Efforts to Enhance Instrumentation to Support Advanced Test Reactor Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    J. L. Rempe; D. L. Knudson; J. E. Daw

    2011-03-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR NSUF facilitates basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors was completed. Based on this review, recommendations were made with respect to what instrumentation is needed at the ATR; and a strategy was developed for obtaining these sensors. In 2009, a report was issued documenting this program’s strategy and initial progress toward accomplishing program objectives. In 2009, a report was issued documenting this instrumentation development strategy and initial progress toward accomplishing instrumentation development program objectives. This document reports progress toward implementing this strategy in 2010.

  11. Numerical Investigation of Entropy Generation in Unsteady MHD Generalized Couette Flow with Variable Electrical Conductivity

    Science.gov (United States)

    Chinyoka, T.; Makinde, O. D.

    2013-01-01

    The thermodynamic second law analysis is utilized to investigate the inherent irreversibility in an unsteady hydromagnetic generalized Couette flow with variable electrical conductivity in the presence of induced electric field. Based on some simplified assumption, the model nonlinear governing equations are obtained and solved numerically using semidiscretization finite difference techniques. Effects of various thermophysical parameters on the fluid velocity, temperature, current density, skin friction, the Nusselt number, entropy generation number, and the Bejan number are presented graphically and discussed quantitatively. PMID:23956691

  12. Non-Nuclear Testing of Compact Reactor Technologies at NASA MSFC

    Science.gov (United States)

    Houts, Michael G.; Pearson, J. Boise; Godfroy, Thomas J.

    2011-01-01

    Safe, reliable, compact, autonomous, long-life fission systems have numerous potential applications, both terrestrially and in space. Technologies and facilities developed in support of these systems could be useful to a variety of concepts. At moderate power levels, fission systems can be designed to operate for decades without the need for refueling. In addition, fast neutron damage to cladding and structural materials can be maintained at an acceptable level. Nuclear design codes have advanced to the stage where high confidence in the behavior and performance of a system can be achieved prior to initial testing. To help ensure reactor affordability, an optimal strategy must be devised for development and qualification. That strategy typically involves a combination of non-nuclear and nuclear testing. Non-nuclear testing is particularly useful for concepts in which nuclear operating characteristics are well understood and nuclear effects such as burnup and radiation damage are not likely to be significant. To be mass efficient, a SFPS must operate at higher coolant temperatures and use different types of power conversion than typical terrestrial reactors. The primary reason is the difficulty in rejecting excess heat to space. Although many options exist, NASA s current reference SFPS uses a fast spectrum, pumped-NaK cooled reactor coupled to a Stirling power conversion subsystem. The reference system uses technology with significant terrestrial heritage while still providing excellent performance. In addition, technologies from the SFPS system could be applicable to compact terrestrial systems. Recent non-nuclear testing at NASA s Early Flight Fission Test Facility (EFF-TF) has helped assess the viability of the reference SFPS and evaluate methods for system integration. In July, 2011 an Annular Linear Induction Pump (ALIP) provided by Idaho National Laboratory was tested at the EFF-TF to assess performance and verify suitability for use in a10 kWe technology

  13. A General Method for Module Automatic Testing in Avionics Systems

    Directory of Open Access Journals (Sweden)

    Li Ma

    2013-05-01

    Full Text Available The traditional Automatic Test Equipment (ATE systems are insufficient to cope with the challenges of testing more and more complex avionics systems. In this study, we propose a general method for module automatic testing in the avionics test platform based on PXI bus. We apply virtual instrument technology to realize the automatic testing and the fault reporting of signal performance. Taking the avionics bus ARINC429 as an example, we introduce the architecture of automatic test system as well as the implementation of algorithms in Lab VIEW. The comprehensive experiments show the proposed method can effectively accomplish the automatic testing and fault reporting of signal performance. It greatly improves the generality and reliability of ATE in avionics systems.

