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Sample records for rod test rig

  1. Large test rigs verify Clinch River control rod reliability

    International Nuclear Information System (INIS)

    Michael, H.D.; Smith, G.G.

    1983-01-01

    The purpose of the Clinch River control test programme was to use multiple full-scale prototypic control rod systems for verifying the system's ability to perform reliably during simulated reactor power control and emergency shutdown operations. Two major facilities, the Shutdown Control Rod and Maintenance (Scram) facility and the Dynamic and Seismic Test (Dast) facility, were constructed. The test programme of each facility is described. (UK)

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

    International Nuclear Information System (INIS)

    Sun, Jongoh; Kim, Gyeongho; Yoo, Yeonsik; Cho, Yeonggarp; Kim, Jong In

    2014-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-15

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

  4. Power ramp tests of MOX fuel rods. HBWR irradiation with the instrument rig, IFA-591

    International Nuclear Information System (INIS)

    Ozawa, Takayuki; Abe, Tomoyuki

    2006-03-01

    Plutonium-uranium mixed oxide (MOX) fuel rods of instrumental rig IFA-591 were ramped in HBWR to study the Advanced Thermal Reactor (ATR) MOX fuel behavior during transient operation and to determine a failure threshold of the MOX fuel rods. Eleven segments were base-irradiated in ATR 'FUGEN' up to 18.4 GWd/t. Zirconium liner claddings were adopted for four segments of them. As the results of non-destructive post irradiation examinations (PIEs) after the base-irradiation and before the ramp tests, no remarkable behavior affecting the integrity of fuel assembly and fuel rod was confirmed. All segments to be used for the ramp tests, which consisted of the multi-step ramp tests and the single-step ramp tests, had instrumentations for in-pile measurements of cladding elongation or plenum pressure, and heated up to the maximum linear power of 58.3-68.4 kW/m without failure. The major results of ramp tests are as follows: There is no difference in PCMI behaviors between two type rods of Zry-2 and Zirconium liner claddings from the in-pile measurements of cladding elongation and plenum pressure. The computations of cladding elongation and inner pressure gave slightly lower elongation and pressure than the in-pile measurements during the ramp-test. However, the cladding relaxation during the power hold was in good agreement, and the fission gas release behavior during cooling down could be evaluated by taking into account the relaxation of contact pressure between pellet and cladding. Although the final power during IFA-591 ramp tests reached the higher linear power than the failure threshold power of UO 2 fuel rods, no indication of fuel failure was observed during the ramp tests. The cladding relaxation due to the creep deformation of the MOX pellets at high temperature could be confirmed at the power steps during the multi-ramp test. The fission gas release due to the emancipation from PCMI stress was observed during the power decreasing. The burn-up dependence could be

  5. A Study on the Sliding/Impact Wear of a Nuclear Fuel Rod in Room Temperature Air: (I) Development of a Test Rig and Characteristic Analysis

    International Nuclear Information System (INIS)

    Lee, Young Ho; Lee, Kang Hee; Kim, Hyung Kyu

    2007-01-01

    A new type of a fretting wear tester has been designed and developed in order to simulate the actual vibration behavior of a nuclear fuel rod for springs/dimples in room temperature. When considering the actual contact condition between fuel rod and spring/dimple, if fretting wear progress due to the Flow-Induced Vibration (FIV) under a specific normal load exerted on the fuel rod by the elastic deformation of the spring, the contacting force between the fuel rod and dimple that were located in the opposite side should be decreased. Consequently, the evaluation of developed spacer grids against fretting wear damage should be performed with the results of a cell unit experiments because the contacting force is one of the most important variables that influence to the fretting wear mechanism. Therefore, it is necessary to develop a new type of fretting test rig in order to simulate the actual contact condition. In this paper, the development procedure of a new fretting wear tester and its performance were discussed in detail

  6. Robotics Test Rig

    International Nuclear Information System (INIS)

    Schuurmans, P.

    2007-01-01

    The experimental Accelerator Driven System XT-ADS is being developed within the European 6th framework programme EUROTRANS using the MYRRHA DRAFT-2 as starting point. The aim for the XT-ADS is to demonstrate the feasibility of the ADS concept at reasonable power levels and to serve as a high performance, multi-purpose experimental irradiation device. One of the fundamental design options that has been taken is to do all maintenance and in-service inspection and repair duties by remote handling. Outside the XT-ADS vessel in a controlled though radio-active environment, remote handling concepts as those already in use at e.g. the Joint European Torus (JET) can be used. Extrapolation to remote handling inside the lead-bismuth eutectic filled main vessel of the XT-ADS is in principle feasible as was shown in a 2003 study performed by Oxford Technologies ltd for the case of MYRRHA. Nevertheless, it is clear that all critical remote handling components need to be qualified for use in liquid LBE. Thus, as a first step, a proof of principle (POP) experimental test rig is require. The principal goal of this work is to identify the critical technological issues that must be resolved to allow operation of remote handling manipulators inside the LBE filled main vessel of the XT-ADS and to propose a concept design and specification catalogue for a proof of principle test rig that is able to experimentally verify the main aspects of manipulator design

  7. Endurance test of DUPIC irradiation test rig-003

    Energy Technology Data Exchange (ETDEWEB)

    Moon, J.S; Yang, M.S.; Lee, C.Y.; Ryu, J.S.; Jeon, H.G

    2001-04-01

    This report presents the pressure drop, vibration and endurance test results for DUPIC Irradiation Test Rig-003 which was design and fabricated by KAERI. From the pressure drop and vibration test results, it is verified that DUPIC Irradiation Test Rig-003 satisfied the limit conditions of HANARO. And, remarkable wear is not observed in DUPIC Irradiation Test Rig-003 during 40 endurance test days.

  8. Tractor accelerated test on test rig

    Directory of Open Access Journals (Sweden)

    M. Mattetti

    2013-09-01

    Full Text Available The experimental tests performed to validate a tractor prototype before its production, need a substantial financial and time commitment. The tests could be reduced using accelerated tests able to reproduce on the structural part of the tractor, the same damage produced on the tractor during real life in a reduced time. These tests were usually performed reproducing a particular harsh condition a defined number of times, as for example using a bumpy road on track to carry out the test in any weather condition. Using these procedures the loads applied on the tractor structure are different with respect to those obtained during the real use, with the risk to apply loads hard to find in reality. Recently it has been demonstrated how, using the methodologies designed for cars, it is possible to also expedite the structural tests for tractors. In particular, automotive proving grounds were recently successfully used with tractors to perform accelerated structural tests able to reproduce the real use of the machine with an acceleration factor higher than that obtained with the traditional methods. However, the acceleration factor obtained with a tractor on proving grounds is in any case reduced due to the reduced speed of the tractors with respect to cars. In this context, the goal of the paper is to show the development of a methodology to perform an accelerated structural test on a medium power tractor using a 4 post test rig. In particular, several proving ground testing conditions have been performed to measure the loads on the tractor. The loads obtained were then edited to remove the not damaging portion of signals, and finally the loads obtained were reproduced in a 4 post test rig. The methodology proposed could be a valid alternative to the use of a proving ground to reproduce accelerated structural tests on tractors.

  9. LWR primary coolant pipe rupture test rig

    International Nuclear Information System (INIS)

    Yoshitoshi, Shyoji

    1978-01-01

    The rupture test rig for primary coolant pipes is constructed in the Japan Atomic Energy Research Institute to verify the reliability of the primary coolant pipes for both PWRs and BWRs. The planned test items consisted of reaction force test, restraint test, whip test, jet test and continuous release test. A pressure vessel of about 4 m 3 volume, a circulating pump, a pressurizer, a heater, an air cooler and the related instrumentation and control system are included in this test rig. The coolant test condition is 160 kg/cm 2 g, 325 deg C for PWR test, and 70 kg/cm 2 g, saturated water and steam for BWR test, 100 ton of test load for the ruptured pipe bore of 8B Schedule 160, and 20 lit/min. discharge during 20 h for continuous release of coolant. The maximum pit internal pressure was estimated for various pipe diameters and time under the PWR and BWR conditions. The spark rupturing device was adopted for the rupture mechanics in this test rig. The computer PANAFACOM U-300 is used for the data processing. This test rig is expected to operate in 1978 effectively for the improvement of reliability of LWR primary coolant pipes. (Nakai, Y.)

  10. Development of TIG Welding System for a Nuclear Fuel Test Rig

    International Nuclear Information System (INIS)

    Joung, Changyoung; Ahn, Sungho; Hong, Jintae; Kim, Kahye

    2013-01-01

    The welding process is one of the most important among the instrumentation processes of the nuclear fuel test rig and rods. To manufacture the nuclear fuel test rig, a precision welding system needs to be fabricated to develop various welding technologies of the fuel test rig and rods jointing the various sensors and end caps on a fuel cladding tube, which is charged with fuel pellets and component parts. Thus, we designed and fabricated the precision welding system consisting of an orbital TIG welder, a low-pressure chamber, and a high-pressure chamber. Using this system, the performance tests were performed with the round and seal spot welds for each welding condition. This paper describes not only the contents for the fabrication of precision TIG welding system but also some results from weld tests using the low-pressure and high-pressure chambers to verify the performance of this system. The TIG welding system was developed to manufacture the nuclear fuel test rig and rods. It has been configured to be able to weld the nuclear fuel test rigs and rods by applying the TIG welder using a low-pressure chamber and a high-pressure chamber. The performance tests using this system were performed with the round and seal spot welds for the welding conditions. The soundness of the orbital TIG welding system was confirmed through performance tests in the low-pressure and high-pressure chambers

  11. Development of TIG Welding System for a Nuclear Fuel Test Rig

    Energy Technology Data Exchange (ETDEWEB)

    Joung, Changyoung; Ahn, Sungho; Hong, Jintae; Kim, Kahye [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    The welding process is one of the most important among the instrumentation processes of the nuclear fuel test rig and rods. To manufacture the nuclear fuel test rig, a precision welding system needs to be fabricated to develop various welding technologies of the fuel test rig and rods jointing the various sensors and end caps on a fuel cladding tube, which is charged with fuel pellets and component parts. Thus, we designed and fabricated the precision welding system consisting of an orbital TIG welder, a low-pressure chamber, and a high-pressure chamber. Using this system, the performance tests were performed with the round and seal spot welds for each welding condition. This paper describes not only the contents for the fabrication of precision TIG welding system but also some results from weld tests using the low-pressure and high-pressure chambers to verify the performance of this system. The TIG welding system was developed to manufacture the nuclear fuel test rig and rods. It has been configured to be able to weld the nuclear fuel test rigs and rods by applying the TIG welder using a low-pressure chamber and a high-pressure chamber. The performance tests using this system were performed with the round and seal spot welds for the welding conditions. The soundness of the orbital TIG welding system was confirmed through performance tests in the low-pressure and high-pressure chambers.

  12. Control rod testing apparatus

    International Nuclear Information System (INIS)

    Gaunt, R.R.; Ashman, C.M.

    1987-01-01

    A control rod testing apparatus is described comprising: a first guide means having a vertical cylindrical opening for grossly guiding a control rod; a second guide means having a vertical cylindrical opening for grossly guiding a control rod. The first and second guide means are supported at axially spaced locations with the openings coaxial; and a substantially cylindrical subassembly having a vertical cylindrical opening therethrough. The subassembly is trapped coaxial with and between the first and second guide means, and the subassembly radially floats with respect to the first and second guide means

  13. A power recirculating test rig for ball screw endurance tests

    Directory of Open Access Journals (Sweden)

    Giberti Hermes

    2016-01-01

    Full Text Available A conceptual design of an innovative test rig for endurance tests of ball screws is presented in this paper. The test rig layout is based on the power recirculating principle and it also allows to overtake the main critical issues of the ball screw endurance tests. Among these there are the high power required to make the test, the lengthy duration of the same and the high loads between the screw and the frame that holds it. The article describes the test rig designed scheme, the kinematic expedients to be adopted in order to obtain the required performance and functionality and the sizing procedure to choose the actuation system.

  14. Endurance test on IR rig for RI production

    International Nuclear Information System (INIS)

    Chung, Heung June; Youn, Y. J.; Han, H. S.; Hong, S. B.; Cho, Y. G.; Ryu, J. S.

    2000-12-01

    This report presents the pressure drop, vibration and endurance test results for IR rig for RI production which were desigened and fabricated by KAERI. From the pressure drop test results, it is noted that the flow rate through the IR rig corresponding to the pressure drop of 200 kPa is measured to be about 3.12 kg/sec. Vibration frequency for the IR rig ranges from 13 to 17 Hz. RMS(Root Mean Square) displacement for the IR rig is less than 30 μm, and the maximum displacement is less than 110μm. These experimental results show that the design criteria of IR rig meet the HANARO limit conditions. Endurance test results show that the appreciable fretting wear for the IR rig does not occur, however tiny trace of wear between contact points is observed

  15. Preliminary nuclear design for test MOX Fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Hyung Kook; Kim, Taek Kyum; Jeong, Hyung Guk; Noh, Jae Man; Cho, Jin Young; Kim, Young Il; Kim, Young Jin; Sohn, Dong Seong

    1997-10-01

    As a part of activity for future fuel development project, test MOX fuel rods are going to be loaded and irradiated in Halden reactor core as a KAERI`s joint international program with Paul Scherrer Institute (PSI). PSI will fabricate test MOX rods with attrition mill device which was developed by KAERI. The test fuel assembly rig contains three MOX rods and three inert matrix rods. One of three MOX rods will be fabricated by BNFL, the other two MOX fuel rods will be manufacturing jointly by KAERI and PSI. Three inert matrix fuel rods will be fabricated with Zr-Y-Er-Pu oxide. Neutronic evaluation was preliminarily performed for test fuel assembly suggested by PSI. The power distribution of test fuel rod in test fuel assembly was analyzed for various fuel rods position in assembly and the depletion characteristic curve for test fuel was also determined. The fuel rods position in test fuel assembly does not effect the rod power distribution, and the proposal for test fuel rods suggested by PSI is proved to be feasible. (author). 2 refs., 13 tabs., 16 figs.

  16. Advisory expert system for test rig operator

    International Nuclear Information System (INIS)

    Zielczynski, P.

    1994-01-01

    The advisory expert system MAESTRO (Modular Advisory Expert System for Test Rig Operator) has been designed to guide the operator of large experimental installation during start-up, steady state and shut down. The installation is located in the research reactor MARIA in the Institute of Atomic Energy in Swierk, Poland. The system acquires and analyses on line signals from installation and performs two tasks in real time: leading the operator and monitoring of the installation (including signal validation). Systems tasks, architecture and knowledge representation concepts are described. The system is based on expert systems techniques what makes in phases of continuous change of process parameters and it has been achieved by special knowledge representation allowing its dynamical modification. (author). 147 refs, 42 figs, 5 tab

  17. A Dynamic Behavior of the Nuclear Test Rig with Coolant using the Fluid-Structural interaction Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Tae-Ho; Hong, Jintae; Ahn, Sung-Ho; Joung, Chang-Young; Jang, Seo-Yun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Yeon, Kon-Whi [Chungnam National University, Daejeon (Korea, Republic of)

    2016-10-15

    In this paper, the dynamic behavior of the test rig in the coolant flow simulator is evaluated by using the 2-way fluid-structural interaction analysis. The maximum value and location of the deformation and equivalent stress in the test rig is confirmed. The fluid-structural interaction analysis is applied to perform the fluid and structural analysis A fluid-structure interaction analysis is used to simulate the relationship between the deformation and hydraulic pressure. There are two types of fluid-structural interaction analysis. One is a 1-way direction analysis in which the hydraulic pressure is calculated by a CFD and transmitted to the surface of the structure, and a structural analysis is then performed. The other is a 2-way direction analysis that is performed by changing the data between the deformation of the structural and pressure of the coolant water for every time step. The location of the maximum deformation of the test rig is the bottom parts of the test rig. It is expected that the equivalent stress of the test rig is occurred. The maximum equivalent stress in the test rig under the circulation of the coolant is 90.1 MPa. The location of the maximum stress in the test rig is the connect part between the fuel rod and flow divider. A safety factor on the test rig is 3, approximately. The deformation motion of the test rig at the bottom part of the test rig is caused about the fluid-induced vibration. A test on the fluid-induced vibration of the test rig will be performed and compared with results of the analysis in further paper.

