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. Safety Evaluation Report related to the renewal of the operating license for the General Electric-Nuclear Test Reactor (GE-NTR) (Docket No. 50-73)

    International Nuclear Information System (INIS)

    1984-09-01

    This Safety Evaluation Report for the application filed by the General Electric Company (GE) for a renewal license number R-33 to continue to operate its research reactor has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is owned and operated by GE and is located in Pleasanton, California. The staff concludes that the reactor can continue to be operated by GE without endangering the health and safety of the public

  3. Design and performance of General Electric boiling water reactor main steam line isolation valves

    International Nuclear Information System (INIS)

    Rockwell, D.A.; van Zylstra, E.H.

    1976-08-01

    An extensive test program has been completed by the General Electric Company in cooperation with the Commonwealth Edison Company on the basic design type of large main steam line isolation valves used on General Electric Boiling Water Reactors. Based on a total of 40 tests under simulated accident conditions covering a wide range of mass flows, mixture qualities, and closing times, it was concluded that the commercially available valves of this basic type will close completely and reliably as required. Analytical methods to predict transient effects in the steam line and valve after postulated breaks were refined and confirmed by the test program

  4. Standard technical specifications for General Electric boiling water reactors

    International Nuclear Information System (INIS)

    1979-08-01

    This Standard Technical Specification (STS) has been structured for the broadest possible use on General Electric plants currently being reviewed for an Operating License. Optional specifications are provided for those features and systems which may be included in individual plant designs but are not generic in their scope of application. This revision of the GE-STS does not typically include requirements which may be added or revised as a result of the NRC staff's further review of the Three Mile Island incident

  5. Standard Technical Specifications for General Electric Boiling Water Reactors (BWR/5)

    International Nuclear Information System (INIS)

    Bottimore, R.R.

    1980-12-01

    The Standard Technical Specifications for General Electric Boiling Water Reactors (GE-STS) is a generic document prepared by the US NRC for use in the licensing process of current General Electric Boiling Water Reactors. The GE-STS sets forth the limits, operating conditions, and other requirements applicable to nuclear reactor facility operation as set forth by Section 50.36 of 10 CFR Part 50 for the protection of the health and safety of the public. The document is revised periodically to reflect current licensing requirements

  6. Generic risk insights for General Electric boiling water reactors

    International Nuclear Information System (INIS)

    Travis, R.; Taylor, J.; Chung, J.

    1991-05-01

    A methodology has been developed to extract generic risk-based information from probabilistic risk assessments (PRAs) of General Electric boiling water rectors and applying the insights gained to plants that have not been subjected to a PRA. The available risk assessments (six plants) were examined to identify the most probable, i.e., dominant accident sequences at each plants. The goal was to include all sequences which represented at least 80% of core damage frequency. If the same plant specific dominant accident sequence appeared within this boundary in at least two plant PRAs, the sequence was considered to be a representative sequence. Eight sequences met this definition. From these sequences, the most important component failures and human error that contributed to each sequence have been prioritized. Guidance is provided to prioritize the representative sequences and modify selected basic events that have been shown to be sensitive to the plant specific design or operating variations of the contributing PRAs. This risk-based guidance can be used for utility and NRC activities including operator training, maintenance, design review, and inspections. 13 refs., 6 tabs

  7. Physical characteristics of GE [General Electric] BWR [boiling-water reactor] fuel assemblies

    International Nuclear Information System (INIS)

    Moore, R.S.; Notz, K.J.

    1989-06-01

    The physical characteristics of fuel assemblies manufactured by the General Electric Company for boiling-water reactors are classified and described. The classification into assembly types is based on the GE reactor product line, the Characteristics Data Base (CDB) assembly class, and the GE fuel design. Thirty production assembly types are identified. Detailed physical data are presented for each assembly type in an appendix. Descriptions of special (nonstandard) fuels are also reported. 52 refs., 1 fig., 6 tabs

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

    International Nuclear Information System (INIS)

    Fletcher, C.D.

    1986-01-01

    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 refs., 17 figs., 1 tab

  9. Preliminary irradiation test results from the Yankee Atomic Electric Company reactor vessel test irradiation program

    International Nuclear Information System (INIS)

    Biemiller, E.C.; Fyfitch, S.; Campbell, C.A.

    1993-01-01

    The Yankee Atomic Electric Company test irradiation program was implemented to characterize the irradiation response of representative Yankee Rowe reactor vessel beltline plate materials and to remove uncertainties in the analysis of existing irradiation data on the Yankee Rowe reactor vessel steel. Plate materials each containing 0.24 w/o copper, but different nickel contents at 0.63 w/o and 0.19 w/o, were heat treated to simulate the Yankee vessel heat treatment (austenitized at 1800 deg F) and to simulate Regulatory Guide 1.99 database materials (austenitized at 1600 deg. F). These heat treatments produced different microstructures so the effect of microstructure on irradiation damage sensitivity could be tested. Because the nickel content of the test plates varied and the copper level was constant, the effect of nickel on irradiation embrittlement was also tested. Correlation monitor material, HSST-02, was included in the program to benchmark the Ford Nuclear Reactor (U. of Michigan Test Reactor) which had never been used for this type of irradiation program. Materials taken from plate surface locations (vs. 1/4T) were included to test whether or not the improved toughness properties of the plate surface layer, resulting from the rapid quench, is maintained after irradiation. If the improved properties are maintained, pressurized thermal shock calculations could utilize this margin. Finally, for one experiment, irradiations were conducted at two irradiation temperatures (500 deg. F and 550 deg. F) to determine the effect of irradiation temperature on embrittlement. The preliminary results of the irradiation program show an increase in T 30 shift of 69 deg. F for a decrease in irradiation temperature of 50 deg. F. The results suggest that for nickel bearing steels, the superior toughness of plate surface material is maintained after irradiation and for the copper content tested, nickel had no apparent effect on irradiation response. No apparent microstructure

  10. Preliminary irradiation test results from the Yankee Atomic Electric Company reactor vessel test irradiation program

    International Nuclear Information System (INIS)

    Biemiller, E.C.; Fyfitch, Stephen; Campbell, C.A.

    1994-01-01

    The Yankee Atomic Electric Company test irradiation program was implemented to characterize the irradiation response of representative Yankee Rowe reactor vessel beltline plate materials and to remove uncertainties in the analysis of existing irradiation data on the Yankee Rowe reactor vessel steel. Plate materials each containing 0.24 w/o copper, but different nickel contents at 0.63 w/o and 0.19 w/o, were heat treated to simulate the Yankee vessel heat treatment (austenitized at 982 o C (1800 o F)) and to simulate Regulatory Guide 1.99 database materials (austenitized at 871 o C (1600 o F)). These heat treatments produced different microstructures so the effect of microstructure on irradiation damage sensitivity could be tested. Because the nickel content of the test plates varied and the copper level was constant, the effect of nickel on irradiation embrittlement was also tested. Correlation monitor material, HSST-02, was included in the program to benchmark the Ford Nuclear Reactor (University of Michigan Test Reactor) which had never been used before for this type of irradiation program. Materials taken from plate surface locations (versus 1/4 T) were included to test whether or not the improved toughness properties of the plate surface layer, resulting from the rapid quench, are maintained after irradiation. If the improved properties are maintained, pressurized thermal shock calculations could utilize this margin. Finally, for one experiment, irradiations were conducted at two irradiation temperatures (260 o C and 288 o C) to determine the effect of irradiation temperature on embrittlement. (Author)

  11. NRC [Nuclear Regulatory Commission] staff evaluation of the General Electric Company Nuclear Reactor Study (''Reed Report'')

    International Nuclear Information System (INIS)

    1987-07-01

    In 1975, the General Electric Company (GE) published a Nuclear Reactor Study, also referred to as ''the Reed Report,'' an internal product-improvement study. GE considered the document ''proprietary'' and thus, under the regulations of the Nuclear Regulatory Commission (NRC), exempt from mandatory public disclosure. Nonetheless, members of the NRC staff reviewed the document in 1976 and determined that it did not raise any significant new safety issues. The staff also reached the same conclusion in subsequent reviews. However, in response to recent inquiries about the report, the staff reevaluated the Reed Report from a 1987 perspective. This re-evaluation, documented in this staff report, concluded that: (1) there are no issues raised in the Reed Report that support a need to curtail the operation of any GE boiling water reactor (BWR); (2) there are no new safety issues raised in the Reed Report of which the staff was unaware; and (3) although certain issues addressed by the Reed Report are still being studied by the NRC and the industry, there is no basis for suspending licensing and operation of GE BWR plants while these issues are being resolved

  12. Data on loss of off-site electric power simulation tests of the high temperature engineering test reactor

    International Nuclear Information System (INIS)

    Takeda, Takeshi; Nakagawa, Shigeaki; Fujimoto, Nozomu; Tachibana, Yukio; Iyoku, Tatsuo

    2002-07-01

    The high temperature engineering test reactor (HTTR), the first high temperature gas-cooled reactor (HTGR) in Japan, achieved the first full power of 30 MW on December 7 in 2001. In the rise-to-power test of the HTTR, simulation tests on loss of off-site electric power from 15 and 30 MW operations were carried out by manual shutdown of off-site electric power. Because helium circulators and water pumps coasted down immediately after the loss of off-site electric power, flow rates of helium and water decreased to the scram points. To shut down the reactor safely, the subcriticality should be kept by the insertion of control rods and the auxiliary cooling system should cool the core continuously avoiding excessive cold shock to core graphite components. About 50 s later from the loss of off-site electric power, the auxiliary cooling system started up by supplying electricity from emergency power feeders. Temperature of hot plenum block among core graphite structures decreased continuously after the startup of the auxiliary cooling system. This report describes sequences of dynamic components and transient behaviors of the reactor and its cooling system during the simulation tests from 15 and 30 MW operations. (author)

  13. Analysis of the General Electric Company swell tests with RELAP4/MOD7

    International Nuclear Information System (INIS)

    Fischer, S.R.; Hendrix, C.E.

    1979-01-01

    The RELAP4/MOD7 nuclear reactor transient analysis code, presently being developed by EG and G Idaho, Inc., will incorporate several significant improvements over earlier versions of RELAP4. As part of the development of RELAP4/MOD7, a thorough assessment of the capability of the code to simulate water reactor LOCA phenomena is being made. This assessment is accomplished in part by comparing results from code calculations with test data from experimental facilities. Simulations of the General Electric Company (GE) level swell tests were performed as part of the code checkout. In these tests, a pressurized vessel partially filled with nearly saturated water was blown down through a simulated break located near the top of the vessel. Comparison of RELAP4 calculations with data from these experiments indicates that the code has the capability to model the unequal phase velocity flow and resulting density gradients that might occur in a BWR steam line break transient. Comparisons of RELAP4 calculations with data from two level swell experiments are presented

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

    Energy Technology Data Exchange (ETDEWEB)

    Carlson, Richard Barney [Idaho National Lab. (INL), Idaho Falls, ID (United States); Scoffield, Don [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bennett, Brion [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    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.

  15. 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.

  16. TU electric reactor model verification

    International Nuclear Information System (INIS)

    Willingham, C.E.; Killgore, M.R.

    1989-01-01

    Power reactor benchmark calculations using the code package CASMO-3/SIMULATE-3 have been performed for six cycles of Prairie Island Unit 1. The reload fuel designs for the selected cycles include gadolinia as a burnable absorber, natural uranium axial blankets, and increased water-to-fuel ratio. The calculated results for both low-power physics tests (boron end points, control rod worths, and isothermal temperature coefficients) and full-power operation (power distributions and boron letdown) are compared to measured plant data. These comparisons show that the TU Electric reactor physics models accurately predict important physics parameters for power reactors

  17. Piping benchmark problems for the General Electric Advanced Boiling Water Reactor

    International Nuclear Information System (INIS)

    Bezler, P.; DeGrassi, G.; Braverman, J.; Wang, Y.K.

    1993-08-01

    To satisfy the need for verification of the computer programs and modeling techniques that will be used to perform the final piping analyses for an advanced boiling water reactor standard design, three benchmark problems were developed. The problems are representative piping systems subjected to representative dynamic loads with solutions developed using the methods being proposed for analysis for the advanced reactor standard design. It will be required that the combined license holders demonstrate that their solutions to these problems are in agreement with the benchmark problem set

  18. Electric-stepping-motor tests for a control-drum actuator of a nuclear reactor

    Science.gov (United States)

    Kieffer, A. W.

    1972-01-01

    Experimental tests were conducted on two stepping motors for application as reactor control-drum actuators. Various control-drum loads with frictional resistances ranging from approximately zero to 40 N-m and inertias ranging from zero to 0.424 kg-sq m were tested.

  19. TRAC development at General Electric

    International Nuclear Information System (INIS)

    Andersen, J.G.M.; Shaug, J.C.; Shiralkar, B.S.

    1987-01-01

    TRAC is a computer code for transient analysis of light water reactors. The BWR version of TRAC has been developed as a result of a close cooperation between General Electric Company and Idaho National Engineering Laboratory. Up through 1985 the development work at General Electric was jointly funded by General Electric, the Nuclear Regulatory Commission and Electric Power Research Institute under the Refill-Reflood and FIST programs. At INEL (which has the main responsibility for the NRC version of TRAC-BWR) this work has led to the development of TRACBD1 and TRACBF1, while at GE, TRACB04 was the final product of the Refill-Reflood and FIST programs. TRAC development has continued at General Electric after the completion of these programs with the evolution of the TRACG code. The purpose of the paper is to describe this work. The TRAC development at General Electric can be divided into two main categories: extended benchmark capability and improved user convenience

  20. Analytical evaluation on loss of off-side electric power simulation of the High Temperature Engineering Test Reactor

    International Nuclear Information System (INIS)

    Takeda, Takeshi; Nakagawa, Shigeaki; Tachibana, Yukio; Takada, Eiji; Kunitomi, Kazuhiko

    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)

  1. Generalities about nuclear reactors

    International Nuclear Information System (INIS)

    Jaouen, C.; Beroux, P.

    2012-01-01

    From Zoe, the first nuclear reactor, till the current EPR, the French nuclear industry has always advanced by profiting from the feedback from dozens of years of experience and operations, in particular by drawing lessons from the most significant events in its history, such as the Fukushima accident. The new generations of reactors must improve safety and economic performance so that the industry maintain its legitimacy and its share in the production of electricity. This article draws the history of nuclear power in France, gives a brief description of the pressurized water reactor design, lists the technical features of the different versions of PWR that operate in France and compares them with other types of reactors. The feedback experience concerning safety, learnt from the major nuclear accidents Three Miles Island (1979), Chernobyl (1986) and Fukushima (2011) is also detailed. Today there are 26 third generation reactors being built in the world: 4 EPR (1 in Finland, 1 in France and 2 in China); 2 VVER-1200 in Russia, 8 AP-1000 (4 in China and 4 in the Usa), 8 APR-1400 (4 in Korea and 4 in UAE), and 4 ABWR (2 in Japan and 2 in Taiwan)

  2. The HTR-10 test reactor project and potential use of HTGR for non-electric application in China

    International Nuclear Information System (INIS)

    Sun Yuliang; Zhong Daxin; Xu Yuanhui; Wu Zhongxin

    1997-01-01

    Coal is the dominant source of energy in China. This use of coal results in two significant problems for China; it is a major burden on the train, road and waterway transportation infrastructures and it is a significant source of environmental pollution. In order to ease the problems caused by the burning of coal and to help reduce the energy supply shortage in China, national policy has directed the development of nuclear power. This includes the erection of nuclear power plants with water cooled reactors and the development of advanced nuclear reactor types, specifically, the high temperature gas cooled reactor (HTGR). The HTGR was chosen for its favorable safety features and its ability to provide high reactor outlet coolant temperatures for efficient power generation and high quality process heat for industrial applications. As the initial modular HTGR development activity within the Chinese High Technology Programme, a 10MW helium cooled test reactor is currently under construction on the site of the Institute of Nuclear Energy Technology northwest of Beijing. This plant features a pebble-bed helium cooled reactor with initial criticality anticipated in 1999. There will be two phases of high temperature heat utilization from the HTR-10. The first phase will utilize a reactor outlet temperature of 700 deg. C with a steam generator providing steam for a steam turbine cycle which works on an electrical/heat co-generation basis. The second phase is planned for a core outlet temperature of 900 deg. C to investigate a steam cycle/gas turbine combined cycle system with the gas turbine and the steam cycle being independently parallel in the secondary side of the plant. This paper provides a review of the technical design, licensing, safety and construction schedule for the HTR-10. It also addresses the potential uses of the HTGR for non-electric applications in China including process steam for the petrochemical industry, heavy oil recovery, coal conversion and

  3. 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

  4. Examination on the testing method for evaluating life of electric wires and cables for nuclear reactors

    International Nuclear Information System (INIS)

    Seguchi, Tadao; Morita, Yosuke; Yoshida, Kenzo

    1984-01-01

    Regarding the method of environmental test on the electric wires and cables used for the safety system in the containment vessels of nuclear power plants, the draft recommendation was issued by the Institute of Electrical Engineers of Japan in 1982. Its contents follow the IEEE Standard of USA, and are composed of the tests on the deterioration in normal operation, the case of LOCA and the prevention of spread of fire. In this report, as to the testing method regarding normal operation, the appropriate method and its basis are described in view of the recent experimental data. In the draft recommendation, the successive method carrying out irradiation after thermal deterioration is adopted, and both testing conditions are given. However, the fundamental problems remain in the propriety of the acceleration of deterioration and the multiplied effect of heat and radiation. The qualitative and quantitative data on these problems have been accumulated in various countries, therefore, the examination of the testing method was carried out based on these data. The dose rate dependence of radiation deterioration, the multiplied effect of radiation and heat, and the correlation of thermal deterioration rate with temperature are discussed. The appropriate method is proposed. (Kako, I.)

  5. 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.

  6. TU Electric reactor physics model verification: Power reactor benchmark

    International Nuclear Information System (INIS)

    Willingham, C.E.; Killgore, M.R.

    1988-01-01

    Power reactor benchmark calculations using the advanced code package CASMO-3/SIMULATE-3 have been performed for six cycles of Prairie Island Unit 1. The reload fuel designs for the selected cycles included gadolinia as a burnable absorber, natural uranium axial blankets and increased water-to-fuel ratio. The calculated results for both startup reactor physics tests (boron endpoints, control rod worths, and isothermal temperature coefficients) and full power depletion results were compared to measured plant data. These comparisons show that the TU Electric reactor physics models accurately predict important measured parameters for power reactors

  7. Conceptual design of an electrical power module for the tokamak fusion test reactor

    International Nuclear Information System (INIS)

    Jassby, D.L.; Bullis, R.; Sedgeley, D.; Caldwell, C.S.; Pettus, W.G.; Schluderberg, D.C.

    1979-01-01

    The TFTR Engineering Test Station (ETS) can support blanket modules with a fusion-neutron view area of 0.5 m/sup 2/. If the TFTR magnetic systems and beam injectors can operate with pulse lengths of 5 s, once every 300 s, the time-averaged neutron power incident on a module will be 1.5 kW, which can be enhanced by a suitable blanket energy multiplier. A preliminary conceptual design of a dual-loop steam-generating power system that can be housed in the ETS has been carried out. The optimal heat transfer fluid in the primary loop is an organic liquid, which allows an operating temperature of 700/degree/F at low pressure. The primary coolant must be preheated electrically to operating temperature. A ballast tank levels the temperature at the steam generator, so that the secondary loop is in steady-state operation. With a natural-uranium blanket multiplier, the time-averaged net electrical power is 1.2 kW(e). 8 refs

  8. 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.

  9. 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.

  10. Nozzle for electric dispersion reactor

    Science.gov (United States)

    Sisson, W.G.; Basaran, O.A.; Harris, M.T.

    1995-11-07

    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. 4 figs.

  11. Nuclear reactor development in China for non-electrical applications

    International Nuclear Information System (INIS)

    Sun Yuliang; Zhong Daxin; Dong Duo; Xu Yuanhui

    1998-01-01

    In parallel to its vigorous program of nuclear power generation, China has attached great importance to the development of nuclear reactors for non-electrical applications. The Institute of Nuclear Energy Technology (INET) in Beijing has been developing technologies of the water-cooled heating reactor and the modular high temperature gas-cooled reactor. In 1989, a 5 MW water cooled test reactor was erected. Currently, an industrial demonstration nuclear heating plant is being projected. Feasibility studies are being made of sea-water desalination using the INET developed nuclear heating reactor as heat source. Also, a 10 MW high temperature gas-cooled test reactor is being constructed at INET in the framework of China's national high-tech program. The paper gives an overview of China's energy market situation. With respect to China's technology development of high temperature gas-cooled reactors and water cooled heating reactors, the paper describes some general requirements on the technical development, reviews the national programs and activities, describes briefly the design and safety features of the reactor concepts, discusses aspects of application potentials. (author)

  12. Real time simulator for material testing reactor

    Energy Technology Data Exchange (ETDEWEB)

    Takemoto, Noriyuki; Imaizumi, Tomomi; Izumo, Hironobu; Hori, Naohiko; Suzuki, Masahide [Japan Atomic Energy Agency, Oarai Research and Development Center, Oarai, Ibaraki (Japan); Ishitsuka, Tatsuo; Tamura, Kazuo [ITOCHU Techno-Solutions Corp., Tokyo (Japan)

    2012-03-15

    Japan Atomic Energy Agency (JAEA) is now developing a real time simulator for a material testing reactor based on Japan Materials Testing Reactor (JMTR). The simulator treats reactor core system, primary and secondary cooling system, electricity system and irradiation facility systems. Possible simulations are normal reactor operation, unusual transient operation and accidental operation. The developed simulator also contains tool to revise/add facility in it for the future development. (author)

  13. Real time simulator for material testing reactor

    International Nuclear Information System (INIS)

    Takemoto, Noriyuki; Imaizumi, Tomomi; Izumo, Hironobu; Hori, Naohiko; Suzuki, Masahide; Ishitsuka, Tatsuo; Tamura, Kazuo

    2012-01-01

    Japan Atomic Energy Agency (JAEA) is now developing a real time simulator for a material testing reactor based on Japan Materials Testing Reactor (JMTR). The simulator treats reactor core system, primary and secondary cooling system, electricity system and irradiation facility systems. Possible simulations are normal reactor operation, unusual transient operation and accidental operation. The developed simulator also contains tool to revise/add facility in it for the future development. (author)

  14. Review of electricity supply failures and plant improvements over 25 years operation of the Harwell materials test reactors

    International Nuclear Information System (INIS)

    Taylor, D.J.

    1986-01-01

    The evolution of the on-site electrical power sources is described, operational experience is reported and shortcomings are identified. Disturbances in the external power supplies to the reactors are listed for the past 25 years and failure probabilities are derived from this historical data. The 132 kV overhead supply to the Harwell site is identified as the source of nearly 90% of the disturbances. (author)

  15. Review of electricity supply failures and plant improvements over 25 years operation of the Harwell materials test reactors

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, D. J. [UKAEA Harwell (United Kingdom)

    1986-02-15

    The evolution of the on-site electrical power sources is described, operational experience is reported and shortcomings are identified. Disturbances in the external power supplies to the reactors are listed for the past 25 years and failure probabilities are derived from this historical data. The 132 kV overhead supply to the Harwell site is identified as the source of nearly 90% of the disturbances. (author)

  16. 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.

  17. Reactors for nuclear electric propulsion

    International Nuclear Information System (INIS)

    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

  18. General Electric Company proposed test and evaluation plan, commercial buildings. National Solar Demonstration Program

    Energy Technology Data Exchange (ETDEWEB)

    None

    1976-04-01

    The general requirements and methods for instrumenting, testing, and evaluating solar HVAC systems forming a part of ERDA's ''Commercial Demonstration Program'' commensurate with ERDA 23A and the Proposed Management Plan 75SDS4270 are defined. Design requirements are specified for the performance of components and subsystems comprising the instrumentation and data gathering system, as well as the support functions required to perform the diagnostic measurements, collection and processing of data, and documentation of reports on solar HVAC system performance, including economic and societal evaluations.

  19. General Electric Nuclear Energy

    International Nuclear Information System (INIS)

    2001-01-01

    The ESBWR is a 1380 MWe boiling water reactor with improved operating safety margins and passive safety systems. He stated that the ESBWR derived from earlier GE plant design certification efforts and is the result of eight years of International cooperative work. He stated that the biggest challenge is to cross the regulatory hurdles associated with the inspections, tests, analyses, and acceptance criteria (ITAAC) and combined license (COL) programs. He further stated that he did not know how long it might take to license the ESBWR, in part, because the last GE design certification took about 8 to 10 years. Dr. Rao also provided a brief overview of the GE Nuclear Advance Liquid Metal S-PRISM design

  20. Reactor operator screening test experiences

    International Nuclear Information System (INIS)

    O'Brien, W.J.; Penkala, J.L.; Witzig, W.F.

    1976-01-01

    When it became apparent to Duquesne Light Company of Pittsburgh, Pennsylvania, that the throughput of their candidate selection-Phase I training-reactor operator certification sequence was something short of acceptable, the utility decided to ask consultants to make recommendations with respect to candidate selection procedures. The recommendation implemented was to create a Nuclear Training Test that would predict the success of a candidate in completing Phase I training and subsequently qualify for reactor operator certification. The mechanics involved in developing and calibrating the Nuclear Training Test are described. An arbitration decision that resulted when a number of International Brotherhood of Electrical Workers union employees filed a grievance alleging that the selection examination was unfair, invalid, not job related, inappropriate, and discriminatorily evaluated is also discussed. The arbitration decision favored the use of the Nuclear Training Test

  1. Reactor recirculation pump test loop

    International Nuclear Information System (INIS)

    Taka, Shusei; Kato, Hiroyuki

    1979-01-01

    A test loop for a reactor primary loop recirculation pumps (PLR pumps) has been constructed at Ebara's Haneda Plant in preparation for production of PLR pumps under license from Byron Jackson Pump Division of Borg-Warner Corporation. This loop can simulate operating conditions for test PLR pumps with 130 per cent of the capacity of pumps for a 1100 MWe BWR plant. A main loop, primary cooling system, water demineralizer, secondary cooling system, instrumentation and control equipment and an electric power supply system make up the test loop. This article describes the test loop itself and test results of two PLR pumps for Fukushima No. 2 N.P.S. Unit 1 and one main circulation pump for HAZ Demonstration Test Facility. (author)

  2. Severe accident testing of electrical penetration assemblies

    International Nuclear Information System (INIS)

    Clauss, D.B.

    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

  3. 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

  4. Advanced Demonstration and Test Reactor Options Study

    International Nuclear Information System (INIS)

    Petti, David Andrew; Hill, R.; Gehin, J.; Gougar, Hans David; Strydom, Gerhard; Heidet, F.; Kinsey, J.; Grandy, Christopher; Qualls, A.; Brown, Nicholas; Powers, J.; Hoffman, E.; Croson, D.

    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'. Advanced reactors are

  5. Development of a general coupling interface for the fuel performance code TRANSURANUS – Tested with the reactor dynamics code DYN3D

    International Nuclear Information System (INIS)

    Holt, L.; Rohde, U.; Seidl, M.; Schubert, A.; Van Uffelen, P.; Macián-Juan, R.

    2015-01-01

    Highlights: • A general coupling interface was developed for couplings of the TRANSURANUS code. • With this new tool simplified fuel behavior models in codes can be replaced. • Applicable e.g. for several reactor types and from normal operation up to DBA. • The general coupling interface was applied to the reactor dynamics code DYN3D. • The new coupled code system DYN3D–TRANSURANUS was successfully tested for RIA. - Abstract: A general interface is presented for coupling the TRANSURANUS fuel performance code with thermal hydraulics system, sub-channel thermal hydraulics, computational fluid dynamics (CFD) or reactor dynamics codes. As first application the reactor dynamics code DYN3D was coupled at assembly level in order to describe the fuel behavior in more detail. In the coupling, DYN3D provides process time, time-dependent rod power and thermal hydraulics conditions to TRANSURANUS, which in case of the two-way coupling approach transfers parameters like fuel temperature and cladding temperature back to DYN3D. Results of the coupled code system are presented for the reactivity transient scenario, initiated by control rod ejection. More precisely, the two-way coupling approach systematically calculates higher maximum values for the node fuel enthalpy. These differences can be explained thanks to the greater detail in fuel behavior modeling. The numerical performance for DYN3D–TRANSURANUS was proved to be fast and stable. The coupled code system can therefore improve the assessment of safety criteria, at a reasonable computational cost

  6. Tokamak engineering test reactor

    International Nuclear Information System (INIS)

    Conn, R.W.; Jassby, D.L.

    1975-07-01

    The design criteria for a tokamak engineering test reactor can be met by operating in the two-component mode with reacting ion beams, together with a new blanket-shield design based on internal neutron spectrum shaping. A conceptual reactor design achieving a neutron wall loading of about 1 MW/m 2 is presented. The tokamak has a major radius of 3.05 m, the plasma cross-section is noncircular with a 2:1 elongation, and the plasma radius in the midplane is 55 cm. The total wall area is 149 m 2 . The plasma conditions are T/sub e/ approximately T/sub i/ approximately 5 keV, and ntau approximately 8 x 10 12 cm -3 s. The plasma temperature is maintained by injection of 177 MW of 200-keV neutral deuterium beams; the resulting deuterons undergo fusion reactions with the triton-target ions. The D-shaped toroidal field coils are extended out to large major radius (7.0 m), so that the blanket-shield test modules on the outer portion of the torus can be easily removed. The TF coils are superconducting, using a cryogenically stable TiNb design that permits a field at the coil of 80 kG and an axial field of 38 kG. The blanket-shield design for the inner portion of the torus nearest the machine center line utilizes a neutron spectral shifter so that the first structural wall behind the spectral shifter zone can withstand radiation damage for the reactor lifetime. The energy attenuation in this inner blanket is 8 x 10 -6 . If necessary, a tritium breeding ratio of 0.8 can be achieved using liquid lithium cooling in the []outer blanket only. The overall power consumption of the reactor is about 340 MW(e). A neutron wall loading greater than 1 MW/m 2 can be achieved by increasing the maximum magnetic field or the plasma elongation. (auth)

  7. Advanced Carbothermal Electric Reactor, Phase I

    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...

  8. Electric hydrogen recombiner special tests

    International Nuclear Information System (INIS)

    Wilson, J.F.

    1975-12-01

    Westinghouse has produced an electric hydrogen recombiner to control hydrogen levels in reactor containments following a postulated loss-of-coolant accident. The recombiner underwent extensive testing for NRC qualification (see WCAP 7709-L and Supplements 1, 2, 3, 4). As a result, WCAP 7709-L and Supplements 1, 2, 3, and 4 have been accepted by the NRC for reference in applications not committed to IEEE-323-1974. Supplement 5 and the next supplement will demonstrate conformance to IEEE-323-1974. This supplement describes additional tests, beyond those necessary to qualify the system, which will be referenced in supplement 6. Each test has demonstrated a considerable margin of safety over required performance. Concurrently, the test results increased the fund of technical information on the electric hydrogen recombiner

  9. Research reactors and materials testing

    International Nuclear Information System (INIS)

    Vidal, H.

    1986-01-01

    Research reactors can be classified in three main groups according to the moderator which is used. Their technical characteristics are given and the three most recent research and materials testing reactors are described: OSIRIS, ORPHEE and the high-flux reactor of Grenoble. The utilization of research reactors is reviewed in four fields of activity: training, fundamental or applied research and production (eg. radioisotopes) [fr

  10. Service to the Electric Utility Industry by the Ford Nuclear Reactor, University of Michigan

    International Nuclear Information System (INIS)

    Burn, R.R.; Simpson, P.A.; Cook, G.M.

    1993-01-01

    Since 1977, the staff of the University of Michigan's Ford Nuclear Reactor has been providing irradiation, testing, analytical, and training services to electric utilities and to suppliers of the nuclear electric utility industry. This paper discusses the reactor's irradiation facilities; reactor programs and utilization; materials testing programs; neutron activation analysis activities; and training programs conducted

  11. Test reactor risk assessment methodology

    International Nuclear Information System (INIS)

    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

  12. Test reactors in the world

    International Nuclear Information System (INIS)

    Corella, M.R.; Gomez Alonso, M.

    1983-01-01

    INFCE work on research reactor core conversion from HEU to LEU, attracted a raising interest on this type of nuclear reactors. In this context, the present work shows a compilation of worldwide research and test nuclear reactors, now in operation, under construction, or planned, as well as decommissioned reactors (tables A to F). Brief descriptions of these reactors are included in tables G to L. In table M a summary view of reactors with power level between 10 and 30 MWt is shown. Attention is focused on that power range, as it has been considered in very preliminar studies for a new research reactor. Almost all data have been obtained from current available bibliography. (author)

  13. Simulator for materials testing reactors

    International Nuclear Information System (INIS)

    Takemoto, Noriyuki; Sugaya, Naoto; Ohtsuka, Kaoru; Hanakawa, Hiroki; Onuma, Yuichi; Hosokawa, Jinsaku; Hori, Naohiko; Kaminaga, Masanori; Tamura, Kazuo; Hotta, Kohji; Ishitsuka, Tatsuo

    2013-06-01

    A real-time simulator for both reactor and irradiation facilities of a materials testing reactor, “Simulator of Materials Testing Reactors”, was developed for understanding reactor behavior and operational training in order to utilize it for nuclear human resource development and to promote partnership with developing countries which have a plan to introduce nuclear power plant. The simulator is designed based on the JMTR (Japan Materials Testing Reactor), and it simulates operation, irradiation tests and various kinds of anticipated operational transients and accident conditions caused by the reactor and irradiation facilities. The development of the simulator was sponsored by the Japanese government as one of the specialized projects of advanced research infrastructure in order to promote basic as well as applied researches. This report summarizes the simulation components, hardware specification and operation procedure of the simulator. (author)

  14. Scyllac fusion test reactor design

    International Nuclear Information System (INIS)

    Dudziak, D.J.; Gerstl, S.A.; Houck, D.L.; Jalbert, R.A.; Krakowski, R.A.; Linford, R.K.; McDonald, T.E.; Rogers, J.D.; Thomassen, K.I.

    1975-01-01

    A general design of the system is given. The implosion heating and compression systems (METS) are described. Tritium handling, shielding and activation of the reactor, and safety and environmental aspects are discussed

  15. General Electric's training program for BWR chemists

    International Nuclear Information System (INIS)

    Osborn, R.N.; Lim, W.

    1981-01-01

    This paper describes the development and implementation of the General Electric boiling water reactor chemistry training program from 1959 to the present. The original intention of this program was to provide practical hands on type training in radiochemistry to BWR chemistry supervisors with fossil station experience. This emphasis on radiochemistry has not changed through the years, but the training has expanded to include the high purity water chemistry of the BWR and has been modified to include new commission requirements, engineering developments and advanced instrumentation. Student and instructor qualifications are discussed and a description of the spin off courses for chemistry technicians and refresher training is presented

  16. Broad-Application Test Reactor

    International Nuclear Information System (INIS)

    Motloch, C.G.

    1992-05-01

    This report is about a new, safe, and operationally efficient DOE reactor of nuclear research and testing proposed for the early to mid- 21st Century. Dubbed the Broad-Application Test Reactor (BATR), the proposed facility incorporates a multiple-application, multiple-mission design to support DOE programs such as naval reactors and space power and propulsion, as well as research in medical, science, isotope, and electronics arenas. DOE research reactors are aging, and implementing major replacement projects requires long lead times. Primary design drivers include safety, low risk, minimum operation cost, mission flexibility, waste minimization, and long life. Scientists and engineers at the Idaho National Engineering Laboratory are evaluating possible fuel forms, structural materials, reactor geometries, coolants, and moderators

  17. Electrical insulators for the theta-pinch fusion reactor

    International Nuclear Information System (INIS)

    Clinard, F.W. Jr.

    1976-01-01

    The five major applications for electrical insulators in the Reference Theta Pinch Reactor are as follows: (1) first-wall insulator, (2) blanket intersegment insulator, (3) graphite encapsulating insulator, (4) implosion coil insulator, and (5) compression coil insulator. Insulator design proposals and some preliminary test results are given for each application

  18. Reactor design for nuclear electric propulsion

    International Nuclear Information System (INIS)

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

    1979-01-01

    Conceptual design studies of a nuclear power plant for electric propulsion of spacecrafts have been on going for several years. An attractive concept which has evolved from these studies and which has been described in previous publications, is a heat-pipe cooled, fast spectrum nuclear reactor that provides 3 MW of thermal energy to out-of-core thermionic converters. The primary motivation for using heat pipes is to provide redundancy in the core cooling system that is not available in gas or liquid-metal cooled reactors. Detailed investigation of the consequences of heat pipe failures has resulted in modifications to the basic reactor design and has led to consideration of an entirely different core design. The new design features an integral laminated core configuration consisting of alternating layers of UO 2 and molybdenum sheets that span the entire diameter of the core. Design characteristics are presented and compared for the two reactors

  19. Nuclear reactors for space electric power

    International Nuclear Information System (INIS)

    Buden, D.

    1978-06-01

    The Los Alamos Scientific Laboratory is studying reactor power plants for space applications in the late 1980s and 1990s. The study is concentrating on high-temperature, compact, fast reactors that can be coupled with various radiation shielding systems and thermoelectric, dynamic, or thermionic electric power conversion systems, depending on the mission. Lifetimes of 7 to 10 yr at full power, at converter operating temperatures of 1275 to 1675 0 K, are being studied. The systems are being designed such that no single-failure modes exist that will cause a complete loss of power. In fact, to meet the long lifetimes, highly redundant design features are being emphasized. Questions have been raised about safety since the COSMOS 954 incident. ''Fail-safe'' means to prevent exposure of the population to radioactive material, meeting the environmental guidelines established by the U.S. Government have been and continue to be a necessary requirement for any space reactor program. The major safety feature to prevent prelaunch and launch radioactive material hazards is not operating the reactor before achieving the prescribed orbit. Design features in the reactor ensure that accidental criticality cannot occur. High orbits (above 400 to 500 nautical miles) have sufficient lifetimes to allow radioactive elements to decay to safe levels. The major proposed applications for satellites with reactors in Earth orbit are in geosynchronous orbit (19,400 nautical miles). In missions at geosynchronous orbit, where orbital lifetimes are practically indefinite, the safety considerations are negligible. Orbits below 400 to 500 nautical miles are the ones where a safety issue is involved in case of satellite malfunction. The potential missions, the question of why reactors are being considered as a prime power candidate, reactor features, and safety considerations will be discussed

  20. Fire safety requirements for electrical cables towards nuclear reactor safety

    International Nuclear Information System (INIS)

    Raju, M.R.

    2002-01-01

    Full text: Electrical power supply forms a very important part of any nuclear reactor. Power supplies have been categorized in to class I, II, III and IV from reliability point. The safety related equipment are provided with highly reliable power supply to achieve the safety of very high order. Vast network of cables in a nuclear reactor are grouped and segregated to ensure availability of power to at least one group under all anticipated occurrences. Since fire can result in failures leading to unavailability of power caused by common cause, both passive and active fire protection methods are adopted in addition to fire detection system. The paper describes the requirement for passive fire protection to electrical cables viz. fire barrier and fire breaks. The paper gives an account of the tests required to standardize the products. Fire safety implementation for cables in research reactors is described

  1. PITR: Princeton Ignition Test Reactor

    International Nuclear Information System (INIS)

    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

  2. Reliability modeling of Clinch River breeder reactor electrical shutdown systems

    International Nuclear Information System (INIS)

    Schatz, R.A.; Duetsch, K.L.

    1974-01-01

    The initial simulation of the probabilistic properties of the Clinch River Breeder Reactor Plant (CRBRP) electrical shutdown systems is described. A model of the reliability (and availability) of the systems is presented utilizing Success State and continuous-time, discrete state Markov modeling techniques as significant elements of an overall reliability assessment process capable of demonstrating the achievement of program goals. This model is examined for its sensitivity to safe/unsafe failure rates, sybsystem redundant configurations, test and repair intervals, monitoring by reactor operators; and the control exercised over system reliability by design modifications and the selection of system operating characteristics. (U.S.)

  3. Modular helium reactor for non-electric applications

    International Nuclear Information System (INIS)

    Shenoy, A.

    1997-01-01

    The high temperature gas-cooled Modular Helium Reactor (MHR) is an advanced, high efficiency reactor system which can play a vital role in meeting the future energy needs of the world by contributing not only to the generation of electric power, but also the non-electric energy traditionally served by fossil fuels. This paper summarizes work done over 20 years, by several people at General Atomics, how the Modular Helium Reactor can be integrated to provide different non-electric applications during Process Steam/Cogeneration for industrial application, Process Heat for transportation fuel development and Hydrogen Production for various energy applications. The MHR integrates favorably into present petrochemical and primary metal process industries, heavy oil recovery, and future shale oil recovery and synfuel processes. The technical fit of the Process Steam/Cogeneration Modular Helium Reactor (PS/C-MHR) into these processes is excellent, since it can supply the required quantity and high quality of steam without fossil superheating. 12 refs, 25 figs, 2 tabs

  4. Diagnosis of electric equipment at the Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    Nguyen Truong Sinh

    1999-01-01

    The Dalat Nuclear Research Reactor (DNRR) is a pool type of its kind in the world: Soviet-designed core and control system harmoniously integrated into the left-over infrastructure of the former American-made TRIGA MARK II reactor, which includes the reactor tank and shielding, graphite reflector, beam tubes and thermal column. The reactor is mainly used for radioisotope and radiopharmaceutical production, elemental analysis using neutron activation techniques, neutron beam exploitation, silicon doping, and reactor physics experimentation. For safe operation of the reactor maintenance work has been carried out for the reactor control and instrumentation, reactor cooling, ventilation, radiomonitoring, mechanical, normal electric supply systems as well as emergency electric diesel generators and the water treatment station. Technical management of the reactor includes periodical maintenance as required by technical specifications, training, re-training and control of knowledge for reactor staff. During recent years, periodic preventive maintenance (PPM) has been carried out for the electric machines of the technological systems. (author)

  5. the JHR Material Testing Reactor

    International Nuclear Information System (INIS)

    Roure, C.; Cornu, B.; Berthet, B.; Simon, E.; Estre, N.; Guimbal, P.; Kinnunen, P.; Kotiluoto, P.

    2013-06-01

    The Jules Horowitz Reactor (JHR) is a European experimental reactor under construction in CEA Cadarache. It will be dedicated to material and fuel irradiation tests, and to medical isotopes production. Non-Destructive nuclear Examinations systems (NDE) will be implemented in pools to analyse the irradiated fuel or tested material in their supporting experimental irradiation devices extracted from the core or its immediate periphery. The Nuclear Measurement Laboratory (NML) of CEA Cadarache is working in collaboration with VTT (Technical Research Centre in Finland) in designing and developing NDE systems implementing gamma-ray spectroscopy and high energy X-ray imaging of the sample and irradiation device. CEA is also designing a neutron radiography system for which NML is working on the detection system. Design studies are performed with Monte Carlo transport codes and specific simulation tools developed by the NML for Xray and neutron imaging. (authors)

  6. 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)

    1999-12-31

    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)

  7. 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)

  8. Reactor transients tests for SNR fuel elements in HFR reactor

    International Nuclear Information System (INIS)

    Plitz, H.

    1989-01-01

    In HFR reactor, fuel pins of LMFBR reactors are putted in irradiation specimen capsules cooled with sodium for reactor transients tests. These irradiation capsules are instrumented and the experiences realized until this day give results on: - Fuel pins subjected at a continual variation of power - melting fuel - axial differential elongation of fuel pins

  9. Cryogenic Electric Motor Tested

    Science.gov (United States)

    Brown, Gerald V.

    2004-01-01

    Technology for pollution-free "electric flight" is being evaluated in a number of NASA Glenn Research Center programs. One approach is to drive propulsive fans or propellers with electric motors powered by fuel cells running on hydrogen. For large transport aircraft, conventional electric motors are far too heavy to be feasible. However, since hydrogen fuel would almost surely be carried as liquid, a propulsive electric motor could be cooled to near liquid hydrogen temperature (-423 F) by using the fuel for cooling before it goes to the fuel cells. Motor windings could be either superconducting or high purity normal copper or aluminum. The electrical resistance of pure metals can drop to 1/100th or less of their room-temperature resistance at liquid hydrogen temperature. In either case, super or normal, much higher current density is possible in motor windings. This leads to more compact motors that are projected to produce 20 hp/lb or more in large sizes, in comparison to on the order of 2 hp/lb for large conventional motors. High power density is the major goal. To support cryogenic motor development, we have designed and built in-house a small motor (7-in. outside diameter) for operation in liquid nitrogen.

  10. Irradiation Facilities at the Advanced Test Reactor

    International Nuclear Information System (INIS)

    S. Blaine Grover

    2005-01-01

    The Advanced Test Reactor (ATR) is the third generation and largest test reactor built in the Reactor Technology Complex (RTC) (formerly known as the Test Reactor Area), located at the Idaho National Laboratory (INL), to study the effects of intense neutron and gamma radiation on reactor materials and fuels. The RTC was established in the early 1950s with the development of the Materials Testing Reactor (MTR), which operated until 1970. The second major reactor was the Engineering Test Reactor (ETR), which operated from 1957 to 1981, and finally the ATR, which began operation in 1967 and will continue operation well into the future. These reactors have produced a significant portion of the world's data on materials response to reactor environments. The wide range of experiment facilities in the ATR and the unique ability to vary the neutron flux in different areas of the core allow numerous experiment conditions to co-exist during the same reactor operating cycle. Simple experiments may involve a non-instrumented capsule containing test specimens with no real-time monitoring or control capabilities. More sophisticated testing facilities include inert gas temperature control systems and pressurized water loops that have continuous chemistry, pressure, temperature, and flow control as well as numerous test specimen monitoring capabilities. There are also apparatus that allow for the simulation of reactor transients on test specimens

  11. Ground testing of an SP-100 prototypic reactor

    International Nuclear Information System (INIS)

    Motwani, K.; Pflasterer, G.R.; Upton, H.; Lazarus, J.D.; Gluck, R.

    1988-01-01

    SP-100 is a space power system which is being developed by GE to meet future space electrical power requirements. The ground testing of an SP-100 prototypic reactor system will be conducted at the Westinghouse Hanford Company site located at Richland, Washington. The objective of this test is to demonstrate the performance of a full scale prototypic reactor system, including the reactor, control system and flight shield. The ground test system is designed to simulate the flight operating conditions while meeting all the necessary nuclear safety requirements in a gravity environment. The goal of the reactor ground test system is to establish confidence in the design maturity of the SP-100 space reactor power system and resolve the technical issues necessary for the development of a flight mission design

  12. Off reactor testings. Technological engineering applicative research

    International Nuclear Information System (INIS)

    Doca, Cezar

    2001-01-01

    By the end of year 2000 over 400 nuclear electro-power units were operating world wide, summing up a 350,000 MW total capacity, with a total production of 2,300 TWh, representing 16% of the world's electricity production. Other 36 units, totalizing 28,000 MW, were in construction, while a manifest orientation towards nuclear power development was observed in principal Asian countries like China, India, Japan and Korea. In the same world's trend one find also Romania, the Cernavoda NPP Unit 1 generating electrical energy into the national system beginning with 2 December 1996. Recently, the commercial contract was completed for finishing the Cernavoda NPP Unit 2 and launching it into operation by the end of year 2004. An important role in developing the activity of research and technological engineering, as technical support for manufacturing the CANDU type nuclear fuel and supplying with equipment the Cernavoda units, was played by the Division 7 TAR of the INR Pitesti. Qualification testings were conducted for: - off-reactor CANDU type nuclear fuel; - FARE tools, pressure regulators, explosion proof panels; channel shutting, as well as functional testing for spare pushing facility as a first step in the frame of the qualification tests for the charging/discharging machine (MID) 4 and 5 endings. Testing facilities are described, as well as high pressure hot/cool loops, measuring chains, all of them fulfilling the requirements of quality assurance. The nuclear fuel off-reactor tests were carried out to determine: strength; endurance; impact, pressure fall and wear resistance. For Cernavoda NPP equipment testings were carried out for: the explosion proof panels, pressure regulators, behaviour to vibration and wear of the steam generation tubings, effects of vibration upon different electronic component, channel shutting (for Cernavoda Unit 2), MID operating at 300 and 500 cycles. A number of R and D programs were conducted in the frame of division 7 TAR of INR

  13. The fast reactor and electricity supply, a utility view

    International Nuclear Information System (INIS)

    Wright, J.K.; Hall, R.S.; Kemmish, W.B.; Thorne, R.T.

    1982-01-01

    The significance of the fast reactor is discussed from the viewpoint of the Central Electricity Generating Board. The need for the fast reactor and a possible timescale for its introduction are examined. It is emphasised that demonstration of the commercial and environmental acceptability of the fuel cycle will be needed before any commitment can be made to fast reactors. (U.K.)

  14. 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.

  15. The direct conversion of heat into electricity in reactors

    International Nuclear Information System (INIS)

    Devin, B.; Bliaux, J.; Lesueur, R.