  14. Thermionic Reactor Design Studies

    Energy Technology Data Exchange (ETDEWEB)

    Schock, Alfred

    1994-08-01

    Paper presented at the 29th IECEC in Monterey, CA in August 1994. The present paper describes some of the author's conceptual designs and their rationale, and the special analytical techniques developed to analyze their (thermionic reactor) performance. The basic designs, first published in 1963, are based on single-cell converters, either double-ended diodes extending over the full height of the reactor core or single-ended diodes extending over half the core height. In that respect they are similar to the thermionic fuel elements employed in the Topaz-2 reactor subsequently developed in the Soviet Union, copies of which were recently imported by the U.S. As in the Topaz-2 case, electrically heated steady-state performance tests of the converters are possible before fueling.

  15. Performance Test of Korea Heat Load Test Facility (KoHLT-EB) for the Plasma Facing Components of Fusion Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Suk-Kwon; Jin, Hyung Gon; Lee, Eo Hwak; Yoon, Jae-Sung; Lee, Dong Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Cho, Seungyon [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    The main components of the plasma facing components (PFCs) in the tokamak are the blanket first wall and divertor, which include the armour materials, the heat sink with the cooling mechanism, and the diagnostics devices for the temperature measurement. The Korea Heat Load Test facility by using electron beam (KoHLT-EB) has been operating for the plasma facing components to develop fusion engineering. This electron beam facility was constructed using a 300 kW electron gun and a cylindrical vacuum chamber. Performance tests were carried out for the calorimetric calibrations with Cu dummy mockup and for the heat load test of large Cu module. For the simulation of the heat load test of each mockup, the preliminary thermal-hydraulic analyses with ANSYS-CFX were performed. For the development of the plasma facing components in the fusion reactors, test mockups were fabricated and tested in the high heat flux test facility. To perform a beam profile test, an assessment of the possibility of electron beam Gaussian power density profile and the results of the absorbed power for that profile before the test starts are needed. To assess the possibility of a Gaussian profile, for the qualification test of the Gaussian heat load profile, a calorimeter mockup and large Cu module were manufactured to simulate real heat. For this high-heat flux test, the Korean high-heat flux test facility using an electron beam system was constructed. In this facility, a cyclic heat flux test will be performed to measure the surface heat flux, surface temperature profile, and cooling capacity.

  16. Methodology comparison for gamma-heating calculations in material-testing reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lemaire, M.; Vaglio-Gaudard, C.; Lyoussi, A. [CEA, DEN, DER, Cadarache F-13108 Saint Paul les Durance (France); Reynard-Carette, C. [Aix Marseille Universite, CNRS, Universite de Toulon, IM2NP UMR 7334, 13397, Marseille (France)

    2015-07-01

    The Jules Horowitz Reactor (JHR) is a Material-Testing Reactor (MTR) under construction in the south of France at CEA Cadarache (French Alternative Energies and Atomic Energy Commission). It will typically host about 20 simultaneous irradiation experiments in the core and in the beryllium reflector. These experiments will help us better understand the complex phenomena occurring during the accelerated ageing of materials and the irradiation of nuclear fuels. Gamma heating, i.e. photon energy deposition, is mainly responsible for temperature rise in non-fuelled zones of nuclear reactors, including JHR internal structures and irradiation devices. As temperature is a key parameter for physical models describing the behavior of material, accurate control of temperature, and hence gamma heating, is required in irradiation devices and samples in order to perform an advanced suitable analysis of future experimental results. From a broader point of view, JHR global attractiveness as a MTR depends on its ability to monitor experimental parameters with high accuracy, including gamma heating. Strict control of temperature levels is also necessary in terms of safety. As JHR structures are warmed up by gamma heating, they must be appropriately cooled down to prevent creep deformation or melting. Cooling-power sizing is based on calculated levels of gamma heating in the JHR. Due to these safety concerns, accurate calculation of gamma heating with well-controlled bias and associated uncertainty as low as possible is all the more important. There are two main kinds of calculation bias: bias coming from nuclear data on the one hand and bias coming from physical approximations assumed by computer codes and by general calculation route on the other hand. The former must be determined by comparison between calculation and experimental data; the latter by calculation comparisons between codes and between methodologies. In this presentation, we focus on this latter kind of bias. Nuclear