  18. ROLLER RIG TESTING AT THE CZECH TECHNICAL UNIVERSITY

    Directory of Open Access Journals (Sweden)

    J. Kalivoda

    2016-08-01

    Full Text Available Purpose. Although the advancements in computer simulation technology have paved way to provide very reliable simulation results, track tests still play an essential role during the process of development and homologation of any railway vehicle. On the other hand, track tests depend on weather conditions, are difficult to organize and are not suitable for testing vehicles in critical situations. On a roller rig, the tested vehicle is longitudinally fixed and a track is replaced by rotating rollers. Such device offer testing of railway vehicle running dynamics in safe and stable laboratory environment. The purpose of an article is to investigate and describe roller rig testing at the Czech technical university in Prague (CTU. Methodology. In the paper it is shown the history of development of the scaled CTU roller rig from the earlier stages until the current projects for which the CTU roller rig is utilized for. The current design of the experimental bogie, roller rig, sensors instrumentation and types of experiments conducted at the CTU roller rig are described in more detail. Findings. Although the differences in vehicle behaviour on a track and a scaled model on a roller rig are not negligible, scaled roller rig experiments are found as a relatively inexpensive way for verification and demonstration of computer simulations results. They are especially useful for verification of multibody system simulations (MBS of entirely new running gear concepts. Originality. The CTU roller rig is currently used for the experiments with active controlled wheelset guidance. According to simulations results published in many papers such systems offer, in principle, better performance compared to conventional passive vehicles. However, utilization and testing of active controlled wheelset guidance on vehicles is still rare. CTU roller rig serves as a tool to verify computer simulations and demonstrate benefits of active wheelset guidance. Practical value

  19. Experimental test results of multi-channel test rig of T1 test section, 5

    International Nuclear Information System (INIS)

    Hino, Ryutaro; Takase, Kazuyuki; Miyamoto, Yoshiaki

    1990-09-01

    Channel blockage test on a fuel column of the high temperature engineering test reactor (HTTR) has been performed under the helium gas atmosphere at a high temperature and a high pressure in order to obtain safety data on flow rate and temperature distributions in the fuel column with the multi-channel test rig of the fuel stack test section (T 1 ) in HENDEL. In the test, one of 12 fuel channels was blockaded to 90% of flow area at the channel inlet. Experimental results showed that the helium gas flow rate in the blockaded channel was 28%∼33% lower than the average flow rate for Reynolds number from 2300 to 14000 in isothermal flow. When simulated fuel rods were heated, the flow rate in the blockaded channel did not decrease down in comparison with the isothermal flow. This is due to that the heat generated in the fuel rods conducts to the other fuel channels in graphite fuel blocks, so that accelerated pressure losses in the fuel channels change with helium gas temperatures. (author)

  20. Nuclear Fuel Test Rod Fabrication for Data Acquisition Test

    International Nuclear Information System (INIS)

    Joung, Chang-Young; Hong, Jin-Tae; Kim, Ka-Hye; Huh, Sung-Ho

    2014-01-01

    A nuclear fuel test rod must be fabricated with precise welding and assembly technologies, and confirmed for their soundness. Recently, we have developed various kinds of processing systems such as an orbital TIG welding system, a fiber laser welding system, an automated drilling system and a helium leak analyzer, which are able to fabricate the nuclear fuel test rods and rigs, and keep inspection systems to confirm the soundness of the nuclear fuel test rods and rids. The orbital TIG welding system can be used with two kinds of welding methods. One can perform the round welding for end-caps of a nuclear fuel test rod by an orbital head mounted in a low-pressure chamber. The other can do spot welding for a pin-hole of a nuclear fuel test rod in a high-pressure chamber to fill up helium gas of high pressure. The fiber laser welding system can weld cylindrical and 3 axis samples such as parts of a nuclear fuel test rod and instrumentation sensors which is moved by an index chuck and a 3 axis (X, Y, Z) servo stage controlled by the CNC program. To measure the real-time temperature change at the center of the nuclear fuel during the irradiation test, a thermocouple should be instrumented at that position. Therefore, a hole needs to be made at the center of fuel pellet to instrument the thermocouple. An automated drilling system can drill a fine hole into a fuel pellet without changing tools or breaking the work-piece. The helium leak analyzer (ASM-380 model of DEIXEN Co.) can check the leak of the nuclear fuel test rod filled with helium gas. This paper describes not only the assembly and fabrication methods used by the process systems, but also the results of the data acquisition test for the nuclear fuel test rod. A nuclear fuel test rod for the data acquisition test was fabricated using the welding and assembling echnologies acquired from previous tests

  1. Nuclear Fuel Test Rod Fabrication for Data Acquisition Test

    Energy Technology Data Exchange (ETDEWEB)

    Joung, Chang-Young; Hong, Jin-Tae; Kim, Ka-Hye; Huh, Sung-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    A nuclear fuel test rod must be fabricated with precise welding and assembly technologies, and confirmed for their soundness. Recently, we have developed various kinds of processing systems such as an orbital TIG welding system, a fiber laser welding system, an automated drilling system and a helium leak analyzer, which are able to fabricate the nuclear fuel test rods and rigs, and keep inspection systems to confirm the soundness of the nuclear fuel test rods and rids. The orbital TIG welding system can be used with two kinds of welding methods. One can perform the round welding for end-caps of a nuclear fuel test rod by an orbital head mounted in a low-pressure chamber. The other can do spot welding for a pin-hole of a nuclear fuel test rod in a high-pressure chamber to fill up helium gas of high pressure. The fiber laser welding system can weld cylindrical and 3 axis samples such as parts of a nuclear fuel test rod and instrumentation sensors which is moved by an index chuck and a 3 axis (X, Y, Z) servo stage controlled by the CNC program. To measure the real-time temperature change at the center of the nuclear fuel during the irradiation test, a thermocouple should be instrumented at that position. Therefore, a hole needs to be made at the center of fuel pellet to instrument the thermocouple. An automated drilling system can drill a fine hole into a fuel pellet without changing tools or breaking the work-piece. The helium leak analyzer (ASM-380 model of DEIXEN Co.) can check the leak of the nuclear fuel test rod filled with helium gas. This paper describes not only the assembly and fabrication methods used by the process systems, but also the results of the data acquisition test for the nuclear fuel test rod. A nuclear fuel test rod for the data acquisition test was fabricated using the welding and assembling echnologies acquired from previous tests.

  2. BURNER RIG TESTING OF A500 C/SiC

    Science.gov (United States)

    2018-03-17

    AFRL-RX-WP-TR-2018-0071 BURNER RIG TESTING OF A500® C /SiC Larry P. Zawada Universal Technology Corporation Jennifer Pierce UDRI...TITLE AND SUBTITLE BURNER RIG TESTING OF A500® C /SiC 5a. CONTRACT NUMBER In-House 5b. GRANT NUMBER 5c. PROGRAM ELEMENT NUMBER 62102F 6...test program characterized the durability behavior of A500® C /SiC ceramic matrix composite material at room and elevated temperature. Specimens were

  3. Development and Initial Testing of the Tiltrotor Test Rig

    Science.gov (United States)

    Acree, C. W., Jr.; Sheikman, A. L.

    2018-01-01

    The NASA Tiltrotor Test Rig (TTR) is a new, large-scale proprotor test system, developed jointly with the U.S. Army and Air Force, to develop a new, large-scale proprotor test system for the National Full-Scale Aerodynamics Complex (NFAC). The TTR is designed to test advanced proprotors up to 26 feet in diameter at speeds up to 300 knots, and even larger rotors at lower airspeeds. This combination of size and speed is unprecedented and is necessary for research into 21st-century tiltrotors and other advanced rotorcraft concepts. The TTR will provide critical data for validation of state-of-the-art design and analysis tools.

  4. Cadmium safety rod thermal tests

    International Nuclear Information System (INIS)

    Thomas, J.K.; Iyer, N.C.; Peacock, H.B.

    1992-01-01

    Thermal testing of cadmium safety rods was conducted as part of a program to define the response of Savannah River Site (SRS) production reactor core components to a hypothetical LOCA leading to a drained reactor tank. The safety rods are present in the reactor core only during shutdown and are not used as a control mechanism during operation; thus, their response to the conditions predicted for the LOCA is only of interest to the extent that it could impact the progression of the accident. This document provides a description of this testing

  5. Test Rig for Valves of Digital Displacement Machines

    DEFF Research Database (Denmark)

    Nørgård, Christian; Christensen, Jeppe Haals; Bech, Michael Møller

    2017-01-01

    A test rig for the valves of digital displacement machines has been developed at Aalborg University. It is composed of a commercial radial piston machine, which has been modified to facilitate Digital Displacement operation for a single piston. Prototype valves have been optimized, designed and m...

  6. Out pile test of a disassembly tool for the intermediate examination of nuclear fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Jintae; Joung, Chang-Young; Ahn, Sung-Ho; Yang, Tae-Ho; Jang, Seo-Yoon; Park, Seung-Jae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The two fuel rod assemblies are assembled with a bayonet coupler, and the non-instrumented fuel rod assembly can be disassembled for intermediate examination. A tool to disassemble the non-instrumented fuel rod assembly from the test rig was developed, and steel wires are connected to the tool to operate release function. In this study, an assembly plug with a quick plug typed bayonet coupler and the accompanying disassembly tool was designed to prevent the interference problem. A test rig mockup was fabricated, and performance test was carried out in the laboratory. And, the out pile test was also carried out in the single channel test loop established in the KAERI. In this study, a modified coupler design to disassemble the non-instrumented fuel rod assembly from the test rig for the intermediate examination was suggested to solve interference problem of previous design. The performance of the modified design was verified by test mockup fabricated with the modified coupler design and accompanied disassembly tool design. Finally, out pile test was carried out in the single channel test loop in the KAERI, and the test rig and the disassembly tool showed good performance and reliability. The developed technique will be useful to the periodic intermediate examination of nuclear fuel rods.

  7. Out pile test of a disassembly tool for the intermediate examination of nuclear fuel rods

    International Nuclear Information System (INIS)

    Hong, Jintae; Joung, Chang-Young; Ahn, Sung-Ho; Yang, Tae-Ho; Jang, Seo-Yoon; Park, Seung-Jae

    2016-01-01

    The two fuel rod assemblies are assembled with a bayonet coupler, and the non-instrumented fuel rod assembly can be disassembled for intermediate examination. A tool to disassemble the non-instrumented fuel rod assembly from the test rig was developed, and steel wires are connected to the tool to operate release function. In this study, an assembly plug with a quick plug typed bayonet coupler and the accompanying disassembly tool was designed to prevent the interference problem. A test rig mockup was fabricated, and performance test was carried out in the laboratory. And, the out pile test was also carried out in the single channel test loop established in the KAERI. In this study, a modified coupler design to disassemble the non-instrumented fuel rod assembly from the test rig for the intermediate examination was suggested to solve interference problem of previous design. The performance of the modified design was verified by test mockup fabricated with the modified coupler design and accompanied disassembly tool design. Finally, out pile test was carried out in the single channel test loop in the KAERI, and the test rig and the disassembly tool showed good performance and reliability. The developed technique will be useful to the periodic intermediate examination of nuclear fuel rods

  8. Testing of elastomer seals using small-size rigs

    International Nuclear Information System (INIS)

    Leeks, C.W.E.; Dunford, B.; Barnfield, J.H.; Gray, I.L.S.

    1997-01-01

    This paper looks at the use of small size seal leakage test rigs to demonstrate the compliance of full size container seals against the IAEA Transport Regulation's limits for activity release for normal transport and accident conditions. The detailed requirements of the regulations are discussed and it is concluded that an appropriate test programme to meet these requirements using only small size test rigs, can normally be set up and carried out on a relatively short time scale. It is important that any small test rigs should be designed to represent the relevant features of the seal arrangement and the overall test programme should cover all of the conditions, specified by the regulations, for the type, classification and contents of the container under consideration. The parameters of elastomer O-rings, which affect their sealing ability, are considered and those which are amenable to small scale testing or have to be modelled at full size are identified. Generally, the seals used in leakage tests have to be modelled with a full size cross-section but can have a reduced peripheral length. (Author)

  9. Test Rig Design and Presentation for a Hydraulic Yaw System

    DEFF Research Database (Denmark)

    Stubkier, Søren; Pedersen, Henrik C.; Andersen, Torben Ole

    2013-01-01

    The design and development of a hydraulic yaw system for multi MWturbines is presented and the concept explained. As part of the development of the new concept a full scale test rig for a 5 MW wind turbine has been designed and constructed. The test rig is presented along with its unique design...... features. The design process is outlined to give insight in the design criteria driving the design. Loads and yaw demands are based on the IEC 61400-1 standard for wind turbine design, and the loads for this examination are extrapolated from the FAST aero elastic design software. The concepts are based...... on a 5 MW offshore turbine. After the system presentation, measurement results are presented to verify the behavior of the system. The loads to the system are applied by torque controlled electrical servo drives, which can add a load of up to 3 MNm to the system. This gives an exact picture of the system...

  10. The laboratory test rig with miniature jet engine to research aviation fuels combustion process

    Directory of Open Access Journals (Sweden)

    Gawron Bartosz

    2015-12-01

    Full Text Available This article presents laboratory test rig with a miniature turbojet engine (MiniJETRig – Miniature Jet Engine Test Rig, that was built in the Air Force Institute of Technology. The test rig has been developed for research and development works aimed at modelling and investigating processes and phenomena occurring in full scale jet engines. In the article construction of a test rig is described, with a brief discussion on the functionality of each of its main components. Additionally examples of measurement results obtained during the realization of the initial tests have been included, presenting the capabilities of the test rig.

  11. Endurance test for IR rig for RI production assembly (test procedure)

    International Nuclear Information System (INIS)

    Chung, Heung June; Ryu, Jeong Soo

    2000-08-01

    This test procedure details the test loop, test method, and test procedure for pressure drop, vibration and endurance test of IR Rig for RI production. From the pressure drop test, the hydraulic design requirements of the capsule are verified. HANARO limit condition is checked and the compatibility with HANARO core is verified. From flow induced vibration test vibration frequency and displacement are investigated. The wear of IR Rig is investigated through endurance test, and these data are used to evaluate the expected wear at maximum resident time of the IR Rig for RI production

  12. Power ramp testing method for PWR fuel rod at research reactor

    International Nuclear Information System (INIS)

    Zhou Yidong; Zhang Peisheng; Zhang Aimin; Gao Yongguang; Wang Huarong

    2003-01-01

    A tentative power ramp test for short PWR fuel rod has been conducted at the Heavy Water Research Reactor (HWRR) in China Institute of Atomic Energy (CIAE). The test fuel rod was cooled by the circulating water in the test loop. The power ramp was realized by moving solid neutron-absorbing screen around the fuel rod. The linear power of the fuel rod increased from 220 W/cm to 340 W/cm with a power ramp rate of 20 W/cm/min. The power of the fuel rod was monitored by both in-core thermal and nuclear measurement sensors in the test rig. This test provides experiences for further developing the power ramp test methods for PWR fuel rods at research reactor. (author)

  13. A New Rig for Testing Textured Surfaces in Pure Sliding Conditions

    DEFF Research Database (Denmark)

    Godi, Alessandro; Grønbæk, J.; Mohaghegh, Kamran

    2013-01-01

    machineries are necessary: a press to provide the normal pressure and a tensile machine to perform the axial movements. The test is calibrated so that the correspondence between the normal pressure and the container advancement is found. Preliminary tests are carried out involving a multifunctional and a fine......Throughout the years, it has become more and more important to find new methods for reducing friction and wear occurrence in machine elements. A possible solution is found in texturing the surfaces under tribological contact, as demonstrated by the development and spread of plateau-honed surface...... for cylinder liners. To prove the efficacy of a particular textured surface, it is paramount to perform experimental tests under controlled laboratory conditions. In this paper, a new test rig simulating pure sliding conditions is presented, dubbed axial sliding test. It presents four major components: a rod...

  14. Testing device for control rod drives

    International Nuclear Information System (INIS)

    Hayakawa, Toshifumi.

    1992-01-01

    A testing device for control rod drives comprises a logic measuring means for measuring an output signal from a control rod drive logic generation circuit, a control means for judging the operation state of a control rod and a man machine interface means for outputting the result of the judgement. A driving instruction outputted from the control rod operation device is always monitored by the control means, and if the operation instruction is stopped, a testing signal is outputted to the control rod control device to simulate a control rod operation. In this case, the output signal of the control rod drive logic generation circuit is held in a control rod drive memory means and intaken into a logic analysis means for measurement and an abnormality is judged by the control means. The stopping of the control rod drive instruction is monitored and the operation abnormality of the control rod is judged, to mitigate the burden of an operator. Further, the operation of the control rod drive logic generation circuit can be confirmed even during a nuclear plant operation by holding the control rod drive instruction thereby enabling to improve maintenance efficiency. (N.H.)

  15. Dynamic Investigation Test-rig on hAptics (DITA)

    International Nuclear Information System (INIS)

    Cannella, F; Olivieri, E; Caldwell, D G; Scalise, L; Memeo, M

    2013-01-01

    Research on tactile sensitivity has been conducted since the last century and many devices have been proposed to study in detail this sense through experimental tests. The sense of touch is essential in every-day life of human beings, but it can also play a fundamental role for the assessment of some neurological disabilities and pathologies. In fact, the level of tactile perception can provide information on the health state of the nervous system. In this paper, authors propose the design and development of a novel test apparatus, named DITA (Dynamic Investigation Test-rig on hAptics), aiming to provide the measurement of the tactile sensitivity trough the determination of the Just Noticeable Difference (JND) curve of a subject. The paper reports the solution adopted for the system design and the results obtained on the set of experiments carried out on volunteers

  16. Core Cutting Test with Vertical Rock Cutting Rig (VRCR)

    Science.gov (United States)

    Yasar, Serdar; Osman Yilmaz, Ali

    2017-12-01

    Roadheaders are frequently used machines in mining and tunnelling, and performance prediction of roadheaders is important for project economics and stability. Several methods were proposed so far for this purpose and, rock cutting tests are the best choice. Rock cutting tests are generally divided into two groups which are namely, full scale rock cutting tests and small scale rock cutting tests. These two tests have some superiorities and deficiencies over themselves. However, in many cases, where rock sampling becomes problematic, small scale rock cutting test (core cutting test) is preferred for performance prediction, since small block samples and core samples can be conducted to rock cutting testing. Common problem for rock cutting tests are that they can be found in very limited research centres. In this study, a new mobile rock cutting testing equipment, vertical rock cutting rig (VRCR) was introduced. Standard testing procedure was conducted on seven rock samples which were the part of a former study on cutting rocks with another small scale rock cutting test. Results showed that core cutting test can be realized successfully with VRCR with the validation of paired samples t-test.

  17. The moisture proof connection of signal cables on test rig instrumentation

    Energy Technology Data Exchange (ETDEWEB)

    Joung, Chang Young; Ahn, Sung Ho; Hong, Jin Tae; Jeong, Hwang Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-10-15

    The rod inner pressure, centerline temperature, coolant temperature, and neutron flux resulting from the irradiation properties of nuclear fuel are an important factor for evaluating nuclear fuel properties in pile. In addition instrumentation and measurement techniques for nuclear fuel are necessary to measure the exact data. Special sensors such as a TC (thermocouple), LVDT (linear variable differential transformer) and SPND (self-powered neutron detector) are instrumented in and out of the fuel rod to measure the various irradiation characteristics of the nuclear fuel. These sensors are made up of the sensor itself and a signal cable. In the instrumentation, an MI (Mineral Insulated) cable used as the signal cable has such properties as high electrical insulation, heat resistance, and mechanical strength. However, it is difficult to handle and treat with care owing to the extremely hard composition, which is made up of weak signal wires and alumina powder in a stainless tube. The sealing of the end tip of the MI cable and extension cable is very important in terms of the insulation resistance to seal the insulator inside the MI cable tube from moisture. To maintain the insulation of sensors and signal cables, the insulation resistance must be checked in accordance with each process throughout the instrumentation and fabrication period. To safely mount the signal cables drawn from a fuel test rig on the terminal block of a junction panel, the MI and extension cables should be easy to connect. Therefore, it is necessary to develop instrumentation technologies of a moisture proof connection process for a fuel test rig. This paper will provide an overview of the work done with moisture proof connection procedures to connect the MI and extension cables to extend the MI cables jointed with the sensor.

