WorldWideScience

Sample records for aerospace system test reactor

  1. Comparison of the Aerospace Systems Test Reactor loss-of-coolant test data with predictions of the 3D-AIRLOCA code

    International Nuclear Information System (INIS)

    This paper compares the predictions of the revised 3D-AIRLOCA computer code to those data available from the Aerospace Systems Test Reactor's (ASTR's) loss-of-coolant-accident (LOCA) tests run in 1964. The theoretical and experimental hot-spot temperature responses compare remarkably well. In the thirteen cases studied, the irradiation powers varied from 0.4 to 8.87 MW; the irradiation times were 300, 1540, 1800, and 104 s. The degrees of agreement between the data and predictions provide an experimental validation of the 3D-AIRLOCA code

  2. Gaseous fuel reactor systems for aerospace applications

    Science.gov (United States)

    Thom, K.; Schwenk, F. C.

    1977-01-01

    Research on the gaseous fuel nuclear rocket concept continues under the programs of the U.S. National Aeronautics and Space Administration (NASA) Office for Aeronautics and Space Technology and now includes work related to power applications in space and on earth. In a cavity reactor test series, initial experiments confirmed the low critical mass determined from reactor physics calculations. Recent work with flowing UF6 fuel indicates stable operation at increased power levels. Preliminary design and experimental verification of test hardware for high-temperature experiments have been accomplished. Research on energy extraction from fissioning gases has resulted in lasers energized by fission fragments. Combined experimental results and studies indicate that gaseous-fuel reactor systems have significant potential for providing nuclear fission power in space and on earth.

  3. A Systems Engineering Approach to Quality Assurance for Aerospace Testing

    Science.gov (United States)

    Shepherd, Christena C.

    2015-01-01

    On the surface, it appears that AS91001 has little to say about how to apply a Quality Management System (QMS) to major aerospace test programs (or even smaller ones). It also appears that there is little in the quality engineering Body of Knowledge (BOK)2 that applies to testing, unless it is nondestructive examination (NDE), or some type of lab or bench testing associated with the manufacturing process. However, if one examines: a) how the systems engineering (SE) processes are implemented throughout a test program; and b) how these SE processes can be mapped to the requirements of AS9100, a number of areas for involvement of the quality professional are revealed. What often happens is that quality assurance during a test program is limited to inspections of the test article; what could be considered a manufacturing al fresco approach. This limits the quality professional and is a disservice to the programs and projects, since there are a number of ways that quality can enhance critical processes, and support efforts to improve risk reduction, efficiency and effectiveness.

  4. Aerospace Systems Monitor Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Proposal Title: Aerospace Systems Monitor PHASE 1 Technical Abstract: This Phase II STTR project will continue development and commercialization of the Aerospace...

  5. Aerospace Systems Monitor Project

    Data.gov (United States)

    National Aeronautics and Space Administration — This Phase I STTR project will demonstrate the Aerospace System Monitor (ASM). This technology transforms the power distribution network in a spacecraft or aircraft...

  6. Current Trends on the Applicability of Ground Aerospace Materials Test Data to Space System Environments

    Science.gov (United States)

    Hirsch, David B.

    2010-01-01

    This slide presentation discusses the application of testing aerospace materials to the environment of space for flammability. Test environments include use of drop towers, and the parabolic flight to simulate the low gravity environment of space.

  7. High intensity acoustic testing to determine structural fatigue life and to improve reliability in nuclear reactor and aerospace structures

    International Nuclear Information System (INIS)

    The author reviews some of the techniques in which high intensity acoustic testing is used in engineering practice. (a) In the nuclear engineering field the simulation of reactor noise due to the CO2 circulator and the use of strain gauges to obtain a response spectrum in order to predict the fatigue life of gas-cooled nuclear reactor structures where a 30 year lifespan is of paramount importance is described. (b) In the satellite field the simulation of the high intensity noise due to the launching rocket motors and the testing of the integrity of the satellite structure and the behaviour of the electronic control system when affected by high intensity acoustic excitation is discussed. The use of acoustic testing to improve the reliability before the launching of the satellite is also considered. (c) In the aircraft and rocket field the generation of high intensity noise to simulate boundary layer pressure fluctuation or turbulence of a flying object or aircraft at various speeds is considered. (Auth.)

  8. Testing of the Micro-Reactor System

    Czech Academy of Sciences Publication Activity Database

    Krystyník, Pavel; Beneš, Ondřej; Klusoň, Petr; Šolcová, Olga

    Praha: Česká společnost průmyslové chemie, 2015, s. 30 /p104./. ISBN 978-80-86238-73-9. [mezinárodní chemicko-technologická konference (ICCT 2015) /3./. Mikulov (CZ), 13.04.2015-15.04.2015] R&D Projects: GA ČR GA15-14228S Institutional support: RVO:67985858 Keywords : micro-reactor technology * heat transfer * testing Subject RIV: CI - Industrial Chemistry, Chemical Engineering

  9. Testing of the Micro-Reactor System

    Czech Academy of Sciences Publication Activity Database

    Beneš, Ondřej; Hanková, Libuše; Klusoň, Petr; Šolcová, Olga

    Bratislava: Slovak Society of Chemical Engineering, 2015 - (Markoš, J.), s. 40 ISBN 978-80-89475-14-8. [International Conference of Slovak Society of Chemical Engineering /42./. Tatranské Matliare (SK), 25.05.2015-29.05.2015] R&D Projects: GA ČR GA15-14228S Institutional support: RVO:67985858 Keywords : micro-reactor technology * testing * partial oxidation Subject RIV: CI - Industrial Chemistry, Chemical Engineering

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  12. Cybersecurity for aerospace autonomous systems

    Science.gov (United States)

    Straub, Jeremy

    2015-05-01

    High profile breaches have occurred across numerous information systems. One area where attacks are particularly problematic is autonomous control systems. This paper considers the aerospace information system, focusing on elements that interact with autonomous control systems (e.g., onboard UAVs). It discusses the trust placed in the autonomous systems and supporting systems (e.g., navigational aids) and how this trust can be validated. Approaches to remotely detect the UAV compromise, without relying on the onboard software (on a potentially compromised system) as part of the process are discussed. How different levels of autonomy (task-based, goal-based, mission-based) impact this remote characterization is considered.

  13. Architecture of the ETR [experimental test reactor] systems code

    International Nuclear Information System (INIS)

    TETRA, a tokamak systems code capable of modeling experimental test reactors (ETRs), was developed in a joint effort by participants of the fusion community. The first version of this code was constructed to model devices similar to the Tokamak Ignition/Burn Engineering Reactor (TIBER) in configuration and design. A major feature of this code is its ability to perform optimization studies. Future work will include broadening the scope of the code, particularly in the area of materials selection, to more accurately simulate tokamak configurations such as the Next European Torus (NET) and the Fusion Engineering Reactor (FER). 18 refs., 2 figs., 4 tabs

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

    International Nuclear Information System (INIS)

    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

  15. Thermal-hydraulic tests for reactor safety system

    International Nuclear Information System (INIS)

    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

  16. Reactor power cutback system test experience at YGN 4

    International Nuclear Information System (INIS)

    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

  17. Fast Shutdown System tests in the Georgia Tech Research Reactor

    International Nuclear Information System (INIS)

    The Fast Shutdown System (FSS) is a new safety system design concept being considered for in installation in the Savannah River (SRS) production reactors. This system is expected to mitigate the consequences of a Design Basis Loss of Coolant Accident, and therefore allow higher operational power levels. A test of this system in the Georgia Tech Research Reactor is proposed to demonstrate the efficacy of this concept. Three tests will be conducted at full power (5MW) and one at low power (100kw). Two full power tests will be conducted with the FSS rod backfilled with one (1) atmosphere of He-4, and one with the rod evacuated. The low power conducted with the FSS rod evacuated. Neutron flux and pressure data will be collected with an independent data acquisition system (DAS). Safety issues associated with the performance of the Fast Shutdown System experiments are addressed in this report. The credible accident scenarios were analyzed using worst case scenarios to demonstrate that no significant nuclear or personnel safety hazards would result from the performance of the proposed experiments

  18. 77 FR 36014 - Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors

    Science.gov (United States)

    2012-06-15

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

  19. New AB-Thermonuclear Reactor for Aerospace

    OpenAIRE

    Bolonkin, Alexander

    2007-01-01

    There are two main methods of nulcear fusion: inertial confinement fusion (ICF) and magnetic confinement fusion (MCF). Existing thermonuclear reactors are very complex, expensive, large, and heavy. They cannot achieve the Lawson creterion. The author offers an innovation. ICF has on the inside surface of the shell-shaped combustion chamber a covering of small Prism Reflectors (PR) and plasma reflector. These prism reflectors have a noteworthy advantage, in comparison with conventional mirror ...

  20. High-Temperature Engineering Test Reactor door valve monitor system

    International Nuclear Information System (INIS)

    This manual describes the detector design features, performance, and operating characteristics of the High-Temperature Engineering Test Reactor (HTTR) Door Valve Monitor System spent-fuel monitor. The HTTR Door Valve Monitor System (HDVM) is installed in the HTTR door valve to provide unattended monitoring data for the transfer of spent fuel through the door valve on the top of the reactor. The system includes a pair of detectors to provide direction of travel and redundancy. The fission product gamma rays are measured using ion chambers (ICs) and the curium neutrons are measured using shielded 3He detectors. There are two ICs and one 3He tube inside each detector package. Gamma-ray and neutron detector (GRAND) electronics supply power to the ICs and 3He tubes, and the data are collected in the GRAND and the Field Works computer. The system is designed to operate unattended with data pickup by the inspectors on a 90-day period. This manual gives the performance and calibration procedures

  1. New AB-Thermonuclear Reactor for Aerospace

    CERN Document Server

    Bolonkin, Alexander

    2007-01-01

    There are two main methods of nulcear fusion: inertial confinement fusion (ICF) and magnetic confinement fusion (MCF). Existing thermonuclear reactors are very complex, expensive, large, and heavy. They cannot achieve the Lawson creterion. The author offers an innovation. ICF has on the inside surface of the shell-shaped combustion chamber a covering of small Prism Reflectors (PR) and plasma reflector. These prism reflectors have a noteworthy advantage, in comparison with conventional mirror and especially with conventional shell: they multi-reflect the heat and laser radiation exactly back into collision with the fuel target capsule (pellet). The plasma reflector reflects the Bremsstrahlung radiation. The offered innovation decreases radiation losses, creates significant radiation pressure and increases the reaction time. The Lawson criterion increases by hundreds of times. The size, cost, and weight of a typical installation will decrease by tens of times. The author is researching the efficiency of these i...

  2. Testing of an advanced thermochemical conversion reactor system

    Energy Technology Data Exchange (ETDEWEB)

    1990-01-01

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

  3. Reliability-based econometrics of aerospace structural systems: Design criteria and test options. Ph.D. Thesis - Georgia Inst. of Tech.

    Science.gov (United States)

    Thomas, J. M.; Hanagud, S.

    1974-01-01

    The design criteria and test options for aerospace structural reliability were investigated. A decision methodology was developed for selecting a combination of structural tests and structural design factors. The decision method involves the use of Bayesian statistics and statistical decision theory. Procedures are discussed for obtaining and updating data-based probabilistic strength distributions for aerospace structures when test information is available and for obtaining subjective distributions when data are not available. The techniques used in developing the distributions are explained.

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

    International Nuclear Information System (INIS)

    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)

  5. Machine intelligence and autonomy for aerospace systems

    Science.gov (United States)

    Heer, Ewald (Editor); Lum, Henry (Editor)

    1988-01-01

    The present volume discusses progress toward intelligent robot systems in aerospace applications, NASA Space Program automation and robotics efforts, the supervisory control of telerobotics in space, machine intelligence and crew/vehicle interfaces, expert-system terms and building tools, and knowledge-acquisition for autonomous systems. Also discussed are methods for validation of knowledge-based systems, a design methodology for knowledge-based management systems, knowledge-based simulation for aerospace systems, knowledge-based diagnosis, planning and scheduling methods in AI, the treatment of uncertainty in AI, vision-sensing techniques in aerospace applications, image-understanding techniques, tactile sensing for robots, distributed sensor integration, and the control of articulated and deformable space structures.

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

    Directory of Open Access Journals (Sweden)

    Hyun-Kook Shin

    2006-08-01

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

  7. Lessons learned from modal testing of aerospace structures

    Science.gov (United States)

    Hunt, David L.; Brillhart, Ralph D.

    1993-02-01

    The primary factors affecting the accuracy and the time required to perform modal tests on aerospace structures are discussed, and the lessons learned from modal tests performed over the past 15 yrs are examined. Case histories of modal testing on aerospace structures are reviewed, including the Galileo satellite and the Space Shuttle solid rocket motor and test stand. Currently recommended approaches to the modal testing are addressed.

  8. Initial testing of the tritium systems at the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, J.L.; Sissingh, R.A.P.; Gentile, C.A.; Rossmassler, R.L.; Walters, R.T.; Voorhees, D.R.

    1993-11-01

    The Tokamak Fusion Test Reactor (TFTR) at Princeton will start its D-T experiments in late 1993, introducing and operating the tokamak with tritium in order to begin the study of burning plasma physics in D-T. Trace tritium injection experiments, using small amounts of tritium will begin in the fall of 1993. In preparation for these experiments, a series of tests with low concentrations of tritium inn deuterium have been performed as an initial qualification of the tritium systems. These tests began in April 1993. This paper describes the initial testing of the equipment in the TFTR tritium facility.

  9. Initial testing of the tritium systems at the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    The Tokamak Fusion Test Reactor (TFTR) at Princeton will start its D-T experiments in late 1993, introducing and operating the tokamak with tritium in order to begin the study of burning plasma physics in D-T. Trace tritium injection experiments, using small amounts of tritium will begin in the fall of 1993. In preparation for these experiments, a series of tests with low concentrations of tritium inn deuterium have been performed as an initial qualification of the tritium systems. These tests began in April 1993. This paper describes the initial testing of the equipment in the TFTR tritium facility

  10. Materials Selection for Aerospace Systems

    Science.gov (United States)

    Arnold, Steven M.; Cebon, David; Ashby, Mike

    2012-01-01

    A systematic design-oriented, five-step approach to material selection is described: 1) establishing design requirements, 2) material screening, 3) ranking, 4) researching specific candidates and 5) applying specific cultural constraints to the selection process. At the core of this approach is the definition performance indices (i.e., particular combinations of material properties that embody the performance of a given component) in conjunction with material property charts. These material selection charts, which plot one property against another, are introduced and shown to provide a powerful graphical environment wherein one can apply and analyze quantitative selection criteria, such as those captured in performance indices, and make trade-offs between conflicting objectives. Finding a material with a high value of these indices maximizes the performance of the component. Two specific examples pertaining to aerospace (engine blades and pressure vessels) are examined, both at room temperature and elevated temperature (where time-dependent effects are important) to demonstrate the methodology. The discussion then turns to engineered/hybrid materials and how these can be effectively tailored to fill in holes in the material property space, so as to enable innovation and increases in performance as compared to monolithic materials. Finally, a brief discussion is presented on managing the data needed for materials selection, including collection, analysis, deployment, and maintenance issues.

  11. Integration of pyrotechnics into aerospace systems

    Science.gov (United States)

    Bement, Laurence J.; Schimmel, Morry L.

    1993-01-01

    The application of pyrotechnics to aerospace systems has been resisted because normal engineering methods cannot be used in design and evaluation. Commonly used approaches for energy sources, such as electrical, hydraulic and pneumatic, do not apply to explosive and pyrotechnic devices. This paper introduces the unique characteristics of pyrotechnic devices, describes how functional evaluations can be conducted, and demonstrates an engineering approach for pyrotechnic integration. Logic is presented that allows evaluation of two basic types of pyrotechnic systems to demonstrate functional margin.

  12. Developing IVHM Requirements for Aerospace Systems

    Science.gov (United States)

    Rajamani, Ravi; Saxena, Abhinav; Kramer, Frank; Augustin, Mike; Schroeder, John B.; Goebel, Kai; Shao, Ginger; Roychoudhury, Indranil; Lin, Wei

    2013-01-01

    The term Integrated Vehicle Health Management (IVHM) describes a set of capabilities that enable sustainable and safe operation of components and subsystems within aerospace platforms. However, very little guidance exists for the systems engineering aspects of design with IVHM in mind. It is probably because of this that designers have to use knowledge picked up exclusively by experience rather than by established process. This motivated a group of leading IVHM practitioners within the aerospace industry under the aegis of SAE's HM-1 technical committee to author a document that hopes to give working engineers and program managers clear guidance on all the elements of IVHM that they need to consider before designing a system. This proposed recommended practice (ARP6883 [1]) will describe all the steps of requirements generation and management as it applies to IVHM systems, and demonstrate these with a "real-world" example related to designing a landing gear system. The team hopes that this paper and presentation will help start a dialog with the larger aerospace community and that the feedback can be used to improve the ARP and subsequently the practice of IVHM from a systems engineering point-of-view.

  13. On the classification of structures, systems and components of nuclear research and test reactors

    International Nuclear Information System (INIS)

    The classification of structures, systems and components of nuclear reactors is a relevant issue related to their design because it is directly associated with their safety functions. There is an important statement regarding quality standards and records that says Structures, systems, and components important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed. The definition of the codes, standards and technical requirements applied to the nuclear reactor design, fabrication, inspection and tests may be seen as the main result from this statement. There are well established guides to classify structures, systems and components for nuclear power reactors such as the Pressurized Water Reactors but one can not say the same for nuclear research and test reactors. The nuclear reactors safety functions are those required to the safe reactor operation, the safe reactor shutdown and continued safe conditions, the response to anticipated transients, the response to potential accidents and the control of radioactive material. So, it is proposed in this paper an approach to develop the classification of structures, systems and components of these reactors based on their intended safety functions in order to define the applicable set of codes, standards and technical requirements. (author)

  14. Thermal hydraulic tests for reactor safety system -Research on the improvement of nuclear safety-

    International Nuclear Information System (INIS)

    The present research aims at the development of the thermal hydraulic verification test technology for the reactor safety system of the conventional and advanced nuclear power plant and the development of the advanced thermal hydraulic measuring techniques. (Author)

  15. The reactor core configuration and important systems related to physics tests of Daya Bay NPP

    International Nuclear Information System (INIS)

    A brief introduction to reactor core configuration and important systems related to physics tests of Daya Bay NPP is given. These systems involve the reactor core system (COR), the full length rod control system (RGL), the in-core instrumentation system (RIC), the out-of-core nuclear instrumentation system (RPN), and the LOCA surveillance system (LSS), the centralized data processing system (KIT) and the test data acquisition system (KDO). In addition, that the adjustment and evaluation of boron concentration related to other systems, for example the reactor coolant system (RCP), the chemical and volume control system (RCV), the reactor boron and water makeup system (REA), the nuclear sampling system (REN) and the reactor control system (RRC), etc. is also described. Analysis of these systems helps not only to familiarize their functions and acquires a deepen understanding for the principle procedure, points for attention and technical key of the core physics tests, but also to further analyze the test results. (3 refs., 11 figs., 1 tab.)

  16. Advances in control system technology for aerospace applications

    CERN Document Server

    2016-01-01

    This book is devoted to Control System Technology applied to aerospace and covers the four disciplines Cognitive Engineering, Computer Science, Operations Research, and Servo-Mechanisms. This edited book follows a workshop held at the Georgia Institute of Technology in June 2012, where the today's most important aerospace challenges, including aerospace autonomy, safety-critical embedded software engineering, and modern air transportation were discussed over the course of two days of intense interactions among leading aerospace engineers and scientists. Its content provide a snapshot of today's aerospace control research and its future, including Autonomy in space applications, Control in space applications, Autonomy in aeronautical applications, Air transportation, and Safety-critical software engineering.

  17. Artificial Immune System Approaches for Aerospace Applications

    Science.gov (United States)

    KrishnaKumar, Kalmanje; Koga, Dennis (Technical Monitor)

    2002-01-01

    Artificial Immune Systems (AIS) combine a priori knowledge with the adapting capabilities of biological immune system to provide a powerful alternative to currently available techniques for pattern recognition, modeling, design, and control. Immunology is the science of built-in defense mechanisms that are present in all living beings to protect against external attacks. A biological immune system can be thought of as a robust, adaptive system that is capable of dealing with an enormous variety of disturbances and uncertainties. Biological immune systems use a finite number of discrete "building blocks" to achieve this adaptiveness. These building blocks can be thought of as pieces of a puzzle which must be put together in a specific way-to neutralize, remove, or destroy each unique disturbance the system encounters. In this paper, we outline AIS models that are immediately applicable to aerospace problems and identify application areas that need further investigation.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-10-01

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

  19. Computational Modeling of Flow Control Systems for Aerospace Vehicles Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Clear Science Corp. proposes to develop computational methods for designing active flow control systems on aerospace vehicles with the primary objective of...

  20. The verifying test of refueling system of the China experimental fast reactor

    International Nuclear Information System (INIS)

    The article introduce the verifying test of refueling system of China Experimental Fast Reactor. The purposes of the test is to check the performance of the equipment of refueling system, and to verify the requirement for the SCADA (Supervisory Control and Data Acquisition) system, and to verify the refueling SCADA system. For these purposes the test platform and device were built. For the first time in China, the simulated automated refueling was realized on the platform. This test has established the base for the test of refueling system on CEFR. (authors)

  1. A sipping test system to the Triga Mark I IPR-R1 reactor

    International Nuclear Information System (INIS)

    The IPR-R1 TRIGA MARK I Research Reactor of the Nuclear Technology Development Centre (CDTN/CNEN-MG) is a tank type reactor of General Atomic Company that has been operating since 1960 at a power of 100 kW. At present there are 63 fuel rods at the reactor core (58 aluminum cladding, 5 stainless steel and 1 stainless steel instrumented). The oldest fuel elements are made with aluminum alloy and the new ones from stainless steel. Some of the old fuel rods present some spots along their lateral fuel plates. These spots are originated by galvanic corrosion between the fuel cladding and the aluminum core grid. To provide an ageing program to the reactor, a sipping tests system will be performed with the reactor fuel. The system intends evaluate the possible presence of 137Cs leaking rate. This work presents the system, the procedure and methodology that will is used to perform the sipping tests with the fuel rods at the reactor core. The results obtained for the 137Cs sipping water activity for some fuel assembly, if any, will be evaluated with the system in operation. A correlation between the possible corrosion and the activity values measured will be realized. (author)

  2. Experience in the maintenance of sodium systems of fast breeder test reactor

    International Nuclear Information System (INIS)

    The fast breeder test reactor (FBTR) is a loop type sodium cooled fast reactor located at Kalpakkam in India and that has been operating for 25 years. The reactor has been operated up to a power level of 18.6 MWt with a sodium outlet temperature of 482 C. degrees. Several modifications were carried out in the sodium systems to improve the plant performance. During the course of operation of the reactor, a number of sodium laden components like pumps, valves, cold traps, rupture disks, level probes, shielding plugs, control rod drive mechanisms, experimental assemblies, piping... were removed for various maintenance, modification and replacement jobs which has given the operators a valuable experience in handling large scale sodium systems. This paper details the special procedures followed during the handling of active and inactive sodium laden components

  3. Y2K issues for real time computer systems for fast breeder test reactor

    International Nuclear Information System (INIS)

    Presentation shows the classification of real time systems related to operation, control and monitoring of the fast breeder test reactor. Software life cycle includes software requirement specification, software design description, coding, commissioning, operation and management. A software scheme in supervisory computer of fast breeder test rector is described with the twenty years of experience in design, development, installation, commissioning, operation and maintenance of computer based supervision control system for nuclear installation with a particular emphasis on solving the Y2K problem

  4. Reactor Simulator Testing

    Science.gov (United States)

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

    2013-01-01

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

  5. 78 FR 64027 - Preoperational Testing of Emergency Core Cooling Systems for Pressurized-Water Reactors

    Science.gov (United States)

    2013-10-25

    ... on June 7, 2011 (76 FR 32878), for a 60-day public comment period. The public comment period closed... published for public comment on June 15, 2012 (77 FR 36014). A total of 45 comments were received on DG-1277... COMMISSION Preoperational Testing of Emergency Core Cooling Systems for Pressurized-Water Reactors...

  6. Efficiency Testing of the Air Cleaning System for a High Temperature Reactor

    International Nuclear Information System (INIS)

    The Los Alamos Ultra High Temperature Reactor Experiment (UHTREX) utilizes a helium-cooled, graphite-moderated reactor, employing refractory fuel elements. Under accident conditions, the effluent that may be released from this reactor requires an air-cleaning system capable of reducing radioactive gas and particulate contaminants to safe levels. Dioctyl phthalate and iodine-131 were used as test aerosols for the HEPA and activated carbon filters, respectively. Methods of aerosol generation and test procedures are detailed for the preinstallation tests of the carbon and in-place testing of the carbon and HEPA filters. The importance of visual inspection of the HEPA filters prior to installation and supervision of filter installation is discussed. In-place tests indicated desirable design changes which would (1) simplify in-place testing procedures, (2) expedite installation and future changing of the filters, and (3) ensure operation of a more efficient system. Problems encountered during in-place testing, recommendations for the design of similar systems, and acceptance criteria used at LASL are discussed. (author)

  7. 'DPS-1 SKODA' diagnostic system for the reactor control rod drives functional and lifetime tests

    International Nuclear Information System (INIS)

    The 'DSP-J SKODA' diagnostic system of the reactor control rod drives (VVER-440, 213 type) is described in this paper. The hardware structure, methods and utility software of the diagnostic system is explained. The main goal of this system is defined: to ensure the functional availability and longer lifetime of modernized drives (15 to 20 years). Experiences from the measurements, evaluation and analysis with the 'DSP-1 SKODA' system in die testing room in SKODA - Bolevec are introduced. The results of functional and lifetime tests of prototype drive reductors are presented. (author)

  8. Test reactor technology

    International Nuclear Information System (INIS)

    The Reactor Development Program created a need for engineering testing of fuels and materials. The Engineering Test Reactors were developed around the world in response to this demand. The design of the test reactors proved to be different from that of power reactors, carrying the fuel elements closer to the threshold of failure, requiring more responsive instrumentation, more rapid control element action, and inherent self-limiting behavior under accident conditions. The design of the experimental facilities to exploit these reactors evolved a new, specialized, branch of engineering, requiring a very high-lvel scientific and engineering team, established a meticulous concern with reliability, the provision for recovery from their own failures, and detailed attention to possible interactions with the test reactors. This paper presents this technology commencing with the Materials Testing Reactor (MTR) through the Fast Flux Test Facility, some of the unique experimental facilities developed to exploit them, but discusses only cursorily the experiments performed, since sample preparation and sample analyses were, and to some extent still are, either classified or proprietary. The Nuclear Engineering literature is filled with this information

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

    International Nuclear Information System (INIS)

    Most test reactors are equipped with shuttle facilities (sometimes called rabbit tubes) whereby small capsules can be inserted into the reactor and retrieved during power operations. With the installation of Hydraulic Shuttle Irradiation System (HSIS) this capability has been restored to the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). The general design and operating principles of this system were patterned after the hydraulic rabbit at Oak Ridge National Laboratory's (ORNL) High Flux Isotope Reactor (HFIR), which has operated successfully for many years. Using primary coolant as the motive medium the HSIS system is designed to simultaneously transport fourteen shuttle capsules, each 16 mm OD x 57 mm long, to and from the B-7 position of the reactor. The B-7 position is one of the higher flux positions in the reactor with typical thermal and fast (>1 Mev) fluxes of 2.8E+14 n/cm2/sec and 1.9E+14 n/cm2/sec respectively. The available space inside each shuttle is approximately 14 mm diameter x 50 mm long. The shuttle containers are made from titanium which was selected for its low neutron activation properties and durability. Shuttles can be irradiated for time periods ranging from a few minutes to several months. The Send and Receive Station (SRS) for the HSIS is located 2.5 m deep in the ATR canal which allows irradiated shuttles to be easily moved from the SRS to a wet loaded cask, or transport pig. The HSIS system first irradiated (empty) shuttles in September 2009 and has since completed a Readiness Assessment in November 2009. The HSIS is a key component of the ATR National Scientific User Facility (NSUF) operated by Battelle Energy Alliance, LLC and is available to a wide variety of university researchers for nuclear fuels and materials experiments as well as medical isotope research and production.

  10. On-Line Information/Measuring System to Support In-Reactor Tests

    International Nuclear Information System (INIS)

    Full text: At the WWR-K reactor there is a universal loop facility (ULF), which is designed to provide the necessary test conditions in the experimental channels of the core. By means of the ULF, necessary environment (nitrogen/helium/vacuum at the predefined pressure and temperature) is created in a channel which houses a sample under studies. To support the reactor tests, the ULF is equipped with the measurement/information system, which makes it possible to provide operators and experimentalists with on-line test-related information. The measurement/information system is a complete set of some technical means (microcontrollers, analog and digital signals modules, power supply units, etc.) and software. (authors)

  11. Risk communication strategy development using the aerospace systems engineering process

    Science.gov (United States)

    Dawson, S.; Sklar, M.

    2004-01-01

    This paper explains the goals and challenges of NASA's risk communication efforts and how the Aerospace Systems Engineering Process (ASEP) was used to map the risk communication strategy used at the Jet Propulsion Laboratory to achieve these goals.

  12. Heat exchanger performance in main cooling system on high temperature test operation at high temperature gas-cooled reactor 'HTTR'

    International Nuclear Information System (INIS)

    High Temperature Engineering Test Reactor (HTTR) of high temperature gas-cooled reactor at Japan Atomic Energy Research Institute achieved the reactor outlet coolant temperature of 950degC for the first time in the world at Apr.19, 2004. To remove generated heat at reactor core and to hold reactor inlet coolant temperature as specified temperature, heat exchangers in HTTR main cooling system should have designed heat exchange performance. In this report, heat exchanger performance is evaluated based on measurement data in high temperature test operation. And it is confirmed the adequacy of heat exchanger designing method by comparison of evaluated value with designed value. (author)

  13. Reactor Simulator Testing Overview

    Science.gov (United States)

    Schoenfeld, Michael P.

    2013-01-01

    Test Objectives Summary: a) Verify operation of the core simulator, the instrumentation & control system, and the ground support gas and vacuum test equipment. b) Examine cooling & heat regeneration performance of the cold trap purification. c) Test the ALIP pump at voltages beyond 120V to see if the targeted mass flow rate of 1.75 kg/s can be obtained in the RxSim. Testing Highlights: a) Gas and vacuum ground support test equipment performed effectively for operations (NaK fill, loop pressurization, and NaK drain). b) Instrumentation & Control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings and ramped within prescribed constraints. It effectively interacted with reactor simulator control model and defaulted back to temperature control mode if the transient fluctuations didn't dampen. c) Cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained which was lower than the predicted 750 K but 156 K higher than the minimum temperature indicating the design provided some heat regeneration. d) ALIP produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-08

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

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

    International Nuclear Information System (INIS)

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

  16. Risk-based management system development for the Advanced Test Reactor

    International Nuclear Information System (INIS)

    A Risk-Based Management System (RBMS) is being developed to facilitate the use of the Advanced Test Reactor (ATR) probabilistic risk assessment to support ATR operation. Most ATR RBMS questions can best be answered using the System Analysis and Risk Assessment System (SARA) developed at the Idaho National Engineering Laboratory. However, some applications may require employment of the other four codes used to develop and report the PRA. These four codes include the Integrated Reliability and Risk Analysis System (IRRAS), SETS, ETA-II, and the Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR). The ATR RBMS will evolve over three years, and will include the results of the Level 3 and external events analysis

  17. Test Results of Reactor Coolant System Natural Circulation using the SMART-ITL Facility

    International Nuclear Information System (INIS)

    In this paper, the Sequence Of Event (SOE) and test conditions of RCS natural circulation test using SMART-ITL are presented, and the major measuring parameters and the test results will be introduced. In this test, the steady state operation satisfied the initial condition of the prescribed test procedure and the boundary conditions were properly simulated. After the RCPs stop, the RCS natural circulation flow was generated by heating in the core region and cooling in the SG heat exchanger region, and the major thermalhydraulic parameters reached at a stable condition. Through this experiment, it has been validated that the SMART-ITL facility can adequately simulate the RCS natural circulation behavior. In addition, it is expected that the experimental data can be used for the code assessment of the TASS/SMR-S code and experiences from this test can be utilized to the subsequent SBLOCA simulation test. SMART (System-integrated Modular Advanced Reactor) is an integral type reactor which major primary components such as the steam generator, the pressurizer, and the reactor coolant pump are installed inside one single reactor vessel and connecting primary pipes are removed. The TASS/SMR-S code is used to perform the performance and safety analysis of the SMART. To evaluate the capability of TASS/SMR-S code on the natural circulation and accident scenarios such as Small-Break Loss of Coolant Accident (SBLOCA) for predicting the thermal-hydraulic phenomena in steady state and transient operation, it is essential to perform a series of validation tests

  18. On-line test of power distribution prediction system for boiling water reactors

    International Nuclear Information System (INIS)

    A power distribution prediction system for boiling water reactors has been developed and its on-line performance test has proceeded at an operating commercial reactor. This system predicts the power distribution or thermal margin in advance of control rod operations and core flow rate change. This system consists of an on-line computer system, an operator's console with a color cathode-ray tube, and plant data input devices. The main functions of this system are present power distribution monitoring, power distribution prediction, and power-up trajectory prediction. The calculation method is based on a simplified nuclear thermal-hydraulic calculation, which is combined with a method of model identification to the actual reactor core state. It has been ascertained by the on-line test that the predicted power distribution (readings of traversing in-core probe) agrees with the measured data within 6% root-mean-square. The computing time required for one prediction calculation step is less than or equal to 1.5 min by an HIDIC-80 on-line computer

  19. Real time simulator for material testing reactor

    International Nuclear Information System (INIS)

    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)

  20. An overview of Ball Aerospace cryogen storage and delivery systems

    Science.gov (United States)

    Marquardt, J.; Keller, J.; Mills, G.; Schmidt, J.

    2015-12-01

    Starting on the Gemini program in the 1960s, Beech Aircraft (now Ball Aerospace) has been designing and manufacturing dewars for a variety of cryogens including liquid hydrogen and oxygen. These dewars flew on the Apollo, Skylab and Space Shuttle spacecraft providing fuel cell reactants resulting in over 150 manned spaceflights. Since Space Shuttle, Ball has also built the liquid hydrogen fuel tanks for the Boeing Phantom Eye unmanned aerial vehicle. Returning back to its fuel cell days, Ball has designed, built and tested a volume-constrained liquid hydrogen and oxygen tank system for reactant delivery to fuel cells on unmanned undersea vehicles (UUVs). Herein past history of Ball technology is described. Testing has been completed on the UUV specific design, which will be described.

  1. Preliminary Test Requirements for the Performance Test of Passive Decay Heat Removal System of Sodium-Cooled Fast Reactor

    International Nuclear Information System (INIS)

    In order to verify the concept of safety grade passive decay removal system PDRC (Passive Decay heat Removal Circuit) of KALIMER-600 and the design features to resolve the design issues for securing the cooling performance, the performance test is implemented. In this report, the preliminary test requirements for using as a guideline to the design of the experimental facility were established. Since the experimental facility should be designed so as to simulate the various thermal- hydraulic phenomena, as closely as possible, to be occurred in reference reactor during the decay heat removal operation, the design characteristics of the reference reactor (KALIMER-600) were analyzed for drawing major constitutive elements to be simulated in the facility. The preliminary test matrix was set up by the analysis of various design basis events and then the key test matrix was determined. Also, the priority for various thermal hydraulic phenomena which should be considered in the design of the experimental facility was determined by analyzing the phenomena for each key test matrix. Based on the analysis, the general design requirements for experimental facility were prepared and the design requirements for fluid systems and instrumentation were established. The test requirements in this report will be reflected in the scaling analysis and the basic design of the experimental facility. The test matrix specified in this report can be modified in the stage of main testing by considering the needs of experiments and circumstances at that time

  2. SP-100 nuclear space power reactor system hardware and testing progress

    International Nuclear Information System (INIS)

    The SP-100 Space Reactor System was established by agencies of the US government as the system of choice to meet the nation's long lifetime, high reliability space power needs in the 10's to 100's of kWe power range. SP-100 is compatible with all power conversion technologies that can utilize reactor coolant temperatures ≤ 1,350 K. The technologies incorporated in SP-100 are directly applicable to earth orbiting satellites, planetary probes or surface power for commercial, military or civil missions. The most significant hardware and testing accomplishments that were made during the past year are reported in this summary paper, including fuel, fabrication technologies, control mechanisms, liquid metal pumps, lithium thaw behavior and characterization, and thermoelectric power conversion

  3. 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.

  4. FASTER test reactor preconceptual design report summary

    International Nuclear Information System (INIS)

    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.

  5. Novel atmospheric extinction measurement techniques for aerospace laser system applications

    Science.gov (United States)

    Sabatini, Roberto; Richardson, Mark

    2013-01-01

    Novel techniques for laser beam atmospheric extinction measurements, suitable for manned and unmanned aerospace vehicle applications, are presented in this paper. Extinction measurements are essential to support the engineering development and the operational employment of a variety of aerospace electro-optical sensor systems, allowing calculation of the range performance attainable with such systems in current and likely future applications. Such applications include ranging, weaponry, Earth remote sensing and possible planetary exploration missions performed by satellites and unmanned flight vehicles. Unlike traditional LIDAR methods, the proposed techniques are based on measurements of the laser energy (intensity and spatial distribution) incident on target surfaces of known geometric and reflective characteristics, by means of infrared detectors and/or infrared cameras calibrated for radiance. Various laser sources can be employed with wavelengths from the visible to the far infrared portions of the spectrum, allowing for data correlation and extended sensitivity. Errors affecting measurements performed using the proposed methods are discussed in the paper and algorithms are proposed that allow a direct determination of the atmospheric transmittance and spatial characteristics of the laser spot. These algorithms take into account a variety of linear and non-linear propagation effects. Finally, results are presented relative to some experimental activities performed to validate the proposed techniques. Particularly, data are presented relative to both ground and flight trials performed with laser systems operating in the near infrared (NIR) at λ = 1064 nm and λ = 1550 nm. This includes ground tests performed with 10 Hz and 20 kHz PRF NIR laser systems in a large variety of atmospheric conditions, and flight trials performed with a 10 Hz airborne NIR laser system installed on a TORNADO aircraft, flying up to altitudes of 22,000 ft.

  6. Proficiency Testing for Evaluating Aerospace Materials Test Anomalies

    Science.gov (United States)

    Hirsch, D.; Motto, S.; Peyton, S.; Beeson, H.

    2006-01-01

    ASTM G 86 and ASTM G 74 are commonly used to evaluate materials susceptibility to ignition in liquid and gaseous oxygen systems. However, the methods have been known for their lack of repeatability. The inherent problems identified with the test logic would either not allow precise identification or the magnitude of problems related to running the tests, such as lack of consistency of systems performance, lack of adherence to procedures, etc. Excessive variability leads to increasing instances of accepting the null hypothesis erroneously, and so to the false logical deduction that problems are nonexistent when they really do exist. This paper attempts to develop and recommend an approach that could lead to increased accuracy in problem diagnostics by using the 50% reactivity point, which has been shown to be more repeatable. The initial tests conducted indicate that PTFE and Viton A (for pneumatic impact) and Buna S (for mechanical impact) would be good choices for additional testing and consideration for inter-laboratory evaluations. The approach presented could also be used to evaluate variable effects with increased confidence and tolerance optimization.

  7. Reliability-based design optimization of multiphysics, aerospace systems

    Science.gov (United States)

    Allen, Matthew R.

    Aerospace systems are inherently plagued by uncertainties in their design, fabrication, and operation. Safety factors and expensive testing at the prototype level traditionally account for these uncertainties. Reliability-based design optimization (RBDO) can drastically decrease life-cycle development costs by accounting for the stochastic nature of the system response in the design process. The reduction in cost is amplified for conceptually new designs, for which no accepted safety factors currently exist. Aerospace systems often operate in environments dominated by multiphysics phenomena, such as the fluid-structure interaction of aeroelastic wings or the electrostatic-mechanical interaction of sensors and actuators. The analysis of such phenomena is generally complex and computationally expensive, and therefore is usually simplified or approximated in the design process. However, this leads to significant epistemic uncertainties in modeling, which may dominate the uncertainties for which the reliability analysis was intended. Therefore, the goal of this thesis is to present a RBDO framework that utilizes high-fidelity simulation techniques to minimize the modeling error for multiphysics phenomena. A key component of the framework is an extended reduced order modeling (EROM) technique that can analyze various states in the design or uncertainty parameter space at a reduced computational cost, while retaining characteristics of high-fidelity methods. The computational framework is verified and applied to the RBDO of aeroelastic systems and electrostatically driven sensors and actuators, utilizing steady-state analysis and design criteria. The framework is also applied to the design of electrostatic devices with transient criteria, which requires the use of the EROM technique to overcome the computational burden of multiple transient analyses.

  8. Standard Test Method for Environmental Resistance of Aerospace Transparencies

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This test method covers determination of the effects of exposure to thermal shock, condensing humidity, and simulated weather on aerospace transparent enclosures. 1.2 This test method is not recommended for quality control nor is it intended to provide a correlation to actual service life. 1.3 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard. 1.3.1 Exceptions—Certain inch-pound units are furnished in parentheses (not mandatory) and certain temperatures in Fahrenheit associated with other standards are also furnished. 1.4 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

  9. Research and Development of Rapid Design Systems for Aerospace Structure

    Science.gov (United States)

    Schaeffer, Harry G.

    1999-01-01

    This report describes the results of research activities associated with the development of rapid design systems for aerospace structures in support of the Intelligent Synthesis Environment (ISE). The specific subsystems investigated were the interface between model assembly and analysis; and, the high performance NASA GPS equation solver software system in the Windows NT environment on low cost high-performance PCs.

  10. Biomedical Application of Aerospace Personal Cooling Systems

    Science.gov (United States)

    Ku, Yu-Tsuan E.; Lee, Hank C.; Montgomery, Leslie D.; Webbon, Bruce W.; Kliss, Mark (Technical Monitor)

    1997-01-01

    Personal thermoregulatory systems which are used by astronauts to alleviate thermal stress during extravehicular activity have been applied to the therapeutic management of multiple sclerosis. However, little information is available regarding the physiologic and circulatory changes produced by routine operation of these systems. The objectives of this study were to compare the effectiveness of two passive and two active cooling vests and to measure the body temperature and circulatory changes produced by each cooling vest configuration. The MicroClimate Systems and the Life Enhancement Tech(LET) lightweight liquid cooling vests, the Steele Vest and LET's Zipper Front Garment were used to cool the chest region of 10 male and female subjects (25 to 55 yr.) in this study. Calf, forearm and finger blood flows were measured using a tetrapolar impedance rheograph. The subjects, seated in an upright position at normal room temperature (approx.22C), were tested for 60 min. with the cooling system operated at its maximum cooling capacity. Blood flows were recorded continuously using a computer data acquisition system with a sampling frequency of 250 Hz. Oral, right and left ear temperatures and cooling system parameters were logged manually every 5 min. Arm, leg, chest and rectal temperatures; heart rate; respiration; and an activity index were recorded continuously on a U.F.I., Inc. Biolog ambulatory monitor. In general, the male and female subjects' oral and ear temperature responses to cooling were similar for all vest configurations tested. Oral temperatures during the recovery period were significantly (Pcooling and recovery periods.