    1964-01-01

    The direct conversion of heat into electricity by thermionic emission in an atomic reactor has been studied with the triple aim of its utilisation: as an energy source for a space device, at the head of a conventional conversion system in power installations, or finally in association with the thermoelectric conversion in very low power installations. The laboratory experiments were mainly orientated towards the electron extraction of metals and compounds and their behaviour at high temperatures. Converters furnishing up to 50 amps at 0. 4 volts with an efficiency close to 10 p. 100 have been constructed in the laboratory; the emitters were heated by electron bombardment and were composed of tungsten covered with an uranium carbide deposit or molybdenum covered with cesium. The main aspects of the coupling between the converter and the reactor have been covered from the point of view of electronics: the influence of the mismatching of the load on the temperature of the emitter and the influence of thermal flux density on the temperature of the emitter and the stability of the converter. Converters using uranium carbide as the electron emitter have been tested in reactors. Tests have been made under dynamic conditions in order to determine the dynamic characteristics. The load matching curves have been constructed and the overall performances of several cells coupled in such a way as to form a reactor rod have been deduced. This information is fundamental to the design of a control system for a thermionic conversion reactor. The problems associated with the reliability of thermionic converters connected in series in the same reactor rod have been examined theoretically. Finally, the absorption isotherms have been drawn at the ambient temperatures for krypton and xenon on activated carbon with the aim of investigating the escape of fission products in a converter. (author) [fr

  16. 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)

  17. Development of a general coupling interface for the fuel performance code transuranus tested with the reactor dynamic code DYN3D

    International Nuclear Information System (INIS)

    Holt, L.; Rohde, U.; Seidl, M.; Schubert, A.; Van Uffelen, P.

    2013-01-01

    Several institutions plan to couple the fuel performance code TRANSURANUS developed by the European Institute for Transuranium Elements with their own codes. One of these codes is the reactor dynamic code DYN3D maintained by the Helmholtz-Zentrum Dresden - Rossendorf. DYN3D was developed originally for VVER type reactors and was extended later to western type reactors. Usually, the fuel rod behavior is modeled in thermal hydraulics and neutronic codes in a simplified manner. The main idea of this coupling is to describe the fuel rod behavior in the frame of core safety analysis in a more detailed way, e.g. including the influence of the high burn-up structure, geometry changes and fission gas release. It allows to take benefit from the improved computational power and software achieved over the last two decades. The coupling interface was developed in a general way from the beginning. Thence it can be easily used also by other codes for a coupling with TRANSURANUS. The user can choose between a one-way as well as a two-way online coupling option. For a one-way online coupling, DYN3D provides only the time-dependent rod power and thermal hydraulics conditions to TRANSURANUS, but the fuel performance code doesn’t transfer any variable back to DYN3D. In a two-way online coupling, TRANSURANUS in addition transfers parameters like fuel temperature and cladding temperature back to DYN3D. This list of variables can be extended easily by geometric and further variables of interest. First results of the code system DYN3D-TRANSURANUS will be presented for a control rod ejection transient in a modern western type reactor. Pre-analyses show already that a detailed fuel rod behavior modeling will influence the thermal hydraulics and thence also the neutronics due to the Doppler reactivity effect of the fuel temperature. The coupled code system has therefore a potential to improve the assessment of safety criteria. The developed code system DYN3D-TRANSURANUS can be used also

  18. Control Performance of General Electric Fuel and Torque Regulator Operating on T31-3 Turbine-Propeller Engine in Sea-Level Test Stand

    Science.gov (United States)

    Oppenheimer, Frank L.; Lazar, James

    1951-01-01

    A .General Electric fuel and torque regulator was tested in conjunction with a T31-3 turbine-propeller engine in the sea-level static test stand at the NACA Lewis laboratory. The engine and control were operated over the entire speed range: 11,000 rpm, nominal flight idle, to 13,000 rpm, full power. Steady-state and transient data were recorded and are presented with a description of the four control loops being used in the system. Results of this investigation indicated that single-lever control operation was satisfactory under conditions of test. Transient data presented showed that turbine-outlet temperature did overshoot maximum operating value on acceleration but that the time duration of overshoot did not exceed approximately 1 second. This temperature limiting resulted from a control on fuel flow as a function of engine speed. Speed and torque first reached their desired values 0.4 second from the time of change in power-setting lever position. Maximum speed overshoot was 3 percent.

  19. Test reactor: basic to U.S. breeder reactor development

    International Nuclear Information System (INIS)

    Miller, B.J.; Harness, A.J.

    1975-01-01

    Long-range energy planning in the U. S. includes development of a national commercial breeder reactor program. U. S. development of the LMFBR is following a conservative sequence of extensive technology development through use of test reactors and demonstration plants prior to construction of commercial plants. Because materials and fuel technology development is considered the first vital step in this sequence, initial U. S. efforts have been directed to the design and construction of a unique test reactor. The Fast Flux Test Facility, FFTF, is a 400 MW(t) reactor with driver fuel locations, open test locations, and closed loops for higher risk experiments. The FFTF will provide a prototypic LMFBR core environment with sufficient instrumentation for detailed core environmental characterization and a testing capability substituted for breeder capability. The unique comprehensive fuel and materials testing capability of the FFTF will be key to achieving long-range objectives of increased power density, improved breeding gain and shorter doubling times. (auth)

  20. Economic aspects of electricity and industrial heat generating reactors

    International Nuclear Information System (INIS)

    Gaussens, J.; Moulle, N.; Dutheil, F.

    1964-01-01

    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) [fr

  1. Proposal of space reactor for nuclear electric propulsion system

    International Nuclear Information System (INIS)

    Nishiyama, Takaaki; Nagata, Hidetaka; Nakashima, Hideki

    2009-01-01

    A nuclear reactor installed in spacecrafts is considered here. The nuclear reactor could stably provide an enough amount of electric power in deep space missions. Most of the nuclear reactors that have been developed up to now in the United States and the former Soviet Union have used uranium with 90% enrichment of 235 U as a fuel. On the other hand, in Japan, because the uranium that can be used is enriched to below 20%, the miniaturization of the reactor core is difficult. A Light-water nuclear reactor is an exception that could make the reactor core small. Then, the reactor core composition and characteristic are evaluated for the cases with the enrichment of the uranium fuel as 20%. We take up here Graphite reactor, Light-water reactor, and Sodium-cooled one. (author)

  2. Ground test facility for nuclear testing of space reactor subsystems

    International Nuclear Information System (INIS)

    Quapp, W.J.; Watts, K.D.

    1985-01-01

    Two major reactor facilities at the INEL have been identified as easily adaptable for supporting the nuclear testing of the SP-100 reactor subsystem. They are the Engineering Test Reactor (ETR) and the Loss of Fluid Test Reactor (LOFT). In addition, there are machine shops, analytical laboratories, hot cells, and the supporting services (fire protection, safety, security, medical, waste management, etc.) necessary to conducting a nuclear test program. This paper presents the conceptual approach for modifying these reactor facilities for the ground engineering test facility for the SP-100 nuclear subsystem. 4 figs

  3. 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.

  4. Safety re-assessment of AECL test and research reactors

    International Nuclear Information System (INIS)

    Winfield, D.J.

    1990-01-01

    Atomic Energy of Canada Limited currently has four operating engineering test/research reactors of various sizes and ages; a new isotope-production reactor Maple-X10, under construction at Chalk River Nuclear Laboratories (CRNL), and a heating demonstration reactor, SDR, undergoing high-power commissioning at Whiteshell Nuclear Research Establishment (WNRE). The company is also performing design studies of small reactors for hot water and electricity production. The older reactors are ZED-2, PTR, NRX, and NRU; these range in age from 42 years (NRX) to 29 years (ZED-2). Since 1984, limited-scope safety re-assessments have been underway on three of these reactors (ZED-2, NRX AND NRU). ZED-2 and PTR are operated by the Reactor Physics Branch; all other reactors are operated by the respective site Reactor Operations Branches. For the older reactors the original safety reports produced were entirely deterministic in nature and based on the design-basis accident concept. The limited scope safety re-assessments for these older reactors, carried out over the past 5 years, have comprised both quantitative probabilistic safety-assessment techniques, such as event tree and fault analysis, and/or qualitative techniques, such as failure mode and effect analysis. The technique used for an individual assessment was dependent upon the specific scope required. This paper discusses the types of analyses carried out, specific insights/recommendations resulting from the analysis, and the plan for future analysis. In addition, during the last four years safety assessments have been carried out on the new isotope-, heat-, and electricity-producing reactors, as part of the safety design review, commissioning and licensing activities

  5. Evaluation of neutronic characteristics of in-pile test reactor for fast reactor safety research

    Energy Technology Data Exchange (ETDEWEB)

    Uto, N.; Ohno, S.; Kawata, N. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1996-09-01

    An extensive research program has been carried out at the Power Reactor and Nuclear Fuel Development Corporation for the safety of future liquid-metal fast breeder reactors to be commercialized. A major part of this program is investigation and planning of advanced safety experiments conducted with a new in-pile safety test facility, which is larger and more advanced than any of the currently existing test reactors. Such a transient safety test reactor generally has unique neutronic characteristics that require various studies from the reactor physics point of view. In this paper, the outcome of the neutronics study is highlighted with presenting a reference core design concept and its performance in regard to the safety test objectives. (author)

  6. Irradiation facilitates at the advanced test reactor

    International Nuclear Information System (INIS)

    Grover, Blaine S.

    2006-01-01

    The Advanced Test Reactor (ATR) is the third generation and largest test reactor built in the Reactor Technology Complex (RTC - formerly known as the Test Reactor Area), located at the Idaho National Laboratory (INL), to study the effects of intense neutron and gamma radiation on reactor materials and fuels. The RTC was established in the early 1950's with the development of the Materials Testing Reactor (MTR), which operated until 1970. The second major reactor was the Engineering Test Reactor (ETR), which operated from 1957 to 1981, and finally the ATR, which began operation in 1967 and will continue operation well into the future. These reactors have produced a significant portion of the world's data on materials response to reactor environments. The wide range of experiment facilities in the ATR and the unique ability to vary the neutron flux in different areas of the core allow numerous experiment conditions to co-exist during the same reactor operating cycle. Simple experiments may involve a non-instrumented capsule containing test specimens with no real-time monitoring or control capabilities. More sophisticated testing facilities include inert gas temperature control systems and pressurized water loops that have continuous chemistry, pressure, temperature, and flow control as well as numerous test specimen monitoring capabilities. There are also apparatus that allow for the simulation of reactor transients on test specimens. The paper has the following contents: ATR description and capabilities; ATR operations, quality and safety requirements; Static capsule experiments; Lead experiments; Irradiation test vehicle; In-pile loop experiments; Gas test loop; Future testing; Support facilities at RTC; Conclusions. To summarize, the ATR has a long history in fuel and material irradiations, and will be fulfilling a critical role in the future fuel and material testing necessary to develop the next generation reactor systems and advanced fuel cycles. The

  7. Design and testing of reactors for 735 kV

    Energy Technology Data Exchange (ETDEWEB)

    Erb, W; Kraaij, D J

    1965-11-01

    The design and testing of five large, single phase shunt reactors rated either 110 or 55 MVAR, supplied for the 735 kV system of the Quebec Hydro Electric Commission which came into operation in the autumn of 1965 are described. As these reactors are permanently connected to the transmission lines, their losses must be considered as being continuously present and must be determined exactly. In addition to the use of a new bridge method, the losses were also measured calorimetrically for the purpose of comparison, the agreement between the two tests being remarkably good. The impulse tests with full wave and chopped wave are subsequently described.

  8. Nuclear science. U.S. electricity needs and DOE's civilian reactor development program

    International Nuclear Information System (INIS)

    England-Joseph, Judy; Allen, Robert E. Jr.; Fitzgerald, Duane; Young, Edward E. Jr.; Leavens, William P.; Bell, Jacqueline

    1990-05-01

    Electricity projections developed by the North American Electric Reliability Council (NERC) appear to be the best available estimates of future U.S. electricity needs. NERC, which represents all segments of the utility industry, forecasts that before 1998 certain regions of the country, particularly in the more heavily populated eastern half of the United States, may experience shortfalls during summer peak demand periods. These forecasts considered the utility companies' plans, as of 1989, to meet electricity needs during the period; these plans include such measures as constructing additional generators and conducting demand management programs. Working closely with the nuclear industry, DOE is supporting the development of several reactor technologies to ensure that nuclear power remains a viable electricity supply option. In fiscal year 1990, DOE's Civilian Reactor Development Program was funded at $253 million. DOE is using these funds to support industry-led efforts to develop light water reactors (LWR), advanced liquid-metal reactors (LMR), and modular high-temperature gas-cooled reactors (MHTGR) that are safe, environmentally acceptable, and economically competitive. The utility company officials we spoke with, all of whom were in the Southeast, generally supported DOE's efforts in developing these technologies. However, most of the officials do not plan to purchase nuclear reactors until after 2000 because of the high costs of constructing nuclear reactors and current public opposition to nuclear power

  9. A General Overview of Electric Road Vehicles

    International Nuclear Information System (INIS)

    Lamblin, Veronique

    2018-01-01

    In July 2017 Nicolas Hulot, the French Minister of Ecological and Inclusive Transition, presented a climate plan featuring an end to electricity generation from coal by 2022, a reduction in the nuclear component of electricity supply by one third, a total ban on the sale of petrol or diesel cars by 2040 and an incentive scheme designed gradually to remove polluting vehicles from the roads. Other European partners are following suit and promoting the spread of electric vehicles (Norway, Germany, Netherlands etc.). Yet is this the panacea that will meet the targets for greenhouse gas reduction in the battle against climate change? Futuribles examines the question in this issue with two articles: the first of these by Pierre Bonnaure, above, assesses the forces driving the spread of electric cars and the impediments to that process; this second article by Veronique Lamblin offers a general over - view of electric road vehicles (passenger cars, heavy good vehicles, bicycles etc.) throughout the world. (author)

  10. Thermionic reactor power conditioner design for nuclear electric propulsion.

    Science.gov (United States)

    Jacobsen, A. S.; Tasca, D. M.

    1971-01-01

    Consideration of the effects of various thermionic reactor parameters and requirements upon spacecraft power conditioning design. A basic spacecraft is defined using nuclear electric propulsion, requiring approximately 120 kWe. The interrelationships of reactor operating characteristics and power conditioning requirements are discussed and evaluated, and the effects on power conditioner design and performance are presented.

  11. Advanced reactor development for non-electric applications

    International Nuclear Information System (INIS)

    Chang, M.H.; Kim, S.H.

    1996-01-01

    Advance in the nuclear reactor technology achieved through nuclear power programs carried out in the world has led nuclear communities to direct its attention to a better and peaceful utilization of nuclear energy in addition to that for power generation. The efforts for non-electric application of nuclear energy has been pursued in a limited number of countries in the world for their special needs. However, those needs and the associated efforts contributed largely to the development and practical realization of advanced reactors characterized by highly improved reactor safety and reliability by deploying the most up-to-date safety technologies. Due mainly to the special purpose of utilization, economic reasons and ease in implementation of new advanced technologies, small and medium reactors have become a major stream in the reactor developments for non-electric applications. The purpose of this paper is to provide, to the interested nuclear society, the overview of the development status and design characteristics of selected advanced nuclear reactors previously developed and/or currently under development specially for non-electric applications. Major design technologies employed in those reactors to enhance the reactor safety and reliability are reviewed to present the underlying principles of the design. Along with the overview, this paper also introduces a development program and major design characteristics of an advanced integral reactor (SMART) for co-generation purpose currently under conceptual development in Korea. (author)

  12. The SPHINX reactor for engineering tests

    International Nuclear Information System (INIS)

    Adamov, E.O.; Artamkin, K.N.; Bovin, A.P.; Bulkin, Y.M.; Kartashev, E.F.; Korneev, A.A.; Stenbok, I.A.; Terekhov, A.S.; Khmel'Shehikov, V.V.; Cherkashov, Y.M.

    1990-01-01

    A research reactor known as SPHINX is under development in the USSR. The reactor will be used mainly to carry out tests on mock-up power reactor fuel assemblies under close-to-normal parameters in experimental loop channels installed in the core and reflector of the reactor, as well as to test samples of structural materials in ampoule and loop channels. The SPHINX reactor is a channel-type reactor with light-water coolant and moderator. Maximum achievable neutron flux density in the experimental channels (cell composition 50% Fe, 50% H 2 O) is 1.1 X 10 15 neutrons/cm 2 · s for fast neutrons (E > 0.1 MeV) and 1.7 X 10 15 for thermal neutrons at a reactor power of 200 MW. The design concepts used represent a further development of the technical features which have met with approval in the MR and MIR channel-type engineering test reactors currently in use in the USSR. The 'in-pond channel' construction makes the facility flexible and eases the carrying out of experimental work while keeping discharges of radioactivity into the environment to a low level. The reactor and all associated buildings and constructions conform to modern radiation safety and environmental protection requirements

  13. Proceedings of the international symposium on materials testing reactors

    International Nuclear Information System (INIS)

    Ishihara, Masahiro; Kawamura, Hiroshi

    2009-01-01

    This report is the Proceedings of the International Symposium on Materials Testing Reactors hosted by Japan Atomic Energy Agency (JAEA). The symposium was held on July 16 to 17, 2008, at the Oarai Research and Development Center of JAEA. This symposium was also held for the 40th anniversary ceremony of Japan Materials Testing Reactor (JMTR) from achieving its first criticality. The objective of the symposium is to exchange the information on current status, future plan and so on among each testing reactors for the purpose of mutual understanding. There were 138 participants from Argentina, Belgium, France, Indonesia, Kazakhstan, Korea, the Russian Federation, Sweden, the United State, Vietnam and Japan. The symposium was divided into four technical sessions and three topical sessions. Technical sessions addressed the general topics of 'status and future plan of materials testing reactors', 'material development for research and testing reactors', irradiation technology (including PIE technology)' and 'utilization with materials testing reactors', and 21 presentations were made. Also the topical sessions addressed 'establishment of strategic partnership', 'management on re-operation work at reactor trouble' and 'basic technology for neutron irradiation tests in MTRs', and panel discussion was made. The 21 of the presented papers are indexed individually. (J.P.N.)

  14. Development of conductor feedthrough module of LV electrical penetration assembly for research reactors

    International Nuclear Information System (INIS)

    Luo Zhiyuan; Wang Guangjin; Zhou Bin

    2007-01-01

    A LV electrical penetration assembly with perfusion sealing conductor feedthrough module was developed, which can be used for the connection of internal and external cables through the wall of the research reactor workshop. The LV electrical penetration assembly was combined with several independent modules. The maintenance and replacement of the assembly can be easily done in service. The sealing of conductor feedthrough module was achieved with the perfusion of self-extinguishing epoxy. The leakage between the conductor feedthrough module and the end plate module was blocked with rubber rings. The result of the leakage test and the electrical performance test for the samples of conductor feedthrough module satisfied the requirement of research reactor. The structure of the new electrical penetration assembly is simple and compact. It can be manufactured with mature technology and cost low price. The performance of the assembly is steady. It can be used widely in research reactors. (authors)

  15. New facilities in Japan materials testing reactor for irradiation test of fusion reactor components

    International Nuclear Information System (INIS)

    Kawamura, H.; Sagawa, H.; Ishitsuka, E.; Sakamoto, N.; Niiho, T.

    1996-01-01

    The testing and evaluation of fusion reactor components, i.e. blanket, plasma facing components (divertor, etc.) and vacuum vessel with neutron irradiation is required for the design of fusion reactor components. Therefore, four new test facilities were developed in the Japan Materials Testing Reactor: an in-pile functional testing facility, a neutron multiplication test facility, an electron beam facility, and a re-weldability facility. The paper describes these facilities

  16. Electrical system regulations of the IEA-R1 reactor

    International Nuclear Information System (INIS)

    Mello, Jose Roberto de; Madi Filho, Tufic

    2013-01-01

    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)

  17. Material test reactor fuel research at the BR2 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dyck, Steven Van; Koonen, Edgar; Berghe, Sven van den [Institute for Nuclear Materials Science, SCK-CEN, Boeretang, Mol (Belgium)

    2012-03-15

    The construction of new, high performance material test reactor or the conversion of such reactors' core from high enriched uranium (HEU) to low enriched uranium (LEU) based fuel requires several fuel qualification steps. For the conversion of high performance reactors, high density dispersion or monolithic fuel types are being developed. The Uranium-Molybdenum fuel system has been selected as reference system for the qualification of LEU fuels. For reactors with lower performance characteristics, or as medium enriched fuel for high performance reactors, uranium silicide dispersion fuel is applied. However, on the longer term, the U-Mo based fuel types may offer a more efficient fuel alternative and-or an easier back-end solution with respect to the silicide based fuels. At the BR2 reactor of the Belgian nuclear research center, SCK-CEN in Mol, several types of fuel testing opportunities are present to contribute to such qualification process. A generic validation test for a selected fuel system is the irradiation of flat plates with representative dimensions for a fuel element. By flexible positioning and core loading, bounding irradiation conditions for fuel elements can be performed in a standard device in the BR2. For fuel element designs with curved plates, the element fabrication method compatibility of the fuel type can be addressed by incorporating a set of prototype fuel plates in a mixed driver fuel element of the BR2 reactor. These generic types of tests are performed directly in the primary coolant flow conditions of the BR2 reactor. The experiment control and interpretation is supported by detailed neutronic and thermal-hydraulic modeling of the experiments. Finally, the BR2 reactor offers the flexibility for irradiation of full size prototype fuel elements, as 200mm diameter irradiation channels are available. These channels allow the accommodation of various types of prototype fuel elements, eventually using a dedicated cooling loop to provide the

  18. Reliability test for reactor internals rejuvenation technology

    International Nuclear Information System (INIS)

    Uchiyama, Junichi

    1998-01-01

    41 transparencies were presented on the subject of 'Reliability test for reactor internals rejuvenation technology'. The items presented give an introduction on the management of plant life in Japan and introduce the Nuclear Power Engineering Corporation (NUPEC). The question of what reliability tests for rejuvenation of reactor internals are is discussed in some detail and an outline of each test is given. Altogether six methods to rejuvenate reactor internals are presented, two of which have already been applied to actual plants. The presentation was supported by many detailed drawings and images

  19. General remarks on fast neutron reactor physics

    International Nuclear Information System (INIS)

    Barre, J.Y.

    1980-01-01

    The main aspects of fast reactor physics, presented in these lecture notes, are restricted to LMFBR's. The emphasis is placed on the core neutronic balance and the burn-up problems. After a brief description of the power reactor main components and of the fast reactor chronology, the fundamental parameters of the one-group neutronic balance are briefly reviewed. Then the neutronic burn-up problems related to the Pu production and to the doubling time are considered

  20. General aspects of CAREM-25 reactor

    International Nuclear Information System (INIS)

    Delmastro, Dario F.; Gomez, Silvia; Ishida, Viviana; Mazzi, Ruben; Santecchia, Alberto; Gomez de Soler, Susana M.

    2000-01-01

    CAREM project consists on the development and design of an advanced nuclear power plant. In order to verify its innovative features the construction of a prototype is planned. In this paper the main technical characteristics of CAREM-25 prototype reactor are presented. This is a very low power innovative reactor (100 M Wth) conceived with new generation design solutions. Based on an indirect cycle integrated light water reactor using enriched uranium, CAREM has some distinctive features that greatly simplify the reactor and also contribute to a high level of safety: -) Integrated primary system; -) Primary system cooling by natural convection; -) Self pressurization; -) and Passive safety systems. (author)

  1. General Aspects of CAREM-25 Reactor

    International Nuclear Information System (INIS)

    Delmastro, Dario; Gomez, S.; Mazzi, R.; Gomez de Soler, S.; Santecchia, A.; Ishida, V.

    2000-01-01

    CAREM project consists on the development and design of an advanced Nuclear Power Plant. In order to verify its innovative features the construction of a prototype is planned. In this paper the main technical characteristics of CAREM-25 prototype reactor are presented. This is a very low power innovative reactor (100MWth) conceived with new generation design solutions. Based on an indirect cycle integrated light water reactor using enriched uranium, CAREM has some distinctive features that greatly simplify the reactor and also contribute to a high level of safety: integrated primary system, primary system cooling by natural convection, selfpressurization, and passive safety systems

  2. Patch Test Negative Generalized Dermatitis.

    Science.gov (United States)

    Spiker, Alison; Mowad, Christen

    2016-01-01

    Allergic contact dermatitis is a common condition in dermatology. Patch testing is the criterion standard for diagnosis. However, dermatitis is not always caused by an allergen, and patch testing does not identify a culprit in every patient. Generalized dermatitis, defined as eczematous dermatitis affecting greater than 3 body sites, is often encountered in dermatology practice, especially patch test referral centers. Management for patients with generalized dermatitis who are patch test negative is challenging. The purpose of this article is to outline an approach to this challenging scenario and summarize the paucity of existing literature on patch test negative generalized dermatitis.

  3. Reactor power control method upon accidents of electrical power system

    International Nuclear Information System (INIS)

    Hirose, Masao.

    1983-01-01

    Purpose: To enable to continue the operation of a BWR type reactor by avoiding the scram while suppressing the reactor power, just after the external disturbance such as earth-trouble in power-transmission network. Method: Steep power drop of an electrical generator is to be detected not only by a current-type power-load-unbalance relay but also with a power-type power-load-unbalance-relay. If steep power-drop was detected by the latter relay, a previously selected control rod is rapidly inserted into the reactor. In this way, in the case where there is a possibility of the reactor scram, the scram can be avoided by suppressing the reactor power, thus the reactor operation can be continued. (Kamimura, M.)

  4. Design and testing of integrated circuits for reactor protection channels

    International Nuclear Information System (INIS)

    Battle, R.E.; Vandermolen, R.I.; Jagadish, U.; Swail, B.K.; Naser, J.

    1995-01-01

    Custom and semicustom application-specific integrated circuit design and testing methods are investigated for use in research and commercial nuclear reactor safety systems. The Electric Power Research Institute and Oak Ridge National Laboratory are working together through a cooperative research and development agreement to apply modern technology to a nuclear reactor protection system. The purpose of this project is to demonstrate to the nuclear industry an alternative approach for new or upgrade reactor protection and safety system signal processing and voting logic. Motivation for this project stems from (1) the difficulty of proving that software-based protection systems are adequately reliable, (2) the obsolescence of the original equipment, and (3) the improved performance of digital processing. A demonstration model for protection system of PWR reactor has been designed and built

  5. Advanced Test Reactor National Scientific User Facility

    International Nuclear Information System (INIS)

    Marshall, Frances M.; Benson, Jeff; Thelen, Mary Catherine

    2011-01-01

    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.

  6. Advanced Test Reactor probabilistic risk assessment

    International Nuclear Information System (INIS)

    Atkinson, S.A.; Eide, S.A.; Khericha, S.T.; Thatcher, T.A.

    1993-01-01

    This report discusses Level 1 probabilistic risk assessment (PRA) incorporating a full-scope external events analysis which has been completed for the Advanced Test Reactor (ATR) located at the Idaho National Engineering Laboratory

  7. 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.

  8. Tokamaks with high-performance resistive magnets: advanced test reactors and prospects for commercial applications

    International Nuclear Information System (INIS)

    Bromberg, L.; Cohn, D.R.; Williams, J.E.C.; Becker, H.; Leclaire, R.; Yang, T.

    1981-10-01

    Scoping studies have been made of tokamak reactors with high performance resistive magnets which maximize advantages gained from high field operation and reduced shielding requirements, and minimize resistive power requirements. High field operation can provide very high values of fusion power density and n tau/sub e/ while the resistive power losses can be kept relatively small. Relatively high values of Q' = Fusion Power/Magnet Resistive Power can be obtained. The use of high field also facilitates operation in the DD-DT advanced fuel mode. The general engineering and operational features of machines with high performance magnets are discussed. Illustrative parameters are given for advanced test reactors and for possible commercial reactors. Commercial applications that are discussed are the production of fissile fuel, electricity generation with and without fissioning blankets and synthetic fuel production

  9. Simulating Neutronic Core Parameters in a Research and Test Reactor

    International Nuclear Information System (INIS)

    Selim, H.K.; Amin, E.A.; Koutb, M.E.

    2011-01-01

    The present study proposes an Artificial Neural Network (ANN) modeling technique that predicts the control rods positions in a nuclear research reactor. The neutron, flux in the core of the reactor is used as the training data for the neural network model. The data used to train and validate the network are obtained by modeling the reactor core with the neutronic calculation code: CITVAP. The type of the network used in this study is the feed forward multilayer neural network with the backpropagation algorithm. The results show that the proposed ANN has good generalization capability to estimate the control rods positions knowing neutron flux for a research and test reactor. This method can be used to predict critical control rods positions to be used for reactor operation after reload

  10. 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.

  11. Artificial Intelligence Research at General Electric

    OpenAIRE

    Sweet, Larry

    1985-01-01

    General Electric is engaged in a broad range of research and development activities in artificial intelligence, with the dual objectives of improving the productivity of its internal operations and of enhancing future products and services in its aerospace, industrial, aircraft engine, commercial, and service sectors. Many of the applications projected for AI within GE will require significant advances in the state of the art in advanced inference, formal logic, and architectures for real-tim...

  12. Status report on nuclear reactors for space electric power

    International Nuclear Information System (INIS)

    Buden, D.

    1978-01-01

    The Los Alamos Scientific Laboratory is studying reactor power plants for space applications in the late 1980s and 1990s. The study is concentrating on high-temperature, compact, fast reactors that can be coupled with various radiation shielding systems and thermoelectric, dynamic, or thermionic electric power conversion systems, depending on the mission. Increased questions have been raised about safety since the COSMOS 954 incident. High orbits (above 400 to 500 nautical miles) have sufficient lifetimes to allow radioactive elements to decay to safe levels. The major proposed applications for satellites with reactors in Earth orbit are in geosynchronous orbit (19,400 nautical miles). In missions at geosynchronous orbit where orbital lifetimes are practically indefinite, the safety considerations are negligible. The potential missions, why reactors are being considered as a prime power candidate, reactor features, and safety considerations are discussed

  13. The advanced test reactor strategic evaluation program

    International Nuclear Information System (INIS)

    Buescher, B.J.

    1989-01-01

    Since the Chernobly accident, the safety of test reactors and irradiation facilities has been critically evaluated from the public's point of view. A systematic evaluation of all safety, environmental, and operational issues must be made in an integrated manner to prioritize actions to maximize benefits while minimizing costs. Such a proactive program has been initiated at the Advanced Test Reactor (ATR). This program, called the Strategic Evaluation Program (STEP), is being conducted for the ATR to provide integrated safety and operational reviews of the reactor against the standards applied to licensed commercial power reactors. This has taken into consideration the lessons learned by the US Nuclear Regulatory Commission (NRC) in its Systematic Evaluation Program (SEP) and the follow-on effort known as the Integrated Safety Assessment Program (ISAP). The SEP was initiated by the NRC to review the designs of older operating nuclear power plants to confirm and document their safety. The ATR STEP objectives are discussed

  14. Non-electric Applications of Fast Reactors

    International Nuclear Information System (INIS)

    Safa, Henri; Borgard, Jean-Marc

    2013-01-01

    Conclusions: → Most of industrial applications (80%) require low temperature heat below 540°C; → Fast Reactors are technically suitable to provide industrial steam at temperatures not accessible by standard LWRs; → As an illustrative example, the application at an oil refinery site has been studied showing the economic benefits; → Nuclear Cogeneration enhances the overall energy efficiency of the power plant; • Nuclear Cogeneration allows massive cut in CO 2 emissions

  15. Experiences in stability testing of boiling water reactors

    International Nuclear Information System (INIS)

    March-Leuba, J.; Otaduy, P.J.

    1986-01-01

    The purpose of this paper is to summarize experiences with boiling water reactor (BWR) stability testing using noise analysis techniques. These techniques have been studied over an extended period of time, but it has been only recently that they have been well established and generally accepted. This paper contains first a review of the problem of BWR neutronic stability, focusing on its physical causes and its effects on reactor operation. The paper also describes the main techniques used to quantify, from noise measurements, the reactor's stability in terms of a decay ratio. Finally, the main results and experiences obtained from the stability tests performed at the Dresden and the Browns Ferry reactors using noise analysis techniques are summarized

  16. A quality assurance program for nuclear power reactor materials tests at the Ford nuclear reactor

    International Nuclear Information System (INIS)

    Burn, R.R.

    1989-01-01

    The University of Michigan Nuclear Reactor Laboratory Quality Assurance Program has been established to assure that materials testing services provided to electric utilities produce accurate results in accordance with industry standards, sound engineering practice, and customer requirements. The program was prepared to comply with applicable requirements of 10CFR50, Appendix B, of the Code of Federal Regulations and a standard of the American National Standards Institute (ANSI), N45.2. The paper discusses the quality assurance program applicability, organization, qualification and training of personnel, material identification and control, examination and testing, measuring and test equipment, nonconforming test equipment, records, audits, and distribution

  17. Reliability tests for reactor internals replacement technology

    International Nuclear Information System (INIS)

    Fujimaki, K.; Uchiyama, J.; Ohtsubo, T.

    2000-01-01

    Structural damage due to aging degradation of LWR reactor internals has been reported in several nuclear plants. NUPEC has started a project to test the reliability of the technology for replacing reactor internals, which was directed at preventive maintenance before damage and repair after damage for the aging degradation. The project has been funded by the Ministry of International Trade and Industry (MITI) of Japan since 1995, and it follows the policy of a report that the MITI has formally issued in April 1996 summarizing the countermeasures to be considered for aging nuclear plants and equipment. This paper gives an outline of the whole test plans and the test results for the BWR reactor internals replacement methods; core shroud, ICM housing, and CRD Housing and stub tube. The test results have shown that the methods were reliable and the structural integrity was appropriate based on the evaluation. (author)

  18. Cosmological tests of general relativity

    International Nuclear Information System (INIS)

    Hut, P.

    1977-01-01

    It is stated that the general relativity theory could be tested on a cosmological scale by measuring the Hubble constant and the deceleration parameter, if, in addition, everything could be known about the matter filling the universe. If, on the other hand, nothing could be presupposed about the matter content of the universe general relativity could not be tested by measuring any number of time derivatives of the scale factor. But upon making the assumption of a universe filled with non-interacting mixture of non-relativistic matter and radiation general relativity can in principle be tested by measuring the first five derivatives of the scale factor. Some general relations are here presented using this assumption. (author)

  19. Advanced Instrumentation for Transient Reactor Testing

    Energy Technology Data Exchange (ETDEWEB)

    Corradini, Michael L.; Anderson, Mark; Imel, George; Blue, Tom; Roberts, Jeremy; Davis, Kurt

    2018-01-31

    Transient testing involves placing fuel or material into the core of specialized materials test reactors that are capable of simulating a range of design basis accidents, including reactivity insertion accidents, that require the reactor produce short bursts of intense highpower neutron flux and gamma radiation. Testing fuel behavior in a prototypic neutron environment under high-power, accident-simulation conditions is a key step in licensing nuclear fuels for use in existing and future nuclear power plants. Transient testing of nuclear fuels is needed to develop and prove the safety basis for advanced reactors and fuels. In addition, modern fuel development and design increasingly relies on modeling and simulation efforts that must be informed and validated using specially designed material performance separate effects studies. These studies will require experimental facilities that are able to support variable scale, highly instrumented tests providing data that have appropriate spatial and temporal resolution. Finally, there are efforts now underway to develop advanced light water reactor (LWR) fuels with enhanced performance and accident tolerance. These advanced reactor designs will also require new fuel types. These new fuels need to be tested in a controlled environment in order to learn how they respond to accident conditions. For these applications, transient reactor testing is needed to help design fuels with improved performance. In order to maximize the value of transient testing, there is a need for in-situ transient realtime imaging technology (e.g., the neutron detection and imaging system like the hodoscope) to see fuel motion during rapid transient excursions with a higher degree of spatial and temporal resolution and accuracy. There also exists a need for new small, compact local sensors and instrumentation that are capable of collecting data during transients (e.g., local displacements, temperatures, thermal conductivity, neutron flux, etc.).

  20. General description of advanced heavy water reactor

    International Nuclear Information System (INIS)

    Kakodkar, A.; Sinha, R.K.; Dhawan, M.L.

    1999-01-01

    Advanced Heavy Water Reactor is a boiling light water cooled, heavy water moderated and vertical pressure tube type reactor with its design optimised for utilisation of thorium for power generation. The core consists of (Th-U 233 )O 2 and (Th-Pu)O 2 fuel with a discharge burn up of 20,000 MWd/Te. This reactor incorporates several features to simplify the design, which eliminate certain systems and components. AHWR design is also optimised for easy replaceability of coolant channels, facilitation of in-service inspection and maintenance and ease of erection. The AHWR design also incorporates several passive systems for performing safety-related functions in the event of an accident. In case of LOCA, emergency coolant is injected through 4 accumulators of 260 m 3 capacity directly into the core. Gravity driven water pool of capacity 6000 m 3 serves to cool the core for 3 days without operator's intervention. Core submergence, passive containment isolation and passive containment cooling are the added features in AHWR. The paper describes the various process systems, core and fuel design, primary components and safety concepts of AHWR. Plant layout and technical data are also presented. The conceptual design of the reactor has been completed, and the detailed design and development is scheduled for completion in the year 2002. (author)

  1. Systematic evaluation program review of NRC Safety Topic VI-10.A associated with the electrical, instrumentation and control portions of the testing of reactor trip system and engineered safety features, including response time for the Dresden station, Unit II nuclear power plant

    International Nuclear Information System (INIS)

    St Leger-Barter, G.

    1980-11-01

    This report documents the technical evaluation and review of NRC Safety Topic VI-10.A, associated with the electrical, instrumentation, and control portions of the testing of reactor trip systems and engineered safety features including response time for the Dresden II nuclear power plant, using current licensing criteria

  2. Needs for development in nondestructive testing for advanced reactor systems

    International Nuclear Information System (INIS)

    McClung, R.W.

    1978-01-01

    The needs for development of nondestructive testing (NDT) techniques and equipment were surveyed and analyzed relative to problem areas for the Liquid-Metal Fast Breeder Reactor, the Molten-Salt Breeder Reactor, and the Advanced Gas-Cooled Reactor. The paper first discusses the developmental needs that are broad-based requirements in nondestrutive testing, and the respective methods applicable, in general, to all components and reactor systems. Next, the requirements of generic materials and components that are common to all advanced reactor systems are examined. Generally, nondestructive techniques should be improved to provide better reliability and quantitativeness, improved flaw characterization, and more efficient data processing. Specific recommendations relative to such methods as ultrasonics, eddy currents, acoustic emission, radiography, etc., are made. NDT needs common to all reactors include those related to materials properties and degradation, welds, fuels, piping, steam generators, etc. The scope of applicability ranges from initial design and material development stages through process control and manufacturing inspection to in-service examination

  3. TESTING OF GAS REACTOR MATERIALS AND FUEL IN THE ADVANCED TEST REACTOR

    International Nuclear Information System (INIS)

    Grover, S.B.

    2004-01-01

    The Advanced Test Reactor (ATR) has long been involved in testing gas reactor materials, and has developed facilities well suited for providing the right conditions and environment for gas reactor tests. This paper discusses the different types of irradiation hardware that have been utilized in past ATR irradiation tests of gas reactor materials. The new Gas Test Loop facility currently being developed for the ATR is discussed and the different approaches being considered in the design of the facility. The different options for an irradiation experiment such as active versus passive temperature control, neutron spectrum tailoring, and different types of lead experiment sweep gas monitors are also discussed. The paper is then concluded with examples of different past and present gas reactor material and fuel irradiations

  4. Testing of Gas Reactor Materials and Fuel in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    S. Blaine Grover

    2004-01-01

    The Advanced Test Reactor (ATR) has long been involved in testing gas reactor materials, and has developed facilities well suited for providing the right conditions and environment for gas reactor tests. This paper discusses the different types of irradiation hardware that have been utilized in past ATR irradiation tests of gas reactor materials. The new Gas Test Loop facility currently being developed for the ATR is discussed and the different approaches being considered in the design of the facility. The different options for an irradiation experiment such as active versus passive temperature control, neutron spectrum tailoring, and different types of lead experiment sweep gas monitors are also discussed. The paper is then concluded with examples of different past and present gas reactor material and fuel irradiations

  5. Startup testing of Romania dual-core test reactor

    International Nuclear Information System (INIS)

    Whittemore, W.L.

    1980-01-01

    Late in 1979 both the Annular Core Pulsed Reactor (ACPR) and the 14-MW steady-state reactor (SSR) were loaded to critical. The fuel loading in both was then carried to completion and low-power testing was conducted. Early in 1980 both reactors successfully underwent high-power testing. The ACPR was operated for several hours at 500 kW and underwent pulse tests culminating in pulses with reactivity insertions of $4.60, peak power levels of about 20,000 MW, energy releases of 100 MW-sec, and peak measured fuel temperatures of 830 deg. C. The SSR was operated in several modes, both with natural convection and forced cooling with one or more pumps. The reactor successfully completed a 120-hr full-power test. Subsequent fuel element inspections confirmed that the fuel has performed without fuel damage or distortion. (author)

  6. Reliability tests for reactor internals rejuvenation technology

    International Nuclear Information System (INIS)

    Fujimaki, Katsumi; Hitoki, Yoichi; Otsubo, Toru; Uchiyama, Junichi

    1998-01-01

    Structural damage due to aging degradation of LWR reactor internals has been reported in several nuclear plants. NUPEC has started a project to test the reliability of the technology for rejuvenating reactor internals which has been funded by the Ministry of International Trade and Industry (MITI) of Japan since 1995. The project follows the policy of a report that the MITI has formally issued in April 1996 summarizing the countermeasures to be considered for aging nuclear plants and equipment. This paper gives an outline of the test plans and results which are directed at preventive maintenance before damage and repair after damage for reactor internals aging degradation. The test results for the replacement methods of ICM housing and BWR core shroud have shown that the methods were reliable and the structural integrity was appropriate based on the evaluation. (author)

  7. Electrical cabling system associated at a nuclear reactor

    International Nuclear Information System (INIS)

    Dejeux, P.; Desfontaines, G.

    1988-01-01

    This cabling system for an electrical device in a nuclear reactor comprises at least a first cable issued of the device, a second cable comprising a first portion, a second portion and a third portion joining the second by a multiple quick fitting connector capable to connect at least ten second portions at ten other third portions of the second cable [fr

  8. Design and testing of integrated circuits for reactor protection channels

    International Nuclear Information System (INIS)

    Battle, R.E.; Vandermolen, R.I.; Jagadish, U.; Swail, B.K.; Naser, J.; Rana, I.

    1995-01-01

    Custom and semicustom application-specific integrated circuit design and testing methods are investigated for use in research and commercial nuclear reactor safety systems. The Electric Power Research Institute and Oak Ridge National Laboratory are working together through a cooperative research and development agreement to apply modern technology to a nuclear reactor protection system. Purpose of this project is to demonstrate to the nuclear industry an alternative approach for new or upgrade reactor protection and safety system signal processing and voting logic. Motivation for this project stems from (1) the difficulty of proving that software-based protection systems are adequately reliable, (2) the obsolescence of the original equipment, and (3) the improved performance of digital processing

  9. Pump selection and application in a pressurized water reactor electric generating plant

    International Nuclear Information System (INIS)

    Kitch, D.M.

    1985-01-01

    Various pump applications utilized in a nuclear pressurized water reactor electric generating plant are described. Emphasis is on pumps installed in the auxiliary systems of the primary nuclear steam supply system. Hydraulic and mechanical details, the ASME Code (Nuclear Design), materials, mechanical seals, shaft design, seismic qualification, and testing are addressed

  10. Reactor group constants and benchmark test

    Energy Technology Data Exchange (ETDEWEB)

    Takano, Hideki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-08-01

    The evaluated nuclear data files such as JENDL, ENDF/B-VI and JEF-2 are validated by analyzing critical mock-up experiments for various type reactors and assessing applicability for nuclear characteristics such as criticality, reaction rates, reactivities, etc. This is called Benchmark Testing. In the nuclear calculations, the diffusion and transport codes use the group constant library which is generated by processing the nuclear data files. In this paper, the calculation methods of the reactor group constants and benchmark test are described. Finally, a new group constants scheme is proposed. (author)

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

    International Nuclear Information System (INIS)

    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

  12. 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.

  13. Implementation of the rapid cross section adjustment approach at General Electric

    International Nuclear Information System (INIS)

    Cowan, C.L.; Kujawski, E.; Protsik, R.

    1978-01-01

    The General Electric rapid cross section adjustment approach was developed to use the shielding factor method for formulating multigroup cross sections. In this approach, space- and composition-dependent cross sections for a particular reactor or shield design are prepared from a generalized cross section library by the use of resonance self-shielding factors, and by the adjustment of elastic scattering cross sections for the local neutron flux spectra. The principal tool in the cross section adjustment package is the data processing code TDOWN. This code was specified to give the user a high degree of flexibility in the analysis of advanced reactor designs. Of particular interest in the analysis of critical experiments is the ability to carry out cell heterogeneity self-shielding calculations using a multiregion equivalence relationship, and the homogenization of the cross sections over the specified cell with the flux weighting obtained from transport theory calculations. Extensive testing of the rapid cross section adjustment approach, including comparisons with Monte Carlo methods, indicated that this approach can be utilized with a high degree of confidence in the design analysis of complex fast reactor systems. 2 figures, 1 table

  14. Nondestructive testing of nuclear reactor components integrity

    International Nuclear Information System (INIS)

    Mala, M.; Miklos, M.

    2011-01-01

    Nuclear energy must respond to current challenges in the energy market. The significant parameters are increase of the nuclear fuel price, closed fuel cycle, reduction and safe and the final disposal of high level radioactive waste. Nowadays, the discussions on suitable energy mix are taking place not only here in Czech Republic, but also in many other European countries. It is necessary to establish an appropriate ratio among the production of electricity from conventional, nuclear and renewable energy sources. Also, it is necessary to find ways how to streamline the economy, central part of the nuclear fuel cycle and thereby to increase the competitiveness of nuclear energy. This streamlining can be carried out by improving utilization of existing nuclear fuel with maintaining a high degree of nuclear facilities safety. Increasing operational reliability and safety together with increasing utilization of nuclear fuel place increasing demands on monitoring of changes during fuel burnup. The potential fuel assembly damages in light water reactors are prevented by the introduction of new procedures and programs of the fuel assembly monitoring. One of them is the Post Irradiation Inspection Program (PIIP) which is a good tool for monitoring of chemical regime impact on the fuel assembly cladding behavior. Main nondestructive techniques that are used at nuclear power plants for the fuel assembly integrity evaluation are ultrasonic measurements, eddy current measurements, radiographic testing, acoustic techniques and others. Ultrasonic system is usual tool for leak fuel rod evaluation and it is also used at Temelin NPP. Since 2009, Temelin NPP has cooperated with Research Center Rez Ltd in frame of PIIP program at both units WWER 1000. This program was established for US VVantage6 fuel assemblies and also it continues for Russian TVSA-T fuel assemblies. (author)

  15. General Relativity: horizons for tests

    Science.gov (United States)

    Yatskiv, Ya. S.; Alexandrov, A. N.; Vavilova, I. B.; Zhdanov, V. I.; Zhuk, A. I.; Kudrya, Yu. N.; Parnovsky, S. L.; Fedorova, E. V.; Khmil, S. V.

    2013-12-01

    Theoretical basis of the General Relativity Theory (GRT), its experimental tests as well as GRT applications are briefly summarized taking into account the results of the last decade. The monograph addresses scientists, post-graduated students, and students specialized in the natural sciences as well as everyone who takes an interest in GRT.

  16. Electrical insulator requirements for mirror fusion reactors

    International Nuclear Information System (INIS)

    Condit, R.H.; Van Konynenburg, R.A.

    1977-01-01

    The requirements for mirror fusion electrical insulators are discussed. Insulators will be required at the neutral beam injectors, injector power supplies, direct converters, and superconducting magnets. Insulators placed at the neutral beam injectors will receive the greatest radiation exposure, 10 14 to 10 16 neutrons/m 2 .s and 0.3 to 3 Gy/s (10 5 to 10 6 R/h) of gamma rays, with shielding. Direct converter insulators may receive the highest temperature (up to 1300 0 K), but low voltage holding requirements. Insulators made from organic materials (e.g., plastics) for the magnet coils may be satisfactory. Immediate conductivity increases of all insulators result from gamma irradiation. With an upper limit to gamma flux exposures of 300 Gy/s in a minimally shielded region, the conductivity could reach 10 -6 S/m. Damage from neutron irradiation may not be serious during several years' exposure. Surface changes in ceramics at the neutral beam injector may be serious. The interior of the injector will contain atomic hydrogen, and sputtering may transfer material away from or onto the ceramic insulators. Unknown and potentially damaging interactions between irradiation, electric fields, temperature gradients, cycling of temperature, surface and joint reactions, sputtering, polarization, and electrotransport in the dielectrics are of concern. Materials research to deal with these problems is needed

  17. 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.; Godfroy, Thomas J.