  18. The Defect Inspection on the Irradiated Fuel Rod by Eddy Current Test

    International Nuclear Information System (INIS)

    Koo, D. S.; Park, Y. K.; Kim, E. K.

    1996-01-01

    The eddy current test(ECT) probe of differential encircling coil type was designed and fabricated, and the optimum condition of ECT was derived for the examination of the irradiated fuel rod. The correlation between ECT test frequency and phase and amplitude was derived by performing the test of the standard rig that includes inner notches, outer notches and through-holes. The defect of through-hole was predicted by ECT at the G33-N2 fuel rod irradiated in the Kori-1 nuclear power reactor. The metallographic examination on the G33-N2 fuel rod was Performed at the defect location predicted by ECT. The result of metallographic examination for the G33-N2 fuel rod was in good agreement with that of ECT. This proves that the evaluation for integrity of irradiated fuel rod by ECT is reliable

  19. Testing of a new morphing trailing edge flap system on a novel outdoor rotating test rig

    DEFF Research Database (Denmark)

    Aagaard Madsen, Helge; Barlas, Athanasios; Løgstrup Andersen, Tom

    2015-01-01

    The morphing trailing edge system or flap system, CRTEF, has been developed over the last 10 years at DTU Wind Energy. After a promising wind tunnel test of the system in 2009 the INDUFLAP project has been carried out from 2011-2014 to transfer the technology from laboratory to industrial...... manufacturing and application. To narrow the gap between wind tunnel testing and full scale prototype testing we developed the rotating test rig. The overall objectives with the rotating test rig are: 1) to test the flap system in a realistic rotating environment with a realistic g-loading; 2) to measure...... the flap performance in real turbulent inflow and 3) to test the flap system in a realistic size and Reynolds number when comparing with full scale applications.. The rotating test rig consists of a 2.2m blade section attached to a 10m boom and mounted on a 100kW turbine platform. It was installed in June...

  20. Test rig overview for validation and reliability testing of shutdown system software

    International Nuclear Information System (INIS)

    Zhao, M.; McDonald, A.; Dick, P.

    2007-01-01

    The test rig for Validation and Reliability Testing of shutdown system software has been upgraded from the AECL Windows-based test rig previously used for CANDU6 stations. It includes a Virtual Trip Computer, which is a software simulation of the functional specification of the trip computer, and a real-time trip computer simulator in a separate chassis, which is used during the preparation of trip computer test cases before the actual trip computers are available. This allows preparation work for Validation and Reliability Testing to be performed in advance of delivery of actual trip computers to maintain a project schedule. (author)

  1. A reciprocating pin-on-plate test-rig for studying friction materials for holding brakes

    DEFF Research Database (Denmark)

    Poulios, Konstantinos; Drago, Nicola; Klit, Peder

    2014-01-01

    -on-plate test-rig for studying the evolution of wear by monitoring the pin height reduction using Eddy-current proximity sensors is presented. Moreover, a new mechanism for recording the friction force is suggested. Apart from the design of the test-rig, friction force and wear rate measurements for two...

  2. Flap testing on the rotating test rig in the INDUFLAP project

    DEFF Research Database (Denmark)

    Barlas, Athanasios; Aagaard Madsen, Helge; Enevoldsen, Karen

    Tests of a prototype Controllable Rubber Trailing Edge Flap (CRTEF) are performed on the rotating test rig at the Risø campus of DTU. The general description and objectives are presented, along with an overview of sensors on the setup and the test cases. The post-processing of data is discussed...

  3. Development of a test rig and its application for validation and reliability testing of safety-critical software

    Energy Technology Data Exchange (ETDEWEB)

    Thai, N D; McDonald, A M [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)

    1996-12-31

    This paper describes a versatile test rig developed by AECL for functional testing of safety-critical software used in the process trip computers of the Wolsong CANDU stations. The description covers the hardware and software aspects of the test rig, the test language and its interpreter, and other major testing software utilities such as the test oracle, sampler and profiler. The paper also discusses the application of the rig in the final stages of testing of the process trip computer software, namely validation and reliability tests. It shows how random test cases are generated, test scripts prepared and automatically run on the test rig. The versatility of the rig is further demonstrated in other types of testing such as sub-system tests, verification of the test oracle, testing of newly-developed test script, self-test and calibration. (author). 5 tabs., 10 figs.

  4. Development of a test rig and its application for validation and reliability testing of safety-critical software

    International Nuclear Information System (INIS)

    Thai, N.D.; McDonald, A.M.

    1995-01-01

    This paper describes a versatile test rig developed by AECL for functional testing of safety-critical software used in the process trip computers of the Wolsong CANDU stations. The description covers the hardware and software aspects of the test rig, the test language and its interpreter, and other major testing software utilities such as the test oracle, sampler and profiler. The paper also discusses the application of the rig in the final stages of testing of the process trip computer software, namely validation and reliability tests. It shows how random test cases are generated, test scripts prepared and automatically run on the test rig. The versatility of the rig is further demonstrated in other types of testing such as sub-system tests, verification of the test oracle, testing of newly-developed test script, self-test and calibration. (author). 5 tabs., 10 figs

  5. Substantial Fatigue Similarity of a New Small-Scale Test Rig to Actual Wheel-Rail System

    NARCIS (Netherlands)

    Naeimi, M.; Li, Z.; Petrov, R.H.; Dollevoet, R.P.B.J.; Sietsma, J.; Wu, J.

    2014-01-01

    The substantial similarity of fatigue mechanism in a new test rig for rolling contact fatigue (RCF) has been investigated. A new reduced-scale test rig is designed to perform controlled RCF tests in wheel-rail materials. The fatigue mechanism of the rig is evaluated in this study using a combined

  6. The Hydraulic Test Procedure for Non-instrumented Irradiation Test Rig of Annular Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dae Ho; Lee, Kang Hee; Shin, Chang Hwan; Park, Chan Kook

    2008-08-15

    This report presents the procedure of pressure drop test, vibration test and endurance test for the non-instrumented rig using the irradiation test in HANARO of advanced PWR annular fuel which were designed and fabricated by KAERI. From the out-pile thermal hydraulic tests, confirm the flow rate at the 200 kPa pressure drop and measure the RMS displacement at this time. And the endurance test is confirmed the wear and the integrity of the non-instrumented rig at the 110% design flow rate. This out-pile test perform the Flow-Induced Vibration and Pressure Drop Experimental Tester(FIVPET) facility. The instruments in FIVPET facility was calibrated in KAERI and the pump and the thermocouple were certified by manufacturer.

  7. Development of a low cost test rig for standalone WECS subject to electrical faults.

    Science.gov (United States)

    Himani; Dahiya, Ratna

    2016-11-01

    In this paper, a contribution to the development of low-cost wind turbine (WT) test rig for stator fault diagnosis of wind turbine generator is proposed. The test rig is developed using a 2.5kW, 1750 RPM DC motor coupled to a 1.5kW, 1500 RPM self-excited induction generator interfaced with a WT mathematical model in LabVIEW. The performance of the test rig is benchmarked with already proven wind turbine test rigs. In order to detect the stator faults using non-stationary signals in self-excited induction generator, an online fault diagnostic technique of DWT-based multi-resolution analysis is proposed. It has been experimentally proven that for varying wind conditions wavelet decomposition allows good differentiation between faulty and healthy conditions leading to an effective diagnostic procedure for wind turbine condition monitoring. Copyright © 2016 ISA. Published by Elsevier Ltd. All rights reserved.

  8. The Necessity of a New Type Test Rig for the Development of an Evaluation Method in Grid Fretting Problems

    International Nuclear Information System (INIS)

    Lee, Young-Ho; Kim, Hyung-Kyu

    2007-01-01

    A grid fretting problem is recognized as one of the most important degradation mechanisms even though the examination results of fretting experiments could be applied to the development and design of spacer grid structures. This is because it is difficult to develop a fretting wear model for a grid fretting problem due to the various wear mechanisms involved according to the mechanical and environmental variables, the contact condition with a spring/dimple and the material properties. A number of spring shapes has been developed in KAERI and their performance tests such as fretting wear, flow-induced vibration (FIV) tests, etc. have been carried out from a part unit to a full assembly scale. From the unit part fretting test results, one of the noticeable results is that the contacting force (normal load) was gradually decreased with increasing number of fretting cycles due to a depth increase and this behavior was closely related to the contacting spring shape. When considering the actual contact condition between a fuel rod and a spring/dimple, if a fretting wear progresses due to a FIV under a specific normal load exerted on the fuel rod by an elastic deformation of the spring, the contacting force between the fuel rod and dimple that are located in the opposite side should be decreased. Consequently, an evaluation of developed spacer grids against fretting wear damage should be performed with the results of 1x1 cell unit experiments because a contacting force is one of the most important variables that influences a fretting wear mechanism. The discussion was focused on the development procedure of a new test rig and its performance by using a 1x1 cell unit test rig. (authors)

  9. The development of fuel pins and material specimens mixed loading irradiation test rig in the experimental fast reactor Joyo. The development of the fuel-material hybrid rig

    International Nuclear Information System (INIS)

    Oyamatsu, Yasuko; Someya, Hiroyuki

    2013-02-01

    In the experimental fast reactor Joyo, there were many tests using the irradiation rigs that it was possible to be set irradiation conditions for each compartment independently. In case of no alternative fuel element to irradiate after unloading the irradiated compartments, the irradiation test was restarted with the dummy compartment which the fuel elements was not mounted. If the material specimens are mounted in this space, it is possible to use the irradiation space effectively. For these reasons, the irradiation rig (hybrid rig) is developed that is consolidated with material specimens compartment and fuel elements compartment. Fuel elements and material specimens differ greatly with heat generation, so that the most important issue in developing of hybrid rig is being able to distribute appropriately the coolant flow which satisfies irradiation conditions. The following is described by this report. (1) It was confirmed that the flow distribution of loading the same irradiation rig with the compartment from which a flow demand differs could be satisfied. (2) It was confirmed that temperature setting range of hybrid rig could be equivalent to that of irradiation condition. (3) By standardizing the coolant entrance structure of the compartment lower part, the prospect which can perform easily recombination of the compartment from which a type differs between irradiation rigs was acquired. (author)

  10. Fuel rod simulator effects in flooding experiments single rod tests

    International Nuclear Information System (INIS)

    Nishida, M.

    1984-09-01

    The influence of a gas filled gap between cladding and pellet on the quenching behavior of a PWR fuel rod during the reflood phase of a LOCA has been investigated. Flooding experiments were conducted with a short length electrically heated single fuel rod simulator surrounded by glass housing. The gap of 0.05 mm width between the Zircaloy cladding and the internal Al 2 O 3 pellets of the rod was filled either wit helium or with argon to vary the radial heat resistance across the gap. This report presents some typical data and an evaluation of the reflood behavior of the fuel rod simulator used. The results show that the quench front propagates faster for increasing heat resistance in the gap between cladding and heat source of the rod. (orig.) [de

  11. Development of Induction Brazing System for Sealing Instrumentation Feed through Part of Nuclear Fuel Test Rig

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Jintae; Kim, Kahye; Heo, Sungho; Ahn, Sungho; Joung, Changyoung; Son, Kwangjae; Jung, Yangil [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-12-15

    To test the performance of nuclear fuels, coolant needs to be circulated through the test rig installed in the test loop. Because the pressure and temperature of the coolant is 15.5 MPa and 300 .deg. C respectively, coolant sealing is one of the most important processes in fabricating a nuclear fuel test rig. In particular, 15 instrumentation cables installed in a test rig pass through the pressure boundary, and brazing is generally applied as a sealing method. In this study, an induction brazing system has been developed using a high frequency induction heater including a vacuum chamber. For application in the nuclear field, BNi2 should be used as a paste, and optimal process variables for Ni brazing have been found by several case studies. The performance and soundness of the brazed components has been verified by a tensile test, cross section test, and sealing performance test.

  12. Joint test rig for tests and calibration of different methods of two-phase mass flow measurement

    International Nuclear Information System (INIS)

    John, H.; Erbacher, F.; Wanner, E.

    1975-01-01

    On behalf of the Federal Ministry of Research and Technology, the Institute of Reactor Components (IRB) has begun building a test rig which will be used for testing and calibrating the methods of measuring non-steady state two-phase mass flows developed by various research agencies. The test rig is designed for the generation of steam-water mixtures of any mixing ratio and a maximum pressure of 160 data. Depending on the mixing ratio, the mass flow will reach a maximum level of 10 to 20 t/h. The conceptual design phase of the test rig has largely been finished, the component ordering phase has begun. (orig.) [de

  13. LOFT fuel rod surface temperature measurement testing

    International Nuclear Information System (INIS)

    Eaton, A.M.; Tolman, E.L.; Solbrig, C.W.

    1978-01-01

    Testing of the LOFT fuel rod cladding surface thermocouples has been performed to evaluate how accurately the LOFT thermocouples measure the cladding surface temperature during a loss-of-coolant accident (LOCA) sequence and what effect, if any, the thermocouple would have on core performance. Extensive testing has been done to characterize the thermocouple design. Thermal cycling and corrosion testing of the thermocouple weld design have provided an expected lifetime of 6000 hours when exposed to reactor coolant conditions of 620 K and 15.9 MPa and to sixteen thermal cycles with an initial temperature of 480 K and peak temperatures ranging from 870 to 1200K. Departure from nucleate boiling (DNB) tests have indicated a DNB penalty (5 to 28% lower) during steady state operation and negligible effects during LOCA blowdown caused by the LOFT fuel rod surface thermocouple arrangement. Experience with the thermocouple design in Power Burst Facility (PBF) and LOFT nonnuclear blowdown testing has been quite satisfactory. Tests discussed here were conducted using both stainless steel and zircaloy-clad electrically heated rod in the LOFT Test Support Facility (LTSF) blowdown simulation loop

  14. The Hydraulic Test Report for Non-instrumented Irradiation Test Rig of DUO-Cooled Annular Pellet

    International Nuclear Information System (INIS)

    Kim, Dae Ho; Lee, Kang Hee; Shin, Chang Hwan; Yang, Yong Sik; Kim, Sun Ki; Bang, Je Geon; Song, Kun Woo

    2007-08-01

    This report presents the results of pressure drop test and vibration test for non-instrumented rig of Advanced PWR DUO-Fuel Annular Pellet which were designed and fabricated by KAERI. From the pressure drop test results, it is noted that the flow velocity across the non-instrumented rig of Advanced PWR DUO-Fuel Annular Pellet corresponding to the pressure drop of 200 kPa is measured to be about 8.30 kg/sec. Vibration frequency results for the non-instrumented rig at the pump spin frequency ranges from 19.0 to 32.0 Hz, RMS(Root Mean Square) displacement for the non-instrumented rig of Advanced PWR DUO-Fuel Annular Pellet is less than 7.25 m, and the maximum displacement is less than 31.27 μm. This test was performed at the FIVPET facility

  15. Verification test of control rod system for HTR-10

    International Nuclear Information System (INIS)

    Zhou Huizhong; Diao Xingzhong; Huang Zhiyong; Cao Li; Yang Nianzu

    2002-01-01

    There are 10 sets of control rods and driving devices in 10 MW High Temperature Gas-cooled Test Reactor (HTR-10). The control rod system is the controlling and shutdown system of HTR-10, which is designed for reactor criticality, operation, and shutdown. In order to guarantee technical feasibility, a series of verification tests were performed, including room temperature test, thermal test, test after control rod system installed in HTR-10, and test of control rod system before HTR-10 first criticality. All the tests data showed that driving devices working well, control rods running smoothly up and down, random position settling well, and exactly position indicating

  16. Grey Rod Test in HANARO Reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-08-15

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

  17. Design and Demonstration of a Test-Rig for Static Performance-Studies of Permanent Magnet Couplings

    DEFF Research Database (Denmark)

    Högberg, Stig; Jensen, Bogi Bech; Bendixen, Flemming Buus

    2013-01-01

    The design and construction of an easy-to-use test-rig for permanent magnet couplings is presented. Static torque of permanent magnet couplings as a function of angular displacement is measured of permanent magnet couplings through an semi-automated test system. The test-rig is capable of measuring...

  18. Development and verification of a reciprocating test rig designed for investigation of piston ring tribology

    DEFF Research Database (Denmark)

    Pedersen, Michael Torben; Imran, Tajammal; Klit, Peder

    2009-01-01

    This paper describes the development and verification of a reciprocating test rig, which was designed to study the piston ring tribology. A crank mechanism is used to generate a reciprocating motion for a moving plate, which acts as the liner. A stationary block acting as the ring package is loaded......, which is suitable for the study of piston ring tribology....

  19. Computational Analysis of a South African Mobile Trailer-Type Medium Sized Tyre Test Rig

    CSIR Research Space (South Africa)

    Sharma, Shikar

    2015-04-01

    Full Text Available To support the South African National Defence Force with their vehicle mobility needs, the CSIR has begun characterising tyres by using a medium, trailer-type, tyre test rig. Two different Pacejka tyre models were generated using two independent...

  20. A compact internal drum test rig for measurements of rolling contact forces between a single tread block and a substrate

    NARCIS (Netherlands)

    Lundberg, O.E.; Kari, L.; Lopez Arteaga, I.