  11. Web-enabled work permit system for fast breeder test reactor

    International Nuclear Information System (INIS)

    The objective of this project is to computerize and web-enable the Work Permit System for the Fast Breeder Test Reactor (FBTR) at IGCAR, Kalpakkam. The existing Work Permit System at FBTR was studied in detail. Since all the formalities were paper-based, the risk of human error in scrutinizing all permits before reactor start-up was high. Compilation of reports (daily, monthly, yearly etc.) was tedious. The work permit system was therefore automated in order to enable the operation group manage the maintenance work carried out in the plant systematically with entries. The entire project was classified into five permit modules -maintenance, transfer, return, cancellation and reissue. Each module takes care of the entry and maintenance of data in their respective fields in their respective tables. The user is also provided with an option to take a hard copy of the report of his/her choice. A client/server based system was designed to web-enable the entire project. The server program was designed using VB 6.0 as the front-end and MS Access database as the back end to store the data. The client software was developed using Active Server Pages and published using personal web server in the Intranet. A number of administrative tools have been incorporated in the software to ensure access security and integrity of the database. An online help feature with search facilities was added to the software. The work permit system software is now already being used at FBTR and has been deemed to be an invaluable aid in empowering the availability of the reactor and determining the performance history of the equipment. (author)

  12. Biomedical Application of Aerospace Personal Cooling Systems

    Science.gov (United States)

    Ku, Yu-Tsuan E.; Lee, Hank C.; Montgomery, Leslie D.; Webbon, Bruce W.; Kliss, Mark (Technical Monitor)

    1997-01-01

    Personal thermoregulatory systems which are used by astronauts to alleviate thermal stress during extravehicular activity have been applied to the therapeutic management of multiple sclerosis. However, little information is available regarding the physiologic and circulatory changes produced by routine operation of these systems. The objectives of this study were to compare the effectiveness of two passive and two active cooling vests and to measure the body temperature and circulatory changes produced by each cooling vest configuration. The MicroClimate Systems and the Life Enhancement Tech(LET) lightweight liquid cooling vests, the Steele Vest and LET's Zipper Front Garment were used to cool the chest region of 10 male and female subjects (25 to 55 yr.) in this study. Calf, forearm and finger blood flows were measured using a tetrapolar impedance rheograph. The subjects, seated in an upright position at normal room temperature (approx.22C), were tested for 60 min. with the cooling system operated at its maximum cooling capacity. Blood flows were recorded continuously using a computer data acquisition system with a sampling frequency of 250 Hz. Oral, right and left ear temperatures and cooling system parameters were logged manually every 5 min. Arm, leg, chest and rectal temperatures; heart rate; respiration; and an activity index were recorded continuously on a U.F.I., Inc. Biolog ambulatory monitor. In general, the male and female subjects' oral and ear temperature responses to cooling were similar for all vest configurations tested. Oral temperatures during the recovery period were significantly (P<0.05) lower than during the control period, approx. 0.2 - 0.5C, for both men and women wearing any of the four different garments. The corresponding ear temperatures were significantly (P<0.05) decreased approx.0.2 - 0.4C by the end of the recovery period. Compared to the control period, no significant differences were found in rectal temperatures during cooling and

  13. Aerospace Power Systems Design and Analysis (APSDA) Tool

    Science.gov (United States)

    Truong, Long V.

    1998-01-01

    The conceptual design of space and/or planetary electrical power systems has required considerable effort. Traditionally, in the early stages of the design cycle (conceptual design), the researchers have had to thoroughly study and analyze tradeoffs between system components, hardware architectures, and operating parameters (such as frequencies) to optimize system mass, efficiency, reliability, and cost. This process could take anywhere from several months to several years (as for the former Space Station Freedom), depending on the scale of the system. Although there are many sophisticated commercial software design tools for personal computers (PC's), none of them can support or provide total system design. To meet this need, researchers at the NASA Lewis Research Center cooperated with Professor George Kusic from the University of Pittsburgh to develop a new tool to help project managers and design engineers choose the best system parameters as quickly as possible in the early design stages (in days instead of months). It is called the Aerospace Power Systems Design and Analysis (APSDA) Tool. By using this tool, users can obtain desirable system design and operating parameters such as system weight, electrical distribution efficiency, bus power, and electrical load schedule. With APSDA, a large-scale specific power system was designed in a matter of days. It is an excellent tool to help designers make tradeoffs between system components, hardware architectures, and operation parameters in the early stages of the design cycle. user interface. It operates on any PC running the MS-DOS (Microsoft Corp.) operating system, version 5.0 or later. A color monitor (EGA or VGA) and two-button mouse are required. The APSDA tool was presented at the 30th Intersociety Energy Conversion Engineering Conference (IECEC) and is being beta tested at several NASA centers. Beta test packages are available for evaluation by contacting the author.

  14. Propulsion Systems for Aircraft. Aerospace Education II. Instructional Unit II.

    Science.gov (United States)

    Elmer, James D.

    This curriculum guide accompanies another publication in the Aerospace Education II series entitled "Propulsion Systems for Aircraft." The guide includes specific guidelines for teachers on each chapter in the textbook. Suggestions are included for objectives (traditional and behavioral), suggested outline, orientation, suggested key points,…

  15. Sensor Selection and Optimization for Health Assessment of Aerospace Systems

    Science.gov (United States)

    Maul, William A.; Kopasakis, George; Santi, Louis M.; Sowers, Thomas S.; Chicatelli, Amy

    2008-01-01

    Aerospace systems are developed similarly to other large-scale systems through a series of reviews, where designs are modified as system requirements are refined. For space-based systems few are built and placed into service these research vehicles have limited historical experience to draw from and formidable reliability and safety requirements, due to the remote and severe environment of space. Aeronautical systems have similar reliability and safety requirements, and while these systems may have historical information to access, commercial and military systems require longevity under a range of operational conditions and applied loads. Historically, the design of aerospace systems, particularly the selection of sensors, is based on the requirements for control and performance rather than on health assessment needs. Furthermore, the safety and reliability requirements are met through sensor suite augmentation in an ad hoc, heuristic manner, rather than any systematic approach. A review of the current sensor selection practice within and outside of the aerospace community was conducted and a sensor selection architecture is proposed that will provide a justifiable, defendable sensor suite to address system health assessment requirements.

  16. Recent palladium membrane reactor development at the tritium systems test assembly

    International Nuclear Information System (INIS)

    The palladium membrane reactor (PMR) is proving to be a simple and effective means for recovering hydrogen isotopes from fusion fuel impurities such as methane and water. This device directly combines two techniques which have long been utilized for hydrogen processing, namely catalytic shift reactions and palladium/silver permeators. A proof-of-principle (PMR) has been constructed and tested at the Tritium Systems Test Assembly of Los Alamos National Laboratory. The first tests with this device showed that is was effective for the proposed purpose. Initial work concluded that a nickel catalyst was an appropriate choice for use in a PMR. More detailed testing of the PMR with such a catalyst was performed and reported in other works. It was shown that a nickel catalyst-packed PMR did, indeed, recover hydrogen from water and methane with efficiencies approaching 100% in a single processing pass. These experiments were conducted over an extended period of time and no failure or need for regeneration was encountered. These positive results have prompted further PMR development. Topics addressed include alternate PMR geometries and initial testing of the PMR with tritium. These are the subjects of this paper

  17. Operation of the tokamak fusion test reactor tritium systems during initial tritium experiments

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, J.L. [Princeton Plasma Physics Laboratory (United States); Gentile, C. [Princeton Plasma Physics Laboratory (United States); Kalish, M. [Princeton Plasma Physics Laboratory (United States); Kamperschroer, J. [Princeton Plasma Physics Laboratory (United States); Kozub, T. [Princeton Plasma Physics Laboratory (United States); LaMarche, P. [Princeton Plasma Physics Laboratory (United States); Murray, H. [Princeton Plasma Physics Laboratory (United States); Nagy, A. [Princeton Plasma Physics Laboratory (United States); Raftopoulos, S. [Princeton Plasma Physics Laboratory (United States); Rossmassler, R. [Princeton Plasma Physics Laboratory (United States); Sissingh, R. [Princeton Plasma Physics Laboratory (United States); Swanson, J. [Princeton Plasma Physics Laboratory (United States); Tulipano, F. [Princeton Plasma Physics Laboratory (United States); Viola, M. [Princeton Plasma Physics Laboratory (United States); Voorhees, D. [Princeton Plasma Physics Laboratory (United States); Walters, R.T. [Princeton Plasma Physics Laboratory (United States)

    1995-03-01

    The high power D-T experiments on the tokamak fusion test reactor (TFTR) at the Princeton Plasma Physics Laboratory commenced in November 1993. During initial operation of the tritium systems a number of start-up problems surfaced and had to be corrected. These were corrected through a series of system modifications and upgrades and by repair of failed or inadequate components. Even as these operational concerns were being addressed, the tritium systems continued to support D-T operations on the tokamak. During the first six months of D-T operations more than 107kCi of tritium were processed successfully by the tritium systems. D-T experiments conducted at TFTR during this period provided significant new data. Fusion power in excess of 9MW was achieved in May 1994. This paper describes some of the early start-up issues, and reports on the operation of the tritium system and the tritium tracking and accounting system during the early phase of TFTR D-T experiments. (orig.).

  18. Performance testing of irradiation facility rabbit system pneumatic reactor RSG-GAS using standard reference material

    International Nuclear Information System (INIS)

    The irradiation facility function test of rabbit system pneumatic (RS-5) has been done using standard reference material SRM 1633 Coal Fly ash through the sending station. Long irradiation of about 4-5 seconds. The results of qualitative analysis showed that the dominant elements listed in the certificate can be detected are Al, Ca, Mg, Si, Na, Ti, V, Mn. But only an element of Mn and Na which has a relative refractive values below 10%. And the other elements have a value relative refractive index 25% - 60%. The significant difference of value was not influenced by the position of irradiation in the reactor facility but due to the influence of the time difference between the sample and the standard count, and the half-life nuclide itself. Overall it can be said that the performance of the irradiation facility pneumatic rabbit system is good, but needs to be tested again by using different standard reference materials, in order to obtain the test results of analysis that can be trusted. (author)

  19. New lidar systems at the German Aerospace Center

    OpenAIRE

    Kaifler, Bernd; Kaifler, Natalie; Büdenbender, Christian; Witschas, Benjamin; Gomez Kabelka, Pau; Rapp, Markus; Mahnke, Peter; Sauder, Daniel; Geyer, Gerhard; Speiser, Jochen

    2015-01-01

    This work gives an overview of the lower-, middle and upper atmosphere lidar projects at the German Aerospace Center (DLR). The Temperature Lidar for Middle Atmosphere research (TELMA) is a combined sodium/Rayleigh/Brillouin-lidar integrated into an 8-foot container. It will provide temperature profiles with high temporal and spatial resolution from near ground level up to approximately 110 km altitude. The lidar system is designed for remote/autonomous operation. First observations with the...

  20. An operating system for future aerospace vehicle computer systems

    Science.gov (United States)

    Foudriat, E. C.; Berman, W. J.; Will, R. W.; Bynum, W. L.

    1984-01-01

    The requirements for future aerospace vehicle computer operating systems are examined in this paper. The computer architecture is assumed to be distributed with a local area network connecting the nodes. Each node is assumed to provide a specific functionality. The network provides for communication so that the overall tasks of the vehicle are accomplished. The O/S structure is based upon the concept of objects. The mechanisms for integrating node unique objects with node common objects in order to implement both the autonomy and the cooperation between nodes is developed. The requirements for time critical performance and reliability and recovery are discussed. Time critical performance impacts all parts of the distributed operating system; e.g., its structure, the functional design of its objects, the language structure, etc. Throughout the paper the tradeoffs - concurrency, language structure, object recovery, binding, file structure, communication protocol, programmer freedom, etc. - are considered to arrive at a feasible, maximum performance design. Reliability of the network system is considered. A parallel multipath bus structure is proposed for the control of delivery time for time critical messages. The architecture also supports immediate recovery for the time critical message system after a communication failure.

  1. Energy Harvesting System for aerospace application

    OpenAIRE

    Ccorimanya Becerra, Hernan Manuel

    2013-01-01

    [ANGLÈS] The Energy Harvesting is really interesting concept nowadays because it consists in get energy from the environment that is already there. Nowadays application and commons such as nodes of a wireless network self-powered, the flexibility to locate them give an interesting advantages to allocate the network in strategic points or even to difficult places. Alternatively, it can be set out from this energy source a sub-system inside the other sub-system more big, that can be self-powere...

  2. The development of a universal diagnostic probe system for Tokamak fusion test reactor

    Science.gov (United States)

    Mastronardi, R.; Cabral, R.; Manos, D.

    1982-01-01

    The Tokamak Fusion Test Reactor (TFTR), the largest such facility in the U.S., is discussed with respect to instrumentation in general and mechanisms in particular. The design philosophy and detailed implementation of a universal probe mechanism for TFTR is discussed.

  3. Valuation of design adaptability in aerospace systems

    Science.gov (United States)

    Fernandez Martin, Ismael

    As more information is brought into early stages of the design, more pressure is put on engineers to produce a reliable, high quality, and financially sustainable product. Unfortunately, requirements established at the beginning of a new project by customers, and the environment that surrounds them, continue to change in some unpredictable ways. The risk of designing a system that may become obsolete during early stages of production is currently tackled by the use of robust design simulation, a method that allows to simultaneously explore a plethora of design alternatives and requirements with the intention of accounting for uncertain factors in the future. Whereas this design technique has proven to be quite an improvement in design methods, under certain conditions, it fails to account for the change of uncertainty over time and the intrinsic value embedded in the system when certain design features are activated. This thesis introduces the concepts of adaptability and real options to manage risk foreseen in the face of uncertainty at early design stages. The method described herein allows decision-makers to foresee the financial impact of their decisions at the design level, as well as the final exposure to risk. In this thesis, cash flow models, traditionally used to obtain the forecast of a project's value over the years, were replaced with surrogate models that are capable of showing fluctuations on value every few days. This allowed a better implementation of real options valuation, optimization, and strategy selection. Through the option analysis model, an optimization exercise allows the user to obtain the best implementation strategy in the face of uncertainty as well as the overall value of the design feature. Here implementation strategy refers to the decision to include a new design feature in the system, after the design has been finalized, but before the end of its production life. The ability to do this in a cost efficient manner after the system

  4. Test system carrier for the ultrasonic testing of the area of connecting nozzles in the case of pressure vessels, in particular reactor pressure vessels from nuclear power plants

    International Nuclear Information System (INIS)

    In the invention at hand a system carrier for the ultrasonic testing of a reactor pressure vessel is described which enables a test for nozzle welds, pipe fitting welds and nozzle edges to be conducted with a single telescope arm. (RW)

  5. Reactor System Design

    International Nuclear Information System (INIS)

    SMART NPP(Nuclear Power Plant) has been developed for duel purpose, electricity generation and energy supply for seawater desalination. The objective of this project IS to design the reactor system of SMART pilot plant(SMART-P) which will be built and operated for the integrated technology verification of SMART. SMART-P is an integral reactor in which primary components of reactor coolant system are enclosed in single pressure vessel without connecting pipes. The major components installed within a vessel includes a core, twelve steam generator cassettes, a low-temperature self pressurizer, twelve control rod drives, and two main coolant pumps. SMART-P reactor system design was categorized to the reactor coe design, fluid system design, reactor mechanical design, major component design and MMIS design. Reactor safety -analysis and performance analysis were performed for developed SMART=P reactor system. Also, the preparation of safety analysis report, and the technical support for licensing acquisition are performed

  6. Testing the reactor charging machine

    International Nuclear Information System (INIS)

    One of the main objective of the R - D technological engineering program devoted to the Fuel Handling System is domestic production of equipment and technology for testing the ends of the reactor charging machine (MID) destined to Cernavoda NPP, beginning with Unit 2. To achieve the objective based on an own design, a bench-scale testing stand of MIDs which can simulate the pressure, flow-rate, and temperature conditions proper to fuel channels in operating CANDU 600 reactors. The main components of this testing facility are: - fuel channels, cold also test sections, allowing the coupling of MID end upwardly and downwardly, corresponding to the direction of the water flow through the channel; - technological installation feeding with light water the testing sections of the facility in thermohydraulic conditions, similar to those in the reactor, allowing the cold and hot testings, respectively, of the MID end; - cold testing installation, water supply and oil control panel, feeding the hydraulic drives of the MID's end during the testings; - fixed bridge and mobile carrier for MID's end positioning against testing sections; - installation for functional testing of MID thrusters, before pre-admission and reception tests; - dedicated tools and devices; - raising and transport mechanical devices for handling and positioning the MID's end upon the carrier; - automation panel for controlling the stand equipment and MID's end; - process computer for conducting on-line tests. MID's end testing implies mainly the following operations: - regulation, calibration and functional testing of the MID thrusters carried out independently on a specialised stand; - regulation and calibration of MID's end sub-assemblages; - carrying out the cold and hot pre-admission tests consisting in automatic performing, without operator intervention, of 12 fuel changes, two of which being successive; - performing the cold and hot reception tests, consisting in automatic accomplishment of 4

  7. Reactor Simulator Integration and Testing

    Science.gov (United States)

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

    2013-01-01

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

  8. Two-phase control absorber development program: out-reactor tests and analysis to establish system operating characteristics

    International Nuclear Information System (INIS)

    The two-phase control absorber system uses a continuously flowing mixture of borated water and oxygen to regulate neutron flux in a reactor core. By varying the flow of water through the absorber element, the density and hence the neutron absorption of the mixture is controlled. The test facility was subjected to a comprehensive experimental program at different operating pressures to establish system operating characteristics so that a conceptual design for a power reactor could be developed. It was possible to establish the density operating range of the absorber, determine the desired water-valve flow characteristic required for constant gain in the flux regulating loop, validate the computer code which would be used for the static calculatons required for the conceptual design of an absorber system for a power reactor, and validate a dynamic, hybrid computer simulation of the two-phase control abosrber. (auth)

  9. Baseline risk assessment of the perched water system at the INEL test reactor area

    International Nuclear Information System (INIS)

    A baseline health risk assessment (HRA) was prepared to evaluate potential risks to human health and the environment posed by the Perched Water System (PWS) at the Test Reactor Area (TRA). The PWS has been designated Operable Unit 2-12, one of the 13 operable units identified at TRA. During the period from 1962 to 1990, a total of 6770 million gal of water were discharged from the TRA to unlined surface ponds. Wastewater discharged to the surface ponds at TRA percolates downward through the surficial alluvium and the underlying basalt bedrock. A resulting shallow perched water zone has formed at the interface between the surficial sediments and the underlying basalt. Further downward movement of groundwater is again impeded by a low-permeability layer of silt, clay, and sand encountered at a depth of ∼150 ft. The deep perched water zone occurs on top of this low-permeability interbed. An evaluation was made as to whether potential risks for the PWS could justify implementing a remedial action. The risk evaluation consisted of two parts, the human health evaluation and the ecological evaluation

  10. A coincidencd logic reactor protection system with automatic and permanent testing

    International Nuclear Information System (INIS)

    Within the context of sodium-cooled fast reactors, the CGEE Alsthom enterprise, under consulting for Novatome, has been engaged with the development of an specific protection system for emergency shutdown situations. System described has been conceived on several stages according to the general organization shown on figure I of the Annex, in such a manner that the exigences and recommendations from the safety regulatory authorities are respected and, at the same time, it is assured a significant reactor operation availability without an spureous rod drop. As an example, a selection of principles, rules and criteria currently applied to the development of a system of this kind is reminded. (J.E. de C.)

  11. Simulator for materials testing reactors

    International Nuclear Information System (INIS)

    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)

  12. DEVELOPMENT AND TESTING OF FAULT-DIAGNOSIS ALGORITHMS FOR REACTOR PLANT SYSTEMS

    Energy Technology Data Exchange (ETDEWEB)

    Grelle, Austin L.; Park, Young S.; Vilim, Richard B.

    2016-06-26

    Argonne National Laboratory is further developing fault diagnosis algorithms for use by the operator of a nuclear plant to aid in improved monitoring of overall plant condition and performance. The objective is better management of plant upsets through more timely, informed decisions on control actions with the ultimate goal of improved plant safety, production, and cost management. Integration of these algorithms with visual aids for operators is taking place through a collaboration under the concept of an operator advisory system. This is a software entity whose purpose is to manage and distill the enormous amount of information an operator must process to understand the plant state, particularly in off-normal situations, and how the state trajectory will unfold in time. The fault diagnosis algorithms were exhaustively tested using computer simulations of twenty different faults introduced into the chemical and volume control system (CVCS) of a pressurized water reactor (PWR). The algorithms are unique in that each new application to a facility requires providing only the piping and instrumentation diagram (PID) and no other plant-specific information; a subject-matter expert is not needed to install and maintain each instance of an application. The testing approach followed accepted procedures for verifying and validating software. It was shown that the code satisfies its functional requirement which is to accept sensor information, identify process variable trends based on this sensor information, and then to return an accurate diagnosis based on chains of rules related to these trends. The validation and verification exercise made use of GPASS, a one-dimensional systems code, for simulating CVCS operation. Plant components were failed and the code generated the resulting plant response. Parametric studies with respect to the severity of the fault, the richness of the plant sensor set, and the accuracy of sensors were performed as part of the validation

  13. Test Facility for SMART Reactor Flow Distribution

    International Nuclear Information System (INIS)

    A Reactor Flow Distribution Test Facilities for SMART, named SCOP (SMART Core Flow and Pressure Test Facility), were designed in order to simulate the distributions of (1) core flow and (2) reactor sectional flow resistance and flow rates. SCOP facility was designed based on the linear scaling law in order to preserve the flow characteristics of the prototype system, which are distributions of flow rate and pressure drop. The reduced scale was selected as a 1/5 of prototype length scale. The nominal flow condition was designed to be similar based on the velocity as that of the SMART reactor, which can minimize the flow distortion in the reduced scale of test facility by maintaining high Re number flow. Test facility includes fluid system, control/instrumentation system, data acquisition system, power system, which were designed to meet the requirement for each system. This report describes the details of the scaling and design features for the test facility

  14. A Knowledge-Based System Developer for aerospace applications

    Science.gov (United States)

    Shi, George Z.; Wu, Kewei; Fensky, Connie S.; Lo, Ching F.

    1993-01-01

    A prototype Knowledge-Based System Developer (KBSD) has been developed for aerospace applications by utilizing artificial intelligence technology. The KBSD directly acquires knowledge from domain experts through a graphical interface then builds expert systems from that knowledge. This raises the state of the art of knowledge acquisition/expert system technology to a new level by lessening the need for skilled knowledge engineers. The feasibility, applicability , and efficiency of the proposed concept was established, making a continuation which would develop the prototype to a full-scale general-purpose knowledge-based system developer justifiable. The KBSD has great commercial potential. It will provide a marketable software shell which alleviates the need for knowledge engineers and increase productivity in the workplace. The KBSD will therefore make knowledge-based systems available to a large portion of industry.

  15. Design and evaluation of heat utilization systems for the high temperature engineering test reactor

    International Nuclear Information System (INIS)

    The primary focus of this CRP was to perform detailed investigation of the high temperature industrial processes that are attainable through incorporation of an HTGR, and for their possible demonstration in the HTTR. The HTGR has the capability to achieve a core outlet temperature approaching 1,000 deg. C in a safe and effective manner. These attributes, coupled with the offer by JAERI to utilize the HTTR, resulted in the initiation of this CRP by the IAEA. High Temperature Engineering Test Reactor (HTTR) utilizes a 30 MW(th) HTGR comprised of 30 fuel columns of hexagonal pin-in-pin graphite block type fuel elements. The fuel consists of UO2 TRISO coated particles with an enrichment of ∼ 6% wt. Relative to the demonstration of high temperature heat applications, the HTTR will be capable of producing 10 MW(th) of heat at 950 deg. C. However, the thermal power for these applications has the potential to be increased up to 30 MW(th) in the future, which may be required for demonstration of gas turbine system components. The HTTR reached initial criticality in November 1998. Initial operational plans includes a series of rise to power tests followed by tests to demonstrate the safety and operational characteristics of the HTTR. In addition to completion of the HTTR demonstration tests, it was recommended that the R and D be performed within the HTTR project. JAERI is encouraged to publicize the results of the HTTR tests and 'lessons learned' from their experiences including potential capabilities of the HTGR for heat applications. The next priority application was determined to be the generation of electricity through the use of the gas turbine. Application of the Brayton Cycle utilizing high temperature helium from a modular HTGR was chosen for development because of its projected benefits as an economic and efficient means for the production of electricity. Evaluation of the remaining high temperature heat utilization applications chosen for investigation resulted

  16. Application of automatic inspection system to nondestructive test of heat transfer tubes of primary pressurized water cooler in the high temperature engineering test reactor. Joint research

    International Nuclear Information System (INIS)

    Heat transfer tubes of a primary pressurized water cooled (PPWC) in the high temperature engineering test reactor (HTTR) form the reactor pressure boundary of the primary coolant, therefore are important from the viewpoint of safety. To establish inspection techniques for the heat transfer tubes of the PPWC, an automatic inspection system was developed. The system employs a bobbin coil probe, a rotating probe for eddy current testing (ECT) and a rotating probe for ultrasonic testing (UT). Nondestructive test of a half of the heat transfer tubes of the PPWC was carried out by the automatic inspection system during reactor shutdown period of the HTTR (about 55% in the maximum reactor power in this paper). The nondestructive test results showed that the maximum signal-to-noise ratio was 1.8 in ECT. Pattern and phase of Lissajous wave, which were obtained for the heat transfer tube of the PPWC, were different from those obtained for the artificially defected tube. In UT echo amplitude of the PPWC tubes inspected was lower than 20% of distance-amplitude calibration curve. Thus, it was confirmed that there was no defect in depth, which was more than the detecting standard of the probes, on the outer surface of the heat transfer tubes of the PPWC inspected. (author)

  17. Study of Reliability Life Test System for Aerospace Relay%航天继电器可靠性寿命试验分析系统的研究

    Institute of Scientific and Technical Information of China (English)

    任立; 余琼; 翟国富

    2009-01-01

    The contact resistance's over-normal situation in the conventional relay life test can't roundly proof the failure of the relay. This paper designs a reliability life test system for aerospace relay based on the multi-parameter test. This system includes four parts: the computer, the main control module, the data collect module and the coil driver module. This system applies to different kinds of the reliability life test with different loads. This system can acquire the real-time parameters like contact resistance, the close time and the over-travel time. Based on the analysis of the parameters, the system can synthetically judge the failure of the relay. Furthermore, this system can work out the failure analysis of the relay, gain the failure model for the further research on the failure mechanism.%在现行继电器寿命试验中,仅依靠接触电阻的超标是不能全面反映继电器的失效问题.就此,本文设计了一种多参数实时采集的航天继电器可靠性寿命试验分析系统.该系统由工控机、主控制单元、数据采集及处理单元和线圈驱动单元组成,适用于不同负载、不同测试数量的航天继电器的可靠性寿命试验.它能实时采集继电器在可靠性寿命试验中的接触电阻、吸合时间、超程时间等特性参数,并对这些参数进行分析、计算和处理,来全面地、综合地判断继电器是否失效.此外,还可对继电器进行失效分析,得出失效模型,为进一步研究继电器失效机理提供了依据.

  18. Demonstration test of the holding stability of the self actuated shutdown system in the experimental fast reactor 'JOYO'

    International Nuclear Information System (INIS)

    Self actuated shutdown system (SASS) with a Curie point electromagnet (CPEM) has been developed for use in a large scale fast breeder reactor (FBR) in order to establish the passive shutdown capability against anticipated transient without scram (ATWS) events. The basic characteristics of SASS have already been investigated by various out-of-pile tests for material elements. As the final stage of the development, the stability of SASS needs to be confirmed under the actual reactor-operational environment with high temperature, high neutron flux, and flowing sodium in order to ensure the high plant availability factor. For this purpose, the demonstration test of holding stability using the reduced-scale experimental equipment of SASS was conducted in the 1st and 2nd operational cycles of the experimental fast reactor JOYO MK-III. As a result of this study, the rod-holding stability and the rod-recovering functions of the driving system to re-connect and pull out the separated control rod were fully confirmed. The results also indicate there is no essential problem for the practical use of SASS about its operational trouble involving the unexpected drop during reactor operation. (author)

  19. L-C Measurement Acquisition Method for Aerospace Systems

    Science.gov (United States)

    Woodard, Stanley E.; Taylor, B. Douglas; Shams, Qamar A.; Fox, Robert L.

    2003-01-01

    This paper describes a measurement acquisition method for aerospace systems that eliminates the need for sensors to have physical connection to a power source (i.e., no lead wires) or to data acquisition equipment. Furthermore, the method does not require the sensors to be in proximity to any form of acquisition hardware. Multiple sensors can be interrogated using this method. The sensors consist of a capacitor, C(p), whose capacitance changes with changes to a physical property, p, electrically connected to an inductor, L. The method uses an antenna to broadcast electromagnetic energy that electrically excites one or more inductive-capacitive sensors via Faraday induction. This method facilitates measurements that were not previously possible because there was no practical means of providing power and data acquisition electrical connections to a sensor. Unlike traditional sensors, which measure only a single physical property, the manner in which the sensing element is interrogated simultaneously allows measurement of at least two unrelated physical properties (e.g., displacement rate and fluid level) by using each constituent of the L-C element. The key to using the method for aerospace applications is to increase the distance between the L-C elements and interrogating antenna; develop all key components to be non-obtrusive and to develop sensing elements that can easily be implemented. Techniques that have resulted in increased distance between antenna and sensor will be presented. Fluid-level measurements and pressure measurements using the acquisition method are demonstrated in the paper.

  20. TRIGA reactor main systems

    International Nuclear Information System (INIS)

    This module describes the main systems of low power (<2 MW) and higher power (≥2 MW) TRIGA reactors. The most significant difference between the two is that forced reactor cooling and an emergency core cooling system are generally required for the higher power TRIGA reactors. However, those TRIGA reactors that are designed to be operated above 3 MW also use a TRIGA fuel that is specifically designed for those higher power outputs (3 to 14 MW). Typical values are given for the respective systems although each TRIGA facility will have unique characteristics that may only be determined by the experienced facility operators. Due to the inherent wide scope of these research reactor facilities construction and missions, this training module covers those systems found at most operating TRIGA reactor facilities but may also discuss non-standard equipment that was found to be operationally useful although not necessarily required. (author)

  1. Active wireless temperature sensors for aerospace thermal protection systems

    Science.gov (United States)

    Milos, Frank S.; Karunaratne, K. S. G.

    2003-07-01

    Vehicle system health diagnostics is an area where major improvements have been identified for potential implementation into the design of new reusable launch vehicles in order to reduce life-cycle costs, to increase safety margins, and to improve mission reliability. NASA Ames is leading the effort to advance inspection and health management technologies for thermal protection systems. This paper summarizes a joint effort by NASA Ames and Korteks to develop active "wireless" sensors that can be embedded in the thermal protection system to monitor subsurface temperature histories. These devices are thermocouples integrated with radio-frequency identification circuits to enable non-contact communication of temperature data through aerospace thermal protection materials. Two generations of prototype sensors are discussed. The advanced prototype collects data from three type-k thermocouples attached to a 25-mm square integrated circuit and can communicate through 7 to 10 cm thickness of thermal protection materials.

  2. Test system mount for ultrasonic testing of the external rotative welding seam of cylindrical construction elements especially in reactor plants

    International Nuclear Information System (INIS)

    The ultra sonic test system is used for testing the external socket welds, tube connection welds and socket edges of a pressure vessel. The test system mount consists of a centered circular frame with a revolving bed track for a transport car. At an axial adjustable sled of the transport car a trailing lever is hinged, ot which the test head is seated cardanically. The trailing lever itself can be pivoted by a piston cylinder system. (DG)

  3. Research on the improvement of nuclear safety -Thermal hydraulic tests for reactor safety system-

    International Nuclear Information System (INIS)

    The present research aims at the development of the thermal hydraulic verification test technology for the safety system of the conventional and advanced nuclear power plant and the development of the advanced thermal hydraulic measuring techniques. In this research, test facilities simulating the primary coolant system and safety system are being constructed for the design verification tests of the existing and advanced nuclear power plant. 97 figs, 14 tabs, 65 refs. (Author)

  4. Research on the improvement of nuclear safety -Thermal hydraulic tests for reactor safety system-

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Moon Kee; Park, Choon Kyung; Yang, Sun Kyoo; Chun, Se Yung; Song, Chul Hwa; Jun, Hyung Kil; Jung, Heung Joon; Won, Soon Yun; Cho, Yung Roh; Min, Kyung Hoh; Jung, Jang Hwan; Jang, Suk Kyoo; Kim, Bok Deuk; Kim, Wooi Kyung; Huh, Jin; Kim, Sook Kwan; Moon, Sang Kee; Lee, Sang Il [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-06-01

    The present research aims at the development of the thermal hydraulic verification test technology for the safety system of the conventional and advanced nuclear power plant and the development of the advanced thermal hydraulic measuring techniques. In this research, test facilities simulating the primary coolant system and safety system are being constructed for the design verification tests of the existing and advanced nuclear power plant. 97 figs, 14 tabs, 65 refs. (Author).

  5. The Effect of Online Systems Analysis Training on Aerospace Industry Business Performance: A Qualitative Study

    Science.gov (United States)

    Burk, Erlan

    2012-01-01

    Aerospace companies needed additional research on technology-based training to verify expectations when enhancing human capital through online systems analysis training. The research for online systems analysis training provided aerospace companies a means to verify expectations for systems analysis technology-based training on business…

  6. Application of acoustic emission as a monitoring system during hydrostatic tests of nuclear reactor components

    International Nuclear Information System (INIS)

    The paper presents the state of art for the surveillance of nuclear reactor components by acoustic emission during hydrostatic tests as obtained during several inspections made by KWU and Battelle-Frankfurt/Main. The following four points are relevant: a) Measures designed to suppress background noise, b) adapted pressure increase rate, c) extensive and practically oriented calibration measurements, d) suitable measuring technique. These necessary preconditions are discussed and results on the wave propagation, location accuracy, attenuation of AE-signals due to geometrical configurations (nozzles) and on the correlation between AE-sources and defects as detected by other NDE-methods presented. Two selected examples of AE-tests on reactor components will demonstrate the results which can be obtained at the present time. These investigations have shown till now that: 1) AE is a sensitive NDE-method, able to detect even very small flaws. The AE-sources lie for the most part in areas of seam welds, welded-on attachments, nozzles, closure studs, or other prominent areas (see rolling track). 2) Indications which were found by AE need not necessarily be detectable by other NDE-methods, e.g. ultrasound. 3) Small leaks can be identified in a short time and can be located within certain limits. An essential point however is that leaks with higher noise level must be sealed off. (orig.)

  7. Trial Destruction Test of Spent Cationic Resins in a Molten Salt Oxidation Reactor System

    International Nuclear Information System (INIS)

    The spent ion-exchange resins have to be disposed of and as such, spent ion-exchange resins are a significant fraction of the combustible organic waste from the nuclear industries. One effective treatment option is incinerating the spent resins to yield ash and gas. However, there are difficulties associated with this approach. One of the criticisms of a high-temperature incinerator is that radioactive and hazardous metals are not retained in the incinerator. In addition, incineration of the cationic exchange resins, which have the sulphurcontaining functional groups of sulfonic acid (-SO3-H+), has revealed significant problems associated with sulfur dioxide (SO2), a primary air pollutant, which must be kept under control. There is therefore the developing need for an alternative destruction process. Molten salt oxidation, or MSO for short, is a promising alternative technology. Molten carbonate filled in a MSO reactor is capable of trapping sulfur during organic destruction. In addition, the relatively low-operation temperature of the MSO reactor reduces the volatility of the radionuclides, compared to the other available high temperature technologies for organics destruction, such as inductively coupled plasma, incineration, plasma arc and microwave heating. Trial destruction tests of spent cationic exchange resins doped with radioactive metal surrogates were performed in this study. Two typical operating parameters, temperature and oxidizing air rate, which significantly affect the organics destruction, were tested to establish the optimum ranges for those parameters

  8. Development of lightweight structural health monitoring systems for aerospace applications

    Science.gov (United States)

    Pearson, Matthew

    This thesis investigates the development of structural health monitoring systems (SHM) for aerospace applications. The work focuses on each aspect of a SHM system covering novel transducer technologies and damage detection techniques to detect and locate damage in metallic and composite structures. Secondly the potential of energy harvesting and power arrangement methodologies to provide a stable power source is assessed. Finally culminating in the realisation of smart SHM structures. 1. Transducer Technology A thorough experimental study of low profile, low weight novel transducers not normally used for acoustic emission (AE) and acousto-ultrasonics (AU) damage detection was conducted. This included assessment of their performance when exposed to aircraft environments and feasibility of embedding these transducers in composites specimens in order to realise smart structures. 2. Damage Detection An extensive experimental programme into damage detection utilising AE and AU were conducted in both composites and metallic structures. These techniques were used to assess different damage mechanism within these materials. The same transducers were used for novel AE location techniques coupled with AU similarity assessment to successfully detect and locate damage in a variety of structures. 3. Energy Harvesting and Power Management Experimental investigations and numerical simulations were undertaken to assess the power generation levels of piezoelectric and thermoelectric generators for typical vibration and temperature differentials which exist in the aerospace environment. Furthermore a power management system was assessed to demonstrate the ability of the system to take the varying nature of the input power and condition it to a stable power source for a system. 4. Smart Structures The research conducted is brought together into a smart carbon fibre wing showcasing the novel embedded transducers for AE and AU damage detection and location, as well as vibration energy

  9. Robust Design Optimization of an Aerospace Vehicle Prolusion System

    Directory of Open Access Journals (Sweden)

    Muhammad Aamir Raza

    2011-01-01

    Full Text Available This paper proposes a robust design optimization methodology under design uncertainties of an aerospace vehicle propulsion system. The approach consists of 3D geometric design coupled with complex internal ballistics, hybrid optimization, worst-case deviation, and efficient statistical approach. The uncertainties are propagated through worst-case deviation using first-order orthogonal design matrices. The robustness assessment is measured using the framework of mean-variance and percentile difference approach. A parametric sensitivity analysis is carried out to analyze the effects of design variables variation on performance parameters. A hybrid simulated annealing and pattern search approach is used as an optimizer. The results show the objective function of optimizing the mean performance and minimizing the variation of performance parameters in terms of thrust ratio and total impulse could be achieved while adhering to the system constraints.

  10. Active Wireless Temperature Sensors for Aerospace Thermal Protection Systems

    Science.gov (United States)

    Milos, Frank S.; Karunaratne, K.; Arnold, Jim (Technical Monitor)

    2002-01-01

    Health diagnostics is an area where major improvements have been identified for potential implementation into the design of new reusable launch vehicles in order to reduce life-cycle costs, to increase safety margins, and to improve mission reliability. NASA Ames is leading the effort to advance inspection and health management technologies for thermal protection systems. This paper summarizes a joint project between NASA Ames and Korteks to develop active wireless sensors that can be embedded in the thermal protection system to monitor sub-surface temperature histories. These devices are thermocouples integrated with radio-frequency identification circuitry to enable acquisition and non-contact communication of temperature data through aerospace thermal protection materials. Two generations of prototype sensors are discussed. The advanced prototype collects data from three type-k thermocouples attached to a 2.54-cm square integrated circuit.

  11. Integration of artificial intelligence and numerical optimization techniques for the design of complex aerospace systems

    International Nuclear Information System (INIS)

    A new software system called Engineous combines artificial intelligence and numerical methods for the design and optimization of complex aerospace systems. Engineous combines the advanced computational techniques of genetic algorithms, expert systems, and object-oriented programming with the conventional methods of numerical optimization and simulated annealing to create a design optimization environment that can be applied to computational models in various disciplines. Engineous has produced designs with higher predicted performance gains that current manual design processes - on average a 10-to-1 reduction of turnaround time - and has yielded new insights into product design. It has been applied to the aerodynamic preliminary design of an aircraft engine turbine, concurrent aerodynamic and mechanical preliminary design of an aircraft engine turbine blade and disk, a space superconductor generator, a satellite power converter, and a nuclear-powered satellite reactor and shield. 23 refs

  12. Micro-Jet Test Facility for Aerospace Propulsion Engineering Education

    OpenAIRE

    López Juste, Gregorio; Montañés García, José Luis; Velázquez, A

    2009-01-01

    This paper describes the methodology that has been developed and implemented at the School ofAeronautics (ETSIA) of the Universidad Politecnica de Madrid (UPM) to familiarize aerospaceengineering students with the operation of real complex jet engine systems. This methodology has atwo-pronged approach: students carry out preparatory work by using, first, a gas turbineperformance prediction numerical code; then they validate their assumptions and results on anexperimental test rig. When lookin...

  13. Progress in design, research and development and testing of safety systems for advanced water cooled reactors. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    The meeting covered the following topics: Developments in design of safety-related heat removal components and systems for advanced water cooled reactors; status of test programmes on heat removal components and systems of new designs; range of validity and extrapolation of test results for the qualification of design/licensing computer models and codes for advanced water cooled reactors; future needs and trends in testing of safety systems for advanced water cooled reactors. Tests of heat removal safety systems have been conducted by various groups supporting the design, testing and certification of advanced water cooled reactors. The Technical Committee concluded that the reported test results generally confirm the predicted performance features of the advanced designs. Refs, figs, tabs

  14. The IBR-2 test reactor

    International Nuclear Information System (INIS)

    Major design criteria, specifications and potential fields of application of the IBR-2 pulsed test reactor (now under construction in Dubna, USSR) are described. The pulsed power bursts will be due to fast periodic reactivity changes by a rotating reflector. The frequency of approximately 100 μs pulsed may be 5, 12.5 or 50 Hz. The IBR-2 reactor will be mostly profitable for slow neutron experiments when investigating solids, nuclei or neutrons themselves using spectroscopic methods. Due to the high peak flux of thermal neutrons (1016-1017 n/cm2xs) the reactor will be superior (for the sort of experiments) to the currently operating SM-2 and HFR high flux steady-state test reactors for many times

  15. Internal fluid mechanics research on supercomputers for aerospace propulsion systems

    Science.gov (United States)

    Miller, Brent A.; Anderson, Bernhard H.; Szuch, John R.

    1988-01-01

    The Internal Fluid Mechanics Division of the NASA Lewis Research Center is combining the key elements of computational fluid dynamics, aerothermodynamic experiments, and advanced computational technology to bring internal computational fluid mechanics (ICFM) to a state of practical application for aerospace propulsion systems. The strategies used to achieve this goal are to: (1) pursue an understanding of flow physics, surface heat transfer, and combustion via analysis and fundamental experiments, (2) incorporate improved understanding of these phenomena into verified 3-D CFD codes, and (3) utilize state-of-the-art computational technology to enhance experimental and CFD research. Presented is an overview of the ICFM program in high-speed propulsion, including work in inlets, turbomachinery, and chemical reacting flows. Ongoing efforts to integrate new computer technologies, such as parallel computing and artificial intelligence, into high-speed aeropropulsion research are described.

  16. Internal computational fluid mechanics on supercomputers for aerospace propulsion systems

    Science.gov (United States)

    Andersen, Bernhard H.; Benson, Thomas J.

    1987-01-01

    The accurate calculation of three-dimensional internal flowfields for application towards aerospace propulsion systems requires computational resources available only on supercomputers. A survey is presented of three-dimensional calculations of hypersonic, transonic, and subsonic internal flowfields conducted at the Lewis Research Center. A steady state Parabolized Navier-Stokes (PNS) solution of flow in a Mach 5.0, mixed compression inlet, a Navier-Stokes solution of flow in the vicinity of a terminal shock, and a PNS solution of flow in a diffusing S-bend with vortex generators are presented and discussed. All of these calculations were performed on either the NAS Cray-2 or the Lewis Research Center Cray XMP.

  17. Testing of the Y-12 Plant Criticality Accident Alarm System detectors at the Sandia Pulsed Reactor Facility

    International Nuclear Information System (INIS)

    The Oak Ridge Y-12 Plant operates its Criticality Accident Alarm System (CAAS) according to the guidance of Standard ANSI/ANS-8.3-1986. This standard requires that the detector shall not fail to initiate an alarm when subjected to a radiation field of at least 0.1 Gy/s (10 rad/s). It also requires that the system shall be designed to immediately detect the minimum accident of concern and shall produce an alarm within one half second of activation. Sixty-three new detectors that use plastic scintillators have been obtained to upgrade the current Y-12 Plant CAAS. To ensure that these detectors can support the above criteria, testing was done using the SPR III reactor at the Sandia Pulsed Reactor Facility

  18. Recent palladium membrane reactor development at the tritium systems test assembly

    International Nuclear Information System (INIS)

    The palladium membrane reactor (PMR) is being investigated as a means for recovering hydrogen isotopes (including tritium) from compounds such as water and methane. Previous work with protiated water and methane showed that this device can be used to obtain high hydrogen recovery efficiencies using a single processing pass and with essentially no waste production. With these successful proof-of-principle results completed, recent work has focused on PMR development. This included studies of various geometries and testing with tritium. The results, which are reported here, have led to a better understanding of the PMR and will lead to the ultimate goal of building a production PMR and putting it into practical tritium processing service. 3 refs., 5 figs., 1 tab

  19. Development of Experimental System for Material Compatibility Test for Ultra-long Cycle Fast Reactor (UCFR)

    International Nuclear Information System (INIS)

    Sodium is a candidate for fast reactor coolants that has been believed to have favorable compatibility with structural materials. However, recent studies showed results which need for a more careful attention at this previous belief. For prolonging the service life time of cladding and structural materials in contact with liquid sodium, more detail analysis methods are needed to examine this material compatibility issue with sodium. As a candidate of liquid metals coolants of Ultra-long Cycle Fast Reactor (UCFR), the compatibility of sodium with cladding materials has to be investigated in detail with long term exposure time. It is known that sodium promotes corrosion in two ways. One is corrosion produced by dissolution of alloy elements into sodium and the other is corrosion produced through a chemical reaction with impurities in sodium (especially, dissolved oxygen). The use of the technique of impedance spectroscopy to measure the electrical impedance response of any oxide layers may be a good experimental tool to this monitoring system. The motivation of current study is to investigate the relationship between the electrochemical behaviors of oxide scales on martensitic and austenitic steels and their corrosion rates in liquid sodium

  20. Material test reactor fuel research at the BR2 reactor

    International Nuclear Information System (INIS)

    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

  1. Potential role of the Fast Flux Test Facility and the advanced test reactor in the U.S. tritium production system

    International Nuclear Information System (INIS)

    The Department of Energy is currently engaged in a dual-track strategy to develop an accelerator and a commercial light water reactor (CLWR) as potential sources of tritium supply. New analysis of the production capabilities of the Fast Flux Test Facility (FFTF) at the Hanford Site argues for considering its inclusion in the tritium supply,system. The use of the FFTF (alone or together with the Advanced Test Reactor [ATR] at the Idaho National Engineering Laboratory) as an integral part of,a tritium production system would help (1) ensure supply by 2005, (2) provide additional time to resolve institutional and technical issues associated with the- dual-track strategy, and (3) reduce discounted total life-cycle'costs and near-tenn annual expenditures for accelerator-based systems. The FFRF would also provide a way to get an early start.on dispositioning surplus weapons-usable plutonium as well as provide a source of medical isotopes. Challenges Associated With the Dual-Track Strategy The Department's purchase of either a commercial reactor or reactor irradiation services faces challenging institutional issues associated with converting civilian reactors to defense uses. In addition, while the technical capabilities of the individual components of the accelerator have been proven, the entire system needs to be demonstrated and scaled upward to ensure that the components work together 1548 as a complete production system. These challenges create uncertainty over the ability of the du2a-track strategy to provide an assured tritium supply source by 2005. Because the earliest the accelerator could come on line is 2007, it would have to operate at maximum capacity for the first few years to regenerate the reserves lost through radioactive decay after 2005

  2. A smart pattern recognition system for the automatic identification of aerospace acoustic sources

    Science.gov (United States)

    Cabell, R. H.; Fuller, C. R.