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-01

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

  18. Operating characteristic analysis of a 400 mH class HTS DC reactor in connection with a laboratory scale LCC type HVDC system

    Science.gov (United States)

    Kim, Sung-Kyu; Kim, Kwangmin; Park, Minwon; Yu, In-Keun; Lee, Sangjin

    2015-11-01

    High temperature superconducting (HTS) devices are being developed due to their advantages. Most line commutated converter based high voltage direct current (HVDC) transmission systems for long-distance transmission require large inductance of DC reactor; however, generally, copper-based reactors cause a lot of electrical losses during the system operation. This is driving researchers to develop a new type of DC reactor using HTS wire. The authors have developed a 400 mH class HTS DC reactor and a laboratory scale test-bed for line-commutated converter type HVDC system and applied the HTS DC reactor to the HVDC system to investigate their operating characteristics. The 400 mH class HTS DC reactor is designed using a toroid type magnet. The HVDC system is designed in the form of a mono-pole system with thyristor-based 12-pulse power converters. In this paper, the investigation results of the HTS DC reactor in connection with the HVDC system are described. The operating characteristics of the HTS DC reactor are analyzed under various operating conditions of the system. Through the results, applicability of an HTS DC reactor in an HVDC system is discussed in detail.

  19. Assessment of the need for noise control research on electric power transformers and reactors. Report No. 4289

    Energy Technology Data Exchange (ETDEWEB)

    Keast, D. N.; Gordon, C. G.

    1980-08-01

    This study was conducted to identify and quantify the needs (if any) for noise control research applicable to electric utility transformers and reactors to comply with quantitative state noise regulations. The study was accomplished by analyzing available published data, by studying a sample of utility substation drawings, and by assessing various noise-control design approaches. No experimental work was done. The study was restricted to outdoor substations. A model was prepared to predict noise from existing US substations. A sample of 658 substation designs from five utilities was analyzed to refine the above model and to provide a detailed analysis of the configurations, capacities, and noise-control features of present US substations. A typical substation was defined. Advanced transformer designs (low-loss core, amorphous core, SF/sub 6/-cooled, vapor-cooled, superconducting) for future substations were reviewed to estimate their noise impacts. Noise abatement options were assessed to define where future noise-control research would be appropriate. It was concluded that: at present, about 5% of the electric utility substations in the US, require an average of 14 dBA of noise reduction to comply with existing noise regulations; estimated cost of compliance is about $200 million; and transformer noise is the dominant problem; current technology can provide the necessary noise control, but it is very costly. Additional research and demonstration programs are recommended to reduce the cost of retrofit noise control treatments for existing substations. It is essential that the electric utility industry be involved in guiding this research.

  20. Residual radioactivity guidelines for the heavy water components test reactor at the Savannah River Site

    Energy Technology Data Exchange (ETDEWEB)

    Owen, M.B. Smith, R.; McNeil, J.

    1997-04-01

    Guidelines were developed for acceptable levels of residual radioactivity in the Heavy Water Components Test Reactor (HWCTR) facility at the conclusion of its decommissioning. Using source terms developed from data generated in a detailed characterization study, the RESRAD and RASRAD-BUILD computer codes were used to calculate derived concentration guideline levels (DCGLs) for the radionuclides that will remain in the facility. The calculated DCGLs, when compared to existing concentrations of radionuclides measured during a 1996 characterization program, indicate that no decontamination of concrete surfaces will be necessary. Also, based on the results of the calculations, activated concrete in the reactor biological shield does not have to be removed, and imbedded radioactive piping in the facility can remain in place. Viewed in another way, the results of the calculations showed that the present inventory of residual radioactivity in the facility (not including that associated with the reactor vessel and steam generators) would produce less than one millirem per year above background to a hypothetical individual on the property. The residual radioactivity is estimated to be approximately 0.04 percent of the total inventory in the facility as of March, 1997. According to the results, the only radionuclides that would produce greater than 0.0.1-millirem per year are Am-241 (0.013 mrem/yr at 300 years), C-14 (0.022 mrem/yr at 1000 years) and U-238 (0.034 mrem/yr at 6000 years). Human exposure would occur only through the groundwater pathways, that is, from water drawn from, a well on the property. The maximum exposure would be approximately one percent of the 4 millirem per year ground water exposure limit established by the U.S. Environmental Protection Agency. 11 refs., 13 figs., 15 tabs.