    2017-01-01

    A novel test rig design is presented which enables detailed studies of the three force components generated in the impact and release phase of rolling contact between a tyre tread block and a substrate. The design of the compact internal drum test rig provides realistic impact and release angles for

  1. The Design and Manufacturing Report of Plug Type Non-Instrumented Rig for Irradiation Test in HANARO OR Hole

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dae Ho; Bang, Je Geon; Lim, Ik Sung; Kim, Sun Ki; Yang, Yong Sik; Song, Kun Woo

    2008-09-15

    This project is developed the plug type non-instrumented irradiation test rig of the advanced nuclear fuel in HANARO for pursuit advanced performance in High Performance Fuel Technology Development as a part Nuclear Mid and Long-term R and D Program. This irradiation rig was confirmed the integrity and HANARO core compatibility by the optimum design and the thermal hydraulic out-pile test in FIVPET. The characteristic of plug type non-instrument rig is to possible irradiation test of variable in-pile condition and reduced the wastes for reusable as function. This plug type non-instrumented rig was satisfied the quality assurance requirements and written out the end of manufacturing report. This plug type non-instrumented rig is adopt to the irradiation test for nuclear fuel irradiation test in HANARO OR hole.

  2. Performance Analysis of Retrofitted Tribo-Corrosion Test Rig for Monitoring In Situ Oil Conditions

    Directory of Open Access Journals (Sweden)

    Arpith Siddaiah

    2017-09-01

    Full Text Available Oils and lubricants, once extracted after use from a mechanical system, can hardly be reused, and should be refurbished or replaced in most applications. New methods of in situ oil and lubricant efficiency monitoring systems have been introduced for a wide variety of mechanical systems, such as automobiles, aerospace aircrafts, ships, offshore wind turbines, and deep sea oil drilling rigs. These methods utilize electronic sensors to monitor the “byproduct effects” in a mechanical system that are not indicative of the actual remaining lifecycle and reliability of the oils. A reliable oil monitoring system should be able to monitor the wear rate and the corrosion rate of the tribo-pairs due to the inclusion of contaminants. The current study addresses this technological gap, and presents a novel design of a tribo-corrosion test rig for oils used in a dynamic system. A pin-on-disk tribometer test rig retrofitted with a three electrode-potentiostat corrosion monitoring system was used to analyze the corrosion and wear rate of a steel tribo-pair in industrial grade transmission oil. The effectiveness of the retrofitted test rig was analyzed by introducing various concentrations of contaminants in an oil medium that usually leads to a corrosive working environment. The results indicate that the retrofitted test rig can effectively monitor the in situ tribological performance of the oil in a controlled dynamic corrosive environment. It is a useful method to understand the wear–corrosion synergies for further experimental work, and to develop accurate predictive lifecycle assessment and prognostic models. The application of this system is expected to have economic benefits and help reduce the ecological oil waste footprint.

  3. Performance Analysis of Retrofitted Tribo-Corrosion Test Rig for Monitoring In Situ Oil Conditions.

    Science.gov (United States)

    Siddaiah, Arpith; Khan, Zulfiqar Ahmad; Ramachandran, Rahul; Menezes, Pradeep L

    2017-09-28

    Oils and lubricants, once extracted after use from a mechanical system, can hardly be reused, and should be refurbished or replaced in most applications. New methods of in situ oil and lubricant efficiency monitoring systems have been introduced for a wide variety of mechanical systems, such as automobiles, aerospace aircrafts, ships, offshore wind turbines, and deep sea oil drilling rigs. These methods utilize electronic sensors to monitor the "byproduct effects" in a mechanical system that are not indicative of the actual remaining lifecycle and reliability of the oils. A reliable oil monitoring system should be able to monitor the wear rate and the corrosion rate of the tribo-pairs due to the inclusion of contaminants. The current study addresses this technological gap, and presents a novel design of a tribo-corrosion test rig for oils used in a dynamic system. A pin-on-disk tribometer test rig retrofitted with a three electrode-potentiostat corrosion monitoring system was used to analyze the corrosion and wear rate of a steel tribo-pair in industrial grade transmission oil. The effectiveness of the retrofitted test rig was analyzed by introducing various concentrations of contaminants in an oil medium that usually leads to a corrosive working environment. The results indicate that the retrofitted test rig can effectively monitor the in situ tribological performance of the oil in a controlled dynamic corrosive environment. It is a useful method to understand the wear-corrosion synergies for further experimental work, and to develop accurate predictive lifecycle assessment and prognostic models. The application of this system is expected to have economic benefits and help reduce the ecological oil waste footprint.

  4. A Study on the Dynamic Analysis of the Nuclear Fuel Test Rig Using 1-Way Fluid-Structure Coupled Analysis

    International Nuclear Information System (INIS)

    Yang, Tae-Ho; Hong, Jin-Tae; Ahn, Sung-Ho; Joung, Chang-Young; Heo, Sung-Ho; Jang, Seo-Yun

    2015-01-01

    1-way fluid-structure coupled analysis is used to estimate the dynamic characteristic of the fuel test rig. the motion at the bottom of the test rig is confirmed. The maximum deformation of the test rig is 0.11 mm. The structural integrity of the test rig is performed by using the comparison with the Von-mises stress of the analysis and yield stress of the material. It is evaluated that the motion at the bottom of the test rig is able to cause other structural problem. Using the 2-way fluid-structural coupled analysis, the structural integrity of the test rig will be performed in further paper. The cooling water with specific flow rate was flowed in the nuclear fuel test rig. The structural integrity of the test rig was affected by the vibration. The fluid-induced vibration test had to be performed to obtain the amplitude of the vibration on the structure. Various test systems was developed. Flow-induced vibration and pressure drop experimental tester was developed in Korea Atomic Energy Research Institute. The vibration test with high fluid flow rate was difficult by the tester. To generate the nuclear fuel test environment, coolant flow simulation system was developed. The scaled nuclear fuel test was able to be performed by the simulation system. The mock-up model of the test rig was used in the simulation system. The mock-up model in the simulation system was manufactured with scaled down full model. In this paper, the fluid induced vibration characteristic of the full model in the nuclear fuel test is studied. The hydraulic pressure on the velocity of the fluid was calculated. The static structure analysis was performed by using the pressure. The structural integrity was assessed using the results of the analysis

  5. TITAN - a 9 MW, 179 bar pressurised water rig

    International Nuclear Information System (INIS)

    Mogford, D.J.; Lee, D.H.

    1987-02-01

    The report describes the TITAN rig built at Winfrith for thermal hydraulic experiments with water at up to 179 bar pressure. A power supply of 9 MW is available. The report describes three typical experiments that show the versatility of the rig. The first is a 25 rod pressurized water reactor fuel bundle critical heat flux experiment, the second is a parallel channel evaporator test and the third is a model jet pump test. (author)

  6. Mechanical Design of a Performance Test Rig for the Turbine Air-Flow Task (TAFT)

    Science.gov (United States)

    Forbes, John C.; Xenofos, George D.; Farrow, John L.; Tyler, Tom; Williams, Robert; Sargent, Scott; Moharos, Jozsef

    2004-01-01

    To support development of the Boeing-Rocketdyne RS84 rocket engine, a full-flow, reaction turbine geometry was integrated into the NASA-MSFC turbine air-flow test facility. A mechanical design was generated which minimized the amount of new hardware while incorporating all test and instrumentation requirements. This paper provides details of the mechanical design for this Turbine Air-Flow Task (TAFT) test rig. The mechanical design process utilized for this task included the following basic stages: Conceptual Design. Preliminary Design. Detailed Design. Baseline of Design (including Configuration Control and Drawing Revision). Fabrication. Assembly. During the design process, many lessons were learned that should benefit future test rig design projects. Of primary importance are well-defined requirements early in the design process, a thorough detailed design package, and effective communication with both the customer and the fabrication contractors.

  7. Threshold Assessment of Gear Diagnostic Tools on Flight and Test Rig Data

    Science.gov (United States)

    Dempsey, Paula J.; Mosher, Marianne; Huff, Edward M.

    2003-01-01

    A method for defining thresholds for vibration-based algorithms that provides the minimum number of false alarms while maintaining sensitivity to gear damage was developed. This analysis focused on two vibration based gear damage detection algorithms, FM4 and MSA. This method was developed using vibration data collected during surface fatigue tests performed in a spur gearbox rig. The thresholds were defined based on damage progression during tests with damage. The thresholds false alarm rates were then evaluated on spur gear tests without damage. Next, the same thresholds were applied to flight data from an OH-58 helicopter transmission. Results showed that thresholds defined in test rigs can be used to define thresholds in flight to correctly classify the transmission operation as normal.

  8. High temperature helium test rig with prestressed concrete pressure vessel

    International Nuclear Information System (INIS)

    Schmidl, H.

    1975-10-01

    The report gives a short description of the joint project prestressed concrete vessel-helium test station as there is the building up of the concrete structure, the system of instrumentation, the data processing, the development of the helium components as well as the testing programs. (author)

  9. Accelerated Bearing Life-time Test Rig Development for Low Speed Data Acquisition

    Directory of Open Access Journals (Sweden)

    Andreas Klausen

    2017-07-01

    Full Text Available Condition monitoring plays an important role in rotating machinery to ensure reliability of the equipment, and to detect fault conditions at an early stage. Although health monitoring methodologies have been thoroughly developed for rotating machinery, low-speed conditions often pose a challenge due to the low signal-to-noise ratio. To this aim, sophisticated algorithms that reduce noise and highlight the bearing faults are necessary to accurately diagnose machines undergoing this condition. In the development phase, sensor data from a healthy and damaged bearing rotating at low-speed is required to verify the performance of such algorithms. A test rig for performing accelerated life-time testing of small rolling element bearings is designed to collect necessary sensor data. Heavy loads at high-speed conditions are applied to the test bearing to wear it out fast. Sensor data is collected in intervals during the test to capture the degeneration features. The main objective of this paper is to provide a detailed overview for the development and analysis of this test rig. A case study with experimental vibration data is also presented to illustrate the efficacy of the developed test rig.

  10. Developmental test report, assessment of XT-70E percussion drill rig operation in tank farms

    International Nuclear Information System (INIS)

    Dougherty, L.F.

    1996-01-01

    The following report documents the testing of the XT-70E percussion drill rig for use in the 241-SX Tank Farm. The test is necessary to support evaluation of the safety and authorization level of the proposed activity of installing up to three new drywells in the 241- SX Tank Farm. The proposed activity plans to install drywells by percussion drilling 7 inch O.D./6 inch I.D. pipe in close proximity of underground storage tanks and associated equipment. The load transmitted from the drill rig's percussion hammer through the ground to the tank structure and equipment is not known and therefore testing is required to ensure the activity is safe and authorized

  11. High temperature corrosion investigation in an oxyfuel combustion test rig

    DEFF Research Database (Denmark)

    Montgomery, Melanie; Bjurman, M.; Hjörnhede, A

    2014-01-01

    Oxyfuel firing and subsequent capture of CO2 is a way to reduce CO2 emissions from coal‐fired boilers. Literature is summarized highlighting results which may contribute to understanding of the corrosion processes in an oxyfuel boiler.Tests were conducted in a 500 kWth oxyfuel test facility...... constructed by Brandenburg Technical University to gain understanding into oxyfuel firing. Two air‐cooled corrosion probes were exposed in this oxyfuel combustion chamber where the fuel was lignite. Gas composition was measured at the location of testing. Various alloys from a 2½ Cr steel, austenitic steels...... to nickel alloys were exposed at set metal temperatures of 570 and 630 °C for 287 h. The specimens were investigated using light optical and scanning electron microscopy and X‐ray diffraction.The deposit on the probe contained predominantly CaSO4 and Fe2O3. Oxide thickness and depth of the precipitated...

  12. The Thermal-hydraulic Performance Test Report for the Non-instrumented Irradiation Test Rig of Annular Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dae Ho; Lee, Kang Hee; Shin, Chang Hwan

    2008-09-15

    This report presents the results of pressure drop test, vibration test and endurance test for the non-instrumented rig using the irradiation test in HANARO of the double cooled annular fuel which were designed and fabricated by KAERI. From the out-pile thermal hydraulic tests, corresponding to the pressure drop of 200 kPa is measured to be about 9.72 kg/sec. Vibration frequency for the non-instrumented rig ranges from 5.0 to 10.7 kg/s. RMS(Root Mean Square) displacement for non-instrumented rig is less than 11.73 m, and the maximum displacement is less than 54.87m. The flow rate for endurance test were 10.5 kg/s, which was 110% of 9.72 kg/s. And the endurance test was carried out for 3 days. The test results found not to the wear and satisfied to the limits of pressure drop, flow rate, vibration and wear in the non-instrumented rig. This test was performed at the FIVPET facility.

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

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  14. Improvement of digital data acquisition system in reflood test rig

    International Nuclear Information System (INIS)

    Sudoh, Takashi; Murao, Yoshio; Niitsuma, Yasushi

    1979-03-01

    The original master digital data acquisition system was designed to collect 30 channels of analog data rapidly and convert them into digital form for recording on a magnetic tape. Due to the increases in the number of channels and the ranges of measurement, an additional acquisition device was needed for the original system. This report descrives the design of the additional data acquisition device and the results of performance tests. The operational manual is attached as an appendix. It was confirmed that the new system satisfied the requirements of system. (author)

  15. Finite element calculation of fields around the end region of a turbine generator test rig

    Energy Technology Data Exchange (ETDEWEB)

    Eastham, J.F.; Rodger, D.; Lai, H.C.; Nouri, H. (Univ. of Bath (United Kingdom))

    1993-03-01

    Under transient conditions, most often caused by faults in the power system, unbalanced load is presented to a turbine generator. This gives rise to airgap fields which do not travel at the speed of the rotor, and cause induced currents which occur in the solid steel surface. This can cause high local heating. The current path is generally in the axial direction of the machine but the distribution in the end region is not so well known. Here, comparisons are drawn between the use of surface impedance elements and volume elements when modeling a test rig using the MEGA package. The test rig is representative of a turbine generator. The work is supported by practical measurements.

  16. Hot corrosion testing of Ni-based alloys and coatings in a modified Dean rig

    Science.gov (United States)

    Steward, Jason Reid

    Gas turbine blades are designed to withstand a variety of harsh operating conditions. Although material and coating improvements are constantly administered to increase the mean time before turbine refurbishment or replacement, hot corrosion is still considered as the major life-limiting factor in many industrial and marine gas turbines. A modified Dean rig was designed and manufactured at Tennessee Technological University to simulate the accelerated hot corrosion conditions and to conduct screening tests on the new coatings on Ni-based superalloys. Uncoated Ni-based superalloys, Rene 142 and Rene 80, were tested in the modified Dean rig to establish a testing procedure for Type I hot corrosion. The influence of surface treatments on the hot corrosion resistance was then investigated. It was found that grit-blasted specimens showed inferior hot corrosion resistance than that of the polished counterpart. The Dean rig was also used to test model MCrAlY alloys, pack cementation NiAl coatings, and electro-codeposited MCrAlY coatings. Furthermore, the hot corrosion attack on the coated-specimens were also assessed using a statistical analysis approach.

  17. A thermal-hydraulic test rig for advanced fast reactor fuel assemblies

    International Nuclear Information System (INIS)

    Rapier, A.C.

    1989-03-01

    A new design of fast reactor fuel assemblies has been proposed in which the pins are supported in grids attached to the wrapper by flexible skirts. Coolant mixing is enhanced by the skirts diverting flow into the cluster of pins at each grid. There are insufficient empirical data available for the detailed design of the skirt or for the input to computer calculations of flow and heat transfer. A test rig to provide these data has been designed and built. (author)

  18. Evaluation of fuel rods behavior - under irradiation test

    International Nuclear Information System (INIS)

    Lameiras, F.S.; Terra, J.L.; Pinto, L.C.M.; Dias, M.S.; Pinheiro, R.B.

    1981-04-01

    By the accompanying of the irradiation of instrumented test fuel rods simulating the operational conditions in reactors, plus the results of post - irradiation exams, tests, evaluation and calibration of analitic modelling of such fuel rods is done. (E.G.) [pt

  19. A universal suspension test rig for electrohydraulic active and passive automotive suspension system

    Directory of Open Access Journals (Sweden)

    Mahmoud Omar

    2017-12-01

    Full Text Available A fully active electro-hydraulic and passive automotive quarter car suspensions with their experimental test-rigs are designed and implemented. Investigation of the active performance compared against the passive is performed experimentally and numerically utilizing SIMULINK's Simscape library. Both systems are modeled as single-degree-of-freedom in order to simplify the validation process. Economic considerations were considered during the rig's implementation. The rig consists of two identical platforms fixed side by side allowing testing two independent suspensions simultaneously. Position sensors for sprung and unsprung masses on both platforms are installed. The road input is introduced by a cam and a roller follower mechanism driven by 1.12 kW single phase induction motor with speed reduction assembly. The active hydraulic cylinder was the most viable choice due to its high power-to-weight ratio. The active control is of the proportional-integral-differential (PID type. Though this technique is quite simple and not new, yet the emphasis of this paper is the engineering, design and implementation of the experimental setup and controller. A successful validation process is performed. Ride comfort significantly improved with active suspension, as shown by the results; 24.8% sprung mass vibration attenuation is achieved. The details of the developed system with the analytical and experimental results are presented. Keywords: Active suspension, Passive suspension, Servo, Hydraulic, Control, PID

  20. Ejected control rod and rods drop measurements during Mochovce startup physical tests

    International Nuclear Information System (INIS)

    Minarcin, Miroslav; Elko, Marek

    1998-01-01

    Paper deals with measurements of asymmetric reactivity insertion into the reactor core that were carried out during physical startup tests of Mochovce Unit 1 in June 1998. Control rods worth measurements with one and two rods s tucked in upper limit and worth measurement of one control rod from group 6 'ejected' from the reactor core are discussed. During the experiments neutron flux was measured by four ionisation chambers (three of them were placed symmetrically around the reactor core). Results of measurements and influence of asymmetric reactivity influence on ionisation chambers response are presented in the paper. (Authors)

  1. A novel test rig to investigate under-platform damper dynamics

    Science.gov (United States)

    Botto, Daniele; Umer, Muhammad

    2018-02-01

    In the field of turbomachinery, vibration amplitude is often reduced by dissipating the kinetic energy of the blades with devices that utilize dry friction. Under-platform dampers, for example, are often placed in the underside of two consecutive turbine blades. Dampers are kept in contact with the under-platform of the respective blades by means of the centrifugal force. If the damper is well designed, vibration of blades instigate a relative motion between the under-platform and the damper. A friction force, that is a non-conservative force, arises in the contact and partly dissipates the vibration energy. Several contact models are available in the literature to simulate the contact between the damper and the under-platform. However, the actual dynamics of the blade-damper interaction have not fully understood yet. Several test rigs have been previously developed to experimentally investigate the performance of under-platform dampers. The majority of these experimental setups aim to evaluate the overall damper efficiency in terms of reduction in response amplitude of the blade for a given exciting force that simulates the aerodynamic loads. Unfortunately, the experimental data acquired on the blade dynamics do not provide enough information to understand the damper dynamics. Therefore, the uncertainty on the damper behavior remains a big issue. In this work, a novel experimental test rig has been developed to extensively investigate the damper dynamic behavior. A single replaceable blade is clamped in the rig with a specific clamping device. With this device the blade root is pressed against a groove machined in the test rig. The pushing force is controllable and measurable, to better simulate the actual centrifugal load acting on the blade. Two dampers, one on each side of the blade, are in contact with the blade under-platforms and with platforms on force measuring supports. These supports have been specifically designed to measure the contact forces on the

  2. Design and simulation of the rotating test rig in the INDUFLAP project

    DEFF Research Database (Denmark)

    Barlas, Thanasis K.; Aagaard Madsen, Helge; Løgstrup Andersen, Tom

    The general description and objectives of the rotating test rig at the Risø campus of DTU are presented, as used for the aeroelastic testing of a controllable rubber trailing edge flap (CRTEF) system in the INDUFLAP project. The design of all new components is presented, including the electrical...... drive, the pitch system, the boom, and the wing/flap section. The overall instrumentation of the components used for the aeroelastic testing is described. Moreover, the aeroelastic model simulating the setup is described, and predictions of steady and dynamic loading along with the aeroelastic analysis...