    1989-01-01

    An intelligent air-noise recognition system is described that uses pattern recognition techniques to distinguish noise signatures of five different types of acoustic sources, including jet planes, propeller planes, a helicopter, train, and wind turbine. Information for classification is calculated using the power spectral density and autocorrelation taken from the output of a single microphone. Using this system, as many as 90 percent of test recordings were correctly identified, indicating that the linear discriminant functions developed can be used for aerospace source identification.

  3. Initial test results of the Omron face cue entry system at the University of Missouri-Rolla Reactor

    International Nuclear Information System (INIS)

    The University of Missouri-Rolla Reactor facility is testing, in collaboration with Omron Transaction Systems, Inc., the Omron Face Cue facial recognition system for access control to its restricted area. The installation of this system is the first of its kind at a security-relevant facility in the U.S. and within the research reactor community. The Face Cue is an on-demand device based on facial recognition and storage technology. The image processing methodology is as follows: (1) facial position detection, (2) background elimination, (3) facial features discrimination via application of a wavelet transform. The extracted facial feature values are compared to the data archived in its database and access is provided upon meeting the authorization criteria. The current test phase consisted of assessing the functionality of the Face Cue during daily use and in terms of its robustness (flexibility) as a function of the following physical parameters: (1) subject's distance away from the Face Cue, (2) ambient lighting conditions, (3) subject's facial orientation, (4) subject's facial expression and (5) peripheral facial features/modifications. The system has operated at nearly 100% reliability during several test intervals with approximately 7,000 entry attempts to date. (author)

  4. 76 FR 41041 - Special Conditions: Gulfstream Aerospace LP (GALP) Model G250 Airplane, Interaction of Systems...

    Science.gov (United States)

    2011-07-13

    ... interaction of control systems and structures. The usual deterministic approach to defining the loads envelope... Administration 14 CFR Part 25 Special Conditions: Gulfstream Aerospace LP (GALP) Model G250 Airplane, Interaction... special conditions are issued for the Gulfstream Aerospace LP (GALP) Model G250 airplane. This...

  5. Effectiveness and resolution of tests for evaluating the performance of cutting fluids in machining aerospace alloys

    DEFF Research Database (Denmark)

    De Chiffre, Leonardo; Axinte, Dragos A.

    2008-01-01

    The paper discusses effectiveness and resolution of five cutting tests (turning, milling, drilling, tapping, VIPER grinding) and their quality output measures used in a multi-task procedure for evaluating the performance of cutting fluids when machining aerospace materials. The evaluation takes...

  6. Advanced test reactor. Testing capabilities and plans

    International Nuclear Information System (INIS)

    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. For future research, some ATR modifications and enhancements are currently planned. In 2007 the US Department of Energy designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR for material testing research by a broader user community. This paper provides more details on some of the ATR capabilities, key design features, experiments, and plants for the NSUF. (author)

  7. Engineering derivatives from biological systems for advanced aerospace applications

    Science.gov (United States)

    Winfield, Daniel L.; Hering, Dean H.; Cole, David

    1991-01-01

    The present study consisted of a literature survey, a survey of researchers, and a workshop on bionics. These tasks produced an extensive annotated bibliography of bionics research (282 citations), a directory of bionics researchers, and a workshop report on specific bionics research topics applicable to space technology. These deliverables are included as Appendix A, Appendix B, and Section 5.0, respectively. To provide organization to this highly interdisciplinary field and to serve as a guide for interested researchers, we have also prepared a taxonomy or classification of the various subelements of natural engineering systems. Finally, we have synthesized the results of the various components of this study into a discussion of the most promising opportunities for accelerated research, seeking solutions which apply engineering principles from natural systems to advanced aerospace problems. A discussion of opportunities within the areas of materials, structures, sensors, information processing, robotics, autonomous systems, life support systems, and aeronautics is given. Following the conclusions are six discipline summaries that highlight the potential benefits of research in these areas for NASA's space technology programs.

  8. SP-100 Program: space reactor system and subsystem investigations

    International Nuclear Information System (INIS)

    For a space reactor power system, a comprehensive safety program will be required to assure that no undue risk is present. This report summarizes the nuclear safety review/approval process that will be required for a space reactor system. The documentation requirements are presented along with a summary of the required contents of key documents. Finally, the aerospace safety program conducted for the SNAP-10A reactor system is summarized. The results of this program are presented to show the type of program that can be expected and to provide information that could be usable in future programs

  9. PITR: Princeton Ignition Test Reactor

    International Nuclear Information System (INIS)

    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

  10. PITR: Princeton Ignition Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    1978-12-01

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

  11. Emergency reactor scram system

    International Nuclear Information System (INIS)

    The present invention provides an emergency reactor scram system capable of shut down a reactor safely upon occurrence of pump trip by improving a passive scram performance for an FBR-type reactor. Namely, a driving motor and an electric generator are connected to a main pump of a primary system. An AC/DC convertor is connected to the electric generator. A shielding plug is disposed to the upper end opening of a reactor container, a control rod drive mechanism is erected on the shielding plug, and an extension pipe is attached to scram magnets of the control rod drive mechanism. The extension pipe is connected to a control rod. The rotation of the shaft of the pump is used as a direct rotator to provide an integrated-type electric generator. The electric generator is electrically connected with the power source of a scram magnet of the emergency scram system. Accordingly, the control rod of the emergency scram system is automatically and rapidly inserted to the reactor core using the power source of the electric generator upon trip of the main pump thereby enabling to scram the reactor safely. (I.S.)

  12. Development of in-service inspection system for core support graphite structures in the high temperature engineering test reactor (HTTR)

    Energy Technology Data Exchange (ETDEWEB)

    Sumita, Junya; Hanawa, Satoshi; Kikuchi, Takayuki; Ishihara, Masahiro [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2003-03-01

    Visual inspection of core support graphite structures using TV camera as in-service inspection and measurement of material characteristics using surveillance test specimens are planned in the High Temperature Engineering Test Reactor (HTTR) to confirm structural integrity of the core support graphite structures. For the visual inspection, in-service inspection system developed from September 1996 to June 1998, and pre-service inspection using the system was carried out. As the result of the pre-service inspection, it was validated that high quality of visual inspection with TV camera can be carried out, and also structural integrity of the core support graphite structures at the initial stage of the HTTR operation was confirmed. (author)

  13. Reactor parameter simulation system

    International Nuclear Information System (INIS)

    A reactor parameter simulation system (RPSS) has been built with the capability of analyzing any reactor signals, decomposing those signals into their deterministic and stochastic components, then reconstructing new, simulated signals that possess the same statistical and correlation structure as the original plant variables. Important uses of the RPSS are for integration with reactor simulation software to provide tools for plant control strategy development, and for safety-study investigations of scenarios that can arise involving signal faults generated from degraded sensors. A third use of the RPSS is for frequency-domain filtering of reactor process variables contaminated with serially correlated noise, which is important for our ongoing development of expert systems for sensor-operability surveillance. 5 refs., 4 figs., 3 tabs

  14. An Approach to the Flammability Testing of Aerospace Materials

    Science.gov (United States)

    Hirsch, David B.

    2012-01-01

    Presentation reviews: (1) Current approach to evaluation of spacecraft materials flammability (2) The need for and the approach to alternative routes (3) Examples of applications of the approach recommended a) Crew Module splash down b) Crew Module depressurization c) Applicability of NASA's flammability test data to other sample configurations d) Applicability of NASA's ground flammability test data to spacecraft environments

  15. Energy Harvesting for Aerospace Structural Health Monitoring Systems

    Science.gov (United States)

    Pearson, M. R.; Eaton, M. J.; Pullin, R.; Featherston, C. A.; Holford, K. M.

    2012-08-01

    Recent research into damage detection methodologies, embedded sensors, wireless data transmission and energy harvesting in aerospace environments has meant that autonomous structural health monitoring (SHM) systems are becoming a real possibility. The most promising system would utilise wireless sensor nodes that are able to make decisions on damage and communicate this wirelessly to a central base station. Although such a system shows great potential and both passive and active monitoring techniques exist for detecting damage in structures, powering such wireless sensors nodes poses a problem. Two such energy sources that could be harvested in abundance on an aircraft are vibration and thermal gradients. Piezoelectric transducers mounted to the surface of a structure can be utilised to generate power from a dynamic strain whilst thermoelectric generators (TEG) can be used to generate power from thermal gradients. This paper reports on the viability of these two energy sources for powering a wireless SHM system from vibrations ranging from 20 to 400Hz and thermal gradients up to 50°C. Investigations showed that using a single vibrational energy harvester raw power levels of up to 1mW could be generated. Further numerical modelling demonstrated that by optimising the position and orientation of the vibrational harvester greater levels of power could be achieved. However using commercial TEGs average power levels over a flight period between 5 to 30mW could be generated. Both of these energy harvesting techniques show a great potential in powering current wireless SHM systems where depending on the complexity the power requirements range from 1 to 180mW.

  16. Energy Harvesting for Aerospace Structural Health Monitoring Systems

    International Nuclear Information System (INIS)

    Recent research into damage detection methodologies, embedded sensors, wireless data transmission and energy harvesting in aerospace environments has meant that autonomous structural health monitoring (SHM) systems are becoming a real possibility. The most promising system would utilise wireless sensor nodes that are able to make decisions on damage and communicate this wirelessly to a central base station. Although such a system shows great potential and both passive and active monitoring techniques exist for detecting damage in structures, powering such wireless sensors nodes poses a problem. Two such energy sources that could be harvested in abundance on an aircraft are vibration and thermal gradients. Piezoelectric transducers mounted to the surface of a structure can be utilised to generate power from a dynamic strain whilst thermoelectric generators (TEG) can be used to generate power from thermal gradients. This paper reports on the viability of these two energy sources for powering a wireless SHM system from vibrations ranging from 20 to 400Hz and thermal gradients up to 50°C. Investigations showed that using a single vibrational energy harvester raw power levels of up to 1mW could be generated. Further numerical modelling demonstrated that by optimising the position and orientation of the vibrational harvester greater levels of power could be achieved. However using commercial TEGs average power levels over a flight period between 5 to 30mW could be generated. Both of these energy harvesting techniques show a great potential in powering current wireless SHM systems where depending on the complexity the power requirements range from 1 to 180mW.

  17. Standardization of shape memory alloy test methods toward certification of aerospace applications

    Science.gov (United States)

    Hartl, D. J.; Mabe, J. H.; Benafan, O.; Coda, A.; Conduit, B.; Padan, R.; Van Doren, B.

    2015-08-01

    The response of shape memory alloy (SMA) components employed as actuators has enabled a number of adaptable aero-structural solutions. However, there are currently no industry or government-accepted standardized test methods for SMA materials when used as actuators and their transition to commercialization and production has been hindered. This brief fast track communication introduces to the community a recently initiated collaborative and pre-competitive SMA specification and standardization effort that is expected to deliver the first ever regulatory agency-accepted material specification and test standards for SMA as employed as actuators for commercial and military aviation applications. In the first phase of this effort, described herein, the team is working to review past efforts and deliver a set of agreed-upon properties to be included in future material certification specifications as well as the associated experiments needed to obtain them in a consistent manner. Essential for the success of this project is the participation and input from a number of organizations and individuals, including engineers and designers working in materials and processing development, application design, SMA component fabrication, and testing at the material, component, and system level. Going forward, strong consensus among this diverse body of participants and the SMA research community at large is needed to advance standardization concepts for universal adoption by the greater aerospace community and especially regulatory bodies. It is expected that the development and release of public standards will be done in collaboration with an established standards development organization.

  18. Reactor protection system

    International Nuclear Information System (INIS)

    The report describes the reactor protection system (RPS-II) designed for use on Babcock and Wilcox 145-, later 177-, and 205-fuel assembly pressurized water reactors. In this system, relays in the trip logic have been replaced by solid state devices. A calculating module for the low DNBR, pump status, and offset trip functions has replaced the overpower trip (based on flow and imbalance), the power/RC pump trip, and the variable low-pressure trip. Included is a description of the changes from the present Oconee-type reactor protection system (RPS-I), a functional and hardware description of the calculating module, a description of the software programmed in the calculating module, and a discussion of the qualification program conducted to ensure that the degree of protection provided by RPS-II is not less than that provided by previously licensed systems supplied by B and W

  19. Operating experience of Fast Breeder Test Reactor

    International Nuclear Information System (INIS)

    made to drain the sodium quickly from the loop to minimize the quantity of sodium leak in case of leak from sodium pipe. Diverse leak detection systems are provided both in primary and secondary circuits to detect sodium leak at the incipient stage itself to take corrective actions. Hydrogen detectors are provided both in sodium and cover gas to detect minute leak of water/ steam into sodium in steam generator (SG) and bring the system to safe configuration. Design provisions are made to protect SG from over pressurization in case of large leaks. This paper details 19 years safety related operating experience of fuel, sodium systems and other systems of the reactor and various safety related tests carried out. (author)

  20. New system for production of reactor medical radionuclides tested with Lu-176

    Czech Academy of Sciences Publication Activity Database

    Seifert, Daniel; Kropáček, Martin; Tomeš, Marek; Kučera, Jan; Lebeda, Ondřej

    2015-01-01

    Roč. 42, S (2015), s. 857-857. ISSN 1619-7070. [28th Annual congress of the European-Association-of-Nuclear-Medicine (EANM). 10.10.2015-14.10.2015, Hamburg] R&D Projects: GA MŠk(CZ) LM2011019 Institutional support: RVO:61389005 Keywords : Lu-177 * radionuclides * reactor Subject RIV: BG - Nuclear, Atomic and Molecular Physics, Colliders

  1. Development and validation of a real-time synthetic aperture focusing technique for ultrasonic testing (SAFT-UT) system for in-service inspection of light water reactors

    International Nuclear Information System (INIS)

    The objectives of the program is to: 1) design, fabricate, and evaluate a real-time flaw detection and characterization system based on synthetic aperture focusing technique for ultrasonic testing (SAFT-UT) for inservice inspection (ISI) of all required light water reactors (LWR) components; 2) establish calibration and field test procedures; 3) demonstrate and validate the system through actual field reactor inspections; and 4) generate an engineering data base to support code acceptance of the real-time SAFT-UT technique. The program scope is defined by the following: 1) conduct laboratory tests to provide engineering data for defining SAFT-UT system performance; 2) complete the development of a special processor to make SAFT a real-time process for ISI application; and 3) fabricate and field test a fieldable real-time SAFT-UT system on nuclear reactor piping, nozzles and pressure vessels

  2. Moon base reactor system

    Science.gov (United States)

    Chavez, H.; Flores, J.; Nguyen, M.; Carsen, K.

    1989-01-01

    The objective of our reactor design is to supply a lunar-based research facility with 20 MW(e). The fundamental layout of this lunar-based system includes the reactor, power conversion devices, and a radiator. The additional aim of this reactor is a longevity of 12 to 15 years. The reactor is a liquid metal fast breeder that has a breeding ratio very close to 1.0. The geometry of the core is cylindrical. The metallic fuel rods are of beryllium oxide enriched with varying degrees of uranium, with a beryllium core reflector. The liquid metal coolant chosen was natural lithium. After the liquid metal coolant leaves the reactor, it goes directly into the power conversion devices. The power conversion devices are Stirling engines. The heated coolant acts as a hot reservoir to the device. It then enters the radiator to be cooled and reenters the Stirling engine acting as a cold reservoir. The engines' operating fluid is helium, a highly conductive gas. These Stirling engines are hermetically sealed. Although natural lithium produces a lower breeding ratio, it does have a larger temperature range than sodium. It is also corrosive to steel. This is why the container material must be carefully chosen. One option is to use an expensive alloy of cerbium and zirconium. The radiator must be made of a highly conductive material whose melting point temperature is not exceeded in the reactor and whose structural strength can withstand meteor showers.

  3. Standard Test Method for Intensity of Scratches on Aerospace Transparent Plastics

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2009-01-01

    1.1 This test method covers the visual inspection of shallow or superficial scratches on the surface of aerospace transparent plastic materials. 1.2 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

  4. The ageless aerospace vehicle: a complex multi-agent structural health management system

    International Nuclear Information System (INIS)

    Full text: Structural health monitoring and management of complex, safety-critical structures such as aerospace vehicles will ultimately require the development of intelligent systems to process the data from large numbers of sensors, to evaluate and diagnose detected damage, to form a prognosis for the damaged structure, and to make decisions regarding remediation or repair of the damage. A complex multi-agent systems approach to the development of such intelligent systems is being investigated, in order to satisfy the requirements of robustness and scalability. This paper reports the current state of development of a laboratory-scale test-bed built to facilitate the development and demonstration of the sensors, sensing strategies and algorithms that will produce the required functionality. This work involves a wide range of physics-related issues in materials science, sensing and complex systems science. Copyright (2005) Australian Institute of Physics

  5. Safety analysis of reactor's cooling system

    International Nuclear Information System (INIS)

    Results of the analysis of reactor's RBMK-1500 coolant system during normal operation mode, hydrodynamic testing and in the case of earthquake are presented. Analysis was performed using RELAP5 code. Calculations showed the most vulnerable place in the reactor's coolant system. It was found that in the case of earthquake the horizontal support system of drum separator could be damaged

  6. Present status of Japan materials testing reactor

    International Nuclear Information System (INIS)

    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)

  7. Sextant: an expert system for transient analysis of nuclear reactors and integral test facilities

    International Nuclear Information System (INIS)

    Expert systems provide a new way of dealing with the computer-aided management of nuclear plants by combining several knowledge bases and reasoning modes together with a set of numerical models for real-time analysis of transients. New development tools are required together with metaknowledge bases handling temporal hypothetical reasoning and planning. They have to be efficient and robust because during a transient, neither measurements nor models, nor scenarios are hold as absolute references. SEXTANT is a general purpose physical analyzer intended to provide a pattern and avoid duplication of general tools and knowledge bases for similar applications. It combines several knowledge bases concerning measurements, models and qualitative behavior of PWR with a mechanism of conjecture-refutation and a set of simplified models matching the current physical state. A prototype is under assessment by dealing with integral test facility transients. For its development, SEXTANT requires a powerful shell. SPIRAL is such a toolkit, oriented towards online analysis of complex processes and already used in several applications

  8. Meeting the Challenges of Exploration Systems: Health Management Technologies for Aerospace Systems With Emphasis on Propulsion

    Science.gov (United States)

    Melcher, Kevin J.; Sowers, T. Shane; Maul, William A.

    2005-01-01

    The constraints of future Exploration Missions will require unique Integrated System Health Management (ISHM) capabilities throughout the mission. An ambitious launch schedule, human-rating requirements, long quiescent periods, limited human access for repair or replacement, and long communication delays all require an ISHM system that can span distinct yet interdependent vehicle subsystems, anticipate failure states, provide autonomous remediation, and support the Exploration Mission from beginning to end. NASA Glenn Research Center has developed and applied health management system technologies to aerospace propulsion systems for almost two decades. Lessons learned from past activities help define the approach to proper ISHM development: sensor selection- identifies sensor sets required for accurate health assessment; data qualification and validation-ensures the integrity of measurement data from sensor to data system; fault detection and isolation-uses measurements in a component/subsystem context to detect faults and identify their point of origin; information fusion and diagnostic decision criteria-aligns data from similar and disparate sources in time and use that data to perform higher-level system diagnosis; and verification and validation-uses data, real or simulated, to provide variable exposure to the diagnostic system for faults that may only manifest themselves in actual implementation, as well as faults that are detectable via hardware testing. This presentation describes a framework for developing health management systems and highlights the health management research activities performed by the Controls and Dynamics Branch at the NASA Glenn Research Center. It illustrates how those activities contribute to the development of solutions for Integrated System Health Management.

  9. Proposal of world network on material testing reactors

    International Nuclear Information System (INIS)

    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. Information retrieval system on reactor test methods and role of methodic information in planning of research in reactor material science field

    International Nuclear Information System (INIS)

    The results of processing of methodic information which is systematized in form of an information retrieval system adapted for needs of researchers in material science field are represented. It permits to optimize planning of development and perfectioning the experimental base for reactor material science. (J.P.)

  11. Reactor system safety assurance

    International Nuclear Information System (INIS)

    The philosophy of reactor safety is that design should follow established and conservative engineering practices, there should be safety margins in all modes of plant operation, special systems should be provided for accidents, and safety systems should have redundant components. This philosophy provides ''defense in depth.'' Additionally, the safety of nuclear power plants relies on ''safety systems'' to assure acceptable response to design basis events. Operating experience has shown the need to study plant response to more frequent upset conditions and to account for the influence of operators and non-safety systems on overall performance. Defense in depth is being supplemented by risk and reliability assessment

  12. Competitive assessment of aerospace systems using system dynamics

    Science.gov (United States)

    Pfaender, Jens Holger

    Aircraft design has recently experienced a trend away from performance centric design towards a more balanced approach with increased emphasis on engineering an economically successful system. This approach focuses on bringing forward a comprehensive economic and life-cycle cost analysis. Since the success of any system also depends on many external factors outside of the control of the designer, this traditionally has been modeled as noise affecting the uncertainty of the design. However, this approach is currently lacking a strategic treatment of necessary early decisions affecting the probability of success of a given concept in a dynamic environment. This suggests that the introduction of a dynamic method into a life-cycle cost analysis should allow the analysis of the future attractiveness of such a concept in the presence of uncertainty. One way of addressing this is through the use of a competitive market model. However, existing market models do not focus on the dynamics of the market. Instead, they focus on modeling and predicting market share through logit regression models. The resulting models exhibit relatively poor predictive capabilities. The method proposed here focuses on a top-down approach that integrates a competitive model based on work in the field of system dynamics into the aircraft design process. Demonstrating such integration is one of the primary contributions of this work, which previously has not been demonstrated. This integration is achieved through the use of surrogate models, in this case neural networks. This enabled not only the practical integration of analysis techniques, but also reduced the computational requirements so that interactive exploration as envisioned was actually possible. The example demonstration of this integration is built on the competition in the 250 seat large commercial aircraft market exemplified by the Boeing 767-400ER and the Airbus A330-200. Both aircraft models were calibrated to existing performance

  13. Development of Modern Safe Systems of Work at the Imperial College Reactor Centre and Their Application to Neutron Detector Testing and Nuclear Training Courses

    International Nuclear Information System (INIS)

    The CONSORT design reactor is owned by, and licensed to, Imperial College of Science, Technology and Medicine, and has been in continuous safe operation since 1965. Today, it is the only civil research reactor left in the United Kingdom. CONSORT is designated a low power research reactor and is rated at 100 kW. The paper concentrates on the issues that have been addressed in ensuring that worker doses are maintained as low as reasonably practicable (ALARP), and describes how robust systems have been implemented to ensure that all assessments are in line with this principle. The principles of risk assessment are applied to all operations within the reactor centre and the framework leading to the establishment of the safe system of work will be outlined and discussed. Two case studies are described in detail as examples, showing the importance to the industry of ensuring that a system is in place to allow the work to continue and summarizes the experience gained in the past few years at Imperial College Reactor Centre. These experiences will provide useful information that managers of similar facilities may wish to consider in formulating their own arrangements. The CONSORT reactor provides an open beam tube facility for the calibration and periodic testing of neutron flux detectors (primarily fission chambers and ion chambers). The first case study outlines the detector testing programme that takes place on the CONSORT reactor using this facility. The second case study relates to the use of this integrated approach to safety management in the teaching environment. With ''safe systems of work'' in place, it is now possible to reinstate the popular student experiment of directly viewing the Cerenkov radiation emitted from the critical CONSORT reactor core. This particular activity had been discontinued for some years on ALARP grounds. (author)

  14. Stack and area tritium monitoring systems for the tokamak fusion test reactor (TFTR)

    Energy Technology Data Exchange (ETDEWEB)

    Pearson, G.G.; Meixler, L.D.; Sissingh, R.A.P.

    1991-01-01

    TFTR Tritium Stack and Area Monitoring Systems have been developed to provide the required level of reliability in a cost effective manner consistent with the mission of the Tritium Handling System on TFTR. Personnel protection, environmental responsibility, and tritium containing system integrity have been the considerations in system design. During the Deuterium-Tritium (D-T) experiments on TFTR, tritium will be used for the first time as one of the fuels. All of the tritium bearing systems will have potentially releasable inventories. Although the tritium inventories (total on-site inventory is limited to 50,000 Ci) are low, the consequences of a release may still be significant. For that reason, a thorough TFTR tritium monitoring program has been initiated. 4 refs., 2 figs.

  15. Reactor design of the SP-100 nuclear assembly test

    International Nuclear Information System (INIS)

    The Nuclear Assembly Test is currently being designed to demonstrate the performance characteristics of a 100-kWe version of the power source for the SP-100 Generic Flight System. Particular emphasis will be placed upon the operation of the prototypical ground test reactor under conditions of high-working temperatures and long life. The key features of the reactor include a small, compact core with component materials consisting of refractory metals and alloys. Because of the unique features of the SP-100 system, extensive use is made of Monte Carlo methods in the design and analysis of the reactor configuration. In addition, detailed testing of the reactor design has been carried out in the Zero Power Physics Reactor facility to provide calibration factors for the principal performance parameters. The key features of the test reactor design are described in this paper

  16. Simulation and tests to individual and coupled models of the reactor vessel simulator and the recirculation system for the SUN-RAH

    International Nuclear Information System (INIS)

    The present project, is continuation of the project presented in the congress SNM-2003. In this new phase of the project, they were carried out adaptive changes to the modeling and implementation of the module of the full superior of the core of the reactor, they were carried out those modeling of the generation of heat as well as of the energy transfer in the one fuel. These models present the main characteristics of the vessel of the one reactor and of the recirculation system, defined by the main phenomena that they intervene in the physical processes, in the previous version the simulation in real time it required of an extremely quick computer and without executing collateral processes. The tests are presented carried out to the different models belonging to the Simulator of the Reactor Vessel and the Recirculation system for the SUN-RAH (University Simulator of Nucleo electric with Boiling Water Reactor), as well as the results hurtled by this tests. In each section the executions of the tests and the corresponding analyses of results are shown for each pattern. Besides the above mentioned, the advantages presented by the Simulator of the reactor vessel and the recirculation system are pointed. (Author)

  17. Standard Test Method for Intensity of Scratches on Aerospace Glass Enclosures

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2009-01-01

    1.1 This test method covers the visual inspection of scratches on the glass surface of aerospace transparent enclosures. 1.2 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard. 1.3 This standard may involve hazardous materials, operations, and equipment. This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

  18. Unusual occurrences in fast breeder test reactor

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  20. TREAT Reactor Control and Protection System

    International Nuclear Information System (INIS)

    The main control algorithm of the Transient Reactor Test Facility (TREAT) Automatic Reactor Control System (ARCS) resides in Read Only Memory (ROM) and only experiment specific parameters are input via keyboard entry. Prior to executing an experiment, the software and hardware of the control computer is tested by a closed loop real-time simulation. Two computers with parallel processing are used for the reactor simulation and another computer is used for simulation of the control rod system. A monitor computer, used as a redundant diverse reactor protection channel, uses more conservative setpoints and reduces challenges to the Reactor Trip System (RTS). The RTS consists of triplicated hardwired channels with one out of three logic. The RTS is automatically tested by a digital Dedicated Microprocessor Tester (DMT) prior to the execution of an experiment. 6 refs., 5 figs., 1 tab

  1. An integrated analytic tool and knowledge-based system approach to aerospace electric power system control

    Science.gov (United States)

    Owens, William R.; Henderson, Eric; Gandikota, Kapal

    1986-10-01

    Future aerospace electric power systems require new control methods because of increasing power system complexity, demands for power system management, greater system size and heightened reliability requirements. To meet these requirements, a combination of electric power system analytic tools and knowledge-based systems is proposed. The continual improvement in microelectronic performance has made it possible to envision the application of sophisticated electric power system analysis tools to aerospace vehicles. These tools have been successfully used in the measurement and control of large terrestrial electric power systems. Among these tools is state estimation which has three main benefits. The estimator builds a reliable database for the system structure and states. Security assessment and contingency evaluation also require a state estimator. Finally, the estimator will, combined with modern control theory, improve power system control and stability. Bad data detection as an adjunct to state estimation identifies defective sensors and communications channels. Validated data from the analytic tools is supplied to a number of knowledge-based systems. These systems will be responsible for the control, protection, and optimization of the electric power system.

  2. NAVSTAR Global Positioning System. (Latest citations from the Aerospace Database)

    Science.gov (United States)

    1998-01-01

    The bibliography contains citations concerning the global system of navigation satellites developed to provide immediate and accurate worldwide three-dimensional positioning by air, land, and sea vehicles equipped with appropriate receiving equipment. Technological forecasting, reliability, performance tests, and evaluations are discussed. Developments and applications of the NAVSTAR system are included.(Contains 50-250 citations and includes a subject term index and title list.)

  3. Startup operational tests of fast reactors

    International Nuclear Information System (INIS)

    This paper is mainly concerned with the experiences of the two main phases of startup operational tests of fast reactors: (1) The general tests and Sodium filling before core loading. (2) The core loading,approach to criticality and power build up operational tests, taking for example a large and middle demonstrating integrated-type fast reactor. (author)

  4. Reactor safety systems

    International Nuclear Information System (INIS)

    The spectrum of possible accidents may become characterized by the 'maximum credible accident', which will/will not happen. Similary, the performance of safety systems in a multitude of situations is sometimes simplified to 'the emergency system will/will not work' or even 'reactors are/ are not safe'. In assessing safety, one must avoid this fallacy of reducing a complicated situation to the simple black-and-white picture of yes/no. Similarly, there is a natural tendency continually to improve the safety of a system to assure that it is 'safe enough'. Any system can be made safer and there is usually some additional cost. It is important to balance the increased safety against the increased costs. (orig.)

  5. Licensing systems for reactor operators in Japan

    International Nuclear Information System (INIS)

    Any person proposing to set up a reactor, whether for commercial power generation or for research, is obliged by law to adopt a ''chief tech-ician'' for the reactor from among those who have passed national examinations for such technicians, for appointment to the supervisory position to ensure safety in reactor operation (System for Reactor Chief Technicians). Candidates for chief technicians first take a written test, followed by an oral test. The written test is given to find out whether or not they have the special knowledge necessary to discharge their possible duties as chief technicians. The oral test is given to see if they have the practical knowledge necessary for reactor operation. After the TMI accident, the Ministry of International Trade and Industry carried out a special inspection of nuclear power plants, leading to the conclusion that it was necessary to make constant efforts to upgrade the skills of the operators. Thus the MITI enforced a new system (System for Reactor Responsible Operators) in July 1980, with the aim of fostering and securing highly qualified responsible operators. The Thermal and Nuclear Power Engineering Society was appointed as the licensing organization in January 1981. It was made obligatory for owners of commercial power reactors to assign persons licensed by the Society as responsible operators of nuclear power reactors. (Nogami, K.)

  6. 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

  7. Testing stand for cosmic gas-cooling fast reactor's sample

    International Nuclear Information System (INIS)

    For carrying out of technical decision and nuclear, radiation and technological safety of gas-cooling space nuclear power plants is elaborating gas-cooling fast reactor's testing stand. In the base of its draft is taken conception of the reactor with filling up type reactor core on the base of ball fuel elements and radial coolant flowing. On the testing stand would suggested carrying out testing for study neutron and physical parameters of gas-cooling reactor, its behaviour under accident simulation. In the reactor core will suggest use carbon nitrides fuel elements with tungsten cover, provides under nominal regime relatively low fission products yield to first contour of device. Construction of fuel element was carrying out on reactor and non reactor testing and its calculated on working resource about 3000 hours. Constructive materials of reactor core have lower melting temperature, that provides organized in good time remove fuel element to containers placed under reactor in case connected with hypothetical accident. In the construction of reactor for seen tree-contours system of heat transfer and its provides multistage system of barriers against fission products yield to environment. tabs.1

  8. Reactor system on barge

    International Nuclear Information System (INIS)

    Floating electrical power plants or power plant barges add new dimensions to utility planners and agencies in the world. Intrinsically safe and economical reactors (ISER) employ steel reactor pressure vessels, which significantly reduce the weight as compared with PIUS, and provide siting versatility including barge-mounted plants. In this paper, the outline of power plant barges and barge-mounted ISERs is described. Besides their mobility, power plant barges have the salient advantages such as short delivery time and better quality control due to the outfitting in shipyards. These power plant barges may be temporarily moored or permanently grounded in shallow water at the centers of industrial complexes or the suitable areas adjacent to them, and satisfy the increasing needs for electric power. A cost-effective and technically perfect barge positioning system should be designed to meet the specific requirement for the location and its condition. Offshore siting away from coast may be applicable only to large plants of 1,000 MWe or more, and inshore siting and coastal or river siting are considered for an ISER-200 barge-mounted plant. The system of a barge-mounted ISER plant is discussed in the case of a floating type and the type on a seismic base isolator. (Kako, I.)

  9. Advanced information processing system - Status report. [for fault tolerant and damage tolerant data processing for aerospace vehicles

    Science.gov (United States)

    Brock, L. D.; Lala, J.

    1986-01-01

    The Advanced Information Processing System (AIPS) is designed to provide a fault tolerant and damage tolerant data processing architecture for a broad range of aerospace vehicles. The AIPS architecture also has attributes to enhance system effectiveness such as graceful degradation, growth and change tolerance, integrability, etc. Two key building blocks being developed by the AIPS program are a fault and damage tolerant processor and communication network. A proof-of-concept system is now being built and will be tested to demonstrate the validity and performance of the AIPS concepts.

  10. Fuel irradiation test plan at the Japan materials testing reactor

    International Nuclear Information System (INIS)

    Development of high performance fuels, which enables burnup extension and high duty uses of light water reactors (LWRs) by means of power up rates and flexible operating cycles, is one of key technical issues for extending the uses for longer periods. Introduction of new design fuel rods with new cladding alloys and wider utilization of mixed oxide fuels is expected in Japan. Fuel irradiation tests for development and safety demonstration are quite important, in order to realize theses progress. Operational management on water chemistry, minimizing the long term degradation of reactor components, could have unfavorable influence on the integrity of the fuel rods. Japanese government and the Japan Atomic Energy Agency have decided to re new the Japan Materials Testing Reactor (JMTR) and to install new test rigs, in order to play an active role solving the issues on the development and the safety of the fuel and the plant aging. Fuel integrity under abnormal transient conditions will be investigated using a special capsule type test rig, which has its own power control system under simulated LWR cooling conditions. Water loops for simulation of high duty operation, e.g. high power, high burnup and high rod internal pressure conditions, are proposed for the development and safety examination of the high performance fuels. Combination of the JMTR tests with simulated reactivity initiated accident tests in the Nuclear Safety Research Reactor and loss of coolant accident tests in hot laboratories would provide a comprehensive data for safety evaluation and design progress of the high performance fuels at extended burnups, covering from the normal to the accident conditions, including abnormal transients

  11. Decommissioning of the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-10-28

    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&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&D effort. In addition, techniques from the decommissioning of fission reactors were adapted to the D&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.

  12. Decommissioning of the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    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

  13. Aerospace propulsion products; high-quality rocket ignition systems for the future

    OpenAIRE

    Van Zon, N.; Nevinskaia, A.

    2013-01-01

    Aerospace Propulsion Products is the leading European company in designing and producing rocket ignition systems and spinoff products. One of their directors, Edwin Vermeulen, gave us an insight on the company and its future. He states that “whatever rocket technology is needed, we have the technology in house to provide the ignition systems”.

  14. Aerospace propulsion products; high-quality rocket ignition systems for the future

    NARCIS (Netherlands)

    Van Zon, N.; Nevinskaia, A.

    2013-01-01

    Aerospace Propulsion Products is the leading European company in designing and producing rocket ignition systems and spinoff products. One of their directors, Edwin Vermeulen, gave us an insight on the company and its future. He states that “whatever rocket technology is needed, we have the technolo

  15. An expert system for integrated structural analysis and design optimization for aerospace structures

    Science.gov (United States)

    1992-01-01

    design optimization tasks in the integrated aerospace structural design process. These expert systems were developed to work in conjunction with procedural finite element structural analysis and design optimization modules (developed in-house at SAT, Inc.). The complete software, AutoDesign, so developed, can be used for integrated 'intelligent' structural analysis and design optimization. The software was beta-tested at a variety of companies, used by a range of engineers with different levels of background and expertise. Based on the feedback obtained by such users, conclusions were developed and are provided.

  16. An expert system for integrated structural analysis and design optimization for aerospace structures

    Science.gov (United States)

    1992-04-01

    design optimization tasks in the integrated aerospace structural design process. These expert systems were developed to work in conjunction with procedural finite element structural analysis and design optimization modules (developed in-house at SAT, Inc.). The complete software, AutoDesign, so developed, can be used for integrated 'intelligent' structural analysis and design optimization. The software was beta-tested at a variety of companies, used by a range of engineers with different levels of background and expertise. Based on the feedback obtained by such users, conclusions were developed and are provided.

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

    Science.gov (United States)

    Kotiluoto, Petri; Wasastjerna, Frej; Kekki, Tommi; Sipilä, Heikki; Banzuzi, Kukka; Kinnunen, Petri; Heikinheimo, Liisa

    2009-08-01

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

  18. Tests of a new CCD-camera based neutron radiography detector system at the reactor stations in Munich and Vienna

    Energy Technology Data Exchange (ETDEWEB)

    Lehmann, E.; Pleinert, H. [Paul Scherrer Inst. (PSI), Villigen (Switzerland); Schillinger, B. [Technische Univ. Muenchen (Germany); Koerner, S. [Atominstitut der Oesterreichischen Universitaeten, Vienna (Austria)

    1997-09-01

    The performance of the new neutron radiography detector designed at PSI with a cooled high sensitive CCD-camera was investigated under real neutronic conditions at three beam ports of two reactor stations. Different converter screens were applied for which the sensitivity and the modulation transfer function (MTF) could be obtained. The results are very encouraging concerning the utilization of this detector system as standard tool at the radiography stations at the spallation source SINQ. (author) 3 figs., 5 refs.

  19. Reactor protection systems of 500 MWe PHWRs

    International Nuclear Information System (INIS)

    The 500 MWe PHWR has two totally independent, diverse, fast acting shutdown system called Shutdown System 1 (SDS 1) and Shutdown System 2 (SDS 2). The trip generation circuitry of SDS 1 and SDS 2 are known as Reactor Protection System 1 (RPS 1) and Reactor Protection System 2 (RPS 2) respectively. Some of the features specific to 500 MWe reactors are Core Over Power Protection System (COPPS) based upon in core Self Powered Neutron Detector (SPND) signals, use of local two out of three coincidence logic and adoption of overlap testing for RPS 2, use of Fine Impulse Testing (FIT) in RPS 2, testing of the final control elements namely electro-magnetic clutch of individual Shutoff Rods (SRs) of SDS 1 and all the fast acting valves of SDS 2, etc. The two shutdown systems have totally separate sets of sensors and associated signal processing circuitry as well as physical arrangements. A separate computerised test and monitoring unit is used for each of the two shutdown systems. Use of Programmable Digital Comparator (PDC) unit exclusively for reactor protection systems, has been adopted. The capability of PDC unit is enhanced and communication links are provided for its integration in over all system. The paper describes the design features of reactor protection systems. (author)

  20. Off reactor testings. Technological engineering applicative research

    International Nuclear Information System (INIS)

    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

  1. Dynamic fiber Bragg gratings based health monitoring system of composite aerospace structures

    Science.gov (United States)

    Panopoulou, A.; Loutas, T.; Roulias, D.; Fransen, S.; Kostopoulos, V.

    2011-09-01

    The main purpose of the current work is to develop a new system for structural health monitoring of composite aerospace structures based on real-time dynamic measurements, in order to identify the structural state condition. Long-gauge Fibre Bragg Grating (FBG) optical sensors were used for monitoring the dynamic response of the composite structure. The algorithm that was developed for structural damage detection utilizes the collected dynamic response data, analyzes them in various ways and through an artificial neural network identifies the damage state and its location. Damage was simulated by slightly varying locally the mass of the structure (by adding a known mass) at different zones of the structure. Lumped masses in different locations upon the structure alter the eigen-frequencies in a way similar to actual damage. The structural dynamic behaviour has been numerically simulated and experimentally verified by means of modal testing on two different composite aerospace structures. Advanced digital signal processing techniques, e.g. the wavelet transform (WT), were used for the analysis of the dynamic response for feature extraction. WT's capability of separating the different frequency components in the time domain without loosing frequency information makes it a versatile tool for demanding signal processing applications. The use of WT is also suggested by the no-stationary nature of dynamic response signals and the opportunity of evaluating the temporal evolution of their frequency contents. Feature extraction is the first step of the procedure. The extracted features are effective indices of damage size and location. The classification step comprises of a feed-forward back propagation network, whose output determines the simulated damage location. Finally, dedicated training and validation activities were carried out by means of numerical simulations and experimental procedures. Experimental validation was performed initially on a flat stiffened panel

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-09-01

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

  3. Standard Test Method for Hail Impact Resistance of Aerospace Transparent Enclosures

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This test method covers the determination of the impact resistance of an aerospace transparent enclosure, hereinafter called windshield, during hailstorm conditions using simulated hailstones consisting of ice balls molded under tightly controlled conditions. 1.2 The values stated in inch-pound units are to be regarded as standard. The values given in parentheses are mathematical conversions to SI units that are provided for information only and are not considered standard. 1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use. For specific hazard statements see Section 7.

  4. Development of integrated programs for Aerospace-vehicle Design (IPAD): Product program management systems

    Science.gov (United States)

    Isenberg, J. M.; Southall, J. W.

    1979-01-01

    The Integrated Programs for Aerospace Vehicle Design (IPAD) is a computing system to support company-wide design information processing. This document presents a brief description of the management system used to direct and control a product-oriented program. This document, together with the reference design process (CR 2981) and the manufacture interactions with the design process (CR 2982), comprises the reference information that forms the basis for specifying IPAD system requirements.

  5. 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.