  1. On Washback of Testing to General English Education

    Institute of Scientific and Technical Information of China (English)

    TangShengping; XuChongning

    2004-01-01

    The educational society identifies that important testing has on teaching and learning great effects which are referred to as washback, Although there is controversy about its benefi cialor harmful effect on teaching and learning, it is generally accepted that testing generates both negative and positive washback, By making a comparison between NMET (the

  2. Testing general relativity: from local to cosmological scales.

    Science.gov (United States)

    Uzan, Jean-Philippe

    2011-12-28

    I summarize various tests of general relativity on astrophysical scales, based on the large-scale structure of the universe but also on other systems, in particular the constants of physics. I emphasize the importance of hypotheses on the geometric structures of our universe while performing such tests and discuss their complementarity as well as their possible extensions.

  3. Quality of routine spirometry tests in Dutch general practices.

    NARCIS (Netherlands)

    Schermer, T.R.J.; Crockett, A.J.; Poels, P.J.P.; Dijke, J.J. van; Akkermans, R.P.; Vlek, H.F.; Pieters, W.R.

    2009-01-01

    BACKGROUND: Spirometry is an indispensable tool for diagnosis and monitoring of chronic airways disease in primary care. AIM: To establish the quality of routine spirometry tests in general practice, and explore associations between test quality and patient characteristics. DESIGN OF STUDY: Analysis

  4. 49 CFR 219.11 - General conditions for chemical tests.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 4 2010-10-01 2010-10-01 false General conditions for chemical tests. 219.11 Section 219.11 Transportation Other Regulations Relating to Transportation (Continued) FEDERAL RAILROAD... for chemical tests. (a) Any employee who performs covered service for a railroad is deemed to...

  5. 40 CFR 59.650 - General testing provisions.

    Science.gov (United States)

    2010-07-01

    ... needed to measure emissions. (c) The specification for gasoline to be used for testing is given in 40 CFR 1065.710. Use the grade of gasoline specified for general testing. Blend this grade of gasoline with reagent grade ethanol in a volumetric ratio of 90.0 percent gasoline to 10.0 percent ethanol. You may...

  6. PAT portable appliance testing : in-service inspection and testing of electrical equipment

    CERN Document Server

    Scaddan, Brian

    2015-01-01

    The Electricity at Work Regulations 1989 requires any electrical system to be constructed, maintained and used in such a manner as to prevent danger. This means that inspection and testing of systems, including portable appliances, is needed in order to determine if maintenance is required. This book explains in clear language what needs to be done and includes expert advice on legislation as well as actual testing. The book contains an appendix providing the electrical fundamentals needed by non-specialists and also has sample questions (with answers) for the C&G 2377 exam. It is also an ideal revision guide for the non-specialist, such as maintenance staff and caretakers who carry out these tasks part-time, alongside their many other duties.

  7. Classical tests of general relativity in brane world models

    Energy Technology Data Exchange (ETDEWEB)

    Boehmer, Christian G [Department of Mathematics and Institute of Origins, University College London, Gower Street, London WC1E 6BT (United Kingdom); De Risi, Giuseppe [Dipartimento di Fisica, Universita degli studi di Bari and Istituto Nazionale di Fisica Nucleare, sez. di Bari, Via G. Amendola 173, 70126 Bari (Italy); Harko, Tiberiu [Department of Physics and Center for Theoretical and Computational Physics, University of Hong Kong, Pok Fu Lam Road (Hong Kong); Lobo, Francisco S N, E-mail: c.boehmer@ucl.ac.u, E-mail: giuseppe.derisi@ba.infn.i, E-mail: harko@hkucc.hku.h, E-mail: flobo@cii.fc.ul.p [Centro de Astronomia e Astrofisica da Universidade de Lisboa, Campo Grande, Ed. C8 1749-016 Lisboa (Portugal)