  3. Lifting devices with minimum effort for testing, maintenance and repair at the example of a lifting rig for core internals

    Energy Technology Data Exchange (ETDEWEB)

    Pache, Martin [Westinghouse Electric Germany GmbH (Germany); Wiesendanger, Robert [Kernkraftwerk Beznau, NOK (Switzerland)

    2008-07-01

    Beznau is a Westinghouse built nuclear power plant in the Aargau area Switzerland. It consists of two PWR units, each providing 365 MWe net capacity. The units were set into operation in 1969 and 1972, respectively, and hold an unlimited license for operation, provided they continue to fulfill current legal and security requirements. Beznau's previous lifting rigs for core internals required a high effort in testing and maintenance. Moreover, a damage to one of the rigs nearly resulted in the inoperability of the rig. However, no element of the load chain was affected, so there was no danger of a crash, but it could have caused an extended outage. Hence, it was decided to replace the lifting rigs with a state-of-the-art functional design that reflects modern requirements on maintenance and testing. Although the plant was built to ASME standards and codes, the new lifting rigs have been designed to German KTA code for lifting devices (KTA 3902 / 3903 for equipment with increased requirements, as per section 4.3 of KTA 3902). Given KTA's demands on periodic testing, one main requirement on the new design was to minimize the testing effort for the new rigs. (orig.)

  4. A durability test rig and methodology for erosion-resistant blade coatings in turbomachinery

    Science.gov (United States)

    Leithead, Sean Gregory

    A durability test rig for erosion-resistant gas turbine engine compressor blade coatings was designed, completed and commissioned. Bare and coated 17-4PH steel V103-profile blades were rotated at up to 11500 rpm and impacted with Garnet sand for 5 hours at an average concentration of 2.51 gm3of air , at a blade leading edge Mach number of 0.50. The rig was determined to be an acceptable first stage axial compressor representation. Two types of 16 microm-thick coatings were tested: Titanium Nitride (TiN) and Chromium-Aluminum-Titanium Nitride (CrAlTiN), both applied using an Arc Physical Vapour Deposition technique at the National Research Council in Ottawa, Canada. A Leithead-Allan-Zhao (LAZ) score was created to compare the durability performance of uncoated and coated blades based on mass-loss and blade dimension changes. The bare blades' LAZ score was set as a benchmark of 1.00. The TiN-coated and CrAlTiN-coated blades obtained LAZ scores of 0.69 and 0.41, respectively. A lower score meant a more erosion-resistant coating. Major modes of blade wear included: trailing edge, leading edge and the rear suction surface. Trailing edge thickness was reduced, the leading edge became blunt, and the rear suction surface was scrubbed by overtip and recirculation zone vortices. It was found that the erosion effects of vortex flow were significant. Erosion damage due to reflected particles was not present due to the low blade solidity of 0.7. The rig is best suited for studying the performance of erosion-resistant coatings after they are proven effective in ASTM standardized testing. Keywords: erosion, compressor, coatings, turbomachinery, erosion rate, blade, experimental, gas turbine engine

  5. Investigation of intracochlear dual actuator stimulation in a scaled test rig

    Directory of Open Access Journals (Sweden)

    van Drunen Wouter J.

    2017-09-01

    Full Text Available For patients suffering from profound hearing loss or deafness still having respectable residual hearing in the low frequency range, the combination of a hearing aid with a cochlear implant results in the best quality of hearing perception (EAS – electric acoustic stimulation. In order to optimize EAS, ongoing research focusses on the integration of these stimuli in a single implant device. Within this study, the performance of piezoelectric actuators, particularly the dual actuator stimulation, in a scaled uncoiled test rig was investigated.

  6. Experimental and Numerical Simulation of Unbalance Response in Vertical Test Rig with Tilting-Pad Bearings

    Directory of Open Access Journals (Sweden)

    Mattias Nässelqvist

    2014-01-01

    Full Text Available In vertically oriented machines with journal bearing, there are no predefined static radial loads, such as dead weight for horizontal rotor. Most of the commercial software is designed to calculate rotordynamic and bearing properties based on machines with a horizontally oriented rotor; that is, the bearing properties are calculated at a static eccentricity. For tilting-pad bearings, there are no existing analytical expressions for bearing parameters and the bearing parameters are dependent on eccentricity and load angle. The objective of this paper is to present a simplified method to perform numerical simulations on vertical rotors including bearing parameters. Instead of recalculating the bearing parameters in each time step polynomials are used to represent the bearing parameters for present eccentricities and load angles. Numerical results are compared with results from tests performed in a test rig. The test rig consists of two guide bearings and a midspan rotor. The guide bearings are 4-pad tilting-pad bearings. Shaft displacement and strains in the bearing bracket are measured to determine the test rig’s properties. The comparison between measurements and simulated results shows small deviations in absolute displacement and load levels, which can be expected due to difficulties in calculating exact bearing parameters.

  7. Small-Scale Testing Rig for Long-Term Cyclically Loaded Monopiles in Cohesionless Soil

    DEFF Research Database (Denmark)

    Roesen, Hanne Ravn; Ibsen, Lars Bo; Andersen, Lars Vabbersgaard

    2012-01-01

    , and the period of the cyclic loading. However, the design guidance on these issues is limited. Thus, in order to investigate the pile behaviour for cyclically long-term loaded monopiles, a test setup for small-scale tests in saturated dense cohesionless soil is constructed and presented in here. The cyclic...... loading is applied mechanically by means of a testing rig, where the important input parameters: mean level, amplitude, number of cycles, and period of the loading can be varied. The results from a monotonic and a cyclic loading test on an open-ended aluminium pile with diameter = 100 mm and embedded...... length = 600 mm proves that the test setup is capable of applying the cyclic long-term loading. The plastic deformations during loading depend not only on the loading applied but also of the relative density of the soil and, thus, the tests are carried out with relative densities of 77-88%, i.e. similar...

  8. Ultrasonic inspection for testing the PWR fuel rod endplug welds

    International Nuclear Information System (INIS)

    Pillet, C.; Destribats, M.T.; Papezyk, F.

    1976-01-01

    A method of ultrasonic testing with local immersion and transversal waves was developed. It is possible to detect defects as the lacks of fusion and penetration and porosity in the PWR fuel rod endplug welds [fr

  9. PHEBUS/test-218, Behaviour of a Fuel Rod Bundle during a Large Break LOCA Transient with a two Peaks Temperature History

    International Nuclear Information System (INIS)

    1987-01-01

    1 - Description of test facility: PHEBUS test facility operated at CEA Research Center Cadarache consists of a pressurized circuit involving pumps, heat exchangers and a blowdown tank - 25 nuclear fuel rod bundle, coupled to a separate driver core; - active length 0.8 m, cosine axial power profile; - pressurized and un-pressurized fuel rods; - controlled cooling conditions at the bundle inlet (blowdown, refill and reflood period); - de-pressurized test rig volume 0.22 m 3 . The following 'as measured' boundary conditions (B.C.) were offered to participants as options with decreasing challenge to their analytical approach: Boundary conditions B.C.0: - full thermal-hydraulic analysis of PHEBUS test rig (was not recommended). Boundary conditions B.C.1: - thermal power level of fuel bundle; - fluid inlet conditions to bundle section. Boundary conditions B.C.2: - local cladding temperatures of rods; - heat transfer coefficients. Boundary conditions B.C.3: - cladding temperatures of rods; - internal pressure of rods. 2 - Description of test: Post-test investigation into the response of a nuclear fuel bundle to a large break loss of coolant accident with respect to - local fuel temperatures, - cladding strain at the time of burst, - time to burst and under given thermal-hydraulic boundary conditions of PHEBUS-test 218

  10. Development Of Test Rig System For Calibration Of Temperature Sensing Fabric

    Directory of Open Access Journals (Sweden)

    Husain Muhammad Dawood

    2017-09-01

    Full Text Available A test rig is described, for the measurement of temperature and resistance parameters of a Temperature Sensing Fabric (TSF for calibration purpose. The equipment incorporated a temperature-controlled hotplate, two copper plates, eight thermocouples, a temperature data-logger and a four-wire high-resolution resistance measuring multimeter. The copper plates were positioned above and below the TSF and in physical contact with its surfaces, so that a uniform thermal environment might be provided. The temperature of TSF was estimated by the measurement of temperature profiles of the two copper plates. Temperature-resistance graphs were created for all the tests, which were carried out over the range of 20 to 50°C, and they showed that the temperature and resistance values were not only repeatable but also reproducible, with only minor variations. The comparative analysis between the temperature-resistance test data and the temperature-resistance reference profile showed that the error in estimation of temperature of the sensing element was less than ±0.2°C. It was also found that the rig not only provided a stable and homogenous thermal environment but also offered the capability of accurately measuring the temperature and resistance parameters. The Temperature Sensing Fabric is suitable for integration into garments for continuous measurement of human body temperature in clinical and non-clinical settings.

  11. Fretting wear of steam generator tubes: high-temperature tests on AECL rig

    International Nuclear Information System (INIS)

    Guerout, F.; Zbinden, M.

    1993-07-01

    The R and DD has undertaken the study of fretting-wear of Alloy 600 S.G. tubes which occurs by contact with migrating items. The test series was performed in Canada at AECL Research (Atomic Energy of Canada Limited) as part of an exchange program. Four types of configuration were envisaged: a tube-to-drilled hole support contact which provides reference results and three types of tube-to-support contacts which simulate the tube fretting-wear induced by a welding rod, a threaded rod and a knife-edge rod support. This programme is completed by the study of the contact between a S.G. tube and a neighbouring S.G. tube which has been broken after plugging. (authors). 1 tab., 3 refs

  12. Sturdy on Orbital TIG Welding Properties for Nuclear Fuel Test Rod

    Energy Technology Data Exchange (ETDEWEB)

    Joung, Changyoung; Hong, Jintae; Kim, Kahye; Huh, Sungho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    We developed a precision TIG welding system that is able to weld the seam between end-caps and a fuel cladding tube for the nuclear fuel test rod and rig. This system can be mainly classified into an orbital TIG welder (AMI, M-207A) and a pressure chamber. The orbital TIG welder can be independently used, and it consists of a power supply unit, a microprocessor, water cooling unit, a gas supply unit and an orbital weld head. In this welder, the power supply unit mainly supplies GTAW power for a welding specimen and controls an arc starting of high frequency, supping of purge gas, arc rotation through the orbital TIG welding head, and automatic timing functions. In addition, the pressure chamber is used to make the welded surface of the cladding specimen clean with the inert gas filled inside the chamber. To precisely weld the cladding tube, a welding process needs to establish a schedule program for an orbital TIG welding. Therefore, the weld tests were performed on a cladding tube and dummy rods under various conditions. This paper describes not only test results on parameters of the purge gas flow rates and the chamber gas pressures for the orbital TIG welding, but also test results on the program establishment of an orbital TIG welding system to weld the fuel test rods. Various welding tests were performed to develop the orbital TIG welding techniques for the nuclear fuel test rod. The width of HAZ of a cladding specimen welded with the identical power during an orbital TIG welding cycle was continuously increased from a welded start-point to a weld end-point because of heat accumulation. The welding effect of the PGFR and CGP shows a relatively large difference for FSS and LSS. Each hole on the cladding specimens was formed in the 1bar CGP with the 20L/min PGFR but not made in the case of the PGFR of 10L/min in the CGP of 2bar. The optimum schedule program of the orbital TIG welding system to weld the nuclear fuel test rod was established through the program

  13. Sturdy on Orbital TIG Welding Properties for Nuclear Fuel Test Rod

    International Nuclear Information System (INIS)

    Joung, Changyoung; Hong, Jintae; Kim, Kahye; Huh, Sungho

    2014-01-01

    We developed a precision TIG welding system that is able to weld the seam between end-caps and a fuel cladding tube for the nuclear fuel test rod and rig. This system can be mainly classified into an orbital TIG welder (AMI, M-207A) and a pressure chamber. The orbital TIG welder can be independently used, and it consists of a power supply unit, a microprocessor, water cooling unit, a gas supply unit and an orbital weld head. In this welder, the power supply unit mainly supplies GTAW power for a welding specimen and controls an arc starting of high frequency, supping of purge gas, arc rotation through the orbital TIG welding head, and automatic timing functions. In addition, the pressure chamber is used to make the welded surface of the cladding specimen clean with the inert gas filled inside the chamber. To precisely weld the cladding tube, a welding process needs to establish a schedule program for an orbital TIG welding. Therefore, the weld tests were performed on a cladding tube and dummy rods under various conditions. This paper describes not only test results on parameters of the purge gas flow rates and the chamber gas pressures for the orbital TIG welding, but also test results on the program establishment of an orbital TIG welding system to weld the fuel test rods. Various welding tests were performed to develop the orbital TIG welding techniques for the nuclear fuel test rod. The width of HAZ of a cladding specimen welded with the identical power during an orbital TIG welding cycle was continuously increased from a welded start-point to a weld end-point because of heat accumulation. The welding effect of the PGFR and CGP shows a relatively large difference for FSS and LSS. Each hole on the cladding specimens was formed in the 1bar CGP with the 20L/min PGFR but not made in the case of the PGFR of 10L/min in the CGP of 2bar. The optimum schedule program of the orbital TIG welding system to weld the nuclear fuel test rod was established through the program

  14. Test Rig for Evaluating Active Turbine Blade Tip Clearance Control Concepts

    Science.gov (United States)

    Lattime, Scott B.; Steinetz, Bruce M.; Robbie, Malcolm G.

    2003-01-01

    Improved blade tip sealing in the high pressure compressor and high pressure turbine can provide dramatic improvements in specific fuel consumption, time-on-wing, compressor stall margin and engine efficiency as well as increased payload and mission range capabilities of both military and commercial gas turbine engines. The preliminary design of a mechanically actuated active clearance control (ACC) system for turbine blade tip clearance management is presented along with the design of a bench top test rig in which the system is to be evaluated. The ACC system utilizes mechanically actuated seal carrier segments and clearance measurement feedback to provide fast and precise active clearance control throughout engine operation. The purpose of this active clearance control system is to improve upon current case cooling methods. These systems have relatively slow response and do not use clearance measurement, thereby forcing cold build clearances to set the minimum clearances at extreme operating conditions (e.g., takeoff, re-burst) and not allowing cruise clearances to be minimized due to the possibility of throttle transients (e.g., step change in altitude). The active turbine blade tip clearance control system design presented herein will be evaluated to ensure that proper response and positional accuracy is achievable under simulated high-pressure turbine conditions. The test rig will simulate proper seal carrier pressure and temperature loading as well as the magnitudes and rates of blade tip clearance changes of an actual gas turbine engine. The results of these evaluations will be presented in future works.

  15. PBF/LOFT Lead Rod Test Program experiment predictions document

    International Nuclear Information System (INIS)

    Varacalle, D.J.; Cox, W.R.; Niebruegge, D.A.; Seiber, S.J.; Brake, T.E.; Driskell, W.E.; Nigg, D.W.; Tolman, E.L.

    1978-12-01

    The PBF/LOFT Lead Rod (LLR) Test Program is being conducted to provide experimental information on the behavior of nuclear fuel under normal and accident conditions in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory. The PBF/LLR tests are designed to simulate the test conditions for the LOFT Power Ascension Tests L2-3 through L2-5. The test program has been designed to provide a parametric evaluation of the LOFT fuel (center and peripheral modules) over a wide range of power. This report presents the experiment predictions for the three four-rod LOCA tests

  16. Ballistic and Cyclic Rig Testing of Braided Composite Fan Case Structures

    Science.gov (United States)

    Watson, William R.; Roberts, Gary D.; Pereira, J. Michael; Braley, Michael S.

    2015-01-01

    FAA fan blade-out certification testing on turbofan engines occurs very late in an engine's development program and is very costly. It is of utmost importance to approach the FAA Certification engine test with a high degree of confidence that the containment structure will not only contain the high-energy debris, but that it will also withstand the cyclic loads that occur with engine spooldown and continued rotation as the non-running engine maintains a low rotor RPM due to forced airflow as the engine-out aircraft returns to an airport. Accurate rig testing is needed for predicting and understanding material behavior of the fan case structure during all phases of this fan blade-out event.