    2010-01-01

    Component level testing of power conversion units proposed for use in fission surface power systems has typically been done using relatively simple electric heaters for thermal input. These heaters do not adequately represent the geometry or response of proposed reactors. As testing of fission surface power systems transitions from the component level to the system level it becomes necessary to more accurately replicate these reactors using reactor simulators. The Direct Drive Gas-Brayton Power Conversion Unit test activity at the NASA Glenn Research Center integrates a reactor simulator with an existing Brayton test rig. The response of the reactor simulator to a change in Brayton shaft speed is shown as well as the response of the Brayton to an insertion of reactivity, corresponding to a drum reconfiguration. The lessons learned from these tests can be used to improve the design of future reactor simulators which can be used in system level fission surface power tests.

  18. Space reactor electric systems: system integration studies, Phase 1 report

    International Nuclear Information System (INIS)

    Anderson, R.V.; Bost, D.; Determan, W.R.; Harty, R.B.; Katz, B.; Keshishian, V.; Lillie, A.F.; Thomson, W.B.

    1983-01-01

    This report presents the results of preliminary space reactor electric system integration studies performed by Rockwell International's Energy Systems Group (ESG). The preliminary studies investigated a broad range of reactor electric system concepts for powers of 25 and 100 KWe. The purpose of the studies was to provide timely system information of suitable accuracy to support ongoing mission planning activities. The preliminary system studies were performed by assembling the five different subsystems that are used in a system: the reactor, the shielding, the primary heat transport, the power conversion-processing, and the heat rejection subsystems. The subsystem data in this report were largely based on Rockwell's recently prepared Subsystem Technology Assessment Report. Nine generic types of reactor subsystems were used in these system studies. Several levels of technology were used for each type of reactor subsystem. Seven generic types of power conversion-processing subsystems were used, and several levels of technology were again used for each type. In addition, various types and levels of technology were used for the shielding, primary heat transport, and heat rejection subsystems. A total of 60 systems were studied

  19. High Flux Materials Testing Reactor (HFR), Petten

    International Nuclear Information System (INIS)

    1975-09-01

    After conversion to burnable poison fuel elements, the High Flux Materials Testing Reactor (HFR) Petten (Netherlands), operated through 1974 for 280 days at 45 MW. Equipment for irradiation experiments has been replaced and extended. The average annual occupation by experiments was 55% as compared to 38% in 1973. Work continued on thirty irradiation projects and ten development activities

  20. Present status of Japan materials testing reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hori, Naohiko; Kaminaga, Masanori; Kusunoki, Tsuyoshi; Ishihara, Masahiro; Niimi, Motoji; Komori, Yoshihiro; Suzuki, Masahide; Kawamura, Hiroshi [Japan Atomic Energy Agency, Oarai Research and Development Center, Oarai, Ibaraki (Japan)

    2012-03-15

    The Japan Materials Testing Reactor (JMTR) in Japan Atomic Energy Agency (JAEA) is a light water cooled tank type reactor with first criticality in March 1968. Owing to the connection between the JMTR and hot laboratory by a canal, easy re-irradiation tests can be conducted with safe and quick transportation of irradiated samples. The JMTR has been applied to fuel/material irradiation examinations for LWRs, HTGR, fusion reactor and RI production. However, the JMTR operation was once stopped in August 2006, and check and review on the reoperation had been conducted by internal as well as external committees. As a result of the discussion, the JMTR reoperation was determined, and refurbishment works started from the beginning of JFY 2007. The refurbishment works have finished in March 2011 taking four years from JFY 2007. Unfortunately, at the end of the JFY 2010 on March 11, the Great-Eastern-Japan-Earthquake occurred, and functional tests before the JMTR restart, such as cooling system, reactor control system and so on, were delayed by the earthquake. Moreover, a detail inspection found some damages such as slight deformation of the truss structure at the roof of the JMTR reactor building. Consequently, the restart of the JMTR will be delayed from June to next October, 2012. Now, the safety evaluation after the earthquake disaster is being carried out aiming at the restart of the JMTR. The renewed JMTR will be started from JFY 2012 and operated for a period of about 20 years until around JFY 2030. The usability improvement of the JMTR, e.g. higher reactor availability, shortening turnaround time to get irradiation results, attractive irradiation cost, business confidence, is also discussed with users as the preparations for re-operation. (author)

  1. Present status of Japan materials testing reactor

    International Nuclear Information System (INIS)

    Hori, Naohiko; Kaminaga, Masanori; Kusunoki, Tsuyoshi; Ishihara, Masahiro; Niimi, Motoji; Komori, Yoshihiro; Suzuki, Masahide; Kawamura, Hiroshi

    2012-01-01

    The Japan Materials Testing Reactor (JMTR) in Japan Atomic Energy Agency (JAEA) is a light water cooled tank type reactor with first criticality in March 1968. Owing to the connection between the JMTR and hot laboratory by a canal, easy re-irradiation tests can be conducted with safe and quick transportation of irradiated samples. The JMTR has been applied to fuel/material irradiation examinations for LWRs, HTGR, fusion reactor and RI production. However, the JMTR operation was once stopped in August 2006, and check and review on the reoperation had been conducted by internal as well as external committees. As a result of the discussion, the JMTR reoperation was determined, and refurbishment works started from the beginning of JFY 2007. The refurbishment works have finished in March 2011 taking four years from JFY 2007. Unfortunately, at the end of the JFY 2010 on March 11, the Great-Eastern-Japan-Earthquake occurred, and functional tests before the JMTR restart, such as cooling system, reactor control system and so on, were delayed by the earthquake. Moreover, a detail inspection found some damages such as slight deformation of the truss structure at the roof of the JMTR reactor building. Consequently, the restart of the JMTR will be delayed from June to next October, 2012. Now, the safety evaluation after the earthquake disaster is being carried out aiming at the restart of the JMTR. The renewed JMTR will be started from JFY 2012 and operated for a period of about 20 years until around JFY 2030. The usability improvement of the JMTR, e.g. higher reactor availability, shortening turnaround time to get irradiation results, attractive irradiation cost, business confidence, is also discussed with users as the preparations for re-operation. (author)

  2. Correlations between power and test reactor data bases

    International Nuclear Information System (INIS)

    Guthrie, G.L.; Simonen, E.P.

    1989-02-01

    Differences between power reactor and test reactor data bases have been evaluated. Charpy shift data has been assembled from specimens irradiated in both high-flux test reactors and low-flux power reactors. Preliminary tests for the existence of a bias between test and power reactor data bases indicate a possible bias between the weld data bases. The bias is nonconservative for power predictive purposes, using test reactor data. The lesser shift for test reactor data compared to power reactor data is interpreted primarily in terms of greater point defect recombination for test reactor fluxes compared to power reactor fluxes. The possibility of greater thermal aging effects during lower damage rates is also discussed. 15 refs., 5 figs., 2 tabs

  3. Electric power from near-term fusion reactors

    International Nuclear Information System (INIS)

    Longhurst, G.R.; Deis, G.A.; Miller, L.G.

    1981-01-01

    This paper examines requirements and possbilities of electric power production on near-term fusion reactors using low temperature cycle technology similar to that used in some geothermal power systems. Requirements include the need for a working fluid with suitable thermodynamics properties and which is free of oxygen and hydrogen to facilitate tritium management. Thermal storage will also be required due to the short system thermal time constants on near-time reactors. It is possbile to use the FED shield in a binary power cycle, and results are presented of thermodynamic analyses of this system

  4. Space-reactor electric systems: subsystem technology assessment

    International Nuclear Information System (INIS)

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

    1983-01-01

    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

  5. Nuclear fuels for material test reactors

    International Nuclear Information System (INIS)

    Ramanathan, L.V.; Durazzo, M.; Freitas, C.T. de

    1982-01-01

    Experimental results related do the development of nuclear fuels for reactors cooled and moderated by water have been presented cylindrical and plate type fuels have been described in which the core consists of U compouns dispersed in an Al matrix and is clad with aluminium. Fabrication details involving rollmilling, swaging or hot pressing have been described. Corrosion and irradiation test results are also discussed. The performance of the different types of fuels indicates that it is possible to locally fabricate fuel plates with U 3 O 8 +Al cores (20% enriched U) for use in operating Brazilian research reactors. (Author) [pt

  6. 78 FR 56594 - Airworthiness Directives; General Electric Company Turbofan Engines

    Science.gov (United States)

    2013-09-13

    ... Airworthiness Directives; General Electric Company Turbofan Engines AGENCY: Federal Aviation Administration (FAA... General Electric Company (GE) GE90-76B, -85B, -90B, -94B, -110B1, and - 115B turbofan engines. This AD was...) Applicability This AD applies to General Electric Company (GE): (1) GE90-76B, -85B, -90B, and -94B turbofan...

  7. 78 FR 72552 - Airworthiness Directives; General Electric Company Turbofan Engines

    Science.gov (United States)

    2013-12-03

    ... Airworthiness Directives; General Electric Company Turbofan Engines AGENCY: Federal Aviation Administration (FAA... General Electric Company model GEnx-2B67 and GEnx-2B67B turbofan engines. This AD was prompted by the... certain serial number General Electric Company (GE) model GEnx-2B67 and GEnx-2B67B turbofan engines. The...

  8. Standard Technical Specifications, General Electric plants, BWR/4

    International Nuclear Information System (INIS)

    1992-09-01

    This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/4, and documents the positions of the Nuclear Regulatory Commission based on the BWR Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power specifications. This report contains three volumes. This document, Volume 2, contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS

  9. Magnox Electric Littlebrook reactor inspection and repair rehearsal facility

    International Nuclear Information System (INIS)

    Barnes, S.A.; Clayton, R.; Gaydon, B.G.; Ramsey, B.H.

    1996-01-01

    Magnox reactors, although designed to be maintenance free during their operational life, have nevertheless highlighted the need for test rig facilities to train operators in the methods and techniques of reactor inspection and repair. The history of the facility for reactor engineering development (FRED) is described and its present role as a repair rehearsal facility noted. Advances in computer graphics may, in future, mean that such operator training will be virtual reality rather than analog reality based; however the need for such rigs to commission techniques and equipment and to establish performance and reliability is likely to continue. (UK)

  10. International benchmark on the natural convection test in Phenix reactor

    International Nuclear Information System (INIS)

    Tenchine, D.; Pialla, D.; Fanning, T.H.; Thomas, J.W.; Chellapandi, P.; Shvetsov, Y.; Maas, L.; Jeong, H.-Y.; Mikityuk, K.; Chenu, A.; Mochizuki, H.; Monti, S.

    2013-01-01

    Highlights: ► Phenix main characteristics, instrumentation and natural convection test are described. ► “Blind” calculations and post-test calculations from all the participants to the benchmark are compared to reactor data. ► Lessons learned from the natural convection test and the associated calculations are discussed. -- Abstract: The French Phenix sodium cooled fast reactor (SFR) started operation in 1973 and was stopped in 2009. Before the reactor was definitively shutdown, several final tests were planned and performed, including a natural convection test in the primary circuit. During this natural convection test, the heat rejection provided by the steam generators was disabled, followed several minutes later by reactor scram and coast-down of the primary pumps. The International Atomic Energy Agency (IAEA) launched a Coordinated Research Project (CRP) named “control rod withdrawal and sodium natural circulation tests performed during the Phenix end-of-life experiments”. The overall purpose of the CRP was to improve the Member States’ analytical capabilities in the field of SFR safety. An international benchmark on the natural convection test was organized with “blind” calculations in a first step, then “post-test” calculations and sensitivity studies compared with reactor measurements. Eight organizations from seven Member States took part in the benchmark: ANL (USA), CEA (France), IGCAR (India), IPPE (Russian Federation), IRSN (France), KAERI (Korea), PSI (Switzerland) and University of Fukui (Japan). Each organization performed computations and contributed to the analysis and global recommendations. This paper summarizes the findings of the CRP benchmark exercise associated with the Phenix natural convection test, including blind calculations, post-test calculations and comparisons with measured data. General comments and recommendations are pointed out to improve future simulations of natural convection in SFRs

  11. Scaling of silent electrical discharge reactors for hazardous organics destruction

    International Nuclear Information System (INIS)

    Coogan, J.J.; Rosocha, L.A.; Brower, M.J.; Kang, M.; Schmidt, C.A.

    1993-01-01

    Silent electrical discharges are used to produce highly reactive free radicals that destroy hazardous compounds entrained in gaseous effluents at ambient gas temperatures and pressures. We have carried out destruction experiments at Los Alamos on a range of volatile organic compounds (VOCs), including trichloroethylene (TCE), carbon tetrachloride, perchloroethylene (PCE), and chlorofluorocarbons (CFCs). We have measured a ''nine-factor'', the amount of energy required to reduce the VOC concentration by a factor of ten. For practical reactor power densities, the ''nine-factor'' can be used to predict the destruction an removal efficiency (DRE) in terms of gas flow rate and the number of reactor modules. This report proposes a modular, stackable architecture for scaling up the reactor throughput

  12. 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.

  13. General directions and recently test modelling results of lithium capillary-pore systems as plasma facing components for tokamak-reactor

    International Nuclear Information System (INIS)

    Evtikhin, V.A.; Lyublinski, I.E.; Vertkov, A.V.; Azizov, E.A.; Mirnov, S.V.; Lazaret, V.B.; Safronov, V.M.

    2003-01-01

    Full text: At present the most promising principal solution of the divertor problem appears to be the use of liquid metals and primarily of lithium Capillary-Pore Systems (CPS) as of plasma facing material. A solid CPS filled with liquid lithium will have high resistance to surface and volume damage because of neutron radiation effects, melting, splashing and thermal stress induced cracking in steady state and during plasma transitions (disruptions, ELMs, VDEs, runaways) to provide the normal operation of divertor target plates and first wall protection elements. These materials would not be the sources of impurities inducing the raise of Z eff and they will not be collected as dust in the divertor area and in ducts. The key directions of experimental investigation of lithium CPS behaviour in first wall and divertor operation simulating conditions are considered. Experiments with lithium CPS in plasma disruption simulation conditions on the hydrogen plasma accelerator MK-200UG (∼10-15 MJ/m 2 , ∼50 μs) have been performed. Shielding lithium plasma layer formation and high stability of these systems have been shown. The new lithium limiter with a thermal regulation system tests on up graded T-11M tokamak (plasma current up to 100 kA, pulse length ∼0.3 s) have been performed. Sorption and desorption of plasma-forming gas, lithium emission into discharge, lithium erosion, limiter deposited power are investigated in this tests

  14. Unusual occurrences in fast breeder test reactor

    International Nuclear Information System (INIS)

    Kapoor, R.P.; Srinivasan, G.; Ellappan, T.R.; Ramalingam, P.V.; Vasudevan, A.T.; Iyer, M.A.K.; Lee, S.M.; Bhoje, S.B.

    2000-01-01

    Fast Breeder Test Reactor (FBTR) is a 40 MWt/13.2 MWe sodium cooled mixed carbide fuelled reactor. Its main aim is to generate experience in the design, construction and operation of fast reactors including sodium systems and to serve as an irradiation facility for the development of fuel and structural materials for future fast reactors. It achieved first criticality in Oct 85 with Mark I core (70% PuC - 30% UC). Steam generator was put in service in Jan 93 and power was raised to 10.5 MWt in Dec 93. Turbine generator was synchronised to the grid in Jul 97. The indigenously developed mixed carbide fuel has achieved a burnup of 44,000 MW-d/t max at a linear heat rating of 320 W/cm max without any fuel clad failure. The commissioning and operation of sodium systems and components have been smooth and performance of major components, viz., sodium pumps, intermediate heat exchangers and once through sodium heated steam generators (SG) have been excellent. There have been three minor incidents of Na/NaK leaks during the past 14 years, which are described in the paper. There have been no incident of a tube leak in SG. However, three incidents of water leaks from water / steam headers have been detailed. The plant has encountered some unusual occurrences, which were critically analysed and remedial measures, in terms of system and procedural modifications, incorporated to prevent recurrence. This paper describes unusual occurrences of fuel handling incident of May 1987, main boiler feed pump seizure in Apr 1992, reactivity transients in Nov 1994 and Apr 1995, and malfunctioning of the core cover plate mechanism in Jul 1995. These incidents have resulted in long plant shutdowns. During the course of investigation, various theoretical and experimental studies were carried out for better understanding of the phenomena and several inspection techniques and tools were developed resulting in enriching the technology of sodium cooled reactors. FBTR has 36 neutronic and process

  15. 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

  16. 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.

  17. Instrumentation to Enhance Advanced Test Reactor Irradiations

    International Nuclear Information System (INIS)

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

    2009-01-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support 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.

  18. Reactor coolant pump type RUV for Westinghouse Electric Company LLC reactor AP1000 TM

    International Nuclear Information System (INIS)

    Baumgarten, S.; Brecht, B.; Bruhns, U.; Fehring, P.

    2010-01-01

    The RUV is a reactor coolant pump, specially designed for the Westinghouse Electric Company LLC AP1000 TM reactor. It is a hermetically sealed, wet winding motor pump. The RUV is a very compact, vertical pump/motor unit, designed to fit into the compartment next to the reactor pressure vessel. Each of the two steam generators has two pump casings welded to the channel head by the suction nozzle. The pump/motor unit consists of a pump part, where a semi-axial impeller/diffuser combination is mounted in a one-piece pump casing. Computational Fluid Dynamics methods combined with various hydraulic tests in a 1:2 scale hydraulic test assure full compliance with the specific customer requirements. A short and rigid shaft, supported by a radial bearing, connects the impeller with the high inertia flywheel. This flywheel consists of a one-piece forged stainless steel cylinder, with an option for several smaller heavy metal cylinders inside. The flywheel is located inside the thermal barrier, which forms part of the pressure boundary. A specific arrangement of cooling water circuits guarantees a homogeneous temperature distribution in and around the flywheel, minimizes the friction losses of the flywheel and protects the motor from hot coolant. The driving torque is transmitted by the motor shaft, which itself is supported by two radial bearings. A three-phase, high-voltage squirrel-cage induction motor generates the driving torque. Due to the wet winding concept it is possible to achieve positive effects regarding motor lifetime. The cooling water is forced through the stator windings and the gap between rotor and stator by an auxiliary impeller. Furthermore, this wet winding motor concept has higher efficiency as compared to a canned motor since there are no eddy current losses. As part of the design process and in addition to the hydraulic scale model, a complete half scale model pump was built. It was used to verify the calculations performed like coast

  19. Advanced test reactor testing experience-past, present and future

    International Nuclear Information System (INIS)

    Marshall, Frances M.

    2006-01-01

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is one of the world's premier test reactors for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The physical configuration of the ATR, a 4-leaf clover shape, allows the reactor to be operated at different power levels in the corner 'lobes' to allow for different testing conditions for multiple simultaneous experiments. The combination of high flux (maximum thermal neutron fluxes of 1E15 neutrons per square centimeter per second and maximum fast [E>1.0 MeV] neutron fluxes of 5E14 neutrons per square centimeter per second) and large test volumes (up to 122 cm long and 12.7 cm diameter) provide unique testing opportunities. The current experiments in the ATR are for a variety of test sponsors - US government, foreign governments, private researchers, and commercial companies needing neutron irradiation services. There are three basic types of test configurations in the ATR. The simplest configuration is the sealed static capsule, which places the capsule in direct contact with the primary coolant. The next level of experiment complexity is an instrumented lead experiment, which allows for active control of experiment conditions during the irradiation. The most complex experiment is the pressurized water loop, in which the test sample can be subjected to the exact environment of a pressurized water reactor. For future research, some ATR modifications and enhancements are currently planned. This paper provides more details on some of the ATR capabilities, key design features, experiments, and future plans

  20. Electrically heated ex-reactor pellet-cladding interaction (PCI) simulations utilizing irradiated Zircaloy cladding

    International Nuclear Information System (INIS)

    Barner, J.O.; Fitzsimmons, D.E.

    1985-02-01

    In a program sponsored by the Fuel Systems Research Branch of the US Nuclear Regulatory Commission, a series of six electrically heated fuel rod simulation tests were conducted at Pacific Northwest Laboratory. The primary objective of these tests was to determine the susceptibility of irradiated pressurized-water reactor (PWR) Zircaloy-4 cladding to failures caused by pellet-cladding mechanical interaction (PCMI). A secondary objective was to acquire kinetic data (e.g., ridge growth or relaxation rates) that might be helpful in the interpretation of in-reactor performance results and/or the modeling of PCMI. No cladding failures attributable to PCMI occurred during the six tests. This report describes the testing methods, testing apparatus, fuel rod diametral strain-measuring device, and test matrix. Test results are presented and discussed

  1. 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 ...

  2. Decommissioning of the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Perry, E.; Chrzanowski, J.; Gentile, C.; Parsells, R.; Rule, K.; Strykowsky, R.; Viola, M.

    2003-01-01

    The Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory was operated from 1982 until 1997. The last several years included operations with mixtures of deuterium and tritium. In September 2002, the three year Decontamination and Decommissioning (D and D) Project for TFTR was successfully completed. The need to deal with tritium contamination as well as activated materials led to the adaptation of many techniques from the maintenance work during TFTR operations to the D and D effort. In addition, techniques from the decommissioning of fission reactors were adapted to the D and D of TFTR and several new technologies, most notably the development of a diamond wire cutting process for complex metal structures, were developed. These techniques, along with a project management system that closely linked the field crews to the engineering staff who developed the techniques and procedures via a Work Control Center, resulted in a project that was completed safely, on time, and well below budget

  3. Developing the MAPLE materials test reactor concept

    International Nuclear Information System (INIS)

    Lee, A.G.; Lidstone, R.F.; Donnelly, J.V.

    1992-05-01

    MAPLE-MTR is a new multipurpose research facility being planned by AECL Research as a possible replacement for the 35-year-old NRU reactor. In developing the MAPLE-MTR concept, AECL is starting from the recent design and licensing experience with the MAPLE-X10 reactor. By starting from technology developed to support the MAPLE-X10 design and adapting it to produce a concept that satisfies the requirements of fuel channel materials testing and fuel irradiation programs, AECL expects to minimize the need for major advances in nuclear technology (e.g., fuel, heat transfer). Formulation of the MAPLE-MTR concept is at an early stage. This report describes the irradiation requirements of the research areas, how these needs are translated into design criteria for the project and elements of the preliminary design concept

  4. Fabrication of Fast Reactor Fuel Pins for Test Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    Karsten, G. [Institute for Applied Reactor Physics, Kernforschungszentrum Karlsruhe, Karlsruhe, Federal Republic of Germany (Germany); Dippel, T. [Institute for Radiochemistry, Kernforschungszentrum Karlsruhe, Karlsruhe, Federal Republic of Germany (Germany); Laue, H. J. [Institute for Applied Reactor Physics, Kernforschungszentrum Karlsruhe, Karlsruhe, Federal Republic of Germany (Germany)

    1967-09-15

    An extended irradiation programme is being carried out for the fuel element development of the Karlsruhe fast breeder project. A very important task within the programme is the testing of plutonium-containing fuel pins in a fast-reactor environment. This paper deals with fabrication of such pins by our laboratories at Karlsruhe. For the fast reactor test positions at present envisaged a fuel with 15% plutonium and the uranium fully enriched is appropriate. Hie mixed oxide is both pelletized and vibro-compacted with smeared densities between 80 and 88% theoretical. The pin design is, for example, such that there are two gas plena at the top and bottom, and one blanket above the fuel with the fuel zone fitting to the test reactor core length. The specifications both for fuel and cladding have been adapted to the special purpose of a fast-breeder reactor - the outer dimensions, the choice of cladding and fuel types, the data used and the kind of tests outline the targets of the development. The fuel fabrication is described in detail, and also the powder line used for vibro-compaction. The source materials for the fuel are oxalate PuO{sub 2} and UO{sub 2} from the UF{sub 6} process. The special problems of mechanical mixing and of plutonium homogeneity have been studied. The development of the sintering technique and grain characteristics for vibratory compactive fuel had to overcome serious problems in order to reach 82-83% theoretical. The performance of the pin fabrication needed a major effort in welding, manufacturing of fits and decontamination of the pin surfaces. This was a stimulation for the development of some very subtle control techniques, for example taking clear X-ray photographs and the tube testing. In general the selection of tests was a special task of the production routine. In conclusion the fabrication of the pins resulted in valuable experiences for the further development of fast reactor fuel elements. (author)

  5. Nuclear reactor power for an electrically powered orbital transfer vehicle

    Science.gov (United States)

    Jaffe, L.; Beatty, R.; Bhandari, P.; Chow, E.; Deininger, W.; Ewell, R.; Fujita, T.; Grossman, M.; Kia, T.; Nesmith, B.

    1987-01-01

    To help determine the systems requirements for a 300-kWe space nuclear reactor power system, a mission and spacecraft have been examined which utilize electric propulsion and this nuclear reactor power for multiple transfers of cargo between low earth orbit (LEO) and geosynchronous earth orbit (GEO). A propulsion system employing ion thrusters and xenon propellant was selected. Propellant and thrusters are replaced after each sortie to GEO. The mass of the Orbital Transfer Vehicle (OTV), empty and dry, is 11,000 kg; nominal propellant load is 5000 kg. The OTV operates between a circular orbit at 925 km altitude, 28.5 deg inclination, and GEO. Cargo is brought to the OTV by Shuttle and an Orbital Maneuvering Vehicle (OMV); the OTV then takes it to GEO. The OTV can also bring cargo back from GEO, for transfer by OMV to the Shuttle. OTV propellant is resupplied and the ion thrusters are replaced by the OMV before each trip to GEO. At the end of mission life, the OTV's electric propulsion is used to place it in a heliocentric orbit so that the reactor will not return to earth. The nominal cargo capability to GEO is 6000 kg with a transit time of 120 days; 1350 kg can be transferred in 90 days, and 14,300 kg in 240 days. These capabilities can be considerably increased by using separate Shuttle launches to bring up propellant and cargo, or by changing to mercury propellant.

  6. Nuclear reactor power for an electrically powered orbital transfer vehicle

    International Nuclear Information System (INIS)

    Jaffe, L.; Beatty, R.; Bhandari, P.

    1987-01-01

    To help determine the systems requirements for a 300-kWe space nuclear reactor power system, a mission and spacecraft have been examined which utilize electric propulsion and this nuclear reactor power for multiple transfers of cargo between low Earth orbit (LEO) and geosynchronous Earth orbit (GEO). A propulsion system employing ion thrusters and xenon propellant was selected. Propellant and thrusters are replaced after each sortie to GEO. The mass of the Orbital Transfer Vehicle (OTV), empty and dry, is 11,000 kg; nominal propellant load is 5000 kg. The OTV operates between a circular orbit at 925 km altitude, 28.5 deg inclination, and GEO. Cargo is brought to the OTV by Shuttle and an Orbital Maneuvering Vehicle (OMV); the OTV then takes it to GEO. The OTV can also bring cargo back from GEO, for transfer by OMV to the Shuttle. OTV propellant is resupplied and the ion thrusters are replaced by the OMV before each trip to GEO. At the end of mission life, the OTV's electric propulsion is used to place it in a heliocentric orbit so that the reactor will not return to Earth. The nominal cargo capability to GEO is 6000 kg with a transit time of 120 days; 1350 kg can be transferred in 90 days, and 14,300 kg in 240 days. These capabilities can be considerably increased by using separate Shuttle launches to bring up propellant and cargo, or by changing to mercury propellant

  7. Proceedings of the international meeting on development, fabrication and application of reduced enrichment fuels for research and test reactors

    International Nuclear Information System (INIS)

    1983-08-01

    Separate abstracts were prepared for each of the papers presented in the following areas: (1) Reduced Enrichment Fuels for Research and Test Reactors (RERTR) Program Status; (2) Fuel Development; (3) Fuel Demonstrations; (4) General Topics; and (5) Specific Reactor Applications

  8. 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.

  9. Electric Energy Consumption of Multi Purpose Reactor GA. Siwabessy During Reactor Operation

    International Nuclear Information System (INIS)

    Koes Indrakoesoema

    2012-01-01

    Electrical power supply of Reactor Center Multi Purpose obtained from PT PLN to 3030 kVA power contracts. Distribution to existing loads in PRSG divided into 3 (three) lines, each of which is supplied through a transformer BHT01, BHT02 and BHT03, each transformer have capacity of 1600 kVA. During reactor operation, only 2 lines that serve loads, each line serve 2 primary pump motor and 2 secondary pump motor. Electrical power for 24 hours for measurement BHT01, the average is 288 kW, for BHT02 is 641 kW and BHT03 is 466 kW. The energy absorbed by each transformer for 24 hours of measurement, for BHT01 is 6.44 MWh, BHT02 absorb 14.8 MWh and BHT03 absorb 10.9 MWh. (author)

  10. Operation and maintenance experience at the General Atomic Company's TRIGA reactor facility at San Diego, California

    International Nuclear Information System (INIS)

    Whittemore, W.L.; Stout, W.A.; Shoptaugh, J.R.; Chesworth, R.H.

    1982-01-01

    Since the startup of the original 250 kW TRIGA Mark I reactor in 1958, General Atomic Company has accumulated nearly 24 years of operation and maintenance experience with this type of reactor. In addition to the nearly 24 years of experience gained on the Mark I, GA has operated the 1.5 MW Advanced Prototype Test Reactor (Mark F) for 22 years and operated a 2 MW below-ground TRIGA Mark III for five years. Information obtained from normal and abnormal operation are presented. (author)

  11. Thermal Hydraulic Tests for Reactor Core Safety

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-06-15

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

  12. The ICRH tokamak fusion test reactor

    International Nuclear Information System (INIS)

    Perkins, F.W.

    1976-01-01

    A Tokamak Fusion Test Reactor where the ion are maintained at Tsub(i) approximately 20keV>Tsub(e) approximately 7keV by ion-cyclotron resonance heating is shown to produce an energy amplification of Q>2 provided the principal ion energy loss channel is via collisional transfer to the electrons. Such a reactor produces 19MW of fusion power to the electrons. Such a reactor produces 19MW of fusion power and requires a 50MHz radio-frequency generator capable of 50MW peak power; it is otherwise compatible with the conceptual design for the Princeton TFTR. The required n tausub(E) values for electrons and ions are respectively ntausub(Ee)>1.5.10 13 cm -3 -sec and ntausub(Ei)>4.10 13 cm -3 -sec. The principal areas where research is needed to establish this concept are: tokamak transport calculations, ICRH physics, trapped-particle instability energy losses, tokamak equilibria with high values of βsub(theta), and, of course, impurities

  13. Safety Assurance for Irradiating Experiments in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    T. A. Tomberlin; S. B. Grover

    2004-11-01

    The Advanced Test Reactor (ATR), located at the Idaho National Engineering and Environmental Laboratory (INEEL), was specifically designed to provide a high neutron flux test environment for conducting a variety of experiments. This paper addresses the safety assurance process for two general types of experiments conducted in the ATR facility and how the safety analyses for experiments are related to the ATR safety basis. One type of experiment is more routine and generally represents greater risks; therefore, this type of experiment is addressed in more detail in the ATR safety basis. This allows the individual safety analysis for this type of experiment to be more standardized. The second type of experiment is defined in more general terms in the ATR safety basis and is permitted under more general controls. Therefore, the individual safety analysis for the second type of experiment tends to be more unique and is tailored to each experiment.

  14. Safety Assurance for Irradiating Experiments in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    T. A. Tomberlin; S. B. Grover

    2004-01-01

    The Advanced Test Reactor (ATR), located at the Idaho National Engineering and Environmental Laboratory (INEEL), was specifically designed to provide a high neutron flux test environment for conducting a variety of experiments. This paper addresses the safety assurance process for two general types of experiments conducted in the ATR facility and how the safety analyses for experiments are related to the ATR safety basis. One type of experiment is more routine and generally represents greater risks; therefore, this type of experiment is addressed in more detail in the ATR safety basis. This allows the individual safety analysis for this type of experiment to be more standardized. The second type of experiment is defined in more general terms in the ATR safety basis and is permitted under more general controls. Therefore, the individual safety analysis for the second type of experiment tends to be more unique and is tailored to each experiment

  15. Reactor power cutback system test experience at YGN 4

    International Nuclear Information System (INIS)

    Chi, Sung Goo; Kim, Se Chang; Seo, Jong Tae; Eom, Young Meen; Wook, Jeong Dae; Choi, Young Boo

    1995-01-01

    YGN 3 and 4 are the nuclear power plants having System 80 characteristics with a rated thermal output of 2815 MWth and a nominal net electrical output of 1040 MWe. YGN 3 achieved commercial operation on March 31, 1995 and YGN 4 completed Power Ascension Test (PAT) at 20%, 50%, 80% and 100% power by September 23, 1995. YGN 3 and 4 design incorporates the Reactor POwer Cutback System (RPCS) which reduces plant trips caused by Loss of Load (LOL)/ Turbine Trip and Loss of One Main Feedwater Pump (LOMFWP). The key design objective of the RPCS is to improve overall plant availability and performance, while minimizing challenges to the plant safety systems. The RPCS is designed to rapidly reduce reactor power by dropping preselected Control Element Assemblies (CEAs) while other NSSS control systems maintain process parameters within acceptable ranges. Extensive RPCS related tests performed during the initial startup of YGN 4 demonstrated that the RPCS can maintain the reactor on-line without opening primary or secondary safety valves and without actuating the Engineered Safety Features Actuation System (ESFAS). It is expected that use of the RPCS at YGN will increase the overall availability of the units and reduce the number of challenges to plant safety systems

  16. WWER type reactor primary loop imitation on large test loop facility in MARIA reactor

    International Nuclear Information System (INIS)

    Moldysh, A.; Strupchevski, A.; Kmetek, Eh.; Spasskov, V.P.; Shumskij, A.M.

    1982-01-01

    At present in Poland in cooperation with USSR a nuclear water loop test facility (WL) in 'MARIA' reactor in Sverke is under construction. The program objective is to investigate processes occuring in WWER reactor under emergency conditions, first of all after the break of the mainprimary loop circulation pipe-line. WL with the power of about 600 kW consists of three major parts: 1) an active loop, imitating the undamaged loops of the WWER reactor; 2) a passive loop assignedfor modelling the broken loop of the WWER reactor; 3) the emergency core cooling system imitating the corresponding full-scale system. The fuel rod bundle consists of 18 1 m long rods. They were fabricated according to the standard WWER fuel technology. In the report some general principles of WWERbehaviour imitation under emergency conditions are given. They are based on the operation experience obtained from 'SEMISCALE' and 'LOFT' test facilities in the USA. A description of separate modelling factors and criteria effects on the development of 'LOCA'-type accident is presented (the break cross-section to the primary loop volume ratio, the pressure differential between inlet and outlet reactor chambers, the pressure drop rate in the loop, the coolant flow rate throuh the core etc.). As an example a comparison of calculated flow rate variations for the WWER-1000 reactor and the model during the loss-of-coolant accident with the main pipe-line break at the core inlet is given. Calculations have been carried out with the use of TECH'-M code [ru

  17. General conditions for electric power supply

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    If it is uncertain whether future power bills will be paid fully, it is admissible to take an action claiming a declaration which states that the electricity rate payment boycotter has no right to non-payment nor a right to withhold payment towards the electricity supply utility, and that the electricity supply utility has the right to stop energy supply because of reduced electricity rate payments effected and/or announced, and to denounce the contract without observing any term of notice. If the electricity buyer reduces a power bill to be paid without any legal grounds, the electricity supply utility has the right to stop power supplies and to denounce the power supply contract without observing any term of notice. The freedom of thought and the freedom of opinion must not be expressed by reducing power bills to be paid. Basic rights discontinue to be effective as soon as a contract or law is broken. A weighing of protected interests is not effected if the exercise of a basic law is unlawful. (orig./HP) [de

  18. Safety analysis calculations for research and test reactors

    Energy Technology Data Exchange (ETDEWEB)

    Chen, S Y; MacDonald, R; MacFarlane, D [Argonne National Laboratory, Argonne, IL (United States)

    1983-08-01

    The goal of the RERTR (Reduced Enrichment in Research and Test Reactor) Program at ANL is to provide technical means for conversion of research and test reactors from HEU (High-Enrichment Uranium) to LEU (Low-Enrichment Uranium) fuels. In exploring the feasibility of conversion, safety considerations are a prime concern; therefore, safety analyses must be performed for reactors undergoing the conversion. This requires thorough knowledge of the important safety parameters for different types of reactors for both HEU and LEU fuel. Appropriate computer codes are needed to predict transient reactor behavior under postulated accident conditions. In this discussion, safety issues for the two general types of reactors i.e., the plate-type (MTR-type) reactor and the rod-type (TRIGA-type) reactor, resulting from the changes associated with LEU vs. HEU fuels, are explored. The plate-type fuels are typically uranium aluminide (UAl{sub x}) compounds dispersed in aluminum and clad with aluminum. Moderation is provided by the water coolant. Self shut-down reactivity coefficients with EU fuel are entirely a result of coolant heating, whereas with LEU fuel there is an additional shut down contribution provided by the direct heating of the fuel due to the Doppler coefficient. In contrast, the rod-type (TRIGA) fuels are mixtures of zirconium hydride, uranium, and erbium. This fuel mixture is formed into rods ( {approx} 1 cm diameter) and clad with stainless steel or Incoloy. In the TRIGA fuel the self-shutdown reactivity is more complex, depending on heating of the fuel rather than the coolant. The two most important mechanisms in providing this feedback are: spectral hardening due to neutron interaction with the ZrH moderator as it is heated and Doppler broadening of resonances in erbium and U-238. Since these phenomena result directly from heating of the fuel, and do not depend on heat transfer to the moderator/coolant, the coefficients are prompt acting. Results of transient

  19. The Concept of the Use of the Marine Reactor Plant in Small Electric Grids

    International Nuclear Information System (INIS)

    Khlopkin, N.; Makarov, V.; Pologikh, B.

    2002-01-01

    In report some aspects of the using marine nuclear reactor are considered for provision of need small non-interconnected power systems, as well as separate settlements and the mining enterprises disposed in regions with a undeveloped infrastructure. Recently for these purposes it is offered to use the nuclear small modular power plants. The required plant power for small electric grids lies within from 1 to several tens of MWe. Module can be collected and tested on machine-building plant, and then delivered in ready type to the working place on some transport, for instance, a barge. Through determined time it's possible to transport a module to the repair shop and also to the point of storage after the end of operation. Marine nuclear reactors on their powers, compactness, mass and size are ideal prototypes for creation of such modules. For instance, building at present floating power unit, intended for functioning in region of the Russian North, based on using reactor plants of nuclear icebreakers. Reliability and safety of the ship reactor are confirmed by their trouble-free operation during approximately 180 reactors-years. Unlike big stationary nuclear plant, working in base mode, power unit with marine reactor wholly capable to work in mode of the loading following. In contrast with reactor of nuclear icebreaker, advisable to increase the core lifetime and to reduce the enrichment of the uranium. This requires more uranium capacity fuel compositions and design of the core. In particular, possible transition from traditional for ship reactor of the channel core to cassette design. Other directions of evolution of the ship reactors, not touching the basic constructive decisions verified by practice, but promoting development of properties of self-security of plant are possible. Among such directions is reduction volumetric power density of a core. (author)

  20. Assessment of residual life of fast breeder test reactor

    International Nuclear Information System (INIS)

    Srinivasan, G.

    2016-01-01

    The Fast Breeder Test Reactor (FBTR) is a loop type sodium cooled fast reactor and has been in operation since 1985. As a part of regulatory requirement for relicensing, residual life assessment had to be carried out. The systems are made of SS 316, and designed for creep and fatigue. The design life for creep is 100,000 h at 550°C. The design fatigue cycle for operation from shutdown to full power varies from component to component. In general, most of the components are designed for 2000 cycles. The reactor has operated mostly below the design temperatures. It is seen that enough creep-fatigue life is available for the non-replaceable, permanent components. The residual life was found to be governed by the residual ductility of the Grid Plate supporting the core after neutron irradiation. Fast flux measurements were carried out at the grid plate location. Samples were irradiated and tensile tested. Results indicate the allowable dpa for a 10% residual ductility criterion as 4.37. This gave a residual life of ~ 6 Effective Full Power Years for the reactor as of Feb 2012. Measures to reduce the neutron dose on the grid plate are being taken. (author)

  1. Correlation of electrical reactor cable failure with materials degradation

    Energy Technology Data Exchange (ETDEWEB)

    Stuetzer, O.M.

    1986-03-01

    Complete circuit failure (shortout) of electrical cables typically used in nuclear power plant containments is investigated. Failure modes are correlated with the mechanical deterioration of the elastomeric cable materials. It is found that for normal reactor operation, electrical cables are reliable and safe over very long periods. During high temperature excursions, however, cables pulled across corners under high stress may short out due to conductor creep. Severe cracking will occur in short times during high temperatures (>150/sup 0/C) and in times of the order of years at elevated temperatures (100/sup 0/C to 140/sup 0/C). A theoretical treatment of stress distribution responsible for creep and for cracking by J.E. Reaugh of Science Applications, Inc. is contained in the Appendix. 29 refs., 32 figs.

  2. Correlation of electrical reactor cable failure with materials degradation

    International Nuclear Information System (INIS)

    Stuetzer, O.M.

    1986-03-01

    Complete circuit failure (shortout) of electrical cables typically used in nuclear power plant containments is investigated. Failure modes are correlated with the mechanical deterioration of the elastomeric cable materials. It is found that for normal reactor operation, electrical cables are reliable and safe over very long periods. During high temperature excursions, however, cables pulled across corners under high stress may short out due to conductor creep. Severe cracking will occur in short times during high temperatures (>150 0 C) and in times of the order of years at elevated temperatures (100 0 C to 140 0 C). A theoretical treatment of stress distribution responsible for creep and for cracking by J.E. Reaugh of Science Applications, Inc. is contained in the Appendix. 29 refs., 32 figs

  3. Integrity assessment of TAPS reactor pressure vessel at extended EOL using surveillance test results

    International Nuclear Information System (INIS)

    Chatterjee, S.; Shah, Priti Kotak

    2008-05-01

    Integrity assessment of pressure vessels of nuclear reactors (RPV) primarily concentrates on the prevention of brittle failure and conditions are defined under which brittle failure can be excluded. Accordingly, two approaches based on Transition Temperature Concept and Fracture Mechanics Concept were adopted using the impact test results of three credible surveillance data sets obtained from the surveillance specimens of Tarapur Atomic Power Station. RT NDT data towards end of life (EOL) were estimated from the impact test results in accordance with the procedures of USNRC Regulatory Guide 1.99, Rev. 2 and were used as primary input for assessment of the vessel integrity. SA302B (nickel modified) steel cladded with stainless steel is used as the pressure vessel material for the two 210 MWe boiling water reactors of the Tarapur Atomic Power Station (TAPS). The reactors were commissioned during the year 1969. The chemical compositions of SA302B (modified) steel used in fabricating the vessel and the specified tensile property and the Charpy impact property requirements of the steel broadly meet ASME specified requirements. Therefore, the pressure temperature limit curves prescribed by General Electric (G.E.) were compared with those as obtained using procedures of ASME Section XII, Appendix G. The tensile and the Charpy impact properties at 60 EFPY of vessel operation as derived from the surveillance specimens even fulfilled the specified requirements for the virgin material of ASME. Integrity assessment carried out using the two approaches indicated the safety of the vessel for continued operation up to 60 EFPY. (author)

  4. SRS reactor stack plume marking tests

    International Nuclear Information System (INIS)

    Petry, S.F.

    1992-03-01

    Tests performed in 105-K in 1987 and 1988 demonstrated that the stack plume can successfully be made visible (i.e., marked) by introducing smoke into the stack breech. The ultimate objective of these tests is to provide a means during an emergency evacuation so that an evacuee can readily identify the stack plume and evacuate in the opposite direction, thus minimizing the potential of severe radiation exposure. The EPA has also requested DOE to arrange for more tests to settle a technical question involving the correct calculation of stack downwash. New test canisters were received in 1988 designed to produce more smoke per unit time; however, these canisters have not been evaluated, because normal ventilation conditions have not been reestablished in K Area. Meanwhile, both the authorization and procedure to conduct the tests have expired. The tests can be performed during normal reactor operation. It is recommended that appropriate authorization and procedure approval be obtained to resume testing after K Area restart

  5. Tests of vacuum interrupters for the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Warren, R.; Parsons, M.; Honig, E.; Lindsay, J.

    1979-04-01

    The Tokamak Fusion Test Reactor (TFTR) project at Princeton University requires the insertion of a resistor in an excited ohmic-heating coil circuit to produce a plasma initiation pulse (PIP). It is expected that the maximum duty for the switching system will be an interruption of 24 kA with an associated recovery voltage of 25 kV. Vacuum interrupters were selected as the most economical means to satisfy these requirements. However, it was felt that some testing of available systems should be performed to determine their reliability under these conditions. Two interrupter systems were tested for over 1000 interruptions each at 24 kA and 25 kV. One system employed special Westinghouse type WL-33552 interrupters in a circuit designed by LASL. This circuit used a commercially available actuator and a minimum size counterpulse bank and saturable reactor. The other used Toshiba type VGB2-D20 interrupters actuated by a Toshiba mechanism in a Toshiba circuit using a larger counterpulse bank and saturable reactor

  6. Improving the fidelity of electrically heated nuclear systems testing using simulated neutronic feedback

    International Nuclear Information System (INIS)

    Bragg-Sitton, Shannon M.; Godfroy, Thomas J.; Webster, Kenny

    2010-01-01

    Nonnuclear test platforms and methodologies can be employed to reduce the overall cost, risk and complexity of testing nuclear systems while allowing one to evaluate the operation of an integrated nuclear system within a reasonable timeframe, providing valuable input to the overall system design. In a nonnuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Standard electric test techniques allow one to fully assess thermal, heat transfer, and stress related attributes of a given system, but these approaches fail 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 testing with nuclear fuel elements installed. 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. This paper summarizes the results of initial system dynamic response testing for two electrically heated reactor concepts: a heat pipe-cooled reactor simulator with integrated heat exchanger and a gas-cooled reactor simulator with integrated Brayton power conversion system. Initial applications apply a simplified reactor kinetics model with either a single or an averaged measured state point. Preliminary results demonstrate the applicability of the dynamic test methodology to any reactor type, elucidating the variation in system response characteristics in different reactor concepts. These results suggest a need to further enhance the dynamic test approach by incorporating a more accurate model of the reactor dynamics and improved hardware instrumentation for better state estimation in application of the

  7. 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)

  8. Review of inservice inspection and nondestructive examination practices at DOE Category A test and research reactors

    International Nuclear Information System (INIS)

    Anderson, M.T.; Aldrich, D.A.

    1990-09-01

    In-service inspection (ISI) programs are used at commercial nuclear power plants for monitoring the pressure boundary integrity of various systems and components to ensure their continued safe operation. The Department of Energy (DOE) operates several test and research reactors. This report represents an evaluation of the ISI and nondestructive examination (NDE) practices at five DOE Category A (> 20 MW thermal) reactors as compared, where applicable, to the current ISI activities of commercial nuclear power facilities. The purpose of an inservice inspection (ISI) program is to establish regular surveillance of safety-related components to ensure their safe and reliable operation. The integrity of materials comprising these components is generally monitored by means of periodic nondestructive examinations (NDE), which, if appropriately performed, provide methods for identifying degradation that could render components unable to perform their intended safety functions. The reactors evaluated during this review were the Experimental Breeder Reactor 2 and the Fast Flux Test Facility (liquid-metal cooled plants), the Advanced Test Reactor and the High Flux Isotopes Reactor (light-water cooled reactors), and the High Flux Beam Reactor (a heavy-water cooled facility). Although these facilities are extremely diverse in design and operation, they all have less stored energy, smaller inventories of radionuclides, and generally, more remote locations than commercial reactors. However, all DOE test and research facilities contain components similar to those of commercial reactors for which continued integrity is important to maintain plant safety. 10 refs., 6 tabs

  9. Decommissioning the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Spampinato, P.T.; Walton, G.R.

    1993-01-01

    The Tokamak Fusion Test Reactor (TFTR) at Princeton Plasma Physics Laboratory (PPPL) will complete its experimental lifetime with a series of deuterium-tritium pulses in 1994. As a result, the machine structures will become radioactive, and vacuum components will also be contaminated with tritium. Dose rate levels will range from less than 1 mr/h for external structures to hundreds of mr/h for the vacuum vessel. Hence, decommissioning operations will range from hands on activities to the use of remotely operated equipment. After 21 months of cool down, decontamination and decommissioning (D and D) operations will commence and continue for approximately 15 months. The primary objective is to render the test cell complex re-usable for the next machine, the Tokamak Physics Experiment (TPX). This paper presents an overview of decommissioning TFTR and discusses the D and D objectives

  10. Portland General Electric Company report on the operating and startup experience with control and instrumentation and electrical systems at the Trojan Nuclear Plant

    International Nuclear Information System (INIS)

    Zimmerman, G.A.

    1977-01-01

    The Trojan Nuclear Plant is an 1178 MWe nuclear plant located on the Columbia River 40 miles northwest of Portland, Oregon. The Nuclear Stream Supply System vendor is Westinghouse with a General Electric turbine generator. The reactor is rated and licensed for 3423 MWt (1178 MWe) and the turbine generator is designed for 3570 MWt(1219 MWe). The startup phase testing of Trojan commenced on November 21, 1975, upon receipt of our NRC Operating License. The startup testing program was completed on May 22, 1976, following 100 hours of full-power operation, at which time a scheduled summer maintenance outage began. Some of the highlights and milestones of the startup testing program are described

  11. 77 FR 4650 - Airworthiness Directives; General Electric Company Turbofan Engines

    Science.gov (United States)

    2012-01-31

    ... Airworthiness Directives; General Electric Company Turbofan Engines AGENCY: Federal Aviation Administration (FAA... General Electric Company (GE) CF6-45 and CF6-50 series turbofan engines with certain low-pressure turbine... Compliance We estimate that this AD will affect 387 CF6-45 and CF6-50 series turbofan engines installed on...