  6. Mobile reactor concepts as applied to testing of compact fusion reactors

    International Nuclear Information System (INIS)

    Compact fusion reactor concepts have recently received increased emphasis because of advantages principally related to their low cost and short development time. Physics experiments are underway and test results are sufficiently encouraging to merit consideration of ignition-type experiments. Since experiments of this nature involve radioactivity, the requirement for test facilities which incorporate remote handling capabilities becomes apparent. One approach to a test facility concept which has particularly attractive features is based on the mobile test reactor concept employing facilities such as are found at the Idaho National Engineering Laboratory (INEL). The mobile reactor test concept was developed in the 1950s and was used extensively in the testing of aircraft nuclear propulsion reactors at the INEL. In this instance, test reactors were assembled on a dolly and were transported to and from test facilities on a four-rail track system. Nuclear operations were conducted from heavily shielded underground control rooms and, for major maintenance operations, the reactors were unplugged and returned to a large, centrally located hot shop. A similar concept is envisioned for compact fusion reactor testing

  7. The mobile reactor concept as applied to testing of compact fusion reactors

    International Nuclear Information System (INIS)

    Compact fusion reactor concepts have recently received increased emphasis because of advantages principally related to their low cost and short development time. Physics experiments are underway and test results are sufficiently encouraging to merit consideration of ''ignition''-type experiments. Since experiments of this nature involve radioactivity, the requirement for test facilities which incorporate remote handling capabilities becomes apparent. One approach to a test facility concept which has particularly attractive features is based on the mobile test reactor concept employing facilities such as are found at the Idaho National Engineering Laboratory (INEL). The mobile reactor test concept was developed in the 1950s and was used extensively in the testing of aircraft nuclear propulsion reactors at the INEL. In this instance, test reactors were assembled on a dolly and were transported to and from test facilities on a fourrail track system. Nuclear operations were conducted from heavily shielded underground control rooms and, for major maintenance operations, the reactors were ''unplugged'' and returned to a large, centrally located ''hot'' shop. A similar concept is envisioned for compact fusion reactor testing

  8. Fiber Bragg Grating Sensor System for Monitoring Smart Composite Aerospace Structures

    Science.gov (United States)

    Moslehi, Behzad; Black, Richard J.; Gowayed, Yasser

    2012-01-01

    Lightweight, electromagnetic interference (EMI) immune, fiber-optic, sensor- based structural health monitoring (SHM) will play an increasing role in aerospace structures ranging from aircraft wings to jet engine vanes. Fiber Bragg Grating (FBG) sensors for SHM include advanced signal processing, system and damage identification, and location and quantification algorithms. Potentially, the solution could be developed into an autonomous onboard system to inspect and perform non-destructive evaluation and SHM. A novel method has been developed to massively multiplex FBG sensors, supported by a parallel processing interrogator, which enables high sampling rates combined with highly distributed sensing (up to 96 sensors per system). The interrogation system comprises several subsystems. A broadband optical source subsystem (BOSS) and routing and interface module (RIM) send light from the interrogation system to a composite embedded FBG sensor matrix, which returns measurand-dependent wavelengths back to the interrogation system for measurement with subpicometer resolution. In particular, the returned wavelengths are channeled by the RIM to a photonic signal processing subsystem based on powerful optical chips, then passed through an optoelectronic interface to an analog post-detection electronics subsystem, digital post-detection electronics subsystem, and finally via a data interface to a computer. A range of composite structures has been fabricated with FBGs embedded. Stress tensile, bending, and dynamic strain tests were performed. The experimental work proved that the FBG sensors have a good level of accuracy in measuring the static response of the tested composite coupons (down to submicrostrain levels), the capability to detect and monitor dynamic loads, and the ability to detect defects in composites by a variety of methods including monitoring the decay time under different dynamic loading conditions. In addition to quasi-static and dynamic load monitoring, the

  9. NASA Engineering Safety Center NASA Aerospace Flight Battery Systems Working Group 2007 Proactive Task Status

    Science.gov (United States)

    Manzo, Michelle A.

    2007-01-01

    In 2007, the NASA Engineering Safety Center (NESC) chartered the NASA Aerospace Flight Battery Systems Working Group to bring forth and address critical battery-related performance/manufacturing issues for NASA and the aerospace community. A suite of tasks identifying and addressing issues related to Ni-H2 and Li-ion battery chemistries was submitted and selected for implementation. The current NESC funded are: (1) Wet Life of Ni-H2 Batteries (2) Binding Procurement (3) NASA Lithium-Ion Battery Guidelines (3a) Li-Ion Performance Assessment (3b) Li-Ion Guidelines Document (3b-i) Assessment of Applicability of Pouch Cells for Aerospace Missions (3b-ii) High Voltage Risk Assessment (3b-iii) Safe Charge Rates for Li-Ion Cells (4) Availability of Source Material for Li-Ion Cells (5) NASA Aerospace Battery Workshop This presentation provides a brief overview of the tasks in the 2007 plan and serves as an introduction to more detailed discussions on each of the specific tasks.

  10. Vibration system identification of Paks and Kozloduy reactor buildings on the basis of the blast test results

    International Nuclear Information System (INIS)

    System identification allows to build mathematical models of a dynamic system based on measured data. System identification is carried out by adjusting parameters within a given model until its output coincides as well as possible with the measured output. The aim of this study is to investigate and model the behavior of complex vibratory systems on the basis of measured excitation and response. The first part of the study describes the theory used in the analysis and the software tools used in the analysis. The second part of the study describes the investigation and modeling of the response of single degree of freedom oscillator excited by sinusoidal and blast excitation. In the third part of the study the system identification of the Kozloduy NPP unit 5 reactor building and Paks NPP unit 1 reactor building is studied and the models are estimated using the method of segmentation of excitation and response. System identification is carried out using MATLAB software by adjusting parameters within a given model until its output coincides as well as possible with the measured output. The types of models used for the were: l) ARX models; 2) ARMAX model; 3) Output-Error (OE) models; 4) Box-Jenkins (BJ) models; 5) State-space models. The model coefficients for different models were calculated using the least-squares and maximum likelihood estimation methods available in MATLAB system identification toolbox. Excitation was in both Paks and Kozloduy case the measured free-field excitation and responses were the vibration responses of the building on the foundation slab level and top of the building. By examining the established models the frequency characteristics of vibration systems were determined with 95 % accuracy and the amplitude response with 80 % accuracy. In case of the steady state response of sinusoidally excited single dof oscillator the modelling gave almost exact results. But in the case of the blast response of the reactor building the obtaining of the

  11. Single-Event Transient Testing of the Crane Aerospace and Electronics SMHF2812D Dual DC-DC Converter

    Science.gov (United States)

    Casey, Megan

    2015-01-01

    The purpose of this testing was to characterize the Crane Aerospace & Electronics (Crane) Interpoint SMHF2812D for single-event transient (SET) susceptibility. These data shall be used for flight lot evaluation, as well as qualification by similarity of the SMHF family of converters, all of which use the same active components.

  12. Kyoto University Reactor diagnostic system

    International Nuclear Information System (INIS)

    For the safety of a nuclear reactor, it is very important that the operators and manager make exact judgement about the various conditions of the nuclear reactor occurring at times. The research is advanced for the purpose of adopting a computer system for the research reactor of Kyoto University (KUR), offering effective information to operators and maintenance workers, making the advice for exactly judging the conditions of the reactor by sufficiently grasping them, consequently, developing the system for increasing the safety of the reactor. For the development of this system, also technical officials took part positively and cooperated in the research and development based on the experience of the maintenance and operation of the research reactor carried out daily. The system comprises the data acquisition part, data base, abnormality diagnostic part, man-machine interface part, and individual dealing part. The abnormality of the reactor is identified by the judgement of operators by referring to the data memorized in the data base, then, the reactor is operated. The constitution of the computer system used is shown. The CPU is a minicomputer ECLIPSE S-140, and the main memory is 512 kB. The auxiliary memories are a fixed disk equipment of 73 MB, two floppy disk equipments and a magnetic tape equipment. Respective subsystems are explained. (Kako, I.)

  13. Reactor monitoring system

    International Nuclear Information System (INIS)

    The present invention concerns a device for monitoring the inside of an FBR type reactor which can not be monitored by a usual optical camera. An ultrasonic camera having an excellent propagating property in a liquid metal sodium is scanned, and reflected waves of the ultrasonic waves are received as signals. The signals are processed by using a virtual realistic feeling (VR) technique such as a head mounting type image display (HMD) and a three dimensional pointing device. With such procedures, the inside of the FBR type reactor can be observed with such a realistic feeling that the inside of the FBR type reactor were seen directly. (I.S.)

  14. 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.

  15. Advanced Test Reactor National Scientific User Facility

    International Nuclear Information System (INIS)

    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.

  16. Obsolescence Challenges for Product-Service Systems in Aerospace and Defence Industry

    OpenAIRE

    Romero Rojo, Francisco Javier; Roy, Rajkumar; Shehab, Essam; Wardle, P. J.

    2009-01-01

    The aerospace and defence industries are moving towards new types of agreement such as availability contracts based on Product-Service System (PSS) business models. Obsolescence has become one of the main problems that will impact on many areas of the system during its life cycle. This paper presents the major challenges to managing obsolescence for availability contracts, identified by means of a comprehensive literature review and several interviews and forums with experts in ob...

  17. Cooling system for reactor container

    International Nuclear Information System (INIS)

    Purpose: To effectively cool a reactor container upon reactor shutdown with no intrusion of metal corrosion products in coolants into the main steam pipe in a BWR type reactor. Constitution: A clean up system comprising a pipeway, a recycling pump, a non-regenerative heat exchanger and a primary coolant purifier and a regenerative heat exchanger is provided branched from a residual heat removing system and the clean up system is connected by way of a valve to a feedwater pipeway, as well as connected by way of the pipeway to the main steam pipeway at the midway of two main steam separation valves outside of the reactor container. This enables to prevent metal corrosion products floating on the surface of reactor water from introducing into the main steam pipe when the pressure vessel is filled with water. Then, since the pressure vessel is filled with primary coolants, the pressure vessel can be cooled uniformly in a short time. (Ikeda, J.)

  18. Integrated Instrumentation and Sensor Systems Enabling Condition-Based Maintenance of Aerospace Equipment

    Directory of Open Access Journals (Sweden)

    Richard C. Millar

    2012-01-01

    Full Text Available The objective of the work reported herein was to use a systems engineering approach to guide development of integrated instrumentation/sensor systems (IISS incorporating communications, interconnections, and signal acquisition. These require enhanced suitability and effectiveness for diagnostics and health management of aerospace equipment governed by the principles of Condition-based maintenance (CBM. It is concluded that the systems engineering approach to IISS definition provided clear benefits in identifying overall system requirements and an architectural framework for categorizing and evaluating alternative architectures, relative to a bottom up focus on sensor technology blind to system level user needs. CBM IISS imperatives identified include factors such as tolerance of the bulk of aerospace equipment operational environments, low intrusiveness, rapid reconfiguration, and affordable life cycle costs. The functional features identified include interrogation of the variety of sensor types and interfaces common in aerospace equipment applications over multiplexed communication media with flexibility to allow rapid system reconfiguration to adapt to evolving sensor needs. This implies standardized interfaces at the sensor location (preferably to open standards, reduced wire/connector pin count in harnesses (or their elimination through use of wireless communications.

  19. Reactor regulating and protection system for a light water reactor

    International Nuclear Information System (INIS)

    Microprocessor based systems are developed for reactor regulation and protection of LWR. A triple modular redundancy approach is followed for the design of this system. This system is functionally partitioned into two sub-systems - Reactor Regulating System (RRS) and Reactor Trip Logic System (RTLS). RRS controls the reactor power as per demand and RTLS generates the reactor trip on abnormal process conditions. This paper describes the details of RRS and RTLS system architecture and fault tolerant and fail-safe features used in the system design. (author)

  20. Safety systems of heavy water reactors and small power reactors

    International Nuclear Information System (INIS)

    After introductional descriptions of heavy water reactors and natural circulation boiling water reactors the safety philosophy and safety systems like ECCS, residual heat removal, protection systems etc., are described. (RW)

  1. Towards Requirements in Systems Engineering for Aerospace IVHM Design

    Data.gov (United States)

    National Aeronautics and Space Administration — Health management (HM) technologies have been employed for safety critical system for decades, but a coherent systematic process to integrate HM into the system...

  2. Development of modern safe systems of work at the Imperial College Reactor Centre and their application to neutron detector testing and nuclear training courses

    International Nuclear Information System (INIS)

    safety hazards. For straight forward not especially hazardous tasks the risk assessment is sufficient to control the operation and provides all the information and instructions for the work to be carried out safely. Operations with a radiological risk, particularly those involving non classified workers in controlled areas have additional written controls in the form of a Written System of Work which goes into details of radiological hazards and safeguards and may involve hold points. The most hazardous tasks (both radiological and conventional) are controlled through the use of Permits to Work. In all cases, operations which involve the use of contractors, require careful assessment of the Contracting Organisation and their plans for the work (method statements etc.) to ensure that the contractors are suitably trained and experienced to do the work. These procedures have allowed the development of the facility requiring man access into a Controlled Area with an open beam tube where doses are maintained ALARP. The CONSORT reactor provides a facility for the calibration and periodic testing of neutron flux detectors (primarily fission chambers and ion chambers). The first case study outlines the detector testing programme which requires the use of a broad range of well thermalised neutron fluxes over eleven orders of magnitude. The facility in which these tests are carried out consists of a beam tube penetrating the graphite thermal column located immediately outside of the tank housing the reactor core. During testing, the detectors are manually loaded into and unloaded from the beam tube with the reactor operating at low power (up to 2kW). This process has been developed, assessed and controlled using the procedures described, in order to ensure that the doses are minimised and acceptable in line with the ALARP principle. On line testing of the detectors is carried out by the manufacturer's employees on site, and involves the control and supervision of these workers

  3. Ceramic multilayer based on ZrB2/SiC system for aerospace applications

    OpenAIRE

    Padovano, Elisa

    2015-01-01

    The work of this PhD thesis is focused on the processing and characterisation of ZrB2/SiC based multilayer materials, produced by tape casting and sintered without pressure assistance for aerospace applications. The multilayer components were processed in order to be used as external part of a thermal protection system; because of they directly face the atmosphere, in order to withstand high temperature, high heat fluxes and oxidizing environment, they have to show good oxidation and thermal ...

  4. Concurrent Systems Engineering in Aerospace: From Excel-based to Model Driven Design

    OpenAIRE

    Schumann, Holger; Wendel, Heinrich; Braukhane, Andy; Berres, Axel; Gerndt, Andreas; Schreiber, Andreas

    2010-01-01

    Concurrent engineering is a modern and very effective discipline of systems engineering. In the European space domain, the European Space Agency is the pioneer in this area and has performed early design studies for 10 years now. The Integrated Design Model (IDM) is still the state of the art in concurrent engineering software environments. It is based on Microsoft Excel, which induces several benefits and drawbacks related to concurrent engineering. The German Aerospace Center has used t...

  5. Development of High-Temperature Sodium Loop System for Materials Compatibility Test for Ultra-long Cycle Fast Reactor (UCFR)

    International Nuclear Information System (INIS)

    Sodium is a candidate for fast reactor coolants that has been believed to have favorable compatibility with structural materials. However, recent studies showed results which need for a more careful attention at this previous belief. For prolonging the service life time of cladding and structural materials in contact with liquid sodium, more detail analysis methods are needed to examine this material compatibility issue with sodium. As a candidate of liquid metals coolants of Ultra-long Cycle Fast Reactor (UCFR), the compatibility of sodium with cladding materials has to be investigated in detail with long term exposure time. It is known that corrosion promotes corrosion in two ways. One is corrosion produced by dissolution of alloy elements into sodium and the other is corrosion produced through a chemical reaction with impurities in sodium, especially dissolved oxygen. The use of the technique of impedance spectroscopy to measure the electrical impedance response of any oxide layers may be a good experimental tool to this monitoring system. The motivation of current study is to investigate the relationship between the electrochemical behaviors of oxide scales on ferritic-martensitic (FM) steel and austenitic steels (as shown in Table I) and their corrosion rates in liquid sodium environment

  6. Towards Requirements in Systems Engineering for Aerospace IVHM Design

    Science.gov (United States)

    Saxena, Abhinav; Roychoudhury, Indranil; Lin, Wei; Goebel, Kai

    2013-01-01

    Health management (HM) technologies have been employed for safety critical system for decades, but a coherent systematic process to integrate HM into the system design is not yet clear. Consequently, in most cases, health management resorts to be an after-thought or 'band-aid' solution. Moreover, limited guidance exists for carrying out systems engineering (SE) on the subject of writing requirements for designs with integrated vehicle health management (IVHM). It is well accepted that requirements are key to developing a successful IVHM system right from the concept stage to development, verification, utilization, and support. However, writing requirements for systems with IVHM capability have unique challenges that require the designers to look beyond their own domains and consider the constraints and specifications of other interlinked systems. In this paper we look at various stages in the SE process and identify activities specific to IVHM design and development. More importantly, several relevant questions are posed that system engineers must address at various design and development stages. Addressing these questions should provide some guidance to systems engineers towards writing IVHM related requirements to ensure that appropriate IVHM functions are built into the system design.

  7. Development and testing of a service concept for integration of simulation in the remote nuclear reactor monitoring system

    International Nuclear Information System (INIS)

    In this dissertation a framework for distributed software services is developed. This allows the easy integration of software components in complex systems. The capabilities of the components are offered as services, which can be consumed by others. Services share a common and generic interface, which increases flexibility and the potential use in systems that have to act, react and communicate flexibly. Services can be suspended, resumed and terminated at run-time. The service framework has been implemented in C++ and uses CORBA as distribution mechanism. It is fully multi-threaded and provides recovery mechanisms in order to safe computation time after a crash. The implementation aims at high flexibility, extensibility, clarity and the use of modern, object-oriented software technologies. The framework is independent of the actual domain and of the data objects that need to be exchanged. The service framework is used in the large-scale Remote Nuclear Reactor Monitoring System of Baden-Wuerttemberg. This mission-critical system simulates the dispersion and deposition of radioactive nuclides and their effects on humans. In the past, the single steps of the simulation had to be co-ordinated by humans. Now, they are provided as automated information services, integrated into the system and are always available. For the system, a special multi-agent system has been employed. The service framework is the backbone of that multi-agent system. Some experiences and insights on the system level are discussed, which cover mainly the distribution of system knowledge and its impact on maintenance costs as well as procedural vs. rule-based problems and approaches. (orig.)

  8. Development of sensor augmented robotic weld systems for aerospace propulsion system fabrication

    Science.gov (United States)

    Jones, C. S.; Gangl, K. J.

    1986-01-01

    In order to meet stringent performance goals for power and reuseability, the Space Shuttle Main Engine was designed with many complex, difficult welded joints that provide maximum strength and minimum weight. To this end, the SSME requires 370 meters of welded joints. Automation of some welds has improved welding productivity significantly over manual welding. Application has previously been limited by accessibility constraints, requirements for complex process control, low production volumes, high part variability, and stringent quality requirements. Development of robots for welding in this application requires that a unique set of constraints be addressed. This paper shows how robotic welding can enhance production of aerospace components by addressing their specific requirements. A development program at the Marshall Space Flight Center combining industrial robots with state-of-the-art sensor systems and computer simulation is providing technology for the automation of welds in Space Shuttle Main Engine production.

  9. Computational simulation for concurrent engineering of aerospace propulsion systems

    Science.gov (United States)

    Chamis, C. C.; Singhal, S. N.

    1993-01-01

    Results are summarized for an investigation to assess the infrastructure available and the technology readiness in order to develop computational simulation methods/software for concurrent engineering. These results demonstrate that development of computational simulation methods for concurrent engineering is timely. Extensive infrastructure, in terms of multi-discipline simulation, component-specific simulation, system simulators, fabrication process simulation, and simulation of uncertainties--fundamental to develop such methods, is available. An approach is recommended which can be used to develop computational simulation methods for concurrent engineering of propulsion systems and systems in general. Benefits and issues needing early attention in the development are outlined.

  10. Computational simulation of concurrent engineering for aerospace propulsion systems

    Science.gov (United States)

    Chamis, C. C.; Singhal, S. N.

    1992-01-01

    Results are summarized of an investigation to assess the infrastructure available and the technology readiness in order to develop computational simulation methods/software for concurrent engineering. These results demonstrate that development of computational simulations methods for concurrent engineering is timely. Extensive infrastructure, in terms of multi-discipline simulation, component-specific simulation, system simulators, fabrication process simulation, and simulation of uncertainties - fundamental in developing such methods, is available. An approach is recommended which can be used to develop computational simulation methods for concurrent engineering for propulsion systems and systems in general. Benefits and facets needing early attention in the development are outlined.

  11. Anthony Pro - Human Automation Interaction in Aerospace Systems Project

    Data.gov (United States)

    National Aeronautics and Space Administration — This proposed project aims to demonstrate the feasibility and utility of a data mining system designed to facilitate the interpretation of information obtained from...

  12. A new SMART sensing system for aerospace structures

    Science.gov (United States)

    Zhang, David C.; Yu, Pin; Beard, Shawn; Qing, Peter; Kumar, Amrita; Chang, Fu-Kuo

    2007-04-01

    It is essential to ensure the safety and reliability of in-service structures such as unmanned vehicles by detecting structural cracking, corrosion, delamination, material degradation and other types of damage in time. Utilization of an integrated sensor network system can enable automatic inspection of such damages ultimately. Using a built-in network of actuators and sensors, Acellent is providing tools for advanced structural diagnostics. Acellent's integrated structural health monitoring system consists of an actuator/sensor network, supporting signal generation and data acquisition hardware, and data processing, visualization and analysis software. This paper describes the various features of Acellent's latest SMART sensing system. The new system is USB-based and is ultra-portable using the state-of-the-art technology, while delivering many functions such as system self-diagnosis, sensor diagnosis, through-transmission mode and pulse-echo mode of operation and temperature measurement. Performance of the new system was evaluated for assessment of damage in composite structures.

  13. Elements of reactor system design

    International Nuclear Information System (INIS)

    When the first commercial nuclear power plants were designed, each plant was treated as a new design problem. However, it became apparent that the full design effort was far too lengthy and costly to be undertaken for each order. The reactor system vendors have therefore developed a series of essentially standard reactor designs. A utility customer is offered that standard design which most closely meets his requirements. Only minor modification are made in order to meet particular local requirements. The reactor design effort for such a plant is generally limited to (a) a verification that the standard system proposed will meet the required specifications and (b) a revision of the safety analysis to take into consideration the features of the particular site. Standard system designs are usually revised on a regular basis to take advantage of new developments and operational experience. It has become customary to refer to the reactor core and entire primary system as the ''nuclear steam supply system''. In the United States, when a reactor vendor supplies a system to a public utility, it is generally only the ''nuclear steam supply system'' and specific auxiliaries which are supplied. The reactor vendor will specify the general requirements of the steam cycle, vapor container and auxiliary systems and safety systems which are not vendor supplied. The detailed design of these systems, as well as the complete structural and electrical design, is normally handled by the utility or an architect-engineer engaged by the utility. The safety analysis is usually conducted by the reactor vendor. As more experience with nuclear systems is gained, it is likely that the larger utilities will assume an expanded role in the design process

  14. Advanced instrumentation for next-generation aerospace propulsion control systems

    Science.gov (United States)

    Barkhoudarian, S.; Cross, G. S.; Lorenzo, Carl F.

    1993-01-01

    New control concepts for the next generation of advanced air-breathing and rocket engines and hypersonic combined-cycle propulsion systems are analyzed. The analysis provides a database on the instrumentation technologies for advanced control systems and cross matches the available technologies for each type of engine to the control needs and applications of the other two types of engines. Measurement technologies that are considered to be ready for implementation include optical surface temperature sensors, an isotope wear detector, a brushless torquemeter, a fiberoptic deflectometer, an optical absorption leak detector, the nonintrusive speed sensor, and an ultrasonic triducer. It is concluded that all 30 advanced instrumentation technologies considered can be recommended for further development to meet need of the next generation of jet-, rocket-, and hypersonic-engine control systems.

  15. Reliability tests for reactor internals replacement technology

    International Nuclear Information System (INIS)

    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)

  16. Aerospace Medicine

    Science.gov (United States)

    Michaud, Vince

    2015-01-01

    NASA Aerospace Medicine overview - Aerospace Medicine is that specialty area of medicine concerned with the determination and maintenance of the health, safety, and performance of those who fly in the air or in space.

  17. Reactor physics and economic aspects of the CANDU reactor system

    International Nuclear Information System (INIS)

    A history of the development of the CANDU system is given along with a fairly detailed description of the 600 MW(e) CANDU reactor. Reactor physics calculation methods are described, as well as comparisons between calculated reactor physics parameters and those measured in research and power reactors. An examination of the economics of CANDU in the Ontario Hydro system and a comparison between fossil fuelled and light water reactors is presented. Some physics, economics and resources aspects are given for both low enriched uranium and thorium-fuelled CANDU reactors. Finally the RβD program in Advanced Fuel Cycles is briefly described

  18. Introduction to System Health Engineering and Management in Aerospace

    Science.gov (United States)

    Johnson, Stephen B.

    2005-01-01

    This paper provides a technical overview of Integrated System Health Engineering and Management (ISHEM). We define ISHEM as "the paper provides a techniques, and technologies used to design, analyze, build, verify, and operate a system to prevent faults and/or minimize their effects." This includes design and manufacturing techniques as well operational and managerial methods. ISHEM is not a "purely technical issue" as it also involves and must account for organizational, communicative, and cognitive f&ms of humans as social beings and as individuals. Thus the paper will discuss in more detail why all of these elements, h m the technical to the cognitive and social, are necessary to build dependable human-machine systems. The paper outlines a functional homework and architecture for ISHEM operations, describes the processes needed to implement ISHEM in the system life-cycle, and provides a theoretical framework to understand the relationship between the different aspects of the discipline. It then derives from these and the social and cognitive bases a set of design and operational principles for ISHEM.

  19. A first system/CFD coupled simulation of a complete nuclear reactor transient using CATHARE2 and TRIO{sub U}. Preliminary validation on the Phénix Reactor Natural Circulation Test

    Energy Technology Data Exchange (ETDEWEB)

    Bavière, R., E-mail: roland.baviere@cea.fr; Tauveron, N., E-mail: nicolas.tauveron@cea.fr; Perdu, F., E-mail: fabien.perdu@cea.fr; Garré, E., E-mail: emile.garre@cea.fr; Li, S., E-mail: simon.li@cea.fr

    2014-10-01

    Highlights: • A system/CFD coupling methodology for thermal-hydraulics analysis. • Application of the model to the Phénix Reactor Natural Circulation Test. • Validation of the methodology against experimental data. - Abstract: The natural circulation test (NCT) was conducted in the Phénix prototype French 580 MWth sodium fast reactor (SFR) in 2009. The main goal of the Phénix NCT is to validate system- and CFD-codes with respect to the establishment of natural circulation in the primary system of a pool type SFR. The present paper describes the calculation of the NCT by coupling the 3D computational fluid dynamics (CFD) code TRIO{sub U} with the best estimate thermal hydraulic system code CATHARE. The coupling methodology and the modeling at the system and at the CFD scales are first presented. A validation of the coupling methodology based on a coupled CATHARE/CATHARE calculation compared to the standard CATHARE predictions is then proposed. In a second step, the results of the TRIO{sub U}/CATHARE calculation are compared both to the available experimental data and to the results of a CATHARE alone computation. These comparisons highlight the effectiveness of coupling CFD- and system-codes for the analysis of plant transients where three-dimensional phenomena play an important role.

  20. A first system/CFD coupled simulation of a complete nuclear reactor transient using CATHARE2 and TRIOU. Preliminary validation on the Phénix Reactor Natural Circulation Test

    International Nuclear Information System (INIS)

    Highlights: • A system/CFD coupling methodology for thermal-hydraulics analysis. • Application of the model to the Phénix Reactor Natural Circulation Test. • Validation of the methodology against experimental data. - Abstract: The natural circulation test (NCT) was conducted in the Phénix prototype French 580 MWth sodium fast reactor (SFR) in 2009. The main goal of the Phénix NCT is to validate system- and CFD-codes with respect to the establishment of natural circulation in the primary system of a pool type SFR. The present paper describes the calculation of the NCT by coupling the 3D computational fluid dynamics (CFD) code TRIOU with the best estimate thermal hydraulic system code CATHARE. The coupling methodology and the modeling at the system and at the CFD scales are first presented. A validation of the coupling methodology based on a coupled CATHARE/CATHARE calculation compared to the standard CATHARE predictions is then proposed. In a second step, the results of the TRIOU/CATHARE calculation are compared both to the available experimental data and to the results of a CATHARE alone computation. These comparisons highlight the effectiveness of coupling CFD- and system-codes for the analysis of plant transients where three-dimensional phenomena play an important role

  1. Startup testing of Romania dual-core test reactor

    International Nuclear Information System (INIS)

    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)

  2. Project, installation and operational tests of a pneumatic system for the IEA-R1 reactor materials

    International Nuclear Information System (INIS)

    Pneumatic Transfer Systems (PTS) are equipment broadly and world widely used for the transport, movement and transfer of diverse types of materials, objects and cargo between two or more environments, near or distant from each other [1]. Due to their flexibility and quickness, the system application is present in several areas, such as medicine (hospitals and clinic analyses laboratories); industry (automobile, metallurgy, iron-making. chemical, food production) commerce (gasoline stations, cinemas, supermarkets, banks, tolls, on-line commerce, casinos); public service (public institutions, courts). In the nuclear field, the PTS has, also, a vast application, highlighting its use in the radioisotope and radiopharmaceuticals of short half life production, such as 67Ga, 201Tl, 18F and 123I-ultra pure. The development of this work is directed to the application of the Pneumatic Transfer System in transport and transfer of materials that will be irradiated in the IEA-R1 reactor, located in the Institute of Energetic and Nuclear Research, IPEN/CNEN-SP, for application of the Neutron Activation Analysis (NAA). (author)

  3. Design of aerospace control systems using fractional PID controller

    Directory of Open Access Journals (Sweden)

    Magdy A.S. Aboelela

    2012-07-01

    Full Text Available The goal is to control the trajectory of the flight path of six degree of freedom flying body model using fractional PID. The design of fractional PID controller for the six degree of freedom flying body is described. The parameters of fractional PID controller are optimized by particle swarm optimization (PSO method. In the optimization process, various objective functions were considered and investigated to reflect both improved dynamics of the missile system and reduced chattering in the control signal of the controller.

  4. Startup of the FFTF sodium cooled reactor. [Acceptance Test Program

    Energy Technology Data Exchange (ETDEWEB)

    Redekopp, R.D.; Umek, A.M.

    1981-03-01

    The Fast Flux Test Facility (FFTF), located on the Department of Energy (DOE) Hanford Reservation near Richland, Washington, is a 3 Loop 400 MW(t) sodium cooled fast reactor with a primary mission to test fuels and materials for development of the Liquid Metal Fast Breeder Reactor (LMFBR). Bringing FFTF to a condition to accomplish this mission is the goal of the Acceptance Test Program (ATP). This program was the mechanism for achieving startup of the FFTF. Highlights of the ATP involving the system inerting, liquid metal and inerted cell testing and initial ascent to full power are discussed.

  5. Models for transient analyses in advanced test reactors

    OpenAIRE

    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 whi...

  6. Reactor group constants and benchmark test

    International Nuclear Information System (INIS)

    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)

  7. Safety Analysis for Power Reactor Protection System

    International Nuclear Information System (INIS)

    The main function of a Reactor Protection System (RPS) is to safely shutdown the reactor and prevents the release of radioactive materials. The purpose of this paper is to present a technique and its application for used in the analysis of safety system of the Nuclear Power Plant (NPP). A more advanced technique has been presented to accurately study such problems as the plant availability assessments and Technical Specifications evaluations that are becoming increasingly important. The paper provides the Markov model for the Reactor Protection System of the NPP and presents results of model evaluations for two testing policies in technical specifications. The quantification of the Markov model provides the probability values that the system will occupy each of the possible states as a function of time.

  8. IECEC '91; Proceedings of the 26th Intersociety Energy Conversion Engineering Conference, Boston, MA, Aug. 4-9, 1991. Vol. 2 - Aerospace power systems, conversion technologies

    International Nuclear Information System (INIS)

    The present volume on energy and the environment discusses space power requirements, space power systems, space power systems hardware, space radioisotope systems, space solar arrays, space solar cells, space station power, and terrestrial applications of aerospace technology. Attention is given to NASA future space power requirements and issues, the design of a battery charger for the NASA EOS Space Platform, in situ carbon dioxide fixation on Mars, and a preliminary design update of the CRAF/Cassini Power Subsystem. Topics addressed include concentrator testing using projected images, solar power satellites and demonstraton platforms from nonterrestrial materials, a mass sensitivity analysis of lunar orbiting beam power systems, and a power-beaming-based infrastructure for space power. Also discussed are fiber-optic sensors for aerospace electrical measurements, the preliminary design of a mobile lunar power supply, advanced power systems for EOS, and Air Force photovoltaic array alternatives

  9. A generalized concept for cost-effective structural design. [Statistical Decision Theory applied to aerospace systems

    Science.gov (United States)

    Thomas, J. M.; Hawk, J. D.

    1975-01-01

    A generalized concept for cost-effective structural design is introduced. It is assumed that decisions affecting the cost effectiveness of aerospace structures fall into three basic categories: design, verification, and operation. Within these basic categories, certain decisions concerning items such as design configuration, safety factors, testing methods, and operational constraints are to be made. All or some of the variables affecting these decisions may be treated probabilistically. Bayesian statistical decision theory is used as the tool for determining the cost optimum decisions. A special case of the general problem is derived herein, and some very useful parametric curves are developed and applied to several sample structures.

  10. Technical Letter Report, An Evaluation of Ultrasonic Phased Array Testing for Reactor Piping System Components Containing Dissimilar Metal Welds, JCN N6398, Task 2A

    Energy Technology Data Exchange (ETDEWEB)

    Diaz, Aaron A.; Cinson, Anthony D.; Crawford, Susan L.; Anderson, Michael T.

    2009-11-30

    Research is being conducted for the U.S. Nuclear Regulatory Commission at the Pacific Northwest National Laboratory to assess the effectiveness and reliability of advanced nondestructive examination (NDE) methods for the inspection of light-water reactor components. The scope of this research encom¬passes primary system pressure boundary materials including dissimilar metal welds (DMWs), cast austenitic stainless steels (CASS), piping with corrosion-resistant cladding, weld overlays, inlays and onlays, and far-side examinations of austenitic piping welds. A primary objective of this work is to evaluate various NDE methods to assess their ability to detect, localize, and size cracks in steel components that challenge standard and/or conventional inspection methodologies. This interim technical letter report provides a summary of a technical evaluation aimed at assessing the capabilities of phased-array (PA) ultrasonic testing (UT) methods as applied to the inspection of small-bore DMW components that exist in the reactor coolant systems (RCS) of pressurized water reactors (PWRs). Operating experience and events such as the circumferential cracking in the reactor vessel nozzle-to-RCS hot leg pipe at V.C. Summer nuclear power station, identified in 2000, show that in PWRs where primary coolant water (or steam) are present under normal operation, Alloy 82/182 materials are susceptible to pressurized water stress corrosion cracking. The extent and number of occurrences of DMW cracking in nuclear power plants (domestically and internationally) indicate the necessity for reliable and effective inspection techniques. The work described herein was performed to provide insights for evaluating the utility of advanced NDE approaches for the inspection of DMW components such as a pressurizer surge nozzle DMW, a shutdown cooling pipe DMW, and a ferritic (low-alloy carbon steel)-to-CASS pipe DMW configuration.

  11. Static and dynamic performance tests of nuclear powered ship Mutsu reactor (report on nuclear ship Mutsu power-up tests)

    International Nuclear Information System (INIS)

    The power-up tests of the Mutsu reactor were performed from March 29th 1990 to December 14th. The tests were divided into six phases: The tests Phase 0 and Phase 1 were done in the state that the ship was moored at the quay of Sekinehama port in March and April; The tests Phase 2, Phase 3, Phase 4, and Phase 5 were done on the Pacific Ocean from July to December. Present report describes the test results on the static and dynamic plant performance. On static plant performance tests, there are 13 test items including measurements of primary system heat balance at low and high power levels, a virgin run of feed water pump with SG steam, a change-over test of steam supply of auxiliary boiler to SG. On the dynamic plant performance, there are 11 test items including a test of reactor power auto-control system, a test of main feed water auto-control system, a test of small load variation, a load increasing test, a turbine trip test, tests of ahead and astern maneuvering, a test of single loop operation, and a reactor scram test. The reactor power for each item's test was increased step by step from zero power to the goal of rated power of 100 %, 36 MWt. In order to confirm proper reactor system performance, criteria were laid down for the static and dynamic tests: for example, (1) reactor scram shall not occur, (2) pressurizer relief valve and steam generator safety valve shall not work, and (3) after the transients reactor systems shall become the steady state without manual adjustment of the reactor control system. The test results satisfied these criteria and some of test data showed that reactor had much more margin in any performance for design. It is verified, therefore, that the Mutsu reactor systems have adequate performances as a marine reactor and that one is capable to respond smoothly and safely to the load of ship's demand. (author)

  12. Plasma reactor waste management systems

    Science.gov (United States)

    Ness, Robert O., Jr.; Rindt, John R.; Ness, Sumitra R.

    1992-01-01

    The University of North Dakota is developing a plasma reactor system for use in closed-loop processing that includes biological, materials, manufacturing, and waste processing. Direct-current, high-frequency, or microwave discharges will be used to produce plasmas for the treatment of materials. The plasma reactors offer several advantages over other systems, including low operating temperatures, low operating pressures, mechanical simplicity, and relatively safe operation. Human fecal material, sunflowers, oats, soybeans, and plastic were oxidized in a batch plasma reactor. Over 98 percent of the organic material was converted to gaseous products. The solids were then analyzed and a large amount of water and acid-soluble materials were detected. These materials could possibly be used as nutrients for biological systems.

  13. Technology Options for a Fast Spectrum Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    D. M. Wachs; R. W. King; I. Y. Glagolenko; Y. Shatilla

    2006-06-01

    Idaho National Laboratory in collaboration with Argonne National Laboratory has evaluated technology options for a new fast spectrum reactor to meet the fast-spectrum irradiation requirements for the USDOE Generation IV (Gen IV) and Advanced Fuel Cycle Initiative (AFCI) programs. The US currently has no capability for irradiation testing of large volumes of fuels or materials in a fast-spectrum reactor required to support the development of Gen IV fast reactor systems or to demonstrate actinide burning, a key element of the AFCI program. The technologies evaluated and the process used to select options for a fast irradiation test reactor (FITR) for further evaluation to support these programmatic objectives are outlined in this paper.

  14. Nuclear reactor measurement system

    International Nuclear Information System (INIS)

    An instrument to detect the temperature and flow-rate of the liquid metal current of a coolant fluid sample from adjacent sub-assemblies of a liquid metal-cooled nuclear reactor is described. It includes three thermocouple hot junctions mounted in series, each intended for exposure to a sample-current from a single sub-assembly, electromagnetic coils being mounted around an induction core which detects variations in the liquid metal flow-rate by deformation of the lines of flux. The instrument may also include a thermocouple to detect the mean temperature of the sample-current of coolant fluid from several sources, the result being that the temperature of the coolant fluid current in a sub-assembly may be inferred from the three temperature readings associated with this sub-assembly

  15. Preliminary Results of an On-Line, Multi-Spectrometer Fission Product Monitoring System to Support Advanced Gas Reactor Fuel Testing and Qualification in the Advanced Test Reactor at the Idaho National Laboratory

    International Nuclear Information System (INIS)

    The Advanced Gas Reactor-1 (AGR-1) experiment is the first experiment in a series of eight separate low enriched uranium (LEU) oxycarbide (UCO) tri-isotropic (TRISO) particle fuel (in compact form) experiments scheduled for placement in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The experiment began irradiation in the ATR with a cycle that reached full power on December 26, 2006 and will continue irradiation for about 2.5 years. During this time six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The goals of the irradiation experiment is to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. This paper presents the preliminary test details of the fuel performance, as measured by the control and acquisition software

  16. Preliminary Results of an On-Line, Multi-Spectrometer Fission Product Monitoring System to Support Advanced Gas Reactor Fuel Testing and Qualification in the Advanced Test Reactor at the Idaho National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Dawn M. Scates; John K. Hartwell; John B. Walter; Mark W. Drigert

    2007-10-01

    The Advanced Gas Reactor -1 (AGR-1) experiment is the first experiment in a series of eight separate low enriched uranium (LEU) oxycarbide (UCO) tri-isotropic (TRISO) particle fuel (in compact form) experiments scheduled for placement in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The experiment began irradiation in the ATR with a cycle that reached full power on December 26, 2006 and will continue irradiation for about 2.5 years. During this time six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The goals of the irradiation experiment is to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. This paper presents the preliminary test details of the fuel performance, as measured by the control and acquisition software.

  17. Integrated software health management for aerospace guidance, navigation, and control systems: A probabilistic reasoning approach

    Science.gov (United States)

    Mbaya, Timmy

    Embedded Aerospace Systems have to perform safety and mission critical operations in a real-time environment where timing and functional correctness are extremely important. Guidance, Navigation, and Control (GN&C) systems substantially rely on complex software interfacing with hardware in real-time; any faults in software or hardware, or their interaction could result in fatal consequences. Integrated Software Health Management (ISWHM) provides an approach for detection and diagnosis of software failures while the software is in operation. The ISWHM approach is based on probabilistic modeling of software and hardware sensors using a Bayesian network. To meet memory and timing constraints of real-time embedded execution, the Bayesian network is compiled into an Arithmetic Circuit, which is used for on-line monitoring. This type of system monitoring, using an ISWHM, provides automated reasoning capabilities that compute diagnoses in a timely manner when failures occur. This reasoning capability enables time-critical mitigating decisions and relieves the human agent from the time-consuming and arduous task of foraging through a multitude of isolated---and often contradictory---diagnosis data. For the purpose of demonstrating the relevance of ISWHM, modeling and reasoning is performed on a simple simulated aerospace system running on a real-time operating system emulator, the OSEK/Trampoline platform. Models for a small satellite and an F-16 fighter jet GN&C (Guidance, Navigation, and Control) system have been implemented. Analysis of the ISWHM is then performed by injecting faults and analyzing the ISWHM's diagnoses.

  18. Power reactor information system (PRIS)

    International Nuclear Information System (INIS)

    Since the very beginning of commercial operation of nuclear power plants, the nuclear power industry worldwide has accumulated more than 5000 reactor years of experience. The IAEA has been collecting Operating Experience data for Nuclear Power Plants since 1970 which were computerized in 1980. The Agency has undertaken to make Power Reactor Information System (PRIS) available on-line to its Member States. The aim of this publication is to provide the users of PRIS from their terminals with description of data base and communication systems and to show the methods of accessing the data

  19. Additive Manufacturing Enabled Ubiquitous Sensing in Aerospace and Integrated Building Systems

    Science.gov (United States)

    Mantese, Joseph

    2015-03-01

    Ubiquitous sensing is rapidly emerging as a means for globally optimizing systems of systems by providing both real time PHM (prognostics, diagnostics, and health monitoring), as well as expanded in-the-loop control. In closed or proprietary systems, such as in aerospace vehicles and life safety or security building systems; wireless signals and power must be supplied to a sensor network via single or multiple data concentrators in an architecture that ensures reliable/secure interconnectivity. In addition, such networks must be robust to environmental factors, including: corrosion, EMI/RFI, and thermal/mechanical variations. In this talk, we describe the use of additive manufacturing processes guided by physics based models for seamlessly embedding a sensor suite into aerospace and building system components; while maintaining their structural integrity and providing wireless power, sensor interrogation, and real-time diagnostics. We detail this approach as it specifically applies to industrial gas turbines for stationary land power. This work is supported through a grant from the National Energy Technology Laboratory (NETL), a division of the Department of Energy.

  20. Analysis of a loss of forced cooling test using the High Temperature Engineering Test Reactor (HTTR)

    International Nuclear Information System (INIS)

    The High Temperature Engineering Test Reactor (HTTR) is the first High Temperature Gas-cooled Reactor (HTGR) built at the Oarai Research and Development Center of JAEA, with a thermal power of 30 MW and a maximum reactor outlet coolant temperature of 950degC (Saito, 1994). Test researches are being conducted using the HTTR to improve HTGR technologies and to collaborate with domestic industries to contribute to foreign projects for acceleration of HTGR development worldwide. To improve HTGR technologies, advanced analysis techniques are being developed using data obtained with the HTTR, which include reactor kinetics, thermal-hydraulics, safety evaluation, and fuel performance evaluation data (including the behavior of fission products). A three-gas-circulators trip test and a vessel-cooling-system stop test were planned as a loss-of-forced-cooling test and demonstrate the inherent safety features of HTGR. The vessel-cooling-system stop test consists of stopping the vessel-cooling-system located outside the reactor pressure vessel (RPV), to remove the residual heat of the reactor core as soon as the three-gas-circulators are tripped. All three-gas-circulators is tripped at 9 MW. The primary coolant flow rate is reduced from the rated 45 t/h to 0 t/h. The control rods are not inserted into the core and the reactor power control system does not operated. A core dynamics analysis of the loss-of-forced-cooling test of the HTTR is performed. Analytical results for the reactor transient during the test are presented in this report. It is determined that the reactor power immediately decreases to the decay heat level due to the negative reactivity feedback effect of the core, even though the reactor shutdown system is not operational, and that the temperature distribution in the core changes slowly because of the high heat capacity due to the large amount of core graphite. Furthermore, the relation between the reactivities (namely, the Doppler, moderator temperature, and

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

    International Nuclear Information System (INIS)

    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)

  2. Reactor Protection Systems. Diverse Approaches

    International Nuclear Information System (INIS)

    Defence in depth design criteria applied to nuclear instrumentation, in particular, to reactor protection systems (RPS), include redundancy, diversity and fail-safe behaviour. Typically, two out of three ('2oo3'), majority-voting systems meet redundancy criteria. A careful analysis of signal levels and polarity and the use of several techniques, such as lives zeros, bias toward safe state, etc. guarantee the same degree of fail-safe behaviour. Diversity criteria, in general, are met by the whole system using more than one method to protect the integrity of reactor (i.e. rod drop plus boron injection), but not for the single instrumentation chain. Moreover, the increasing information needs of supervision systems encourage the use of digital instrumentation in RPS; if the digital instrumentation has software based implementation, the diversity requirement will be mandatory for the instrumentation of each system. In the paper, three possible configurations of the first protection system (rod drop) are analysed. The first one is the traditional hardware approach, the second one is a software based system, and the last one is a proposed mix system. For all configurations, a redundant system two out of four ('2oo4') is assumed. Availability and reliability points of view are taken into account. The proposed mix system is explained in full detail. A discussion about programmable logic and its considerations are introduced. A CPLD based system in a research reactor (RA1) and its functionality are explained. (author)

  3. Standard Test Method for Bird Impact Testing of Aerospace Transparent Enclosures

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This test method covers conducting bird impact tests under a standard set of conditions by firing a packaged bird at a stationary transparency mounted in a support structure. 1.2 The values stated in inch-pound units are to be regarded as standard. The values given in parentheses are mathematical conversions to SI units that are provided for information only and are not considered standard. 1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use. For specific hazard statements, see Section 8.