  17. Optimization of inverse model identification for multi-axial test rig control

    Directory of Open Access Journals (Sweden)

    Müller Tino

    2016-01-01

    Full Text Available Laboratory testing of multi-axial fatigue situations improves repeatability and allows a time condensing of tests which can be carried out until component failure, compared to field testing. To achieve realistic and convincing durability results, precise load data reconstruction is necessary. Cross-talk and a high number of degrees of freedom negatively affect the control accuracy. Therefore a multiple input/multiple output (MIMO model of the system, capturing all inherent cross-couplings is identified. In a first step the model order is estimated based on the physical fundamentals of a one channel hydraulic-servo system. Subsequently, the structure of the MIMO model is optimized using correlation of the outputs, to increase control stability and reduce complexity of the parameter optimization. The identification process is successfully applied to the iterative control of a multi-axial suspension rig. The results show accurate control, with increased stability compared to control without structure optimization.

  18. NASA GRC's High Pressure Burner Rig Facility and Materials Test Capabilities

    Science.gov (United States)

    Robinson, R. Craig

    1999-01-01

    The High Pressure Burner Rig (HPBR) at NASA Glenn Research Center is a high-velocity. pressurized combustion test rig used for high-temperature environmental durability studies of advanced materials and components. The facility burns jet fuel and air in controlled ratios, simulating combustion gas chemistries and temperatures that are realistic to those in gas turbine engines. In addition, the test section is capable of simulating the pressures and gas velocities representative of today's aircraft. The HPBR provides a relatively inexpensive. yet sophisticated means for researchers to study the high-temperature oxidation of advanced materials. The facility has the unique capability of operating under both fuel-lean and fuel-rich gas mixtures. using a fume incinerator to eliminate any harmful byproduct emissions (CO, H2S) of rich-burn operation. Test samples are easily accessible for ongoing inspection and documentation of weight change, thickness, cracking, and other metrics. Temperature measurement is available in the form of both thermocouples and optical pyrometery. and the facility is equipped with quartz windows for observation and video taping. Operating conditions include: (1) 1.0 kg/sec (2.0 lbm/sec) combustion and secondary cooling airflow capability: (2) Equivalence ratios of 0.5- 1.0 (lean) to 1.5-2.0 (rich), with typically 10% H2O vapor pressure: (3) Gas temperatures ranging 700-1650 C (1300-3000 F): (4) Test pressures ranging 4-12 atmospheres: (5) Gas flow velocities ranging 10-30 m/s (50-100) ft/sec.: and (6) Cyclic and steady-state exposure capabilities. The facility has historically been used to test coupon-size materials. including metals and ceramics. However complex-shaped components have also been tested including cylinders, airfoils, and film-cooled end walls. The facility has also been used to develop thin-film temperature measurement sensors.

  19. Vibration features of an 180 kW maglev circulator test rig

    International Nuclear Information System (INIS)

    Su Jiageng; Li Hongwei; Shi Qian; Sha Honglei; Yu Suyuan

    2015-01-01

    The helium circulator is the key equipment to drive the helium gas flowing in the primary loop for energy exchange in HTGR. Active magnetic bearings (AMB) have been considered as an alternative to replace traditional mechanical bearings in the helium circulator. Such contactless bearings do not have frictional wear and can be used to suppress vibration in rotor-dynamic applications. It is necessary to study the vibration characteristics of the maglev helium circulator to guarantee the reactor safety. Therefore, a maglev circulator test rig was built. The power of the circulator is 180 kW and the maximum speed is 17000 rpm. For the time being, the test atmosphere is air. In this paper the test rig was introduced. Vibration test work of the maglev circulator was also carried out. The measuring points were arranged at the seat because the seat vibration level is important to evaluate the machine noise. The measuring points were also arranged at the base of the circulator housing to better study the vibration characteristics. The vibrations were measured by the LC-8024 multichannel machinery diagnoses system. At each measuring point the vibrations were detected in three directions (X, Y and Z) with the vibration acceleration sensors. The test speeds varied from 1000 rpm to 17000 rpm with an increase of 1000 rpm each time. The vibration values of the seat are from 89.5 dB at 1000 rpm to 113.3 dB at 17000 rpm. The test results showed that the maglev circulator exhibits good vibration properties. This work will offer important theoretical base and engineering experience to explore the high-speed helium circulator in HTGR. (author)

  20. Modern challenges for flow investigations in model hydraulic turbines on classical test rig

    International Nuclear Information System (INIS)

    Deschênes, C; Houde, S; Aeschlimann, V; Fraser, R; Ciocan, G D

    2014-01-01

    The BulbT project involved several investigations of flow phenomena in different parts of a model bulb turbine installed on the test rig of Laval University Laboratory. The aim is to create a comprehensive data base in order to increase the knowledge of the flow phenomena in this type of turbines and to validate or improve numerical flow simulation strategies. This validation being based on a kinematic comparison between experimental and numerical data, the project had to overcome challenges to facilitate the use of the experimental data for that purpose. Many parameters were checked, such as the test bench repeatability, the intrusiveness of a priori non-intrusive methods, the geometry of the runner and draft tube. This paper illustrates how some of those problematic were solved

  1. The heater system monitoring and control of the fuelling machines test rig fluid

    International Nuclear Information System (INIS)

    Iorga, C.; Iorga, H.

    2016-01-01

    The thermo-mechanical hot loop (HL) of the testing rig for the fuelling machines (F/Ms) represents a set of facilities and equipment that perform the pressure, temperature and flow thermo-hydraulic parameters similar to those from the fuel channel for CANDU 600 reactor types. The 2.1 MW electric heater (EH), part of the HL, working under the conditions of a pressure vessel (110 bars) and provides an average temperature of 300°C of the working agent. The monitoring equipment implemented aims to simultaneously control the temperature for each of the 12 modules that compose the EH, without influencing the work logic of the display/recording and protecting existing equipment. This paper presents the structure of the monitoring equipment and its performance obtained after performing the functional tests. (authors)

  2. {open_quote}Nintendo Rig{close_quote} lets two men do work of three on traditional servicing rig

    Energy Technology Data Exchange (ETDEWEB)

    Rintoul, B.

    1996-01-01

    New well servicing rig saves costs and increases safety by using a robot derrickman. The rigs is called the Nintendo Rig, taking the name from the joystick that controls the robot on the racking board 25 feet above the ground. An automated tong/slip package permanently mounted on the front of the rig handles pipe and rods on the ground.

  3. Numerical Application of a Stick-Slip Control and Experimental Analysis using a Test Rig

    Directory of Open Access Journals (Sweden)

    Pereira Leonardo D.

    2018-01-01

    Full Text Available Part of the process of exploration and development of an oil field consists of the drilling operations for oil and gas wells. Particularly for deep water and ultra deep water wells, the operation requires the control of a very flexible structure which is subjected to complex boundary conditions such as the nonlinear interactions between drill bit and rock formation and between the drill string and borehole wall. Concerning this complexity, the stick-slip phenomenon is a major component related to the torsional vibration and it can excite both axial and lateral vibrations. With these intentions, this paper has the main goal of confronting the torsional vibration problem over a test rig numerical model using a real-time conventional controller. The system contains two discs in which dry friction torques are applied. Therefore, the dynamical behaviour were analysed with and without controlling strategies.

  4. CFD simulation and experimental analysis of erosion in a slurry tank test rig

    Directory of Open Access Journals (Sweden)

    Bart Hans-Jörg

    2013-04-01

    Full Text Available Erosion occurring in equipment dealing with liquid-solid mixtures such as pipeline parts, slurry pumps, liquid-solid stirred reactors and slurry mixers in various industrial applications results in operational failure and economic costs. A slurry erosion tank test rig is designed and was built to investigate the erosion rates of materials and the influencing parameters such as flow velocity and turbulence, flow angle, solid particle concentration, particles size distribution, hardness and target material properties on the material loss and erosion profiles. In the present study, a computational fluid dynamics (CFD tool is used to simulate the erosion rate of sample plates in the liquid-solid slurry mixture in a cylindrical tank. The predictions were made in a steady state and also transient manner, applying the flow at the room temperature and using water and sand as liquid and solid phases, respectively. The multiple reference frame method (MRF is applied to simulate the flow behavior and liquid-solid interactions in the slurry tank test rig. The MRF method is used since it is less demanding than sliding mesh method (SM and gives satisfactory results. The computational domain is divided into three regions: a rotational or MRF zone containing the mixer, a rotational zone (MRF containing the erosion plates and a static zone (outer liquid zone. It is observed that changing the MRF zone diameter and height causes a very low impact on the results. The simulated results were obtained for two kinds of hard metals namely stainless steel and ST-50 under some various operating conditions and are found in good agreement with the experimental results.

  5. Biannular Airbreathing Nozzle Rig (BANR) facility checkout and plug nozzle performance test data

    Science.gov (United States)

    Cummings, Chase B.

    2010-09-01

    The motivation for development of a supersonic business jet (SSBJ) platform lies in its ability to create a paradigm shift in the speed and reach of commercial, private, and government travel. A full understanding of the performance capabilities of exhaust nozzle configurations intended for use in potential SSBJ propulsion systems is critical to the design of an aircraft of this type. Purdue University's newly operational Biannular Airbreathing Nozzle Rig (BANR) is a highly capable facility devoted to the testing of subscale nozzles of this type. The high accuracy, six-axis force measurement system and complementary mass flowrate measurement capabilities of the BANR facility make it rather ideally suited for exhaust nozzle performance appraisal. Detailed accounts pertaining to methods utilized in the proper checkout of these diagnostic capabilities are contained herein. Efforts to quantify uncertainties associated with critical BANR test measurements are recounted, as well. Results of a second hot-fire test campaign of a subscale Gulfstream Aerospace Corporation (GAC) axisymmetric, shrouded plug nozzle are presented. Determined test article performance parameters (nozzle thrust efficiencies and discharge coefficients) are compared to those of a previous test campaign and numerical simulations of the experimental set-up. Recently acquired data is compared to published findings pertaining to plug nozzle experiments of similar scale and operating range. Suggestions relating to the future advancement and improvement of the BANR facility are provided. Lessons learned with regards to test operations and calibration procedures are divulged in an attempt to aid future facility users, as well.

  6. A unique laboratory test rig reduces the need for offshore tests to combat calcium naphthenate deposition in oilfield process equipment.

    Energy Technology Data Exchange (ETDEWEB)

    Mediaas, Heidi; Grande, Knut; Hustad, Britt-Marie; Hoevik, Kim Reidar; Kummernes, Hege; Nergaard, Bjoern; Vindstad, Jens Emil

    2006-03-15

    Producing and refining high-TAN crude oils introduces a number of challenges, among which calcium naphthenate deposition in process facilities is the most serious production issue. Until recently, the only option for studying chemicals and process parameters in order to prevent naphthenate deposition has been field tests. Statoil has now developed a small scale pilot plant where these experiments can be performed in the laboratory at Statoil's Research and Technology Center in Trondheim, Norway. The results from the pilot plant are in full agreement with the extensive naphthenate experience obtained from almost 9 years operation of the Heidrun oilfield. The design and operational procedures for this test facility are based on the recent discovery by Statoil and ConocoPhillips of the ARN acid. The ARN acid is a prerequisite for calcium naphthenate deposition. The new continuous flow pilot plant, the Naphthenate Rig, is used to develop new environmental friendly naphthenate inhibitors and to optimize process operating conditions. Since it operates on real crudes the need for field tests in qualifying new naphthenate inhibitors is reduced. To the best of our knowledge, the rig is the first of its kind in the world. (Author)

  7. Construction and testing of a 'fork' type central rod

    International Nuclear Information System (INIS)

    Guicherd, R.; Meunier, C.

    1964-01-01

    Aim of the control rod: - Reduction of the flux peaks - Increase of the efficiency - simplification of the mechanics - without modification of the grid nor of the standard fuel plates. Results of the tests: - Reduction of the flux peaks by 29 per cent - Increase of the efficiency by 20 per cent - The suppleness of the plates and the decrease in the number of parts results in greater safety during operation. Conclusions: Study of a generalized adaptation of this type of rod which presents definite advantages. (authors) [fr

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-03-15

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

  9. Hemispheric Correlates of the Rod-And-Frame Test.

    Science.gov (United States)

    Berlin, Donna F.; Languis, Marlin L.

    1981-01-01

    Right-handed sixth graders were administered the WISC Block Design and verbal and nonverbal versions of the Rod-and-Frame Test (RFT), measuring field dependence/independence. Results seemed to reflect a right hemisphere processing for the nonverbal RFT and a possible sex bias against girls in its traditional verbal administration. (Author/SJL)

  10. Measuring deformation of Fuel pin in a Nuclear Fuel Test Rig

    Energy Technology Data Exchange (ETDEWEB)

    Heo, S. H.; Yang, T. H.; Hong, J. T.; Joung, C. Y.; Ahn, S. H.; Jang, S. Y.; Kim, J. H. [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    In this study, an LVDT core for measuring the longitudinal displacement of fuel pellets and clad was designed and produced. A signal processing method for the prepared core was investigated. The Nuclear Fuel Test Rig is used to observe changes in the characteristics of the fuel according to the neutron irradiation at HANARO (High-flux Advanced Neutron Application Reactor), which is a research reactor. Which are the strain and internal temperature of the irradiated nuclear fuel and the internal pressure of fuel due to fission gas, the characteristics of the fuel are measured using various sensors such as a thermocouple, SPND and LVDT. In this study, two shaped LVDT (Linear Variable Differential Transformer) cores for displacement measurements were designed and manufactured in order to measure the displacement of a fuel pellet and cladding tube using LVDT sensors for measuring electrical signals by converting the physical variation such as the force and displacement into a linear motion. In addition, signals from the manufactured LVDT sensor were collected and calibrated. Moreover, a method for obtaining the displacement in the core according to the sensing signal was planned. A derived equation can used to predict the change in the position of core. A following study should be conducted to test the output signal and real variation of out-pile system. For further work, a performance verification is required for an in-pile irradiation test.

  11. Modelling the nonlinear behaviour of an underplatform damper test rig for turbine applications

    Science.gov (United States)

    Pesaresi, L.; Salles, L.; Jones, A.; Green, J. S.; Schwingshackl, C. W.

    2017-02-01

    Underplatform dampers (UPD) are commonly used in aircraft engines to mitigate the risk of high-cycle fatigue failure of turbine blades. The energy dissipated at the friction contact interface of the damper reduces the vibration amplitude significantly, and the couplings of the blades can also lead to significant shifts of the resonance frequencies of the bladed disk. The highly nonlinear behaviour of bladed discs constrained by UPDs requires an advanced modelling approach to ensure that the correct damper geometry is selected during the design of the turbine, and that no unexpected resonance frequencies and amplitudes will occur in operation. Approaches based on an explicit model of the damper in combination with multi-harmonic balance solvers have emerged as a promising way to predict the nonlinear behaviour of UPDs correctly, however rigorous experimental validations are required before approaches of this type can be used with confidence. In this study, a nonlinear analysis based on an updated explicit damper model having different levels of detail is performed, and the results are evaluated against a newly-developed UPD test rig. Detailed linear finite element models are used as input for the nonlinear analysis, allowing the inclusion of damper flexibility and inertia effects. The nonlinear friction interface between the blades and the damper is described with a dense grid of 3D friction contact elements which allow accurate capturing of the underlying nonlinear mechanism that drives the global nonlinear behaviour. The introduced explicit damper model showed a great dependence on the correct contact pressure distribution. The use of an accurate, measurement based, distribution, better matched the nonlinear dynamic behaviour of the test rig. Good agreement with the measured frequency response data could only be reached when the zero harmonic term (constant term) was included in the multi-harmonic expansion of the nonlinear problem, highlighting its importance

  12. Design and commission of an experimental test rig to apply a full-scale pressure load on composite sandwich panels representative of an aircraft secondary structure

    International Nuclear Information System (INIS)

    Crump, D A; Dulieu-Barton, J M; Savage, J

    2010-01-01

    This paper describes the design of a test rig, which is used to apply a representative pressure load to a full-scale composite sandwich secondary aircraft structure. A generic panel was designed with features to represent those in the composite sandwich secondary aircraft structure. To provide full-field strain data from the panels, the test rig was designed for use with optical measurement techniques such as thermoelastic stress analysis (TSA) and digital image correlation (DIC). TSA requires a cyclic load to be applied to a structure for the measurement of the strain state; therefore, the test rig has been designed to be mounted on a standard servo-hydraulic test machine. As both TSA and DIC require an uninterrupted view of the surface of the test panel, an important consideration in the design is facilitating the optical access for the two techniques. To aid the test rig design a finite element (FE) model was produced. The model provides information on the deflections that must be accommodated by the test rig, and ensures that the stress and strain levels developed in the panel when loaded in the test rig would be sufficient for measurement using TSA and DIC. Finally, initial tests using the test rig have shown it to be capable of achieving the required pressure and maintaining a cyclic load. It was also demonstrated that both TSA and DIC data can be collected from the panels under load, which are used to validate the stress and deflection derived from the FE model

  13. Experience in handling core subassemblies in sodium cooled reactor KNK and test rigs

    International Nuclear Information System (INIS)

    Althaus; Jansing; Kesseler; Kirchner; Menck

    1974-01-01

    Compared with a water cooled reactor plant a sodium cooled reactor plant presents a number of problems which result from the specific nature of sodium. These problems that must be faced during all handling operations are mainly: 1. The rapid reaction of sodium in air requires handling to be done only under cover gas. 2. The temperature of all sodium-wetted components is to be kept above the melting point of sodium. 3. Poor draining of removed reactor components due to the high surface tension of sodium and the associated danger of dripping radioactive sodium may produce radiation or contamination problems. 4. Sodium is not transparent. The sum of these and further influences dictate that the general handling usually is carried out without visual means, though a method is under development in the USA to use ultrasonic for under sodium 'viewing'. These limitations to sodium component handling are applicable to all sodium reactor plants, several of which are discussed in this report. After the description of the handling systems of the KNK plant now operating at Karlsruhe, the experience with the SNR test rig and finally the handling systems for SNR 300 and SNR 2 are discussed

  14. Model of ASTM Flammability Test in Microgravity: Iron Rods

    Science.gov (United States)

    Steinberg, Theodore A; Stoltzfus, Joel M.; Fries, Joseph (Technical Monitor)

    2000-01-01

    There is extensive qualitative results from burning metallic materials in a NASA/ASTM flammability test system in normal gravity. However, this data was shown to be inconclusive for applications involving oxygen-enriched atmospheres under microgravity conditions by conducting tests using the 2.2-second Lewis Research Center (LeRC) Drop Tower. Data from neither type of test has been reduced to fundamental kinetic and dynamic systems parameters. This paper reports the initial model analysis for burning iron rods under microgravity conditions using data obtained at the LERC tower and modeling the burning system after ignition. Under the conditions of the test the burning mass regresses up the rod to be detached upon deceleration at the end of the drop. The model describes the burning system as a semi-batch, well-mixed reactor with product accumulation only. This model is consistent with the 2.0-second duration of the test. Transient temperature and pressure measurements are made on the chamber volume. The rod solid-liquid interface melting rate is obtained from film records. The model consists of a set of 17 non-linear, first-order differential equations which are solved using MATLAB. This analysis confirms that a first-order rate, in oxygen concentration, is consistent for the iron-oxygen kinetic reaction. An apparent activation energy of 246.8 kJ/mol is consistent for this model.