  12. 78 FR 72567 - Airworthiness Directives; General Electric Company Turbofan Engines

    Science.gov (United States)

    2013-12-03

    ... Airworthiness Directives; General Electric Company Turbofan Engines AGENCY: Federal Aviation Administration (FAA... General Electric Company (GE) GE90-110B1 and -115B turbofan engines. This AD was prompted by multiple... turbofan engines with variable bypass valve (VBV) actuator fuel supply tube, part number (P/N) 2165M22P01...

  13. 78 FR 76045 - Airworthiness Directives; General Electric Company Turbofan Engines

    Science.gov (United States)

    2013-12-16

    ... Airworthiness Directives; General Electric Company Turbofan Engines AGENCY: Federal Aviation Administration (FAA... (AD) for General Electric Company (GE) GE90-110B1 and GE90-115B turbofan engines with certain high... turbofan engines with high pressure compressor (HPC) rotor stage 2-5 spools, part numbers (P/Ns) 351-103...

  14. 78 FR 38195 - Airworthiness Directives; General Electric Company Turbofan Engines

    Science.gov (United States)

    2013-06-26

    ... Directives; General Electric Company Turbofan Engines AGENCY: Federal Aviation Administration (FAA), DOT... all General Electric Company (GE) GE90-110B1 and GE90-115B turbofan engines. This emergency AD was.... owners and operators of these GE90-110B1 and GE90-115B turbofan engines. This action was prompted by...

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

    International Nuclear Information System (INIS)

    Vejvoda, E.J.; Pottmeyer, J.A.; DeLorenzo, D.S.; Weyns-Rollosson, M.I.; Duncan, D.R.

    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

  16. 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.

  17. The behavior of fission products during nuclear rocket reactor tests

    International Nuclear Information System (INIS)

    Bokor, P.C.; Kirk, W.L.; Bohl, R.J.

    1991-01-01

    Fission product release from nuclear rocket propulsion reactor fuel is an important consideration for nuclear rocket development and application. Fission product data from the last six reactors of the Rover program are collected in this paper to provide as basis for addressing development and testing issues. Fission product loss from the fuel will depend on fuel composition and reactor design and operating parameters. During ground testing, fission products can be contained downstream of the reactor. The last Rover reactor tested, the Nuclear Furnance, was mated to an effluent clean-up system that was effective in preventing the discharge of fission products into the atmosphere

  18. Standard Technical Specifications, General Electric plants, BWR/4

    International Nuclear Information System (INIS)

    1992-09-01

    This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/6, and documents the positions of the Nuclear Regulatory Commission based on the B ampersand W Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. 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. This document Volume 3, contains the Bases for Sections 3.4--3.10 of the improved STS

  19. Standard Technical Specifications, General Electric Plants, BWR/6

    International Nuclear Information System (INIS)

    1992-09-01

    This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/6, and documents the positions of the Nuclear Regulatory Commission based on the B ampersand W Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. 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. This document 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

  20. Standard Technical Specifications, General Electric plants, BWR/4

    International Nuclear Information System (INIS)

    1992-09-01

    This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/6, and documents the positions of the Nuclear Regulatory Commission based on the BWR Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. 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. This document 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. Standard Technical Specifications, General Electric Plants, BWR/6

    International Nuclear Information System (INIS)

    1992-09-01

    This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/4, and documents the positions of the Nuclear Regulatory Commission based on the BWR Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. 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

  2. Standard Technical Specifications, General Electric plants, BWR/6

    International Nuclear Information System (INIS)

    1992-09-01

    This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/6, and documents the positions of the Nuclear Regulatory Commission based on the B ampersand W Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. 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. This document 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

  3. NRC review of passive reactor design certification testing programs: Overview and regulatory perspective

    International Nuclear Information System (INIS)

    Levin, A.E.

    1993-01-01

    Reactor vendors are developing new designs for future deployment, including open-quotes passiveclose quotes light water reactors (LWRs), such as General Electric's (G.E.'s) simplified boiling water reactor (SBWR) and Westinghouse's AP600, which depend primarily on inherent processes, such as national convection and gravity feed, for safety injection and emergency core cooling. The U.S. Nuclear Regulatory Commission (NRC) has implemented a new process, certification of standardized reactor designs, for licensing these Plants. Part 52 of Title 10 of the Code of Federal Regulations (10CFR52) contains the requirements that vendors must meet for design certification. One important section, 10CFR52.47, reads open-quotes Certification of a standard design which . . . utilizes simplified, inherent, passive, or other innovative means to accomplish its safety functions will be granted only if: (1) The performance of each safety feature of the design has been demonstrated through either analysis, appropriate test programs, experience, or a combination thereof; (2) Interdependent effects among the safety features have been found acceptable by analysis, appropriate test programs, experience, or a combination thereof; and (3) Sufficient data exist on the safety features of the design to assess the analytical tools used for safety analyses. . . . The vendors have initiated programs to test innovative features of their designs and to develop data bases needed to validate their analytical codes, as required by the design certification rule. Accordingly, the NRC is reviewing and evaluating the vendors programs to ensure that they address adequately key issues concerning safety system performance. This paper provides an overview of the NRC's review process and regulatory perspective

  4. Automated reactor protection testing saves time and avoids errors

    International Nuclear Information System (INIS)

    Raimondo, E.

    1990-01-01

    When the Pressurized Water Reactor units in the French 900MWe series were designed, the instrumentation and control systems were equipped for manual periodic testing. Manual reactor protection system testing has since been successfully replaced by an automatic system, which is also applicable to other instrumentation testing. A study on the complete automation of process instrumentation testing has been carried out. (author)

  5. TREAT [Transient Reactor Test Facility] reactor control rod scram system simulations and testing

    International Nuclear Information System (INIS)

    Solbrig, C.W.; Stevens, W.W.

    1990-01-01

    Air cylinders moving heavy components (100 to 300 lbs) at high speeds (above 300 in/sec) present a formidable end-cushion-shock problem. With no speed control, the moving components can reach over 600 in/sec if the air cylinder has a 5 ft stroke. This paper presents an overview of a successful upgrade modification to an existing reactor control rod drive design using a computer model to simulate the modified system performance for system design analysis. This design uses a high speed air cylinder to rapidly insert control rods (278 lb moved 5 ft in less than 300 msec) to scram an air-cooled test reactor. Included is information about the computer models developed to simulate high-speed air cylinder operation and a unique new speed control and end cushion design. A patent application is pending with the US Patent ampersand Trade Mark Office for this system (DOE case number S-68,622). The evolution of the design, from computer simulations thru operational testing in a test stand (simulating in-reactor operating conditions) to installation and use in the reactor, is also described. 6 figs

  6. Corrosion of spent Advanced Test Reactor fuel

    International Nuclear Information System (INIS)

    Lundberg, L.B.; Croson, M.L.

    1994-01-01

    The results of a study of the condition of spent nuclear fuel elements from the Advanced Test Reactor (ATR) currently being stored underwater at the Idaho National Engineering Laboratory (INEL) are presented. This study was motivated by a need to estimate the corrosion behavior of dried, spent ATR fuel elements during dry storage for periods up to 50 years. The study indicated that the condition of spent ATR fuel elements currently stored underwater at the INEL is not very well known. Based on the limited data and observed corrosion behavior in the reactor and in underwater storage, it was concluded that many of the fuel elements currently stored under water in the facility called ICPP-603 FSF are in a degraded condition, and it is probable that many have breached cladding. The anticipated dehydration behavior of corroded spent ATR fuel elements was also studied, and a list of issues to be addressed by fuel element characterization before and after forced drying of the fuel elements and during dry storage is presented

  7. Electrical characterization and an equivalent circuit model of a microhollow cathode discharge reactor

    International Nuclear Information System (INIS)

    Taylan, O.; Berberoglu, H.

    2014-01-01

    This paper reports the electrical characterization and an equivalent circuit of a microhollow cathode discharge (MHCD) reactor in the self-pulsing regime. A MHCD reactor was prototyped for air plasma generation, and its current-voltage characteristics were measured experimentally in the self-pulsing regime for applied voltages from 2000 to 3000 V. The reactor was modeled as a capacitor in parallel with a variable resistor. A stray capacitance was also introduced to the circuit model to represent the capacitance of the circuit elements in the experimental setup. The values of the resistor and capacitors were recovered from experimental data, and the proposed circuit model was validated with independent experiments. Experimental data showed that increasing the applied voltage increased the current, self-pulsing frequency and average power consumption of the reactor, while it decreased the peak voltage. The maximum and the minimum voltages obtained using the model were in agreement with the experimental data within 2.5%, whereas the differences between peak current values were less than 1%. At all applied voltages, the equivalent circuit model was able to accurately represent the peak and average power consumption as well as the self-pulsing frequency within the experimental uncertainty. Although the results shown in this paper was for atmospheric air pressures, the proposed equivalent circuit model of the MHCD reactor could be generalized for other gases at different pressures.

  8. General safety orientations of the Jules Horowitz Reactor Project (JHRP)

    International Nuclear Information System (INIS)

    Tremodeux, P.; Fiorini, G.L.

    2000-01-01

    After a brief reminder of the JHR purpose, the document outlines the General Safety related Orientations/Recommendations used for the design and the safety assessment of the facility. As far as the JHR design is new, the safety philosophy adopted for this reactor will be as consistent as possible with that recommended for future (power...) reactors. The general recommendations developed in the paper are: the general nuclear safety approach for the design, operation and analysis with, in particular, the adoption of the Defence In Depth principle; the general safety objectives in terms of radiological consequences; the use of Probabilistic Safety Studies; quality assurance. The 'Defence in Depth' concept using amongst others the 'Barrier' principle remains the basis of the JHR safety. 'Defence In Depth' is applied both to design and operation. Its adequacy is checked during the safety assessment and the paper gives the technical recommendations that should allow the designer to implement this concept into the final design. Built mainly for experimental irradiation the JHR facilities will be handled according to conventional or new operation rules which could put materials under stress and entail handling errors. Specific recommendations are defined to take into account the corresponding peculiarities; they are discussed in the paper. The safety design of the JHR takes into account the experience accumulated through the CEA experimental irradiation programmes, which represents several dozen reactor years; the consultation of CEA reactor facilities operators is ongoing. The corresponding feedback is shortly described. Recommendations related to maintenance and associated operation are indicated as well as those regarding the human factor. Details are given on the JHR safety practical implementation through the CEA/DRN Safety approach. Details of the corresponding Safety Objectives are also discussed. Finally the designer position on the role of probabilistic safety

  9. Proposal of world network on material testing reactors

    International Nuclear Information System (INIS)

    Takemoto, Noriyuki; Izumo, Hironobu; Hori, Naohiko; Ishitsuka, Etsuo; Ishihara, Masahiro

    2011-01-01

    Establishment of an international cooperation system of worldwide testing reactor network (world network) is proposed in order to achieve efficient facility utilization and provide high quality irradiation data by role sharing of irradiation tests with materials testing reactors in the world. As for the first step, mutual understanding among materials testing reactors is thought to be necessary. From this point, an international symposium on materials testing reactors (ISMTR) was held to construct the world network from 2008, and a common understanding of world network has begun to be shared. (author)

  10. Fast reactor safety testing in Transient Reactor Test (TREAT) in the 1980s

    International Nuclear Information System (INIS)

    Wright, A.E.; Dutt, D.S.; Harrison, L.J.

    1990-01-01

    Several series of fast reactor safety tests were performed in TREAT during the 1980s. These focused on the transient behavior of full-length oxide fuels (US reference, UK reference, and US advanced design) and on modern metallic fuels. Most of the tests addressed fuel behavior under transient overpower or loss-of-flow conditions. The test series were the PFR/TREAT tests; the RFT, TS, CDT, and RX series on oxide fuels; and the M series on metallic fuels. These are described in terms of their principal results and relevance to analyses and safety evaluation. 4 refs., 3 tabs

  11. Core test reactor shield cooling system analysis

    International Nuclear Information System (INIS)

    Larson, E.M.; Elliott, R.D.

    1971-01-01

    System requirements for cooling the shield within the vacuum vessel for the core test reactor are analyzed. The total heat to be removed by the coolant system is less than 22,700 Btu/hr, with an additional 4600 Btu/hr to be removed by the 2-inch thick steel plate below the shield. The maximum temperature of the concrete in the shield can be kept below 200 0 F if the shield plug walls are kept below 160 0 F. The walls of the two ''donut'' shaped shield segments, which are cooled by the water from the shield and vessel cooling system, should operate below 95 0 F. The walls of the center plug, which are cooled with nitrogen, should operate below 100 0 F. (U.S.)

  12. The Advanced Test Reactor Strategic Evaluation Program

    International Nuclear Information System (INIS)

    Buescher, B.J.

    1990-01-01

    A systematic evaluation of safety, environmental, and operational issues has been initiated at the Advanced Test Reactor (ATR). This program, the Strategic Evaluation Program (STEP), provides an integrated review of safety and operational issues against the standards applied to licensed commercial facilities. In the review of safety issues, 18 deviations were identified which required prompt attention. Resolution of these items has been accelerated in the program. An integrated living schedule is being developed to address the remaining findings. A risk evaluation is being performed on the proposed corrective actions and these actions will then be formally ranked in order of priority based on considerations of safety and operational significance. Once the final ranking is completed, an integrated schedule will be developed, which will include considerations of availability of funding and operating schedule. 3 refs., 2 figs

  13. Probability of a surface rupture offset beneath a nuclear test reactor

    International Nuclear Information System (INIS)

    Reed, J.W.; Meehan, R.L.; Crellin, G.L.

    1981-01-01

    A probabilistic analysis was conducted to determine the likelihood of a surface rupture offset of any size beneath the 50 megawatt General Electric Test Reactor (GETR), which is located at the Vallecitos Nuclear Center near Pleasanton, California. Geologic faults have been observed at the GETR site. These faults may be due to surface folds, landslides, or deep tectonic movement. They are referred to in the paper as 'existing faults;' however, use of this term does not imply that they are tectonic in origin. The objective of the analysis was to evaluate whether a conservative estimate of the probability of occurrence of a future fault movement is sufficiently low so that movement beneath the reactor building need not be considered as a design basis event. The reactor building is located between two existing faults which are approximately 1320 feet apart. If a fault movement occurs in the future, it is conservatively assumed to occur either on the existing faults or between the faults, or on a fault(s) and between the two faults at the same time. The probabilistic model included the possibility of movements occurring due to unknown, undiscovered faults in the region. For this part, movements were assumed to occur according to a Poisson process. For the possibility of new faults occurring due to the two existing faults, a hazard function was used which increases with time since the last offset. (orig./RW)

  14. JENDL-3.3 thermal reactor benchmark test

    International Nuclear Information System (INIS)

    Akie, Hiroshi

    2001-01-01

    Integral tests of JENDL-3.2 nuclear data library have been carried out by Reactor Integral Test WG of Japanese Nuclear Data Committee. The most important problem in the thermal reactor benchmark testing was the overestimation of the multiplication factor of the U fueled cores. With several revisions of the data of 235 U and the other nuclides, JENDL-3.3 data library gives a good estimation of multiplication factors both for U and Pu fueled thermal reactors. (author)

  15. Final Physics Report for the Engineering Test Reactor

    International Nuclear Information System (INIS)

    Wolfe, I. B.

    1956-01-01

    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

  16. Automated testing of reactor protection instrumentation made easy

    International Nuclear Information System (INIS)

    Iborra, A.; De Marcos, F.; Pastor, J.A.; Alvarez, B.; Jimenez, A.; Mesa, E.; Alsonso, L.; Regidor, J.J.

    1997-01-01

    Maintenance and testing of reactor protection systems is an important cause of unplanned reactor trips. Automated testing is the answer because it minimises test times and reduces human error. The GAMA I system, developed and implemented at Vandellos II in Spain, has the added advantage that it uses visual programming, which means that changing the software does not need specialist programming skills. (author)

  17. Nuclear reactor vessel surface inspecting technique applying electric resistance probe

    International Nuclear Information System (INIS)

    Yamaguchi, T.; Enami, K.; Yoshioka, M.

    1975-01-01

    A new technique for inspecting the inner surface of the PWR type nuclear reactor vessel by use of an electric resistance probe is introduced, centering on a data processing system. This system is composed of a mini-computer, a system typewriter, an interface unit, a D-A converter and controller, and X-Y recorder and others. Its functions are judging flaws and making flaw detection maps. In order to judge flaws by flaw detection signals, three kinds of flaw judging methods have been developed. In case there is a flaw, its position and depth are calculated and listed on the system typewriter. The flaw detection maps are expressed in four kinds of modes and they are displayed on the X-Y recorder. (auth.)

  18. Analysis and evaluation of ZPPR critical experiments for a 100 kilowatt-electric space reactor

    International Nuclear Information System (INIS)

    McFarlane, H.F.; Collins, P.J.; Carpenter, S.G.; Olsen, D.N.; Smith, D.M.; Schaefer, R.W.; Doncals, R.A.; Andre, S.V.; Porter, C.A.; Cowan, C.L.; Stewart, S.L.; Protsik, R.

    1990-01-01

    ZPPR critical experiments were used for physics testing the reactor design of the SP-100, a 100-kW thermoelectric LMR that is being developed to provide electrical power for space applications. These tests validated all key physics characteristics of the design, including the ultimate safety in the event of a launch or re-entry accident. Both the experiments and the analysis required the use of techniques not previously needed for fast reactor designs. A few significant discrepancies between the experimental and calculated results leave opportunities for further reductions in the mass of the SP-100. An initial investigation has been made into application of the ZPPR-20 results, along with those of other relevant integral data, to the SP-100 design

  19. 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.

  20. Advanced Test Reactor outage risk assessment

    International Nuclear Information System (INIS)

    Thatcher, T.A.; Atkinson, S.A.

    1997-01-01

    Beginning in 1997, risk assessment was performed for each Advanced Test Reactor (ATR) outage aiding the coordination of plant configuration and work activities (maintenance, construction projects, etc.) to minimize the risk of reactor fuel damage and to improve defense-in-depth. The risk assessment activities move beyond simply meeting Technical Safety Requirements to increase the awareness of risk sensitive configurations, to focus increased attention on the higher risk activities, and to seek cost-effective design or operational changes that reduce risk. A detailed probabilistic risk assessment (PRA) had been performed to assess the risk of fuel damage during shutdown operations including heavy load handling. This resulted in several design changes to improve safety; however, evaluation of individual outages had not been performed previously and many risk insights were not being utilized in outage planning. The shutdown PRA provided the necessary framework for assessing relative and absolute risk levels and assessing defense-in-depth. Guidelines were written identifying combinations of equipment outages to avoid. Screening criteria were developed for the selection of work activities to receive review. Tabulation of inherent and work-related initiating events and their relative risk level versus plant mode has aided identification of the risk level the scheduled work involves. Preoutage reviews are conducted and post-outage risk assessment is documented to summarize the positive and negative aspects of the outage with regard to risk. The risk for the outage is compared to the risk level that would result from optimal scheduling of the work to be performed and to baseline or average past performance

  1. 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.

  2. International Experience with Fast Reactor Operation & Testing

    International Nuclear Information System (INIS)

    Sackett, John I.; Grandy, C.

    2013-01-01

    Conclusion: • Worldwide experience with fast reactors has demonstrated the robustness of the technology and it stands ready for worldwide deployment. • The lessons learned are many and there is danger that what has been learned will be forgotten given that there is little activity in fast reactor development at the present time. • For this reason it is essential that knowledge of fast reactor technology be preserved, an activity supported in the U.S. as well as other countries

  3. Imperfection detection probability at ultrasonic testing of reactor vessels

    International Nuclear Information System (INIS)

    Kazinczy, F. de; Koernvik, L.Aa.

    1980-02-01

    The report is a lecture given at a symposium organized by the Swedish nuclear power inspectorate on February 1980. Equipments, calibration and testing procedures are reported. The estimation of defect detection probability for ultrasonic tests and the reliability of literature data are discussed. Practical testing of reactor vessels and welded joints are described. Swedish test procedures are compared with other countries. Series of test data for welded joints of the OKG-2 reactor are presented. Future recommendations for testing procedures are made. (GBn)

  4. Results of assembly test of HTTR reactor internals

    International Nuclear Information System (INIS)

    Maruyama, S.; Saikusa, A.; Shiozawa, S.; Tsuji, N.; Miki, T.

    1996-01-01

    The assembly test of the HTTR actual reactor internals had been carried out at the works, prior to their installation in the actual reactor pressure vessel(RPV) at the construction site. The assembly test consists of several items such as examining fabricating precision of each component and alignment of piled-up structures, measuring circumferential coolant velocity profile in the passage between the simulated RPV and the reactor internals as well as under the support plates, measuring by-pass flow rate through gaps between the reactor internals, and measuring the binding force of the core restraint mechanism. Results of the test showed good performance of the HTTR reactor internals. Installation of the reactor internals in the actual RPV was started at the construction site of HTTR in April, 1995. In the installation process, main items of the assembly test at the works were repeated to investigate the reproducibility of installation. (author). 5 refs, 11 figs

  5. 78 FR 19983 - Airworthiness Directives; General Electric Company Turbofan Engines

    Science.gov (United States)

    2013-04-03

    ... Airworthiness Directives; General Electric Company Turbofan Engines AGENCY: Federal Aviation Administration (FAA... Electric Company (GE) CF34-8C and CF34-8E turbofan engines with certain part numbers (P/N) of operability...-8E6, and CF34-8E6A1 turbofan engines, with an operability bleed valve (OBV) part number (P/N...

  6. 78 FR 50320 - Airworthiness Directives; General Electric Company Turbofan Engines

    Science.gov (United States)

    2013-08-19

    ... Airworthiness Directives; General Electric Company Turbofan Engines AGENCY: Federal Aviation Administration (FAA... Electric Company (GE) model GEnx-2B67B turbofan engines with booster anti-ice (BAI) air duct, part number...-2B67 turbofan engine be removed from the Applicability section of this AD. The commenters noted that...

  7. 77 FR 3088 - Airworthiness Directives; General Electric Company Turbofan Engines

    Science.gov (United States)

    2012-01-23

    ... Airworthiness Directives; General Electric Company Turbofan Engines AGENCY: Federal Aviation Administration (FAA... Electric Company (GE) CF34-10E series turbofan engines. This AD was prompted by a report of heavy wear... turbofan engines installed on airplanes of U.S. registry. We also estimate that it will take about 8 work...

  8. Local electric dipole moments: A generalized approach.

    Science.gov (United States)

    Groß, Lynn; Herrmann, Carmen

    2016-09-30

    We present an approach for calculating local electric dipole moments for fragments of molecular or supramolecular systems. This is important for understanding chemical gating and solvent effects in nanoelectronics, atomic force microscopy, and intensities in infrared spectroscopy. Owing to the nonzero partial charge of most fragments, "naively" defined local dipole moments are origin-dependent. Inspired by previous work based on Bader's atoms-in-molecules (AIM) partitioning, we derive a definition of fragment dipole moments which achieves origin-independence by relying on internal reference points. Instead of bond critical points (BCPs) as in existing approaches, we use as few reference points as possible, which are located between the fragment and the remainder(s) of the system and may be chosen based on chemical intuition. This allows our approach to be used with AIM implementations that circumvent the calculation of critical points for reasons of computational efficiency, for cases where no BCPs are found due to large interfragment distances, and with local partitioning schemes other than AIM which do not provide BCPs. It is applicable to both covalently and noncovalently bound systems. © 2016 Wiley Periodicals, Inc. © 2016 Wiley Periodicals, Inc.

  9. 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

  10. 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

  11. EPR, a GEN 3 Reactor providing a competitive electricity cost

    International Nuclear Information System (INIS)

    Salhi, Othman

    2006-01-01

    Since the very beginning of the development of what was to become the EPR, several European entities were involved. The French and German safety authorities expressed that reinforced safety was compulsory. Additional measures were then included to prevent the occurrence of events likely to damage the core, and reduce the possibility of exposure of operating and maintenance personnel. However, not with standing these safety related features resulting from the requirements of the safety authorities, we will focus today on another group of entities that were key players in EPR development: the Utilities. The Utilities voiced their need for a competitive electricity produced and a competitive nuclear reactor. The tradeoff was then to reach both targets in a unique product: a safer and more competitive NPP. Today, the EPR presents features that enable our clients to compete with the cheapest fossil-based electricity production plants. Increased thermal efficiency is obtained both through a higher steam pressure and through careful optimization of the secondary system thermal cycle

  12. Possible Future Role of Small and Medium Sized Reactors (SMRs) in Countries with Small and Medium Electricity Grids

    International Nuclear Information System (INIS)

    Alujevic, L.

    2016-01-01

    Small and Medium Sized Reactors (SMRs) could have enormous potential as options for enhancing the energy supply security, as well as providing a lower capital investment compared to conventional Nuclear Power Plants (NPPs). The trend in SMR development has been towards design certification of small modular reactors, defined as advanced reactors that produce electric power up to 300 MW(e), designed to be built in factories and shipped to utilities for installation as demand arises. The factory-built small modular reactors aim to reduce lengthy construction times while simultaneously increasing quality, thereby minimizing the costs associated with the current time for construction that spans 5 to 8 years. SMR designs include water-cooled reactors, high temperature gas cooled reactors, as well as liquid metal cooled reactors with fast neutron spectrum. Also, many are designed to be emplaced below ground level, giving a high resistance to terrorist threats. The projected timelines of readiness for deployment of SMRs generally range from the present to 2025 - 2030. Currently there are more than 45 SMR designs under development for different application issues. This paper will try to elaborate the benefits and drawbacks of SMRs, as well as describe a couple of designs. Furthermore, some timelines and cost estimates will be provided, depending on the data currently available. Taking all that into account, the conclusion will try to ascertain the suitability of SMRs for Countries with Small and Medium Electricity Grids, namely Croatia. (author).

  13. 78 FR 47534 - Airworthiness Directives; General Electric Company Turbofan Engines

    Science.gov (United States)

    2013-08-06

    ... Airworthiness Directives; General Electric Company Turbofan Engines AGENCY: Federal Aviation Administration (FAA... directive (AD) 2013-14-51 for General Electric Company (GE) GE90-110B1 and GE90-115B turbofan engines with... all known U.S. owners and operators of GE90-110B1 and GE90-115B turbofan engines. AD 2013-14-51...

  14. 78 FR 24671 - Airworthiness Directives; General Electric Company Turbofan Engines

    Science.gov (United States)

    2013-04-26

    ... Airworthiness Directives; General Electric Company Turbofan Engines AGENCY: Federal Aviation Administration (FAA... certain General Electric Company (GE) CF6-80C2 series turbofan engines. That AD currently requires.../B1F/B2F/B4F/B6F/B7F/D1F turbofan engines with any of the following installed: (1) Fuel tube, part...

  15. Composite electric generator equipped with steam generator for heating reactor coolant

    International Nuclear Information System (INIS)

    Watabe, Masaharu; Soman, Yoshindo; Kawanishi, Kohei; Ota, Masato.

    1997-01-01

    The present invention concerns a composite electric generator having coolants, as a heating source, of a PWR type reactor or a thermonuclear reactor. An electric generator driving gas turbine is disposed, and a superheater using a high temperature exhaust gas of the gas turbine as a heating source is disposed, and main steams are superheated by the superheater to elevate the temperature at the inlet of the turbine. This can increase the electric generation capacity as well as increase the electric generation efficiency. In addition, since the humidity in the vicinity of the exit of the steam turbine is reduced, occurrence of loss and erosion can be suppressed. When cooling water of the thermonuclear reactor is used, the electric power generated by the electric generator driven by the gas turbine can be used upon start of the thermonuclear reactor, and it is not necessary to dispose a large scaled special power source in the vicinity, which is efficient. (N.H.)

  16. The combined use of test reactor experiments and power reactor tests for the development of PCI-resistant fuel

    International Nuclear Information System (INIS)

    Junkrans, S.; Vesterlund, G.; Vaernild, O.

    1980-01-01

    The theme of this paper is that for development of PCI-resistant fuel acceptable from the commercial and licensing aspects, extensive and time-consuming work is needed both in a test reactor and in power reactors. The test reactor is necessary for ramp testing to power levels not allowed in power reactors and with the aim of generating fuel failures. It is also used for other special irradiation experiments. The access to power reactors is necessary to generate information on performance in a real LWR core and to incubate at a reasonable cost the large amount of rods required for test reactor ramping. Selected results from the ASEA-ATOM work are used to support these conclusions. (author)

  17. Development of electrically insulating self-healing coatings in vanadium alloys for lithium fusion reactor

    International Nuclear Information System (INIS)

    1999-01-01

    Problems on electrically insulating self-healing coatings (SHC) on vanadium alloys for lithium fusion reactor systems are considered. In particular, the SHC stability and radiation resistance in lithium and effect of magnetic field on the efficiency of the TNR lithium systems are studied. New technological methods for application of self-healing coatings and study on their properties are developed. The vanadium-lithium materials testing in pile loops for solution of the above problems under conditions of the lithium TNR is described [ru

  18. 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.

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

    International Nuclear Information System (INIS)

    Stallmann, F.W.; Wang, J.A.; Kam, F.B.K.

    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

  20. Review of the general atomic experimental fusion power reactor initial conceptual design

    International Nuclear Information System (INIS)

    Baker, C.C.; Sager, P.H. Jr.; Harder, C.R.

    1976-01-01

    The primary objective of the Experimental Power Reactor (EPR) is to provide the necessary interface between physics experiments and the first demonstration power plants. Since economically viable tokamak-type reactors may well have to be very high Q devices (ratio of fusion power out to power into the plasma), it will be essential for a tokamak demonstration reactor to operate at or near ignition conditions. Thus, it is believed that one of the primary objectives of the EPR must be to fully model the behavior of a D-T burning plasma required in the reactor of a demonstration plant. Therefore, a major objective of the EPR should be to achieve ignition conditions. In addition to demonstrating the ability to ignite and control a D-T plasma, it is also desirable that the EPR should produce, or at least demonstrate the ability to produce, a small amount of net electrical power. These objectives should be accomplished at a reasonable cost; this implies achieving a sufficiently high β (ratio of plasma pressure to magnetic field pressure). It is believed that noncircular cross section tokamaks offer the best chance of realizing these objectives. Consequently, noncircular cross sections are a major design feature of the General Atomic EPR

  1. Space reactor fuel element testing in upgraded TREAT

    International Nuclear Information System (INIS)

    Todosow, M.; Bezler, P.; Ludewig, H.; Kato, W.Y.

    1993-01-01

    The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., is a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. Initial results suggests that full-scale PBR elements could be tested at an average energy deposition of ∼60--80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperture limit, average energy deposition of ∼100 MW/L may be achievable

  2. Space reactor fuel element testing in upgraded TREAT

    Science.gov (United States)

    Todosow, Michael; Bezler, Paul; Ludewig, Hans; Kato, Walter Y.

    1993-01-01

    The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., is a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. Initial results suggests that full-scale PBR elements could be tested at an average energy deposition of ˜60-80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperture limit, average energy deposition of ˜100 MW/L may be achievable.

  3. Reactor/Brayton power systems for nuclear electric spacecraft

    Science.gov (United States)

    Layton, J. P.

    1980-01-01

    Studies are currently underway to assess the technological feasibility of a nuclear-reactor-powered spacecraft propelled by electric thrusters. This vehicle would be capable of performing detailed exploration of the outer planets of the solar system during the remainder of this century. The purpose of this study was to provide comparative information on a closed cycle gas turbine power conversion system. The results have shown that the performance is very competitive and that a 400 kWe space power system is dimensionally compatible with a single Space Shuttle launch. Performance parameters of system mass and radiator area were determined for systems from 100 to 1000 kWe. A 400 kWe reference system received primary attention. The components of this system were defined and a conceptual layout was developed with encouraging results. The preliminary mass determination for the complete power system was very close to the desired goal of 20 kg/kWe. Use of more advanced technology (higher turbine inlet temperature) will substantially improve system performance characteristics.

  4. Advanced High-Temperature Reactor for Production of Electricity and Hydrogen: Molten-Salt-Coolant, Graphite-Coated-Particle-Fuel

    International Nuclear Information System (INIS)

    Forsberg, C.W.

    2002-01-01

    The objective of the Advanced High-Temperature Reactor (AHTR) is to provide the very high temperatures necessary to enable low-cost (1) efficient thermochemical production of hydrogen and (2) efficient production of electricity. The proposed AHTR uses coated-particle graphite fuel similar to the fuel used in modular high-temperature gas-cooled reactors (MHTGRs), such as the General Atomics gas turbine-modular helium reactor (GT-MHR). However, unlike the MHTGRs, the AHTR uses a molten salt coolant with a pool configuration, similar to that of the PRISM liquid metal reactor. A multi-reheat helium Brayton (gas-turbine) cycle, with efficiencies >50%, is used to produce electricity. This approach (1) minimizes requirements for new technology development and (2) results in an advanced reactor concept that operates at essentially ambient pressures and at very high temperatures. The low-pressure molten-salt coolant, with its high heat capacity and natural circulation heat transfer capability, creates the potential for (1) exceptionally robust safety (including passive decay-heat removal) and (2) allows scaling to large reactor sizes [∼1000 Mw(e)] with passive safety systems to provide the potential for improved economics

  5. 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.

  6. Development of dust removal system using static electricity for fusion experimental reactors

    Energy Technology Data Exchange (ETDEWEB)

    Onozuka, Masanori; Ueda, Yasutoshi; Oda, Yasushi; Takahashi, Kenji [Mitsubishi Heavy Industries Ltd., Tokyo (Japan); Seki, Yasushi; Aoki, Isao; Ueda, Shuzo; Kurihara, Ryoichi

    1997-11-01

    Tests to collect and transport metallic and non-metallic dust particles have been conducted using static electricity in a vacuum environment to investigate the applicability of a static electricity dust removal system for fusion experimental reactors. The dust particles are charged by electrostatic induction, floated and collected due to the Coulomb force generated by the AC electric field. They are then transported due to the gradient force induced by the electric curtain of the non-uniform travelling-wave electric field. Using a fully insulated electrode with a single-phase AC voltage up to 15 kV, aluminum and carbon dust were successfully collected. The highest collection rates for the aluminum and carbon dust were around 30 and 2 g/min, respectively. The linear-type electrodes, using as high as 22 kV of the three-phase AC voltage, transported aluminum dust up to an angle of 60deg. Applying a guide electrode to the linear-type electrode, the transportation rate was approximately doubled and almost constant at every angle, including a 90deg angle. The system transported aluminum dust up to the rate of 13 g/min. The influence of the 0.15 T magnetic field on the dust collection and transportation efficiencies was found to be negligible. (author)

  7. Development of dust removal system using static electricity for fusion experimental reactors

    International Nuclear Information System (INIS)

    Onozuka, Masanori; Ueda, Yasutoshi; Oda, Yasushi; Takahashi, Kenji; Seki, Yasushi; Aoki, Isao; Ueda, Shuzo; Kurihara, Ryoichi.

    1997-01-01

    Tests to collect and transport metallic and non-metallic dust particles have been conducted using static electricity in a vacuum environment to investigate the applicability of a static electricity dust removal system for fusion experimental reactors. The dust particles are charged by electrostatic induction, floated and collected due to the Coulomb force generated by the AC electric field. They are then transported due to the gradient force induced by the electric curtain of the non-uniform travelling-wave electric field. Using a fully insulated electrode with a single-phase AC voltage up to 15 kV, aluminum and carbon dust were successfully collected. The highest collection rates for the aluminum and carbon dust were around 30 and 2 g/min, respectively. The linear-type electrodes, using as high as 22 kV of the three-phase AC voltage, transported aluminum dust up to an angle of 60deg. Applying a guide electrode to the linear-type electrode, the transportation rate was approximately doubled and almost constant at every angle, including a 90deg angle. The system transported aluminum dust up to the rate of 13 g/min. The influence of the 0.15 T magnetic field on the dust collection and transportation efficiencies was found to be negligible. (author)

  8. 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.

  9. 77 FR 26607 - Energy Conservation Program: Test Procedures for Electric Motors and Small Electric Motors

    Science.gov (United States)

    2012-05-04

    ... Conservation Program: Test Procedures for Electric Motors and Small Electric Motors; Final Rules #0;#0;Federal... Procedures for Electric Motors and Small Electric Motors AGENCY: Office of Energy Efficiency and Renewable... electric motors and small electric motors. That supplemental proposal, along with an earlier proposal from...

  10. Research reactors for power reactor fuel and materials testing - Studsvik's experience

    International Nuclear Information System (INIS)

    Grounes, M.

    1998-01-01

    Presently Studsvik's R2 test reactor is used for BWR and PWR fuel irradiations at constant power and under transient power conditions. Furthermore tests are performed with defective LWR fuel rods. Tests are also performed on different types of LWR cladding materials and structural materials including post-irradiation testing of materials irradiated at different temperatures and, in some cases, in different water chemistries and on fusion reactor materials. In the past, tests have also been performed on HTGR fuel and FBR fuel and materials under appropriate coolant, temperature and pressure conditions. Fuel tests under development include extremely fast power ramps simulating some reactivity initiated accidents and stored energy (enthalpy) measurements. Materials tests under development include different types of in-pile tests including tests in the INCA (In-Core Autoclave) facility .The present and future demands on the test reactor fuel in all these cases are discussed. (author)

  11. Power reactors in Member States. 1978 edition

    International Nuclear Information System (INIS)

    1978-01-01

    The computer-based reactor listing gives information on reactor core characteristics and plant systems for all power reactors in operation under construction and planned. The following two tables are included to give a general picture of the overall situation: Reactor types and net electrical power; Reactor units and net electrical power by country and cumulated by year

  12. 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.

  13. Rise-to-power test in High Temperature Engineering Test Reactor. Test progress and summary of test results up to 30 MW of reactor thermal power

    International Nuclear Information System (INIS)

    Nakagawa, Shigeaki; Fujimoto, Nozomu; Shimakawa, Satoshi

    2002-08-01

    The High Temperature Engineering Test Reactor (HTTR) is a graphite moderated and gas cooled reactor with the thermal power of 30 MW and the reactor outlet coolant temperature of 850degC/950degC. Rise-to-power test in the HTTR was performed from April 23rd to June 6th in 2000 as phase 1 test up to 10 MW in the rated operation mode, from January 29th to March 1st in 2001 as phase 2 test up to 20 MW in the rated operation mode and from April 14th to June 8th in 2001 as phase 3 test up to 20 MW in the high temperature test the mechanism of the reactor outlet coolant temperature becomes 850degC at 30 MW in the rated operation mode and 950degC in the high temperature test operation mode. Phase 4 rise-to-power test to achieve the thermal reactor power of 30 MW started on October 23rd in 2001. On December 7th in 2001 it was confirmed that the thermal reactor power and the reactor outlet coolant temperature reached to 30 MW and 850degC respectively in the single loaded operation mode in which only the primary pressurized water cooler is operating. Phase 4 test was performed until March 6th in 2002. JAERI (Japan Atomic Energy Research Institute) obtained the certificate of the pre-operation test from MEXT (Ministry of Education Culture Sports Science and Technology) after all the pre-operation tests by MEXT were passed successfully with the reactor transient test at an abnormal event as a final pre-operation test. From the test results of the rise-up-power test up to 30 MW in the rated operation mode, performance of the reactor and cooling system were confirmed, and it was also confirmed that an operation of reactor facility can be performed safely. Some problems to be solved were found through the tests. By solving them, the reactor operation with the reactor outlet coolant temperature of 950degC will be achievable. (author)

  14. Start-up test of the prototype heavy water reactor 'FUGEN', (1)

    International Nuclear Information System (INIS)

    Ando, Hideki; Kawahara, Toshio

    1982-01-01

    The advanced thermal prototype reactor ''Fugen'' is a heavy water-moderated, boiling light water-cooled power reactor with electric output of 165 MW, which has been developed since 1966 as a national project. The start-up test was begun in March, 1978, being scheduled for about one year, and in March, 1979, it passed the final pre-use inspection and began the full scale operation. In this paper, the result of the start-up test of Fugen is reported. From the experience of the start-up test of Fugen, the following matters are important for the execution of start-up test. 1) Exact testing plan and work schedule, 2) the organization to perform the test, 3) the rapid evaluation of test results and the reflection to next testing plan, and 4) the reflection of test results to rated operation, regular inspection and so on. In the testing plan, the core characteristics peculiar to Fugen, and the features of heavy water-helium system, control system and other equipment were added to the contents of the start-up test of BWRs. The items of the start-up test were reactor physics test, plant equipment performance test, plant dynamic characteristic test, chemical and radiation measurement, and combined test. The organization to perform the start-up test, and the progress and the results of the test are reported. (Kako, I.)

  15. Design optimization of general arrangement in Korean next generation reactor

    International Nuclear Information System (INIS)

    Kim, S. H.; Jung, D. W.; Choi, Y. B.; Cho, S. J.

    1999-01-01

    In order to optimize the general arrangement(GA) of Korean Next Generation Reactor (KNGR), field opinions in domestic nuclear power plants have been collected, and the bench marking on UCN No.1,2 which were estimated to be the most excellent in view of operability and maintenance has been accomplished. Through this work, design optimization items for GA were reviewed. Major items to be selected for optimization are summarized as follows; 'Expanding the compound building function and the mezzanine floor concept in the auxiliary building', 'Including the diesel generator building to the auxiliary building', 'Change of the equipment removal method in the auxiliary building'. With these GA design optimization, the auxiliary building boundary will be improved as a complete rectangular type. The power block volume except the changing effect to the single containment structure will be reduced to about 10% in comparison with that of in KNGR phase II

  16. Refurbishing the BR2 materials testing reactor

    International Nuclear Information System (INIS)

    Baugnet, J.M.; Dekeyser, J.; Gubel, P.

    1995-01-01

    SCK/CEN is refurbishing its BR2 reactor to allow its further operation during the next 15 years; in doing so, it chooses to keep BR2 available for future scientific and technological irradiation programs within an international context. (author) 2 figs

  17. Reactor primary pumps dynamic balancing test

    International Nuclear Information System (INIS)

    Lu Qunxian

    2002-01-01

    Reactor primary Pump is the important equipment in the primary circuit, its working quality would directly influence the safety and operation of nuclear power plant. The author describes that the primary pump vibration status, vibration fault diagnosis and dynamic balancing process on site have been performed since commercial operation of DA YA BAY Nuclear Power plant

  18. Design criteria for the electrical system in advanced passive reactors. Special features of the AP-600 Reactor

    International Nuclear Information System (INIS)

    Moraleda Lopez, A.

    1997-01-01

    The design of the electrical system of an Passive Advanced Reactor is determined by the concept of passive actuation of safety systems, simplification of process systems and optimisation of equipment performance. The system that results from these criteria is very different to those designed for present plants. The main differences are: No class 1E alternating current systems No emergency diesel generators Fewer safety and non-safety class electricity consumers System for continuous monitoring of battery status Use of electronic speed regulators for reactor feedwater pump motors Outsite battery backup safety power supply Motor-operated valves are the only safety electrical actuators Portable power supply for post 72 hour equipment This paper develops these concepts and applies them to the AP-600 project and describes the electrical system of this type of plant. (Author)

  19. 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...

  20. Summary - Advanced high-temperature reactor for hydrogen and electricity production

    International Nuclear Information System (INIS)

    Forsberg, Charles W.

    2001-01-01

    Historically, the production of electricity has been assumed to be the primary application of nuclear energy. That may change. The production of hydrogen (H 2 ) may become a significant application. The technology to produce H 2 using nuclear energy imposes different requirements on the reactor, which, in turn, may require development of new types of reactors. Advanced High Temperature reactors can meet the high temperature requirements to achieve this goal. This alternative application of nuclear energy may necessitate changes in the regulatory structure

  1. Procedure for qualification of electric equipment installed in containments for pressurized water reactors subject to accident conditions

    International Nuclear Information System (INIS)

    1991-11-01

    This generic norm is usable for electrical equipment installed in containment building of PWR subject to accidental conditions. She defines the qualification methods and the general rules usable for the test specifications of qualification for these materials

  2. Requalification of SPERT [Special Power Excursion Reactor Test] pins for use in university reactors

    International Nuclear Information System (INIS)

    Snelgrove, J.L.; Domagala, R.F.; Dates, L.R.

    1986-12-01

    A series of nondestructive and destructive examinations have been performed on a representative sample of stainless steel-clad UO 2 fuel pins procured in the early-to-mid 1960s for the SPERT program. These examinations were undertaken in order to requalify the SPERT pins for use in converting university research reactors from the use of highly enriched uranium to the use of low-enriched uranium. The requalification program included visual and dimensional inspections of fuel pins and fuel pellets, radiographic inspections of welds, fill gas analyses, and chemical and spectrographic analyses of fuel and cladding materials. In general all attributes tested were within or very close to specified values, although some weld defects not covered by the original specifications were found. 1 ref., 4 figs., 11 tabs

  3. Interlaboratory computational comparisons of critical fast test reactor pin lattices

    International Nuclear Information System (INIS)

    Mincey, J.F.; Kerr, H.T.; Durst, B.M.

    1979-01-01

    An objective of the Consolidated Fuel Reprocessing Program's (CFRP) nuclear engineering group at Oak Ridge National Laboratory (ORNL) is to ensure that chemical equipment components designed for the reprocessing of spent LMFBR fuel (among other fuel types) are safe from a criticality standpoint. As existing data are inadequate for the general validation of computational models describing mixed plutonium--uranium oxide systems with isotopic compositions typical of LMFBR fuel, a program of critical experiments has been initiated at the Battelle Pacific Northwest Laboratories (PNL). The first series of benchmark experiments consisted of five square-pitched lattices of unirradiated Fast Test Reactor (FTR) fuel moderated and reflected by light water. Calculations of these five experiments have been conducted by both ORNL/CFRP and PNL personnel with the purpose of exploring how accurately various computational models will predict k/sub eff/ values for such neutronic systems and if differences between k/sub eff/ values obtained with these different models are significant

  4. Integral test of JENDL-3.3 for thermal reactors

    International Nuclear Information System (INIS)

    Okumura, Keisuke; Mori, Takamasa

    2003-01-01

    Criticality benchmark testing was carried out for 59 experiments in various thermal reactors using a continues-energy Monte Carlo code MVP and its different libraries generated from JENDL-3.2, JENDL-3.3, JEF-2.2 and ENDF/B-VI (R8). From the benchmark results, we can say JENDL-3.3 generally gives better k eff values compared with other nuclear data libraries. However, further modification of JENDL-3.3 is expected to solve the following problems: 1) systematic underestimation of k eff depending on 235 U enrichment for the cores with low (less than 3wt.%) enriched uranium fueled cores, 2) dependence of C/E value of k eff on neutron spectrum and plutonium composition for MOX fueled cores. These are common problems for all of the nuclear data libraries used in this study. (author)

  5. Non-electric applications of pool-type nuclear reactors

    International Nuclear Information System (INIS)

    Adamov, E.O.; Cherkashov, Yu.M.; Romenkov, A.A.

    1997-01-01

    This paper recommends the use of pool-type light water reactors for thermal energy production. Safety and reliability of these reactors were already demonstrated to the public by the long-term operation of swimming pool research reactors. The paper presents the design experience of two projects: Apatity Underground Nuclear Heating Plant and Nuclear Sea-Water Desalination Plant. The simplicity of pool-type reactors, the ease of their manufacturing and maintenance make this type of a heat source attractive to the countries without a developed nuclear industry. (author). 6 figs, 1 tab

  6. Testing plutonium fuel assembly production for fast-neutron reactors

    International Nuclear Information System (INIS)

    Nougues, B.; Benhamou, A.; Bertothy, G.; Lepetit, H.

    1975-01-01

    The main characteristics of plutonium fuel elements for fast breeder reactors justify specific test procedures and special techniques. The specific tests relating to the Pu content consist of Pu enrichment and distribution tests, determination of the O/M ratio and external contamination tests. The specific tests performed on fuel configuration are: testing of sintered pellet diameter, testing of pin welding and checking of internal assmbly [fr

  7. Development and testing of control rod drives for ship reactors

    International Nuclear Information System (INIS)

    Bruelheide, K.; Mundt, D.; Peters, C.-H.; Manthey, H.-J.

    1978-01-01

    The following paper deals with the development and testings of a new control rod drive design for marine reactors. Starting from the good operating experience with the advanced pressurized water reactor (FDR) of the NS OTTO HAHN a control rod drive system with an hermetically sealed drive principle was developed. A prototype control rod drive system was put through extensive tests and developed ready for standard production at the 'Gesellschaft fuer Kernenergieverwertung in Schiffbau und Schiffahrt'

  8. Reactor calculation benchmark PCA blind test results

    International Nuclear Information System (INIS)

    Kam, F.B.K.; Stallmann, F.W.