  4. Heavy Water Components Test Reactor Decommissioning

    International Nuclear Information System (INIS)

    The Heavy Water Components Test Reactor (HWCTR) Decommissioning Project was initiated in 2009 as a Comprehensive Environmental Response, Compensation and Liability Act (CERCLA) Removal Action with funding from the American Recovery and Reinvestment Act (ARRA). This paper summarizes the history prior to 2009, the major D and D activities, and final end state of the facility at completion of decommissioning in June 2011. The HWCTR facility was built in 1961, operated from 1962 to 1964, and is located in the northwest quadrant of the Savannah River Site (SRS) approximately three miles from the site boundary. The HWCTR was a pressurized heavy water test reactor used to develop candidate fuel designs for heavy water power reactors. In December of 1964, operations were terminated and the facility was placed in a standby condition as a result of the decision by the U.S. Atomic Energy Commission to redirect research and development work on heavy water power reactors to reactors cooled with organic materials. For about one year, site personnel maintained the facility in a standby status, and then retired the reactor in place. In the early 1990s, DOE began planning to decommission HWCTR. Yet, in the face of new budget constraints, DOE deferred dismantlement and placed HWCTR in an extended surveillance and maintenance mode. The doors of the reactor facility were welded shut to protect workers and discourage intruders. In 2009 the $1.6 billion allocation from the ARRA to SRS for site footprint reduction at SRS reopened the doors to HWCTR - this time for final decommissioning. Alternative studies concluded that the most environmentally safe, cost effective option for final decommissioning was to remove the reactor vessel, both steam generators, and all equipment above grade including the dome. The transfer coffin, originally above grade, was to be placed in the cavity vacated by the reactor vessel and the remaining below grade spaces would be grouted. Once all above equipment

  5. In-reactor testing of ionic thermometers

    International Nuclear Information System (INIS)

    Ionic thermometers have been tested in a nuclear reactor with attention to the steepness of the ionic conductivity jump and the influence of a glass container on the accuracy of the temperature measurements. It was found that, at the neutron fluxes up to 1.5 x 1018 m-2 s-1 (thermal) and 3 x 1018 m-2 s-1 (fast) in a light water reactor, the change of conductivity jump slope is negligible or nil for an ionic thermometer filled by HgI2, i.e., at 256.0 +- 0.2 0C. The need to use boron-free glass was confirmed. The impact on the accuracy of the temperature point indication in a nuclear reactor core is discussed, as well as obvious inertness of the melting process mechanism to the intense irradiation field

  6. Tokamak Fusion Test Reactor Project

    International Nuclear Information System (INIS)

    Where the project's major contracts for buildings and equipment have experienced significant increases in cost, even though the contracts were for fixed prices, IG concluded that a major reason for the increases was the failure to adequately specify requirements prior to awarding the contracts. IG recommended that the Department should develop guidelines on what constitutes an adequate specification for a fixed-price contract and what controls should be placed over change orders. The Assistant Secretary for Management and Administration indicated in his comments to the draft report that an upcoming revision of the Accounting Practices and Procedures Handbook would establish controls over reallocations of construction funds to operating funds. Recommendations also propose that particular attention be given to ensuring that an agreed-upon plan and budget for completing the project is established, that a performance measurement system be implemented by the Princeton Plasma Physics Laboratory, and that improvements be made in the laboratory's Cost/Schedule Performance Report. Improvements are also needed in the quality assurance and safety programs of the Princeton Plasma Physics Laboratory. A number of recommendations from previous quality-assurance and safety reviews, performed by personnel from the Princeton Plasma Physics Laboratory and the Department of Energy, have not been implemented. Comments to the draft report also address these outstanding issues

  7. Experience with the generating plant at fast breeder test reactor

    International Nuclear Information System (INIS)

    Fast Breeder Test Reactor (FBTR) is a 40 MWth/13.2 MW(e) sodium cooled, loop type, mixed carbide-fuelled reactor. Its main aim is to gain experience in the design, construction and operation of fast reactors including sodium systems and generating systems and to serve as an irradiation facility for development of fuel and structural materials for future fast reactors. It achieved first criticality in Oct.1985 with Mark-I core (70 % PUC-30 % UC). FBTR heat transport system consists of two primary sodium loops, two secondary sodium loops and one common tertiary steam and water circuit. Heat generated in the reactor core is transported to the tertiary loop by primary and secondary sodium loops. The steam water system mainly consists of a once through steam generator, which produces super heated steam at a pressure of 120 bars and temperature of 480 degC, feed water system and condensate system. The steam produced is supplied to a condensing turbine. The turbine in turn is coupled to an alternator. The steam generator was put in service in Jan.1993 and turbine generator was synchronized to the grid in July 1997. The paper presents operating experience with generating plant consisting of steam water circuit, condensing turbine and its associated systems and the alternator, various modifications carried out to improve system reliability and availability and certain incidents taken place in the generating plant. (author)

  8. Safety system of reactor container

    International Nuclear Information System (INIS)

    The safety system of the present invention can shut down a BWR type reactor safely without operator's aid even upon occurrence of an abnormal state. Namely, a pressure/temperature measuring and controlling device is disposed to a dry well in the safety system of a reactor container incorporating a pressure vessel, a space between a dry wall and a wet well and a pressure suppression chamber. Operation signals sent from pipelines of an emergency reactor core cooling system delivered from the pressure vessel are inputted to the pressure/temperature measuring and controlling device. Output signals of the pressure/temperature measuring and controlling device are inputted to a spray device. With such procedures, when actuation of dry well spray is required upon loss of coolants accident, necessity for the actuation of the spray can be judged based on the pressure, temperature in the dry well, reactor water level and a state of operation and duration of abnormal state of other ECCS system using the pressure/temperature measuring and controlling device disposed in the dry well. If actuation of spray is required, the dry wall spray is automatically actuated to reduce pressure and temperature in the container. (I.S.)

  9. Spray system of reactor container

    International Nuclear Information System (INIS)

    A BWR type reactor comprises a pressure accumulation tank for temporary storing fire-extinguishing system water by making a branch between a fire-extinguishing pump and a reactor container spray, an inlet valve and an outlet valve of the pressure accumulation tank disposed on a pipeline at the upstream and downstream of the branch of the pipeline, a pressure accumulation tank pressure gauge, a valve controller for opening the outlet valve by the pressure high signal of the pressure gauge of the pressure accumulation tank and closing the outlet valve by a pressure low signal of the pressure gauge of the pressure accumulation tank, and a valve for isolating equipments described above from the fire-extinguishing system and the container spray system. The pressure accumulation tank is disposed to a water injection facility while using the fire-extinguishing system water in common for preventing failures due to overpressure or overheating of the reactor container upon occurrence of a severe accident. Gaseous radioactive materials in the dry well can be removed efficiently by the spray while maintaining the cooling performance for the reactor container by the intermittent spraying. Then, scrubbing effect of a pressure suppression pool can be improved thereby enabling to reduced radiation released to the environments. (N.H.)

  10. The foundation and analysis of polarized magnetic system's unified mathematical model in aerospace electromagnetic relay

    Institute of Scientific and Technical Information of China (English)

    2007-01-01

    Polarized magnetic system has a series of features, such as small volume, light weight, low power consumption, high sensitivity, quick movement and so on, widely used in the products of the military aerospace electromagnetic relay. The typical polarized magnetic system has mainly four structures and its simplified equivalent magnetic circuits model is the base of the design of the electromagnetic relay with permanent magnet. In the past, the analysis method that people used was difficult to build the unified mathematical models, which divided the work gap magnetic flux into "permanent magnet flux" and "electromagnetic flux". Based on the analysis method of the work gap magnetic voltage, this paper founds the unified mathematical model of the polarized magnetic system and divides the attractive torque into permanent magnet torque, polarized torque and electromagnetic torque through the energy balance formula. It analyses the influence of permanent magnet sizes on permanent magnet torque, polarized torque and electromagnetic torque through the energy balance formula and the conclusions can direct the design of aerospace electromagnetic relay with permanent magnet.

  11. Power Reactor Information System (PRIS)

    International Nuclear Information System (INIS)

    The IAEA has been collecting Operating Experience data for Nuclear Power Plants of the IAEA Member States since 1970. In order to facilitate an analysis of nuclear power plant performance as well as to produce relevant publications, all previously collected data supplied from the questionnaires were computerized in 1980 and the Power Reactor Information System was implemented. PRIS currently contains production records for the years up to and including 1990 and about 98% of the reactors-years operating experience in the world is contained in PRIS. (orig.)

  12. Innovations in dynamic test restraint systems

    Science.gov (United States)

    Fuld, Christopher J.

    1990-01-01

    Recent launch system development programs have led to a new generation of large scale dynamic tests. The variety of test scenarios share one common requirement: restrain and capture massive high velocity flight hardware with no structural damage. The Space Systems Lab of McDonnell Douglas developed a remarkably simple and cost effective approach to such testing using ripstitch energy absorbers adapted from the sport of technical rockclimbing. The proven system reliability of the capture system concept has led to a wide variety of applications in test system design and in aerospace hardware design.

  13. An Approach to Evaluate Precision and Inter-Laboratory Variability of Flammability Test Methods for Aerospace Materials

    Science.gov (United States)

    Hirsch, David; Beeson, Harold D.

    2005-01-01

    Materials selection for spacecraft is based on conventional flammability or ignition sensitivity acceptance tests. Current procedures for determining the inter-laboratory repeatability and reproducibility of aerospace materials flammability tests are not considering the dependence of data variability on test conditions and consequently attempts to characterize the precision of these methods were not successful. The inter-laboratory data variability is determined with tests conducted under arbitrary conditions, which consequently may not provide sufficient information to enable adequate determination of a method's precision. For evaluating the precision of NASA's flammability test methods, the protocol recommended includes selecting critical parameters and determining the 50% failure point by considering the specific failure criteria of each method using the critical parameter as a variable. Upon performing inter-laboratory round robin testing using this approach, the laboratories' performance could be evaluated by comparing the repeatability of the 50% failure point and/or the repeatability of critical conditions where the probabilities of passing and failing are unity, i.e., the transition zone repeatability. When a sufficient amount of data has been acquired with this method, an adequate estimation of precision of aerospace materials flammability test methods will be possible.

  14. Design of the Advanced Gas Reactor Fuel Experiments for Irradiation in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover

    2005-10-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight particle fuel tests in the Advanced Test Reactor (ATR) located at the newly formed Idaho National Laboratory (INL) to support development of the next generation Very High Temperature Reactor (VHTR) in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The experiments will be irradiated in an inert sweep gas atmosphere with on-line temperature monitoring and control combined with on-line fission product monitoring of the sweep gas. The final design phase has just been completed on the first experiment (AGR-1) in this series and the support systems and fission product monitoring system that will monitor and control the experiment during irradiation. This paper discusses the development of the experimental hardware and support system designs and the status of the experiment.

  15. Job-mix modeling and system analysis of an aerospace multiprocessor.

    Science.gov (United States)

    Mallach, E. G.

    1972-01-01

    An aerospace guidance computer organization, consisting of multiple processors and memory units attached to a central time-multiplexed data bus, is described. A job mix for this type of computer is obtained by analysis of Apollo mission programs. Multiprocessor performance is then analyzed using: 1) queuing theory, under certain 'limiting case' assumptions; 2) Markov process methods; and 3) system simulation. Results of the analyses indicate: 1) Markov process analysis is a useful and efficient predictor of simulation results; 2) efficient job execution is not seriously impaired even when the system is so overloaded that new jobs are inordinately delayed in starting; 3) job scheduling is significant in determining system performance; and 4) a system having many slow processors may or may not perform better than a system of equal power having few fast processors, but will not perform significantly worse.

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

    International Nuclear Information System (INIS)

    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

  17. Reactor limit control system

    International Nuclear Information System (INIS)

    The very extensive use of limitations in the operational field between protection system and closed-loop controls is an important feature of German understanding of operational safety. The design of limitations is based on very large activities in the computational field but mostly on the high level of the plant-wide own commissioning experience of a turnkey contractor. Limitations combine intelligence features of closed-loop controls with the high availability of protection systems. (orig.)

  18. Automatic power control system for 235 MWe atomic power reactor

    International Nuclear Information System (INIS)

    The paper highlights the essential features of the design, fabrication and testing of microprocessor based reactor power regulating system of Narora Atomic Power Plant (NAPP) and Kakrapar Atomic Power Plant (KAPP). The improved system design at KAPP employs the reactor power control based on neutron flux signal after correction. The control system responses have been presented and compared with the responses using a reactor functional simulator. A new fault tolerant reactor regulating system has been designed using a dual active and hot stand-by microprocessor system to improve operational reliability. (author). 1 ref., 8 figs

  19. Fabrication, testing, and qualification of reactor graphites

    International Nuclear Information System (INIS)

    The work performed under the HBK project for development and testing of reactor graphites could have recourse to results and experience already gained in Great Britain, in the F.R.G., the USA, and the Netherlands. The specific problems to be tackled by the HBK project activities result from the particularly exacting requirements with regard to behaviour under irradiation that are to be met by the graphite reflector for the THTR follower plant. From a great number of candidate graphites, selected for testing and evaluation, the extensive irradiation experiments revealed a variety of graphites best suited to the various tasks in mind, as defined by the operational conditions. The tests examined radiation-induced changes of linear dimension, E-module, thermal expansion, and heat conductivity, as well as radiation-induced creep and corrosion in reactor graphites under specified normal and under accident conditions. The work performed also includes tests for defining design criteria for reactor graphite components. The goals have been achieved, but further work will be necessary, as new requirements are taking shape in the course of current THTR follower plant development. (orig.)

  20. Validation of reactor core protection system

    International Nuclear Information System (INIS)

    Reactor COre Protection System (RCOPS), an advanced core protection calculator system, is a digitized one which provides core protection function based on two reactor core operation parameters, Departure from Nucleate Boiling Ratio (DNBR) and Local Power Density (LPD). It generates a reactor trip signal when the core condition exceeds the DNBR or LPD design limit. It consists of four independent channels adapted a two-out-of-four trip logic. System configuration, hardware platform and an improved algorithm of the newly designed core protection calculator system are described in this paper. One channel of RCOPS was implemented as a single channel facility for this R and D project where we performed final integration software testing. To implement custom function blocks, pSET is used. Software test is performed by two methods. The first method is a 'Software Module Test' and the second method is a 'Software Unit Test'. New features include improvement of core thermal margin through a revised on-line DNBR algorithm, resolution of the latching problem of control element assembly signal and addition of the pre-trip alarm generation. The change of the on-line DNBR calculation algorithm is considered to improve the DNBR net margin by 2.5%-3.3%. (author)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-09-01

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

  3. 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.

  4. Instrumentation to Enhance Advanced Test Reactor Irradiations

    International Nuclear Information System (INIS)

    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.

  5. Simulation of the TREAT-Upgrade Automatic Reactor Control System

    International Nuclear Information System (INIS)

    This paper describes the design of the Automatic Reactor Control System (ARCS) for the Transient Reactor Test Facility (TREAT) Upgrade. A simulation was used to facilitate the ARCS design and to completely test and verify its operation before installation at the TREAT facility

  6. A Conceptual Aerospace Vehicle Structural System Modeling, Analysis and Design Process

    Science.gov (United States)

    Mukhopadhyay, Vivek

    2007-01-01

    A process for aerospace structural concept analysis and design is presented, with examples of a blended-wing-body fuselage, a multi-bubble fuselage concept, a notional crew exploration vehicle, and a high altitude long endurance aircraft. Aerospace vehicle structures must withstand all anticipated mission loads, yet must be designed to have optimal structural weight with the required safety margins. For a viable systems study of advanced concepts, these conflicting requirements must be imposed and analyzed early in the conceptual design cycle, preferably with a high degree of fidelity. In this design process, integrated multidisciplinary analysis tools are used in a collaborative engineering environment. First, parametric solid and surface models including the internal structural layout are developed for detailed finite element analyses. Multiple design scenarios are generated for analyzing several structural configurations and material alternatives. The structural stress, deflection, strain, and margins of safety distributions are visualized and the design is improved. Over several design cycles, the refined vehicle parts and assembly models are generated. The accumulated design data is used for the structural mass comparison and concept ranking. The present application focus on the blended-wing-body vehicle structure and advanced composite material are also discussed.

  7. Demonstration and verification of on-line core evaluation system in the startup test of heavy water reactor, FUGENug

    International Nuclear Information System (INIS)

    Validity of FUGEN on-line core performance evaluation system, ATROPOS, was verified through the startup test. The main verification results are as follows; (1) The estimation error of the power distribution was less than 3% at any power levels. (2) The differences were about 4% between channel flow rate converted from the measured pressure drop and that estimated by ATROPOS. (3) The uncertainties of the thermal operations limits were evaluated to be about 6%. (4) The core thermal power can be predicted accurately and the prediction depends on the Dopper reactivity coefficient to be used. (5) The prediction errors of the thermal operation limits were less than 4% within the specified region. (6) The whole core power distribution was predicted with about 6% error excluding core peripheral segments. (orig.)

  8. Hydraulic characteristics of the N Reactor core and reactor cooling system

    International Nuclear Information System (INIS)

    In conjunction with the NUSAR program, the need was recognized for well substantiated pressure drop correlations for the N Reactor core to support in-depth safety analysis consistent with currently-available technology. Additionally, it was considered desirable to reconfirm the hydraulic characteristics of the reactor coolant system in the light of improved understanding of the hydraulic features of the current reactor fuel loading. The report summarizes the results of laboratory tests and analysis accomplished to meet the above objectives

  9. Fast Breeder Test Reactor: 15 years of operating experience

    International Nuclear Information System (INIS)

    Fast Breeder Test Reactor (FBTR) is a 40 MWt/13.2 MWe sodium cooled, loop type, mixed carbide-fuelled reactor. Its main aim is to gain experience in the design, construction and operation of fast reactors including sodium systems and to serve as an irradiation facility for development of fuel and structural materials for future fast reactors. It achieved first criticality in Oct 1985 with Mark I core (70% PuC - 30% UC). Steam generator was put in service in Jan 1993 and power was raised to 10.5 MWt in Dec 1993. Turbine generator was synchronized to the grid in Jul 1997. The indigenously developed mixed carbide fuel has achieved a peak burn up of 88,000 MWd/t till now at a linear heat rating of 320 W/cm and reactor power of 13.4 MWt without any fuel-clad failure. The paper presents operating and decontamination experience, performance of fuel, steam generator and sodium circuits, certain unusual occurrences encountered by the plant and various improvements carried out in reactor systems to enhance plant availability. (author)

  10. Development of a Dynamically Configurable, Object-Oriented Framework for Distributed, Multi-modal Computational Aerospace Systems Simulation

    Science.gov (United States)

    Afjeh, Abdollah A.; Reed, John A.

    2003-01-01

    The following reports are presented on this project:A first year progress report on: Development of a Dynamically Configurable,Object-Oriented Framework for Distributed, Multi-modal Computational Aerospace Systems Simulation; A second year progress report on: Development of a Dynamically Configurable, Object-Oriented Framework for Distributed, Multi-modal Computational Aerospace Systems Simulation; An Extensible, Interchangeable and Sharable Database Model for Improving Multidisciplinary Aircraft Design; Interactive, Secure Web-enabled Aircraft Engine Simulation Using XML Databinding Integration; and Improving the Aircraft Design Process Using Web-based Modeling and Simulation.

  11. Design and Testing of a Labview- Controlled Catalytic Packed- Bed Reactor System For Production of Hydrocarbon Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Street, J.; Yu, F.; Warnock, J.; Wooten, J.; Columbus, E.; White, M. G.

    2012-05-01

    Gasified woody biomass (producer gas) was converted over a Mo/H+ZSM-5 catalyst to produce gasolinerange hydrocarbons. The effect of contaminants in the producer gas showed that key retardants in the system included ammonia and oxygen. The production of gasoline-range hydrocarbons derived from producer gas was studied and compared with gasoline-range hydrocarbon production from two control syngas mixes. Certain mole ratios of syngas mixes were introduced into the system to evaluate whether or not the heat created from the exothermic reaction could be properly controlled. Contaminant-free syngas was used to determine hydrocarbon production with similar mole values of the producer gas from the gasifier. Contaminant-free syngas was also used to test an ideal contaminant-free synthesis gas situation to mimic our particular downdraft gasifier. Producer gas was used in this study to determine the feasibility of using producer gas to create gasoline-range hydrocarbons on an industrial scale using a specific Mo/H+ZSM-5 catalyst. It was determined that after removing the ammonia, other contaminants poisoned the catalyst and retarded the hydrocarbon production process as well.

  12. Reactor vessel annealing system

    Science.gov (United States)

    Miller, Phillip E.; Katz, Leonoard R.; Nath, Raymond J.; Blaushild, Ronald M.; Tatch, Michael D.; Kordalski, Frank J.; Wykstra, Donald T.; Kavalkovich, William M.

    1991-01-01

    A system for annealing a vessel (14) in situ by heating the vessel (14) to a defined temperature, composed of: an electrically operated heater assembly (10) insertable into the vessel (14) for heating the vessel (14) to the defined temperature; temperature monitoring components positioned relative to the heater assembly (10) for monitoring the temperature of the vessel (14); a controllable electric power supply unit (32-60) for supplying electric power required by the heater assembly (10); a control unit (80-86) for controlling the power supplied by the power supply unit (32-60); a first vehicle (2) containing the power supply unit (32-60); a second vehicle (4) containing the control unit (80-86); power conductors (18,22) connectable between the power supply unit (32-60) and the heater unit (10) for delivering the power supplied by the power supply unit (32-60) to the heater assembly (10); signal conductors (20,24) connectable between the temperature monitoring components and the control unit (80-86) for delivering temperature indicating signals from the temperature monitoring components to the control unit (80-86); and control conductors (8) connectable between the control unit (80-86) and the power supply unit (32-60) for delivering to the power supply unit (32-60) control signals for controlling the level of power supplied by the power supply unit (32-60) to the heater assembly (10).

  13. Performance tests for integral reactor nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-02-15

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

  14. Performance tests for integral reactor nuclear fuel

    International Nuclear Information System (INIS)

    An integral type reactor SMART plans to utilize metallic Zr-U fuel which is Zr-based alloy with 34∼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

  15. Irradiation test of diagnostic components for ITER application in a fission reactor, Japan Materials Testing Reactor

    International Nuclear Information System (INIS)

    Radiation effects on components and materials will be one of the most serious technological issues in fusion systems realizing burning plasmas. Especially, diagnostic components, which should play crucial roles to control plasmas and to understand physics of burning plasmas, will be exposed to high-flux neutrons and gamma-rays. Dynamic radiation effects will affects performance of components substantially from beginning of exposure to radiation environments, and accumulated radiation effects will gradually degrade their functioning abilities in the course of their services. High-power-density fission reactors will be only realistic tools to simulate the irradiation environments expected in burning-plasma fusion machines such as the ITER, at present. Some key diagnostic components, namely magnetic coils, bolometers, and optical fibers, were irradiation-tested in a fission reactor, JMTR, to evaluate their performances under heavy irradiation environments. Results indicate that the ITER-relevant diagnostic components could be developed in time, though there are still some technological problems to overcome. (author)

  16. Irradiation test of diagnostic components for ITER application in a fission reactor, Japan Materials Testing Reactor

    International Nuclear Information System (INIS)

    Radiation effects on components and materials will be one of the most serious technological issues in fusion systems realizing burning plasmas. Especially, diagnostic components, which should play crucial roles to control plasmas and to understand physics of burning plasmas, will be exposed to high-flux neutrons and gamma-rays. Dynamics radiation effects will affects performance of components substantially from beginning of exposure to radiation environments, and accumulated radiation effects will gradually degrade their functioning abilities in the course of their services. High-power-density fission reactors will be only realistic tools to simulate the irradiation environments expected in burning-plasma fusion machines such as the ITER, at present. Some key diagnostic components, namely magnetic coils, bolometers, and optical fibers, were irradiation-tested in a fission reactor, JMTR, to evaluate their performances under heavy irradiation environments. Results indicate that the ITER-relevant diagnostic components could be developed in time, though there are still some technological problems to overcome. (author)

  17. Systemization of Design and Analysis Technology for Advanced Reactor

    International Nuclear Information System (INIS)

    The present study is performed to establish the base for the license application of the original technology by systemization and enhancement of the technology that is indispensable for the design and analysis of the advanced reactors including integral reactors. Technical reports and topical reports are prepared for this purpose on some important design/analysis methodology; design and analysis computer programs, structural integrity evaluation of main components and structures, digital I and C systems and man-machine interface design. PPS design concept is complemented reflecting typical safety analysis results. And test plans and requirements are developed for the verification of the advanced reactor technology. Moreover, studies are performed to draw up plans to apply to current or advanced power reactors the original technologies or base technologies such as patents, computer programs, test results, design concepts of the systems and components of the advanced reactors. Finally, pending issues are studied of the advanced reactors to improve the economics and technology realization

  18. Developing the MAPLE materials test reactor concept

    International Nuclear Information System (INIS)

    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

  19. TRIGA reactor dynamics: Frequency response tests

    International Nuclear Information System (INIS)

    In this work, the results of frequency response tests conducted on ITU TRIGA Reactor are presented. To conduct the experiments, a special 'micro control rod' and its submersible stepping-motor drive mechanism was designed and constructed. The experiments cover a frequency range of 0.002 - 2 Hz., and 0.02, 4, 200 kW nominal power levels. Zero-power and at-power reactivity to % power transfer functions are presented as gain, and phase shift vs. frequency diagrams. Low power response is in close agreement with the point reactor zero-power transfer function. Response at 200 kW is studied with the help of a Nyquist diagram, and found to be stable. An elaboration on the main features of the feedback mechanism is also given. Power to reactivity feedback was measured to be just about 1.5 cent / % power change. (authors)

  20. Development of telerobotic systems for reactor decommissioning, (3)

    International Nuclear Information System (INIS)

    This paper describes the telerobotic system for reactor decommissioning in the scope of engineering demonstration of dismantling radioactive reactor internals of an experimental boiling water power reactor JPDR. The total system consists of a telerobotic manipulator system equipped with a multi-functional amphibious slave manipulator with a load capacity of 25 daN, a chain-driven transport system, and a computer-assisted monitoring and control system. Preceding to the application of the telerobotic system to actual dismantling operation, a mockup test was performed of dismantling the simulated reactor internals of actual-size by the method of underwater plasma arc cutting in order to study the performance of the telerobotic system in a realistic environment. The system was then successfully applied to dismantling the actual reactor internals according to the JPDR decommissioning program. (author)

  1. Pneumatic transport systems for TRIGA reactors

    International Nuclear Information System (INIS)

    Main parameters and advantages of pneumatically operated systems, primarily those operated by gas pressure are discussed. The special irradiation ends for the TRIGA reactor are described. To give some idea of the complexity of some modern systems, the author presents the large system currently operating at the National Bureau of Standards in Washington. In this system, 13 stations are located throughout the radiochemistry laboratories and three irradiation ends are located in the reactor, which is a 14-megawatt unit. The system incorporates practically every fail-safe device possible, including ball valves located on all capsule lines entering the reactor area, designed to close automatically in the event of a reactor scram, and at that time capsules within the reactor would be diverted by means of switches located on the inside of the reactor wall. The whole system is under final control of a permission control panel located in the reactor control room. Many other safety accessories of the system are described

  2. Reactor technology: power conversion systems and reactor operation and maintenance

    International Nuclear Information System (INIS)

    The use of advanced fuels permits the use of coolants (organic, high pressure helium) that result in power conversion systems with good thermal efficiency and relatively low cost. Water coolant would significantly reduce thermal efficiency, while lithium and salt coolants, which have been proposed for DT reactors, will have comparable power conversion efficiencies, but will probably be significantly more expensive. Helium cooled blankets with direct gas turbine power conversion cycles can also be used with DT reactors, but activation problems will be more severe, and the portion of blanket power in the metallic structure will probably not be available for the direct cycle, because of temperature limitations. A very important potential advantage of advanced fuel reactors over DT fusion reactors is the possibility of easier blanket maintenance and reduced down time for replacement. If unexpected leaks occur, in most cases the leaking circuit can be shut off and a redundant cooling curcuit will take over the thermal load. With the D-He3 reactor, it appears practical to do this while the reactor is operating, as long as the leak is small enough not to shut down the reactor. Redundancy for Cat-D reactors has not been explored in detail, but appears feasible in principle. The idea of mobile units operating in the reactor chamber for service and maintenance of radioactive elements is explored

  3. Synergistic effects of zinc borate and aluminium trihydroxide on flammability behaviour of aerospace epoxy system

    Directory of Open Access Journals (Sweden)

    2009-06-01

    Full Text Available The flame retardancy of mono-component epoxy resin (RTM6, widely used for aerospace composites, treated with zinc borate (ZB, aluminium trihydroxide (ATH and their mixtures at different concentrations have been investigated by morphological and thermal characterization. Cone calorimeter data reveal that combustion behaviour, heat release rate peak (PHRR and heat release rate average (HRR Average of RTM6 resin decrease substantially when synergistic effects of zinc borate and aluminium trihydroxide intervene. Thermogravimetric (TGA results and analysis of the residue show that addition higher than 20% w/w of ZB, ATH, and their mixture greatly promotes RTM6 char formation acting as a barrier layer for the fire development. Depending upon the different used flame additives, SEM micrographs indicate that the morphology of residual char could vary from a compact amalgam-like structure, for the RTM6+ZB system, to a granular structure, characterized by very small particles of degraded resin and additive for the ATH.

  4. Natural convection test in Phenix reactor and associated CATHARE calculation

    International Nuclear Information System (INIS)

    The Phenix sodium cooled fast reactor started operation in 1973 and was stopped in 2009. Before the reactor was definitively stopped, final tests were performed, including a natural convection test in the primary circuit. One objective of this natural convection test in Phenix reactor is the qualification of plant dynamic codes as CATHARE code for future safety studies. The paper firstly describes the Phenix reactor primary circuit. The initial test conditions and the detailed transient scenario are presented. Then, the CATHARE modelling of the Phenix primary circuit is described. The whole transient scenario is calculated, including the nominal state, the steam generators dry out, the scram, the onset of natural convection in the primary circuit and the natural convection phases. The CATHARE calculations are compared to the Phenix measurements. A particular attention is paid to the significant decrease of the core power before the scram. Then, the evolution of main components inlet and outlet temperatures is compared. The need of coupling a system code with a CFD code to model the 3D behaviour of large pools is pointed out. This work is in progress. (author)

  5. A Model-Based Approach to Engineering Behavior of Complex Aerospace Systems

    Science.gov (United States)

    Ingham, Michel; Day, John; Donahue, Kenneth; Kadesch, Alex; Kennedy, Andrew; Khan, Mohammed Omair; Post, Ethan; Standley, Shaun

    2012-01-01

    One of the most challenging yet poorly defined aspects of engineering a complex aerospace system is behavior engineering, including definition, specification, design, implementation, and verification and validation of the system's behaviors. This is especially true for behaviors of highly autonomous and intelligent systems. Behavior engineering is more of an art than a science. As a process it is generally ad-hoc, poorly specified, and inconsistently applied from one project to the next. It uses largely informal representations, and results in system behavior being documented in a wide variety of disparate documents. To address this problem, JPL has undertaken a pilot project to apply its institutional capabilities in Model-Based Systems Engineering to the challenge of specifying complex spacecraft system behavior. This paper describes the results of the work in progress on this project. In particular, we discuss our approach to modeling spacecraft behavior including 1) requirements and design flowdown from system-level to subsystem-level, 2) patterns for behavior decomposition, 3) allocation of behaviors to physical elements in the system, and 4) patterns for capturing V&V activities associated with behavioral requirements. We provide examples of interesting behavior specification patterns, and discuss findings from the pilot project.

  6. 75 FR 3141 - Airworthiness Directives; AVOX Systems and B/E Aerospace Oxygen Cylinder Assemblies, as Installed...

    Science.gov (United States)

    2010-01-20

    ...-16049 (74 FR 63063, December 2, 2009). That AD applies to certain AVOX Systems and B/E Aerospace oxygen... ``significant rule'' under the DOT Regulatory Policies and Procedures (44 FR 11034, February 26, 1979); and 3....13 by removing amendment 39-16049 (74 FR 63063, December 2, 2009) and adding the following new...

  7. The decommissioning of the KEMA suspension test reactor

    International Nuclear Information System (INIS)

    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 UO2/ ThO2 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-06-30

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

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

    International Nuclear Information System (INIS)

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

  10. Nuclear reactor internals construction and failed fuel rod detection system

    International Nuclear Information System (INIS)

    A system is provided for determining during operation of a nuclear reactor having fluid pressure operated control rod mechanisms the exact location of a fuel assembly with a defective fuel rod. The construction of the reactor internals is simplified in a manner to facilitate the testing for defective fuel rods and the reduce the cost of producing the upper internals of the reactor. 13 claims, 10 drawing figures

  11. System-integrated modular advanced reactor (SMART)

    International Nuclear Information System (INIS)

    coolant pump (MCP) and pressurizer (PZR), are installed in a single reactor vessel assembly (RVA). The conceptual and basic designs of SMART with a desalination system were completed in March of 1999 and March of 2002, respectively. SMART development has been conducted under the nuclear research and development programme supported by the Ministry of Science and Technology (MOST) of the Republic of Korea and thus KAERI and MOST are the principal stakeholders. The SMART design focuses on the enhancement of safety and improvement of the reliability as well as the economics. For these purposes, highly advanced design features enhancing the safety, reliability, performance, and operability were introduced into the SMART design. Advanced design features should be proven or qualified by experience, testing, or analysis and, if possible, the equipment should be designed according to approved standards. Some fundamental thermal-hydraulic experiments were carried out during the design concept development to assure the fundamental behaviour of major concepts of the SMART systems. Various thermal-hydraulic and mechanical tests are in progress and planned. In addition, overall SMART performance will be demonstrated through the SMART pilot plant construction and operation

  12. Propose Reactor Control and Monitoring System for RTP

    International Nuclear Information System (INIS)

    Reactor control and monitoring system is a one of the important features used in reactor. The control and monitoring must come together to provide safety, excellent performance and reliable in nuclear reactor technology application. Objectives of this technical paper are to design and propose reactor control system and reactor monitoring system in Research Reactor (RTP) for Reactor Upgrading Project. (author)

  13. SRS reactor stack plume marking tests

    International Nuclear Information System (INIS)

    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

  14. Computerized reactor pressure vessel materials information system

    International Nuclear Information System (INIS)

    A computerized information system for storage and retrieval of reactor pressure vessel materials data was established, as part of Task Action Plan A-11, Reactor Vessel Materials Toughness. Data stored in the system are necessary for evaluating the resistance of reactor pressure vessels to flaw-induced fracture. This report includes (1) a description of the information system; (2) guidance on accessing the system; and (3) a user's manual for the system

  15. 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 ...

  16. Reactor vessel model flow tests for 145-fuel assembly core

    International Nuclear Information System (INIS)

    Hydraulic tests on a one-sixth-scale model of a two-loop pressurized water reactor with 145 fuel assemblies are described. Core inlet and outlet flow distributions and reactor vessel pressure drop were investigated. The core inlet flow distribution was developed to be independent of the flow conditions in the inlet annulus. A flow distribution system, consisting of several flow splitters in the inlet annulus and a spherical plate flow distributor in the lower head region, was developed to obtain a symmetric and stable core inlet flow distribution. A minimum core inlet flow factor of 0.99 was established in the core. Reactor vessel unrecoverable pressure drops were measured on the model to predict losses that will occur in the prototype

  17. Special tests and destructive physical analyses as used by the Aerospace Corporation with nickel-hydrogen cells

    Science.gov (United States)

    Zimmerman, A. H.; Quinzio, M. V.; Thaller, L. H.

    1992-01-01

    The destructive physical analysis (DPA) of electrochemical devices is an important part of the overall test. Specific tests were developed to investigate the degradation mode or the failure mechanism that surfaces during the course of a cell being assembled, acceptance tested, and life-cycle tested. The tests that have been developed are peculiar to the cell chemistry under investigation. Tests are often developed by an individual or group of researchers as a result of their particular interest in an unresolved failure mechanism or degradation mode. A series of production, operational, and storage issues that were addressed by the Electrochemistry Group at The Aerospace Corporation are addressed. As a result of these investigations, as well as associated research studies carried out to develop a clearer understanding of the nickel oxyhydroxide electrode, a series of unique and useful specialized tests were developed. Some of these special tests were assembled to describe the methods that were found to be particularly useful in resolving a wide spectrum of manufacturing, operational, and storage issues related to nickel-hydrogen cells. The general methodology of these tests is given here with references listed to provide the reader with a more detailed understanding of the tests. The tests are classified according to the sequencing, starting with the impregnation of the nickel plaque material and culminating with the storage of completed cells. The details of the wet chemical procedures that were found to be useful because of their accuracy and reproducibility are given. The equations used to make the appropriate calculations are listed.

  18. Functional systems of a pressurized water reactor

    International Nuclear Information System (INIS)

    The main topics, discussed in the present paper, are: - Principle design of the reactor coolant system - reactor pressure vessel with internals - containment design - residual heat removal and emergency cooling systems - nuclear component cooling systems - emergency feed water systems - plant electric power supply system. (orig./RW)

  19. Emerging Needs for Pervasive Passive Wireless Sensor Networks on Aerospace Vehicles

    Science.gov (United States)

    Wilson, William C.; Juarez, Peter D.

    2014-01-01

    NASA is investigating passive wireless sensor technology to reduce instrumentation mass and volume in ground testing, air flight, and space exploration applications. Vehicle health monitoring systems (VHMS) are desired on all aerospace programs to ensure the safety of the crew and the vehicles. Pervasive passive wireless sensor networks facilitate VHMS on aerospace vehicles. Future wireless sensor networks on board aerospace vehicles will be heterogeneous and will require active and passive network systems. Since much has been published on active wireless sensor networks, this work will focus on the need for passive wireless sensor networks on aerospace vehicles. Several passive wireless technologies such as microelectromechanical systems MEMS, SAW, backscatter, and chipless RFID techniques, have all shown potential to meet the pervasive sensing needs for aerospace VHMS applications. A SAW VHMS application will be presented. In addition, application areas including ground testing, hypersonic aircraft and spacecraft will be explored along with some of the harsh environments found in aerospace applications.

  20. Piping installation for reactor heavy water system

    International Nuclear Information System (INIS)

    Characteristics and main installation steps for the piping of the reactor heavy water loop system were introduced in this paper. According to the system design, equipment accommodation and spot management, main issues with effect on the quality and schedule of pipeline installation were analyzed. Accordingly, some solutions were put forward, which included: work allocation should be made clear in documents; installation preparative such as design checkup and technology communication should be prepared completely; requirements of system cleaning, test items in every experiment, inspection in work and equipment maintenance should be considered in the system design; perfect documents distribution system and stock plan should be built; technology requirements and quality assurance should be claimed in contracts; quality should be controlled by way of external evidence, inspection in manufactory, exterior quality assurance examination, and test during consignment; series of management procedure should be established in detail. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-03-01

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

  2. Thermal Hydraulic Integral Effect Tests for Pressurized Water Reactors

    International Nuclear Information System (INIS)

    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∼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∼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

  3. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-06-30

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

  5. Verification tests performed for development of an integral type reactor

    International Nuclear Information System (INIS)

    SMART is an integral type reactor with innovative design features aimed at achieving a highly enhanced safety and improved economics. The SMART design is based on proven reactor design technologies with the use of new advanced design features. Most of the design features implemented into the SMART have been proven, however the advanced design features implemented into the SMART should be proven by testing. Various thermal hydraulic experiments have been carried out and also planned to assure the fundamental behavior of major concepts of the SMART and to prove the performance of the systems with new innovative technologies. This paper describes the thermal hydraulic test program for the SMART development and briefly discusses the typical test results. (author)

  6. Properties of ultrasonic testing systems

    International Nuclear Information System (INIS)

    For a long time, ultrasonic testing of reactor components and plants whose safety had to meet high demands, lacked definitions of the required properties of the ultrasonic testing system. The standard draft DIN 25 450 states demands on the ultrasonic testing unit and the test heads and recommends measuring methods to determine their properties. With test units and test heads meeting the demands of the draft a better reproducibility of the test is obtained than before; the improved test statement results in an increased safety during production and operation of components and plants. (orig./HP)

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

    Science.gov (United States)

    2013-10-24

    ...) System, which support functions for alternate water injection during station blackout. ] II. Additional..., was published in the Federal Register on June 15, 2012 (77 FR 36014), for a 60-day public...

  8. Development of Advanced Verification and Validation Procedures and Tools for the Certification of Learning Systems in Aerospace Applications

    Science.gov (United States)

    Jacklin, Stephen; Schumann, Johann; Gupta, Pramod; Richard, Michael; Guenther, Kurt; Soares, Fola

    2005-01-01

    Adaptive control technologies that incorporate learning algorithms have been proposed to enable automatic flight control and vehicle recovery, autonomous flight, and to maintain vehicle performance in the face of unknown, changing, or poorly defined operating environments. In order for adaptive control systems to be used in safety-critical aerospace applications, they must be proven to be highly safe and reliable. Rigorous methods for adaptive software verification and validation must be developed to ensure that control system software failures will not occur. Of central importance in this regard is the need to establish reliable methods that guarantee convergent learning, rapid convergence (learning) rate, and algorithm stability. This paper presents the major problems of adaptive control systems that use learning to improve performance. The paper then presents the major procedures and tools presently developed or currently being developed to enable the verification, validation, and ultimate certification of these adaptive control systems. These technologies include the application of automated program analysis methods, techniques to improve the learning process, analytical methods to verify stability, methods to automatically synthesize code, simulation and test methods, and tools to provide on-line software assurance.

  9. Acoustic emission for on-line reactor monitoring: results of intermediate vessel test monitoring and reactor hot functional

    International Nuclear Information System (INIS)

    The objective of the acoustic emission (AE)/flaw characterization program presented is to develop use of the AE method on a continuous basis during operation and during hydrotest, to detect and analyze flow growth in reactor pressure vessels and primary piping. The program scope is described by three primary areas of effort: develop a method to identify crack growth AE signals; develop a relationship between measured AE and crack growth; demonstrate the total concept through off-reactor vessel tests; and, on-reactor monitoring. The laboratory speciments used to determine fundamental feasibility of program objectives were ASTM A533 B, Class 1 steel. The ZB-1 vessel test is described, and the results are presented. Reactor hot functional testing was done on the Watts Bar Unit 1. Evidence shows that AE from cracking in inaccessible parts of the reactor system such as the vessel beltline should be detectable

  10. Conceptual design study of a scyllac fusion test reactor

    International Nuclear Information System (INIS)

    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

  11. Conceptual design study of a scyllac fusion test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Thomassen, K.I. (comp.)

    1975-07-01

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

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

    International Nuclear Information System (INIS)

    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)

  13. Corrosion of spent Advanced Test Reactor fuel

    International Nuclear Information System (INIS)

    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

  14. Modeling of Control System of Tajoura Reactor Using Apros

    International Nuclear Information System (INIS)

    This paper is a part of a project that simulated, using the Apros (Advanced Process Simulator) software, the Tajoura research reactor (TRR). This part of the project aimed at the modeling of the control systems of the reactor, where important control parameters, as detailed in the following section, had been studied and simulated by Apros: These parameters are tested for various values, and it was made sure the reactor is responding properly to the change of different conditions. In particular, the reactor scrams automatically whenever a scram condition is reached via any of the above parameters. In addition, it is possible to manually scram the reactor when needed, as is the case in the simulated reactor. In the end of the paper, we give many plots derived from Apros, that illustrate the work of the system and the response to different parameter changes. (author

  15. A Diagnostic Approach for Electro-Mechanical Actuators in Aerospace Systems

    Data.gov (United States)

    National Aeronautics and Space Administration — Electro-mechanical actuators (EMA) are finding increasing use in aerospace applications, especially with the trend towards all all-electric aircraft and spacecraft...