    2010-09-21

    The classical tests of general relativity (perihelion precession, deflection of light and the radar echo delay) are considered for several spherically symmetric static vacuum solutions in brane world models. Generally, the spherically symmetric vacuum solutions of the brane gravitational field equations have properties quite distinct as compared to the standard black hole solutions of general relativity. As a first step a general formalism that facilitates the analysis of general relativistic Solar System tests for any given spherically symmetric metric is developed. It is shown that the existing observational Solar System data on the perihelion shift of Mercury, on the light bending around the Sun (obtained using long-baseline radio interferometry), and ranging to Mars using the Viking lander constrain the numerical values of the parameters of the specific models.

  8. Simulation and tests to individual and coupled models of the reactor vessel simulator and the recirculation system for the SUN-RAH; Simulacion y pruebas a modelos individuales y acoplados del simulador de la vasija del reactor y el sistema de recirculacion para el SUN-RAH

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez S, R.A. [UNAM, Laboratorio de Analisis de Ingenieria de Reactores Nucleares, DEPFI, Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)]. e-mail: rsanchez_15@yahoo.com.mx

    2004-07-01

    The present project, is continuation of the project presented in the congress SNM-2003. In this new phase of the project, they were carried out adaptive changes to the modeling and implementation of the module of the full superior of the core of the reactor, they were carried out those modeling of the generation of heat as well as of the energy transfer in the one fuel. These models present the main characteristics of the vessel of the one reactor and of the recirculation system, defined by the main phenomena that they intervene in the physical processes, in the previous version the simulation in real time it required of an extremely quick computer and without executing collateral processes. The tests are presented carried out to the different models belonging to the Simulator of the Reactor Vessel and the Recirculation system for the SUN-RAH (University Simulator of Nucleo electric with Boiling Water Reactor), as well as the results hurtled by this tests. In each section the executions of the tests and the corresponding analyses of results are shown for each pattern. Besides the above mentioned, the advantages presented by the Simulator of the reactor vessel and the recirculation system are pointed. (Author)

  9. A high resolution pneumatic stepping actuator for harsh reactor environments

    Science.gov (United States)

    Tippetts, Thomas B.; Evans, Paul S.; Riffle, George K.

    1993-01-01

    A reactivity control actuator for a high-power density nuclear propulsion reactor must be installed in close proximity to the reactor core. The energy input from radiation to the actuator structure could exceed hundreds of W/cc unless low-cross section, low-absorptivity materials are chosen. Also, for post-test handling and subsequent storage, materials should not be used that are activated into long half-life isotopes. Pneumatic actuators can be constructed from various reactor-compatible materials, but conventional pneumatic piston actuators generally lack the stiffness required for high resolution reactivity control unless electrical position sensors and compensated electronic control systems are used. To overcome these limitations, a pneumatic actuator is under development that positions an output shaft in response to a series of pneumatic pulses, comprising a pneumatic analog of an electrical stepping motor. The pneumatic pulses are generated remotely, beyond the strong radiation environment, and transmitted to the actuator through tubing. The mechanically simple actuator uses a nutating gear harmonic drive to convert motion of small pistons directly to high-resolution angular motion of the output shaft. The digital nature of this actuator is suitable for various reactor control algorithms but is especially compatible with the three bean salad algorithm discussed by Ball et al. (1991).

  10. Neural network setpoint control of an advanced test reactor experiment loop simulation

    Energy Technology Data Exchange (ETDEWEB)

    Cordes, G.A.; Bryan, S.R.; Powell, R.H.; Chick, D.R.

    1990-09-01

    This report describes the design, implementation, and application of artificial neural networks to achieve temperature and flow rate control for a simulation of a typical experiment loop in the Advanced Test Reactor (ATR) located at the Idaho National Engineering Laboratory (INEL). The goal of the project was to research multivariate, nonlinear control using neural networks. A loop simulation code was adapted for the project and used to create a training set and test the neural network controller for comparison with the existing loop controllers. The results for three neural network designs are documented and compared with existing loop controller action. The neural network was shown to be as accurate at loop control as the classical controllers in the operating region represented by the training set. 9 refs., 28 figs., 2 tabs.