  15. Heat transfer and friction on smooth and rough test rods

    International Nuclear Information System (INIS)

    Franken, W.M.P.; Hoogland, H.; Deijman, P.

    1977-06-01

    Results are reported on heat transfer and pressure drop tests on one smooth and nine rough test rods in an annular geometry. The wall roughness consisted of transversal ribs with various roughness pitches, rib heights and rib widths. The tests were performed with air as coolant under a wide range of experimental conditions: 10 5 5 , 1.1 2. Special attention has been given to the effect of variation of the physical coolant properties over the flow cross section. This effect could be described by the power function (Tsub(w)/Tsub(b))sup(-0.3l) in additional systematic variation of the heat transfer could be recognized, dependent on the coolant temperature level. The experimental results were correlated by the equation St = C(Tsub(in)) Resup(-0.2) Prsup(-0.6) (Tsub(w)/Tsub(b)sup(-0.31). Values of C(Tsub(in)) are given in tabular form. The thermal entrance effect has been measured on various test rods. A substantial reduction of the heat transfer coefficient was almost constant along the rough test rods

  16. Analysis of rig test data for an axial/centrifugal compressor in the 12 kg/sec

    Science.gov (United States)

    Owen, A. K.

    1994-01-01

    Extensive testing was done on a T55-L-712 turboshaft engine compressor in a compressor test rig at TEXTRON/Lycoming. These rig tests will be followed by a series of engine tests to occur at the NASA Lewis Research Center beginning in the last quarter of 1993. The goals of the rig testing were: (1) map the steady state compressor operation from 20 percent to 100 percent design speed, (2) quantify the effects of compressor bleed on the operation of the compressor, and (3) explore and measure the operation of the compressor in the flow ranges 'beyond' the normal compressor stall line. Instrumentation consisted of 497 steady state pressure sensors, 153 temperature sensors and 34 high response transducers for transient analysis in the pre- and post-stall operating regime. These measurements allow for generation of detailed stage characteristics as well as overall mapping. Transient data is being analyzed for the existence of modal disturbances at the front face of the compression system ('stall precursors'). This paper presents some preliminary results of the ongoing analysis and a description of the current and future program plans. It will primarily address the unsteady events at the front face of the compression system that occur as the system transitions from steady state to unsteady (stall/surge) operation.

  17. Control rod drive mechanism stator loss of coolant test

    International Nuclear Information System (INIS)

    Besel, L.; Ibatuan, R.

    1977-04-01

    This report documents the stator loss of coolant test conducted at HEDL on the lead unit Control Rod Drive Mechanism (CRDM) in February, 1977. The purpose of the test was to demonstrate scram capability of the CRDM with an uncooled stator and to obtain a time versus temperature curve of an uncooled stator under power. Brief descriptions of the test, hardware used, and results obtained are presented in the report. The test demonstrated that the CRDM could be successfully scrammed with no anomalies in both the two-phase and three-phase stator winding hold conditions after the respective equilibrium stator temperatures had been obtained with no stator coolant

  18. PBF/LOFT Lead Rod Test Program experiment operating specification

    International Nuclear Information System (INIS)

    Varacalle, D.J. Jr.

    1978-11-01

    The PBF/LOFT Lead Rod (LLR) Test Program is being conducted to provide experimental information on the behavior of nuclear fuel under normal and accident conditions in the Power Burst Facility at the Idaho National Engineering Laboratory. Understanding the behavior of light-water reactors (LWR) under loss-of-coolant conditions is a major objective of the NRC Reactor Safety Research Program. The Loss of Fluid Test (LOFT) facility is the major testing facility to evaluate the systems response of an LWR over a wide range of Loss of Coolant Experment (LOCE) conditions. As such, the LOFT core is intended to be used for sequential LOCE tests provided no significant fuel rod failures occur. The PFB/LLR tests are designed to simulate the test conditions for the LOFT Power Ascension Tests L2-2 through L2-5. The test program has been designed to provide a parametric evaluation of the LOFT fuel over a wide range of power. Thus, a relatively accurate assessment of the state of the LOFT core after the completion of each subtest and the anticipated effect of the next test can be obtained by utilizing a combination of LLR test data and analytical predictions. Specifications for the test program are presented

  19. Remote helium leak test of the DUPIC fuel rod

    International Nuclear Information System (INIS)

    Kim, W. K; Kim, S. S.; Lim, S. P.; Lee, J. W.; Yang, M. S.

    1998-01-01

    DUPIC(Direct Use of spent PWR fuel In CANDU reactor) is one of dry reprocessing fuel cycles to reuse irradiated PWR fuel in CANDU power plant. DUPIC fuel is so radioactive that DUPIC fuel is remotely fabricated at hot cell such as IMEF hot cell in which radiation is shielded and remote operation is possible. In this study, Helium leakage has been tested for the simulated DUPIC fuel rod manufactured by Nd:YAG laser end-cap welding at simulated hot cell. The remote inspection technique has been developed to evaluate the soundness of DUPIC fuel fabricated through new processes. Vacuum chamber has been developed to be remotely operated by manipulators at hot cell. As the result of remote test, Helium leakage of DUPIC fuel rod is around background level, CANDU specification has been satisfied. In the result of the study, remote test has been successfully performed at the simulated hot cell, and the soundness of DUPIC fuel rod welded by Nd:YAG laser has been confirmed

  20. Test requirement for PIE of HANARO irradiated fuel rod

    International Nuclear Information System (INIS)

    Lim, I. C.; Cho, Y. G.

    2000-06-01

    Since the first criticality of HANARO reached in Feb. of 1995, the rod type U 3 Si-A1 fuel imported from AECL has been used. From the under-water fuel inspection which has been conducted since 1997, a ballooning-rupture type abnormality was observed in several fuel rods. In order to find the root cause of this abnormality and to find the resolution, the post irradiation examination(PIE) was proposed as the best way. In this document, the information from the under-water inspection as well as the PIE requirements are described. Based on the information in this document, a detail test plan will be developed by the project team who shall conduct the PIE

  1. Development of Test Rig for Robotization of Mining Technological Processes - Oversized Rock Breaking Process Case

    Science.gov (United States)

    Pawel, Stefaniak; Jacek, Wodecki; Jakubiak, Janusz; Zimroz, Radoslaw

    2017-12-01

    Production chain (PCh) in underground copper ore mine consists of several subprocesses. From our perspective implementation of so called ZEPA approach (Zero Entry Production Area) might be very interesting [16]. In practice, it leads to automation/robotization of subprocesses in production area. In this paper was investigated a specific part of PCh i.e. a place when cyclic transport by LHDs is replaced with continuous transport by conveying system. Such place is called dumping point. The objective of dumping points with screen is primary classification of the material (into coarse and fine material) and breaking oversized rocks with hydraulic hammer. Current challenges for the underground mining include e.g. safety improvement as well as production optimization related to bottlenecks, stoppages and operational efficiency of the machines. As a first step, remote control of the hydraulic hammer has been introduced, which not only transferred the operator to safe workplace, but also allowed for more comfortable work environment and control over multiple technical objects by a single person. Today literature analysis shows that current mining industry around the world is oriented to automation and robotization of mining processes and reveals technological readiness for 4th industrial revolution. The paper is focused on preliminary analysis of possibilities for the use of the robotic system to rock-breaking process. Prototype test rig has been proposed and experimental works have been carried out. Automatic algorithms for detection of oversized rocks, crushing them as well as sweeping and loosening of material have been formulated. Obviously many simplifications have been assumed. Some near future works have been proposed.

  2. Study of the Parametric Performance of Solid Particle Erosion Wear under the Slurry Pot Test Rig

    Directory of Open Access Journals (Sweden)

    S.R. More

    2017-12-01

    Full Text Available Stainless Steel (SS 304 is commonly used material for slurry handling applications like pipelines, valves, pumps and other equipment's. Slurry erosion wear is a common problem in many engineering applications like process industry, thermal and hydraulic power plants and slurry handling equipments. In this paper, experimental investigation of the influence of solid particle size, impact velocity, impact angle and solid concentration parameters in slurry erosion wear behavior of SS 304 using slurry pot test rig. In this study the design of experiments was considered using Taguchi technique. A comparison has been made for the experimental and Taguchi technique results. The erosion wear morphology was studied using micro-graph obtained by scanning electron microscope (SEM analysis. At shallow impact angle 30°, the material removal pattern was observed in the form of micro displacing, scratching and ploughing with plastic deformation of the material. At 60° impact angle, mixed type of micro indentations and pitting action is observed. At normal impact angle 90°, the material removal pattern was observed in form of indentation and rounded lips. It is found that particle velocity was the most influence factor than impact angle, size and solid concentration. From this investigation, it can be concluded that the slurry erosion wear is minimized by controlling the slurry flow velocity which improves the service life of the slurry handling equipments. From the comparison of experimental and Taguchi experimental design results it is found that the percentage deviation was very small with a higher correlation coefficient (r2 0.987 which is agreeable.

  3. Materials and boiler rig testing to support chemical cleaning of once-through AGR boilers

    International Nuclear Information System (INIS)

    Tice, D.R.; Platts, N.; Raffel, A.S.; Rudge, A.

    2002-01-01

    chosen. This paper describes the results of the latter two activities: additional materials testing necessary to support a full scale plant clean and the trial clean on a model boiler rig replicating a single boiler tube together with the pre- and post-clean thermohydraulic behaviour. (authors)

  4. Rig supervisors

    International Nuclear Information System (INIS)

    Nordt, D.P.; Stone, M.S.

    1992-01-01

    This paper helps prepare the inexperienced rig supervisor to manage a drilling operation. It outlines operational-knowledge requirements and optimization concepts for improving drilling performance and lowering drilling costs. It gives guidelines on safety and environmental responsibilities, and provides recommendations on work tools, leadership, and communication

  5. Joint test rig for testing and calibrating of different methods of two-phase mass flow measurement

    International Nuclear Information System (INIS)

    Reimann, J.; Arnold, G.; Chung, M.; Hahn, H.; John, H.; Mueller, S.; Wanner, E.

    1977-01-01

    The start-up of the steady-state steam-water loop has been finished. The planned maximal values of the mass flow rate as function of quality and pressure are reached. The components for the steady-state air-water loop have been ordered, the loop has been built up, first function tests have been carried out. Because of the additional work of the extension for air-water flows, the blowdown test rig was delayed. Calculations for the security of the pressure vessel have begun. During the experiments the knowledge of the flow regime and the apparent density is essential. To detect flow regime, impedance probes were developed and have been tested in steam-water flows at pressures up to 150 at. The probe signals can be adjointed to flow patterns even in those cases when high speed movies could not be interpreted definitely. To measure the apparent density a multiple γ-beam densitometer is developed. The collimator block and the mounting support for the γ-source were manufactured, the shielding and cooling of the scintillator has begun. (orig./RW) [de

  6. Design review and analysis for a Pratt and Whitney fluid-film bearing and seal testing rig

    Science.gov (United States)

    Childs, Dara W.

    1994-01-01

    A design review has been completed for a Pratt and Whitney (P&W)-designed fluid-film bearing and annular-seal test rig to be manufactured and installed at George C. Marshall Space Flight Center (MSFC). Issues covered in this study include: (1) the capacity requirements of the drive unit; (2) the capacity and configuration of the static loading system; (3) the capacity and configuration of the dynamic excitation system; (4) the capacity, configuration, and rotordynamic stability of a test bearing, support bearings, and shaft; and (5) the characteristics and configuration of the measurement transducers and data channels.

  7. Ambient Pressure Test Rig Developed for Testing Oil-Free Bearings in Alternate Gases and Variable Pressures

    Science.gov (United States)

    Bauman, Steven W.

    1990-01-01

    The Oil-Free Turbomachinery research team at the NASA Glenn Research Center is conducting research to develop turbomachinery systems that utilize high-speed, high temperature foil (air) bearings that do not require an oil lubrication system. Such systems combine the most advanced foil bearings from industry with NASA-developed hightemperature solid-lubricant technology. New applications are being pursued, such as Oil- Free turbochargers, auxiliary power units, and turbine propulsion systems for aircraft. An Oil-Free business jet engine, for example, would be simpler, lighter, more reliable, and less costly to purchase and maintain than current engines. Another application is NASA's Prometheus mission, where gas bearings will be required for the closed-cycle turbine based power-conversion system of a nuclear power generator for deep space. To support these applications, Glenn's Oil-Free Turbomachinery research team developed the Ambient Pressure Test Rig. Using this facility, researchers can load and heat a bearing and evaluate its performance with reduced air pressure to simulate high altitude conditions. For the nuclear application, the test chamber can be purged with gases such as helium to study foil gas bearing operation in working fluids other than air.

  8. Status of prototype rupture disc testing in the large leak test rig

    International Nuclear Information System (INIS)

    Amos, J.C.

    1979-09-01

    The prototype CRBRP double membrane rupture disc assembly is being performance tested in conjunction with the LLTR Series II Large Leak Program. In May 1979, the double membrane disc assembly was inadvertently activated during sodium system pressure instrument calibration. This experience indicated that the rupture disc burst at essentially the design burst pressure when a gradually increasing state pressure was applied. The area of membrane opening was found to be about 25 to 30% of the cross-sectional area. In July 1979, the disc assembly was again tested (this time in a single membrane configuration) in conjunction with the first LLTR Series II Test A-1a (inert gas injection). Test data indicated that the disc burst in about 35 ms at essentially the design burst pressure with an opening of about 30% of the cross-sectional area. The pressure immediately downstream of the disc dropped below atmospheric pressure following the rupture tube event (releasing high pressure nitrogen into sodium) for about 1.5 seconds before increasing to a maximum of 30 psig. This behavior raises a question on the adequacy of a downstream pressure device for rapid sensing of disc rupture and initiating plant shutdown following a large SWR event. 14 figures

  9. Power ramp tests of high burnup BWR segment rods

    International Nuclear Information System (INIS)

    Hayashi, H.; Etoh, Y.; Tsukuda, Y.; Shimada, S.; Sakurai, H.

    2002-01-01

    Lead use assemblies (LUAs) of high burnup 8x8 fuel design for Japanese BWRs were irradiated up to 5 cycles in Fukushima Daini Nuclear Power Station No. 2 Unit. Segment rods were installed in LUAs and used for power ramp tests in Japanese Material Test Reactor (JMTR). Post irradiation examinations (PIEs) of segment rods were carried out at Nippon Nuclear Fuel Development Co., Ltd. before and after ramp tests. Maximum linear heat rates of LUAs were kept above 300 W/cm in the first cycle, above 250 W/cm in the second and third cycles and decreased to 200 W/cm in the fourth cycle and 80 W/cm in the fifth cycle. The integrity of high burnup 8x8 fuel was confirmed up to the bundle burnup of 48 GWd/t after 5 cycles of irradiation. Systematic and high quality data were collected through detailed PIEs. The main results are as follows. The oxide on the outer surface of cladding tubes was uniform and its thickness was less than 20 micro-meter after 5 cycles of irradiation and was almost independent of burnup. Hydrogen contents in cladding tubes were less than 150 ppm after 5 cycles of irradiation, although hydrogen contents increased during the fourth and fifth irradiation cycles. Mechanical properties of cladding tubes were on the extrapolated line of previous data up to 5 cycles of irradiation. Fission gas release rates were in the low level (mainly less than 6%) up to 5 cycles of irradiation due to the design to decrease pellet temperature. Pellet-cladding bonding layers were observed after the third cycle and almost full bonding was observed after the fifth cycle. Pellet volume increased with burnup in proportion to solid swelling rate up to the forth cycle. After the fifth cycle, slightly higher pellet swelling was confirmed. Power ramp tests were carried out and satisfactory performance of Zr-lined cladding tube was confirmed up to 60 GWd/t (segment average burnup). One segment rod irradiated for 3 cycles failed by a single step ramp test at terminal ramp power of 614 W

  10. Method for determining detailed rod worth profiles at low power in the fast test reactor

    International Nuclear Information System (INIS)

    Sevenich, R.A.

    1975-08-01

    A method for obtaining a detailed rod worth profile at low power for a slow control rod insertion is presented. The accuracy of the method depends on a preparatory experiment in which the test rod is dropped quickly to yield, upon analysis, the magnitude of the rod worth and an effective source value. These numbers are employed to initialize the inverse kinetics analysis for the slow insertion. Corrections for changes in detection efficiency are not included for the simulated experiments. (U.S.)

  11. Seismic tests in sodium of the SPX-1 primary pump shaft carried out in the CPV-1 test rig at ENEA-Brasimone

    International Nuclear Information System (INIS)

    Contardi, T.; Rapezzi, L.; Le Coz, P.; Tigeot, Y.; Partiti, C.; Zola, M.; Denimal, P.