  16. Titanium production for aerospace applications

    OpenAIRE

    Vinicius A. R. Henriques

    2009-01-01

    Titanium parts are ideally suited for advanced aerospace systems because of their unique combination of high specific strength at both room temperature and moderately elevated temperature, in addition to excellent general corrosion resistance. The objective of this work is to present a review of titanium metallurgy focused on aerospace applications, including developments in the Brazilian production of titanium aimed at aerospace applications. The article includes an account of the evolution ...

  17. Trends in large-scale testing of reactor structures

    International Nuclear Information System (INIS)

    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)

  18. Emergency reactor core cooling system of BWR type reactor

    International Nuclear Information System (INIS)

    The present invention provides an emergency reactor core cooling system which can reduce a capacity of a power source required upon occurrence of emergency, extending an start-up time of an emergency reactor core cooling system (ECCA) to provide a plant endurable to a common factor accident and can provide time margin up to the start-up time. Namely, the system of the present invention comprises a division I equipped with an isolation condenser (IC), an after-heat removing system (low pressure system)(LPFL/RHR) and an emergency gas turbine generator (GT), a division II equipped with a diesel driving water injection system (high pressure system)(HDIS), LPFL/RHR, and GT, and a division III equipped with a reactor isolation time cooling system (high pressure system)(ARCIC), LPFL/RHR and GT. With such a constitution, since the IC, HDIS and ARCIC are used in combination as a high pressure system, an electromotive pump required to be operated upon high pressure state can be saved. In addition, if a static reactor cooling system (PCCS) is adopted and is provided with a back-up function for LPFL/RHR with respect to heat removal of the container upon occurrence of an accident, the countermeasure for occurrence of severe accidents can be enhanced. (I.S.)

  19. Development of a core follow calculational system for research reactors

    International Nuclear Information System (INIS)

    Over the last few years a comprehensive Pressurized Water Reactor and Materials Testing Reactor core analysis code system based on modern reactor physics methods has been under development by the Atomic Energy Corporation of South Africa. This system, known as OSCAR-3, will incorporate a customized graphical user interface and data management system to ensure user-friendliness and good quality control. The system has now reached the stage of development where it can be used for practical MTR core analyses. This paper describes the current capabilities of the components of the OSCAR-3 package, their integration within the package, and outlines future developments. 10 refs., 1 tab., 1 fig

  20. Static and dynamic performance tests of nuclear powered ship Mutsu reactor (report on nuclear ship Mutsu power-up tests)

    Energy Technology Data Exchange (ETDEWEB)

    Ishida, Toshihisa; Kusunoki, Tsuyoshi; Ochiai, Masa-aki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Tanaka, Yoshimi; Inoue, Kimio; Yao, Toshiaki; Kamai, Satoshi; Kitamura, Toshikatsu

    1992-08-01

    The power-up tests of the Mutsu reactor were performed from March 29th 1990 to December 14th. The tests were divided into six phases: The tests Phase 0 and Phase 1 were done in the state that the ship was moored at the quay of Sekinehama port in March and April; The tests Phase 2, Phase 3, Phase 4, and Phase 5 were done on the Pacific Ocean from July to December. Present report describes the test results on the static and dynamic plant performance. On static plant performance tests, there are 13 test items including measurements of primary system heat balance at low and high power levels, a virgin run of feed water pump with SG steam, a change-over test of steam supply of auxiliary boiler to SG. On the dynamic plant performance, there are 11 test items including a test of reactor power auto-control system, a test of main feed water auto-control system, a test of small load variation, a load increasing test, a turbine trip test, tests of ahead and astern maneuvering, a test of single loop operation, and a reactor scram test. The reactor power for each item`s test was increased step by step from zero power to the goal of rated power of 100 %, 36 MWt. In order to confirm proper reactor system performance, criteria were laid down for the static and dynamic tests: for example, (1) reactor scram shall not occur, (2) pressurizer relief valve and steam generator safety valve shall not work, and (3) after the transients reactor systems shall become the steady state without manual adjustment of the reactor control system. The test results satisfied these criteria and some of test data showed that reactor had much more margin in any performance for design. It is verified, therefore, that the Mutsu reactor systems have adequate performances as a marine reactor and that one is capable to respond smoothly and safely to the load of ship`s demand. (author).

  1. Static and dynamic performance tests of nuclear powered ship Mutsu reactor (report on nuclear ship Mutsu power-up tests)

    Energy Technology Data Exchange (ETDEWEB)

    Ishida, Toshihisa; Kusunoki, Tsuyoshi; Ochiai, Masa-aki (Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment); Tanaka, Yoshimi; Inoue, Kimio; Yao, Toshiaki; Kamai, Satoshi; Kitamura, Toshikatsu.

    1992-08-01

    The power-up tests of the Mutsu reactor were performed from March 29th 1990 to December 14th. The tests were divided into six phases: The tests Phase 0 and Phase 1 were done in the state that the ship was moored at the quay of Sekinehama port in March and April; The tests Phase 2, Phase 3, Phase 4, and Phase 5 were done on the Pacific Ocean from July to December. Present report describes the test results on the static and dynamic plant performance. On static plant performance tests, there are 13 test items including measurements of primary system heat balance at low and high power levels, a virgin run of feed water pump with SG steam, a change-over test of steam supply of auxiliary boiler to SG. On the dynamic plant performance, there are 11 test items including a test of reactor power auto-control system, a test of main feed water auto-control system, a test of small load variation, a load increasing test, a turbine trip test, tests of ahead and astern maneuvering, a test of single loop operation, and a reactor scram test. The reactor power for each item's test was increased step by step from zero power to the goal of rated power of 100 %, 36 MWt. In order to confirm proper reactor system performance, criteria were laid down for the static and dynamic tests: for example, (1) reactor scram shall not occur, (2) pressurizer relief valve and steam generator safety valve shall not work, and (3) after the transients reactor systems shall become the steady state without manual adjustment of the reactor control system. The test results satisfied these criteria and some of test data showed that reactor had much more margin in any performance for design. It is verified, therefore, that the Mutsu reactor systems have adequate performances as a marine reactor and that one is capable to respond smoothly and safely to the load of ship's demand. (author).

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-01

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

  3. Conceptual design of a Bitter-magnet toroidal-field system for the ZEPHYR Ignition Test Reactor

    International Nuclear Information System (INIS)

    The following problems are described and discussed: (1) parametric studies - these studies examine among other things the interdependence of throat stresses, plasma parameters (margins of ignition) and stored energy. The latter is a measure of cost and is minimized in the present design; (2) magnet configuration - the shape of the plates are considered in detail including standard turns, turns located at beam ports, diagnostic and closure flanges; (3) ripple computation - this section describes the codes by which ripple is computed; (4) field diffusion and nuclear heating - the effect of magnetic field diffusion on heating is considered along with neutron heating. Current, field and temperature profiles are computed; (5) finite element analysis - the two and three dimensional finite element codes are described and the results discussed in detail; (6) structures engineering - this considers the calculation of critical stresses due to toroidal and overturning forces and discusses the method of constraint of these forces. The Materials Testing Program is also discussed; (7) fabrication - the methods available for the manufacture of the constituent parts of the Bitter plates, the method of assembly and remote maintenance are summarized

  4. Developing an ultrasonic NDE system for a research reactor tank

    International Nuclear Information System (INIS)

    Ultrasonic testing is one of the established tools for routine in-service inspection of reactor tanks. As part of the preventive maintenance of the IRR2 reactor, an ultrasonic scanning system was developed for the inspection of the reactor tank wall. Here, we present the main features of the special equipment developed for this task. In addition, we describe the procedure used for validating the inspection method. A special apparatus was developed for the ultrasonic scanning of a research reactor tank wall, the operation of which was practiced using a full-scale mock-up. The inspection technique was validated using a variety of flaws that were unknown to the operators

  5. Post reactor researches of fuel pins, tested under alternating NEMF reactor functioning modes

    International Nuclear Information System (INIS)

    Changing of rod ceramic fuel pins state under their exploitation conditions changing influence at alternating of three-mode nuclear energy-moving facility reactor functioning has been examined. There are presented the results of researches of fuel pins, tested in the reactor IRGIT and RA, firstly under moving mode, then - under energy mode of minor power of NEMF reactor. (author)

  6. Nuclear Reactor RA Safety Report, Vol. 5, Reactor cooling systems

    International Nuclear Information System (INIS)

    RA reactor cooling system enable cooling during normal operation and under possible accidental conditions and include: technical water system, heavy water system, helium gas system, system for heavy water purification and emergency cooling system. Primary cooling system is a closed heavy water circulation system. Heavy water system is designed to enable permanent circulation and twofold function of heavy water. In the upward direction of cooling it has a coolant role and in the downward direction it is the moderator. Separate part of the primary coolant loop is the system for heavy water purification. This system uses distillation and ion exchange processes

  7. Feedback compensation network design for KAPP reactor regulating system

    International Nuclear Information System (INIS)

    The reactor power regulating system based on coolant differential temperature across the boiler is inaccurate and sluggish because of the transport delays and time constants associated with temperature measurement. Moreover the control system cannot correct promptly the disturbances transmitted by the secondary system. Above problems can be easily overcome by the reactor control system based on neutron flux measured by the out of core ionisation chambers. The report describes the design and analysis of feedback compensation network based on neutron flux measurement. Closed loop system stability analysis of Kakrapar Atomic Power Plant has been made based on linearised transfer function models of sub-system, to achieve good gain and phase control margins. The control system responses have been tested using reactor functional simulator. The design has been verified by sampled data system analysis using Z transform of the reactor mathematical model. (author). 21 refs., 1 tab

  8. 23rd June 2010 - University of Bristol Head of the Aerospace Engineering Department and Professor of Aerospace Dynamics N. Lieven visiting CERN control centre with Beams Department Head P. Collier, visiting the LHC superconducting magnet test hall with R. Veness and CMS control centre with Collaboration Spokesperson G. Tonelli and CMS User J. Goldstein.

    CERN Multimedia

    Jean-Claude Gadmer

    2010-01-01

    23rd June 2010 - University of Bristol Head of the Aerospace Engineering Department and Professor of Aerospace Dynamics N. Lieven visiting CERN control centre with Beams Department Head P. Collier, visiting the LHC superconducting magnet test hall with R. Veness and CMS control centre with Collaboration Spokesperson G. Tonelli and CMS User J. Goldstein.

  9. Army Gas-Cooled Reactor Systems Program. Operation of ML-1 reactor skid in GCRE: safety evaluation report

    Energy Technology Data Exchange (ETDEWEB)

    None

    1964-10-01

    The operation of the ML-1 reactor skid in the modified GCRE facility, utilizing the GCRE reactor coolant circulating and heat removal systems, is described. An evaluation of the safety considerations associated with this mode of operation indicates that the consequences of the maximum credible accident are less severe than those previously approved for operation of the ML-1 reactor at the ML-1 test site or for operation of the GCRE-I reactor in the GCRE facility.

  10. Biofouling and microbial corrosion problem in the thermo-fluid heat exchanger and cooling water system of a nuclear test reactor.

    Science.gov (United States)

    Rao, T S; Kora, Aruna Jyothi; Chandramohan, P; Panigrahi, B S; Narasimhan, S V

    2009-10-01

    This article discusses aspects of biofouling and corrosion in the thermo-fluid heat exchanger (TFHX) and in the cooling water system of a nuclear test reactor. During inspection, it was observed that >90% of the TFHX tube bundle was clogged with thick fouling deposits. Both X-ray diffraction and Mossbauer analyses of the fouling deposit demonstrated iron corrosion products. The exterior of the tubercle showed the presence of a calcium and magnesium carbonate mixture along with iron oxides. Raman spectroscopy analysis confirmed the presence of calcium carbonate scale in the calcite phase. The interior of the tubercle contained significant iron sulphide, magnetite and iron-oxy-hydroxide. A microbiological assay showed a considerable population of iron oxidizing bacteria and sulphate reducing bacteria (10(5) to 10(6) cfu g(-1) of deposit). As the temperature of the TFHX is in the range of 45-50 degrees C, the microbiota isolated/assayed from the fouling deposit are designated as thermo-tolerant bacteria. The mean corrosion rate of the CS coupons exposed online was approximately 2.0 mpy and the microbial counts of various corrosion causing bacteria were in the range 10(3) to 10(5) cfu ml(-1) in the cooling water and 10(6) to 10(8) cfu ml(-1) in the biofilm. PMID:20183117

  11. Thermal-Hydraulic Analysis of an Experimental Reactor Cavity Cooling System with Air. Part I: Experiments; Part II: Separate Effects Tests and Modeling

    International Nuclear Information System (INIS)

    This experimental study investigates the thermal hydraulic behavior and the heat removal performance for a scaled Reactor Cavity Cooling System (RCCS) with air. A quarter-scale RCCS facility was designed and built based on a full-scale General Atomics (GA) RCCS design concept for the Modular High Temperature Gas Reactor (MHTGR). The GA RCCS is a passive cooling system that draws in air to use as the cooling fluid to remove heat radiated from the reactor pressure vessel to the air-cooled riser tubes and discharged the heated air into the atmosphere. Scaling laws were used to preserve key aspects and to maintain similarity. The scaled air RCCS facility at UW-Madison is a quarter-scale reduced length experiment housing six riser ducts that represent a 9.5° sector slice of the full-scale GA air RCCS concept. Radiant heaters were used to simulate the heat radiation from the reactor pressure vessel. The maximum power that can be achieved with the radiant heaters is 40 kW with a peak heat flux of 25 kW per meter squared. The quarter-scale RCCS was run under different heat loading cases and operated successfully. Instabilities were observed in some experiments in which one of the two exhaust ducts experienced a flow reversal for a period of time. The data and analysis presented show that the RCCS has promising potential to be a decay heat removal system during an accident scenario.

  12. Thermal-Hydraulic Analysis of an Experimental Reactor Cavity Cooling System with Air. Part I: Experiments; Part II: Separate Effects Tests and Modeling

    Energy Technology Data Exchange (ETDEWEB)

    Corradin, Michael [Univ. of Wisconsin, Madison, WI (United States). Dept. of Engineering Physics; Anderson, M. [Univ. of Wisconsin, Madison, WI (United States). Dept. of Engineering Physics; Muci, M. [Univ. of Wisconsin, Madison, WI (United States). Dept. of Engineering Physics; Hassan, Yassin [Texas A & M Univ., College Station, TX (United States); Dominguez, A. [Texas A & M Univ., College Station, TX (United States); Tokuhiro, Akira [Univ. of Idaho, Moscow, ID (United States); Hamman, K. [Univ. of Idaho, Moscow, ID (United States)

    2014-10-15

    This experimental study investigates the thermal hydraulic behavior and the heat removal performance for a scaled Reactor Cavity Cooling System (RCCS) with air. A quarter-scale RCCS facility was designed and built based on a full-scale General Atomics (GA) RCCS design concept for the Modular High Temperature Gas Reactor (MHTGR). The GA RCCS is a passive cooling system that draws in air to use as the cooling fluid to remove heat radiated from the reactor pressure vessel to the air-cooled riser tubes and discharged the heated air into the atmosphere. Scaling laws were used to preserve key aspects and to maintain similarity. The scaled air RCCS facility at UW-Madison is a quarter-scale reduced length experiment housing six riser ducts that represent a 9.5° sector slice of the full-scale GA air RCCS concept. Radiant heaters were used to simulate the heat radiation from the reactor pressure vessel. The maximum power that can be achieved with the radiant heaters is 40 kW with a peak heat flux of 25 kW per meter squared. The quarter-scale RCCS was run under different heat loading cases and operated successfully. Instabilities were observed in some experiments in which one of the two exhaust ducts experienced a flow reversal for a period of time. The data and analysis presented show that the RCCS has promising potential to be a decay heat removal system during an accident scenario.

  13. Development of multi-functional telerobotic systems for reactor dismantlement

    International Nuclear Information System (INIS)

    This report summarizes technological features of advanced telerobotic systems for reactor dismantling application developed at the Japan Atomic Energy Research Institute. Taking into consideration the special environmental conditions in reactor dismantling, major effort was made to develop multifunctional telerobotic system of high reliability which can be used to perform various complex tasks in an unstructured environment and operated in an easy and flexible manner. The system development was carried out through constructing three systems in seccession; a light-duty and a heavy-duty system as a prototype system for engineering test in cold environment, and a demonstration system for practical on-site application to dismantling highly radioactive reactor internals of an experimental boiling water reactor JPDR (Japan Power Demonstration Reactor). Each system was equipped with one or two amphibious manipulators which can be operated in either a push-button manual, a bilateral master-slave, a teach-and-playback or a programmed control mode. Different scheme was adopted in each system at designing the manipulator, transporter and man-machine interface so as to compare their advantages and disadvantages. According to the JPDR decommissioning program, the demonstration system was successfully operated to dismantle a portion of the radioactive reactor internals of the JPDR, which used underwater plasma arc cutting method and proved the usefulness of the multi-functional telerobotic system for reducing the occupational hazards and enhancing the work efficiency in the course of dismantling highly radioactive reactor components. (author)

  14. Tandem Mirror Reactor Systems Code (Version I)

    International Nuclear Information System (INIS)

    A computer code was developed to model a Tandem Mirror Reactor. Ths is the first Tandem Mirror Reactor model to couple, in detail, the highly linked physics, magnetics, and neutronic analysis into a single code. This report describes the code architecture, provides a summary description of the modules comprising the code, and includes an example execution of the Tandem Mirror Reactor Systems Code. Results from this code for two sensitivity studies are also included. These studies are: (1) to determine the impact of center cell plasma radius, length, and ion temperature on reactor cost and performance at constant fusion power; and (2) to determine the impact of reactor power level on cost

  15. Radiation exposure: Cytogenetic tests. Chernobyl reactor accident

    International Nuclear Information System (INIS)

    Forty test subjects who, either during or after the reactor accident of Chernobyl (26th April 1986), stayed at a building site at Shlobin 150 km away, were examined for spontaneously occurring as well as mitomycin C-induced Sister Chromatid Exchanges (SCE). The building site staff, who underwent a whole-body radionuclide count upon their return to Austria (June through September 1986), were used for the cytogenetic tests. The demonstration of the SCE was made from whole-blood cultures by the fluorescence/Giemse technique. At last 20 Metaphases of the 2nd mitotic cycle were evaluated per person. The radiation doses of the test subjects were calculated by adding the external exposure determined on the building site, the estimated thyroid dose through I-131, and the measured incorporation of Cs-134 and Cs-137. The subjects were divided into two groups for statistical analysis: One was a more exposed group (proven stay at Shlobin between 26th April and 31st May 1986, mostly working in the open air) and the other a less exposed group for comparison (staying at Shlobin from 1st Juni 1986 and working mainly indoors). (orig.)

  16. Modeling the Behaviour of an Advanced Material Based Smart Landing Gear System for Aerospace Vehicles

    International Nuclear Information System (INIS)

    The last two decades have seen a substantial rise in the use of advanced materials such as polymer composites for aerospace structural applications. In more recent years there has been a concerted effort to integrate materials, which mimic biological functions (referred to as smart materials) with polymeric composites. Prominent among smart materials are shape memory alloys, which possess both actuating and sensory functions that can be realized simultaneously. The proper characterization and modeling of advanced and smart materials holds the key to the design and development of efficient smart devices/systems. This paper focuses on the material characterization; modeling and validation of the model in relation to the development of a Shape Memory Alloy (SMA) based smart landing gear (with high energy dissipation features) for a semi rigid radio controlled airship (RC-blimp). The Super Elastic (SE) SMA element is configured in such a way that it is forced into a tensile mode of high elastic deformation. The smart landing gear comprises of a landing beam, an arch and a super elastic Nickel-Titanium (Ni-Ti) SMA element. The landing gear is primarily made of polymer carbon composites, which possess high specific stiffness and high specific strength compared to conventional materials, and are therefore ideally suited for the design and development of an efficient skid landing gear system with good energy dissipation characteristics. The development of the smart landing gear in relation to a conventional metal landing gear design is also dealt with

  17. A Diagnostic Approach for Electro-Mechanical Actuators in Aerospace Systems

    Science.gov (United States)

    Balaban, Edward; Saxena, Abhinav; Bansal, Prasun; Goebel, Kai Frank; Stoelting, Paul; Curran, Simon

    2009-01-01

    Electro-mechanical actuators (EMA) are finding increasing use in aerospace applications, especially with the trend towards all all-electric aircraft and spacecraft designs. However, electro-mechanical actuators still lack the knowledge base accumulated for other fielded actuator types, particularly with regard to fault detection and characterization. This paper presents a thorough analysis of some of the critical failure modes documented for EMAs and describes experiments conducted on detecting and isolating a subset of them. The list of failures has been prepared through an extensive Failure Modes and Criticality Analysis (FMECA) reference, literature review, and accessible industry experience. Methods for data acquisition and validation of algorithms on EMA test stands are described. A variety of condition indicators were developed that enabled detection, identification, and isolation among the various fault modes. A diagnostic algorithm based on an artificial neural network is shown to operate successfully using these condition indicators and furthermore, robustness of these diagnostic routines to sensor faults is demonstrated by showing their ability to distinguish between them and component failures. The paper concludes with a roadmap leading from this effort towards developing successful prognostic algorithms for electromechanical actuators.

  18. SMART Reactor Flow Distribution Test (SCOP-E-01)

    International Nuclear Information System (INIS)

    A Reactor Flow Distribution Test Facilities for SMART, named SCOP (SMART Core Flow and Pressure Test Facility), were designed in order to simulate the distributions of (1) core flow and (2) reactor sectional flow resistance and flow rates. This report summaries and analyzes the SCOP-E-01 Test which simulated the reactor internal flow distribution under the steady state conditions with the same flow rate at each loop. The primary parameters, which are represented by static/differential pressure, flow rate and temperature were found to satisfy well the requirement of instrumentation and uncertainties. In order to evaluate overall quality of test results, various secondary parameters were selected and analyzed, which shows that the quality of data are good. From the various hydraulic data representing the hydraulics of the SMART reactor, the soundness and performance of the reactor design can be demonstrated. The test data will be utilized as boundary conditions for the thermal margin analysis of SMART reactor

  19. A comparative analysis of user preference-based and existing knowledge management systems attributes in the aerospace industry

    Science.gov (United States)

    Varghese, Nishad G.

    Knowledge management (KM) exists in various forms throughout organizations. Process documentation, training courses, and experience sharing are examples of KM activities performed daily. The goal of KM systems (KMS) is to provide a tool set which serves to standardize the creation, sharing, and acquisition of business critical information. Existing literature provides numerous examples of targeted evaluations of KMS, focusing on specific system attributes. This research serves to bridge the targeted evaluations with an industry-specific, holistic approach. The user preferences of aerospace employees in engineering and engineering-related fields were compared to profiles of existing aerospace KMS based on three attribute categories: technical features, system administration, and user experience. The results indicated there is a statistically significant difference between aerospace user preferences and existing profiles in the user experience attribute category, but no statistically significant difference in the technical features and system administration attribute categories. Additional analysis indicated in-house developed systems exhibit higher technical features and user experience ratings than commercial-off-the-self (COTS) systems.

  20. Device for reactor control system

    International Nuclear Information System (INIS)

    A device for nuclear reactor control system is described. The device comprises a channel with control column of neutron absorbing liquid displacer and drain throttle. To increase the reliability and stabilization the control in the flow of liquid, the displacer is fixed to the bar with the help of a rod which length is not less than the half of the core height. The displacer occupies the lower section of the core and divides the column of liquid in two parts consisting of the control column above the displacer and protective column below the drain throttle. The proposed device provides the control of energy distribution along the core height and can be used for leveling energy distribution field or its shaping. A reliable operation of the device is insured, in particular, the stability of such important characteristics as the position and height of the column of liquid, the magnitude of introduced reactivity, the range of controlled parameters

  1. Replacement of the Advanced Test Reactor control room

    International Nuclear Information System (INIS)

    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

  2. A linear model of the Fast Breeder Test Reactor Plant

    International Nuclear Information System (INIS)

    A linear analysis of the Fast Breeder Test Reactor System, consisting of the reactor, intermediate heat exchanger, steam generator and connected piping is presented. The problem of variable boundaries in the steam generator is reduced to a problem of fixed boundaries by dividing the steam generator into six zones. Based upon this, one can obtain the transfer function of any input/output combination. Starting with the time domain non-linear partial differential equations, the problem is reduced to a system of linear equations in complex variables, which can be solved basically by Gaussian elimination process. The results of this work will be useful in determining a suitable control scheme for waterflow in the steam generator and the control parameters. (auth.)

  3. Reactor coolant pump monitoring and diagnostic system

    International Nuclear Information System (INIS)

    In order to reliably and safely operate a nuclear power plant, it is necessary to continuously monitor the performance of numerous subsystems to confirm that the plant state is within its prescribed limits. An important function of a properly designed monitoring system is the detection of incipient faults in all subsystems (with the avoidance of false alarms) coupled with an information system that provides the operators with fault diagnosis, prognosis of fault progression and recommended (either automatic or prescriptive) corrective action. In this paper, such a system is described that has been applied to reactor coolant pumps. This system includes a sensitive pattern-recognition technique based upon the sequential probability ratio test (SPRT) that detects incipient faults from validated signals, an expert system embodying knowledge bases on pump and sensor performance, extensive hypertext files containing operating and emergency procedures as well as pump and sensor information and a graphical interface providing the operator with easily perceived information on the location and character of the fault as well as recommended corrective action. This system is in the prototype stage and is currently being validated utilizing data from a liquid-metal cooled fast reactor (EBR-II). 3 refs., 4 figs

  4. Summary of aerospace and nuclear engineering activities

    Science.gov (United States)

    1988-01-01

    The Texas A&M Nuclear and Aerospace engineering departments have worked on five different projects for the NASA/USRA Advanced Design Program during the 1987/88 year. The aerospace department worked on two types of lunar tunnelers that would create habitable space. The first design used a heated cone to melt the lunar regolith, and the second used a conventional drill to bore its way through the crust. Both used a dump truck to get rid of waste heat from the reactor as well as excess regolith from the tunneling operation. The nuclear engineering department worked on three separate projects. The NEPTUNE system is a manned, outer-planetary explorer designed with Jupiter exploration as the baseline mission. The lifetime requirement for both reactor and power-conversion systems was twenty years. The second project undertaken for the power supply was a Mars Sample Return Mission power supply. This was designed to produce 2 kW of electrical power for seven years. The design consisted of a General Purpose Heat Source (GPHS) utilizing a Stirling engine as the power conversion unit. A mass optimization was performed to aid in overall design. The last design was a reactor to provide power for propulsion to Mars and power on the surface. The requirements of 300 kW of electrical power output and a mass of less than 10,000 Rg were set. This allowed the reactor and power conversion unit to fit within the Space Shuttle cargo bay.

  5. Design of ventilation system of reactor building for CARR engineering

    International Nuclear Information System (INIS)

    The ventilation rate of the reactor building is determined by the calculation results of radiation protection, which has referred to the demands of code, the experience of German FRMII reactor designing, and the pollution level as well as the personnel residence of each room. Direct ventilation system is used in the reactor building, and each storey has a separate ventilation and air cleaning system. Airtight quick isolation valves, of which the in-out-leakage rate is 0 under the reactor building tightness test pressure (12.5kPa), are set on the air ducts which through the sealing boundary of operating hall. This measure can guarantee the radioactive matters against leaking into outside spaces through the ventilation ducts during accident conditions of the reactor. Direct-connected steel fans and integral stainless steel air cleaning equipments are chosen in the system designing. (authors)

  6. Knowledge maturity as a means to support decision making during product-service systems development projects in the aerospace sector

    OpenAIRE

    Johansson, Christian; Hicks, Ben; Larsson, Andreas; Bertoni, Marco

    2011-01-01

    Streamlining new product development forces companies to make decisions on preliminary information. This paper considers this challenge within the context of project management in the aerospace sector, and in particular the development of product-service systems.  The concept of knowledge maturity is explored as a means to provide practical decision support, which increases decision makers' awareness of the knowledge base and supports cross-boundary discussions on the perceived maturity of av...

  7. Ohmically heated toroidal experiment (OHTE) mobile ignition test reactor facility concept study

    International Nuclear Information System (INIS)

    This report presents the results of a study to evaluate the use of an existing nuclear test complex at the Idaho National Engineering Laboratory (INEL) for the assembly, testing, and remote maintenance of the ohmically heated toroidal experiment (OHTE) compact reactor. The portable reactor concept is described and its application to OHTE testing and maintenance requirements is developed. Pertinent INEL facilities are described and several test system configurations that apply to these facilities are developed and evaluated

  8. Expert system for fast reactor diagnostic

    International Nuclear Information System (INIS)

    A general description of expert systems is given. The operation of a fast reactor is reviewed. The expert system to the diagnosis of breakdowns limited to the reactor core. The structure of the system is described: specification of the diagnostics; structure of the data bank and evaluation of the rules; specification of the prediagnostics and evaluation; explanation of the diagnostics; time evolution of the system; comparison with other expert systems. Applications to some cases of faults are finally presented

  9. New Sensors for Irradiation Testing at Materials and Test Reactors

    International Nuclear Information System (INIS)

    Enhanced instrumentation, capable of providing real-time measurements of parameters during fuels and material irradiations, is required to support irradiation testing requested by US nuclear research programs. For example, several research programs funded by the US Department of Energy (US DOE) are emphasizing the use of first principle models to characterize the performance of fuels and materials. To facilitate this approach, high fidelity, real-time data are essential to demonstrate the performance of these new fuels and materials during irradiation testing. Furthermore, sensors that obtain such data in US MTRs, such as the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL), must be miniature, reliable, and able to withstand high fluxes and high temperatures. Depending on program requirements, sensors may need to obtain data in inert gas, pressurized water, or liquid metal environments. To address these needs, INL has developed and deployed several new sensors to support irradiation testing in US DOE programs. The paper identifies the sensors currently available to support higher flux US MTR irradiations. Recent results and products from sensor research and development are highlighted. In particular, progress in deploying enhanced in-pile sensors for detecting temperature, elongation, and thermal conductivity is emphasized. Finally, initial results from research to evaluate the viability of ultrasonic and fiber optic technologies for irradiation testing are summarized. (author)

  10. Applicability of Aerospace Materials Ground Flammability Test Data to Spacecraft Environments Theory and Applied Technologies

    Science.gov (United States)

    Hirsch, David; Williams, Jim; Beeson, Harold

    2009-01-01

    This slide presentation reviews the use of ground test data in reference to flammability to spacecraft environments. It reviews the current approach to spacecraft fire safety, the challenges to fire safety that the Constellation program poses, the current trends in the evaluation of the Constellation materials flammability, and the correlation of test data from ground flammability tests with the spacecraft environment. Included is a proposal for testing and the design of experiments to test the flammability of materials under similar spacecraft conditions.

  11. Method of testing fuel assemblies for nuclear reactors

    International Nuclear Information System (INIS)

    The stresses occurring in the fuel assemblies are simulated by power excursions. For this purpose the fuel assembly is placed in the neutron field of a test reactor and for a short time can be exposed to the much higher neutron field of a pulsed reactor. One possibility of design provides for the test and the pulsed reactor lying one above the other, separated by a neutron absorber and penetrated by a common irradiation channel. The fuel assembly then is to be moved from the position in the test reactor to the position in the pulsed reactor. The other possibility is to make the irradiation duct pass along the gap between both reactors and, by means of a tube-shaped absorber, open one or the other irradiation field. (DG)

  12. Simulation and tests to individual and coupled models of the reactor vessel simulator and the recirculation system for the SUN-RAH; Simulacion y pruebas a modelos individuales y acoplados del simulador de la vasija del reactor y el sistema de recirculacion para el SUN-RAH

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez S, R.A. [UNAM, Laboratorio de Analisis de Ingenieria de Reactores Nucleares, DEPFI, Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)]. e-mail: rsanchez_15@yahoo.com.mx

    2004-07-01

    The present project, is continuation of the project presented in the congress SNM-2003. In this new phase of the project, they were carried out adaptive changes to the modeling and implementation of the module of the full superior of the core of the reactor, they were carried out those modeling of the generation of heat as well as of the energy transfer in the one fuel. These models present the main characteristics of the vessel of the one reactor and of the recirculation system, defined by the main phenomena that they intervene in the physical processes, in the previous version the simulation in real time it required of an extremely quick computer and without executing collateral processes. The tests are presented carried out to the different models belonging to the Simulator of the Reactor Vessel and the Recirculation system for the SUN-RAH (University Simulator of Nucleo electric with Boiling Water Reactor), as well as the results hurtled by this tests. In each section the executions of the tests and the corresponding analyses of results are shown for each pattern. Besides the above mentioned, the advantages presented by the Simulator of the reactor vessel and the recirculation system are pointed. (Author)

  13. Research Reactor Power Control System Design by MATLAB/SIMULINK

    Energy Technology Data Exchange (ETDEWEB)

    Baang, Dane; Suh, Yong Suk; Kim, Young Ki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Im, Ki Hong [Samsung Electronics, Suwon (Korea, Republic of)

    2013-07-01

    In this study it is presented that MATLAB/SIMULINK can be efficiently used for modeling and power control system design for research reactors. The presented power control system deals with various functions including reactivity control, signals processing, reactivity calculation, alarm request generation, etc., thus it is required to test all the software logic using proper model for reactor, control rods, and field instruments. In MATLAB/SIMULINK tool, point kinetics, thermal model, control absorber rod model, and other instrument models were developed based on reactor parameters and known properties of each component or system. The software for power control system was invented and linked to the model to test each function. From the simulation result it is shown that the power control performance and other functions of the system can be easily tested and analyzed in the proposed simulation structure.

  14. Research Reactor Power Control System Design by MATLAB/SIMULINK

    International Nuclear Information System (INIS)

    In this study it is presented that MATLAB/SIMULINK can be efficiently used for modeling and power control system design for research reactors. The presented power control system deals with various functions including reactivity control, signals processing, reactivity calculation, alarm request generation, etc., thus it is required to test all the software logic using proper model for reactor, control rods, and field instruments. In MATLAB/SIMULINK tool, point kinetics, thermal model, control absorber rod model, and other instrument models were developed based on reactor parameters and known properties of each component or system. The software for power control system was invented and linked to the model to test each function. From the simulation result it is shown that the power control performance and other functions of the system can be easily tested and analyzed in the proposed simulation structure

  15. Near term test plan using HTTR (high temperature engineering test reactor)

    Energy Technology Data Exchange (ETDEWEB)

    Takada, Shoji, E-mail: takada.shoji@jaea.go.jp [HTTR Reactor Engineering Section, Department of HTTR, Oarai Research and Development Center, Japan Atomic Energy Agency, Narita, Oarai, Higashi-ibaraki, Ibaraki 311-1393 (Japan); Iigaki, Kazuhiko; Shinohara, Masanori; Tochio, Daisuke; Shimazaki, Yosuke; Ono, Masato; Yanagi, Shunki [HTTR Reactor Engineering Section, Department of HTTR, Oarai Research and Development Center, Japan Atomic Energy Agency (JAEA) (Japan); Nishihara, Tetsuo [Policy Department and Administration Department, JAEA (Japan); Fukaya, Yuji [HTGR Design Group, Small-Sized HTGR Research and Development Division, Nuclear Hydrogen and Heat Application Research Center, JAEA (Japan); Goto, Minoru [HTGR Safety Evaluation Group, Small-Sized HTGR Research and Development Division, Nuclear Hydrogen and Heat Application Research Center, JAEA (Japan); Tachibana, Yukio [HTGR Design Group, Small-Sized HTGR Research and Development Division, Nuclear Hydrogen and Heat Application Research Center, JAEA (Japan); Sawa, Kazuhiro [HTTR Reactor Engineering Section, Department of HTTR, Oarai Research and Development Center, Japan Atomic Energy Agency (JAEA) (Japan)

    2014-05-01

    JAEA has carried out research and development to establish the technical basis of high temperature gas cooled reactors (HTGRs) using HTTR. In order to connect hydrogen production system to HTTR, it is necessary to ensure the stability of plant dynamics when the thermal-load of the system is lost. Thermal-load fluctuation test is planned to demonstrate the stable reactor dynamics and to gain the test data for validation of the plant dynamics code. It will be confirmed that the reactor become stable state during a part of removed heat at HTTR heat-sink is lost. A temperature coefficient of reactivity is one of the important parameters for core dynamics calculations for safety analysis, and changes with burnup because of variance of fuel compositions. Measurement of temperature coefficient of reactivity has been conducted by HTTR to confirm the validity of the calculated temperature coefficient of reactivity. A loss of forced cooling (LOFC) test using HTTR has been carried out to verify the inherent safety of HTGR under the condition of loss of forced cooling while the reactor shut-down system disabled.

  16. Reactor inventory monitoring system for Angra-1 reactor

    International Nuclear Information System (INIS)

    This work describes the project of Reactor Inventory Monitoring System, which will be installed in Angra I Nuclear Power Plant. The inventory information is important to the operators take corrective actions in case of an incident that may cause a failure in the core cooling. (author)

  17. Nuclear reactors transients identification and classification system

    International Nuclear Information System (INIS)

    This work describes the study and test of a system capable to identify and classify transients in thermo-hydraulic systems, using a neural network technique of the self-organizing maps (SOM) type, with the objective of implanting it on the new generations of nuclear reactors. The technique developed in this work consists on the use of multiple networks to do the classification and identification of the transient states, being each network a specialist at one respective transient of the system, that compete with each other using the quantization error, that is a measure given by this type of neural network. This technique showed very promising characteristics that allow the development of new functionalities in future projects. One of these characteristics consists on the potential of each network, besides responding what transient is in course, could give additional information about that transient. (author)

  18. Intelligent test integration system

    Science.gov (United States)

    Sztipanovits, J.; Padalkar, S.; Rodriguez-Moscoso, J.; Kawamura, K.; Purves, B.; Williams, R.; Biglari, H.

    1988-01-01

    A new test technology is described which was developed for space system integration. The ultimate purpose of the system is to support the automatic generation of test systems in real time, distributed computing environments. The Intelligent Test Integration System (ITIS) is a knowledge based layer above the traditional test system components which can generate complex test configurations from the specification of test scenarios.

  19. Decontamination and Decommissioning of the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-11-01

    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.

  20. LIGHT WATER REACTOR ACCIDENT TOLERANT FUELS IRRADIATION TESTING

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-01

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

  1. High temperature indentation tests on fusion reactor candidate materials

    International Nuclear Information System (INIS)

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

  2. Aerospace Environmental Technology Conference

    Science.gov (United States)

    Whitaker, A. F. (Editor)

    1995-01-01

    The mandated elimination of CFC's, Halons, TCA, and other ozone depleting chemicals and specific hazardous materials has required changes and new developments in aerospace materials and processes. The aerospace industry has been involved for several years in providing product substitutions, redesigning entire production processes, and developing new materials that minimize or eliminate damage to the environment. These activities emphasize replacement cleaning solvents and their application verifications, compliant coatings including corrosion protection systems, and removal techniques, chemical propulsion effects on the environment, and the initiation of modifications to relevant processing and manufacturing specifications and standards. The Executive Summary of this Conference is published as NASA CP-3297.

  3. Frontier Aerospace Opportunities

    Science.gov (United States)

    Bushnell, Dennis M.

    2014-01-01

    Discussion and suggested applications of the many ongoing technology opportunities for aerospace products and missions, resulting in often revolutionary capabilities. The, at this point largely unexamined, plethora of possibilities going forward, a subset of which is discussed, could literally reinvent aerospace but requires triage of many possibilities. Such initial upfront homework would lengthen the Research and Development (R&D) time frame but could greatly enhance the affordability and performance of the evolved products and capabilities. Structural nanotubes and exotic energetics along with some unique systems approaches are particularly compelling.

  4. Thermal stratification in nuclear reactor piping system

    International Nuclear Information System (INIS)

    Thermal stratification, cycling, and striping (TASCS) issue has drawn attention recently because of the incidents at several nuclear plants relative to thermal fatigue in piping systems connected to the main coolant piping. U.S. nuclear utilities are addressing the issue in response to the concerns. In particular, the Electric Power Research Institute (EPRI) initiated a major research program to resolve the TASCS issue. In Phase 1 research, a methodology and program have been developed to conduct detailed research into mechanisms which lead to fatigue in nominally stagnant piping systems near the reactor coolant piping. Three key efforts from the Phase 1 program are described in this paper. First, the line evaluation methodology is described, which also leads to requirements for the Phase 2 program. Second, tests to investigate interaction between main coolant piping and stagnant attached lines by turbulence penetration are described. Turbulence penetration into unisolable lines, or the transport of turbulence into stagnant piping from the reactor coolant system (RCS) line, represents a mechanism for carrying hot RCS water into regions filled with colder water. The possibility of stratification of the two fluids (and the resultant stresses) are the reason for developing an understanding of the turbulence penetration process. Lastly, results of an evaluation to develop a loading definition for thermal striping are included. Future plans and prospective results are also discussed. (author)

  5. Structural materials challenges for advanced reactor systems

    Science.gov (United States)

    Yvon, P.; Carré, F.

    2009-03-01

    Key technologies for advanced nuclear systems encompass high temperature structural materials, fast neutron resistant core materials, and specific reactor and power conversion technologies (intermediate heat exchanger, turbo-machinery, high temperature electrolytic or thermo-chemical water splitting processes, etc.). The main requirements for the materials to be used in these reactor systems are dimensional stability under irradiation, whether under stress (irradiation creep or relaxation) or without stress (swelling, growth), an acceptable evolution under ageing of the mechanical properties (tensile strength, ductility, creep resistance, fracture toughness, resilience) and a good behavior in corrosive environments (reactor coolant or process fluid). Other criteria for the materials are their cost to fabricate and to assemble, and their composition could be optimized in order for instance to present low-activation (or rapid desactivation) features which facilitate maintenance and disposal. These requirements have to be met under normal operating conditions, as well as in incidental and accidental conditions. These challenging requirements imply that in most cases, the use of conventional nuclear materials is excluded, even after optimization and a new range of materials has to be developed and qualified for nuclear use. This paper gives a brief overview of various materials that are essential to establish advanced systems feasibility and performance for in pile and out of pile applications, such as ferritic/martensitic steels (9-12% Cr), nickel based alloys (Haynes 230, Inconel 617, etc.), oxide dispersion strengthened ferritic/martensitic steels, and ceramics (SiC, TiC, etc.). This article gives also an insight into the various natures of R&D needed on advanced materials, including fundamental research to investigate basic physical and chemical phenomena occurring in normal and accidental operating conditions, lab-scale tests to characterize candidate materials

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

    International Nuclear Information System (INIS)

    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

  7. 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.