    1980-01-01

    Further improvement in calculational procedures or a combination of calculations and measurements is necessary to attain 10 to 15% (1 sigma) accuracy for neutron exposure parameters (flux greater than 0.1 MeV, flux greater than 1.0 MeV, and dpa). The calculational modeling of power reactors should be benchmarked in an actual LWR plant to provide final uncertainty estimates for end-of-life predictions and limitations for plant operations. 26 references, 14 figures, 6 tables

  9. Reactor calculation benchmark PCA blind test results

    Energy Technology Data Exchange (ETDEWEB)

    Kam, F.B.K.; Stallmann, F.W.

    1980-01-01

    Further improvement in calculational procedures or a combination of calculations and measurements is necessary to attain 10 to 15% (1 sigma) accuracy for neutron exposure parameters (flux greater than 0.1 MeV, flux greater than 1.0 MeV, and dpa). The calculational modeling of power reactors should be benchmarked in an actual LWR plant to provide final uncertainty estimates for end-of-life predictions and limitations for plant operations. 26 references, 14 figures, 6 tables.

  10. Review of fast reactor operating experience gained in 1998 in Russia. General trends of future fast reactor development

    International Nuclear Information System (INIS)

    Poplavski, V.M.; Ashurko, Y.M.; Zverev, K.V.; Sarayev, O.M.; Oshkanov, N.N.; Korol'kov, A.S.

    1999-01-01

    Review of the general state of nuclear power in Russia as for 1998 is given in brief in the paper. Results of operation of BR-10, BOR-60 and BN-600 fast reactors are presented as well as of scientific and technological escort of the BN-350 reactor. The paper outlines the current status and prospects of South-Urals and Beloyarskaya power unit projects with the BN-800 reactors. The main planned development trends on fast reactors are described concerning both new projects and R and D works. (author)

  11. 77 FR 48110 - Airworthiness Directives; General Electric Company Turbofan Engines

    Science.gov (United States)

    2012-08-13

    ... Company Turbofan Engines AGENCY: Federal Aviation Administration (FAA), DOT. ACTION: Notice of proposed... certain General Electric Company (GE) CF6-80C2 series turbofan engines. The existing AD requires... 2000-04-14, Amendment 39-11597 (65 FR 10698, February 29, 2000), for all GE CF6-80C2 series turbofan...

  12. 76 FR 64844 - Airworthiness Directives; General Electric Company Turbofan Engines

    Science.gov (United States)

    2011-10-19

    ... Company Turbofan Engines AGENCY: Federal Aviation Administration (FAA), DOT. ACTION: Notice of proposed... General Electric Company (GE) CF6-45 and CF6-50 series turbofan engines with certain low-pressure turbine... series turbofan engines with certain LPT rotor stage 3 disks installed. That AD requires initial and...

  13. 77 FR 58471 - Airworthiness Directives; General Electric Company Turbofan Engines

    Science.gov (United States)

    2012-09-21

    ... Airworthiness Directives; General Electric Company Turbofan Engines AGENCY: Federal Aviation Administration (FAA.../P1, GEnx-1B75/P1, GEnx- 2B67, and GEnx-2B67B turbofan engines. This AD requires initial and... this AD will affect 11 GE GEnx turbofan engines installed on airplanes of U.S. registry. We also...

  14. 77 FR 76977 - Airworthiness Directives; General Electric Company Turbofan Engines

    Science.gov (United States)

    2012-12-31

    ... Company Turbofan Engines AGENCY: Federal Aviation Administration (FAA), DOT. ACTION: Supplemental notice... proposed airworthiness directive (AD) for certain General Electric Company (GE) CF6-80C2 series turbofan... part 39 to include an AD that would apply to certain GE CF6-80C2 series turbofan engines. That NPRM...

  15. Remote maintenance of in-vessel components in Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Loesser, G.D.; Heitzenroeder, P.; Kungl, D.; Dylla, H.F.; Cerdan, G.

    1990-01-01

    The Tokamak Fusion Test Reactor (TFTR) will generate a total of 3 x 10 21 neutrons during its planned D-T operational period. A maintenance manipulator has been designed and tested to minimize personnel radiation during in-vessel maintenance activities. Its functions include visual inspection, first-wall tile replacement, cleaning, diagnostics calibrations and leak detection. To meet these objectives, the TFTR maintenance manipulator is required to be operable in the TFTR high vacuum environment, typically -8 torr, ( -6 Pa). Geometrically, the manipulator must extend 180 0 in either direction around the torus to assure complete coverage of the vessel first-wall. The manipulator consists of a movable carriage, and movable articulated link sections which are driven by electrical actuators. The boom has vertical load capacity of 455 kg and lateral load capacity of 46 kg. The boom can either be fitted with a general inspection arm or dextrous slave arms. The general inspection arm is designed to hold the leak detector and an inspection camera; it is capable of rotation along two axes and has a linkage system which permits motion normal to the vacuum vessel wall. All systems except the dextrous slave arms are operable in a vacuum. (author)

  16. Utilization of fission reactors for fusion engineering testing

    International Nuclear Information System (INIS)

    Deis, G.A.; Miller, L.G.

    1985-01-01

    Fission reactors can be used to conduct some of the fusion nuclear engineering tests identified in the FINESSE study. To further define the advantages and disadvantages of fission testing, the technical and programmatic constraints on this type of testing are discussed here. This paper presents and discusses eight key issues affecting fission utilization. Quantitative comparisons with projected fusion operation are made to determine the technical assets and limitations of fission testing. Capabilities of existing fission reactors are summarized and compared with technical needs. Conclusions are then presented on the areas where fission testing can be most useful

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

    International Nuclear Information System (INIS)

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

    1992-07-01

    This report describes the EG ampersand 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

  18. A design study of high electric power for fast reactor cooled by supercritical light water

    International Nuclear Information System (INIS)

    Koshizuka, Seiichi

    2000-03-01

    In order to evaluate the possibility to achieve high electric power by a fast reactor with supercritical light water, the design study was carried out on a large fast reactor core with high coolant outlet temperature (SCFR-H). Since the reactor coolant circuit uses once-through direct cycle where all feedwater flows through the core to the turbine at supercritical pressure, it is possible to design much simpler and more compact reactor systems and to achieve higher thermal efficiency than those of current light water reactors. The once-through direct cycle system is employed in current fossil-fired power plants. In the present study, three types of core were designed. The first is SCFR-H with blankets cooled by ascending flow, the second is SCFR-H with blankets cooled by descending flow and the third is SCFR-H with high thermal power. Every core was designed to achieve the thermal efficiency over 43%, positive coolant density reactivity coefficient and electric power over 1600 MW. Core characteristics of SCFR-Hs were compared with those of SCLWR-H (electric power: 1212 MW), which is a thermal neutron spectrum reactor cooled and moderated by supercritical light water, with the same diameter of the reactor pressure vessel. It was shown that SCFR-H could increase the electric power about 1.7 times maximally. From the standpoint of the increase of a reactor thermal power, a fast reactor has advantages as compared with a thermal neutron reactor, because it can increase the power density by adopting tight fuel lattices and eliminating the moderator region. Thus, it was concluded that a reactor cooled by supercritical light water could further improve the cost competitiveness by using a fast neutron spectrum and achieving a higher thermal power. (author)

  19. Conceptual design for simulator of irradiation test reactors

    International Nuclear Information System (INIS)

    Takemoto, Noriyuki; Ohto, Tsutomu; Magome, Hirokatsu; Izumo, Hironobu; Hori, Naohiko

    2012-03-01

    A simulator of irradiation test reactors has been developed since JFY 2010 for understanding reactor behavior and for upskilling in order to utilize a nuclear human resource development (HRD) and to promote partnership with developing countries which have a plan to introduce nuclear power plant. The simulator is designed based on the JMTR, one of the irradiation test reactors, and it simulates operation, irradiation tests and various kinds of accidents caused by the reactor and irradiation facility. The development of the simulator is sponsored by the Japanese government as one of the specialized projects of advanced research infrastructure in order to promote basic as well as applied researches. The training using the simulator will be started for the nuclear HRD from JFY 2012. This report summarizes the result of the conceptual design of the simulator in JFY 2010. (author)

  20. Analysis of design floor response spectra and testing of the electrical systems

    International Nuclear Information System (INIS)

    Ambriashvili, Y.

    1996-01-01

    This report covers the following activities as foreseen according to the working plan of 'Atmoenergoproject': analysis of calculated floor response spectra used during the design of Kozloduy NPP and comparison with other spectra recommended for this NPP; analysis of floor response spectrum for the most important systems (reactor, main coolant loop, electrical systems); tests of main electrical systems and analysis of the results on seismic stability of those systems. Results of the response spectra analysis are given, some of the electrical systems are identified by the Kozloduy authorities to be analyzed in future according to the results of the test on seismicity

  1. IEEE Std 383-1974: IEEE standard for type test of Class IE electric cables, field splices, and connections for nuclear power generating stations

    International Nuclear Information System (INIS)

    Anon.

    1992-01-01

    This standard provides direction for establishing type tests which may be used in qualifying Class 1E electric cables, field splices, and other connections for service in nuclear power generating stations. General guidelines for qualifications are given in IEEE Std 323-1974, Standard for Qualifying Class IE Electric Equipment for Nuclear Power Generating Stations. Categories of cables covered are those used for power control and instrumentation services. Though intended primarily to pertain to cable for field installation, this guide may also be used for the qualification of internal wiring of manufactured devices. This guide does not cover cables for service within the reactor vessel

  2. Education for university students, high school teachers and the general public using the Kinki University Reactor

    International Nuclear Information System (INIS)

    Tsuruta, T.

    2007-01-01

    Atomic Energy Research Institute of Kinki University is equipped with a nuclear reactor which is called UTR-KINKI. UTR is the abbreviation for University Teaching and Research Reactor. The reactor is the first one installed in Japanese universities. Though the reactor is owned and operated by Kinki University, its use is widely open to scientists and students from other universities and research institutions. The reactor is made the best of teaching instrument for the training of high school teachers. In addition, the reactor is utilized for general public education concerning atomic energy. (author)

  3. Research and development activities for reactor decommissioning. Developing technology of Fuji Electric Co., Ltd

    International Nuclear Information System (INIS)

    Shirakawa, Masahiro; Takaya, Jyunichi; Mizukoshi, Seiji; Hosoda, Hiroshi; Tomizuka, Chiaki; Funaguchi, Susumu; Ito, Katsuhito

    1997-01-01

    Fuji Electric Co., Ltd. is conducting decommissioning R and D for commercial reactor, especially for gas cooled reactor since the construction of the Tokai-1 power station of JAPCO, in the field of system engineering, residual radioactivity evaluation, dismantling of core internals, remote handling, treatment and disposal of radioactive waste, and radioactivity measurement. These R and D have been performed mainly under contract of JAPCO and JAERI. This paper gives a summary of the present status and future plan concerning technical development for decommissioning of nuclear reactor by Fuji Electric Co., Ltd. (author)

  4. Current and prospective fuel test programmes in the MIR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Izhutov, A.L.; Burukin, A.V.; Iljenko, S.A.; Ovchinnikov, V.A.; Shulimov, V.N.; Smirnov, V.P. [State Scientific Centre of Russia Research Institute of Atomic Reactors, Ulyanovsk region (Russian Federation)

    2007-07-01

    MIR reactor is a heterogeneous thermal reactor with a moderator and a reflector made of metal beryllium, it has a channel-type design and is placed in a water pool. MIR reactor is mainly designed for testing fragments of fuel elements and fuel assemblies (FA) of different nuclear power reactor types under normal (stationary and transient) operating conditions as well as emergency situations. At present six test loop facilities are being operated (2 PWR loops, 2 BWR loops and 2 steam coolant loops). The majority of current fuel tests is conducted for improving and upgrading the Russian PWR fuel, these tests involve issues such as: -) long term tests of short-size rods with different modifications of cladding materials and fuel pellets; -) further irradiation of power plant re-fabricated and full-size fuel rods up to achieving 80 MW*d/kg U; -) experiments with leaking fuel rods at different burnups and under transient conditions; -) continuation of the RAMP type experiments at high burnup of fuel; and -) in-pile tests with simulation of LOCA and RIA type accidents. Testing of the LEU (low enrichment uranium) research reactor fuel is conducted within the framework of the RERTR programme. Upgrading of the gas cooled and steam cooled loop facilities is scheduled for testing the HTGR fuel and sub-critical water-cooled reactor, correspondingly. The present paper describes the major programs of the WWER high burn-up fuel behavior study in the MIR reactor, capabilities of the applied techniques and some results of the performed irradiation tests. (authors)

  5. Thermal reactor benchmark tests on JENDL-2

    International Nuclear Information System (INIS)

    Takano, Hideki; Tsuchihashi, Keichiro; Yamane, Tsuyoshi; Akino, Fujiyoshi; Ishiguro, Yukio; Ido, Masaru.

    1983-11-01

    A group constant library for the thermal reactor standard nuclear design code system SRAC was produced by using the evaluated nuclear data JENDL-2. Furthermore, the group constants for 235 U were calculated also from ENDF/B-V. Thermal reactor benchmark calculations were performed using the produced group constant library. The selected benchmark cores are two water-moderated lattices (TRX-1 and 2), two heavy water-moderated cores (DCA and ETA-1), two graphite-moderated cores (SHE-8 and 13) and eight critical experiments for critical safety. The effective multiplication factors and lattice cell parameters were calculated and compared with the experimental values. The results are summarized as follows. (1) Effective multiplication factors: The results by JENDL-2 are considerably improved in comparison with ones by ENDF/B-IV. The best agreement is obtained by using JENDL-2 and ENDF/B-V (only 235 U) data. (2) Lattice cell parameters: For the rho 28 (the ratio of epithermal to thermal 238 U captures) and C* (the ratio of 238 U captures to 235 U fissions), the values calculated by JENDL-2 are in good agreement with the experimental values. The rho 28 (the ratio of 238 U to 235 U fissions) are overestimated as found also for the fast reactor benchmarks. The rho 02 (the ratio of epithermal to thermal 232 Th captures) calculated by JENDL-2 or ENDF/B-IV are considerably underestimated. The functions of the SRAC system have been continued to be extended according to the needs of its users. A brief description will be given, in Appendix B, to the extended parts of the SRAC system together with the input specification. (author)

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

    CERN Multimedia

    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).

  7. Reactor numerical simulation and hydraulic test research

    International Nuclear Information System (INIS)

    Yang, L. S.

    2009-01-01

    In recent years, the computer hardware was improved on the numerical simulation on flow field in the reactor. In our laboratory, we usually use the Pro/e or UG commercial software. After completed topology geometry, ICEM-CFD is used to get mesh for computation. Exact geometrical similarity is maintained between the main flow paths of the model and the prototype, with the exception of the core simulation design of the fuel assemblies. The drive line system is composed of drive mechanism, guide bush assembly, fuel assembly and control rod assembly, and fitted with the rod level indicator and drive mechanism power device

  8. Tokamak fusion test reactor. Final design report

    International Nuclear Information System (INIS)

    1978-08-01

    Detailed data are given for each of the following areas: (1) system requirements, (2) the tokamak system, (3) electrical power systems, (4) experimental area systems, (5) experimental complex, (6) neutral beam injection system, (7) diagnostic system, and (8) central instrumentation control and data acquisition system

  9. HFR irradiation testing of light water reactor (LWR) fuel

    International Nuclear Information System (INIS)

    Markgraf, J.F.W.

    1985-01-01

    For the materials testing reactor HFR some characteristic information with emphasis on LWR fuel rod testing capabilities and hot cell investigation is presented. Additionally a summary of LWR fuel irradiation programmes performed and forthcoming programmes are described. Project management information and a list of publications pertaining to LWR fuel rod test programmes is given

  10. SMORN-III benchmark test on reactor noise analysis methods

    International Nuclear Information System (INIS)

    Shinohara, Yoshikuni; Hirota, Jitsuya

    1984-02-01

    A computational benchmark test was performed in conjunction with the Third Specialists Meeting on Reactor Noise (SMORN-III) which was held in Tokyo, Japan in October 1981. This report summarizes the results of the test as well as the works made for preparation of the test. (author)

  11. Permeated defect detecting test method and device in reactor

    International Nuclear Information System (INIS)

    Sakurai, Yoshishige.

    1996-01-01

    The present invention provides a method of and a device capable of performing a test for entire inner surfaces of the reactor upon periodical inspection of a BWR type reactor while sufficiently taking countermeasures for radiation rays into consideration. Namely, the present invention comprises following steps. (1) A provisional step for taking a shroud head of a reactor core shroud and incore structural components above and below the shroud out of the reactor, discharging reactor water and water tightly closing openings such as reactor wall perforation holes, (2) a pretreatment step for washing exposed inner surfaces of the reactor and peeling deteriorated materials, (3) a first drying step for drying portions washed and peeled in the step (2), (4) a permeation step for applying a permeation liquid of a defect detecting medium on the exposed inner surfaces of the reactor, (5) a permeation liquid removing step for removing the an excess permeation liquid in the step (4), (6) a second drying step for drying corresponding portions after performing the step (5), and (7) a flaw detecting step for optically observing the corresponding portions after performing the step (6) and detecting flaws. (I.S.)

  12. General areas needing chemical competence to support reactor operation

    International Nuclear Information System (INIS)

    Proksch, E.; Bildstein, H.

    1963-01-01

    Chemical competence is needed not only for the development of new types of reactors but also for the start-up and safe operation of reactors. The activities of chemistry and chemical engineering cover a number of fields, namely chemical analysis, radiochemical analysis, corrosion research, radiolysis of water and water purification. The author reviews fields in reactor operation and maintenance in which chemical competence is needed. (author). 9 refs

  13. Multifrequency tests in the EBR-II reactor plant

    International Nuclear Information System (INIS)

    Feldman, E.E.; Mohr, D.; Gross, K.C.

    1989-01-01

    A series of eight multifrequency tests was conducted on the Experimental Breeder Reactor II. In half of the tests a control rod was oscillated and in the other half the controller input voltage to the intermediate-loop-sodium pump was perturbed. In each test the input disturbance consisted of several superimposed single-frequency sinusoidal harmonics of the same fundamental. The tests are described along with the theoretical and practical aspects of their development and design. Samples of measured frequency responses are also provided for both the reactor and the power plant. 22 refs., 5 figs., 2 tabs

  14. Possibilities for power reactor structural material and fuel testing in reactor RA; Mogucnosti reaktora RA za testiranje konstrukcionih materijala i goriva energetskih reaktora

    Energy Technology Data Exchange (ETDEWEB)

    Martinc, R; Lazarevic, Dj; Stefanovic, D; Cupac, S; Pesic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1978-05-15

    Nuclear reactor RA at Vinca has been designed as a high flux general purpose research reactor. Among other it was intended to play a role of material testing reactor. A scope of activities of Material Laboratory and Reactor RA Department of Boris Kidric Institute is presented in this report. Reactor RA capacity for reactor structural material and fuel irradiation is also described. The increase of RA reactor irradiation capacity is based on the improvement of VISA type fuel channel for fast neutron irradiations, as well as on the general neutron flux increase, due to introduction of highly enriched uranium fuel into reactor core and the advanced in-core fuel management. The irradiation capacities described allow for the reactor material and fuel testing to the considerable extent. Istrazivacki reaktor RA u Vinci je projektovan kao visokofluksni istrazivacki reaktor opste namene. Pored ostalog, on je namenjen i za testiranje reaktorskih konstrukcionih materijala i goriva. U radu je dat pregled aktivnosti Laboratorije za materijale IBK i reaktora RA na tom podrucju, kao i opis povecanih mogucnosti reaktora RA za ozracivanje reaktorskih materijala i goriva u cilju njihovog testiranja. Povecanje mogucnosti reaktora RA zasniva se na usavrsavanju specijalnog gorivnog kanala tipa VISA (za ozracivanje materijala brzim neutronima), kao i na opstem povecanju neutronskog fluksa na osnovu uvodjenja i nacina koriscenja visokoobogacenog uranskog goriva u reaktoru RA. Opisane mogucnosti reaktora RA dozvoljavaju u znatnoj meri ispitivanje konstrukcionih materijala i goriva energetskih reaktora.

  15. EBR-2 [Experimental Breeder Reactor-2], IFR [Integral Fast Reactor] prototype testing programs

    International Nuclear Information System (INIS)

    Lehto, W.K.; Sackett, J.I.; Lindsay, R.W.; Planchon, H.P.; Lambert, J.D.B.

    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

  16. EBR-2 [Experimental Breeder Reactor-2] test programs

    International Nuclear Information System (INIS)

    Sackett, J.I.; Lehto, W.K.; Lindsay, R.W.; Planchon, H.P.; Lambert, J.D.B.; Hill, D.J.

    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, advanced control system development, plant diagnostics development and component testing. This paper discusses EBR-2 as the IFR prototype and the associated testing programs. 29 refs

  17. Licensing experience of the HTR-10 test reactor

    International Nuclear Information System (INIS)

    Sun, Y.; Xu, Y.

    1996-01-01

    A 10MW high temperature gas-cooled test reactor (HTR-10) is now being projected by the Institute of Nuclear Energy Technology within China's National High Technology Programme. The Construction Permit of HTR-10 was issued by the Chinese nuclear licensing authority around the end of 1994 after a period of about one year of safety review of the reactor design. HTR-10 is the first high temperature gas-cooled reactor (HTGR) to be constructed in China. The purpose of this test reactor project is to test and demonstrate the technology and safety features of the advanced modular high temperature reactor design. The reactor uses spherical fuel elements with coated fuel particles. The reactor unit and the steam generator unit are arranged in a ''side-by-side'' way. Maximum fuel temperature under the accident condition of a complete loss of coolant is limited to values much lower than the safety limit set for the fuel element. Since the philosophy of the technical and safety design of HTR-10 comes from the high temperature modular reactor design, the reactor is also called the Test Module. HTR-10 represents among others also a licensing challenge. On the one side, it is the first helium reactor in China, and there are less licensing experiences both for the regulator and for the designer. On the other side, the reactor design incorporates many advanced design features in the direction of passive or inherent safety, and it is presently a world-wide issue how to treat properly the passive or inherent safety design features in the licensing safety review. In this presentation, the licensing criteria of HTR-10 are discussed. The organization and activities of the safety review for the construction permit licensing are described. Some of the main safety issues in the licensing procedure are addressed. Among these are, for example, fuel element behaviour, source term, safety classification of systems and components, containment design. The licensing experiences of HTR-10 are of

  18. Tritium experience in the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Skinner, C.H.; Blanchard, W.; Hosea, J.; Mueller, D.; Nagy, A.; Hogan, J.

    1998-01-01

    Tritium management is a key enabling element in fusion technology. Tritium fuel was used in 3.5 years of successful deuterium-tritium (D-T) operations in the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory. The D-T campaign enabled TFTR to explore the transport, alpha physics, and MHD stability of a reactor core. It also provided experience with tritium retention and removal that highlighted the importance of these issues in future D-T machines. In this paper, the authors summarize the tritium retention and removal experience in TFTR and its implications for future reactors

  19. Tests for validation of fast neutron reactors safety

    International Nuclear Information System (INIS)

    Nagata, T.; Yamashita, H.

    2001-01-01

    Japanese scientific research and design enterprises in cooperation with industrial and power generating corporations implement a project on creating a fast neutron reactor of the ultimate safety. One of the basic expected results from such a development is creation of a reactor core structure that is able to eliminate recriticality occurrence in the course of reactor accident involving fuel melting. One of the possible ways to solve this problem is to include pipes (meant for specifying directed (controlled) molten fuel relocation) into fuel assembly structure. In the course of conduction and subsequent implementation of such a design the basic issue is to experimentally confirm the operating capacity of FA having such a structure and that is called FAIDUS. Within EAGLE Project on experimental basis of IAE NNC RK an activity has been started on preparation and conduction of out-of-pile and in-pile tests. During tests a sodium coolant will be used. Studies are conducted by NNC RK in cooperation with the Japanese corporations JAPC and JNC. Basic objective of out-of-pile tests was to obtain preliminary information on fuel relocation behavior under conditions simulating accident involving melting of core consisting of FAIDUS FA, which will help to clarify simulation criteria and to develop the most optimum structure of the experimental channel for reactor experiments conduction. The basic objective of in-pile tests was the experimental confirmation of operating capacity of FAIDUS FA model under reactor conditions. According to the program two tests are planned to be performed at IGR reactor: tests for validation of fast neutron reactor safety, and out-of-pile tests at EAGLE experimental facility without sodium coolant

  20. System Description of the Electrical Power Supply System for the ATLAS Integral Test Loop

    International Nuclear Information System (INIS)

    Moon, S. K.; Park, J. K.; Kim, Y. S.; Song, C. H.; Baek, W. P.

    2007-02-01

    An integral effect test loop for pressurized water reactors (PWRs), the ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation), is constructed by Thermal-Hydraulics Safety Research Team in Korea Atomic Energy Research Institute (KAERI). The ATLAS facility has been designed to have the length scale of 1/2 and area scale of 1/144 compared with the reference plant, APR1400. This report describes the design and technical specifications of the electrical power supply system which supplies the electrical powers to core heater rods, other heaters, various pumps and other systems. The electrical power supply system had acquired the final approval on the operation from the Korea Electrical Safety Corporation. During performance tests for the operation and control, the electrical power supply system showed completely acceptable operation and control performance

  1. Nuclear electric capacity expansion in Mexico: system effects of reactor size and cost

    International Nuclear Information System (INIS)

    Thayer, G.R.; Abbey, D.S.; Hardie, R.W.; Enriquez, R.P.; Uria, E.G.

    1984-01-01

    Mexico's electrical generation capacity could more than double over the next ten years - from about 15 GWe currently to as much as 35 GWe in 1990. While new capacity additions will be predominantly oil-fired in the 1980's, nuclear power will become increasingly important in the 1990's. This study investigated the appropriate size of new, nuclear capacity additions by assessing the implications of installing different size reactors into Mexico's electrical grid. Included in the assessments of reactor sizes are estimates of electrical generation costs and comparisons of the effective load-carrying capability of a 10 GWe nuclear capacity expansion

  2. Strain measurement in and analysis for hydraulic test of CPR1000 reactor pressure vessel

    International Nuclear Information System (INIS)

    Zhou Dan; Zhuang Dongzhen

    2013-01-01

    The strain measurement in hydraulic test of CPR1000 reactor pressure vessel performed in Dongfang Heavy Machinery Co., Ltd. is introduced. The detail test scheme and method was introduced and the measurement results of strain and stress was given. Meanwhile the finite element analysis was performed for the pressure vessel, which was generally matched with the measurement results. The reliability of strain measurement was verified and the high strength margin of vessel was shown, which would give a good reference value for the follow-up hydraulic tests and strength analysis of reactor pressure vessel. (authors)

  3. SP-100 reactor disassembly remote handling test program

    International Nuclear Information System (INIS)

    Wilson, C.E.; Potter, J.D.; Maiden, G.E.; Vader, D.P.

    1991-01-01

    This paper is presented as an overview of the remote handling equipment validation testing, which will be conducted before installation and use in the ground engineering test facility. This equipment will be used to defuel the SP-100 reactor core after removing it from the Test Assembly following nuclear testing. A series of full scale mock-up operational tests will be conducted at a Hanford Site facility to verify equipment design, operation, and capabilities

  4. Small Modular Reactors for Countries with Small to Medium Electric Grids - An Economically Sensible Solution

    International Nuclear Information System (INIS)

    Young, P.

    2012-01-01

    There has recently been a renewed interest throughout the world in small nuclear units for generating electricity and for other applications. A report by the World Nuclear Association discussing the advantages of small modular nuclear reactors (SMRs) over traditional nuclear reactor designs, states that ''modern small reactors for power generation are expected to have greater simplicity of design, economy of mass production, and reduced siting costs. Many are also designed for a high level of passive or inherent safety in the event of malfunction.'' Since the inception of nuclear power, the size of reactor units has grown from under 100 MWe to more than 1600 MWe. Today, due partly to the high capital cost of large power reactors and partly to the need to service small electricity grids, there is a move to develop smaller units. These may be built individually or as modules in a larger plant. SMRs are a good fit in markets where anticipated electricity demand is projected to increase incrementally, because SMRs could be built in series as needed. SMRs might be particularly attractive in countries that currently rely on diesel generators for producing electricity. Small reactors could make economic sense because of the high cost of diesel generation compared to the low marginal cost of producing electricity from nuclear energy. (Keeping in mind the initial investment costs and the need to establish a national regulatory program.) Some SMR designs are fabricated in a factory and then delivered to the site. This could be a solution for markets that lack the qualified engineers and skilled craft workers needed to construct large reactors on site. This paper will provide an overview of the types and attributes of SMRs in use or under development worldwide, describe the similarities and important differences between designs, discuss potential applications for SMRs, including baseload electricity generation, electricity generation for remote locations and areas with

  5. 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.

  6. Integral test of JENDL-3.3 for fast reactors

    International Nuclear Information System (INIS)

    Chiba, Gou

    2003-01-01

    An integral test of JENDL-3.3 was performed for fast reactors. Various types of fast reactors were analyzed. Calculation values of the nuclear characteristics were greatly especially affected by the revisions of the cross sections of U-235 capture and elastic scattering reactions. The C/E values were improved for ZPPR cross where plutonium is mainly fueled, but not for BFS cores where uranium is mainly fueled. (author)

  7. Research on the Reliability Testing of Electrical Automation Control Equipment

    OpenAIRE

    Yongjie Luo

    2014-01-01

    According to the author’s many years’ work experience, this paper first discusses the concepts of electrical automation control equipment reliability testing, and then analyzes the test method of electrical automation control equipment reliability testing, finally, on this basis, this article discusses how to determine the reliability test method of electrical automation control equipment. Results of this study will provide a useful reference for electrical automation control equipment reliab...

  8. Heat removal performance of auxiliary cooling system for the high temperature engineering test reactor during scrams

    International Nuclear Information System (INIS)

    Takeda, Takeshi; Tachibana, Yukio; Iyoku, Tatsuo; Takenaka, Satsuki

    2003-01-01

    The auxiliary cooling system of the high temperature engineering test reactor (HTTR) is employed for heat removal as an engineered safety feature when the reactor scrams in an accident when forced circulation can cool the core. The HTTR is the first high temperature gas-cooled reactor in Japan with reactor outlet gas temperature of 950 degree sign C and thermal power of 30 MW. The auxiliary cooling system should cool the core continuously avoiding excessive cold shock to core graphite components and water boiling of itself. Simulation tests on manual trip from 9 MW operation and on loss of off-site electric power from 15 MW operation were carried out in the rise-to-power test up to 20 MW of the HTTR. Heat removal characteristics of the auxiliary cooling system were examined by the tests. Empirical correlations of overall heat transfer coefficients were acquired for a helium/water heat exchanger and air cooler for the auxiliary cooling system. Temperatures of fluids in the auxiliary cooling system were predicted on a scram event from 30 MW operation at 950 degree sign C of the reactor outlet coolant temperature. Under the predicted helium condition of the auxiliary cooling system, integrity of fuel blocks among the core graphite components was investigated by stress analysis. Evaluation results showed that overcooling to the core graphite components and boiling of water in the auxiliary cooling system should be prevented where open area condition of louvers in the air cooler is the full open

  9. TIBER engineering test reactor (ETR) startup scenarios

    International Nuclear Information System (INIS)

    Blackfield, D.T.; Perkins, L.J.

    1987-01-01

    A time-dependent Tokamak Systems Code (TTSC) has been developed and used to examine various inductively driven startup scenarios for the TIBER reactor. Radially averaged particle and energy balance equations are solved. In addition, time varying currents in the PF and OH coils are determined from MHD equilibrium and volt-seconds considerations. Less than 20 MW of auxiliary power deposited in the electrons is required to obtain steady-state operations. For this scenario, less than 10% of the total volt-seconds capability is consumed during startup and the currents in the PF and OH coils do not appear to exceed stress limits. For every volt-second saved during startup, the burn time can be extended 14 seconds. 4 refs., 6 figs., 3 tabs

  10. Hybrid Electric Vehicle Testing | Transportation Research | NREL

    Science.gov (United States)

    Hybrid Electric Vehicle Evaluations Hybrid Electric Vehicle Evaluations How Hybrid Electric Vehicles Work Hybrid electric vehicles combine a primary power source, an energy storage system, and an is used to propel the vehicle during normal drive cycles. The batteries supply additional power for

  11. Loss-of-Fluid Test findings in pressurized water reactor core's thermal-hydraulic behavior

    International Nuclear Information System (INIS)

    Russell, M.

    1983-01-01

    This paper summarizes the pressurized water reactor (PWR) core's thermal-hydraulic behavior findings from experiments performed at the Loss-of-Fluid Test (LOFT) Facility at the Idaho National Engineering Laboratory. The potential impact of these findings on the safety and economics of PWR's generation of electricity is also discussed. Reviews of eight important findings in the core's physical behavior and in experimental methods are presented with supporting evidence

  12. Processing test of an upgraded mechanical design for PERMCAT reactor

    International Nuclear Information System (INIS)

    Borgognoni, Fabio; Demange, David; Doerr, Lothar; Tosti, Silvano; Welte, Stefan

    2010-01-01

    The PERMCAT membrane reactor is a coaxial combination of a Pd/Ag permeator membrane and a catalyst bed. This device has been proposed for processing fusion reactor plasma exhaust gas. A stream containing tritium (up to 1% of tritium in different chemical forms such as water, methane or molecular hydrogen) is decontaminated in the PERMCAT by counter-current isotopic swamping with protium. Different mechanical designs of the membrane reactor have been proposed to improve robustness and lifetime. The ENEA membrane reactor uses a permeator tube with a length of about 500 mm produced via cold-rolling and diffusion welding of Pd/Ag thin foils: two stainless steel pre-tensioned bellows have been applied to the Pd/Ag tube in order to avoid any significant compressive and bending stresses due to the permeator tube elongation consequent to the hydrogen uptake. An experimental test campaign has been performed using this reactor in order to assess the influence of different operating parameters and to evaluate the overall performance (decontamination factor). Tests have been carried out on two reactor prototypes: a defect-free membrane with complete (infinite) hydrogen selectivity and not perm-selective membrane. In this last case, the study has been aimed at verifying the behaviour of the PERMCAT devices under non-normal (accidental) conditions in the view of providing information for future safety analysis. The paper will present the specific mechanical design and the experimental results of tests based on isotopic exchange between H 2 O and D 2 .

  13. Testing of the multi-application small light water reactor (MASLWR) passive safety systems

    International Nuclear Information System (INIS)

    Reyes, Jose N.; Groome, John; Woods, Brian G.; Young, Eric; Abel, Kent; Yao, You; Yoo, Yeon Jong

    2007-01-01

    Experimental thermal hydraulic research has been conducted at Oregon State University for the purpose of assessing the performance of a new reactor design concept, the multi-application small light water reactor (MASLWR). The MASLWR is a pressurized light water reactor design with a net output of 35 MWe that uses natural circulation in both normal and transient operation. Due to its small size, portability and modularity, the MASLWR design is well suited to help fill the potential need for grid appropriate reactor designs for smaller electricity grids as may be found in developing or remote regions. The purpose of the OSU MASLWR test facility is to assess the operation of the MASLWR under normal full operating pressure and full temperature conditions and to assess the passive safety systems under transient conditions. The data generated by the testing program will be used to assess computer code calculations and to provide a better understanding of the thermal-hydraulic phenomena in the design of the MASLWR NSSS. During this testing program, four tests were conducted at the OSU MASLWR test facility. These tests included one design basis accident and one beyond design basis accident. During the performance of these tests, plant operations to include start up, normal operation and shut down evolutions were demonstrated successfully

  14. Black hole based tests of general relativity

    International Nuclear Information System (INIS)

    Yagi, Kent; Stein, Leo C

    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 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. (paper)

  15. Testing of a transport cask for research reactor spent fuel

    International Nuclear Information System (INIS)

    Mourao, Rogerio P.; Silva, Luiz Leite da; Miranda, Carlos A.; Mattar Neto, Miguel; Quintana, Jose F.A.; Saliba, Roberto O.; Novara, Oscar E.

    2011-01-01

    Since the beginning of the last decade three Latin American countries which operate research reactors - Argentina, Brazil and Chile - have been joining efforts to improve the regional capability in the management of spent fuel elements from the reactors operated in the region. As a step in this direction, a packaging for the transport of irradiated fuel from research reactors was designed by a tri-national team and a half-scale model for MTR fuel constructed in Argentina and tested in Brazil. Two test campaigns have been carried out so far, covering both normal conditions of transportation and hypothetical accident conditions. Although the specimen has not successfully performed the tests, its overall performance was considered very satisfactory, and improvements are being introduced to the design. A third test sequence is planned for 2011. (author)

  16. Radwaste volume reduction and solidification by General Electric

    International Nuclear Information System (INIS)

    Green, T.A.; Weech, M.E.; Miller, G.P.; Eberle, J.W.

    1982-01-01

    Since 1978 General Electric has been actively engaged in developing a volume reduction and solidifcation system or treatment of radwaste generated in commercial nuclear power plants. The studies have been aimed at defining an integrated system that would be directly responsive to the rapid evolving needs of the industry for the volume reduction and solidification of low-level radwaste. The resulting General Electric Volume Reduction System (GEVRS) is an integrated system based on two processes: the first uses azeotropic distillation technology and is called AZTECH, and the second is controlled-air incineration...called INCA. The AZTECH process serves to remove water from concentrated salt solutions, ion exchange resins and filter sludge slurries and then encapsulates the dried solids into a dense plastic product. The INCA unit serves to reduce combustible wastes to ashes suitable for encapsulation into the same plastic product produced by AZTECH

  17. Ageing management practice in Fast Breeder Test Reactor

    International Nuclear Information System (INIS)

    Srinivasan, G.; Ramanathan, V.; Swaminathan, P.R.; Babu, A.; Rajasekarappa, E.; Rajendran, B.; Ramalingam, P.V.

    2006-01-01

    Fast Breeder Test Reactor is a 40 MWt, sodium cooled, PuC-UC fuelled fast reactor, located at Kalpakkam, India. The reactor went critical in October 85 with Mark I core rated for 10.5 MWt at a peak LHR of 320 W/cm. The reactor core was progressively enlarged and TG was synchronized to the grid in July 97. The present core has 41 fuel subassemblies rated for 15.7 MWt at a peak LHR of 320 W/cm. The reactor has so far been operated for 33000 h and has seen 660 EFPD of operation corresponding to peak LHR of 320 W/cm. The peak burnup reached by the carbide fuel is 127 GWd/t, without any fuel clad failure. The four sodium pumps have been operating satisfactorily for a cumulative time of more than 5,00,000 h. Creep, fatigue and fluence govern the life of the nuclear systems. Because of the reduced power and temperature at which the reactor has so far been operated, there is little ageing of the nuclear systems. The life of the nuclear components is being monitored by periodic surveillance. Periodic assessment of the fluence seen by reactor components is being made. The conventional systems have been in service for the past 19 years. Civil structures are 25 years old. These have been maintained by periodic preventive maintenance and replacement / repair wherever required. This paper details the various ageing management practices in FBTR. (author)

  18. Automation of Electrical Cable Harnesses Testing

    Directory of Open Access Journals (Sweden)

    Zhuming Bi

    2017-12-01

    Full Text Available Traditional automated systems, such as industrial robots, are applied in well-structured environments, and many automated systems have a limited adaptability to deal with complexity and uncertainty; therefore, the applications of industrial robots in small- and medium-sized enterprises (SMEs are very limited. The majority of manual operations in SMEs are too complicated for automation. The rapidly developed information technologies (IT has brought new opportunities for the automation of manufacturing and assembly processes in the ill-structured environments. Note that an automation solution should be designed to meet the given requirements of the specified application, and it differs from one application to another. In this paper, we look into the feasibility of automated testing for electric cable harnesses, and our focus is on some of the generic strategies for the improvement of the adaptability of automation solutions. Especially, the concept of modularization is adopted in developing hardware and software to maximize system adaptability in testing a wide scope of products. A proposed system has been implemented, and the system performances have been evaluated by executing tests on actual products. The testing experiments have shown that the automated system outperformed manual operations greatly in terms of cost-saving, productivity and reliability. Due to the potential of increasing system adaptability and cost reduction, the presented work has its theoretical and practical significance for an extension for other automation solutions in SMEs.

  19. Achieving salt-cooled reactor goals: economics, variable electricity, no major fuel failures - 15118

    International Nuclear Information System (INIS)

    Forsberg, C.

    2015-01-01

    The Fluoride-salt-cooled High-temperature Reactor (FHR) with a Nuclear air-Brayton Combined Cycle (NACC) and Firebrick Resistance-Heated Energy Storage (FIRES) is a new reactor concept. The FHR uses High-Temperature Gas-cooled Reactor (HTGR) coated-particle fuel and liquid-salt coolants originally developed for molten salt reactors (MSRs) where the fuel was dissolved in the coolant. The FIRES system consists of high-temperature firebrick heated to high temperatures with electricity at times of low electric prices. For a modular FHR operating with a base-load 100 MWe output, the station output can vary from -242 MWe to +242 MWe. The FHR can be built in different sizes. The reactor concept was developed using a top-down approach: markets, requirements, reactor design. The goals are: (1) increase plant revenue by 50 to 100% relative to base-load nuclear plants with capital costs similar to light-water reactors, (2) enable a zero-carbon nuclear renewable electricity grid, and (3) no potential for major fuel failure and thus no potential for major radionuclide offsite releases in a beyond-design-basis accident (BDBA). The basis for the goals and how they may be achieved is described

  20. Supply of appropriate nuclear technology for the developing world: small power reactors for electricity generation

    International Nuclear Information System (INIS)

    Heising-Goodman, C.D.

    1981-01-01

    This paper reviews the supply of small nuclear power plants (200 to 500 MWe electrical generating capacity) available on today's market, including the pre-fabricated designs of the United Kingdom's Rolls Royce Ltd and the French Alsthom-Atlantique Company. Also, the Russian VVER-440 conventionally built light-water reactor design is reviewed, including information on the Soviet Union's plans for expansion of its reactor-building capacity. A section of the paper also explores the characteristics of LDC electricity grids, reviewing methods available for incorporating larger plants into smaller grids as the Israelis are planning. Future trends in reactor supply and effects on proliferation rates are also discussed, reviewing the potential of the Indian 220 MWe pressurised heavy-water reactor, South Korean and Jananese potential for reactor exports in the Far East, and the Argentine-Brazilian nuclear programme in Latin America. This study suggests that small reactor designs for electrical power production and other applications, such as seawater desalination, can be made economical relative to diesel technology if traditional scaling laws can be altered by adopting and standardising a pre-fabricated nuclear power plant design. Also, economy can be gained if sufficient attention is concentrated on the design, construction and operating experience of suitably sized conventionally built reactor systems. (author)

  1. Reduced enrichment for research and test reactors: Proceedings

    International Nuclear Information System (INIS)

    1993-07-01

    The 15th annual Reduced Enrichment for Research and Test Reactors (RERTR) international meeting was organized by Ris oe 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

  2. 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.

  3. Testing General Relativity with Pulsar Timing

    Directory of Open Access Journals (Sweden)

    Stairs Ingrid H.

    2003-01-01

    Full Text Available Pulsars of very different types, including isolated objects and binaries (with short- and long-period orbits, and white-dwarf and neutron-star companions provide the means to test both the predictions of general relativity and the viability of alternate theories of gravity. This article presents an overview of pulsars, then discusses the current status of and future prospects for tests of equivalence-principle violations and strong-field gravitational experiments.

  4. Proving Test on the Reliability for Reactor Containment Vessel

    International Nuclear Information System (INIS)

    Takumi, K.; Nonaka, A.

    1988-01-01

    NUPEC (Nuclear Power Engineering Test Center) has started an eight-year project of Proving Test on the Reliability for Reactor Containment Vessel since June 1987. The objective of this project is to confirm the integrity of containment vessels under severe accident conditions. This paper shows the outline of this project. The test Items are (1) Hydrogen mixing and distribution test, (2) Hydrogen burning test, (3) Iodine trapping characteristics test, and (4) Structural behavior test. Based on the test results, computer codes are verified and as the results of analysis and evaluation by the computer codes, containment integrity is to be confirmed

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

    International Nuclear Information System (INIS)

    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

  6. Performance tests of the reactor containment structures of HTTR

    International Nuclear Information System (INIS)

    Sakaba, Nariaki; Iigaki, Kazuhiko; Kawaji, Satoshi; Iyoku, Tatsuo

    1998-03-01

    The containment structures of the HTTR consist of the reactor containment vessel (CV), service area (SA) and emergency air purification system, which minimize the release of FPs in the postulated accidents with FP release from the reactor facilities. The CV is designed to withstand the temperature and pressure transients and to be leak-tight within the specified leakage limit even in the case of a rupture of the primary concentric hot gas duct. The pressure of inside of the SA should be maintained slightly lower than that of atmosphere by the emergency air purification system. The radioactive materials are released from the stack to environment via the emergency air purification system under the accident condition. Then the emergency air purification system should remove airborne radio-activities and should maintain proper pressure in the SA. We established the method to measure leak rate of the CV with closed reactor coolant pressure boundary although it is normally measured under opened reactor coolant pressure boundary as employed in LWRs. The CV leak rate test was carried out by the newly developed method and the expected performance was obtained. The SA and emergency air purification system were also confirmed by the performance test. We concluded that the reactor containment structures were fabricated to minimize the release of FPs in the postulated accidents with FP release from the reactor facilities. (author)

  7. Preliminary design studies on the Broad Application Test Reactor

    International Nuclear Information System (INIS)

    Terry, W.J.; Terry, W.K.; Ryskamp, J.M.; Jahshan, S.N.; Fletcher, C.D.; Moore, R.L.; Leyse, C.F.; Ottewitte, E.H.; Motloch, C.G.; Lacy, J.M.

    1992-08-01

    This report describes progress made at the Idaho National Engineering Laboratory during the first three quarters of Fiscal Year (FY) 1992 on the Laboratory-Directed Research and Development (LDRD) project to perform preliminary design studies on the Broad Application Test Reactor (BATR). This work builds on the FY-92 BATR studies, which identified anticipated mission and safety requirements for BATR and assessed a variety of reactor concepts for their potential capability to meet those requirements. The main accomplishment of the FY-92 BATR program is the development of baseline reactor configurations for the two conventional conceptual test reactors recommended in the FY-91 report. Much of the present report consists of descriptions and neutronics and thermohydraulics analyses of these baseline configurations. In addition, we considered reactor safety issues, compared the consequences of steam explosions for alternative conventional fuel types, explored a Molten Chloride Fast Reactor concept as an alternate BATR design, and examined strategies for the reduction of operating costs. Work planned for the last quarter of FY-92 is discussed, and recommendations for future work are also presented

  8. Inductive testing of reactor pressure vessels

    International Nuclear Information System (INIS)

    Bergh, H.

    1987-01-01

    In Service Inspection of Reactor Pressure Vessels is mostly done with ultrasonics. Using special 2 crystal-probes good detectability is achieved for near surface defects. The problem is to detect closely spaced cracks, to decide if the defects are surface braking and, if not, to decide the remaining ligament. The purpose of this study is to investigate to what extent Eddy Current can solve these problems. Detecting surfacebreaking cracks and fields of cracks can be done using conventional Eddy Current techniques. Mapping of closely spaced cracks requires a small probe and a high frequency. Measurement of depths a larger probe, a lower frequency and knowledge of the crackfield since 2 closely spaced shallow cracks might be mistaken for one deep crack. Depths of singel cracks can be measured down to 7-8 mm. In closely spaced crackfields the depths can not be measured. The measurement is mostly based on amplitude. For not surface breaking defects the problem is to decide the ligament, i.e. the distance between surface and cracktip. To achieve good penetration a large probe, low frequency and high energy or pulsed energy is used. Ligament up to 4 mm can be measured with good accuracy. The measurements is mostly based on phase. Noise, which originates from rough surface, varied material structure and lift off, can be reduced using multi frequency mix, probe design and scanning pattern. (author)

  9. Chinese nuclear heating test reactor and demonstration plant

    International Nuclear Information System (INIS)

    Wang Dazhong; Ma Changwen; Dong Duo; Lin Jiagui

    1992-01-01

    In this report the importance of nuclear district heating is discussed. From the viewpoint of environmental protection, uses of energy resources and transport, the development of nuclear heating in China is necessary. The development program of district nuclear heating in China is given in the report. At the time being, commissioning of the 5 MW Test Heating Reactor is going on. A 200 MWt Demonstration Plant will be built. In this report, the main characteristics of these reactors are given. It shows this type of reactor has a high inherent safety. Further the report points out that for this type of reactor the stability is very important. Some experimental results of the driving facility are included in the report. (orig.)

  10. The Test Reactor Embrittlement Data Base (TR-EDB)

    International Nuclear Information System (INIS)

    Stallmann, F.W.; Kam, F.B.K.; Wang, J.A.