  11. Improving the AGR Fuel Testing Power Density Profile Versus Irradiation-Time in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gray S. Chang; David A. Petti; John T. Maki; Misti A. Lillo

    2009-05-01

    The Very High Temperature gas-cooled Reactor (VHTR), which is currently being developed, achieves simplification of safety through reliance on ceramic-coated fuel particles. Each TRISO-coated fuel particle has its own containment which serves as the principal barrier against radionuclide release under normal operating and accident conditions. These fuel particles, in the form of graphite fuel compacts, are currently undergoing a series of irradiation tests in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) to support the Advanced Gas-Cooled Reactor (AGR) fuel qualification program. A representive coated fuel particle with an 235U enrichment of 19.8 wt% was used in this analysis. The fuel burnup analysis tool used to perform the neutronics study reported herein, couples the Monte Carlo transport code MCNP, with the radioactive decay and burnup code ORIGEN2. The fuel burnup methodology known as Monte-Carlo with ORIGEN2 (MCWO) was used to evaluate the AGR experiment assembly and demonstrate compliance with ATR safety requirements. For the AGR graphite fuel compacts, the MCWO-calculated fission power density (FPD) due to neutron fission in 235U is an important design parameter. One of the more important AGR fuel testing requirements is to maintain the peak fuel compact temperature close to 1250°C throughout the proposed irradiation campaign of 550 effective full power days (EFPDs). Based on the MCWO-calculated FPD, a fixed gas gap size was designed to allow regulation of the fuel compact temperatures throughout the entire fuel irradiation campaign by filling the gap with a mixture of helium and neon gases. The chosen fixed gas gap can only regulate the peak fuel compact temperature in the desired range during the irradiation test if the ratio of the peak power density to the time-dependent low power density (P/T) at 550 EFPDs is less than 2.5. However, given the near constant neutron flux within the ATR driver core and the depletion of 235U in

  12. 78 FR 75961 - Energy Conservation Program: Test Procedures for Electric Motors

    Science.gov (United States)

    2013-12-13

    ... for Electric Motors; Final Rule #0;#0;Federal Register / Vol. 78 , No. 240 / Friday, December 13, 2013... Program: Test Procedures for Electric Motors AGENCY: Office of Energy Efficiency and Renewable Energy... efficiency test procedures for electric motors to allow currently unregulated motors to be tested...

  13. 76 FR 9374 - Proposed Extension of Existing Information Collection; Examinations and Testing of Electrical...

    Science.gov (United States)

    2011-02-17

    ....820 Electrical work; troubleshooting and testing; 75.821 Testing, examination and maintenance; 75.900... effectiveness of their electrical maintenance programs, to determine that the required tests and examinations... effectiveness of the coal mine operator's electrical maintenance programs. By comparing the records with...

  14. Flow tests of a single fuel element coolant channel for a compact fast reactor for space power

    Science.gov (United States)

    Springborn, R. H.

    1971-01-01

    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.

  15. Verifying seismic design of nuclear reactors by testing. Volume 1: test plan

    Energy Technology Data Exchange (ETDEWEB)

    1979-07-20

    This document sets forth recommendations for a verification program to test the ability of operational nuclear power plants to achieve safe shutdown immediately following a safe-shutdown earthquake. The purpose of the study is to develop a program plan to provide assurance by physical demonstration that nuclear power plants are earthquake resistant and to allow nuclear power plant operators to (1) decide whether tests should be conducted on their facilities, (2) specify the tests that should be performed, and (3) estimate the cost of the effort to complete the recommended test program.