  8. Skoda computational system for WWER reactors

    International Nuclear Information System (INIS)

    The code system that is used in Skoda Plzen for reactor safety analysis, in-core fuel management studies, criticality assessment of storage pools etc. for WWER-type reactor lattices, and some results of its validation, are described in this paper. (author). 13 refs, 10 figs, 7 tabs

  9. New technology for reactor protection system of CAREM reactor

    International Nuclear Information System (INIS)

    The use of FPGA in safety functions in a nuclear power plant, increase the reliability of software based systems, without loose any of the function required by the supervision and control systems. In this work the architecture of a Reactor Protection System is described, it use four independent measurement channels in 2 oo 4 configuration, each channel is based on diverse approach in 1 oo 2 configuration, the reliability of this system is near the same than the hardwired logic, with full performance like software based system. (author)

  10. IAEA data base system for nuclear research reactors (RRDB)

    International Nuclear Information System (INIS)

    The IAEA Data Base System for Nuclear Research Reactors (RRDB) User's Guide is intended for the user who wishes to understand the concepts and operation of the RRDB system. The RRDB is a computerized system recording administrative, operational and technical data on all the nuclear research reactors currently operating, under construction, planned or shut down in IAEA Member States. The data is received by the IAEA from reactor centres on magnetic tapes or as responses to questionnaires. All the data on research, training, test and radioactive isotope production reactors and critical assemblies is stored on the RRDB system. A full set of RRDB programs (in NATURAL) are contained at the back of this Guide

  11. Microprocessor tester for the treat upgrade reactor trip system

    International Nuclear Information System (INIS)

    The upgrading of the Transient Reactor Test (TREAT) Facility at ANL-Idaho has been designed to provide additional experimental capabilities for the study of core disruptive accident (CDA) phenomena. In addition, a programmable Automated Reactor Control System (ARCS) will permit high-power transients up to 11,000 MW having a controlled reactor period of from 15 to 0.1 sec. These modifications to the core neutronics will improve simulation of LMFBR accident conditions. Finally, a sophisticated, multiply-redundant safety system, the Reactor Trip System (RTS), will provide safe operation for both steady state and transient production operating modes. To insure that this complex safety system is functioning properly, a Dedicated Microprocessor Tester (DMT) has been implemented to perform a thorough checkout of the RTS prior to all TREAT operations

  12. Operating and test experience with Experimental Breeder Reactor number 2 (EBR-II), the Integral Fast Reactor (IFR) prototype

    International Nuclear Information System (INIS)

    The Experimental Breeder Reactor number 2 (EBR-II) has operated for 30 years, the longest for any liquid metal cooled reactor (LMR) power plant in the world. Given the scope of what has been developed and demonstrated over those years, it is arguably the most successful test reactor operation ever. Tests have been carried out on virtually every fast reactor fuel type. The reactor itself has been extensively studied. The most dramatic safety tests, conducted on 3 April, 1986, showed that an LMR with metallic fuel could safely accommodate loss of flow or loss of heat-sink without scram. EBR-II operated as the Integral Fast Reactor (IFR) prototype, demonstrating important innovations in safety, plant design, fuel design and actinide recycle. The ability to accommodate anticipated transients without scram passively resulted in significant simplification of the reactor plant, primarily through less reliance on emergency power and not having to require the secondary sodium or steam systems to be safety grade. These features have been quantified in a probabilistic risk assessment (PRA) conducted for EBR-II, demonstrating considerable safety advantages over other reactor concepts. Fundamental to the superior safety and operating characteristics of this reactor is the metallic U-Pu-Zr alloy fuel. Performance of the fuel has been fully proven: achieved burnup levels exceed 20 at.% in the lead test assemblies. A complete set of fuel performance and safety limits has been developed and was carried forward in formal safety documents supporting conversion of the core to IFR fuel. The last major demonstration planned was to assess the performance of recycled actinides in the fuel and to confirm that passive safety characteristics are maintained with recycled actinide fuel in the core. (author)

  13. Clinch River Breeder Reactor secondary control rod system

    International Nuclear Information System (INIS)

    The shutdown system for the Clinch River Breeder Reactor (CRBR) includes two independent systems--a primary and a secondary system. The Secondary Control Rod System (SCRS) is a new design which is being developed by General Electric to be independent from the primary system in order to improve overall shutdown reliability by eliminating potential common-mode failures. The paper describes the status of the SCRS design and fabrication and testing activities. Design verification testing on the component level is largely complete. These component tests are covered with emphasis on design impact results. A prototype unit has been manufactured and system level tests in sodium have been initiated

  14. Development and Deployment of an Aerospace Recommended Practice (ARP) Compliant Measurement System for nvPM Certification Measurements of Aircraft Engines - Current Status.

    Science.gov (United States)

    Whitefield, P. D.; Hagen, D. E.; Lobo, P.; Miake-Lye, R. C.

    2015-12-01

    The Society of Automotive Engineers (SAE) Aircraft Exhaust Emissions Measurement Committee (E-31) has published an Aerospace Information Report (AIR) 6241 detailing the sampling system for the measurement of non-volatile particulate matter (nvPM) from aircraft engines (SAE 2013). The system is designed to operate in parallel with existing International Civil Aviation Organization (ICAO) Annex 16 compliant combustion gas sampling systems used for emissions certification from aircraft engines captured by conventional (Annex 16) gas sampling rakes (ICAO, 2008). The SAE E-31 committee is also working to ballot an Aerospace Recommended Practice (ARP) that will provide the methodology and system specification to measure nvPM from aircraft engines. The ARP is currently in preparation and is expected to be ready for ballot in 2015. A prototype AIR-compliant nvPM measurement system - The North American Reference System (NARS) has been built and evaluated at the MSTCOE under the joint sponsorship of the FAA, EPA and Transport Canada. It has been used to validate the performance characteristics of OEM AIR-compliant systems and is being used in engine certification type testing at OEM facilities to obtain data from a set of representative engines in the fleet. The data collected during these tests will be used by ICAO/CAEP/WG3/PMTG to develop a metric on which on the regulation for nvPM emissions will be based. This paper will review the salient features of the NARS including: (1) emissions sample transport from probe tip to the key diagnostic tools, (2) the mass and number-based diagnostic tools for nvPM mass and number concentration measurement and (3) methods employed to assess the extent of nvPM loss throughout the sampling system. This paper will conclude with a discussion of the recent results from inter-comparison studies conducted with other US - based systems that gives credence to the ARP's readiness for ballot.

  15. Distributed expert systems for nuclear reactor control

    International Nuclear Information System (INIS)

    A network of distributed expert systems is the heart of a prototype supervisory control architecture developed at the Oak Ridge National Laboratory (ORNL) for an advanced multimodular reactor. Eight expert systems encode knowledge on signal acquisition, diagnostics, safeguards, and control strategies in a hybrid rule-based, multiprocessing and object-oriented distributed computing environment. An interactive simulation of a power block consisting of three reactors and one turbine provides a realistic, testbed for performance analysis of the integrated control system in real-time. Implementation details and representative reactor transients are discussed

  16. Design of test blanket system for ITER module testing

    International Nuclear Information System (INIS)

    Test blanket systems to be installed in ITER for developing demo blankets have been investigated. One of the main engineering goals of ITER is to test tritium breeding blankets relevant to a power reactor. The test foreseen on modules include the demonstration of a breeding capability that would lead to tritium self-sufficiency in a reactor and extraction of a high grade heat suitable for electricity generation. To accomplish these goals, several ITER equatorial ports are available to test the test blanket systems, both in the basic performance phase (BPP) and the enhanced performance phase (EPP). Test blanket systems for water-cooled and helium-cooled type DEMO blankets with ceramic breeders, developed in Japan have been designed. The design activities include the neutronics, thermal and hydraulic analyses, and mechanical configuration considering port sharing, cooling systems and tritium recovery systems, and test blanket system compatible with the current ITER design has been developed. (author)

  17. Testing of HTR UO2 TRISO fuels in AVR and in material test reactors

    International Nuclear Information System (INIS)

    The German High Temperature Reactor Fuel Development Program successfully developed, licensed and manufactured many thousands of spherical fuel elements that were used to power the experimental AVR reactor and the commercial THTR reactor. In the 1970s, this program extended the performance envelope of HTR fuels by developing and qualifying the TRISO-coated particle system. Irradiation testing in real-time AVR tests and accelerated MTR tests demonstrated the superior manufacturing process of this fuel and its irradiation performance. In the 1980s, another program direction change was made to a low enriched UO2 TRISO-coated particle system coupled with high-quality manufacturing specifications designed to meet new HTR plant design needs. These needs included requirements for inherent safety under normal operation and accident conditions. Again, the German fuel development program met and exceeded these challenges by manufacturing and qualifying the low-enriched UO2 TRISO-fuel system for HTR systems with steam generation, gas-turbine systems and very high temperature process heat applications. Fuel elements were manufactured in production scale facilities that contained near defect free UO2 TRISO coated particles, homogeneously distributed within a graphite matrix with very low levels of uranium contamination. Good irradiation performance for these elements was demonstrated under normal operating conditions to 12% FIMA and under accident conditions not exceeding 1600 °C

  18. Development and validation of a real-time SAFT-UT [synthetic aperture focusing technique for ultrasonic testing] system for the inspection of light water reactor components: Annual report, October 1985-September 1986

    International Nuclear Information System (INIS)

    The Pacific Northwest Laboratory is working to design, fabricate, and evaluate a real-time flaw detection and characterization system based on the synthetic aperture focusing technique for ultrasonic testing (SAFT-UT). The system is designed to perform inservice inspection of light-water reactor components. Included objectives of this program for the Nuclear Regulatory Commission are to develop procedures for system calibration and field operation, to validate the system through laboratory and field inspections, and to generate an engineering data base to support ASME Code acceptance of the technology. This progress report covers the programmatic work from October 1985 through September 1986. 45 figs., 8 tabs

  19. NASA Aerospace Flight Battery Program: Recommendations for Technical Requirements for Inclusion in Aerospace Battery Procurements. Volume 1, Part 2

    Science.gov (United States)

    Jung, David S.; Manzo, Michelle A.

    2010-01-01

    This NASA Aerospace Flight Battery Systems Working Group was chartered within the NASA Engineering and Safety Center (NESC). The Battery Working Group was tasked to complete tasks and to propose proactive work to address battery related, agency-wide issues on an annual basis. In its first year of operation, this proactive program addressed various aspects of the validation and verification of aerospace battery systems for NASA missions. Studies were performed, issues were discussed and in many cases, test programs were executed to generate recommendations and guidelines to reduce risk associated with various aspects of implementing battery technology in the aerospace industry. This document contains Part 2 - Volume I: Recommendations for Technical Requirements for Inclusion in Aerospace Battery Procurements of the program's operations.

  20. NASA Non-Flow-Through PEM Fuel Cell System for Aerospace Applications

    Science.gov (United States)

    Araghi, Koorosh R.

    2011-01-01

    NASA is researching passive NFT Proton Exchange Membrane (PEM) fuel cell technologies for primary fuel cell power plants in air-independent applications. NFT fuel cell power systems have a higher power density than flow through systems due to both reduced parasitic loads and lower system mass and volume. Reactant storage still dominates system mass/volume considerations. NFT fuel cell stack testing has demonstrated equivalent short term performance to flow through stacks. More testing is required to evaluate long-term performance.

  1. Proceedings of the international symposium on materials testing reactors

    International Nuclear Information System (INIS)

    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.)

  2. EPR/PTFE dosimetry for test reactor environments

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

  3. EPR/PTFE dosimetry for test reactor environments

    International Nuclear Information System (INIS)

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

  4. Application of Shuttle Remote Manipulator System technology to the replacement of fuel channels in the Pickering CANDU reactor

    International Nuclear Information System (INIS)

    Spar Aerospace Limited of Toronto was the prime contractor to the National Research Council of Canada for the design and development of the Shuttle Remote Manipulator (SRMS). Spar is presently under contract to Ontario Hydro to design and build a Remote Manipulation Control System to replace the fuel channels in the Pickering A Nuclear Generating Station. The equipment may be used to replace the fuel channels in six other early generation CANDU reactors

  5. Development, utilization, and future prospects of materials test reactors

    International Nuclear Information System (INIS)

    Reactor radiation affects the chemical and physical properties of materials. These changes can be very drastic in certain cases. Special test reactors have therefore been built since the 1950's and specific skills were developed to expose materials specimens to the precise irradiation conditions required. Materials testing reactors are those research reactor facilities which are designed and operated predominantly for studies into radiation damage. About a dozen plants in European communities (EC) Member States and in the US can be identified in this category, with 5 to 100 MW fission power and neutron fluxes between 5 x 1013 and 1015 cm-2s-1. The paper elaborates common aspects of development, utilization, and future prospects of US and EC materials testing reactors, and indicates the most significant differences

  6. Controls in new construction reactors-factory testing of the non-safety portion of the Lungmen nuclear power plant distributed control system

    International Nuclear Information System (INIS)

    The construction permit for Taipower's Lungmen Nuclear Units 1 and 2, two ABWR plants, was issued on March 17, 1999[1], The construction of these units is progressing actively at site. The digital I and C system supplied by GE, which is designated as the Distributed Control and Information System (DCIS) in this project, is being implemented primarily at one vendor facility. In order to ensure the reliability, safety and availability of the DCIS, it is required to comprehensively test the whole DCIS in factory. This article describes the test requirements and acceptance criteria for functional testing of the Non-Safety Distributed Control and Information system (DCIS) for Taiwan Power's Lungmen Units 1 and 2 GE selected Invensys as the equipment supplier for this Non-Safety portion of DCIS. The DCIS system of the Lungmen Units is a physically distributed control system. Field transmitters are connected to hard I/O terminal inputs on the Invensys I/A system. Once the signal is digitized on FBMs (Field Bus Modules) in Remote Multiplexing Units (RMUs), the signal is passed into an integrated control software environment. Control is based on the concept of compounds and blocks where each compound is a logical collection of blocks that performs a control function. Each point identified by control compound and block can be individually used throughout the DCIS system by referencing its unique name. In the Lungmen Project control logic and HSI (Human System Interface) requirements are divided into individual process systems called MPLs (Master Parts List). Higher-level Plant Computer System (PCS) algorithms access control compounds and blocks in these MPLs to develop functions. The test requirements and acceptance criteria for the DCIS system of the Lungmen Project are divided into three general categories (see 1,2,3 below) of verification, which in turn are divided into several specific tests: 1. DCIS System Physical Checks a) RMU Test - To confirm that the hard I

  7. Stack Monitoring System At PUSPATI TRIGA Reactor

    International Nuclear Information System (INIS)

    This paper describes the current Stack Monitoring System at PUSPATI TRIGA Reactor (RTP) building. A stack monitoring system is a continuous air monitor placed at the reactor top for monitoring the presence of radioactive gaseous in the effluent air from the RTP building. The system consists of four detectors that provide the reading for background, particulate, Iodine and Noble gas. There is a plan to replace the current system due to frequent fault of the system, thus thorough understanding of the current system is required. Overview of the whole system will be explained in this paper. Some current results would be displayed and moving forward brief plan would be mentioned. (author)

  8. Research on Fiber Optic Gyroscope Test Data Management System

    Directory of Open Access Journals (Sweden)

    Hongxia Cai

    2013-05-01

    Full Text Available FOG is a new type of angular velocity transducer; it is widely used in aviation, aerospace, marine and other fields. During FOG R & D, the test work costs long time, there are many test data in FOG life cycle, including structured data and unstructured data. This paper analyzed the FOG R & D process, and classified the test data. The paper also analyzed the test data management requirements and pointed out the main problems in the test data management. Based on this, test data management methods and test data management system architecture are given in this paper. Finally, a test data management system with B / S structure is developed.

  9. Small test SDHW systems

    DEFF Research Database (Denmark)

    Vejen, Niels Kristian

    1999-01-01

    Three small test SDHW systems was tested in a laboratory test facility.The three SDHW systems where all based on the low flow principe and a mantle tank but the design of the systems where different.......Three small test SDHW systems was tested in a laboratory test facility.The three SDHW systems where all based on the low flow principe and a mantle tank but the design of the systems where different....

  10. Molecular ecology of anaerobic reactor systems

    DEFF Research Database (Denmark)

    Hofman-Bang, H. Jacob Peider; Zheng, D.; Westermann, Peter; Ahring, Birgitte Kiær; Raskin, L.

    2003-01-01

    Anaerobic reactor systems are essential for the treatment of solid and liquid wastes and constitute a core facility in many waste treatment plants. Although much is known about the basic metabolism in different types of anaerobic reactors, little is known about the microbes responsible for these ...... specific nucleic acid probes are discussed and exemplified by studies of anaerobic granular sludge, biofilm and digester systems...... malfunctions of anaerobic digesters occasionally experienced, leading to sub-optimal methane production and wastewater treatment. Using a variety of molecular techniques, we are able to determine which microorganisms are active, where they are active, and when they are active, but we still need to determine...... abundance of each microbe in anaerobic reactor systems by rRNA probing. This chapter focuses on various molecular techniques employed and problems encountered when elucidating the microbial ecology of anaerobic reactor systems. Methods such as quantitative dot blot/fluorescence in-situ probing using various...

  11. Theoretical Rationale of Heating Block for Testing Bench of Aerospace Crafts Thermal Protection Elements

    Science.gov (United States)

    Petrova, Anna A.; Reznik, Sergey V.

    2016-02-01

    The theoretical rationale for the structural layout of a testing bench with zirconium dioxide heating elements on the basis of modelling radiative-conductive heat transfer are presented. The numerical simulation of radiative-conductive heat transfer for the two-dimensional scaled model of the testing segment with the finite-element analysis software package Ansys 15.0 are performed. The simulation results showed that for the selected layout of the heaters the temperature non-uniformity along the length of the sample over time will not exceed 3 % even at a temperature of 2000 K.

  12. University of Florida training reactor. Annual progress report, September 1, 1984-August 31, 1985

    International Nuclear Information System (INIS)

    This annual progress report of the University of Florida Training Reactor discusses: reactor operation; personnel; modifications made to the reactors; reactor maintenance; and testing of reactor systems

  13. Thermionic reactor electric propulsion system requirements.

    Science.gov (United States)

    Mondt, J. F.; Sawyer, C. D.; Schaupp, R. W.

    1972-01-01

    Results of mission analysis, system analysis and mission engineering studies to find a single nuclear electric propulsion (NEP) system which would be applicable for a broad range of unmanned outer planet missions. The NEP system studied uses an in-core nuclear thermionic reactor as the electric power source and mercury bombardment ion engines for propulsion. Many requirements, which are imposed on the NEP system by the mission, were determined from the studies in the process of trying to find a single NEP system for many missions. It is concluded that a single thermionic reactor NEP system could be useful for a broad range of unmanned outer planet missions. The thermionic reactor NEP system should have a power level in the range from 70 to 120 kWe, a system specific weight of approximately 30 kg/kWe, and a full power output capability of 20,000 hr.

  14. Scanning tunneling microscope assembly, reactor, and system

    Energy Technology Data Exchange (ETDEWEB)

    Tao, Feng; Salmeron, Miquel; Somorjai, Gabor A

    2014-11-18

    An embodiment of a scanning tunneling microscope (STM) reactor includes a pressure vessel, an STM assembly, and three spring coupling objects. The pressure vessel includes a sealable port, an interior, and an exterior. An embodiment of an STM system includes a vacuum chamber, an STM reactor, and three springs. The three springs couple the STM reactor to the vacuum chamber and are operable to suspend the scanning tunneling microscope reactor within the interior of the vacuum chamber during operation of the STM reactor. An embodiment of an STM assembly includes a coarse displacement arrangement, a piezoelectric fine displacement scanning tube coupled to the coarse displacement arrangement, and a receiver. The piezoelectric fine displacement scanning tube is coupled to the coarse displacement arrangement. The receiver is coupled to the piezoelectric scanning tube and is operable to receive a tip holder, and the tip holder is operable to receive a tip.

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

    International Nuclear Information System (INIS)

    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.)

  16. Reactor water level control system

    International Nuclear Information System (INIS)

    A BWR type reactor comprises a control valve disposed in a reactor water draining pipelines and undergoing an instruction to control the opening degree, an operation board having a setting device for generating the instruction and a control board for giving the instruction generated by the setting device to the control valve. The instruction is supplied from the setting device to the control valve by way of a control circuit to adjust the opening degree of the control valve thereby controlling the water level in the reactor. In addition, a controller generating an instruction independent of the setting device and a signal transmission channel for signal-transmitting the instruction independent of the control circuit are disposed, to connect the controller electrically to the signal transmission. The signal transmission channel and the control circuit are electrically connected to the control valve switchably with each other. Since instruction can be given to the control valve even at a periodical inspection or modification when the setting device and the control circuit can not be used, the reactor water level can be controlled automatically. Then, operator's working efficiency upon inspection can be improved remarkably. (N.H.)

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

    International Nuclear Information System (INIS)

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

  18. Development of an automatic reactor inspection system

    International Nuclear Information System (INIS)

    Using recent technologies on a mobile robot computer science, we developed an automatic inspection system for weld lines of the reactor vessel. The ultrasonic inspection of the reactor pressure vessel is currently performed by commercialized robot manipulators. Since, however, the conventional fixed type robot manipulator is very huge, heavy and expensive, it needs long inspection time and is hard to handle and maintain. In order to resolve these problems, we developed a new automatic inspection system using a small mobile robot crawling on the vertical wall of the reactor vessel. According to our conceptual design, we developed the reactor inspection system including an underwater inspection robot, a laser position control subsystem, an ultrasonic data acquisition/analysis subsystem and a main control subsystem. We successfully carried out underwater experiments on the reactor vessel mockup, and real reactor ready for Ulchine nuclear power plant unit 6 at Dusan Heavy Industry in Korea. After this project, we have a plan to commercialize our inspection system. Using this system, we can expect much reduction of the inspection time, performance enhancement, automatic management of inspection history, etc. In the economic point of view, we can also expect import substitution more than 4 million dollars. The established essential technologies for intelligent control and automation are expected to be synthetically applied to the automation of similar systems in nuclear power plants

  19. Flexible Conversion Ratio Fast Reactor Systems Evaluation

    International Nuclear Information System (INIS)

    Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. Both concepts have cores treated at 2400 MWt placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchanges coupling a primary coolant to a compact and efficient supercritical CO2 Brayton cycle power conversion system. Decay heat is removed passively using an enhanced Reactor Vessel Auxiliary Cooling System and a Passive Secondary Auxiliary Cooling System. The most important findings were that (1) it is feasible to design the lead-cooled and salt-cooled reactor with the flexible conversion ratio (CR) in the range of CR=0 and CR=1 n a manner that achieves inherent reactor shutdown in unprotected accidents, (2) the salt-cooled reactor requires Lithium thermal Expansion Modules to overcome the inherent salt coolant's large positive coolant temperature reactivity coefficient, (3) the preferable salt for fast spectrum high power density cores is NaCl-Kcl-MgCl2 as opposed to fluoride salts due to its better thermal-hydraulic and neutronic characteristics, and (4) both reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor

  20. Flexible Conversion Ratio Fast Reactor Systems Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Neil Todreas; Pavel Hejzlar

    2008-06-30

    Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. Both concepts have cores reated at 2400 MWt placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchanges coupling a primary coolant to a compact and efficient supercritical CO2 Brayton cycle power conversion system. Decay heat is removed passively using an enhanced Reactor Vessel Auxiliary Cooling System and a Passive Secondary Auxiliary Cooling System. The most important findings were that (1) it is feasible to design the lead-cooled and salt-cooled reactor with the flexible conversion ratio (CR) in the range of CR=0 and CR=1 n a manner that achieves inherent reactor shutdown in unprotected accidents, (2) the salt-cooled reactor requires Lithium thermal Expansion Modules to overcme the inherent salt coolant's large positive coolant temperature reactivity coefficient, (3) the preferable salt for fast spectrum high power density cores is NaCl-Kcl-MgCl2 as opposed to fluoride salts due to its better themal-hydraulic and neutronic characteristics, and (4) both reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor.

  1. Improved Quality in Aerospace Testing Through the Modern Design of Experiments

    Science.gov (United States)

    DeLoach, R.

    2000-01-01

    This paper illustrates how, in the presence of systematic error, the quality of an experimental result can be influenced by the order in which the independent variables are set. It is suggested that in typical experimental circumstances in which systematic errors are significant, the common practice of organizing the set point order of independent variables to maximize data acquisition rate results in a test matrix that fails to produce the highest quality research result. With some care to match the volume of data required to satisfy inference error risk tolerances, it is possible to accept a lower rate of data acquisition and still produce results of higher technical quality (lower experimental error) with less cost and in less time than conventional test procedures, simply by optimizing the sequence in which independent variable levels are set.

  2. Partnership Opportunities with AFRC for Wireless Systems Flight Testing

    Science.gov (United States)

    Hang, Richard

    2015-01-01

    The presentation will overview the flight test capabilities at NASA Armstrong Flight Research Center (AFRC), to open up partnership collaboration opportunities for Wireless Community to conduct flight testing of aerospace wireless technologies. Also, it will brief the current activities on wireless sensor system at AFRC through SBIR (Small Business Innovation Research) proposals, and it will show the current areas of interest on wireless technologies that AFRC would like collaborate with Wireless Community to further and testing.

  3. Standard Test Method for Shear Strength of Fusion Bonded Polycarbonate Aerospace Glazing Material

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    1995-01-01

    1.1 This test method determines the shear yield strength Fsy and shear ultimate strength Fsu of fusion bonds in polycarbonate by applying torsional shear loads to the fusion-bond line. This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

  4. Design and safety consideration in the High-Temperature Engineering Test Reactor (HTTR)

    International Nuclear Information System (INIS)

    The budget for construction of the High-Temperature Engineering Test Reactor (HTTR) was recently committed by the Government in Japan. The HTTR is a test reactor with thermal output of 30 MW and reactor outlet coolant temperature of 950 deg. C at high temperature test operation. The HTTR plant uses a pin-in-block design core and will be used as an experience leading to high temperature applications. Several major important safety considerations are adopted in the design of the HTTR. These are as follows: 1) A coated particle fuel must not be failed during a normal reactor operation and an anticipated operational occurrence; 2) Two independent and diverse reactor shut-down systems are provided in order to shut down the reactor safely and reliably in any condition; 3) Back-up reactor cooling systems which are safety ones are provided in order to remove residual heat of reactor in any condition; 4) Multiple barriers and countermeasures are provided to contain fission products such as a containment, pressure gradient between the primary and secondary cooling circuit and so on, though coated particle fuels contain fission products with high reliability; 5) The functions of materials used in the primary cooling circuit are separated to be pressure-resisting and heat-resisting in order to resolve material problems and maintain high reliability. The detailed design of the HTTR was completed with extensive accumulation of material data and component tests. (author)

  5. Large break LOCA experiment at reactor thermohydraulic test loop

    International Nuclear Information System (INIS)

    The experiments of large break LOCA in the reactor Thermohydraulic Test Loop (UUTR) has been done. The experiments were held at hot leg side by use of accumulator safety injection system and without accumulator. Two experiments were done without activating the high and low pressure safety injection system, while make up system was activated until the experiments were stopped. The test were done at 1 MWt power with about 9,4 kg/sec primary coolant flow rate, and pressure 154 bar. The phenomena were only be limited on the effect of accumulator to the system during LOCA. The results of the experiments indicated a similar system depressurization phenomena for both with and without accumulator was activated. The system trip were happened at very closely different time for both at around 10 t h cycle. The significant difference were that in the experiments were the accumulator was activated, the system depressurization was going more slowly than without the accumulator, and the rods and the fluid temperature were more lower. In the case, the water injection system from the accumulator was able to reduced the rod and the cooling temperature as long as the water inventory is available

  6. Molecular ecology of anaerobic reactor systems

    DEFF Research Database (Denmark)

    Hofman-Bang, H. Jacob Peider; Zheng, D.; Westermann, Peter;

    2003-01-01

    Anaerobic reactor systems are essential for the treatment of solid and liquid wastes and constitute a core facility in many waste treatment plants. Although much is known about the basic metabolism in different types of anaerobic reactors, little is known about the microbes responsible...... to the abundance of each microbe in anaerobic reactor systems by rRNA probing. This chapter focuses on various molecular techniques employed and problems encountered when elucidating the microbial ecology of anaerobic reactor systems. Methods such as quantitative dot blot/fluorescence in-situ probing using various...... and malfunctions of anaerobic digesters occasionally experienced, leading to sub-optimal methane production and wastewater treatment. Using a variety of molecular techniques, we are able to determine which microorganisms are active, where they are active, and when they are active, but we still need to determine...

  7. Technological innovations for FBR reactor cooling system

    International Nuclear Information System (INIS)

    The fast breeder reactor (FBR) is expected to be commercialized early in the 21st century. In order to realize this goal, technological innovations are desired in order to extensively enhance economic performance, and improvement of the reactor cooling system is of primary importance in this regard. Over the past 10 years, Toshiba has developed a succession of new technologies in the field of reactor cooling systems, including a compact type intermediate heat exchanger (IHX), an integral once-through type steam generator (SG), a double-wall-tube type steam generator, and a sodium-immersed high-temperature type electromagnetic pump (EMP). As a synthesis of the fruits of such research and development we have formulated innovative concepts for a reactor cooling system and its constituent components. These advances in research and development activities will significantly contribute to the commercialization of FBRs. (author)

  8. Miniature fiber Bragg grating sensor interrogator (FBG-Transceiver) system for use in aerospace and automotive health monitoring systems

    Science.gov (United States)

    Mendoza, Edgar A.; Kempen, Cornelia; Panahi, Allan; Lopatin, Craig

    2007-09-01

    Fiber Bragg grating sensors (FBGs) have gained rapid acceptance in aerospace and automotive structural health monitoring applications for the measurement of strain, stress, vibration, acoustics, acceleration, pressure, temperature, moisture, and corrosion distributed at multiple locations within the structure using a single fiber element. The most prominent advantages of FBGs are: small size and light weight, multiple FBG transducers on a single fiber, and immunity to radio frequency interference. A major disadvantage of FBG technology is that conventional state-of-the-art fiber Bragg grating interrogation systems are typically bulky and heavy bench top instruments that are assembled from off-the-shelf fiber optic and optical components integrated with a signal electronics board into an instrument console. Based on the need for a compact FBG interrogation system, this paper describes recent progress towards the development of a miniature fiber Bragg grating sensor interrogator (FBG-Transceiver TM) system based on multi-channel integrated optic sensor (InOSense) microchip technology. The hybrid InOSense microchip technology enables the integration of all of the functionalities, both passive and active, of conventional bench top FBG sensor interrogators systems, packaged in a miniaturized, low power operation, 2-cm x 5-cm small form factor (SFF) package suitable for the long-term structural health monitoring in applications where size, weight, and power are critical for operation. The sponsor of this program is NAVAIR under a DOD SBIR contract.

  9. Status and future plan of Japan materials testing reactor

    International Nuclear Information System (INIS)

    The Japan Materials Testing Reactor (JMTR) of Japan Atomic Energy Agency (JAEA) is a light water cooling tank typed reactor. JMTR has been used for fuel and material irradiation studies for LWRs, HTGR, fusion reactor and RI production. Since the JMTR is connected with hot laboratory through the canal, re-irradiation tests can conduct easily by safety and quick transportation of irradiation samples. First criticality was achieved in March 1968, and operation was stopped from August, 2006 for the refurbishment. The reactor facilities are refurbished during four years from the beginning of FY 2007, and necessary examination and work are carrying out on schedule. The renewed and upgraded JMTR will start from FY 2011 and operate for a period of about 20 years (until around FY 2030). The usability improvement of the JMTR, such as higher reactor available factor, shortening turnaround time to get irradiation results, attractive irradiation cost, business confidence, is also discussing as the preparations for re-operation. (author)

  10. Electronically controled mechanical seal for aerospace applications -- Part 1: Design, analysis, and steady state tests

    Science.gov (United States)

    Salant, Richard F.; Wolff, Paul; Navon, Samuel

    1994-01-01

    An electronically-controlled mechanial seal, for use as the purge gas seal in a liquid oxygen turbopump, has been designed, analyzed, and built. The thickness of the lubricating film between the faces is controlled by adjusting the coning of the carbon face. This is done by applying a voltage across a piezoelectric element to which the carbon face is bound. Steady state tests have shown that the leakage rate (and film thickness) can be adjusted over a substantial range, utilizing the available range of voltage.

  11. PC based systems for measurements in reactor physics

    International Nuclear Information System (INIS)

    A set of applications was developed, utilizing the advantages of coupling acquisition cards to PCs, fructifying the experience in reactor physics measurements. The main objectives were: on-line work with advanced codes of gamma spectra analysis, performing on-line corrections for neutron detectors with complex dynamics and dynamic processing of signals from ratemeter and frequency-meter lines. The advanced date processing is carried out at foreground level while the data acquisition and primary processing are carried out at background level. The achieved software covers pulse-height analyzers, ratemeters, frequency meters, multichannel counting ratemeters and noise analysis. By making use of a simple hardware and by increasing the weight of software we improved the performance of generally used acquisition cards, fulfilling, at the same time, the requirements for reactor physics accurate measurements. In-reactor experiments or experiments using neutron sources were used for testing the PC based systems as well as for adjusting their parameters. Extensions to reactor control/safety systems are conceived as developing models systems, because they proved to be versatile tools for testing physics and safety principles. Thus, this work represents an interface between reactor physics and reactor instrumentation and control engineering

  12. [AVIATION MEDICINE: THEORETICAL CONCEPTS AND FOCAL FUNDAMENTAL AND PRACTICAL ISSUES (for the 80th anniversary of the Research Test Center of Aerospace Medicine and Military Ergonomics)].

    Science.gov (United States)

    Zhdanko, I M; Pisarev, A A; Vorona, A A; Lapa, V V; Khomenko, M N

    2015-01-01

    The article discloses postulates of theoretical concepts that make the methodological basis for addressing the real-world aviation medicine challenges of humanizing aviator's environment, labor content and means, and health and performance maintenance. Under consideration are focal fundamental and practical issues arising with the technological progress in aviation and dealt with at the AF CRI Research Test Center of Aerospace Medicine and Military Ergonomics. PMID:26087580

  13. Improving the proliferation resistance of research and test reactors

    International Nuclear Information System (INIS)

    Elimination, or substantial reduction, of the trade in 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. A program is now beginning in the U.S. to develop the necessary fuel technology, but several years of work will be needed. Accordingly, as an immediate interim step, the U.S. is proposing to convert existing research and test reactors (and new designs) from the use of 90-93% enriched fuel to the use of 30-45% enriched fuel wherever this can be done without unacceptable reactor performance degradation

  14. Reactor safety research program at Thai test facility

    International Nuclear Information System (INIS)

    Thermal-hydraulics, Hydrogen, Aerosol and Iodine (Thai) aims at providing experimental database for the verification and validation of Lumped Parameter (Lp) and Computational Fluid Dynamics (CFD) codes with 3-dimensional capabilities. Since its construction in 2000, Thai facility has been engaged in the field of reactor safety in the frame of various national (Thai I: 2000-2003, Thai II: 2003-2006, Thai III: 2006-2009, Thai IV: 2009-2012) and international programs (OECD-Thai: 2007-2009). Additionally, experimental data has been provided for several international standard problems (ISP 41, 46, 47 and 49) code validation exercises. Experiments performed in Thai facility cover a wide spectrum or reactor safety relevant issues by investigating separate and coupled-phenomenon experiments under design basis accident and severe-accident-typical scenarios. Experiments are performed in close co-operation with AREVA Erlangen and Grs Koln. Experimental configuration and the operating conditions in Thai vessel typical of those for PWR, BWR and High Temperature Gas Cooled Reactor can be produced thanks to its modular structure, appropriate feeding/generation devices for gases (H2, He, Steam, N2, etc.), Aerosol (inert and hygroscopic), Iodine Radiotracer, and advanced instrumentation. Experiments also cover investigation of passive safety systems, e.g. commercial Par for H2 mitigation in phenomenon orientated experiments to enhance the confidence in the performance of passive mitigation systems during severe accident scenarios and also to establish a common database accessible by a large research community to support further development and validation of the Lp and CFD codes with 3-dimensional capabilities. This paper summarizes experimental investigations made in Thai test facility to investigate issues related to the thermal-hydraulics, fission product (aerosol, iodine) transport and their interaction with containment walls (deposition, resuspension) and passive safety

  15. Irradiation testing of miniature fuel plates for the RERTR program. [Reduced Enrichment Research and Test Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Senn, R.L.; Martin, M.M.

    1981-07-01

    An irradiation test facility, which provides a test bed for irradiating a variety of miniature fuel plates (miniplates) for the Reduced Enrichment Research and Test Reactors (RERTR) program, has been placed into operation. These tests screen various candidate fuel materials on their suitability for replacing the highly enriched uranium fuel materials currently used by the world's test and research reactors with a lower enrichment fuel material, without significantly degrading reactor operating characteristics and power levels. The use of low uranium enrichment of about 20% /sup 235/U in place of highly enriched fuel for these reactors would reduce the potential for /sup 235/U diversion. The irradiation test facility, designated as HFED, is operating in core position E-7 in the Oak Ridge Research Reactor (ORR), a 30-MW water-moderated reactor. The miniplates will achieve burnups of up to approx. 2.2 x 10/sup 27/ fissions/m/sup 3/ of fuel.

  16. Utilization of fission reactors for fusion engineering testing

    Energy Technology Data Exchange (ETDEWEB)

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

    1985-02-08

    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. Reactor vessel stud closure system

    International Nuclear Information System (INIS)

    A quick-acting stud tensioner apparatus for enabling the loosening or tightening of a stud nut on a reactor vessel stud. The apparatus is adapted to engage the vessel stud by closing a gripper around an upper end of the vessel stud when the apparatus is seated on the stud. Upon lifting the apparatus, the gripper releases the vessel stud so that the apparatus can be removed

  18. Hydraulic stand for testing the models of nuclear reactors

    International Nuclear Information System (INIS)

    Paper describes the basic design and hydraulic characteristics of the MR experimental bench including equipment and coolant circulation circuit, two-loop reactor hydraulic model and measuring system. The bench enables to investigate into thermal and hydraulic characteristics of models of two- and four-loop vessel reactors. Paper dwells upon research and training use of the bench

  19. Recent reactor testing and experience with gamma thermometers

    International Nuclear Information System (INIS)

    Recent experience with gamma thermometers for light water reactors has primarily been in the Framatome reactors operated by Electricite de France. Other recent testing has taken place at Oak Ridge National Laboratory and the Otto Hahn ship reactor. Earlier experience with gamma thermometers was in heavy water reactors at Savannah River and Halden. This paper presents recent data from the light water reactor (LWR) programs. The principles of design and operation of the Radcal gamma thermometer were presented in ''Gamma Thermometer Developments for Light Water Reactors'', Leyse and Smith1. Observations from LWRs confirm the earlier experience from heavy water reactors that the gamma thermometer units give signals which are proportional to the power of surrounding fuel rods and virtually independent of exposure, surrounding poison and other conditions which affect signals of neutron sensitive devices. After 200 sensor-years in EdF reactors, there has been no change in the sensitivity of the devices. Nonetheless, the Radcal units can be recalibrated in-reactor by the introduction of electrical heating via a heater cable imbedded in the device. Algorithms and signal processing software have been developed to interpret and display the gamma thermometer signals. The results of this processing are illustrated here

  20. Reactor calculation benchmark PCA blind test results

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  2. Development of research reactor simulator and its application to dynamic test-bed

    International Nuclear Information System (INIS)

    We developed a real-time simulator for 'High-flux Advanced Neutron Application ReactOr (HANARO), and the Jordan Research and Training Reactor (JRTR). The main purpose of this simulator is operator training, but we modified this simulator into a dynamic test-bed (DTB) to test the functions and dynamic control performance of reactor regulating system (RRS) in HANARO or JRTR before installation. The simulator hardware consists of a host computer, 6 operator stations, a network switch, and a large display panel. The software includes a mathematical model that implements plant dynamics in real-time, an instructor station module that manages user instructions, and a human machine interface module. The developed research reactor simulators are installed in the Korea Atomic Energy Research Institute nuclear training center for reactor operator training. To use the simulator as a dynamic test-bed, the reactor regulating system modeling software of the simulator was replaced by actual RRS cabinet, and was interfaced using a hard-wired and network-based interface. RRS cabinet generates control signals for reactor power control based on the various feedback signals from DTB, and the DTB runs plant dynamics based on the RRS control signals. Thus the Hardware-In-the-Loop Simulation between RRS and the emulated plant (DTB) has been implemented and tested in this configuration. The test result shows that the developed DTB and actual RRS cabinet works together simultaneously resulting in quite good dynamic control performances. (author)

  3. Extended Cooling System for High Power Reactors

    International Nuclear Information System (INIS)

    In-vessel retention (IVR) of core melt is a key severe accident management strategy adopted by some operating nuclear power plants (NPPs) and proposed for advanced light water reactors (LWRs). However, it is not clear that currently proposed external reactor vessel cooling (ERVC) could provide sufficient heat removal for higher power reactors. This paper proposes a dual retention strategy to realize fail-proof defense-in-depth in the APR1400 (Advanced Power Reactor 1400 MWe) and the OPR 1000 (Optimized Power Reactor 1000 MWe). The dual retention has the advantage of IVR-ERVC as well as ex-vessel cooling (EVC) strategies. The multilateral, multidisciplinary project calls for national and international cutting-edge technologies to research and produce (R and P) the D2R2 (Duel Retention Demonstration Reactor) equipped with OASIS (Optimized Advanced Safety Injection System) and ROSIS (Reactor Outer Safety Injection System) to cope with design-basis accidents and beyond in a coherent, continual, comprehensive manner. The enterprise aims to develop the design-basis and severe accident engineering solutions. The enterprise aims to develop the design-basis and severe accident engineering solutions. The former embraces ISAIAH (Injection System Annular Interactive Aero Hydrodynamics) and MESIAH (Methodical Evaluation System Interactive Aero Hydrodynamics). The latter comprises GODIVA (Geo metrics of Direct Injection Versatile Arrangement), SONATA (Simulation of Narrow Annular Thermomechanical Arrest or), TOCATA (Termination of Corium Ablation Thermal Attack) and STRADA (Solution to Reactor Advanced Design Alternatives). D2R2 will contribute to enhancement of both safety and economics for an advanced high power particular and nuclear power in general

  4. Breeder reactor fuel fabrication system development

    International Nuclear Information System (INIS)

    Significant progress has been made in the design and development of remotely operated breeder reactor fuel fabrication and support systems (e.g., analytical chemistry). These activities are focused by the Secure Automated Fabrication (SAF) Program sponsored by the Department of Energy to provide: a reliable supply of fuel pins to support US liquid metal cooled breeder reactors and at the same time demonstrate the fabrication of mixed uranium/plutonium fuel by remotely operated and automated methods

  5. Fast reactor systems for deep sea research

    International Nuclear Information System (INIS)

    Fast reactor (FR) systems have been studied as power units for unmanned bases and research submersibles to monitor various phenomena and as a thermal source for the unmanned base to feed useful microorganisms in the deep sea region. The systems, which are set in pressure hulls, comprise of the FR's and secondary gas loops. Concepts and arrangements of the systems are presented. (author)

  6. In-Research Reactor Tests for SCWR Fuel Verifications

    International Nuclear Information System (INIS)

    The Supercritical water cooled reactors (SCWRs) are essentially light water reactors (LWRs) operating at higher pressure and temperature. The SCWRs achieve high thermal efficiency (i.e., about 45% vs. about 35% efficiency for advanced LWRs) and are simpler plants as the need for many of the traditional LWR components is eliminated. The SCWRs build upon two proven technologies, the LWR and the supercritical coal-fired boiler. The main mission of the SCWR is production of low-cost electricity. Thus the SCWR is also suited for hydrogen generation with electrolysis, and can support the development of the hydrogen economy in the near term. In this paper, the SCWR fuel performance verification tests are reviewed. Based on this review results, in-research reactor verification tests to be performed in a fuel test loop through the international joint program are proposed. In addition, capsule tests and fuel test loop tests to be performed in HANARO are also proposed

  7. Micro-specimen testing techniques for evaluating nuclear reactor materials

    International Nuclear Information System (INIS)

    In the initial construction of nuclear power plant nuclear materials not only have to be high quality in mechanical properties and fracture resistant characteristics, but also considerations have to be given to weakness cause and continued safe operation of power reactor. Recognizing the importance of integrity evaluation test material samples are provided under monitoring program in reactor for evaluation of reactor material property. But because of limited space and necessity of a homogeneous irradiation environment a very limited quantity of micro specimen is provided. The existing test method of toughness property and fracture resistance requires pre-determined size specimen. Therefore, it is very difficult to evaluate those properties by limited micro-specimen provided under monitoring program. In this paper the test technologies of micro-specimen, which can be utilized to evaluate material integrity of reactors in operation, are reviewed. (Hong, J. S.)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-11-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-06-01

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

  13. Tokamak fusion test reactor. Final design report

    International Nuclear Information System (INIS)

    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

  14. Decommissioning of nuclear reactor systems

    International Nuclear Information System (INIS)

    The decision-making process involving the decommissioning of the British graphite-moderated, gas-cooled Magnox power stations is complex. There are timing, engineering, waste disposal, cost and lost generation capacity factors and the ultimate uptake of radiation dose to consider and, bearing on all of these, the overall decision of when and how to proceed with decommissioning may be heavily weighed by political and public tolerance dimensions. These factors and dimensions are briefly reviewed with reference to the ageing Magnox nuclear power stations, of which Berkeley and Hunterston A are now closed down and undergoing the first stages of decommissioning and Trawsfynydd, although still considered as available capacity, has had both reactors closed down since February 1991 and is awaiting substantiation and acceptance of a revised reactor pressure vessel safety case. Although the other first-generation Magnox power station at Hinkley Point, Bradwell, Dungeness and Sizewell are operational, it is most doubtful that these stations will be able to eke out a generating function for much longer. It is concluded that the British nuclear industry has adopted a policy of deferred decommissioning, that is delaying the process of complete dismantlement of the radioactive components and assemblies for at least one hundred years following close-down of the plant. (Author)

  15. Dual-core TRIGA research and materials testing reactor

    International Nuclear Information System (INIS)

    General Atomic Company is under contract from the Romanian Institute for Nuclear Technologies to design, fabricate, and install a research reactor in support of the Romanian National Program for Power Reactor Development. The goal was to select a design concept that provided reasonably high neutron fluxes for long term testing of various fuel-cladding-coolant combinations and also provide high performance pulsing for transient testing of fuel specimens. An effective solution was achieved by the selection of a 14 MW steady-state TRIGA reactor for high flux endurance testing, and an Annular Core Pulsing Reactor (ACPR) for high performance pulsing testing, with both reactors mounted in the same reactor tank and operated independently. The fuel bundles for the steady-state reactor consist of 25 uranium-zirconium hydride rods clad in stainless steel arranged in a square 5 x 5 array. The steady-state core is provided with downflow cooling at a rate of approximately 275 gpm/bundle. Bundle flow tests will be performed with both heated and unheated models. The core will be optimized for peak thermal neutron flux and reactivity lifetime within the constraint of a peak fuel meat temperature of 7500C. The operation of the steady-state reactor at a power level of 14 MW will yield peak unperturbed thermal neutron fluxes in the central experiment position of 2.9 x 1014 n/cm2-sec. The corresponding fast neutron flux (less than 1.125 keV) will be 2.6 x 1014 nv. (U.S.)