    1993-01-01

    The Test Reactor Embrittlement Data Base (TR-EDB) is part of an ongoing program to collect test data from materials irradiations to aid in the research and evaluation of embrittlement prediction models that are used to assure the safety of pressure vessels in power reactors. This program is being funded by the US Nuclear Regulatory Commission (NRC) and has resulted in the publication of the Power Reactor Embrittlement Data Base (PR-EDB) whose second version is currently being released. The TR-EDB is a compatible collection of data from experiments in materials test reactors. These data contain information that is not obtainable from surveillance results, especially, about the effects of annealing after irradiation. Other information that is only available from test reactors is the influence of fluence rates and irradiation temperatures on radiation embrittlement. The first version of the TR-EDB will be released in fall of 1993 and contains published results from laboratories in many countries. Data collection will continue and further updates will be published

  11. 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…

  12. Models for transient analyses in advanced test reactors

    International Nuclear Information System (INIS)

    Gabrielli, Fabrizio

    2011-01-01

    Several strategies are developed worldwide to respond to the world's increasing demand for electricity. Modern nuclear facilities are under construction or in the planning phase. In parallel, advanced nuclear reactor concepts are being developed to achieve sustainability, minimize waste, and ensure uranium resources. To optimize the performance of components (fuels and structures) of these systems, significant efforts are under way to design new Material Test Reactors facilities in Europe which employ water as a coolant. Safety provisions and the analyses of severe accidents are key points in the determination of sound designs. In this frame, the SIMMER multiphysics code systems is a very attractive tool as it can simulate transients and phenomena within and beyond the design basis in a tightly coupled way. This thesis is primarily focused upon the extension of the SIMMER multigroup cross-sections processing scheme (based on the Bondarenko method) for a proper heterogeneity treatment in the analyses of water-cooled thermal neutron systems. Since the SIMMER code was originally developed for liquid metal-cooled fast reactors analyses, the effect of heterogeneity had been neglected. As a result, the application of the code to water-cooled systems leads to a significant overestimation of the reactivity feedbacks and in turn to non-conservative results. To treat the heterogeneity, the multigroup cross-sections should be computed by properly taking account of the resonance self-shielding effects and the fine intra-cell flux distribution in space group-wise. In this thesis, significant improvements of the SIMMER cross-section processing scheme are described. A new formulation of the background cross-section, based on the Bell and Wigner correlations, is introduced and pre-calculated reduction factors (Effective Mean Chord Lengths) are used to take proper account of the resonance self-shielding effects of non-fuel isotopes. Moreover, pre-calculated parameters are applied

  13. An Electric Field Test Using the MRI

    Czech Academy of Sciences Publication Activity Database

    Fiala, P.; Bartušek, Karel

    2008-01-01

    Roč. 4, č. 7 (2008), s. 701-705 ISSN 1931-7360 Institutional research plan: CEZ:AV0Z20650511 Keywords : MRI * electric field * numerical modeling Subject RIV: JA - Electronics ; Optoelectronics, Electrical Engineering

  14. Vibration tests on some models of PEC reactor core elements

    International Nuclear Information System (INIS)

    Bonacina, G.; Castoldi, A.; Zola, M.; Cecchini, F.; Martelli, A.; Vincenzi, D.

    1982-01-01

    This paper describes the aims of the experimental tests carried out at ISMES, within an agreement with the Department of Fast Reactors of ENEA, on some models of the elements of PEC Fast Nuclear Reactor Core in the frame of the activities for the seismic verification of the PEC core. The seismic verification is briefly described with particular attention to the problems arising from the shocks among the various elements during an earthquake, as well as the computer code used, the purpose and the techniques used to perform tests, some results and the first comparison between the theory and the experimental data

  15. EMERIS: an advanced information system for a materials testing reactor

    International Nuclear Information System (INIS)

    Adorjan, F.; Buerger, L.; Lux, I.; Mesko, L.; Szabo, K.; Vegh, J.; Ivanov, V.V.; Mozhaev, A.A.; Yakovlev, V.V.

    1990-06-01

    The basic features of the Materials Testing Reactor of IAE, Moscow (MR) Information System (EMERIS) are outlined. The purpose of the system is to support reactor and experimental test loop operators by a flexible, fully computerized and user-friendly tool for the aquisition, analysis, archivation and presentation of data obtained during operation of the experimental facility. High availability of EMERIS services is ensured by redundant hardware and software components, and by automatic configuration procedure. A novel software feature of the system is the automatic Disturbance Analysis package, which is aimed to discover primary causes of irregularities occurred in the technology. (author) 2 refs.; 2 figs

  16. Progress on qualification testing methodology study of electric cables

    International Nuclear Information System (INIS)

    Yoshida, K.; Seguchi, T.; Okada, S.; Ito, M.; Kusama, Y.; Yagi, T.; Yoshikawa, M.

    1983-01-01

    Many instrumental, control and power cables are installed in nuclear power plants, and these cables contain a large amount of organic polymers as insulating and jacketing materials. They are exposed to radiation at high dose rate, steam at high temperature and chemical (or water) spray simultaneously when a LOCA occurs. Under such conditions, the polymers tend to lose their original properties. For reactor safety, the cables should be functional even if they are subjected to a loss-of-coolant accident (LOCA) at the end of their intended service life. In Japan, cable manufacturers qualify their cables according to the proposed test standard issued from IEEJ in 1982, but the standard still has many unsolved problems or uncertainties which have been dealt with tentatively through the manufacturer-user's agreement. The objectives of this research are to study the methodologies for qualification testing of electric wires and cables, and to provide the improved technical bases for modification of the standard. Research activities are divided into the Accident (LOCA) Testing Methodology and the Accelerated Aging Methodology

  17. 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.

  18. Improving the proliferation resistance of research and test reactors

    International Nuclear Information System (INIS)

    Lewis, R.A.

    1978-01-01

    Elimination, or substantial reduction, of the trade in unirradiated highly-enriched fuel elements for research and test reactors would significantly reduce the proliferation risk associated with the current potential for diversion of these materials. To this end, it is the long-term goal of U.S. policy to fuel all new and existing research and test reactors with uranium of less-than-20% enrichment (but substantially greater than natural) excepting, perhaps, only a small number of high-power, high-performance, reactors. The U.S. development program for enrichment reduction in research and test reactor designs currently using 90-93% enriched uranium is based on the practical criterion that enrichment reduction should not cause significant flux performance (flux per unit power) or burnup performance degradation relative to the unmodified reactor design. To first order, this implies the requirement that the 235 U loading in the reduced-enrichment fuel elements be the same as the 235 U loading in the 90-93% enriched fuel elements. This can be accomplished by substitution of higher uranium density fuel technology for currently-used fuel technology in the fuel meat volume of the current fuel element design and/or by increasing the usable fuel meat volume. For research and test reactors of power greater than 5-10 megawatts, fuel technology does not currently exist that would permit enrichment reductions to below 20% utilizing this criterion. A program is now beginning in the U.S. to develop the necessary fuel technology. Currently-proven fuel technology is capable, however, of accommodating enrichment reductions to the 30-45% range (from 90-93%) for many reactors in the 5-50MW range. Accordingly the U.S. is proposing to convert existing reactors (and new designs) in the 5-50MW range from the use of highly-enriched fuel to the use of 30-45% enriched fuel, and reactors of less that about 5MW to less-than-20% enrichment, wherever this can be done without significant

  19. Wiring installation for electric devices above the roof slab of a nuclear reactor

    International Nuclear Information System (INIS)

    Jahnke, S.

    1986-01-01

    The wiring installation is situated inside the nuclear reactor building. It includes, associated to electric devices, a first cable which extends from the device to a fixed connector arranged above the cover. A second cable is connected to the said fixed connector and to a connector fixed on a plate situated out of the reactor. According to the present invention each second cable has several sections. A first section can be connected to the said fixed connector situated above the cover and to a fixed lead-in connector of a fluid-tight conduit above the reactor core. A second section is inside the conduit. A third section can be connected to a lead-out connector fixed on the plate which is out of the reactor. The invention applies more particularly to pressurized water nuclear reactors [fr

  20. The Integral Fast Reactor concept: Today's hope for tomorrow's electrical energy needs

    International Nuclear Information System (INIS)

    Dwight, C.C.; Phipps, R.D.

    1989-01-01

    Acid rain and the greenhouse effect are getting more attention as their impacts on the environment become evident around the world. Substantial evidence indicates that fossil fuel combustion for electrical energy production activities is a key cause of those problems. A change in electrical energy production policy is essential to a stable, healthy environment. That change is inevitable, it's just a matter of when and at what cost. Vision now, instead of reaction later, both in technological development and public perception, will help to limit the costs of change. The Integral Fast Reactor (IFR) is a visionary concept developed by Argonne National Laboratory that involves electrical energy production through fissioning of heavy metals by fast neutrons in a reactor cooled by liquid sodium. Physical characteristics of the coolant and fuel give the reactor impressive characteristics of inherent and passive safety. Spent fuel is pyrochemically reprocessed and returned to the reactor in the IFR's closed fuel cycle. Advantages in waste management are realized, and the reactor has the potential for breeding, i.e., producing as much or more fuel than it uses. This paper describes the IFR concept and shows how it is today's hope for tomorrow's electrical energy needs. 14 refs., 1 fig., 1 tab

  1. Modular Electric Propulsion Test Bed Aircraft, Phase I

    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....

  2. Technical management on commissioning test of nuclear heating reactor

    International Nuclear Information System (INIS)

    Zhang Yajun; Su Qingshan

    1999-01-01

    The commissioning is the last construction stage of a nuclear heating project. The commissioning quality will directly affect on the safe operation and availability of the heating reactor. The author presents the whole test process until the completion of the test report from the point of test documents, including the preparation and execution of the test, the management of the various unexpected events during the test. And it will be emphatically discussed that the managing procedures of the various unexpected events during the test, including temporary control change, setpoint change, unexpected events and design change

  3. Analysis of severe accidents on fast reactor test loop

    International Nuclear Information System (INIS)

    Cenerini, R.; Verzelletti, G.; Curioni, S.

    1975-01-01

    The Pec reactor is a sodium cooled fast reactor which is being designed for the primary purpose of accomodating closed sodium cooled test loops for the developmental and proof testing of fast reactor fuel assemblies. The test loops are located in the central test region of reactor. The basic function for which the loop is designed is burn-up to failure testing of fuel under advanced performance conditions. It is therefore necessary to design the loop for failure conditions. Basically two types of accidents can occur within the loops: rupture of gas plenum in the fuel pins and coolant starvation. Explosive tests on Pec loop, whose first set is described in this report, are devoted to investigate the effects of an accidental energy release on loop containment. The loop model reproduces in the test section the prototype dimensions in radial scale 1:1. Using a wire explosive charge of 300mm, the height of test section is sufficient for determining the containment capability of the loop that has a nearly constant deformation in a length of. 3-4 time the diameter. The inertial effects of the coolant column are reproduced by two tubes at the extremities of test section, closed with top plugs. Some tests has been performed by wrapping around the test section four layers of steel wire in order to evaluate the influence on the containment of tungsten wire that is foreseen in prototype loop. The influence of the coolant around the loop was evaluated by inserting the model in water. Dummy sub-assemblies was used and explosive substitutes the central rods. Piezoelectric pressure transducers were mounted on the three plugs and radial deformation was measured directly at different height. From experiments performed it resulted the importance of harmonic wires and inertial reaction of external water on loop containment; maximum containable energy is about 50 Cal with E.1 explosive

  4. Nuclear reactor capable of electric power generation during in-service inspection

    International Nuclear Information System (INIS)

    Nakamura, Shinsuke; Nogami, Hitoshi.

    1992-01-01

    The nuclear power plant according to the present invention can generate electric power even in a period when one of a pair of reactors is put to in-service inspection. That is, the nuclear power plant of the present invention comprises a system constitution of two nuclear reactors each of 50% thermal power and one turbine power generator of 100% electric power. Further, facilities of various systems relevant to the two reactors each of 50% thermal power, as a pair, are used in common as much as possible in order to reduce the cost for construction and maintenance/ inspection. Further, a reactor building and a turbine building disposed in adjacent with each for paired two reactors each of 50% thermal power are arranged vertically. This arrangement can facilitate the common use of the facilities for various systems and equipments to attain branching and joining of fluids in reactor feed water systems and main steam system pipelines easily with low pressure loss and low impact shocks. The facility utilization factor of such reactors is remarkably improved by doubling the period of continuous power generation. As a result, economic property is remarkably improved. (I.S.)

  5. Reactor protection system with automatic self-testing and diagnostic

    International Nuclear Information System (INIS)

    Gaubatz, D.C.

    1996-01-01

    A reactor protection system is disclosed having four divisions, with quad redundant sensors for each scram parameter providing input to four independent microprocessor-based electronic chassis. Each electronic chassis acquires the scram parameter data from its own sensor, digitizes the information, and then transmits the sensor reading to the other three electronic chassis via optical fibers. To increase system availability and reduce false scrams, the reactor protection system employs two levels of voting on a need for reactor scram. The electronic chassis perform software divisional data processing, vote 2/3 with spare based upon information from all four sensors, and send the divisional scram signals to the hardware logic panel, which performs a 2/4 division vote on whether or not to initiate a reactor scram. Each chassis makes a divisional scram decision based on data from all sensors. Automatic detection and discrimination against failed sensors allows the reactor protection system to automatically enter a known state when sensor failures occur. Cross communication of sensor readings allows comparison of four theoretically ''identical'' values. This permits identification of sensor errors such as drift or malfunction. A diagnostic request for service is issued for errant sensor data. Automated self test and diagnostic monitoring, sensor input through output relay logic, virtually eliminate the need for manual surveillance testing. This provides an ability for each division to cross-check all divisions and to sense failures of the hardware logic. 16 figs

  6. 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.

  7. UF6 breeder reactor power plants for electric power generation

    International Nuclear Information System (INIS)

    Rust, J.H.; Clement, J.D.; Hohl, F.

    1976-01-01

    The reactor concept analyzed is a 233 UF 6 core surrounded by a molten salt (Li 7 F, BeF 2 , ThF 4 ) blanket. Nuclear survey calculations were carried out for both spherical and cylindrical geometries. A maximum breeding ratio of 1.22 was found. Thermodynamic cycle calculations were performed for a variety of Rankine cycles. Optimization of a Rankine cycle for a gas core breeder reactor employing an intermediate heat exchanger gave a maximum efficiency of 37 percent. A conceptual design is presented along with a system layout for a 1000 MW stationary power plant. The advantages of the GCBR are as follows: (1) high efficiency, (2) simplified on-line reprocessing, (3) inherent safety considerations, (4) high breeding ratio, (5) possibility of burning all or most of the long-lived nuclear waste actinides, and (6) possibility of extrapolating the technology to higher temperatures and MHD direct conversion

  8. Technical concept for a test of geologic storage of spent reactor fuel in the climax granite, Nevada Test Site

    International Nuclear Information System (INIS)

    Ramspott, L.D.; Ballou, L.B.; Carlson, R.C.; Montan, D.N.; Butkovich, T.R.; Duncan, J.E.; Patrick, W.C.; Wilder, D.G.; Brough, W.G.; Mayr, M.C.

    1979-01-01

    We plan to emplace spent fuel assemblies from an operating commercial nuclear reactor in the Climax granite at the US Department of Energy's Nevada Test Site. In this generic test, 11 canisters of spent fuel will be emplaced with 6 electrical simulator canisters in a storage drift 420 m below in surface and their effects compared. Two adjacent drifts will contain electrical heaters, operated to simulate the temperature-stress-displacement fields of a large repository. We describe the test objectives, the technical issues, the site, the preoperational measurement program, thermal and mechanical response calculations, the characteristics of the spent fuel, the field instrumentation and data-acquisition systems, and the system for handling the spent fuel

  9. Technical concept for test of geologic storage of spent reactor fuel in the Climax granite, Nevada Test Site

    International Nuclear Information System (INIS)

    Ramspott, L.D.; Ballou, L.B.; Carlson, R.C.; Montan, D.N.; Butkovich, T.R.; Duncan, J.E.; Patrick, W.C.; Wilder, D.G.; Brough, W.G.; Mayr, M.C.

    1979-01-01

    The Spent Fuel Test in the Climax granite at the Nevada Test Site is a generic test in which spent fuel assemblies from an operating commercial nuclear reactor are emplaced at, and retrieved from, a plausible waste repository depth in a typical granite. Eleven canisters of spent fuel are emplaced in a storage drift 420 m below the surface along with six electrical simulator canisters. Two adjacent drifts contain electrical heaters which are operated so as to simulate the initial five years of the temperature-stress-displacement fields of a large repository. The site is described, and the pre-operational measurement program and characteristics of the spent fuel are given. Both thermal and mechanical response calculations are summarized. The field instrumentation and data acquisition systems are described, as well as the system for handling the spent fuel

  10. 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...

  11. RELAP5 kinetics model development for the Advanced Test Reactor

    International Nuclear Information System (INIS)

    Judd, J.L.; Terry, W.K.

    1990-01-01

    A point-kinetics model of the Advanced Test Reactor has been developed for the RELAP5 code. Reactivity feedback parameters were calculated by a three-dimensional analysis with the PDQ neutron diffusion code. Analyses of several hypothetical reactivity insertion events by the new model and two earlier models are discussed. 3 refs., 10 figs., 6 tabs

  12. Tokamak Fusion Test Reactor neutral beam injection system vacuum chamber

    International Nuclear Information System (INIS)

    Pedrotti, L.R.

    1977-01-01

    Most of the components of the Neutral Beam Lines of the Tokamak Fusion Test Reactor (TFTR) will be enclosed in a 50 cubic meter box-shaped vacuum chamber. The chamber will have a number of unorthodox features to accomodate both neutral beam and TFTR requirements. The design constraints, and the resulting chamber design, are presented

  13. Testing of research reactor fuel in the high flux reactor (Petten)

    International Nuclear Information System (INIS)

    Guidez, J.; Markgraf, J.W.; Sordon, G.; Wijtsma, F.J.; Thijssen, P.J.M.; Hendriks, J.A.

    1999-01-01

    The two types of fuel most frequently used by the main research reactors are metallic: highly enriched uranium (>90%) and silicide low enriched uranium ( 3 . However, a need exists for research on new reactor fuel. This would permit some plants to convert without losses in flux or in cycle length and would allow new reactor projects to achieve higher possibilities especially in fluxes. In these cases research is made either on silicide with higher density, or on other types of fuel (UMo, etc.). In all cases when new fuel is proposed, there is a need, for safety reasons, to test it, especially regarding the mechanical evolution due to burn-up (swelling, etc.). Initially, such tests are often made with separate plates, but lately, using entire elements. Destructive examinations are often necessary. For this type of test, the High Flux Reactor, located in Petten (The Netherlands) has many specific advantages: a large core, providing a variety of interesting positions with high fluence rate; a downward coolant flow simplifies the engineering of the device; there exists easy access with all handling possibilities to the hot-cells; the high number of operating days (>280 days/year), together with the high flux, gives a possibility to reach quickly the high burn-up needs; an experienced engineering department capable of translating specific requirements to tailor-made experimental devices; a well equipped hot-cell laboratory on site to perform all necessary measurements (swelling, γ-scanning, profilometry) and all destructive examinations. In conclusion, the HFR reactor readily permits experimental research on specific fuels used for research reactors with all the necessary facilities on the Petten site. (author)

  14. Use of small reactors as an alternative to supply electricity to Baja California Sur

    International Nuclear Information System (INIS)

    Alonso, G.; Portes, E.; Ramirez, J. R.; Ortega, G.

    2016-09-01

    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)

  15. Processing test of an upgraded mechanical design for PERMCAT reactor

    Energy Technology Data Exchange (ETDEWEB)

    Borgognoni, Fabio, E-mail: fabio.borgognoni@enea.i [Associazione ENEA-Euratom sulla Fusione, C.R. ENEA Frascati, Via E. Fermi 45, Frascati, Roma I-00044 (Italy); Demange, David; Doerr, Lothar [Forschungszentrum Karlsruhe GmbH, Institute for Technical Physics, Tritium Laboratory Karlsruhe, Postfach 3640, D-76021 Karlsruhe (Germany); Tosti, Silvano [Associazione ENEA-Euratom sulla Fusione, C.R. ENEA Frascati, Via E. Fermi 45, Frascati, Roma I-00044 (Italy); Welte, Stefan [Forschungszentrum Karlsruhe GmbH, Institute for Technical Physics, Tritium Laboratory Karlsruhe, Postfach 3640, D-76021 Karlsruhe (Germany)

    2010-12-15

    The PERMCAT membrane reactor is a coaxial combination of a Pd/Ag permeator membrane and a catalyst bed. This device has been proposed for processing fusion reactor plasma exhaust gas. A stream containing tritium (up to 1% of tritium in different chemical forms such as water, methane or molecular hydrogen) is decontaminated in the PERMCAT by counter-current isotopic swamping with protium. Different mechanical designs of the membrane reactor have been proposed to improve robustness and lifetime. The ENEA membrane reactor uses a permeator tube with a length of about 500 mm produced via cold-rolling and diffusion welding of Pd/Ag thin foils: two stainless steel pre-tensioned bellows have been applied to the Pd/Ag tube in order to avoid any significant compressive and bending stresses due to the permeator tube elongation consequent to the hydrogen uptake. An experimental test campaign has been performed using this reactor in order to assess the influence of different operating parameters and to evaluate the overall performance (decontamination factor). Tests have been carried out on two reactor prototypes: a defect-free membrane with complete (infinite) hydrogen selectivity and not perm-selective membrane. In this last case, the study has been aimed at verifying the behaviour of the PERMCAT devices under non-normal (accidental) conditions in the view of providing information for future safety analysis. The paper will present the specific mechanical design and the experimental results of tests based on isotopic exchange between H{sub 2}O and D{sub 2}.

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

    International Nuclear Information System (INIS)

    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

  17. 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.

  18. Research and Test Reactor Fuel Elements (RTRFE)

    International Nuclear Information System (INIS)

    Pace, Brett W.; Marinak, Edward A.

    1999-01-01

    BWX Technologies Inc. (BWXT) has experienced several production improvements over the past year. The homogeneity yields in 4.8 gU/cc U 3 Si 2 plates have increased over last year's already high yields. Through teamwork and innovative manufacturing techniques, maintaining high quality surface finishes on plates and elements is becoming easier and less expensive. Currently, BWXT is designing a fabrication development plan to reach a fuel loading of 9 gU/cc within 2 - 4 years. This development will involve a step approach requested by ANL to produce plates using U-8Mo at a loading of 6 gU/cc first and qualify the fuel at those levels. In achieving the goal of a very high-density fuel loading of 9 gU/cc, BWXT is considering employing several new, state of the art, ultrasonic testing techniques for fuel core evaluation. (author)

  19. Feasibility study of the Dragon reactor for HTGR fuel testing

    International Nuclear Information System (INIS)

    Wallroth, C.F.

    1975-01-01

    The Organization of European Community Development (OECD) Dragon high-temperature reactor project has performed HTGR fuel and fuel element testing for about 10 years. To date, a total of about 250 fuel elements have been irradiated and the test program continues. The feasibility of using this test facility for HTGR fuel testing, giving special consideration to U. S. needs, is evaluated. A detailed description for design, preparation, and data acquisition of a test experiment is given together with all possible options on supporting work, which could be carried out by the experienced Dragon project staff. 11 references. (U.S.)

  20. NEPTUNIX, a general program of simulation applied to nuclear reactors

    International Nuclear Information System (INIS)

    Bonnemay, A.; Dansac Bon, V.

    1978-01-01

    Most simulation languages admit an incremental description and involve explicit integration algorithms. NEPTUNIX is a simulation language directly admitting algebraic differential equations under an implicit form, and it involves a very efficient implicit integration method with variable step and order. NEPTUNIX is a tool used for building large systems models in the field of nuclear reactors [fr

  1. Human factors evaluation of the engineering test reactor control room

    International Nuclear Information System (INIS)

    Banks, W.W.; Boone, M.P.

    1981-03-01

    The Reactor and Process Control Rooms at the Engineering Test Reactor were evaluated by a team of human factors engineers using available human factors design criteria. During the evaluation, ETR, equipment and facilities were compared with MIL-STD-1472-B, Human Engineering design Criteria for Military Systems. The focus of recommendations centered on: (a) displays and controls; placing displays and controls in functional groups; (b) establishing a consistent color coding (in compliance with a standard if possible); (c) systematizing annunciator alarms and reducing their number; (d) organizing equipment in functional groups; and (e) modifying labeling and lines of demarcation

  2. Applicability of Avery's coupled reactor theory to estimate subcriticality of test region in two region system

    International Nuclear Information System (INIS)

    Kugo, Teruhiko

    1992-01-01

    The author examined the validity to estimate the subcriticality of a test region in a coupled reactor system using only measurable quantities on the basis of Avery's coupled reactor theory. For the purpose, we analyzed coupled reactor experiments performed at the Tank-type Critical Assembly in Japan Atomic Energy Research Institute by using two region systems and evaluated the subcriticality of the test region through a numerical study. Coupling coefficients were redefined at the quasi-static state because their definitions by Avery were not clear. With the coupling coefficients obtained by the numerical calculation, the multiplication factor of the test region was evaluated by two formulas; one for the evaluation using only the measurable quantities and the other for the accurate evaluation which contains the terms dropped in the former formula by assuming the unchangeableness for the perturbation induced in a driver region. From the comparison between the results of the evaluations, it was found that the estimation using only the measurable quantities is valid only for the coupled reactor system where the subcriticality of the test region was very small within a few dollars in reactivity. Consequently, it is concluded that the estimation using only the measurable quantities is not applicable to a general coupled reactor system. (author)

  3. Closed Loop In-Reactor Assembly (CLIRA): a fast flux test facility test vehicle

    International Nuclear Information System (INIS)

    Oakley, D.J.

    1978-01-01

    The Closed Loop In-Reactor Assembly (CLIRA) is a test vehicle for in-core material and fuel experiments in the Fast Flux Test Facility (FFTF). The FFTF is a fast flux nuclear test reactor operated for the Department of Energy (DOE) by Westinghouse Hanford Company in Richland, Washington. The CLIRA is a removable/replaceable part of the Closed Loop System (CLS) which is a sodium coolant system providing flow and temperature control independent of the reactor coolant system. The primary purpose of the CLIRA is to provide a test vehicle which will permit testing of nuclear fuels and materials at conditions more severe than exist in the FTR core, and to isolate these materials from the reactor core

  4. Power unit with GT-MHR reactor plant for electricity production and district heating

    International Nuclear Information System (INIS)

    Kiryushin, A.L.; Kodochigov, N.G.; Kuzavkov, N.G.; Golovko, V.F.

    2000-01-01

    Modular helium reactor with the gas turbine (GT-MHR) is a perspective power reactor plant for the next century. The project reactor is based on experience of operation more than 50 gas-cooled reactors on carbon dioxide and helium, and also on subsequent achievements in the field of realization direct gas turbine Brayton cycle. To the beginning of 90 years, achievements in technology of gas turbines, highly effective recuperators and magnetic bearings made it possible to start development of the reactor plant project combining a safe modular gas cooled reactor and a power conversion system, realizing the highly effective Brayton cycle. The conceptual project of the commercial GT-MHR reactor plant fulfilled in 1997 by joint efforts of international firms, combines a safe modular reactor with an annular active core of prismatic fuel blocks and a power conversion system with direct gas turbine cycle. The efficiency of GT-MHR gas turbine cycle at level of about 48% makes it competitive in the electricity production market in comparison with any fossil or nuclear power stations

  5. Reactor vessel dismantling at the high flux materials testing reactor Petten

    International Nuclear Information System (INIS)

    Tas, A.; Teunissen, G.

    1986-01-01

    The project of replacing the reactor vessel of the high flux materials testing reactor (HFR) originated in 1974 when results of several research programs confirmed severe neutron embrittlement of aluminium alloys suggesting a limited life of the existing facility. This report describes the dismantling philosophy and organisation, the design of special underwater equipment, the dismantling of the reactor vessel and thermal column, and the conditioning and shielding activities resulting in a working area for the installation of the new vessel with no access limitations due to radiation. Finally an overview of the segmentation, waste disposal and radiation exposure is given. The total dismantling, segmentation and conditioning activities resulted in a total collective radiation dose of 300 mSv. (orig.) [de

  6. 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).

  7. Replacement of core components in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    Durney, J.L.; Croucher, D.W.

    1990-01-01

    The core internals of the Advanced Test Reactor are subjected to very high neutron fluences resulting in significant aging. The most irradiated components have been replaced on several occasions as a result of the neutron damage. The surveillance program to monitor the aging developed the needed criteria to establish replacement schedules and maximize the use of the reactor. The methods to complete the replacements with minimum radiation exposures to workers have been developed using the experience gained from each replacement. The original design of the reactor core and associated components allows replacements to be completed without special equipment. The plant has operated for about 20 years and is expected to continue operation for at least and additional 25 years. Aging evaluations are in progress to address additional replacements that may be needed during this period

  8. Replacement of core components in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    Durney, J.L.; Croucher, D.W.

    1989-01-01

    The core internals of the Advanced Test Reactor are subjected to very high neutron fluences resulting in significant aging. The most irradiated components have been replaced on several occasions as a result of the neutron damage. The surveillance program to monitor the aging developed the needed criteria to establish replacement schedules and maximize the use of the reactor. Methods to complete the replacements with minimum radiation exposures to workers have been developed using the experience gained from each replacement. The original design of the reactor core and associated components allows replacements to be completed without special equipment. The plant has operated for about 20 years and will continue operation for perhaps another 20 years. Aging evaluations are in program to address additional replacements that may be needed during this extended time period. 3 figs

  9. Reduced enrichment for research and test reactors: Proceedings

    International Nuclear Information System (INIS)

    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 3 was by then in routine use, illustrated how far work has progressed

  10. 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.

  11. Education and training by utilizing irradiation test reactor simulator

    International Nuclear Information System (INIS)

    Eguchi, Shohei; Koike, Sumio; Takemoto, Noriyuki; Tanimoto, Masataka; Kusunoki, Tsuyoshi

    2016-01-01

    The Japan Atomic Energy Agency, at its Japan Materials Testing Reactor (JMTR), completed an irradiation test reactor simulator in May 2012. This simulator simulates the operation, irradiation test, abnormal transient change during operation, and accident progress events, etc., and is able to perform operation training on reactor and irradiation equipment corresponding to the above simulations. This simulator is composed of a reactor control panel, process control panel, irradiation equipment control panel, instructor control panel, large display panel, and compute server. The completed simulator has been utilized in the education and training of JMTR operators for the purpose of the safe and stable operation of JMTR and the achievement of high operation rate after resuming operation. For the education and training, an education and training curriculum has been prepared for use in not only operation procedures at the time of normal operation, but also learning of fast and accurate response in case of accident events. In addition, this simulator is also being used in operation training for the purpose of contributing to the cultivation of human resources for atomic power in and out of Japan. (A.O.)

  12. Situation of test and research reactors' spent fuels

    International Nuclear Information System (INIS)

    Shimizu, Kenichi; Uchiyama, Junzo; Sato, Hiroshi

    1996-01-01

    The U.S. DOE decided a renewal Off-Site Fuel Policy for stopping to spread a highly enriched uranium which was originally enriched at the U.S., the policy declared that to receive all HEU spent fuels from Test and Research reactors in all the world. In Japan, under bilateral agreement of cooperation between the government of the United States and the government of Japan concerning peaceful uses of nuclear energy, the highly enriched uranium of Test and Research Reactors' fuels was purchased from the U.S. and the fuels had been manufactured in Japan, America, Germany and France. On the other hand, a former president of the U.S. J. Carter proposed that to convert the fuels from HEU to LEU concerning a nonproliferation of nuclear materials in 1978, and Japan absolutely supported this policy. Under this condition, the U.S. stopped to receive the spent fuels from the other countries concerning legal action to the Off-Site Fuels Policy. As a result, the spent fuels are increasing, and to cross to each reactor's storage capacity, and if this policy start, a faced crisis of Test and Research Reactors will be avoided. (author)

  13. The decommissioning of the KEMA suspension test reactor

    International Nuclear Information System (INIS)

    Spruyt, A.; Peters, D.; Loon, W.M.G.M. van; Boekschoten, H.J.C.; Brugman, H.

    1991-01-01

    In this report the decommissioning of the KEMA Suspension Test Reactor (KSTR) is described. This reactor was a 1 MWth aqueous homo-geneous nuclear reactor in which a suspension of a mixed oxide UO 2 / ThO 2 in light water was circulated in a closed loop through a sphere-shaped core vessel. The reactor, located on KEMA premises, made 150 MW of heat during its critical periods. Dismantling of this reactor, with its many connected subsystems, meant the mastering of activated components which were also contaminated on inner surfaces caused by small fuel deposits (alpha contaminants) and fission products (beta, gamma contaminants). A description is given of the save removal of the fuel, the remote dismantling of systems and components and the disposal of steel scrap and other materials. Important features are the measures to be taken and provisions needed for safe handling, for the reduction of the radiation dose for the working team and the prevention of spreading of activity over the working area and the environment. It has been demonstrated that safe dismantling and disposal of such systems can be achieved. Experience gained at KEMA for the proper dismantling and for safety measures to be taken for workers and the environment can be made available for similar dismantling projects. A cost break-down is included in the report. (author). 22 refs.; 52 figs.; 12 tabs

  14. 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......, a large portion of voltage drop was caused by the ohmic (electrolyte) resistance of the medium present between two electrodes, although the two electrodes were closely positioned (about 3 cm distance; internal resistance = 35 ± 2 Ω). The open circuit potential (0.393 V vs. a standard hydrogen electrode...

  15. Electrical measuring device for a high temperature reactor

    International Nuclear Information System (INIS)

    Elter, C.; Handel, H.; Schoening, J.; Schmitt, H.

    1982-01-01

    The device for measuring the low or high neutron flux during start-up or at load is accommodated in an armoured guide tube projecting into the floor. A gas-tight capsule is formed as the measuring column with outer dome with a lid solidly connected by a flange to the armoured tube situated on the side wall of the concrete reactor vessel, together with the armoured guide tube. Two shielding shutters prevent the passage of radiation through the armoured tube. (DG) [de

  16. Accelerated irradiation test of gundremmingen reactor vessel trepan material

    International Nuclear Information System (INIS)

    Hawthorne, J.R.

    1992-08-01

    Initial mechanical properties tests of beltline trepanned from the decommissioned KRB-A pressure vessel and archive material irradiated in the UBR test reactor revealed a major anomaly in relative radiation embrittlement sensitivity. Poor correspondence of material behavior in test vs. power reactor environments was observed for the weak test orientation (ASTL C-L) whereas correspondence was good for the strong orientation (ASTM C-L). To resolve the anomaly directly, Charpy-V specimens from a low (essentially-nil) fluence region of the vessel were irradiated together with archive material at 279 degrees C in the UBR test reactor. Properties tests before UBR irradiation revealed a significant difference in 41-J transition temperature and upper shelf energy level between the materials. However, the materials exhibited essentially the same radiation embrittlement sensitivity (both orientations), proving that the anomaly is not due to a basic difference in material irradiation resistances. Possible causes of the original anomaly and the significance to NRC Regulatory Guide 1.99 are discussed

  17. Accelerated irradiation test of Gundremmingen reactor vessel trepan material

    Energy Technology Data Exchange (ETDEWEB)

    Hawthorne, J.R. [Materials Engineering Associates, Inc., Lanham, MD (United States)

    1992-08-01

    Initial mechanical properties tests of beltline trepanned from the decommissioned KRB-A pressure vessel and archive material irradiated in the UBR test reactor revealed a major anomaly in relative radiation embrittlement sensitivity. Poor correspondence of material behavior in test vs. power reactor environments was observed for the weak test orientation (ASTL C-L) whereas correspondence was good for the strong orientation (ASTM C-L). To resolve the anomaly directly, Charpy-V specimens from a low (essentially-nil) fluence region of the vessel were irradiated together with archive material at 279{degrees}C in the UBR test reactor. Properties tests before UBR irradiation revealed a significant difference in 41-J transition temperature and upper shelf energy level between the materials. However, the materials exhibited essentially the same radiation embrittlement sensitivity (both orientations), proving that the anomaly is not due to a basic difference in material irradiation resistances. Possible causes of the original anomaly and the significance to NRC Regulatory Guide 1.99 are discussed.

  18. Rupture tests with reactor pressure vessel head models

    International Nuclear Information System (INIS)

    Talja, H.; Keinaenen, H.; Hosio, E.; Pankakoski, P.H.; Rahka, K.

    2003-01-01

    In the LISSAC project (LImit Strains in Severe ACcidents), partly funded by the EC Nuclear Fission and Safety Programme within the 5th Framework programme, an extensive experimental and computational research programme is conducted to study the stress state and size dependence of ultimate failure strains. The results are aimed especially to make the assessment of severe accident cases more realistic. For the experiments in the LISSAC project a block of material of the German Biblis C reactor pressure vessel was available. As part of the project, eight reactor pressure vessel head models from this material (22 NiMoCr 3 7) were tested up to rupture at VTT. The specimens were provided by Forschungszentrum Karlsruhe (FzK). These tests were performed under quasistatic pressure load at room temperature. Two specimens sizes were tested and in half of the tests the specimens contain holes describing the control rod penetrations of an actual reactor pressure vessel head. These specimens were equipped with an aluminium liner. All six tests with the smaller specimen size were conducted successfully. In the test with the large specimen with holes, the behaviour of the aluminium liner material proved to differ from those of the smaller ones. As a consequence the experiment ended at the failure of the liner. The specimen without holes yielded results that were in very good agreement with those from the small specimens. (author)

  19. 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.; 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.; Levin, Y.; Levine, B. M.; Li, T. G. F.; Libson, A.; Littenberg, T. B.; Lockerbie, N. A.; Logue, J.; Lombardi, A. L.; London, L. T.; Lord, J. E.; Lorenzini, M.; Loriette, V.; Lormand, M.; Losurdo, G.; Lough, J. D.; Lousto, C. O.; Lovelace, G.; Lück, H.; Lundgren, A. P.; Luo, J.; Lynch, R.; Ma, Y.; MacDonald, T.; Machenschalk, B.; MacInnis, M.; Macleod, D. M.; Magaña-Sandoval, F.; Magee, R. M.; Mageswaran, M.; Majorana, E.; Maksimovic, I.; Malvezzi, V.; Man, N.; Mandel, I.; Mandic, V.; Mangano, V.; Mansell, G. L.; Manske, M.; Mantovani, M.; Marchesoni, F.; Marion, F.; Márka, S.; Márka, Z.; Markosyan, A. S.; Maros, E.; Martelli, F.; Martellini, L.; Martin, I. W.; Martin, R. M.; Martynov, D. V.; Marx, J. N.; Mason, K.; Masserot, A.; Massinger, T. J.; Masso-Reid, M.; Matichard, F.; Matone, L.; Mavalvala, N.; Mazumder, N.; Mazzolo, G.; McCarthy, R.; McClelland, D. E.; McCormick, S.; McGuire, S. C.; McIntyre, G.; McIver, J.; McManus, D. J.; McWilliams, S. T.; Meacher, D.; Meadors, G. D.; Meidam, J.; Melatos, A.; Mendell, G.; Mendoza-Gandara, D.; Mercer, R. A.; Merilh, E.; Merzougui, M.; Meshkov, S.; Messenger, C.; Messick, C.; Meyers, P. M.; Mezzani, F.; Miao, H.; Michel, C.; Middleton, H.; Mikhailov, E. E.; Milano, L.; Miller, J.; Millhouse, M.; Minenkov, Y.; Ming, J.; Mirshekari, S.; Mishra, C.; Mitra, S.; Mitrofanov, V. P.; Mitselmakher, G.; Mittleman, R.; Moggi, A.; Mohan, M.; Mohapatra, S. R. P.; Montani, M.; Moore, B. C.; Moore, C. J.; Moraru, D.; Moreno, G.; Morriss, S. R.; Mossavi, K.; Mours, B.; Mow-Lowry, C. M.; Mueller, C. L.; Mueller, G.; Muir, A. W.; Mukherjee, Arunava; Mukherjee, D.; Mukherjee, S.; Mukund, N.; Mullavey, A.; Munch, J.; Murphy, D. J.; Murray, P. G.; Mytidis, A.; Nardecchia, I.; Naticchioni, L.; Nayak, R. K.; Necula, V.; Nedkova, K.; Nelemans, G.; Neri, M.; Neunzert, A.; Newton, G.; Nguyen, T. T.; Nielsen, A. B.; Nissanke, S.; Nitz, A.; Nocera, F.; Nolting, D.; Normandin, M. E.; Nuttall, L. K.; Oberling, J.; Ochsner, E.; O'Dell, J.; Oelker, E.; Ogin, G. H.; Oh, J. J.; Oh, S. H.; Ohme, F.; Oliver, M.; Oppermann, P.; Oram, Richard J.; O'Reilly, B.; O'Shaughnessy, R.; Ottaway, D. J.; Ottens, R. S.; Overmier, H.; Owen, B. J.; Pai, A.; Pai, S. A.; Palamos, J. R.; Palashov, O.; Palomba, C.; Pal-Singh, A.; Pan, H.; Pan, Y.; Pankow, C.; Pannarale, F.; Pant, B. C.; Paoletti, F.; Paoli, A.; Papa, M. A.; Paris, H. R.; Parker, W.; Pascucci, D.; Pasqualetti, A.; Passaquieti, R.; Passuello, D.; Patricelli, B.; Patrick, Z.; Pearlstone, B. L.; Pedraza, M.; Pedurand, R.; Pekowsky, L.; Pele, A.; Penn, S.; Perreca, A.; Pfeiffer, H. P.; Phelps, M.; Piccinni, O.; Pichot, M.; Piergiovanni, F.; Pierro, V.; Pillant, G.; Pinard, L.; Pinto, I. M.; Pitkin, M.; Poggiani, R.; Popolizio, P.; Post, A.; Powell, J.; Prasad, J.; Predoi, V.; Premachandra, S. S.; Prestegard, T.; Price, L. R.; Prijatelj, M.; Principe, M.; Privitera, S.; Prix, R.; Prodi, G. A.; Prokhorov, L.; Puncken, O.; Punturo, M.; Puppo, P.; Pürrer, M.; Qi, H.; Qin, J.; Quetschke, V.; Quintero, E. A.; Quitzow-James, R.; Raab, F. J.; Rabeling, D. S.; Radkins, H.; Raffai, P.; Raja, S.; Rakhmanov, M.; Rapagnani, P.; Raymond, V.; Razzano, M.; Re, V.; Read, J.; Reed, C. M.; Regimbau, T.; Rei, L.; Reid, S.; Reitze, D. H.; Rew, H.; Reyes, S. D.; Ricci, F.; Riles, K.; Robertson, N. A.; Robie, R.; Robinet, F.; Rocchi, A.; Rolland, L.; Rollins, J. G.; Roma, V. J.; Romano, R.; Romanov, G.; Romie, J. H.; Rosińska, D.; Rowan, S.; Rüdiger, A.; Ruggi, P.; Ryan, K.; Sachdev, S.; Sadecki, T.; Sadeghian, L.; Salconi, L.; Saleem, M.; Salemi, F.; Samajdar, A.; Sammut, L.; Sanchez, E. J.; Sandberg, V.; Sandeen, B.; Sanders, J. R.; Sassolas, B.; Sathyaprakash, B. S.; Saulson, P. R.; Sauter, O.; Savage, R. L.; Sawadsky, A.; Schale, P.; Schilling, R.; Schmidt, J.; Schmidt, P.; Schnabel, R.; Schofield, R. M. S.; Schönbeck, A.; Schreiber, E.; Schuette, D.; Schutz, B. F.; Scott, J.; Scott, S. M.; Sellers, D.; Sengupta, A. S.; Sentenac, D.; Sequino, V.; Sergeev, A.; Serna, G.; Setyawati, Y.; Sevigny, A.; Shaddock, D. A.; Shah, S.; Shahriar, M. S.; Shaltev, M.; Shao, Z.; Shapiro, B.; Shawhan, P.; Sheperd, A.; Shoemaker, D. H.; Shoemaker, D. M.; Siellez, K.; Siemens, X.; Sigg, D.; Silva, A. D.; Simakov, D.; Singer, A.; Singer, L. P.; Singh, A.; Singh, R.; Singhal, A.; Sintes, A. M.; Slagmolen, B. J. J.; Smith, J. R.; Smith, N. D.; Smith, R. J. E.; Son, E. J.; Sorazu, B.; Sorrentino, F.; Souradeep, T.; Srivastava, A. K.; Staley, A.; Steinke, M.; Steinlechner, J.; Steinlechner, S.; Steinmeyer, D.; Stephens, B. C.; Stone, R.; Strain, K. A.; Straniero, N.; Stratta, G.; Strauss, N. A.; Strigin, S.; Sturani, R.; Stuver, A. L.; Summerscales, T. Z.; Sun, L.; Sutton, P. J.; Swinkels, B. L.; Szczepańczyk, M. J.; Tacca, M.; Talukder, D.; Tanner, D. B.; Tápai, M.; Tarabrin, S. P.; Taracchini, A.; Taylor, R.; Theeg, T.; Thirugnanasambandam, M. P.; Thomas, E. G.; Thomas, M.; Thomas, P.; Thorne, K. A.; Thorne, K. S.; Thrane, E.; Tiwari, S.; Tiwari, V.; Tokmakov, K. V.; Tomlinson, C.; Tonelli, M.; Torres, C. V.; Torrie, C. I.; Töyrä, D.; Travasso, F.; Traylor, G.; Trifirò, D.; Tringali, M. C.; Trozzo, L.; Tse, M.; Turconi, M.; Tuyenbayev, D.; Ugolini, D.; Unnikrishnan, C. S.; Urban, A. L.; Usman, S. A.; Vahlbruch, H.; Vajente, G.; Valdes, G.; Vallisneri, M.; van Bakel, N.; van Beuzekom, M.; van den Brand, J. F. J.; Van Den Broeck, C.; Vander-Hyde, D. C.; van der Schaaf, L.; van Heijningen, J. V.; van Veggel, A. A.; Vardaro, M.; Vass, S.; Vasúth, M.; Vaulin, R.; Vecchio, A.; Vedovato, G.; Veitch, J.; Veitch, P. J.; Venkateswara, K.; 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.; LIGO Scientific; Virgo Collaborations

    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.