  16. General Electric Company proposed management plan, commercial buildings, National Solar Demonstration Program

    Energy Technology Data Exchange (ETDEWEB)

    None

    1976-04-01

    The National Solar Demonstration Plan (NSDP) for the solar heating and cooling of non-residential buildings is designed to meet the goals and objectives of ERDA 23A. The plan has been formulated for a matrix of two hundred (200) commercial demonstration projects. This report contains the General Electric recommendations for the NSDP pertaining to the management structure, demonstration implementation, schedules, data dissemination, and resource requirements. Closely associated with these major planning elements are the related elements in the other recent General Electric reports submitted under the SHACOB Phase I contract. (WDM)

  17. Generalized Mei Conserved Quantity of Mei Symmetry for Mechanico-electrical Systems with Nonholonomic Controllable Constraints

    Institute of Scientific and Technical Information of China (English)

    XIA Li-Li; ZHAO Xian-Lin

    2009-01-01

    On the basis of the total time derivative along the trajectory, we study the generalized Mei conserved quantity of Mei symmetry for mechanico-electrical systems with nonholonomie controllable constraints. Firstly, the definition and criterion of Mei symmetry for mechanico-electrical systems with nonholonomic controllable constraints are presented. Secondly, a coordination function is introduced, and the conditions of existence of generalized Mei conserved quantity as well as the forms are proposed. Lastly, an example is given to illustrate the application of the results.

  18. 40 CFR 80.59 - General test fleet requirements for vehicle testing.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 16 2010-07-01 2010-07-01 false General test fleet requirements for... test fleet requirements for vehicle testing. (a) The test fleet must consist of only 1989-91 MY... test fleet shall have no fewer than 4,000 miles of accumulated mileage prior to being included in...

  19. Monte Carlo Simulation Study of a Differential Calorimeter Measuring the Nuclear Heating in Material Testing Reactors

    Science.gov (United States)

    Amharrak, H.; Reynard-Carette, C.; Lyoussi, A.; Carette, M.; Brun, J.; De Vita, C.; Fourmentel, D.; Villard, J.-F.; Guimbal, P.

    2016-02-01

    The nuclear heating measurements in Material Testing Reactors (MTRs) are crucial for the study of nuclear materials and fuels under irradiation. The reference measurements of this nuclear heating are especially performed by a differential calorimeter including a graphite sample material. Then these measurements are used for other materials, other geometries, or other experimental conditions in order to predict the nuclear heating and thermal conditions induced in the irradiation devices. This paper will present new simulations with MCNP Monte-Carlo transport code to determine the gamma heating profile inside the calorimeter. The whole complex geometry of the sensor has been considered. We use as an input source in the model, the photon spectra calculated in various positions of CARMEN-1 irradiation program in OSIRIS reactor. After a description of the differential calorimeter device, the MCNP modeling used for the calculations of radial profile of nuclear heating inside the calorimeter elements will be introduced. The obtained results of different simulations will be detailed and discussed in this paper. The charged particle equilibrium inside the calorimeter elements will be studied. Then we will focus on parametric studies of the various components of the calorimeter. The influence of source type will be also took into account. Moreover the influence of the material used for the sample will be described.

  20. Monte Carlo Simulation Study of a Differential Calorimeter Measuring the Nuclear Heating in Material Testing Reactors

    Directory of Open Access Journals (Sweden)

    Amharrak H.

    2016-01-01

    Full Text Available The nuclear heating measurements in Material Testing Reactors (MTRs are crucial for the study of nuclear materials and fuels under irradiation. The reference measurements of this nuclear heating are especially performed by a differential calorimeter including a graphite sample material. Then these measurements are used for other materials, other geometries, or other experimental conditions in order to predict the nuclear heating and thermal conditions induced in the irradiation devices. This paper will present new simulations with MCNP Monte-Carlo transport code to determine the gamma heating profile inside the calorimeter. The whole complex geometry of the sensor has been considered. We use as an input source in the model, the photon spectra calculated in various positions of CARMEN-1 irradiation program in OSIRIS reactor. After a description of the differential calorimeter device, the MCNP modeling used for the calculations of radial profile of nuclear heating inside the calorimeter elements will be introduced. The obtained results of different simulations will be detailed and discussed in this paper. The charged particle equilibrium inside the calorimeter elements will be studied. Then we will focus on parametric studies of the various components of the calorimeter. The influence of source type will be also took into account. Moreover the influence of the material used for the sample will be described.