  16. Results of the BREST-300 type reactor model fuel elements testing in the IGR reactor

    International Nuclear Information System (INIS)

    Testings of BREST-300 type fast reactor's model fuel elements with nitride fuel in the lead coolant in the central experimental channel of IGR reactor were carried out. In the testing the regime of non-controlled power burst was simulated. In the result of testing the seal failure of fuel elements with 2 % and 10 % 235U enrichment has been occurred, and fragmentation of the part of fuel pellets at interaction with coolant has been taken place. During the reactor testing the measurements and registration of experimental parameters (temperature of fuel, shell, coolant; pressure in fuel elements and testing ampoule; power release in the reactor) were conducted. The physical study of the 'fuel element - ampoule - reactor' was carried out, after-start-up spectrometric and material testing studies, calculated evaluation of temperature fields parameters in the testing ampoule were examined as well. Calculated and experimental values of breaking down specific power releases in the fuel are obtained. The assessment of both fuel fragmentation rate and it character is carried out. Distribution of fuel fragmentation within experimental ampoule volume is studied

  17. Integrated infrastructure initiatives for material testing reactor innovations

    International Nuclear Information System (INIS)

    Highlights: → The EU FP7 MTR+I3 project has initiated a durable cooperation between MTR operators. → Improvements in irradiation test device technology and instrumentation were achieved. → Professional training efforts were streamlined and best practices were exchanged. → A framework has been set up to coordinate and optimize the use of MTRs in the EU. - Abstract: The key goal of the European FP6 project MTR+I3 was to build a durable cooperation between Material Testing Reactor (MTR) operators and relevant laboratories that can maintain European leadership with updated capabilities and competences regarding reactor performances and irradiation technology. The MTR+I3 consortium was composed of 18 partners with a high level of expertise in irradiation-related services for all types of nuclear plants. This project covered activities that foster integration of the MTR community involved in designing, fabricating and operating irradiation devices through information exchange, know-how cross-fertilization, exchanges of interdisciplinary personnel, structuring of key-technology suppliers and professional training. The network produced best practice guidelines for selected irradiation activities. This project allowed to launch or to improve technical studies in various domains dealing with irradiation test device technology, experimental loop designs and instrumentation. Major results are illustrated in this paper. These concern in particular: on-line fuel power determination, neutron screen optimization, simulation of transmutation process, power transient systems, water chemistry and stress corrosion cracking, fission gas measurement, irradiation behaviour of electronic modules, mechanical loading under irradiation, high temperature gas loop technology, heavy liquid metal loop development and safety test instrumentation. One of the major benefits of this project is that, starting from a situation of fragmented resources in a strongly competitive sector, it has

  18. Irradiation of the First Advanced Gas Reactor Fuel Development and Qualification Experiment in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover; David A. Petti

    2008-10-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the United States Department of Energy’s lead laboratory for nuclear energy development. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The experiments, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control for each capsule. The swept gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation. The design of the first experiment (designated AGR-1) was completed in 2005, and the fabrication and assembly of the test train as well as the support systems and fission product monitoring system that monitor and control the experiment during irradiation were completed in September 2006. The experiment was inserted in the ATR in December 2006, and is serving as a shakedown test of the multi-capsule experiment design that will be used in the subsequent irradiations as well as a test of the early variants of the fuel produced under this program. The experiment test train as well as the monitoring, control, and data collection systems are discussed and the status of the experiment is provided.

  19. Acoustic emission for on-line reactor monitoring: results of intermediate vessel test monitoring and reactor hot functional testing

    International Nuclear Information System (INIS)

    This article discusses a program designed to develop the use of acoustic emission (AE) methods for continuous surveillance to detect and evaluate flaw growth in reactor pressure boundaries. Technology developed in the laboratory for identifying AE from crack growth and for using AE information to estimate flaw severity is now being evaluated on an intermediate vessel test and on a reactor facility. A vessel, designated ZB-1, has been tested under fatigue loadings with simulated reactor conditions at Mannheim, West Germany, in collaboration with the German Materialpruefungsanstalt (MPA), Stuttgart. Fatigue cracking from machined flaws and in a fabrication weld were both detected clearly by AE. AE data were measured on a US nuclear reactor (Watts Bar, Unit 1) during hot functional preservice testing. This demonstrated that coolant flow noise is a manageable problem and that AE can be detected under operational coolant flow and temperature conditions. (author)

  20. Static tests of the propulsion system. [Propfan Test Assessment program

    Science.gov (United States)

    Withers, C. C.; Bartel, H. W.; Turnberg, J. E.; Graber, E. J.

    1987-01-01

    Advanced, highly-loaded, high-speed propellers, called propfans, are promising to revolutionize the transport aircraft industry by offering a 15- to 30-percent fuel savings over the most advanced turbofans without sacrificing passenger comfort or violating community noise standards. NASA Lewis Research Center and industry have been working jointly to develop the needed propfan technology. The NASA-funded Propfan Test Assessment (PTA) Program represents a key element of this joint program. In PTA, Lockheed-Georgia, working in concert with Hamilton Standard, Rohr Industries, Gulfstream Aerospace, and Allison, is developing a propfan propulsion system which will be mounted on the left wing of a modified Gulfstream GII aircraft and flight tested to verify the in-flight characteristics of a 9-foot diameter, single-rotation propfan. The propfan, called SR-7L, was designed and fabricated by Hamilton Standard under a separate NASA contract. Prior to flight testing, the PTA propulsion system was static tested at the Rohr Brown Field facility. In this test, propulsion system operational capability was verified and data was obtained on propfan structural response, system acoustic characteristics, and system performance. This paper reports on the results of the static tests.

  1. Exploration of a Capability-Focused Aerospace System of Systems Architecture Alternative with Bilayer Design Space, Based on RST-SOM Algorithmic Methods

    OpenAIRE

    2014-01-01

    In defense related programs, the use of capability-based analysis, design, and acquisition has been significant. In order to confront one of the most challenging features of a huge design space in capability based analysis (CBA), a literature review of design space exploration was first examined. Then, in the process of an aerospace system of systems design space exploration, a bilayer mapping method was put forward, based on the existing experimental and operating data. Finally, the feasibil...

  2. T2 Response Time Analysis and Test of Nuclear Power Plant Reactor Protection System%核电厂数字化反应堆保护系统T2响应时间分析及测试

    Institute of Scientific and Technical Information of China (English)

    马刚; 康礼鸿

    2015-01-01

    Reactor Protection System is very important safety system for nuclear power plant DCS I&C system. For the safety of nuclear power plant, the response time of RPS has the strict requisition, ,therefore it is necessary to evaluate the response time of RPS. In this paper the structure of nuclear power plant reactor protection system is briefly introduced, the scope of T2 response time test is given, and theoretical analysis of the response time of RPS is conducted. Test method for T2 response time is proposed,test principle of T2 Response time is established. How to test the response time is also introduced by using the VP link as the test devices. By taking the conditions for response time test of reactor trip due to SG1 Low-Low level of NPP Unit 1&2 as example, the actual response time executed in project is introduced in detail, then the test document about response time test isoutput, and the test result is recorded and analyzed.%反应堆保护系统是核电厂数字化仪表控制系统中重要的安全系统,是DCS的重要组成部分。为了核电站的安全,对保护系统的响应时间有严格的要求,有必要对响应时间进行评价,本文简要介绍了核电站反应堆保护系统的结构,给出了T2响应时间测试范围,并对反应堆保护系统的响应时间进行理论分析,给出了T2响应时间测试方法,建立了响应时间测试原理,介绍了VP Link作为测试装置如何进行响应时间测试。以某核电厂1&2机组的SG1水位低低导致紧急停堆响应时间测试工况为例,详细介绍了实际响应时间测试工作,给出了响应时间测试的输出文件清单,并对测试结果进行记录和分析。

  3. The ''CAMERA'' test facility in the OSIRIS reactor

    International Nuclear Information System (INIS)

    CAMERA is an irradiation installation conceived to measure under neutronic flux and continuously the dimension variations of a fuel pencil of PWR reactors. The device, set in the periphery of the OSIRIS reactor, can receive new, preirradiated or reconstituted pencils. The principles of measurements is explained. Then, a brief description of the installation is given: in-pile part; out-of-pile part; connections. The technical characteristics of the installation are presented. A first qualification test of the installation under flux has been carried out at the end of the first semester 1984 in the OSIRIS reactor

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

    Energy Technology Data Exchange (ETDEWEB)

    Wright, Steven A.; Lipinski, Ronald J.; Vernon, Milton E.; Sanchez, Travis

    2006-04-01

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

  5. Laser fusion power reactor system (LFPRS)

    Energy Technology Data Exchange (ETDEWEB)

    Kovacik, W. P.

    1977-12-19

    This report gives detailed information for each of the following areas: (1) reference concept description, (2) nuclear design, (3) structural design, (4) thermal and fluid systems design, (5) materials design and analysis, (6) reactor support systems and balance of plant, (7) instrumentation and control, (8) environment and safety, (9) economics assessment, and (10) development requirements. (MOW)

  6. Principles of the reactor code system RHEIN

    International Nuclear Information System (INIS)

    A description is given of the principles of the reactor code system RHEIN which is applied in connection with a BESM6-type computer. In transfering data between the components of the system external storage is used. The programme passage is controlled by the input data. (author)

  7. Laser fusion power reactor system (LFPRS)

    International Nuclear Information System (INIS)

    This report gives detailed information for each of the following areas: (1) reference concept description, (2) nuclear design, (3) structural design, (4) thermal and fluid systems design, (5) materials design and analysis, (6) reactor support systems and balance of plant, (7) instrumentation and control, (8) environment and safety, (9) economics assessment, and (10) development requirements

  8. Software system for reactor physics analyses

    International Nuclear Information System (INIS)

    The paper presents the working stage of the development of the HEXAB-3DI - RADMAGRU Code System for calculation of important neutron physics characteristics in the WWER-1000 reactor cores. It gives a notion about the system functions and structure, as well as the new organization of calculation and feedback procedures. (author)

  9. The reactor core analysis code CITATION-1000VP for High Temperature Engineering Test Reactor

    International Nuclear Information System (INIS)

    Reactor core analysis with full core model has been necessary for the High Temperature Engineering Test Reactor (HTTR) design. The CITATION-1000VP code has been developed to enable reactor core analysis of HTTR with full core model through extending the number of zones and meshes, and enhancing the calculation speed of CITATION code. This report describes the program changes for extending the number of zones and meshes, and for vectorization. The maximum number of zones and meshes becomes 999 and 500, respectively. The calculation speed is enhanced up to 21 times. (author)

  10. Electromechanical drive for a reactor control system

    International Nuclear Information System (INIS)

    The invention is related to control systems of nuclear researche swimming pool-type reactors. The presented electromechanical drive for a nuclear reactor control system consists of an electromagnetic grip of control element with magnet power supply cable, drum and flexible element, e.g., wire rope. Two branches of the rope which are separated from the electromagnet to the core and the drum form the closed loop. To decrease the dimensions of the drive, the magnet power supply cable serves as a loop flexible element which goes from the electromagnet to the core. For a particular reactor the drive, made according to the invention is 100 mm shorter and 20 mm narrower as compared with the known one, and that is rather significant in cases when a drive is to be installed directly on a control system channel

  11. Management system requirements for small reactors

    International Nuclear Information System (INIS)

    This abstract identifies the management system requirements for the life cycle of small reactors from initial conception through completion of decommissioning. For small reactors, the requirements for management systems remain the same as those for 'large' reactors regardless of the licensee' business model and objectives. The CSA N-Series of standards provides an interlinked set of requirements for the management of nuclear facilities. CSA N286 provides overall direction to management to develop and implement sound management practices and controls, while other CSA nuclear standards provide technical requirements and guidance that support the management system. CSA N286 is based on a set of principles. The principles are then supported by generic requirements that are applicable to the life cycle of nuclear facilities. CNSC regulatory documents provide further technical requirements and guidance. (author)

  12. 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.

  13. Integral test of JENDL-3.3 for fast reactors

    International Nuclear Information System (INIS)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-02-01

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

  15. Testing of rubber O-rings for R-5 reactor

    International Nuclear Information System (INIS)

    This paper summarises the results of various experiments and tests conducted for the selection of suitable O-rings to be used in R-5 reactor. O-rings of various elastomeric compositions obtained from different manufacturers were tested for heat aging, fluid aging, radiation stability and specific gravity. They were irradiated to various dose levels at Apsara Reactor. The changes in axial thickness and hardness, after each test, were measured and results were correlated. The tests reveal that O-rings made of ethylene-propylene rubber are the best suited for the use in R-5 reactor. The O-rings made of nitrile rubber are also good. Neoprene rubber O-rings were found unsuitable mainly because of their low radiation resistance. (author)

  16. Monitoring test piece for a reactor pressure vessel

    International Nuclear Information System (INIS)

    Purpose: To obtain a test piece capable of measurement for neutron exposure ranging 0.1 -- 2 MeV in a reactor pressure vessel Constitution: Fissionable materials causing nuclear fission by fast neutrons are contained within a sealed container in addition to a test piece for monitoring the change in the mechanical properties and a monitor wire for measuring the neutron dose. If uranium 238 and thorium 232 are selected as the fissionable materials, for instance, they cause nuclear fission by the reaction with neutrons of higher than about 2 MeV and about 0.2 MeV respectively. Then, after the stop of the reactor operation, the monitoring test piece is taken out from the reactor pressure vessel to determine the radioactivity, whereby the neutron dose within the energy range of 0.1 - 2 MeV applied to the fissionable materials of the test piece can be estimated with ease. (Horiuchi, T.)

  17. Content-Based Document Recommender System for Aerospace Grey Literature: System Design

    Directory of Open Access Journals (Sweden)

    K. Nageswara Rao

    2011-06-01

    Full Text Available Recorded knowledge in the form of manuscripts, print documents, microforms, CD-ROMs, computer files, etc., is increasing exponentially. In order to locate and access relevant information from this vast amount of literature, efforts have been made from time and on to develop various tools and techniques. Early evolved tools/techniques include: library catalogues, indexes, concordances and so on. Information Retrieval Systems played a vital role in the field of information & librarianship to find relevant information from vast number of documents. Recently intelligent agents gained importance as they are able to query databases and resources on Internet, remote library catalogs thereby reducing information overload on the user. Another technology that alleviates the information overload problem is the filtering or recommender systems. The purpose of development of recommender systems is to provide useful and most relevant recommendations or suggestions from number of available alternatives. The present study aimed at design & development of Content-based Document Recommender System (CODORS to retrieve most relevant technical documents without necessarily matching title terms and closely related to a particular search term(s as opposed to general Online Public Access Catalog (OPAC search results. The developed CODORS converts terms expressed by the user in natural language automatically into subject descriptors, carry on search, ranks and retrieves documents. http://dx.doi.org/10.14429/djlit.31.3.1046

  18. Hardware Specific Integration Strategy for Impedance-Based Structural Health Monitoring of Aerospace Systems

    Science.gov (United States)

    Owen, Robert B.; Gyekenyesi, Andrew L.; Inman, Daniel J.; Ha, Dong S.

    2011-01-01

    The Integrated Vehicle Health Management (IVHM) Project, sponsored by NASA's Aeronautics Research Mission Directorate, is conducting research to advance the state of highly integrated and complex flight-critical health management technologies and systems. An effective IVHM system requires Structural Health Monitoring (SHM). The impedance method is one such SHM technique for detection and monitoring complex structures for damage. This position paper on the impedance method presents the current state of the art, future directions, applications and possible flight test demonstrations.

  19. Balancing passive and active systems for evolutionary water cooled reactors

    International Nuclear Information System (INIS)

    Advanced concepts of the water-cooled reactors are intended to improve safety, economics and public perception of nuclear power. The potential inclusion of new passive means in addition or instead of traditional active systems is being considered by nuclear plant designers to reach these goals. With respect to plant safety, application of the passive means is mainly intended to simplify the safety systems and to improve their reliability, to mitigate the effect of human errors and equipment malfunction. However, some clear drawbacks and the limited experience and testing of passive systems may raise additional questions that have to be addressed in the design process for each advanced reactor. Therefore the plant designer should find a reasonable balance of active and passive means to effectively use their advantages and compensate their drawbacks. Some considerations that have to be taken into account when balancing active/passive means in advanced water-cooled reactors are discussed in this paper. (author)

  20. Digital instrumentation system for nuclear research reactors

    International Nuclear Information System (INIS)

    This work describes a proposal for a system of nuclear instrumentation and safety totally digital for the Argonauta Reactor. The system divides in the subsystems: channel of pulses, channel of current, conventional instrumentation and safety system. The connection of the subsystems is made through redundant double local net, using the protocol modbus/rtu. So much the channel of pulses, the current channel and safety's system use modules operating in triple redundancy. (author)

  1. Upgraded reactor systems for enhanced safety at TRIGA-INR

    International Nuclear Information System (INIS)

    After almost three decades of operation of stationary TRIGA 14MW with systems provided and installed at reactor first start-up, it appeared obvious that an extended modernization program is required, both for enhancing the nuclear safety and to expand the facility lifetime. A first step has been achieved through complete HEU to LEU core conversion, meaning also core refuelling possibility for the future. Systems that have been subjected to the upgrading program are: control rods, radiation monitoring, data acquisition and processing, ventilation, irradiation devices, and above all, the outstanding modernization of the I and C system, including a brand new reactor control desk. Taking into account own and research reactors community operation experience, IAEA guides and recommendations, the basic requirement for the Instrumentation and Control System is the separation between safety and operation components, in order to decrease human error consequences and avoid common cause failures. Modernization did not cover any sensor replacement, but preserve the present scram logic and conditions (as given and approved in the Safety Report and Licensed Limits and Conditions) The entire modernization program is performed according to QA system. Out of intrinsic nuclear safety enhancement, enhanced population and environment protection is a concern and an expected result of the program. Upgrading the overall performances of the reactor and extending its operational lifetime, the Reactor Department of Institute will be able to perform competitive irradiation tests for nuclear fuel and materials, and to continue to develop nuclear investigation techniques or isotope production. (author)

  2. The Search for Nonflammable Solvent Alternatives for Cleaning Aerospace Oxygen Systems

    Science.gov (United States)

    Mitchell, Mark; Lowrey, Nikki

    2012-01-01

    Oxygen systems are susceptible to fires caused by particle and nonvolatile residue (NVR) contaminants, therefore cleaning and verification is essential for system safety. . Cleaning solvents used on oxygen system components must be either nonflammable in pure oxygen or complete removal must be assured for system safety. . CFC -113 was the solvent of choice before 1996 because it was effective, least toxic, compatible with most materials of construction, and non ]reactive with oxygen. When CFC -113 was phased out in 1996, HCFC -225 was selected as an interim replacement for cleaning propulsion oxygen systems at NASA. HCFC-225 production phase-out date is 01/01/2015. HCFC ]225 (AK ]225G) is used extensively at Marshall Space Flight Center and Stennis Space Center for cleaning and NVR verification on large propulsion oxygen systems, and propulsion test stands and ground support equipment. . Many components are too large for ultrasonic agitation - necessary for effective aqueous cleaning and NVR sampling. . Test stand equipment must be cleaned prior to installation of test hardware. Many items must be cleaned by wipe or flush in situ where complete removal of a flammable solvent cannot be assured. The search for a replacement solvent for these applications is ongoing.

  3. Dose management in decommissioning the PLUTO Materials Testing Reactor at Harwell

    International Nuclear Information System (INIS)

    This paper outlines the aspects of decommissioning small and medium sized facilities, which lead to dose management problems. The dose management system, consisting of a work management data base and local dose control system developed for the decommissioning of PLUTO materials testing reactor at AEA Harwell is described. The effectiveness of the system and future developments are discussed. (author)

  4. Reactor pressure vessel design of the high temperature engineering test reactor

    International Nuclear Information System (INIS)

    The reactor pressure vessel (RPV) of the HTTR is 5.5m (inside diameter), 13.2m (inside height), and 122mm (shell thickness). The RPV contains core components, reactor internals, reactivity control system, etc.2 1/4Cr-1Mo steel is chosen as the material for RPV. The temperature reaches about 400 deg. C at normal operation. The fluence of the RPV is estimated to be less than 1 x 1017n/cm2 (E > 1MeV) and so irradiation embrittlement is negligible, but temper embrittlement is not negligible. For the purpose of reducing embrittlement, content of some elements must be limited in the 2 1/4Cr-1Mo steel for the RPV; embrittlement parameters, J-factor and X-bar are used.In this paper, design and structure of the RPV are reviewed first. Fabrication procedure of the RPV and its special feature are described. Material data on the 2 1/4Cr-1Mo steel manufactured for the RPV, especially the embrittlement parameters, J-factor and X-bar , and nil-ductility transition temperatures, TNDT, by drop weight tests, are shown. In-service inspection and results of R and Ds are also described

  5. Reactor fault simulation at the closure of the Windscale advanced gas-cooled reactor: analysis of reactor transient tests

    International Nuclear Information System (INIS)

    The testing of fault transient analysis methods by direct simulation of fault sequences on a commercial reactor is clearly excluded on safety and economic grounds. The closure of the Windscale prototype advanced gas-cooled reactor (WAGR) therefore offered a unique opportunity to test fault study methods under extreme conditions relatively unfettered by economic constraints, although subject to appropriate safety regulations. One aspect of these important experiments was a series of reactor transient tests. The objective of these reactor transients was to increase confidence in the fault study computer models used for commercial AGR safety assessment by extending their range of validation to cover large amplitude and fast transients in temperature, power and flow, relevant to CAGR faults, and well beyond the conditions achievable experimentally on commercial reactors. A large number of tests have now been simulated with the fault study code KINAGRAX. Agreement with measurement is very good and sensitivity studies show that such discrepancies as exist may be due largely to input data errors. It is concluded that KINAGRAX is able to predict steady state conditions and transient amplitudes in both power and temperature to within a few percent. (author)

  6. 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.

  7. Reduced enrichment for research and test reactors: Proceedings

    International Nuclear Information System (INIS)

    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

  8. Mobile leak testing system

    International Nuclear Information System (INIS)

    The design and implementation are described of a mobile testing unit ULTRATEST M for helium leak tests. The equipment has been developed by Leybold-Heraeus GmbH in Cologne and is in-built in a Mercedes-Benz 208 van. The equipment is designed for the operative use in assembly and construction of nuclear power plants and its throughput is sufficient for checking the whole upper reactor block. It may also be used for removing defects of vacuum equipment requiring a high level of tightness or equally demanding equipment used in the chemical industry. Experience with the equipment is described. (B.S.)

  9. Investigation of the heavy water distillation system at the RA reactor

    International Nuclear Information System (INIS)

    The heavy water distillation system was tested because this was not done before the reactor start-up. Detailed inspection of the system components showed satisfactory results. Leak testing was done as well as the testing of the instrumentation which enables reliable performance of the system. Performance testing was done with ordinary water and later 2700 l of heavy water from the reactor was purified, decreasing the activity by 45%

  10. U.S. aerospace industry opinion of the effect of computer-aided prediction-design technology on future wind-tunnel test requirements for aircraft development programs

    Science.gov (United States)

    Treon, S. L.

    1979-01-01

    A survey of the U.S. aerospace industry in late 1977 suggests that there will be an increasing use of computer-aided prediction-design technology (CPD Tech) in the aircraft development process but that, overall, only a modest reduction in wind-tunnel test requirements from the current level is expected in the period through 1995. Opinions were received from key spokesmen in 23 of the 26 solicited major companies or corporate divisions involved in the design and manufacture of nonrotary wing aircraft. Development programs for nine types of aircraft related to test phases and wind-tunnel size and speed range were considered.

  11. Conceptual design of a uranyl nitrate fueled reactor for the destructive testing of liquid metal fast breeder reactor fuel subassemblies

    International Nuclear Information System (INIS)

    A preliminary design of a uranyl nitrate test reactor is developed, with emphasis placed on the core neutronics and cross section development. ENDF/B-IV cross section data and the AMPX system were used to develop a 25 group neutron cross section library. A series of one-dimensional transport calculations were made in order to arrive at a reference design. Power densities of 16.5 Kw/1 appear to be attainable in the 217 pin FFTF test subassembly, with a peak neutron flux in the test zone of 2.4 x 1014 n/cm2-sec. Other engineering features pertinent to the overall system design are discussed, including: (1) corrosion, (2) treatment of radiolytic gas, (3) heat removal, and (4) reactor control

  12. Development of a fiber optic health monitoring system for aerospace applications

    Institute of Scientific and Technical Information of China (English)

    2007-01-01

    This paper describes our research activity involved in the identification, development and test of a prototype SHM system constituted by optical sensing nodes to measure both temperature and strain on ultra high temperature ceramics (UHTC) materials up to 1000 ℃. Commercially available optic devices can operate up to 550 ℃. To raise temperature limit up to 1000 ℃, custom devices, mainly under development for scientific applications, have been identified. A prototype SHM system has been developed adopting a FBG sensor for temperature measurement and an EFPI sensor in sapphire fiber for strain measurement. The preliminary findings from thermo-mechanical tests indicate that former SHM system is capable of accurately measuring strain at elevated temperatures on UHTC materials.

  13. Operating safety experience of fast breeder test reactor

    International Nuclear Information System (INIS)

    Full text: Operational safety criteria for nuclear reactors are very stringent and it is essential to incorporate adequate inherent and engineered safety features in the design to ensure safe operation of the reactor. Commissioning and operation of FBTR, being first of its kind in India based on nuclear chain reaction maintained by fast neutrons and use of high temperature liquid sodium as coolant, was a challenging task. Safe operation of the reactor for the past 17 years with good performance of sodium systems and the indigenous plutonium rich carbide fuel, touching a burn up level of 100 GWd/t has underlined the high level of design and operation competence achieved

  14. Reactor building integrity testing: A novel approach at Gentilly 2 - principles and methodology

    International Nuclear Information System (INIS)

    In 1987, Hydro-Quebec embarked on an ambitious development program to provide the Gentilly 2 nuclear power station with an effective, yet practical reactor building Integrity Test. The Gentilly 2 Integrity Test employs an innovative approach based on the reference volume concept. It is identified as the Temperature Compensation Method (TCM) System. This configuration has been demonstrated at both high and low test pressure and has achieved extraordinary precision in the leak rate measurement. The Gentilly 2 design allows the Integrity Test to be performed at a nominal 3 kPa(g) test pressure during an (11) hour period with the reactor at full power. The reactor building Pressure Test by comparison, is typically performed at high pressure 124 kPa(g)) in a 7 day window during an annual outage. The Integrity Test was developed with the goal of demonstrating containment availability. Specifically it was purported to detect a leak or hole in the 'bottled-up' reactor building greater in magnitude than an equivalent pipe of 25 mm diameter. However it is considered feasible that the high precision of the Gentilly 2 TCM System Integrity Test and a stable reactor building leak characteristic will constitute sufficient grounds for the reduction of the Pressure Test frequency. It is noted that only the TCM System has, to this date, allowed a relevant determination of the reactor building leak rate at a nominal test pressure of 3 kPa(g). Classical method tests at low pressure have lead to inconclusive results due to the high lack of precision

  15. Compilation and development of K-6 aerospace materials for implementation in NASA spacelink electronic information system

    Science.gov (United States)

    Blake, Jean A.

    1987-01-01

    Spacelink is an electronic information service to be operated by the Marshall Space Flight Center. It will provide NASA news and educational resources including software programs that can be accessed by anyone with a computer and modem. Spacelink is currently being installed and will soon begin service. It will provide daily updates of NASA programs, information about NASA educational services, manned space flight, unmanned space flight, aeronautics, NASA itself, lesson plans and activities, and space program spinoffs. Lesson plans and activities were extracted from existing NASA publications on aerospace activities for the elementary school. These materials were arranged into 206 documents which have been entered into the Spacelink program for use in grades K-6.

  16. Performance and testing of refractory alloy clad fuel elements for space reactors

    International Nuclear Information System (INIS)

    Two fast reactor irradiation tests, SP-1 and SP-2, provide a unique and self-consistent data set with which to evaluate the technical feasibility of potential fuel systems for the SP-100 space reactor. Fuel pins fabricated with leading cladding candidates (Nb-1Zr, PWC-11, and Mo-13Re) and fuel forms (UN and UO2) are operated at temperatures typical of those expected in the SP-100 design. The first US fast reactor irradiated, refractory alloy clad fuel pins, from the SP-1 test, reached 1 at .% burnup in EBR-II in March 1985. At that time selected pins were discharged for interim examination. These examinations confirmed the excellent performance of the Nb-1Zr clad uranium oxide and uranium nitride fuel elements, which are the baseline fuel systems for two SP-100 reactor concepts

  17. Sensitivity Test of RIA in 5MW Research Reactor during Startup

    International Nuclear Information System (INIS)

    During startup operation, control absorber rods are not located in the equal critical position since they can be manually controlled by an operator without position limitation. Therefore, the power peaking factor in this control mode becomes larger due to the skewed power shape, making the accident consequence worse. In research reactor, the reactor protection system (RPS) has linear power trip and power lograte trip for a safe shutdown of reactor in the accident, and the occurrence of those trips depend both on the initial reactor power and the reactivity insertion rate. Therefore, with a series of sensitive analyses, we identified the most severe combination of initial conditions among the various initial reactor powers and reactivity insertion rates. The model reactor in this analysis is a 5MW pooltype research reactor having two different operation modes; a power operation and a training operation.. Since the accident occurs during startup of the reactor, the training mode without a forced convection results in more severe consequences in a view of fuel integrity. Therefore, the inadvertent withdrawal of a control rod during a startup of training operation is analyzed as a limiting case of the accident. Sensitivity tests with combinations of different initial reactor powers and reactivity insertion rates are performed for an inadvertent CAR withdrawal during startup of the training operation

  18. Experimental capabilities of the transient reactor test (TREAT) facility

    International Nuclear Information System (INIS)

    The TREAT facility was designed and built in the 1950s to provide a transient reactor for conducting safety experiments on reactor fuels. Throughout its almost 40-year history, it has proven to be a safe, reliable, and versatile facility, compiling a distinguished record of successful experiments. Several major improvements to the facility have been made, including an expansion of the building and of equipment handling capability, and enlargement of the access hole above the core, rearrangement of the reactor's control rods to provide more-uniform flux profiles, installation of improved reactor computer-control systems, a feedback system that safely allows real-time changes in power transients depending upon events occurring in the experiment, and several upgrades in the fast neutron hodoscope for improved experiment-fuel-motion diagnostics. The original TREAT fuel is still in use, however, since it appears to have no degradation from its many years of service

  19. Inductive testing of reactor pressure vessels

    International Nuclear Information System (INIS)

    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)

  20. Preliminary design studies on the Broad Application Test Reactor

    International Nuclear Information System (INIS)

    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

  1. Review of Operation and Maintenance Support Systems for Research Reactors

    International Nuclear Information System (INIS)

    Operation support systems do not directly control the plant but it can aid decision making itself by obtaining and analyzing large amounts of data. Recently, the demand of research reactor is growing and the need for operation support systems is increasing, but it has not been applied for research reactors. This study analyzes operation and maintenance support systems of NPPs and suggests appropriate systems for research reactors based on analysis. In this paper, operation support systems for research reactors are suggested by comparing with those of power reactors. Currently, research reactors do not cover special systems in order to improve safety and operability in comparison with power reactors. Therefore we expect to improve worth to use by introducing appropriate systems for research reactors. In further research, we will develop an appropriate system such as applications or tools that can be applied to the research reactor

  2. Reactor shutdown system of prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Full text: The shutdown system of PFBR is designed to assure a very high reliability by employing well known principles of redundancy, diversity and independence. The failure probability of the shutdown system limited to -6/ ry. Salient features of the shutdown system are: Two independent shutdown systems, each of them able to accommodate an additional single failure and made up of a trip system and an associated absorber rod group. Diversity between trip systems, rods and mechanisms. Initiation of SCRAM by two diverse physical parameters of the two shutdown systems for design events leading potentially to unacceptable conditions is the core. The first group of nine rods called control and safety rods (CSR) is used for both shutdown as well as power regulation. The second group consisting of three rods known as diverse safety rods (DSR) is used only for shutdown. Diversity between the two groups is ensured by varying the operating conditions of the electromagnets and the configurations of the mobile parts. The reactivity worth of the absorber rods have been chosen such that each group of rods would ensure cold shutdown on SCRAM even when the most reactive rod of the group fails to drop. Together the two groups ensure a shutdown margin of 5000 pcm. The speed and individual rod worth of the CSR is chosen from operational and safety considerations during reactor start up and raising of power. Required drop time of rods during SCRAM depends on the incident considered. For a severe reactivity incident of 3 $/s this has to be limited to 1s and is ensured by limiting electromagnet response time and facilitating drop by gravity. Design safety limits for core components have been determined and SCRAM parameters have been identified by plant dynamic analysis to restrict the temperatures of core components within the limits. The SCRAM parameters are distributed between the two systems appropriately. Fault tree analysis of the system has been carried out to determine the

  3. An Overview of Performance Characteristics, Experiences and Trends of Aerospace Engine Bearings Technologies

    Institute of Scientific and Technical Information of China (English)

    Ebert Franz-Josef

    2007-01-01

    In this paper, the operating conditions, technical requirements, performance characteristics, design ideas, application experiences and development trends of aerospace engine bearings, including material technology, integration design and reliability, are reviewed. The development history of aerospace engine bearing is recalled briefly at first. Then today's material technologies and the high bearing performances of the bearings obtained through the new materials are introduced, which play important rolls in the aeroengine bearing developments. The integration design ideas and practices are explained to indicate its significant advantages and importance to the aerospace engine bearings. And the reliability of the shaft-bearing system is pointed out and treated as the key requirement with goals for both engine and bearing. Finally, as it is believed that the correct design comes from practice, the pre-qualification rig testing conducted by FAG Aerospace GmbH & Co. KG is briefly illustrated as an example. All these lead to the development trends of aerospace engine bearings from different aspects.

  4. Quantification of structural materials for reactor systems: synergy's in materials for fusion/fission reactors and advanced fission reactor

    International Nuclear Information System (INIS)

    In nuclear technology a lot of experience has been accumulated meanwhile from reactor programmes for ferritic alloys, austenitic steels and Ni-based alloys as main component materials during R and D, design, construction and operation. Generally materials are a key issue for a safe and reliable operation of -NPPs. Many grades investigated are of interest for the design of GenIVs and fusion reactors. Synergisms of materials, material technologies, mechanical data, corrosion and other topics -for the qualification of materials for nuclear systems are generally discussed and information on a qualification procedure is compiled. Also some lessons learned from fabrication, test programmes or operation of NPPs are provided. A special problem is the fusion system because a final validation for alloy performance in the long term will need irradiation under realistic -fusion condition anticipated in a high-energetic, fusion-specific intense neutron source such as (IFMIF), the International Fusion Materials Irradiation Facility. (author)

  5. Functional description of the dynamically operating reactor protection system

    International Nuclear Information System (INIS)

    For detection of accidents and irritation of engineering safety systems there is required a reactor protection system meeting highest demands for reliability. These demands are met by redundant systems where all faults are automatically recorded or detected by means of testing. Clear design of circuits, good maintenance properties, and extensive inspections and acceptance tests of subassemblies and entire circuit will also contribute to make the system reliable. In detail there is presented: functional disposition, analog, logic and relay (generating output signals) sections, types of subassemblies, elementary diagrams. (orig.)

  6. Completely modular Thermionic Reactor Ion Propulsion System (TRIPS)

    Science.gov (United States)

    Peelgren, M. L.; Kikin, G. M.; Sawyer, C. D.

    1972-01-01

    The nuclear reactor powered ion propulsion system described is an advanced completely modularized system which lends itself to development of prototype and/or flight type components without the need for complete system tests until late in the development program. This modularity is achieved in all of the subsystems and components of the electric propulsion system including (1) the thermionic fuel elements, (2) the heat rejection subsystem (heat pipes), (3) the power conditioning modules, and (4) the ion thrusters. Both flashlight and external fuel type in-core thermionic reactors are considered as the power source. The thermionic fuel elements would be useful over a range of reactor power levels. Electrical heated acceptance testing in their flight configuration is possible for the external fuel case. Nuclear heated testing by sampling methods could be used for acceptance testing of flashlight fuel elements. The use of heat pipes for cooling the collectors and as a means of heat transport to the radiator allows early prototype or flight configuration testing of a small module of the heat rejection subsystem as opposed to full scale liquid metal pumps and radiators in a large vacuum chamber. The power conditioner (p/c) is arranged in modules with passive cooling.

  7. A stochastic study of coupled reactor systems

    International Nuclear Information System (INIS)

    The neutronic behaviour of a system of two loosely coupled reactor cores is investigated on the basis of a stochastic formulation, by the development of a four points model. The mathematical development is explained. Both real and imaginary parts of the core-to-core neutron cross sepctral density show good agreement with those reported for expermental noise. (U.K.)

  8. Hybrid Molten Salt Reactor (HMSR) System Study

    Energy Technology Data Exchange (ETDEWEB)

    Woolley, Robert D [PPPL; Miller, Laurence F [PPPL

    2014-04-01

    Can the hybrid system combination of (1) a critical fission Molten Salt Reactor (MSR) having a thermal spectrum and a high Conversion Ratio (CR) with (2) an external source of high energy neutrons provide an attractive solution to the world's expanding demand for energy? The present study indicates the answer is an emphatic yes.

  9. Absorber rod drive system for nuclear reactors

    International Nuclear Information System (INIS)

    The invention concerns a continuously operating drive system for BWR's. A case tube is used to accomodate all long parts, which can be mounted from above and from below. It can be coupled to the guide tube situated in the reactor pressure vessel. (HP)

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

    International Nuclear Information System (INIS)

    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

  11. Reactor Physics Tests for the Full Power Operation of HANARO

    International Nuclear Information System (INIS)

    The initial criticality of HANARO was achieved on the Feb. 8th of 1995. As HANARO is a unique reactor, there were difficulties to get a license to its full power operation, in which the design power of HANARO is 30 MW. There were two operation license conditions that limited the operation power to 80% of the design power. They were resolved in 2003 and the power ascension tests were conducted for the full power operation. This paper presents the several reactor physics tests for the power ascension to the full power of HANARO

  12. A Data Acquisition System (DAS) for marine and ecological research from aerospace technology

    Science.gov (United States)

    Johnson, R. A.

    1972-01-01

    The efforts of researchers at Mississippi State University to utilize space-age technology in the development of a self-contained, portable data acquisition system for use in marine and ecological research are presented. The compact, lightweight data acquisition system is capable of recording 14 variables in its present configuration and is suitable for use in either a boat, pickup truck, or light aircraft. This system will provide the acquisition of reliable data on the structure of the environment and the effect of man-made and natural activities on the observed phenomenon. Utilizing both self-contained analog recording and a telemetry transmitter for real-time digital readout and recording, the prototype system has undergone extensive testing.

  13. Regulatory aspects of reactor shutdown systems

    International Nuclear Information System (INIS)

    Provision of shutdown system is primary and essential requirement for ensuring safety of a nuclear reactor. The shutdown function has to be performed reliably and adequately as and when called for. The reactor design must establish and provide the shutdown system with required reactivity worth, the required reactivity insertion rate and assure adequate shutdown margin. Reliability of the shutdown system must be assured by proper system design and by provision of redundancy and diversity. For reliable operation of shutdown system it is essential that the quality assurance requirements are identified and met during all the stages of design, fabrication, commissioning and operation. This paper highlights relevant regulatory requirements laid down by Atomic Energy Regulatory Board (AERB) in its safety codes on design, operation as well as on quality assurance of nuclear power plants. The paper also elaborates some of the activities which should be performed for effective compliance of the requirements. (author)

  14. The CANDU Reactor System: An Appropriate Technology.

    Science.gov (United States)

    Robertson, J A

    1978-02-10

    CANDU power reactors are characterized by the combination of heavy water as moderator and pressure tubes to contain the fuel and coolant. Their excellent neutron economy provides the simplicity and low costs of once-through natural-uranium fueling. Future benefits include the prospect of a near-breeder thorium fuel cycle to provide security of fuel supply without the need to develop a new reactor such as the fast breeder. These and other features make the CANDU system an appropriate technology for countries, like Canada, of intermediate economic and industrial capacity. PMID:17788102

  15. Documents needed for obtaining the operation licence for the HERBE system at the RB reactor

    International Nuclear Information System (INIS)

    Documents included in this volume are needed for obtaining the operation licence for the coupled fast-thermal system HERBE constructed at the RB reactor. It contains the following chapters: description of the system; nuclear calculations; performed changes at the RB reactor; proofs about static and dynamic stability of the built construction; normal operation regime of HERBE; accident analysis; dosimetry data; additional instructions and regulations for reactor operation; program of start-up; program for testing the HERBE system

  16. Application of Hastelloy X in Gas-Cooled Reactor Systems

    DEFF Research Database (Denmark)

    Brinkman, C. R.; Rittenhouse, P. L.; Corwin, W.R.;

    1976-01-01

    Hastelloy X, an Ni--Cr--Fe--Mo alloy, may be an important structural alloy for components of gas-cooled reactor systems. Expected applications of this alloy in the High-Temperature Gas-Cooled Reactor (HTGR) are discussed, and the development of interim mechanical properties and supporting data...... extensive amount of information has been generated on this material at Oak Ridge National Laboratory and elsewhere concerning behavior in air, which is reviewed. However, only limited data are available from tests conducted in helium. Comparisons of the fatigue and subcritical growth behavior in air between...... Hastelloy X and a number of other structural alloys are given....

  17. Reactor alarm system development and application issues

    International Nuclear Information System (INIS)

    The new hardware and software technologies, and the need in research reactors for assistance systems in operation and maintenance, have given an appropriate background to develop a computer based system named ''Reactor Alarm System'' (RAS). RAS is a software package, user oriented, with emphasis on production, experiments and maintenance goals. It is designed to run on distributed systems conformed with microcomputers under QNX operating system. RAS main features are: a) Alarm Panel Display; b) Alarm Page; c) Alarm Masking and Inhibition; d) Alarms Color and Attributes; e) Condition Classification; and f) Arrangement Presentation. RAS design allows it to be installed as a part of a computer based Supervision and Control System in new installations or retrofit existing reactor instrumentation systems. The analysis of human factors during development stage and successive user feedback from different applications, brought out several RAS improvements: a) Multiple-copy alarm summaries; b) Improved alarm handling; c) Extended dictionary; and d) Enhanced hardware availability. It has proved successful in providing new capabilities for operators, and also has shown the continuous increase of user-demands, reflecting the expectations placed today on computer-based systems. (author). 6 figs, 1 tabs

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

    International Nuclear Information System (INIS)

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

  19. Large scale replacement of fuel channels in the Pickering CANDU reactor using a man-in-the-loop remote control system

    International Nuclear Information System (INIS)

    Spar Aerospace Limited of Toronto is presently under contract to Ontario Hydro to design a Remote Manipulation and Control System (RMCS) to be used during the large scale replacement of the fuel channels in the Pickering A Nuclear Generating Station. The system is designed to support the replacement of all 390 fuel channels in each of the four reactors at the Pickering A station in a safe manner that minimizes worker radiation exposure and unit outage time

  20. Reactor power system deployment and startup

    Science.gov (United States)

    Wetch, J. R.; Nelin, C. J.; Britt, E. J.; Klein, G.

    1985-01-01

    This paper addresses issues that should receive further examination in the near-term as concept selection for development of a U.S. space reactor power system is approached. The issues include: the economics, practicality and system reliability associated with transfer of nuclear spacecraft from low earth shuttle orbits to operational orbits, via chemical propulsion versus nuclear electric propulsion; possible astronaut supervised reactor and nuclear electric propulsion startup in low altitude Shuttle orbit; potential deployment methods for nuclear powered spacecraft from Shuttle; the general public safety of low altitude startup and nuclear safe and disposal orbits; the question of preferred reactor power level; and the question of frozen versus molten alkali metal coolant during launch and deployment. These issues must be considered now because they impact the SP-100 concept selection, power level selection, weight and size limits, use of deployable radiators, reliability requirements, and economics, as well as the degree of need for and the urgency of developing space reactor power systems.