  20. 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; 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; Levin, Y; Levine, B M; Li, T G F; Libson, A; Littenberg, T B; Lockerbie, N A; Logue, J; Lombardi, A L; London, L T; Lord, J E; Lorenzini, M; Loriette, V; Lormand, M; Losurdo, G; Lough, J D; Lousto, C O; Lovelace, G; Lück, H; Lundgren, A P; Luo, J; Lynch, R; Ma, Y; MacDonald, T; Machenschalk, B; MacInnis, M; Macleod, D M; Magaña-Sandoval, F; Magee, R M; Mageswaran, M; Majorana, E; Maksimovic, I; Malvezzi, V; Man, N; Mandel, I; Mandic, V; Mangano, V; Mansell, G L; Manske, M; Mantovani, M; Marchesoni, F; Marion, F; Márka, S; Márka, Z; Markosyan, A S; Maros, E; Martelli, F; Martellini, L; Martin, I W; Martin, R M; Martynov, D V; Marx, J N; Mason, K; Masserot, A; Massinger, T J; Masso-Reid, M; Matichard, F; Matone, L; Mavalvala, N; Mazumder, N; Mazzolo, G; McCarthy, R; McClelland, D E; McCormick, S; McGuire, S C; McIntyre, G; McIver, J; McManus, D J; McWilliams, S T; Meacher, D; Meadors, G D; Meidam, J; Melatos, A; Mendell, G; Mendoza-Gandara, D; Mercer, R A; Merilh, E; Merzougui, M; Meshkov, S; Messenger, C; Messick, C; Meyers, P M; Mezzani, F; Miao, H; Michel, C; Middleton, H; Mikhailov, E E; Milano, L; Miller, J; Millhouse, M; Minenkov, Y; Ming, J; Mirshekari, S; Mishra, C; Mitra, S; Mitrofanov, V P; Mitselmakher, G; Mittleman, R; Moggi, A; Mohan, M; Mohapatra, S R P; Montani, M; Moore, B C; Moore, C J; Moraru, D; Moreno, G; Morriss, S R; Mossavi, K; Mours, B; Mow-Lowry, C M; Mueller, C L; Mueller, G; Muir, A W; Mukherjee, Arunava; Mukherjee, D; Mukherjee, S; Mukund, N; Mullavey, A; Munch, J; Murphy, D J; Murray, P G; Mytidis, A; Nardecchia, I; Naticchioni, L; Nayak, R K; Necula, V; Nedkova, K; Nelemans, G; Neri, M; Neunzert, A; Newton, G; Nguyen, T T; Nielsen, A B; Nissanke, S; Nitz, A; Nocera, F; Nolting, D; Normandin, M E; Nuttall, L K; Oberling, J; Ochsner, E; O'Dell, J; Oelker, E; Ogin, G H; Oh, J J; Oh, S H; Ohme, F; Oliver, M; Oppermann, P; Oram, Richard J; O'Reilly, B; O'Shaughnessy, R; Ottaway, D J; Ottens, R S; Overmier, H; Owen, B J; Pai, A; Pai, S A; Palamos, J R; Palashov, O; Palomba, C; Pal-Singh, A; Pan, H; Pan, Y; Pankow, C; Pannarale, F; Pant, B C; Paoletti, F; Paoli, A; Papa, M A; Paris, H R; Parker, W; Pascucci, D; Pasqualetti, A; Passaquieti, R; Passuello, D; Patricelli, B; Patrick, Z; Pearlstone, B L; Pedraza, M; Pedurand, R; Pekowsky, L; Pele, A; Penn, S; Perreca, A; Pfeiffer, H P; Phelps, M; Piccinni, O; Pichot, M; Piergiovanni, F; Pierro, V; Pillant, G; Pinard, L; Pinto, I M; Pitkin, M; Poggiani, R; Popolizio, P; Post, A; Powell, J; Prasad, J; Predoi, V; Premachandra, S S; Prestegard, T; Price, L R; Prijatelj, M; Principe, M; Privitera, S; Prix, R; Prodi, G A; Prokhorov, L; Puncken, O; Punturo, M; Puppo, P; Pürrer, M; Qi, H; Qin, J; Quetschke, V; Quintero, E A; Quitzow-James, R; Raab, F J; Rabeling, D S; Radkins, H; Raffai, P; Raja, S; Rakhmanov, M; Rapagnani, P; Raymond, V; Razzano, M; Re, V; Read, J; Reed, C M; Regimbau, T; Rei, L; Reid, S; Reitze, D H; Rew, H; Reyes, S D; Ricci, F; Riles, K; Robertson, N A; Robie, R; Robinet, F; Rocchi, A; Rolland, L; Rollins, J G; Roma, V J; Romano, R; Romanov, G; Romie, J H; Rosińska, D; Rowan, S; Rüdiger, A; Ruggi, P; Ryan, K; Sachdev, S; Sadecki, T; Sadeghian, L; Salconi, L; Saleem, M; Salemi, F; Samajdar, A; Sammut, L; Sanchez, E J; Sandberg, V; Sandeen, B; Sanders, J R; Sassolas, B; Sathyaprakash, B S; Saulson, P R; Sauter, O; Savage, R L; Sawadsky, A; Schale, P; Schilling, R; Schmidt, J; Schmidt, P; Schnabel, R; Schofield, R M S; Schönbeck, A; Schreiber, E; Schuette, D; Schutz, B F; Scott, J; Scott, S M; Sellers, D; Sengupta, A S; Sentenac, D; Sequino, V; Sergeev, A; Serna, G; Setyawati, Y; Sevigny, A; Shaddock, D A; Shah, S; Shahriar, M S; Shaltev, M; Shao, Z; Shapiro, B; Shawhan, P; Sheperd, A; Shoemaker, D H; Shoemaker, D M; Siellez, K; Siemens, X; Sigg, D; Silva, A D; Simakov, D; Singer, A; Singer, L P; Singh, A; Singh, R; Singhal, A; Sintes, A M; Slagmolen, B J J; Smith, J R; Smith, N D; Smith, R J E; Son, E J; Sorazu, B; Sorrentino, F; Souradeep, T; Srivastava, A K; Staley, A; Steinke, M; Steinlechner, J; Steinlechner, S; Steinmeyer, D; Stephens, B C; Stone, R; Strain, K A; Straniero, N; Stratta, G; Strauss, N A; Strigin, S; Sturani, R; Stuver, A L; Summerscales, T Z; Sun, L; Sutton, P J; Swinkels, B L; Szczepańczyk, M J; Tacca, M; Talukder, D; Tanner, D B; Tápai, M; Tarabrin, S P; Taracchini, A; Taylor, R; Theeg, T; Thirugnanasambandam, M P; Thomas, E G; Thomas, M; Thomas, P; Thorne, K A; Thorne, K S; Thrane, E; Tiwari, S; Tiwari, V; Tokmakov, K V; Tomlinson, C; Tonelli, M; Torres, C V; Torrie, C I; Töyrä, D; Travasso, F; Traylor, G; Trifirò, D; Tringali, M C; Trozzo, L; Tse, M; Turconi, M; Tuyenbayev, D; Ugolini, D; Unnikrishnan, C S; Urban, A L; Usman, S A; Vahlbruch, H; Vajente, G; Valdes, G; Vallisneri, M; van Bakel, N; van Beuzekom, M; van den Brand, J F J; Van Den Broeck, C; Vander-Hyde, D C; van der Schaaf, L; van Heijningen, J V; van Veggel, A A; Vardaro, M; Vass, S; Vasúth, M; Vaulin, R; Vecchio, A; Vedovato, G; Veitch, J; Veitch, P J; Venkateswara, K; 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.

  1. RIA testing capability of the transient reactor test facility

    International Nuclear Information System (INIS)

    Crawford, D.C.; Swanson, R.W.

    1999-01-01

    The advent of high-burnup fuel implementation in LWRs has generated international interest in high-burnup LWR fuel performance. Recent testing under simulated RIA conditions has demonstrated that certain fuel designs fail at peak fuel enthalpy values that are below existing regulatory criteria. Because many of these tests were performed with non-prototypically aggressive test conditions (i.e., with power pulse widths less than 10 msec FWHM and with non-protoypic coolant configurations), the results (although very informative) do not indisputably identify failure thresholds and fuel behavior. The capability of the TREAT facility to perform simulated RIA tests with prototypic test conditions is currently being evaluated by ANL personnel. TREAT was designed to accommodate test loops and vehicles installed for in-pile transient testing. During 40 years of TREAT operation and fuel testing and evaluation, experimenters have been able to demonstrate and determine the transient behavior of several types of fuel under a variety of test conditions. This experience led to an evolution of test methodology and techniques which can be employed to assess RIA behavior of LWR fuel. A pressurized water loop that will accommodate RIA testing of LWR and CANDU-type fuel has completed conceptual design. Preliminary calculations of transient characteristics and energy deposition into test rods during hypothetical TREAT RIA tests indicate that with the installation of a pressurized water loop, the facility is quite capable of performing prototypic RIA testing. Typical test scenarios indicate that a simulated RIA with a 72 msec FWHM pulse width and energy deposition of 1200 kJ/kg (290 cal/gm) is possible. Further control system enhancements would expand the capability to pulse widths as narrow as 40 msec. (author)

  2. Present status and future perspective of research and test reactors in JAERI

    International Nuclear Information System (INIS)

    Baba, Osamu; Kaieda, Keisuke

    1999-01-01

    Since 1957, Japan Atomic Energy Research Institute (JAERI) has constructed several research and test reactors to fulfil 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 reached first criticality and is waiting for the power-up test. This paper introduce these reactors and describe their present operational status. The recent tendency of utilization and future perspectives are also reported. (author)

  3. Present status and future perspective of research and test reactors in JAERI

    Energy Technology Data Exchange (ETDEWEB)

    Baba, Osamu [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Kaieda, Keisuke

    1999-08-01

    Since 1957, Japan Atomic Energy Research Institute (JAERI) has constructed several research and test reactors to fulfil 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 reached first criticality and is waiting for the power-up test. This paper introduce these reactors and describe their present operational status. The recent tendency of utilization and future perspectives are also reported. (author)

  4. Study on personnel qualification for non-destructive tests in the field of reactor safety

    International Nuclear Information System (INIS)

    Trusch, K.; Wuestenberg, H.

    1977-01-01

    The training system for non-destructive testing is described, and the available and necessary personnel is analyzed; the personnel required for reactor safety problems is treated separately. On this basis, the subjects discussed in the study - available personnel, personnel requirements, training, training requirements, and suggestions for realisation - are treated in a general manner to begin with and afterwards with a view to specific problems of reactor safety. The methods employed are adapted to this situation. To obtain the necessary empirical data, questionnaires were set up and distributed, and experts in selected business companies and institutions were interviewed who work in the field of reactor safety or do same training in non-destructive testing. (orig.) [de

  5. Neutron importance and the generalized Green function for the conventionally critical reactor with normalized neutron distribution

    International Nuclear Information System (INIS)

    Khromov, V.V.

    1978-01-01

    The notion of neutron importance when applied to nuclear reactor statics problems described by time-independent homogeneous equations of neutron transport with provision for normalization of neutron distribution is considered. An equation has been obtained for the function of neutron importance in a conditionally critical reactor with respect to an arbitrary nons linear functional determined for the normalized neutron distribution. Relation between this function and the generalized Green function of the selfconjugated operator of the reactor equation is determined and the formula of small perturbations for the functionals of a conditionally critical reactor is deduced

  6. General scheme of research reactor mainly for production of fission 99Mo

    International Nuclear Information System (INIS)

    Shen Feng; Liu Xingmin; Wu Xiaochun; Sun Zheng; Guo Chunqiu; Yi Dayong

    2012-01-01

    On the basis of the analysis for current circumstance and development tendency of research reactor mainly for 99 Mo production in the world, the design idea of this sort of research reactor was proposed. By the optimization and basic design, the general design parameters of the reactor were analyzed and testified. The evaluation of output activities of 99 Mo and the analysis of economics were conducted on the basically assumption. It is argued that the economics of this reactor is improved dramatically while the safety is ensured by the analysis. (authors)

  7. Conceptual design of a demonstration reactor for electric power generation

    International Nuclear Information System (INIS)

    Asaoka, Y.; Hiwatari, R.; Okano, K.; Ogawa, Y.; Ise, H.; Nomoto, Y.; Kuroda, T.; Mori, S.; Shinya, K.

    2005-01-01

    Conceptual study on a demonstration plant for electric power generation, named Demo-CREST, was conducted based on the consideration that a demo-plant should have capacities both (1) to demonstrate electric power generation in a plant scale with moderate plasma performance, which will be achieved in the early stage of the ITER operation, and foreseeable technologies and materials and (2) to have a possibility to show an economical competitiveness with advanced plasma performance and high performance blanket systems. The plasma core was optimized to be a minimum size for both net electric power generation with the ITER basic plasma parameters and commercial-scale generation with advance plasma parameters, which would be attained by the end of ITER operation. The engineering concept, especially the breeding blanket structure and its maintenance scheme, is also optimized to demonstrate the tritium self-sustainability and maintainability of in-vessel components. Within the plasma performance as planned in the present ITER program, the net electric power from 0 MW to 500 MW is possible with the basic blanket system under the engineering conditions of maximum magnetic field 16 T, NBI system efficiency 50%, and NBI current drive power restricted to 200 MW. Capacities of stabilization of reversed shear plasma and the high thermal efficiency are additional factors for optimization of the advanced blanket. By replacing the blanket system with the advanced one of higher thermal efficiency, the net electric power of about 1000 MW is also possible so that the economic performance toward the commercial plant can be also examined with Demo-CREST. (author)

  8. A precise extragalactic test of General Relativity.

    Science.gov (United States)

    Collett, Thomas E; Oldham, Lindsay J; Smith, Russell J; Auger, Matthew W; Westfall, Kyle B; Bacon, David; Nichol, Robert C; Masters, Karen L; Koyama, Kazuya; van den Bosch, Remco

    2018-06-22

    Einstein's theory of gravity, General Relativity, has been precisely tested on Solar System scales, but the long-range nature of gravity is still poorly constrained. The nearby strong gravitational lens ESO 325-G004 provides a laboratory to probe the weak-field regime of gravity and measure the spatial curvature generated per unit mass, γ. By reconstructing the observed light profile of the lensed arcs and the observed spatially resolved stellar kinematics with a single self-consistent model, we conclude that γ = 0.97 ± 0.09 at 68% confidence. Our result is consistent with the prediction of 1 from General Relativity and provides a strong extragalactic constraint on the weak-field metric of gravity. Copyright © 2018 The Authors, some rights reserved; exclusive licensee American Association for the Advancement of Science. No claim to original U.S. Government Works.

  9. 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.

  10. Heat resistant/radiation resistant cable and incore structure test device for FBR type reactor

    International Nuclear Information System (INIS)

    Tanimoto, Hajime; Shiono, Takeo; Sato, Yoshimi; Ito, Kazumi; Sudo, Shigeaki; Saito, Shin-ichi; Mitsui, Hisayasu.

    1995-01-01

    A heat resistant/radiation resistant coaxial cable of the present invention comprises an insulation layer, an outer conductor and a protection cover in this order on an inner conductor, in which the insulation layer comprises thermoplastic polyimide. In the same manner, a heat resistant/radiation resistant power cable has an insulation layer comprising thermoplastic polyimide on a conductor, and is provided with a protection cover comprising braid of alamide fibers at the outer circumference of the insulation layer. An incore structure test device for an FBR type reactor comprises the heat resistant/radiation resistant coaxial cable and/or the power cable. The thermoplastic polyimide can be extrusion molded, and has excellent radiation resistant by the extrusion, as well as has high dielectric withstand voltage, good flexibility and electric characteristics at high temperature. The incore structure test device for the FBR type reactor of the present invention comprising such a cable has excellent reliability and durability. (T.M.)

  11. Generalized saddle point condition for ignition in a tokamak reactor with temperature and density profiles

    International Nuclear Information System (INIS)

    Mitari, O.; Hirose, A.; Skarsgard, H.M.

    1989-01-01

    In this paper, the concept of a generalized ignition contour map, is extended to the realistic case of a plasma with temperature and density profiles in order to study access to ignition in a tokamak reactor. The generalized saddle point is found to lie between the Lawson and ignition conditions. If the height of the operation path with Goldston L-mode scaling is higher than the generalized saddle point, a reactor can reach ignition with this scaling for the case with no confinement degradation effect due to alpha-particle heating. In this sense, the saddle point given in a general form is a new criterion for reaching ignition. Peaking the profiles for the plasma temperature and density can lower the height of the generalized saddle point and help a reactor to reach ignition. With this in mind, the authors can judge whether next-generation tokamaks, such as Compact Ignition Tokamak, Tokamak Ignition/Burn Experimental Reactor, Next European Torus, Fusion Experimental Reactor, International Tokamak Reactor, and AC Tokamak Reactor, can reach ignition with realistic profile parameters and an L-mode scaling law

  12. A European view of the use of nuclear reactors for applications other than electricity generation

    International Nuclear Information System (INIS)

    Marsham, T.N.; Brierley, G.

    Energy demands and temperature ranges employed by heat-consuming industrial processes are analyzed. Matching heat demand to reactor size is a problem. Emphasis is placed on HTGR's providing heat in the range 300-800 deg C. Further non-electrical uses of nuclear power, like nuclear ship propulsion, are analyzed. (E.C.B.)

  13. An electrical pulse hydride injector (EPHI) for reactor fueling and tritium handling applications

    International Nuclear Information System (INIS)

    Azizov, E.A.; Kareev, Yu.A.; Savotkin, A.N.; Frunze, V.V.; Penzhorn, R.D.; Glugla, M.

    1995-01-01

    An electrical pulse hydride injector (EPHI) has been developed for reactor fuelling as well as for handling of hydrogen isotopes in facilities operating with tritium. Salient features of the EPHI are the accuracy with which the fuelling rate can be controlled and the avoidance of a pressurized ballast. The generator is simple and allows for safe operation with tritium. (orig.)

  14. Integrated leak rate test results of JOYO reactor containment vessel

    International Nuclear Information System (INIS)

    Tamura, M.; Endo, J.

    1982-02-01

    Integrated leak rate tests of JOYO after the reactor coolant system had been filled with sodium have been performed two times since 1978 (February 1978 and December 1979). The tests were conducted with the in-containment sodium systems, primary argon cover gas system and air conditioning systems operating. Both the absolute pressure method and the reference chamber method were employed during the test. The results of both tests confirmed the functioning of the containment vessel, and leak rate limits were satisfied. In Addition, the adequancy of the test instrumentation system and the test method was demonstrated. Finally the plant conditions required to maintain reasonable accuracy for the leak rate testing of LMFBR were established. In this paper, the test conditions and the test results are described. (author)

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

    International Nuclear Information System (INIS)

    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 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

  16. General considerations in fuel management for thermal reactors

    International Nuclear Information System (INIS)

    Tyror, J.G.; Fayers, F.J.

    1971-07-01

    By fuel management we mean the strategy for fuelling and refuelling a reactor together with any associated absorber movements. It incorporates (a) decisions made about the timing of fuel loading operations; (b) choice of enrichments to be loaded; (c) selection of sites at which reloading occurs; (d) programming of control rods and any other reactivity control facilities such as soluble or burnable poisons; and (e) evaluation of the resulting fuel element performance consequences. The topic of fuel management is thus a vast and vital one. It embraces most of the various aspects of core performance and determines many of a reactor's design characteristics. In this paper we review what to us appear to be some of the important issues in this important field

  17. General considerations for neutron capture therapy at a reactor facility

    International Nuclear Information System (INIS)

    Binney, S.E.

    2001-01-01

    In addition to neutron beam intensity and quality, there are also a number of other significant criteria related to a nuclear reactor that contribute to a successful neutron capture therapy (NCT) facility. These criteria are classified into four main categories: Nuclear design factors, facility management and operations factors, facility resources, and non-technical factors. Important factors to consider are given for each of these categories. In addition to an adequate neutron beam intensity and quality, key requirements for a successful neutron capture therapy facility include necessary finances to construct or convert a facility for NCT, a capable medical staff to perform the NCT, and the administrative support for the facility. The absence of any one of these four factors seriously jeopardizes the overall probability of success of the facility. Thus nuclear reactor facility management considering becoming involved in neutron capture therapy, should it be proven clinically successful, should take all these factors into consideration. (author)

  18. 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.

  19. 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.

  20. Mechanical behaviour of the reactor vessel support of a pressurized water reactor: tests and analysis

    International Nuclear Information System (INIS)

    Bolvin, M.; L'huby, Y.; Quillico, J.J.; Humbert, J.M.; Thomas, J.P.; Hugenschmitt, R.

    1985-08-01

    The PWR reactor vessel is supported by a steel ring laying on the reactor pit. This support has to ensure a good behaviour of the vessel in the event of accidental conditions (earthquake and pipe rupture). A new evolution of the evaluation methods of the applied forces has shown a significant increase in the design loads used until now. In order to take into account these new forces, we carried out a test on a representative mock-up of the vessel support (scale 1/6). This test was performed by CEA, EDF and FRAMATOME. Several static equivalent forces were applied on the experimental mock-up. Displacements and strains were simultaneously recorded. The results of the test have enabled to justify the design of the pit and the ring, to show up a wide safety margin until the collapse of the structures and to check our hypothesis about the transmission of the forces between the ring and the pit

  1. Manufacturing and material properties of forgings for reactor pressure vessel of high temperature engineering test reactor

    International Nuclear Information System (INIS)

    Sato, I.; Suzuki, K.

    1994-01-01

    For the reactor pressure vessel (RPV) of high temperature engineering test reactor (HTTR) which has been developed by Japan Atomic Energy Research Institute (JAERI), 2 1/4Cr-1Mo steel is used first in the world. Material confirmation test has been carried out to demonstrate good applicability of forged low Si 2 1/4Cr-1Mo steel to the RPV of HTTR. Recently, JSW has succeeded in the manufacturing of large size ring forgings and large size forged cover dome integrated with nozzles for stand pipe for the RPV. This paper describes the results of the material confirmation test as well as the manufacturing and material properties of the large forged cover dome integrated with nozzles for stand pipe. (orig.)

  2. Generation of net electric power with a tokamak reactor under foreseeable physical and engineering conditions

    International Nuclear Information System (INIS)

    Hiwatari, R.; Asaoka, Y.; Okano, K.; Yoshida, T.; Tomabechi, K.

    2004-01-01

    This study reveals for the first time the plasma performance required for a tokamak reactor to generate net electric power under foreseeable engineering conditions. It was found that the reference plasma performance of the ITER inductive operation mode with β N = 1.8, HH = 1.0, andf nGW 0.85 had sufficient potential to achieve the electric break-even condition (net electric power P e net = 0MW) under the following engineering conditions: machine major radius 6.5m ≤ R p ≤ 8.5m, the maximum magnetic field on TF coils B tmax = 16 T, thermal efficiency η e 30%, and NBI system efficiency η NBI = 50%. The key parameters used in demonstrating net electric power generation in tokamak reactors are β N and fη GW . ≥ 3.0 is required for P e net ∼ 600MW with fusion power P f ∼ 3000MW. On the other hand, fη GW ≥ 1.0 is inevitable to demonstrate net electric power generation, if high temperatures, such as average temperatures of T ave > 16 keV, cannot be selected for the reactor design. To apply these results to the design of a tokamak reactor for demonstrating net electric power generation, the plasma performance diagrams on the Q vs P f (energy multiplication factor vs fusion power) space for several major radii (i.e. 6.5, 7.5, and 8.5 m) were depicted. From these figures, we see that a design with a major radius R p ∼ 7.5m seems preferable for demonstrating net electric power generation when one aims at early realization of fusion energy. (author)

  3. Thermal-hydraulic tests for reactor safety system

    International Nuclear Information System (INIS)

    Chun, Se Young; Chung, Moon Ki; Baek, Won Pil

    2002-05-01

    Tests for the safety depressurization system, Sparger adopted for the Korean next generation reactor, APR1400 are carried out for several geometries with the B and C (Blowdown and Condensation) facility in the condition of high temperature and pressure and with a small test facility in the condition of atmospheric temperature and pressure. Tests for the critical heat flux are performed with the RCS(Reactor Coolant System) facility as well as with the Freon CHF Loop in the condition of high temperature and pressure. The atmospheric temperature and pressure facility is utilized for development of the high standard thermal hydraulic measurement technology. The optical method is developed to measure the local thermal-hydraulic behavior for the single and two-phase boiling phenomena

  4. Reactor physics tests of TRIGA Mark-II Reactor in Ljubljana

    International Nuclear Information System (INIS)

    Ravnik, M.; Mele, I.; Trkov, A.; Rant, J.; Glumac, B.; Dimic, V.

    2008-01-01

    TRIGA Mark-II Reactor in Ljubljana was recently reconstructed. The reconstruction consisted mainly of replacing the grid plates, the control rod mechanisms and the control unit. The standard type control rods were replaced by the fuelled follower type, the central grid location (A ring) was adapted for fuel element insertion, the triangular cutouts were introduced in the upper plate design. However, the main novelty in reactor physics and operational features of the reactor was the installation of a pulse rod. Having no previous operational experience in pulsing, a detailed and systematic sequence of tests was defined in order to check the predicted design parameters of the reactor with measurements. The following experiments are treated in this paper: initial criticality, excess reactivity measurements, control rod worth measurement, fuel temperature distribution, fuel temperature reactivity coefficient, pulse parameters measurement (peak power, prompt energy, peak temperature). Flux distributions in steady state and pulse mode were measured as well, however, they are treated only briefly due to the volume of the results. The experiments were performed with completely fresh fuel of 12 w% enriched Standard Stainless Steel type. The core configuration was uniform (one fuel element type, including fuelled followers) and compact (no irradiation channels or gaps), as such being particularly convenient for testing the computer codes for TRIGA reactor calculations. Comparison of analytical predictions, obtained with WIMS, SLXTUS, TRIGAP and PULSTRI codes to measured values showed agreement within the error of the measurement and calculation. The paper has the following contents: 1. Introduction; 2. Steady State Experiments; 2.1. Core loading and critical experiment; 2.2. Flux range determination for tests at zero power; 2.3. Digital reactivity meter checkout; 2.4. Control rod worth measurements; 2.5. Excess reactivity measurement; 2.6. Thermal power calibration; 2

  5. 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)

  6. Enhanced In-Pile Instrumentation at the Advanced Test Reactor

    Science.gov (United States)

    Rempe, Joy L.; Knudson, Darrell L.; Daw, Joshua E.; Unruh, Troy; Chase, Benjamin M.; Palmer, Joe; Condie, Keith G.; Davis, Kurt L.

    2012-08-01

    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 provides an update on 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 real-time 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.

  7. Study on the leak rate test for HANARO reactor building

    International Nuclear Information System (INIS)

    Choi, Y. S.; Kim, Y. K.; Kim, M. J.; Park, J. M.; Woo, J. S.

    2002-01-01

    The reactor building of HANARO adopts the confinement concept, which allows a certain amount of air leakage. In order to restrict the air leakage through the confinement boundary, negative pressure of at least 2.5 mmWG is maintained in normal operating condition while maintaining 25 mmWG of negative pressure in abnormal condition, the inside air filtered by a train of charcoal filter is released to the atmosphere through the stack. In this situation, if the emergency ventilation system is not operable, the reactor building is isolated from the outside then the trapped air inside will be leaked out through the building by ground release concept. As the leak rate may be affected by an effect of wind velocity outside the reactor building, the air tightness of confinement should be maintained to limit the leak rate below the allowable value. The local leak rate test method was used since the beginning of the commissioning until July 1999. However it has been pointed out as a defect that the method is so susceptible to the change of temperature and atmospheric pressure during testing. For more accurate leak rate testing, we have introduced a new test method. We have periodically carried out the new leak rate testing and the results indicate that the bad effect by the temperature and atmospheric pressure change is considerably reduced, which gives more stable leak rate measurement

  8. Design of high temperature Engineering Test Reactor (HTTR)

    International Nuclear Information System (INIS)

    Saito, Shinzo; Tanaka, Toshiyuki; Sudo, Yukio

    1994-09-01

    Construction of High Temperature Engineering Test Reactor (HTTR) is now underway to establish and upgrade basic technologies for HTGRs and to conduct innovative basic research at high temperatures. The HTTR is a graphite-moderated and helium gas-cooled reactor with 30 MW in thermal output and outlet coolant temperature of 850degC for rated operation and 950degC for high temperature test operation. It is planned to conduct various irradiation tests for fuels and materials, safety demonstration tests and nuclear heat application tests. JAERI received construction permit of HTTR reactor facility in February 1990 after 22 months of safety review. This report summarizes evaluation of nuclear and thermal-hydraulic characteristics, design outline of major systems and components, and also includes relating R and D result and safety evaluation. Criteria for judgment, selection of postulated events, major analytical conditions for anticipated operational occurrences and accidents, computer codes used in safety analysis and evaluation of each event are presented in the safety evaluation. (author)

  9. Tests of the RBMK-1500 reactor fuel assemblies in the Leningrad reactor

    International Nuclear Information System (INIS)

    Aden, V.C.; Varovin, I.A.; Vorontsov, B.A.

    1981-01-01

    Test of fuel assemblies of the RBMK-1500 reactor is conducted in the reactor of the Leningrad NPP unit 2 for proving the calculational values of critical power of the RBMK-1500 reactor fuel assemblies adopted in design. The experiment presupposes the maximal approximation of the fuel assembly operation parameters to the calculational critical parameters without bringing into the mode of heat transfer crisis. The experiments are carried out at 500, 850 and 900 MW(el) of the reactor. The maximal channel power made up 472 kW at 20.5 t/h coolant flow rate and 49% mass steam content at the outlet of the channel. It was concluded that there was supply up to the heat transfer crisis in all the investigated modes. Data of temperature measurings of the fuel element cans, readings of the devices of the failure control system of the fuel element cans and external inspection of the assemblies after the tests testify to it [ru

  10. Model tests and numerical analysis on restoring force characteristics of reactor buildings

    International Nuclear Information System (INIS)

    Uchiyama, Y.; Suzuki, S.; Akino, K.

    1987-01-01

    Seismic shear walls of nuclear reactor buildings are composed of cylindrical, truncated cone-shape, box-shape, irregular polygonal walls or its combination and they are generally heavily reinforced concrete (RC) walls. So the elasto-plastic behaviors of those RC structures in ultimate regions have many unsolved and may be considered as especially important factors for explaining nonlinear response of nuclear reactor buildings. Following these research demands, the authors have prepared a nonlinear F.E.M. code called ''SANREF'' and made an extensive study for the restoring force characteristics of the inner concrete structures (I/C) of a PWR-type containment vessel and the principal seismic shear walls of a BWR-type reactor building by some series of reduced model tests and simulation analysis for the tests results. The detailed objectives of this study can be summarized as follows: (1) Examine the effectiveness of the configurations of shear walls, reinforcement ratios, shear span ratios (M/Qd) and vertical axial stress by ''partial model test'' which simulates some independent shear walls of the PWR-type and BWR-type reactor buildings. (2) Obtain fundamental data of restoring force characteristics of the complex shaped RC structures by ''composite model test'' which models are composed of the partial model test specimens. (3) Verify the applicability of analytical methods and constitutive modelings in SANREF code for complex shaped RC structures through nonlinear simulation analysis for the composite model test

  11. Decontamination and Decommissioning of the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Perry, E.; Chrzanowski, J.; Rule, K.; Viola, M.; Williams, M.; Strykowsky, R.

    1999-01-01

    The Tokamak Fusion Test Reactor (TFTR) is a one-of-a-kind, tritium-fueled fusion research reactor that ceased operation in April 1997. The Decontamination and Decommissioning (D and D) of the TFTR is scheduled to occur over a period of three years beginning in October 1999. This is not a typical Department of Energy D and D Project where a facility is isolated and cleaned up by ''bulldozing'' all facility and hardware systems to a greenfield condition. The mission of TFTR D and D is to: (a) surgically remove items which can be re-used within the DOE complex, (b) remove tritium contaminated and activated systems for disposal, (c) clear the test cell of hardware for future reuse, (d) reclassify the D-site complex as a non-nuclear facility as defined in DOE Order 420.1 (Facility Safety) and (e) provide data on the D and D of a large magnetic fusion facility. The 100 cubic meter volume of the donut-shaped reactor makes it the second largest fusion reactor in the world. The record-breaking deuterium-tritium experiments performed on TFTR resulted in contaminating the vacuum vessel with tritium and activating the materials with 14 Mev neutrons. The total tritium content within the vessel is in excess of 7,000 Curies while dose rates approach 75 mRem/hr. These radiological hazards along with the size and shape of the Tokamak present a unique and challenging task for dismantling

  12. The fast breeder reactor and the electricity provision system

    International Nuclear Information System (INIS)

    1975-01-01

    Due to inflexible planning, the electricity provision system is bound to generate demands for big centralized power plants, that is, nuclear power plants. A lack of feed-back from producer to consumer results in a technical distribution net-work that cannot react to big changes in supply and demand. This makes it particularly difficult to include alternative energy sources with fluctuating energy production

  13. Very High Efficiency Reactor (VHER) Concepts for Electrical Power Generation and Hydrogen Production

    International Nuclear Information System (INIS)

    PARMA JR, EDWARD J.; PICKARD, PAUL S.; SUO-ANTTILA, AHTI JORMA

    2003-01-01

    The goal of the Very High Efficiency Reactor study was to develop and analyze concepts for the next generation of nuclear power reactors. The next generation power reactor should be cost effective compared to current power generation plant, passively safe, and proliferation-resistant. High-temperature reactor systems allow higher electrical generating efficiencies and high-temperature process heat applications, such as thermo-chemical hydrogen production. The study focused on three concepts; one using molten salt coolant with a prismatic fuel-element geometry, the other two using high-pressure helium coolant with a prismatic fuel-element geometry and a fuel-pebble element design. Peak operating temperatures, passive-safety, decay heat removal, criticality, burnup, reactivity coefficients, and material issues were analyzed to determine the technical feasibility of each concept

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

    International Nuclear Information System (INIS)

    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

  15. Safety analysis calculations for research and test reactors

    International Nuclear Information System (INIS)

    Chen, S.Y.; MacDonald, R.; MacFarlane, D.

    1983-01-01

    Safety issues for the two general types of reactors, i.e., the plate-type (MTR-type) reactor and the rod-type (TRIGA-type) reactor, resulting from the changes associated with LEU vs HEU fuels, are explored. The plate-type fuels are typically uranium aluminide (UAl/sub x/) compounds dispersed in aluminum and clad with aluminum. Moderation is provided by the water coolant. Self shut-down reactivity coefficients with HEU fuel are entirely a result of coolant heating, whereas with LEU fuel there is an additional shut down contribution provided by the direct heating of the fuel due to the Doppler coefficient. In contrast, the rod-type (TRIGA) fuels are mixtures of zirconium hydride, uranium, and erbium. This fuel mixture is formed into rods (approx. 1 cm diameter) and clad with stainless steel or Incoloy. In the TRIGA fuel the self-shutdown reactivity is more complex, depending on heating of the fuel rather than the coolant. Results of transient calculations performed with existing computer codes, most suited for each type of reactor, are presented

  16. Trends in large-scale testing of reactor structures

    International Nuclear Information System (INIS)

    Blejwas, T.E.

    2003-01-01

    Large-scale tests of reactor structures have been conducted at Sandia National Laboratories since the late 1970s. This paper describes a number of different large-scale impact tests, pressurization tests of models of containment structures, and thermal-pressure tests of models of reactor pressure vessels. The advantages of large-scale testing are evident, but cost, in particular limits its use. As computer models have grown in size, such as number of degrees of freedom, the advent of computer graphics has made possible very realistic representation of results - results that may not accurately represent reality. A necessary condition to avoiding this pitfall is the validation of the analytical methods and underlying physical representations. Ironically, the immensely larger computer models sometimes increase the need for large-scale testing, because the modeling is applied to increasing more complex structural systems and/or more complex physical phenomena. Unfortunately, the cost of large-scale tests is a disadvantage that will likely severely limit similar testing in the future. International collaborations may provide the best mechanism for funding future programs with large-scale tests. (author)

  17. 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.

  18. Performance demonstration experience for reactor pressure vessel shell ultrasonic testing

    International Nuclear Information System (INIS)

    Zado, V.

    1998-01-01

    The most ultrasonic testing techniques used by many vendors for pressurized water reactor (PWR) examinations were based on American Society of Mechanical Engineers 'Boiler and Pressurized Vessel Code' (ASME B and PV Code) Sections XI and V. The Addenda of ASME B and PV Code Section XI, Edition 1989 introduced Appendix VIII - 'Performance Demonstration for Ultrasonic Examination Systems'. In an effort to increase confidence in performance of ultrasonic testing of the operating nuclear power plants in United States, the ultrasonic testing performance demonstration examination of reactor vessel welds is performed in accordance with Performance Demonstration Initiative (PDI) program which is based on ASME Code Section XI, Appendix VIII requirements. This article provides information regarding extensive qualification preparation works performed prior EPRI guided performance demonstration exam of reactor vessel shell welds accomplished in January 1997 for the scope of Appendix VIII, Supplements IV and VI. Additionally, an overview of the procedures based on requirements of ASME Code Section XI and V in comparison to procedure prepared for Appendix VIII examination is given and discussed. The samples of ultrasonic signals obtained from artificial flaws implanted in vessel material are presented and results of ultrasonic testing are compared to actual flaw sizes. (author)

  19. Advanced In-pile Instrumentation for Material and Test Reactors

    International Nuclear Information System (INIS)

    Rempe, J.L.; Knudson, D.L.; Daw, J.E.; Unruh, T.C.; Chase, B.M.; Davis, K.L.; Palmer, A.J.; Schley, R.S.

    2013-06-01

    The US Department of Energy sponsors the Advanced Test Reactor (ATR) National Scientific User Facility (NSUF) program to promote U.S. research 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, advancing U.S. energy security needs. A key component of the ATR NSUF effort is to design, develop, and deploy new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. This paper describes the strategy developed by the Idaho National Laboratory (INL) for identifying instrumentation needed for ATR irradiation tests and the program initiated to obtain these sensors. New sensors developed from this effort are identified; and the progress of other development efforts is summarized. As reported in this paper, INL staff is currently involved in several tasks to deploy real-time length and flux detection sensors, and efforts have been initiated to develop a crack growth test rig. Tasks evaluating 'advanced' technologies, such as fiber-optics based length detection and ultrasonic thermometers are also underway. In addition, specialized sensors for real-time detection of temperature and thermal conductivity are not only being provided to NSUF reactors, but are also being provided to several international test reactors. (authors)

  20. Advanced In-Pile Instrumentation for Materials Testing Reactors

    Science.gov (United States)

    Rempe, J. L.; Knudson, D. L.; Daw, J. E.; Unruh, T. C.; Chase, B. M.; Davis, K. L.; Palmer, A. J.; Schley, R. S.

    2014-08-01

    The U.S. Department of Energy sponsors the Advanced Test Reactor (ATR) National Scientific User Facility (NSUF) program to promote U.S. research 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, advancing U.S. energy security needs. A key component of the ATR NSUF effort is to design, develop, and deploy new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. This paper describes the strategy developed by the Idaho National Laboratory (INL) for identifying instrumentation needed for ATR irradiation tests and the program initiated to obtain these sensors. New sensors developed from this effort are identified, and the progress of other development efforts is summarized. As reported in this paper, INL researchers are currently involved in several tasks to deploy real-time length and flux detection sensors, and efforts have been initiated to develop a crack growth test rig. Tasks evaluating `advanced' technologies, such as fiber-optics based length detection and ultrasonic thermometers, are also underway. In addition, specialized sensors for real-time detection of temperature and thermal conductivity are not only being provided to NSUF reactors, but are also being provided to several international test reactors.

  1. Advanced Test Reactor National Scientific User Facility Partnerships

    International Nuclear Information System (INIS)

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

    2012-01-01

    In 2007, the United States Department of Energy designated the Advanced Test Reactor (ATR), located at Idaho National Laboratory, as a National Scientific User Facility (NSUF). This designation made test space within the ATR and post-irradiation examination (PIE) equipment at INL available for use by researchers via a proposal and peer review process. The goal of the ATR NSUF is to provide researchers with the best ideas access to the most advanced test capability, regardless of the proposer's physical location. Since 2007, the ATR NSUF has expanded its available reactor test space, and obtained access to additional PIE equipment. Recognizing that INL may not have all the desired PIE equipment, or that some equipment may become oversubscribed, the ATR NSUF established a Partnership Program. This program enables and facilitates user access to several university and national laboratories. So far, seven universities and one national laboratory have been added to the ATR NSUF with capability that includes reactor-testing space, PIE equipment, and ion beam irradiation facilities. With the addition of these universities, irradiation can occur in multiple reactors and post-irradiation exams can be performed at multiple universities. In each case, the choice of facilities is based on the user's technical needs. Universities and laboratories included in the ATR NSUF partnership program are as follows: (1) Nuclear Services Laboratories at North Carolina State University; (2) PULSTAR Reactor Facility at North Carolina State University; (3) Michigan Ion Beam Laboratory (1.7 MV Tandetron accelerator) at the University of Michigan; (4) Irradiated Materials at the University of Michigan; (5) Harry Reid Center Radiochemistry Laboratories at University of Nevada, Las Vegas; (6) Characterization Laboratory for Irradiated Materials at the University of Wisconsin-Madison; (7) Tandem Accelerator Ion Beam. (1.7 MV terminal voltage tandem ion accelerator) at the University of Wisconsin

  2. The BR2 materials testing reactor. Past, ongoing and under-study upgradings

    Energy Technology Data Exchange (ETDEWEB)

    Baugnet, J M; Roedt, Ch de; Gubel, P; Koonen, E [Centre d' Etude de I' Energie Nucleaire, Studiecentrum voor Kernenergie, C.E.N./S.C.K., Mol (Belgium)

    1990-05-01

    The BR2 reactor (Mol, Belgium) is a high-flux materials testing reactor. The fuel is 93% {sup 235}U enriched uranium. The nominal power ranges from 60 to 100 MW. The main features of the design are the following: 1) maximum neutron flux, thermal: 1.2 x 10{sup 15} n/cm{sup 2} s; fast (E > 0.1 MeV) : 8.4 x 10{sup 14} n /cm{sup 2} s; 2) great flexibility of utilization: the core configuration and operation mode can be adapted to the experimental loading; 3) neutron spectrum tailoring; 4) availability of five 200 mm diameter channels besides the standard channels (84 mm diameter); 5) access to the top and bottom covers of the reactor authorizing the irradiation of loops. The reactor is used to study the behaviour of fuel elements and structural materials intended for future nuclear power stations of several types (fission and fusion). Irradiations are carried out in connection with performance tests up to very high burn-up or neutron fluence as well as for safety experiments, power cycling experiments, and generally speaking, tests under off-normal conditions. Irradiations for nuclear transmutation (production of high specific activity radio-isotopes and transplutonium elements), neutron-radiography, use of beam tubes for physics studies, and gamma irradiations are also carried out. The BR2 is used in support of Belgian programs, at the request of utilities, industry and universities and in the framework of international agreements. The paper reviews the past and ongoing upgrading and enhancement of reactor capabilities as well as those under study or consideration, namely with regard to: reactor equipment, fuel elements, irradiation facilities, reactor operation conditions and long-term strategy. (author)

  3. Thermal Hydraulic Integral Effect Tests for Pressurized Water Reactors

    International Nuclear Information System (INIS)

    Baek, Won Pil; Song, C. H.; Kim, Y. S.

    2007-02-01

    The objectives of the project are to construct a thermal-hydraulic integral effect test facility and to perform the tests for design, operation, and safety regulation of pressurized water reactors. In the first phase of this project (1997.8∼2002.3), the basic technology for thermal-hydraulic integral effect tests was established and the basic design of the test facility was accomplished. In the second phase (2002.4∼2005.2), an optimized design of the ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) was established and the construction of the facility was almost completed. In the third phase (2005.3∼2007.2), the construction and commission tests of the ATLAS are to be completed and some first-phase tests are to be conducted

  4. Core Seismic Tests for a Sodium-Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Gyeong Hoi; Lee, J. H

    2007-01-15

    This report describes the results of the comparison of the core seismic responses between the test and the analysis for the reduced core mock-up of a sodium-cooled fast reactor to verify the FAMD (Fluid Added Mass and Damping) code and SAC-CORE (Seismic Analysis Code for CORE) code, which implement the application algorithm of a consistent fluid added mass matrix including the coupling terms. It was verified that the narrow fluid gaps between the duct assemblies significantly affect the dynamic characteristics of the core duct assemblies and it becomes stronger as a number of duct increases within a certain level. As conclusion, from the comparison of the results between the tests and the analyses, it is verified that the FAMD code and the SAC-CORE code can give an accurate prediction of a complex core seismic behavior of the sodium-cooled fast reactor.

  5. Removal of the Materials Test Reactor overhead working reservoir

    International Nuclear Information System (INIS)

    Lunis, B.C.

    1975-10-01

    Salient features of the removal of an excessed contaminated facility, the Materials Test Reactor (MTR) overhead working reservoir (OWR) from the Test Reactor Area to the Radioactive Waste Management Complex at the Idaho National Engineering Laboratory are described. The 125-ton OWR was an overhead 160,000-gallon-capacity tank approximately 193 feet high which supplied cooling water to the MTR. Radiation at ground level beneath the tank was 5 mR/hr and approximately 600 mR/hr at the exterior surface of the tank. Sources ranging from 3 R/hr to in excess of 500 R/hr exist within the tank. The tank interior is contaminated with uranium, plutonium, and miscellaneous fission products. The OWR was lowered to ground level with the use of explosive cutters. Dismantling, decontamination, and disposal were performed by Aerojet Nuclear Company maintenance forces

  6. Fuels for research and test reactors, status review: July 1982

    International Nuclear Information System (INIS)

    Stahl, D.

    1982-12-01

    A thorough review is provided on nuclear fuels for steady-state thermal research and test reactors. The review was conducted to provide a documented data base in support of recent advances in research and test reactor fuel development, manufacture, and demonstration in response to current US policy on availability of enriched uranium. The review covers current fabrication practice, fabrication development efforts, irradiation performance, and properties affecting fuel utilization, including thermal conductivity, specific heat, density, thermal expansion, corrosion, phase stability, mechanical properties, and fission-product release. The emphasis is on US activities, but major work in Europe and elsewhere is included. The standard fuel types discussed are the U-Al alloy, UZrH/sub x/, and UO 2 rod fuels. Among new fuels, those given major emphasis include H 3 Si-Al dispersion and UO 2 caramel plate fuels

  7. 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.

  8. General Vehicle Test Plan (GVTP) for Urban Rail Transit Cars

    Science.gov (United States)

    1977-09-01

    The General Vehicle Test Plan provides a system for general vehicle testing and for documenting and utilizing data and information in the testing of urban rail transit cars. Test procedures are defined for nine categories: (1) Performance; (2) Power ...

  9. Scheduling and recording of reactor maintenance and testing by computer

    International Nuclear Information System (INIS)

    Gray, P.L.

    1975-01-01

    The use of a computer program, Maintenance Information and Control (MIAC), at the Savannah River Laboratory (SRL) assists a small operating staff in maintaining three research reactors and a subcritical facility. The program schedules and defines preventive maintenance, schedules required periodic tests, logs repair and cost information, specifies custodial and service responsibilities, and provides equipment maintenance history, all with a minimum of record-keeping

  10. Fracture toughness testing of a reactor grade graphite

    Energy Technology Data Exchange (ETDEWEB)

    Roeding, M.; Klein, G.; Schiffers, H.; Nickel, H.

    1976-03-15

    Fracture mechanics is a well established tool for the assessment of brittle fracture in metallic structural materials. In this paper an attempt is made to apply fracture mechanics to a reactor-grade graphite. The effect of several test parameters on the stress intensity factor was measured; this was found to lie in the range 25 and 50 N/mm/sup -3/2/. The results are discussed in terms of the well known mechanical characteristics of graphite.

  11. Facility for in-reactor creep testing of fuel cladding

    International Nuclear Information System (INIS)

    Kohn, E.; Wright, M.G.

    1976-11-01

    A biaxial stress creep test facility has been designed and developed for operation in the WR-1 reactor. This report outlines the rationale for its design and describes its construction and the operating experience with it. The equipment is optimized for the determination of creep data on CANDU fuel cladding. Typical results from Zr-2.5 wt% Nb fuel cladding are used to illustrate the accuracy and reliability obtained. (author)

  12. Diamond Wire Cutting of the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Keith Rule; Erik Perry; Robert Parsells

    2003-01-01

    The Tokamak Fusion Test Reactor (TFTR) is a one-of-a-kind, tritium-fueled fusion research reactor that ceased operation in April 1997. As a result, decommissioning commenced in October 1999. The 100 cubic meter volume of the donut-shaped reactor makes it the second largest fusion reactor in the world. The deuterium-tritium experiments resulted in contaminating the vacuum vessel with tritium and activating the materials with 14 MeV neutrons. The total tritium content within the vessel is in excess of 7,000 Curies, while dose rates approach 50 mRem/hr. These radiological hazards along with the size of the tokamak present a unique and challenging task for dismantling. Engineers at the Princeton Plasma Physics Laboratory (PPPL) decided to investigate an alternate, innovative approach for dismantlement of the TFTR vacuum vessel: diamond wire cutting technology. In August 1999, this technology was successfully demonstrated and evaluated on vacuum vessel surrogates. Subsequently, the technology was improved and redesigned for the actual cutting of the vacuum vessel. Ten complete cuts were performed in a 6-month period to complete the removal of this unprecedented type of DandD (Decontamination and Decommissioning) activity

  13. Evolution of general design requirements for french pressurized water reactors

    International Nuclear Information System (INIS)

    Gros, G.; Jalouneix, J.; Rollinger, F.

    1988-10-01

    The design of French pressurized water reactors is based first on deterministic principles, using the well-known defense in depth concept. This safety approach, basically reflected current American practice at that time, which consisted notably in designing engineered safeguard systems capable of limiting the consequences of accidents assumed to be credible despite the preventive measures taken. Further reflections have led to complete this approach, resulting in modifications to regulatory practice, mainly related to better practical assimilation of the problems arising during plant unit operation and reactor control after an accident and to the determination to enhance the overall consistency of the safety approach. As regards system redundancy, it should be noted that common cause failures can result in the total loss of a redundant system. System redundancy aspects will be dealt with in Chapter 2. As regards study of design basis accidents, attention was focused on the human intervention stage following automatic activation of protection and safeguard systems. This resulted, for all plant units, in the revision of operating procedures, accompanied by examination of the means required for their implementation. These subjects will be discussed in Chapter 3. Finally, as regards equipment classification, the range of equipment subjected to particular requirements, formerly limited to design basis safety classified equipment, was enlarged to include important for safety equipment. This subject will be dealt with in Chapter 5

  14. Electrical, thermal and abusive tests on lithium thionyl chloride cells

    Science.gov (United States)

    Frank, H. A.

    1980-04-01

    Electrical characterizations, thermal characterizations, and outer limits tests of lithium thionyl chloride cells are discussed. Graphs of energy density vs power density and heat rate vs time are presented along with results of forced reversal and high rate discharge tests.

  15. Laboratory scale tests of electrical impedence tomography

    International Nuclear Information System (INIS)

    Binley, A; Daily, W; LaBredcque, D; Ramirez, A.

    1998-01-01

    Electrical impedance tomographs (magnitude and phase) of known, laboratory-scale targets are reported. Three methods are used to invert electrical impedance data and their tomographs compared. The first method uses an electrical resistance tomography (ERT) algonthm (designed for DC resistivity inversion) to perform impedance magnitude inversion and a linearized perturbation approach (PA) to invert the imaginary part. The second approximate method compares ERT magnitude inversions at two frequencies and uses the frequency effect (FE) to compute phase tomographs. The third approach, electrrcal impedance tomography (EIT), employs fully complex algebra to account for the real and imaginary components of electrical impedance data. The EIT approach provided useful magnitude and phase images for the frequency range of 0.0625 to 64 Hz; images for higher frequencies were not reliable. Comparisons of the ERT and EIT magnitude images show that both methods provided equivalent results for the water blank, copper rod and PVC rod targets. The EIT magnitude images showed better spatial resolutron for a sand-lead mixture target. Phase images located anomalies of both high and low contrast IP and provided better spatial resolution than the magnitude images. When IP was absent from the data, the EIT algorithm reconstructed phase values consistent with the data noise levels

  16. Seismic proving test of PWR reactor containment vessel

    International Nuclear Information System (INIS)

    Akiyama, H.; Yoshikawa, T.; Tokumaru, Y.

    1987-01-01

    The seismic reliability proving tests of nuclear power plant facilities are carried out by Nuclear Power Engineering Test Center (NUPEC), using the large-scale, high-performance vibration of Tadotsu Engineering Laboratory, and sponsored by the Ministry of International Trade and Industry (MITI). In 1982, the seismic reliability proving test of PWR containment vessel started using the test component of reduced scale 1/3.7 and the test component proved to have structural soundness against earthquakes. Subsequently, the detailed analysis and evaluation of these test results were carried out, and the analysis methods for evaluating strength against earthquakes were established. Whereupon, the seismic analysis and evaluation on the actual containment vessel were performed by these analysis methods, and the safety and reliability of the PWR reactor containment vessel were confirmed

  17. Study of reactor Brayton power systems for nuclear electric spacecraft

    Science.gov (United States)

    1979-01-01

    The feasibility of using Brayton power systems for nuclear electric spacecraft was investigated. The primary performance parameters of systems mass and radiator area were determined for systems from 100 to 1000 kW sub e. Mathematical models of all system components were used to determine masses and volumes. Two completely independent systems provide propulsion power so that no single-point failure can jeopardize a mission. The waste heat radiators utilize armored heat pipes to limit meteorite puncture. The armor thickness was statistically determined to achieve the required probability of survival. A 400 kW sub e reference system received primary attention as required by the contract. The components of this system were defined and a conceptual layout was developed with encouraging results. An arrangement with redundant Brayton power systems having a 1500 K (2240 F) turbine inlet temperature was shown to be compatible with the dimensions of the space shuttle orbiter payload bay.

  18. Applications of nuclear reactor power systems to electric propulsion missions.

    Science.gov (United States)

    Schaupp, R. W.; Sawyer, C. D.

    1971-01-01

    The performance of nuclear electric propulsion systems (NEP) has been evaluated for a wide variety of missions in an attempt to establish the commonality of NEP system requirements. Emphasis was given to those requirements and system characteristics that serve as guidelines for current technology development programs. Various interactions and tradeoffs between NEP system and mission parameters are described. The results show that the most significant factors in selecting NEP system size are launch mode (direct or spiral escape) and, to a weaker extent, launch vehicle capability. Other factors such as mission, payload, and thrust time constraints, have little influence, thus allowing one NEP system to be used for many missions. The results indicated that a 100 kWe NEP would be suitable for most direct escape missions and a 250 kWe NEP system would be suitable for more demanding missions that use the spiral escape mode.

  19. Expert systems for analysis and management assistance of nuclear reactor electrical power supplies

    International Nuclear Information System (INIS)

    Evrard, J.M.; Souchet, Y.

    1988-01-01

    Electrical power supplies of nuclear plants are very complex systems. They are studied in many ways: failure consequences, probabilistic risk assessement, failure diagnosis, corrective actions in case of incident ... Knowledge base technology (expert systems) is very suited to solve these problems. A common structural representation can generate specific functional representations; thus we get coherency and easy evolution. This paper shows a two facet methodology: the plant description (both static and dynamic) and reasoning about it. Current applications are developed for pressurized water reactors and gas cooled reactors, using the SPIRAL expert system shell buil at CEA [fr

  20. Device for providing a leak-tight penetration for electric cables through a reactor vault roof

    International Nuclear Information System (INIS)

    Eyral, M.; Mahe, A.

    1979-01-01

    The device for providing a cable penetration through the vault roof of a liquid sodium cooled fast reactor comprises a vertical tube closed at the top end by a flange-plate. Electric cables connected to measuring and detecting instruments are passed through the flange-plate which is joined to the reactor vault roof in leak-tight manner and enclosed within a removable hood. At least one horizontal plate is mounted within the vertical tube and provided with orifices for the leak-tight passage of the cables. Cable storage reels are placed within the tube and can be locked in position or released by controlled mechanical means

  1. Pressure test method for reactor pressure vessel in construction field

    International Nuclear Information System (INIS)

    Takeda, Masakado; Ushiroda, Koichi; Miyahara, Ryohei; Takano, Hiroshi; Matsuura, Tadashi; Sato, Keiya.

    1998-01-01

    Plant constitutional parts as targets of both of a primary pressure test and a secondary pressure test are disposed in communication with a reactor pressure vessel, and a pressure of the primary pressure test is applied to the targets of both tests, so that the primary pressure test and the second pressure test are conducted together. Since the number of pressure tests can be reduced to promote construction, and the number of workers can also be reduced. A pressure exceeding the maximum pressure upon use is applied to the pressure vessel after disposing the incore structures, to continuously conduct the primary pressure test and the secondary pressure test joined together and an incore flowing test while closing the upper lid of the pressure vessel as it is in the construction field. The number of opening/closing of the upper lid upon conducting every test can be reduced, and since the pressure resistance test is conducted after arranging circumference conditions for the incore flowing test, the tests can be conducted collectively also in view of time. (N.H.)

  2. 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.

  3. Design, in-sodium testing and performance evaluation of annular linear induction pump for a sodium cooled fast reactor

    International Nuclear Information System (INIS)

    Nashine, B.K.; Rao, B.P.C.

    2014-01-01

    Highlights: • Derivation of applicable design equations. • Design of an annular induction pump based on these equations. • Testing of the designed pump in a sodium test facility. • Performance evaluation of the designed pump. - Abstract: Annular linear induction pumps (ALIPs) are used for pumping electrically conducting liquid metals. These pumps find wide application in fast reactors since the coolant in fast reactors is liquid sodium which a good conductor of electricity. The design of these pumps is usually done using equivalent circuit approach in combination with numerical simulation models. The equivalent circuit of ALIP is similar to that of an induction motor. This paper presents the derivation of equivalent circuit parameters using first principle approach. Sodium testing of designed ALIP using the equivalent circuit approach is also described and experimental results of the testing are presented. Comparison between experimental and analytical calculations has also been carried out. Some of the reasons for variation have also been listed in this paper

  4. Proceedings of the 4th international symposium on material testing reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ishihara, Masahiro; Suzuki, Masahide [Japan Atomic Energy Agency, Oarai Research and Development Center, Oarai, Ibaraki (Japan)

    2012-03-15

    This report is the Proceedings of the fourth International Symposium on Material Testing Reactors hosted by Japan Atomic Energy Agency (JAEA). The first symposium was held on 2008, at the Oarai Research and Development Center of JAEA, the second, 2009, Idaho National Laboratory (INL) of United States and the third 2010, Nuclear Research Institute (NRI) in Czech Republic to exchange information for deep mutual understanding of material testing reactors. The fourth symposium was originally scheduled to be held INVAP in Argentina. However, the aftermath of volcanic explosion at Chili forced the symposium to change place. Total 111 participants attended from Argentina, Belgium, France, Germany, Indonesia, Malasia, Korea, South Africa, Switzerland, the United State and Japan. This symposium addressed the general topics of 'status and future plan of material testing reactors', 'advancement of irradiation technology', 'expansion of industry use(RI)', 'facility, upgrade, aging management', 'new generation MTR', 'advancement of PIE technology', 'development of advanced driver fuel', and 'nuclear human resource development(HRD) for next generation', and 39 presentations were made. Furthermore, three topics, 'Necessity of cooperation for Mo-99 production by (n,gamma) reaction', 'Necessity of standardization of irradiation technology' and 'Conceptual design of next generation materials testing reactor by collaboration', were selected and discussed. (author)

  5. Proceedings of the 4th international symposium on material testing reactors

    International Nuclear Information System (INIS)

    Ishihara, Masahiro; Suzuki, Masahide

    2012-03-01

    This report is the Proceedings of the fourth International Symposium on Material Testing Reactors hosted by Japan Atomic Energy Agency (JAEA). The first symposium was held on 2008, at the Oarai Research and Development Center of JAEA, the second, 2009, Idaho National Laboratory (INL) of United States and the third 2010, Nuclear Research Institute (NRI) in Czech Republic to exchange information for deep mutual understanding of material testing reactors. The fourth symposium was originally scheduled to be held INVAP in Argentina. However, the aftermath of volcanic explosion at Chili forced the symposium to change place. Total 111 participants attended from Argentina, Belgium, France, Germany, Indonesia, Malasia, Korea, South Africa, Switzerland, the United State and Japan. This symposium addressed the general topics of 'status and future plan of material testing reactors', 'advancement of irradiation technology', 'expansion of industry use(RI)', 'facility, upgrade, aging management', 'new generation MTR', 'advancement of PIE technology', 'development of advanced driver fuel', and 'nuclear human resource development(HRD) for next generation', and 39 presentations were made. Furthermore, three topics, 'Necessity of cooperation for Mo-99 production by (n,gamma) reaction', 'Necessity of standardization of irradiation technology' and 'Conceptual design of next generation materials testing reactor by collaboration', were selected and discussed. (author)

  6. Advanced marine reactor MRX and application to nuclear barge supplying electricity and heat

    International Nuclear Information System (INIS)

    Ishida, Toshihisa; Kusunoki, Tsuyoshi; Odano, Naoteru; Yoritsune, Tsutomu; Fukuhara, Yoshifumi; Ochiai, Masa-aki

    2000-01-01

    The basic design concept of an advanced marine reactor MRX has been established with adoption of several new technologies. The MRX is an integral-type PWR with 100 MWt aimed basically for use of ship propulsion. Adoption of a water-filled containment together with the integral type reactor makes the reactor light-weight and compact greatly. A engineered safety system is a simplified passive system, function of which is confirmed by the safety analysis. The MRX can be applied to an energy supply system of electricity and heat co-generation by installing it on a barge. Concept of a nuclear barge with the MRX of 334 MWt output is presented for use of supplying electricity, fresh water and hot water. Combined system of electric generation and desalination with the RO process can deliver variable output of electricity and fresh water according a demand. Latent heat of the exhausted steam from the turbine can be used effectively to raise the temperature of cold water as heat supply. (author)

  7. Increased sharing of renewable energies in the electricity production system: what impact on the reactor fleet?

    International Nuclear Information System (INIS)

    Cany, C.; Devezeaux de Lavergne, J.G.; Mansilla, C.; Mathonniere, G.

    2017-01-01

    This article presents the flexibility of an individual reactor and of the complete fleet of reactors as a means to cope with the variability of renewable energies like solar or wind energies. Flexibility means the ability for load following and this ability is limited by both safety rules and limits on the release of radionuclides in the environment. The flexibility of the fleet depends on individual reactor flexibility but also on organisational and economic constraints. The participation of a reactor to load following depends on: its availability (not in maintenance or testing phase), its position in the cycle, the positioning of its scheduled shutdowns and the minimization of the volume of effluents. The study presents the future need of flexibility for the reactor fleet as the shares of wind and solar energies increase in the French energy mix. (A.C.)

  8. In-pile instrumentation improvements for fuel irradiations in test reactor

    International Nuclear Information System (INIS)

    Blanc, J.Y.; Bernard, J.L.; Estrade, J.; Geoffroy, G.

    1996-01-01

    Knowledge of fuel limits and safety margins in normal and off-normal transients in nuclear power plants remains a constant preoccupation for electricity producers and fuel manufacturers. Accurate determination of such limits, through fuel irradiation testing in the OSIRIS reactor at Saclay is closely linked to the reliability of appropriate instrumentation techniques. Two paths are currently followed to obtain short experimental rods: segmented fuel coming directly from power plants, or re-fabrication of rods in hot cells with our FABRICE process. It can be associated with instrumentation such as fuel centerline thermocouple in annular pellets, pressure transducer or fission gas release measurement by gamma-spectrometry using helium sweeping, in analytic experiments. Our present development, to be implemented in 1993, is the the centerline instrumentation of a fuel column with solid pellets. Inserting the thermocouple requires a cold drilling machine, using CO 2 freezing of broken UO 2 (with liquid nitrogen). During the fuel rod irradiation itself, we try to lower the uncertainties associated to power determination, using thermal balance or neutronic calibration, or even gamma spectrometry. A description of the new test train designed for the ISABELLE water loop in OSIRIS is given, with special emphasis on instrumentation: a LVDT for measuring fuel rod elongation and eventual clad failure, and increased number and better localization of thermocouples and SPDN. The third part is devoted to the measurements by optical microdensitometry of neutron radiographs of the fuel pellet dish modification after irradiation. Dishes are generally disappearing through thermal and mechanical deformation of the pellet, and this can eventually be modelized to better understand pellet-cladding mechanical interaction. (author). 3 refs, 5 figs

  9. The Advanced Test Reactor Irradiation Facilities and Capabilities

    International Nuclear Information System (INIS)

    S. Blaine Grover; Raymond V. Furstenau

    2007-01-01

    The Advanced Test Reactor (ATR) is one of the world's premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. The ATR is a very versatile facility with a wide variety of experimental test capabilities for providing the environment needed in an irradiation experiment. These different capabilities include passive sealed capsule experiments, instrumented and/or temperature-controlled experiments, and pressurized water loop experiment facilities. The ATR has enhanced capabilities in experiment monitoring and control systems for instrumented and/or temperature controlled experiments. The control systems utilize feedback from thermocouples in the experiment to provide a custom blended flowing inert gas mixture to control the temperature in the experiments. Monitoring systems have also been utilized on the exhaust gas lines from the experiment to monitor different parameters, such as fission gases for fuel experiments, during irradiation. ATR's unique control system provides axial flux profiles in the experiments, unperturbed by axially positioned control components, throughout each reactor operating cycle and over the duration of test programs requiring many years of irradiation. The ATR irradiation positions vary in diameter from 1.6 cm (0.625 inches) to 12.7 cm (5.0 inches) over an active core length of 122 cm (48.0 inches). Thermal and fast neutron fluxes can be adjusted radially across the core depending on the needs of individual test programs. This paper will discuss the different irradiation capabilities available and the cost/benefit issues related to each capability. Examples of different experiments will also be discussed to demonstrate the use of the capabilities and facilities at ATR for performing irradiation experiments

  10. Shaking table qualification tests of mechanical and electrical components

    International Nuclear Information System (INIS)

    Jurukovski, D.

    1993-01-01

    This presentation covers the experience of the Institute of Earthquake Engineering and Engineering Seismology, Skopje, Republic of Macedonia in seismic qualification of mechanical components by shaking table testing. The characteristics of the biaxial seismic and single component shaking tables used at the Institute are given. Some examples of the experience from performed test for reactor components are included

  11. Implosion and staging systems for a Scyllac Fusion Test Reactor

    International Nuclear Information System (INIS)

    Gribble, R.F.; Linford, R.K.; Thomassen, K.I.

    1976-01-01

    The implosion heating and adiabatic compression processes will be separated in future theta pinch devices. The circuit to achieve the fast implosion heating and power crowbar (staging) for the Scyllac Fusion Test Reactor is described here. The plasma is very tightly coupled to the circuit and presents a varying inductive load. Computer-aided circuit designs which achieve a programmed magnetic field waveform are described. The field approximates a two-step waveform, on-off-on, which is ideal for achieving the large initial plasma radius needed for stability. The components for the circuits have been developed and are being tested in experiments at Los Alamos

  12. Implosion and staging systems for a Scyllac fusion test reactor

    International Nuclear Information System (INIS)

    Gribble, R.F.; Linford, R.K.; Thomassen, K.I.

    1975-01-01

    The implosion heating and adiabatic compression processes will be separated in future theta pinch devices. The circuit to achieve the fast implosion heating and power crowbar (staging) for the Scyllac Fusion Test Reactor is described here. The plasma is very tightly coupled to the circuit and presents a varying inductive load. Computer-aided circuit designs which achieve a programmed magnetic field waveform are described. The field approximates a two-step waveform, on-off-on, which is ideal for achieving the large initial plasma radius needed for stability. The components for the circuits have been developed and are being tested in experiments at Los Alamos. (auth)

  13. Conceptual design study of a scyllac fusion test reactor

    International Nuclear Information System (INIS)

    Thomassen, K.I.

    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

  14. 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.

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

    International Nuclear Information System (INIS)

    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. Electric vehicle test report Cutler-Hammer Corvette

    Science.gov (United States)

    1981-01-01

    Vehicles were characterized for the state of the art assessment of electric vehicles. The vehicle evaluated was a Chevrolet Corvette converted to electric operation. 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 emphasis was on the electrical portion of the drive train, although some analysis and discussion of the mechanical elements are included. 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.

  17. Operation, test, research and development of the High Temperature Engineering Test Reactor (HTTR). FY2013

    International Nuclear Information System (INIS)

    2014-12-01

    The High Temperature Engineering Test Reactor (HTTR), a graphite-moderated and helium gas-cooled reactor with 30MW of thermal power, constructed at the Oarai Research and Development Center of the Japan Atomic Energy Agency (JAEA) is the first high-temperature gas-cooled reactor (HTGR) in Japan. The HTTR was attained at the full power operation of 30MW in December 2001 and achieved the 950degC of outlet coolant temperature at the outside the reactor pressure vessel in June 2004. To establish and upgrade basic technologies for HTGRs, we have obtained demonstration test data necessary for several R and Ds, and accumulated operation and maintenance experience of HTGRs throughout the HTTR's operation such as rated power operations, safety demonstration tests and long-term high temperature operations, and so on. In fiscal year 2013, we started to prepare the application document of reactor installation license for the HTTR to prove conformity with the new research reactor's safety regulatory requirements taken effect from December 2013. We had been making effort to restart the HTTR which was stopped since the 2011 when the Pacific coast of Tohoku Earthquake (2011.3.11) occurred. This report summarizes activities and results of HTTR operation, maintenance, and several R and Ds, which were carried out in the fiscal year 2013. (author)

  18. Operation, test, research and development of the High Temperature Engineering Test Reactor (HTTR). FY2014

    International Nuclear Information System (INIS)

    2016-02-01

    The High Temperature Engineering Test Reactor (HTTR), a graphite-moderated and helium gas-cooled reactor with 30 MW of thermal power, constructed at the Oarai Research and Development Center of the Japan Atomic Energy Agency is the first high-temperature gas-cooled reactor (HTGR) in Japan. The HTTR was attained at the full power operation of 30 MW in December 2001 and achieved the 950degC of coolant outlet temperature at outside of the reactor pressure vessel in June 2004. To establish and upgrade basic technologies for HTGRs, we have obtained demonstration test data necessary for several R and Ds, and accumulated operation and maintenance experience of HTGRs throughout the HTTR's operation such as rated power operations, safety demonstration tests and long-term high temperature operations, and so on. In fiscal year 2014, we started to apply the application document of reactor installation license for the HTTR to prove conformity with the new research reactor's safety regulatory requirements taken effect from December 2013. We had been making effort to restart the HTTR which was stopped since the 2011 by the Pacific coast of Tohoku Earthquake. This report summarizes activities and results of HTTR operation, maintenance, and several R and Ds, which were carried out in the fiscal year 2014. (author)

  19. Manipulator for testing a top-opened reactor pressure vessel

    International Nuclear Information System (INIS)

    Bauer, R.; Kastl, H.

    1991-01-01

    The design is described of a manipulator to be inserted into the inside of reactor pressure vessels opened at the top. The main components of the manipulator include a fixed column protruding into the pressure vessel and a support which is slidable on the column and carries the bearing component for the measuring, testing, inspection and repair instruments. The device includes a driving equipment for the support as well as the power supply for the sets accommodated on the support, with the aim to reduce the failure rate of the manipulator as a whole, shorten the time necessary for its assembling and thus the time of staying in the reactor pressure vessel and, at the same time, make its maintenance and operation easier. (Z.S.). 13 figs

  20. ITER: a technology test bed for a fusion reactor

    International Nuclear Information System (INIS)

    Huguet, M.; Green, B.J.

    1996-01-01

    The ITER Project aims to establish nuclear fusion as an energy source that has potential safety and environmental advantages, and to develop the technologies required for a fusion reactor. ITER is a collaborative project between the European Union, Japan, the Russian Federation and the United States of America. During the current phase of the Project, an R and D programme of about 850 million dollars is underway to develop the technologies required for ITER. This technological effort should culminate in the construction of the components and systems of the ITER machine and its auxiliaries. The main areas of technological development include the first wall and divertor technology, the blanket technology and tritium breeding, superconducting magnet technology, pulsed power technology and remote handling. ITER is a test bed and an essential step to establish the technology of future fusion reactors. Many of the ITER technologies are of potential interest to other fields and their development is expected to benefit the industries involved. (author)

  1. Replacement of the Advanced Test Reactor control room

    International Nuclear Information System (INIS)

    Durney, J.L.; Klingler, W.B.

    1989-01-01

    The control room for the Advanced Test Reactor has been replaced to provide modern equipment utilizing current standards and meeting the current human factors requirements. The control room was designed in the early 1960 era and had not been significantly upgraded since the initial installation. The replacement did not change any of the safety circuits or equipment but did result in replacement of some of the recorders that display information from the safety systems. The replacement was completed in concert with the replacement of the control room simulator which provided important feedback on the design. The design successfully incorporates computer-based systems into the display of the plant variables. This improved design provides the operator with more information in a more usable form than was provided by the original design. The replacement was successfully completed within the scheduled time thereby minimizing the down time for the reactor. 1 fig., 1 tab

  2. Replacement of the Advanced Test Reactor control room

    International Nuclear Information System (INIS)

    Durney, J.L.; Klingler, W.B.

    1990-01-01

    The control room for the Advanced Test Reactor has been replaced to provide modern equipment utilizing current standards and meeting the current human factors requirements. The control room was designed in the early 1960 era and had not been significantly upgraded since the initial installation. The replacement did not change any of the safety circuits or equipment but did result in replacement of some of the recorders that display information from the safety systems. The replacement was completed in concert with the replacement of the control room simulator which provided important feedback on the design. The design successfully incorporates computer-based systems into the display of the plant variables. This improved design provides the operator with more information in a more usable form than was provided by the original design. The replacement was successfully completed within the scheduled time thereby minimizing the down time for the reactor

  3. Air leakage test of reactor hall using tracer technique

    International Nuclear Information System (INIS)

    Yang Yanqiu; Yang Liang; Yang Tongzai

    2011-01-01

    The leakage ratios of three related reactor halls were tested by sulfur hexafluoride gaseous tracer technique. Moreover, the accumulation intensities of leak gas and its retention time in some important working rooms, the crossroads of corridors and anteroom of the building were detected. The results show that the air leakage ratios of the three reactor halls are (7.30±0.16) x 10 -4 , (1.88±0.12) x 10 -4 and (2.07±0.07) x 10 -4 h -1 . The leak gas accumulates in all the detected working rooms fast, and the retention time to various rooms is about 5 h. The heaviest intensities are in the clothes change rooms on the first floor. However, the retention time to the crossroads and the anteroom is about 10 h, and the accumulation intensities are much small. (authors)

  4. Alteration in reactor installations (Unit 1 and 2 reactor facilities) in the Hamaoka Nuclear Power Station of The Chubu Electric Power Co., Inc. (report)

    International Nuclear Information System (INIS)

    1982-01-01

    A report by the Nuclear Safety Commission to the Ministry of International Trade and Industry concerning the alteration in Unit 1 and 2 reactor facilities in the Hamaoka Nuclear Power Station, Chubu Electric Power Co., Inc., was presented. The technical capabilities for the alteration of reactor facilities in Chubu Electric Power Co., Inc., were confirmed to be adequate. The safety of the reactor facilities after the alteration was confirmed to be adequate. The items of examination made for the confirmation of the safety are as follows: reactor core design (nuclear design, mechanical design, mixed reactor core), the analysis of abnormal transients in operation, the analysis of various accidents, the analysis of credible accidents for site evaluation. (Mori, K.)

  5. The Jules Horowitz Reactor (JHR), a European Material Testing Reactor (MTR), with extended experimental capabilities

    International Nuclear Information System (INIS)

    Ballagny, A.; Bergamaschi, Y.; Bouilloux, Y.; Bravo, X.; Guigon, B.; Rommens, M.; Tremodeux, P.

    2003-01-01

    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 14 ncm -2 s -1 and a fast flux of 6,4.10 14 ncm -2 s -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 = deplacement 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. (author)

  6. Enhancing reactor availability factor by diagnostic monitoring and data acquisition of electrical equipments

    International Nuclear Information System (INIS)

    Singh, G.

    2006-01-01

    Electrical energy has made significant contribution to rapid growth of industrial activity in the country. Development and improvement of energy conversion devices or electrical apparatus have supported the growth. Reliability is probably the most important factor in electrical supply system, not only to give uninterrupted service but to provide an economic supply. Regular diagnostic testing of electrical equipments will make a significant contribution to the reliability of electrical supply. The purpose of diagnostic monitoring is to recognize the development of faults at an early stage, which consequently allows greater freedom to schedule the outages resulting in lower downtime and lower capitalized losses. The insulation constitutes the heart of any electrical/power equipment. The insulation in power equipment in normal condition undergoes certain changes in the physical, chemical, electrical and mechanical properties. The change with respect to time in the presence of an influencing factor, more often a stress (electrical) is referred to as ageing. The deterioration of insulating material plays an important role in the assessing the condition of electrical equipments. The systematic diagnostic tests are also part of the maintenance program to ensure the continued serviceability of electrical equipments, by replacing or repairing the components likely to fail, as revealed by the test. Diagnostic tests are carried out on various electrical equipments for detection of incipient fault, location and judging their severity. (author)

  7. 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.

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

    International Nuclear Information System (INIS)

    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

    2008-01-01

    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: (1) Identifies pre-conceptual design requirements; (2) Develops test loop equipment schematics and layout; (3) Identifies space allocations for each of the facility functions, as required; (4) Develops a pre-conceptual site layout including transportation, parking and support structures, and railway systems; (5) Identifies pre-conceptual utility and support system needs; and (6) Establishes pre-conceptual electrical one-line drawings and schedule for development of power needs

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

    International Nuclear Information System (INIS)

    Saikusa, Akio; Nakagawa, Shigeaki; Fujimoto, Nozomu; Tachibana, Yukio; Iyoku, Tatsuo

    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. Operating experiences since rise-to-power test in high temperature engineering test reactor (HTTR)

    International Nuclear Information System (INIS)

    Tochio, Daisuke; Watanabe, Shuji; Motegi, Toshihiro; Kawano, Shuichi; Kameyama, Yasuhiko; Sekita, Kenji; Kawasaki, Kozo

    2007-03-01

    The rise-to-power test of the High Temperature Engineering Test Reactor (HTTR) was actually started in April 2000. The rated thermal power of 30MW and the rated reactor outlet coolant temperature of 850degC were achieved in the middle of Dec. 2001. After that, the reactor thermal power of 30MW and the reactor outlet coolant temperature of 950degC were achieved in the final rise-to-power test in April 2004. After receiving the operation licensing at 850degC, the safety demonstration tests have conducted to demonstrate inherent safety features of the HTGRs as well as to obtain the core and plant transient data for validation of safety analysis codes and for establishment of safety design and evaluation technologies. This paper summarizes the HTTR operating experiences for six years from start of the rise-to-power test that are categorized into (1) Operating experiences related to advanced gas-cooled reactor design, (2) Operating experiences for improvement of the performance, (3) Operating experiences due to fail of system and components. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-07-01

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

  12. Verification Test of Hydraulic Performance for Reactor Coolant Pump

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sang Jun; Kim, Jae Shin; Ryu, In Wan; Ko, Bok Seong; Song, Keun Myung [Samjin Ind. Co., Seoul (Korea, Republic of)

    2010-01-15

    According to this project, basic design for prototype pump and model pump of reactor coolant pump and test facilities has been completed. Basic design for prototype pump to establish structure, dimension and hydraulic performance has been completed and through primary flow analysis by computational fluid dynamics(CFD), flow characteristics and hydraulic performance have been established. This pump was designed with mixed flow pump having the following design requirements; specific velocity(Ns); 1080.9(rpm{center_dot}m{sup 3}/m{center_dot}m), capacity; 3115m{sup 3}/h, total head ; 26.3m, pump speed; 1710rpm, pump efficiency; 77.0%, Impeller out-diameter; 349mm, motor output; 360kw, design pressure; 17MPaG. The features of the pump are leakage free due to no mechanical seal on the pump shaft which insures reactor's safety and law noise level and low vibration due to no cooling fan on the motor which makes eco-friendly product. Model pump size was reduced to 44% of prototype pump for the verification test for hydraulic performance of reactor coolant pump and was designed with mixed flow pump and canned motor having the following design requirements; specific speed(NS); 1060.9(rpm{center_dot}m{sup 3}/m{center_dot}m), capacity; 539.4m{sup 3}/h, total head; 21.0m, pump speed; 3476rpm, pump efficiency; 72.9%, Impeller out-diameter; 154mm, motor output; 55kw, design pressure; 1.0MPaG. The test facilities were designed for verification test of hydraulic performance suitable for pump performance test, homologous test, NPSH test(cavitation), cost down test and pressure pulsation test of inlet and outlet ports. Test tank was designed with testing capacity enabling up to 2000m{sup 3}/h and design pressure 1.0MPaG. Auxiliary pump was designed with centrifugal pump having capacity; 1100m{sup 3}/h, total head; 42.0m, motor output; 190kw

  13. 76 FR 70166 - Electrical Standards for Construction and General Industry; Extension of the Office of Management...

    Science.gov (United States)

    2011-11-10

    ...] Electrical Standards for Construction and General Industry; Extension of the Office of Management and Budget... contained in the Electrical Standards for Construction (29 CFR part 1926, Subpart K) and for General... maintenance of electric utilization equipment that prevent death and serious injuries among construction and...

  14. The approximate thermal-model-testing method for non-stationary temperature fields in central zones of fast reactor assemblies

    International Nuclear Information System (INIS)

    Mikhin, V.I.; Matukhin, N.M.

    2000-01-01

    The approach to generalization of the non-stationary heat exchange data for the central zones of the nuclear reactor fuel assemblies and the approximate thermal-model-testing criteria are proposed. The fuel assemblies of fast and water-cooled reactors with different fuel compositions have been investigated. The reason of the non-stationary heat exchange is the fuel-energy-release time dependence. (author)

  15. 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.

  16. Main refurbishment activities on electronic and electrical equipment for the FRG-1 research reactor

    International Nuclear Information System (INIS)

    Blom, K.H.; Krull, W.

    1997-01-01

    As GKSS intends to operate the research reactor FRG-1 safely and reliably for many years to come, the plant is constantly refurbished and upgraded both in the interests of safety and operational reasons. The following electronic and electrical systems have been replaced or improved since 1990: Information and signalling systems; Emergency power plant (permit applied for); External and internal lightning protection system; Reactor protection system (in part); Safety lighting; Alarm and staff locating system; Control room telephone system; Closed-circuit television system; Beam tube controls; Storage plant for radioactive liquid waste; Ambient dose rate measuring system; Meteorological measuring system; Control and measuring system for the primary cooling circuit; Control rod drives; Control rod control system; Soft start for the secondary pumps; Control and switching devices for the emergency power plant; Trailing cable installation for the reactor bridge; Main-voltage distribution systems/cable routes. (author). 13 figs, 1 tab

  17. Design study of electrical power supply system for tokamak fusion power reactor

    International Nuclear Information System (INIS)

    1977-01-01

    Design study of the electrical power supply system for a 2000MWt Tokamak-type fusion reactor has been carried out. The purposes are to reveal and study problems in the system, leading to a plan of the research and development. Performed were study of the electrical power supply system and design of superconducting inductive energy storages and power switches. In study of the system, specification and capability of various power supplies for the fusion power reactor and design of the total system with its components were investigated. For the superconducting inductive energy storages, material choice, design calculation, and structural design were conducted, giving the size, weight and performance. For thyristor switches, circuit design in the parallel / series connection of element valves and cooling design were studied, providing the size and weight. (auth.)

  18. The roles of ozone and zeolite on reactive dye degradation in electrical discharge reactors.

    Science.gov (United States)

    Peternel, L; Kusic, H; Koprivanac, N; Locke, B R

    2006-05-01

    In this study high voltage pulsed corona electrical discharge advanced oxidation processes (AOPs) were applied to bleach and degrade C.I. Reactive Green 8 and C.I. Reactive Red 45 organic dyes in water solutions. Two types of hybrid gas/liquid high voltage electrical discharge (corona) reactors, known as hybrid series and hybrid parallel were studied. The difference between these reactors relates to electrode configuration, which affects the amounts of ozone, hydrogen peroxide and hydroxyl radicals produced. Experiments were conducted using dye concentrations of 20 mgl(-1) and 75 mgl(-1), with and without NH4ZSM5 zeolite addition in order to determine possible effects of added solid particles to total process efficiency. The role of ozone in combination with zeolites was assessed through comparative direct ozonation experiments with ozone supplied by an ozone generator. UV/VIS spectrophotometric measurements and measurements of total organic carbon (TOC) were used for the determination of decolorization and mineralization rates.

  19. Small nuclear power reactor emergency electric power supply system reliability comparative analysis

    International Nuclear Information System (INIS)

    Bonfietti, Gerson

    2003-01-01

    This work presents an analysis of the reliability of the emergency power supply system, of a small size nuclear power reactor. Three different configurations are investigated and their reliability analyzed. The fault tree method is used as the main tool of analysis. The work includes a bibliographic review of emergency diesel generator reliability and a discussion of the design requirements applicable to emergency electrical systems. The influence of common cause failure influences is considered using the beta factor model. The operator action is considered using human failure probabilities. A parametric analysis shows the strong dependence between the reactor safety and the loss of offsite electric power supply. It is also shown that common cause failures can be a major contributor to the system reliability. (author)

  20. [Generalized neonatal screening based on laboratory tests].

    Science.gov (United States)

    Ardaillou, Raymond; Le Gall, Jean-Yves

    2006-11-01

    Implementation of a generalized screening program for neonatal diseases must obey precise rules. The disease must be severe, recognizable at an early stage, amenable to an effective treatment, detectable with a non expensive and widely applicable test; it must also be a significant public health problem. Subjects with positive results must be offered immediate treatment or prevention. All screening programs must be regularly evaluated. In France, since 1978, a national screening program has been organized by a private association ("Association française pour le dépistage et la prévention des handicaps de l'enfant") and supervised by the "Caisse nationale d'assurance maladie" and "Direction Générale de la Sante". Five diseases are now included in the screening program: phenylketonuria, hypothyroidism, congenital adrenal hyperplasia, cystic fibrosis and sickle cell disease (the latter only in at-risk newborns). Toxoplasmosis is a particular problem because only the children of mothers who were not tested during the pregnancy or who seroconverted are screened. Neonatal screening for phenylketonuria and hypothyrodism is unanimously recommended. Screening for congenital adrenal hyperplasia is approved in most countries. Cases of sickle cell disease and cystic fibrosis are more complex because--not all children who carry the mutations develop severe forms;--there is no curative treatment;--parents may become anxious, even though the phenotype is sometimes mild or even asymptomatic. Supporters of screening stress the benefits of early diagnosis (which extends the life expectancy of these children, particularly in the case of sickle cell disease), the fact that it opens up the possibility of prenatal screening of future pregnancies, and the utility of informing heterozygous carriers identified by familial screening. Neonatal screening for other diseases is under discussion. Indeed, technical advances such as tandem mass spectrometry make it possible to detect about 50

  1. Integral test of JENDL-3.3 on fast reactors

    International Nuclear Information System (INIS)

    Chiba, Gou; Hazama, Taira

    2003-05-01

    An integral test has been carried out to evaluate a performance of evaluated nuclear data library JENDL-3.3, which was newly released, in a view of applying neutronics analyses of fast reactors. Japan Nuclear Cycle Development Institute has a large amount of data of critical assembly experiments (ZPPR, BFS, MOZART and FCA) and power reactor tests (JOYO). The database was utilized in this test. In plutonium loaded cores, an improvement was observed about 0.3% ε k in criticality and 5% in the non-leakage term of sodium void reactivity by a revision form JENDL-3.2 to -3.3. These results shoed that the revision is valid in plutonium loaded cores. In uranium loaded cores, dependence of C/E values on control rod position became smaller in control rod worth in ZPPR cores. On the other hand, C/E values became worse both in criticality (0.6%εk) and in sodium void reactivity (30%) in BFS cores. The main cause was a revision of uranium-235 capture cross section, and it could not be concluded whether the revision is valid or not in uranium loaded cores. It is necessary to carry out a validation test at other independent critical experiments in which uranium fuel is used. (author)

  2. DESIGN SAFETY PROBLEMS OF NUCLEAR REACTORS IN SPACE FOR ELECTRICAL POWER

    Energy Technology Data Exchange (ETDEWEB)

    Pickler, D A

    1963-06-15

    A general treatment is presented of some of the problems in the design safety of reactors which are to be operated in space. The basic requirements of these reachigh temperatures. The usual concept of a space reactor is described briefly, and the hazards of an assumed unmanned vehicle with an enriched-U-fueled reactor are examined during its launching, orbit, and reentry. Graphs are given for the dose vs distance downwind for an excursion of 100 Mw-sec, for the activity vs time after shutdown of a reactor which has been operated for 5 yr at 100 kw(t), and for the altitude vs orbital lifetime. Apparent conflicts between the basic requirements are discussed. (D.L.C.)

  3. STG-ET: DLR electric propulsion test facility

    Directory of Open Access Journals (Sweden)

    Andreas Neumann

    2017-04-01

    Full Text Available DLR operates the High Vacuum Plume Test Facility Göttingen – Electric Thrusters (STG-ET. This electric propulsion test facility has now accumulated several years of EP-thruster testing experience. Special features tailored to electric space propulsion testing like a large vacuum chamber mounted on a low vibration foundation, a beam dump target with low sputtering, and a performant pumping system characterize this facility. The vacuum chamber is 12.2m long and has a diameter of 5m. With respect to accurate thruster testing, the design focus is on accurate thrust measurement, plume diagnostics, and plume interaction with spacecraft components. Electric propulsion thrusters have to run for thousands of hours, and with this the facility is prepared for long-term experiments. This paper gives an overview of the facility, and shows some details of the vacuum chamber, pumping system, diagnostics, and experiences with these components.

  4. High flux testing reactor Petten. Replacement of the reactor vessel and connected components. Overall report

    International Nuclear Information System (INIS)

    Chrysochoides, N.G.; Cundy, M.R.; Von der Hardt, P.; Husmann, K.; Swanenburg de Veye, R.J.; Tas, A.

    1985-01-01

    The project of replacing the HFR originated in 1974 when results of several research programmes confirmed severe neutron embrittlement of aluminium alloys suggesting a limited life of the existing facility. This report contains the detailed chronology of events concerning preparation and execution of the replacement. After a 14 months' outage the reactor resumed routine operation on 14th February, 1985. At the end of several years of planning and preparation the reconstruction proceded in the following steps: unloading of the old core, decay of short-lived radioactivity in December 1983, removal of the old tank and of its peripheral equipment in January-February 1984, segmentation and waste disposal of the removed components in March-April, decontamination of the pools, bottom penetration overhauling in May-June, installation of the new tank and other new components in July-September, testing and commissioning, including minor modifications in October-December, and, trials runs and start-up preparation in January-February 1985. The new HFR Petten features increased and improved experimental facilities. Among others the obsolete thermal columns was replaced by two high flux beam tubes. Moreover the new plant has been designed for future increases of reactor power and neutron fluxes. For the next three to four years the reactor has to cope with a large irradiation programme, claiming its capacity to nearly 100%

  5. Electrical Stimulation of the Ventral Tegmental Area Induces Reanimation from General Anesthesia

    Science.gov (United States)

    Solt, Ken; Van Dort, Christa J.; Chemali, Jessica J.; Taylor, Norman E.; Kenny, Jonathan D.; Brown, Emery N.

    2014-01-01

    BACKGROUND Methylphenidate or a D1 dopamine receptor agonist induce reanimation (active emergence) from general anesthesia. We tested whether electrical stimulation of dopaminergic nuclei also induces reanimation from general anesthesia. METHODS In adult rats, a bipolar insulated stainless steel electrode was placed in the ventral tegmental area (VTA, n = 5) or substantia nigra (SN, n = 5). After a minimum 7-day recovery period, the isoflurane dose sufficient to maintain loss of righting was established. Electrical stimulation was initiated and increased in intensity every 3 min to a maximum of 120μA. If stimulation restored the righting reflex, an additional experiment was performed at least 3 days later during continuous propofol anesthesia. Histological analysis was conducted to identify the location of the electrode tip. In separate experiments, stimulation was performed in the prone position during general anesthesia with isoflurane or propofol, and the electroencephalogram was recorded. RESULTS To maintain loss of righting, the dose of isoflurane was 0.9% ± 0.1 vol%, and the target plasma dose of propofol was 4.4 μg/ml ± 1.1 μg/ml (mean ± SD). In all rats with VTA electrodes, electrical stimulation induced a graded arousal response including righting that increased with current intensity. VTA stimulation induced a shift in electroencephalogram peak power from δ (anesthesia with isoflurane or propofol. These results are consistent with the hypothesis that dopamine release by VTA, but not SN, neurons induces reanimation from general anesthesia. PMID:24398816

  6. Synchrotron radiation losses in Engineering Test Reactors (ETRs)

    International Nuclear Information System (INIS)

    Uckan, N.A.

    1987-11-01

    In next-generation Engineering Test Reactors (ETRs), one major objective is envisioned to be a long-pulse or steady-state burn using noninductive current drive. At the high temperatures needed for efficient current drive, synchrotron radiation could represent a large power loss, especially if wall reflectivity (R) is very low. Many INTOR-class ETR designs [Fusion Engineering Reactor (FER), Next European Torus (NET), OTR, Tokamak Ignition/Burn Engineering Reactor (TIBER), etc.] call for carbon-covered surfaces for which wall reflectivity is uncertain. Global radiation losses are estimated for these devices using empirical expressions given by Trubnikov (and others). Various operating scenarios are evaluated under the assumption that the plasma performance is limited by either the density limit (typical of the ignition phase) or the beta limit (typical of the current drive phase). For a case with ≥90% wall reflectivity, synchrotron radiation is not a significant contribution to the overall energy balance (the ratio of synchrotron to alpha power is less than 10 to 20%, even at ∼ 30 keV) and thus should not adversely alter performance in these devices. In extreme cases with 0% wall reflectivity, the ratio of synchrotron radiation to alpha power may approach 30 to 60% (depending on the device and limiting operating scenario), adversely affecting the performance characteristics. 12 refs., 7 tabs

  7. 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

  8. 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.

  9. 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.

  10. Development of large insulator rings for the TOKAMAK Fusion Test Reactor

    International Nuclear Information System (INIS)

    Brown, T.; Tobin, A.

    1977-01-01

    Research and development leading to the manufacture of large ceramic insulator rings for the TFTR (TOKAMAK Fusion Test Reactor). Material applictions, fabrication approach and testing activities are highlighted

  11. 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.

  12. Seismically induced accident sequence analysis of the advanced test reactor

    International Nuclear Information System (INIS)

    Khericha, S.T.; Henry, D.M.; Ravindra, M.K.; Hashimoto, P.S.; Griffin, M.J.; Tong, W.H.; Nafday, A.M.

    1991-01-01

    A seismic probabilistic risk assessment (PRA) was performed for the Department of Energy (DOE) Advanced Test Reactor (ATR) as part of the external events analysis. The risk from seismic events to the fuel in the core and in the fuel storage canal was evaluated. The key elements of this paper are the integration of seismically induced internal flood and internal fire, and the modeling of human error rates as a function of the magnitude of earthquake. The systems analysis was performed by EG ampersand G Idaho, Inc. and the fragility analysis and quantification were performed by EQE International, Inc. (EQE)

  13. Reduced enrichment for research and test reactors: Proceedings

    International Nuclear Information System (INIS)

    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

  14. Feasibility study of full-reactor gas core demonstration test

    Science.gov (United States)

    Kunze, J. F.; Lofthouse, J. H.; Shaffer, C. J.; Macbeth, P. J.

    1973-01-01

    Separate studies of nuclear criticality, flow patterns, and thermodynamics for the gas core reactor concept have all given positive indications of its feasibility. However, before serious design for a full scale gas core application can be made, feasibility must be shown for operation with full interaction of the nuclear, thermal, and hydraulic effects. A minimum sized, and hence minimum expense, test arrangement is considered for a full gas core configuration. It is shown that the hydrogen coolant scattering effects dominate the nuclear considerations at elevated temperatures. A cavity diameter of somewhat larger than 4 ft (122 cm) will be needed if temperatures high enough to vaporize uranium are to be achieved.

  15. Technology issues for decommissioning the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Spampinato, P.T.; Walton, G.R.

    1994-01-01

    The approach for decommissioning the Tokamak Fusion Test Reactor has evolved from a conservative plan based on cutting up and burying all of the systems, to one that considers the impact tritium contamination will have on waste disposal, how large size components may be used as their own shipping containers, and even the possibility of recycling the materials of components such as the toroidal field coils and the tokamak structure. In addition, the project is more carefully assessing the requirements for using remotely operated equipment. Finally, valuable cost database is being developed for future use by the fusion community

  16. 309 plutonium recycle test reactor ion exchanger vault deactivitation report

    International Nuclear Information System (INIS)

    Griffin, P.W.

    1996-03-01

    This report documents the deactivation of the ion exchanger vault at the 309 Plutonium Recycle Test Reactor (PRTR) Facility in the 300 Area. The vault deactivation began in May 1995 and was completed in June 1995. The final site restoration and shipment of the low-level waste for disposal was finished in September 1995. The ion exchanger vault deactivation project involved the removal and disposal of twelve ion exchangers and decontaminating and fixing of residual smearable contamination on the ion exchanger vault concrete surfaces

  17. Adaptation of electrical conductivity test for Moringa oleifera seeds

    Directory of Open Access Journals (Sweden)

    Maria Luiza de Souza Medeiros

    2017-09-01

    Full Text Available This study aimed to adapt and test the efficiency of electrical conductivity methodology test in quality evaluation of Moringa oleifera Lam seeds. For physiological characterization four seed sets were evaluated by tests of germination, seedlings emergency, speed of emergency index, emergency first count, seedlings length and dry mass and cold test. The electrical conductivity test was carried out at 25 °C for 4, 8, 12, 16 and 24 h of immersion in 75 or 125 mL of distilled water using 25 or 50 seeds. A completely randomized design was used. The best results were obtained when using 50 seeds immersed in 75 mL or 125 mL of distilled water for 4 h. The electrical conductivity test adapted to moringa seeds was efficient in ranking sets of different vigor levels. The test may be efficiently used for physiological quality evaluation of moringa seeds.

  18. Reactivity control system of the high temperature engineering test reactor

    International Nuclear Information System (INIS)

    Tachibana, Yukio; Sawahata, Hiroaki; Iyoku, Tatsuo; Nakazawa, Toshio

    2004-01-01

    The reactivity control system of the high temperature engineering test reactor (HTTR) consists of a control rod system and a reserve shutdown system. During normal operation, reactivity is controlled by the control rod system, which consists of 32 control rods (16 pairs) and 16 control rod drive mechanisms except for the case when the center control rods are removed to perform an irradiation test. In an unlikely event that the control rods fail to be inserted, reserve shutdown system is provided to insert pellets of neutron-absorbing material into the core. Alloy 800H is chosen for the metallic parts of the control rods. Because the maximum temperature of the control rods reaches about 900 deg. C at reactor scrams, structural design guideline and design material data on Alloy 800H are needed for the high temperature design. The design guideline for the HTTR control rod is based on ASME Code Case N-47-21. Design material data is also determined and shown in this paper. Observing the guideline, temperature and stress analysis were conducted; it can be confirmed that the target life of the control rods of 5 years can be achieved. Various tests conducted for the control rod system and the reserve shutdown system are also described

  19. Testing the generalized partial credit model

    OpenAIRE

    Glas, Cornelis A.W.

    1996-01-01

    The partial credit model (PCM) (G.N. Masters, 1982) can be viewed as a generalization of the Rasch model for dichotomous items to the case of polytomous items. In many cases, the PCM is too restrictive to fit the data. Several generalizations of the PCM have been proposed. In this paper, a generalization of the PCM (GPCM), a further generalization of the one-parameter logistic model, is discussed. The model is defined and the conditional maximum likelihood procedure for the method is describe...

  20. Testing and development of electric vehicle batteries for EPRI Electric Transportation Program

    Science.gov (United States)

    1985-11-01

    Argonne National Laboratory conducted an electric-vehicle battery testing and development program for the Electric Power Research Institute. As part of this program, eighteen battery modules previously developed by Johnson Controls, Inc. were tested. This type of battery (EV-2300 - an improved state-of-the-art lead-acid battery) was designed specifically for improved performance, range, and life in electric vehicles. In order to obtain necessary performance data, the batteries were tested under various duty cycles typical of normal service. This program, supported by the Electric Power Research Institute, consisted of three tasks: determination of the effect of cycle life vs peak power and rest period, determination of the impact of charge method on cycle life, and evaluation of the EV-2300 battery system. Two supporting studies were also carried out: one on thermal management of electric-vehicle batteries and one on enhanced utilization of active material in lead-acid batteries.