    1988-01-01

    Dynamic tests were carried out by ISMES, on behalf of ENEA and CEA and in co-operation with FIAT/TTG, on a SPX-1 primary pump shaft. These tests were conducted, mainly in sodium, in the CPV-1 test rig at the ENEA Brasimone Center. The excitation was applied to the flange supporting the hydrostatic bearing. After some preliminary analysis performed in absence of liquid sodium and at ambient temperature, the following tests were performed on the rig filled with sodium at operating temperature: (A) sine sweeps between 1 and 15 Hz, (B) ambient vibration investigation, and (C) seismic tests with a SSE acceleration time-history (20 s duration) calculated by CEA at hydrostatic bearing level. Two sets of seismic tests were carried out, each time increasing amplitudes up to 70% of SSE. This value was not exceeded for safety reasons and actuator power limit. The first set of tests began in nominal operating conditions; when 70% of SSE was reached, pressure feed to hydrostatic bearing was reduced lowering its effective support. This simulated a larger earthquake. The second set of tests was representative of SPX-1 pump actual operating conditions, because both hydrostatic bearing pressure and shaft rotating speed were simultaneously reduced following the primary pump characteristic curve. The tests allowed the SPX-1 pump rotating set to be widely qualified. Among the main results, it is worth noting that the stiffness of the hydrostatic bearing system was generally compatible with seismic requirements. Finally, it is worth pointing out that, in order to allow the above-mentioned tests to be carried out, a full seismic qualification of the CPV-1 test rig was necessary: thus, this rig might be used in the future for further seismic tests on LMFBR components and systems in sodium. (author). Figs and tabs

  12. Testing remotely operated module technique for Wackersdorf reprocessing plant at Lahde test rig

    International Nuclear Information System (INIS)

    Leister, P.; Schroeder, G.; Boehme, G.

    1986-01-01

    The FEMO technique represents a plant concept which makes it possible to carry out the repair of high and medium activity wet chemical stages of the process by remote handling without direct access by staff. For this purpose, the apparatus of this step of the process is arranged modularly in large cells, so that movable large handling devices such as cranes and manipulator systems can replace process components subject to wear via the process modules. The machine room of the former coal-fired power station Heyden I at Lahde was, after removal of the turbines and generators, converted to a hall in which the following test areas were accommodated: FEMO cell section with 10 positions for module, cell wall mock-up and wall penetration, module mounting area, module measuring position, workplace for service area, training position, welding position and FEMO control position. (orig./HP) [de

  13. Sodium test of the Super-Phenix full size primary pump shaft on the CPV-1 test rig at ENEA-Brasimone

    International Nuclear Information System (INIS)

    Contardi, T.; Rapezzi, L.; Partiti, C.; Zola, M.; Denimal, P.

    1984-01-01

    Tests on FBR Superphenix primary pump shaft were performed within the sodium-cooled FBR common research and development programs provided for by the cooperation agreement between ENEA and CEA. These tests were performed in CPV-1 plant ENEA - Brasimone Energy Research Center. The CPV-1 rig was built by FIAT-TTG and reproduces the reactor operating conditions (sodium-temperature and level, shaft inclination, etc..). Furthermore, CPV-1 rig's most interesting feature is its possibility to apply seismic stresses to test section by means of an oleodynamic actuator. Pivoterie-1 test section was made by JEUMONT-SCHNEIDER which built Superphenix pumps too; it was given to ENEA by FIAT-TTG. Seismic tests were performed with the cooperation of ISMES and FIAT-TTG. (author)

  14. The experimental development and performance test of the pneumatic control-rod drive for the THTR

    International Nuclear Information System (INIS)

    Lange, G.; Boehlo, D.; Heim, H.; Kleine-Tebbe, A.

    1976-01-01

    Reactor control and shutdown of the THTR is accomplished by two independent systems, the first consisting of 36 absorber rods penetrating the graphite reflector region surrounding the core, the second consisting of 42 absorber rods that insert directly into the pebble bed core. This paper describes the design development and testing of the pneumatic rod drives used for movement of the 42 core control rods. The core control rods have two functions: the first, for reactor safety purposes, provides for adequate safe shutdown of the reactor under cold conditions; the second, for operational purposes, provides for compensation of slow changes in reactivity. The safety and operational functions for each absorber rod are respectively carried out by a long-stroke-piston pneumatic drive and by a stepping-piston pneumatic drive, both of these independent, helium-driven drives being incorporated in the rod drive unit for each control rod. To study the performance of the rod drive, a complete prototype control rod and rod drive unit was built and tested under simulated reactor operational conditions. Operational experience under helium temperatures and pressures was gained and the drives were tested under stress and simulated accident conditions. The reliability of this system has been demonstrated to licensing authorities and to the customer. The programme will be completed with the commissioning tests of drives for the THTR-300 reactor. (author)

  15. Drilling rig

    Energy Technology Data Exchange (ETDEWEB)

    Galiopa, A A; Yegorov, E K

    1981-01-04

    A drilling rig is proposed which contains a tower, lifter in the form of n infinite chain, and mobile rotator with holding device connected to the chain, and pipe holder. In order to accelerate the auxiliary operations to move the drilling string and unloaded rotator, the rotator is equipped with a clamp with means for transverse connection of it to both branches of the chain, while the pipe holders equipped with a clamp with means of connecting it to one of the branches of the chain.

  16. Test rig with active damping control for the simultaneous evaluation of vibration control and energy harvesting via piezoelectric transducers

    International Nuclear Information System (INIS)

    Perfetto, S; Rohlfing, J; Infante, F; Mayer, D; Herold, S

    2016-01-01

    Piezoelectric transducers can be used to harvest electrical energy from structural vibrations in order to power continuously operating condition monitoring systems local to where they operate. However, excessive vibrations can compromise the safe operation of mechanical systems. Therefore, absorbers are commonly used to control vibrations. With an integrated device, the mechanical energy that otherwise would be dissipated can be converted via piezoelectric transducers. Vibration absorbers are designed to have high damping factors. Hence, the integration of transducers would lead to a low energy conversion. Efficient energy harvesters usually have low damping capabilities; therefore, they are not effective for vibration suppression. Thus, the design of an integrated device needs to consider the two conflicting requirements on the damping. This study focuses on the development of a laboratory test rig with a host structure and a vibration absorber with tunable damping via an active relative velocity feedback. A voice coil actuator is used for this purpose. To overcome the passive damping effects of the back electromagnetic force a novel voltage feedback control is proposed, which has been validated both in simulation and experimentally. The aim of this study is to have a test rig ready for the introduction of piezo-transducers and available for future experimental evaluations of the damping effect on the effectiveness of vibration reduction and energy harvesting efficiency. (paper)

  17. Test rig with active damping control for the simultaneous evaluation of vibration control and energy harvesting via piezoelectric transducers

    Science.gov (United States)

    Perfetto, S.; Rohlfing, J.; Infante, F.; Mayer, D.; Herold, S.

    2016-09-01

    Piezoelectric transducers can be used to harvest electrical energy from structural vibrations in order to power continuously operating condition monitoring systems local to where they operate. However, excessive vibrations can compromise the safe operation of mechanical systems. Therefore, absorbers are commonly used to control vibrations. With an integrated device, the mechanical energy that otherwise would be dissipated can be converted via piezoelectric transducers. Vibration absorbers are designed to have high damping factors. Hence, the integration of transducers would lead to a low energy conversion. Efficient energy harvesters usually have low damping capabilities; therefore, they are not effective for vibration suppression. Thus, the design of an integrated device needs to consider the two conflicting requirements on the damping. This study focuses on the development of a laboratory test rig with a host structure and a vibration absorber with tunable damping via an active relative velocity feedback. A voice coil actuator is used for this purpose. To overcome the passive damping effects of the back electromagnetic force a novel voltage feedback control is proposed, which has been validated both in simulation and experimentally. The aim of this study is to have a test rig ready for the introduction of piezo-transducers and available for future experimental evaluations of the damping effect on the effectiveness of vibration reduction and energy harvesting efficiency.

  18. Fabrication of the instrumented fuel rods for the 3-Pin Fuel Test Loop at HANARO

    International Nuclear Information System (INIS)

    Sohn, Jae Min; Park, Sung Jae; Shin, Yoon Tag; Lee, Jong Min; Ahn, Sung Ho; Kim, Soo Sung; Kim, Bong Goo; Kim, Young Ki; Lee, Ki Hong; Kim, Kwan Hyun

    2008-09-01

    The 3-Pin Fuel Test Loop(hereinafter referred to as the '3-Pin FTL') facility has been installed at HANARO(High-flux Advanced Neutron Application Reactor) and the 3-Pin FTL is under a test operation. The purpose of this report is to fabricate the instrumented fuel rods for the 3-Pin FTL. The fabrication of these fuel rods was based on experiences and technologies of the instrumented fuel rods for an irradiation fuel capsule. The three instrumented fuel rods of the 3-Pin FTL have been designed. The one fuel rod(180 .deg. ) was designed to measure the centerline temperature of the nuclear fuels and the internal pressure of the fuel rod, and others(60 .deg. and 300 .deg. ) were designed to measure the centerline temperature of the fuel pellets. The claddings were made of the reference material 1 and 2 and new material 1 and 2. And nuclear fuel was used UO 2 (2.0w/o) pellet type with large grain and standard grain. The major procedures of fabrication are followings: (1) the assembling and weld of fuel rods with the pellet mockups and the sensor mockups for the qualification tests, (2) the qualification tests(dimension measurements, tensile tests, metallography examinations and helium leak tests) of weld, (3) the assembling and weld of instrumented fuel rods with the nuclear pellets and the sensors for the irradiation test, and (4) the qualification tests(the helium leak test, the dimensional measurement, electric resistance measurements of sensors) of test fuel rods. Satisfactory results were obtained for all the qualification tests of the instrumented fuel rods for the 3-Pin FTL. Therefore the three instrumented fuel rods for the 3-Pin FTL have been fabricated successfully. These will be installed in the In-Pile Section of 3-Pin FTL. And the irradiation test of these fuel rods is planned from the early next year for about 3 years at HANARO

  19. Flame photometric detection of sodium leaks: Tests on a fullscale model for the control gear sodium rig

    International Nuclear Information System (INIS)

    Grundy, B.R.; Knowles, P.

    1971-01-01

    The proposed arrangement for detecting sodium leaks from the large flanges of the Control Gear Sodium Rig (Test Section No. 8, MCTR) at REML is to jacket then in a secondary containment from which air samples will be continuously pumped. Pipework feeds the air to a flame photometer which responds if soditun is present. To prove that sodium smoke could be transferred through the system, tests were performed on a fullscale model by burning small amounts of sodium in different jackets. Large signals free from fluctuations were obtained in all tests, peak response occurring in 2 1/4 minutes or less. The signal quickly cleared after isolating the appropriate vessel. A waiting period of several hours was sufficient to reduce the signal to zero, no cleaning of pipework, etc being necessary. In contrast, samples of two lagging materials heated to 400 °C gave no response with the photometer at maximum sensitivity. (author)

  20. LOCA scenario tests of irradiated fuel rod specimens

    International Nuclear Information System (INIS)

    Scott, Harold

    2004-01-01

    Full text: The NRC's cladding performance program at Argonne National Laboratory (ANL) is testing fueled high-burnup segments subjected to LOCA integral phenomena. The data are provided to NRC and the nuclear industry for their independent assessment of the adequacy of licensing criteria for LOCA events. The tests are being conducted with high-burnup 30 cm segments from Limerick (9x9 Zry-2) and H.B. Robinson (15x15 Zry-4) reactors. Prior to testing, sibling samples are characterized with respect to fuel morphology, fuel-cladding bond, cladding oxide layer thickness, hydrogen content and high-temperature steam oxidation kinetics. Specimens that survive quench are subjected to four-point bend tests, followed by local diametral compression tests. The retention of post-quench ductility is a more limiting requirement than surviving thermal stresses during quench. Companion tests are conducted with unirradiated cladding to generate baseline data for comparison with the high-burnup fuel results. LOCA integral tests have the following sequential steps: stabilization of temperature, internal pressure and steam flow at 300 C, ramping of temperature (∼5C/s) through ballooning and burst to 1204 C, hold at 1204 C for 1-5 minutes, slow-cooling (∼3C/s) to 800 C, and water quenching at ∼800C. Two high-burnup tests were completed in 2002 with Limerick BWR rod segments: ramp to burst in argon followed by slow cooling; and the LOCA test with 5-minute hold time at 1204 C, followed by slow cooling. With the exception of burst-opening shape, results for burst temperature, burst pressure, burst length, and ballooning strain profile are more similar to, than different from, results for unirradiated Zry-2 cladding exposed to the same time-temperature history. The 3rd Limerick test with quench was performed in December 2003, and a 4th Limerick test was performed in March 2004. Tests on high-burnup Robinson PWR fuel segments are scheduled to begin in June 2004. The presentation points

  1. Study on anti-seismic test of control rod driving system suspended by magnetic force

    International Nuclear Information System (INIS)

    Zhang Zhihua; Qian Dazhi; Xu Xianqi; Huang Hongwen; Zhang Zhengming; Wu Xinxin; Hu Xiao

    2012-01-01

    To verify the stability, reliability and security function in extreme conditions, the anti-seismic test of control rod drive line was conducted. Drop-time of control rod drive line in different earthquake intensities was got. The response and strain values of control rod drive line acceleration on SL-1, SL-2 level were measured. Safety functions of control rod drive line were validated in different work conditions. Anti-seismic test data shows that the driving system can keep the structure's integrality and realize operation function under OBE and SSE. (authors)

  2. Water hammer and column separation due to accidental simultaneous closure of control valves in a large scale two-phase flow experimental test rig

    NARCIS (Netherlands)

    Bergant, A.; Westende, van 't J.M.C.; Koppel, T.; Gale, J.; Hou, Q.; Pandula, Z.; Tijsseling, A.S.

    2010-01-01

    A large-scale pipeline test rig at Deltares, Delft, The Netherlands has been used for filling and emptying experiments. Tests have been conducted in a horizontal 250 mm diameter PVC pipe of 258 m length with control valves at the downstream and upstream ends. This paper investigates the accidental

  3. Heat transfer in a seven-rod test bundle with supercritical pressure water (1). Experiments

    International Nuclear Information System (INIS)

    Ezato, Koichiro; Seki, Yohji; Dairaku, Masayuki; Suzuki, Satoshi; Enoeda, Mikio; Akiba, Masato; Mori, H.; Oka, Y.

    2009-01-01

    Heat transfer experiments in a seven-rod test bundle with supercritical pressure water has been carried out. The pressure drop and heat transfer coefficients (HTCs) in the test section are evaluated. In the present limited conditions, difference between HTCs at the surface facing the sub-channel center and those at the surface in the narrowest region between rods is not observed. (author)

  4. Characterization of the lubricity of bio-oil/diesel fuel blends by high frequency reciprocating test rig

    International Nuclear Information System (INIS)

    Xu, Yufu; Wang, Qiongjie; Hu, Xianguo; Li, Chuan; Zhu, Xifeng

    2010-01-01

    The diesel fuel was mixed with the rice husk bio-oil using some emulsifiers based on the theory of Hydrophile-Lipophile Balance (HLB). The lubricity of the bio-oil/diesel fuel blend was studied on a High Frequency Reciprocating Test Rig (HFRR) according to ASTM D 6079-2004. The microscopic topography and chemical composition on the worn surface were analyzed respectively using scanning electron microscopy (SEM) and energy dispersive spectrometer (EDS). The profile and surface roughness of the rubbed trace were measured using a profilometer. The chemical group and composition were studied by a Fourier transform infrared spectrometry (FTIR). The results showed that the lubrication ability of the present fuel blend was better than that of the Chinese conventional diesel fuel (number zero). However, the anti-corrosion and anti-wear properties of the fuel blend were not satisfactory in comparison with those of conventional diesel fuel.

  5. Report on the evaluation of the tritium producing burnable absorber rod lead test assembly. Revision 1

    International Nuclear Information System (INIS)

    1997-03-01

    This report describes the design and fabrication requirements for a tritium-producing burnable absorber rod lead test assembly and evaluates the safety issues associated with tritium-producing burnable absorber rod irradiation on the operation of a commercial light water reactor. The report provides an evaluation of the tritium-producing burnable absorber rod design and concludes that irradiation can be performed within U.S. Nuclear Regulatory Commission regulations applicable to a commercial pressurized light water reactor

  6. Rig it right! Maya animation rigging concepts

    CERN Document Server

    O'Hailey, Tina

    2013-01-01

    Rigging a character can be a complicated undertaking. Move from a bi-pedal character to a quad- or poly-pedal and, well, things just got real. Where do you begin? Unlike all of those button-pushing manuals out there, Rig it Right! breaks down rigging so that you can achieve a fundamental understanding of the concept, allowing you to rig more intuitively in your own work. Veteran animation professor Tina O'Hailey will get you up and rigging in a matter of hours with step-by-step tutorials covering multiple animation control types, connection methods, interactive skinning, Blend

  7. A comparison between burn-out data for 19-rod cluster test-sections cooled by Freon-12 at 155 lb/in2 (abs), and by water at 1000 lb/in2 in vertical upflow

    International Nuclear Information System (INIS)

    Stevens, G.F.; Wood, R.W.

    1966-01-01

    Previous experiments on the Winfrith Freon Rig have produced scaling factors which relate these Freon experiments to the corresponding experiments in water with an accuracy of about 10%. It has also been found that the Freon rig is accurate, economical and easy to use. The scaling factors so obtained have now been tested against data for 19-rod clusters which had previously been tested at Columbia University. This report presents the results of the rod cluster tests in which comparison is made between Freon-12 and water for three test-sections which differ in the means of spacing the individual rods. All the test-sections were heated uniformly with respect to length, but had a radial flux depression of nominally 0.70/1.0. The results provide strong evidence that the scaling factor method using Freon-12 at 155 lb/in 2 (abs) is a useful technique for predicting the behaviour at burn-out of complicated test-sections cooled by boiling water at 1000 lb/in 2 with only one-eighteenth of the power required for the water experiment. In particular, the Freon tests reproduce closely the relative burn-out powers previously measured in water. It has also been found that repeated rebuilding of a nominally unchanged cluster from the same components can produce burn-out powers differing by ± 6%. This new result illustrates the power and value of the Freon technique. (author)

  8. "Fan-Tip-Drive" High-Power-Density, Permanent Magnet Electric Motor and Test Rig Designed for a Nonpolluting Aircraft Propulsion Program

    Science.gov (United States)

    Brown, Gerald V.; Kascak, Albert F.

    2004-01-01

    A scaled blade-tip-drive test rig was designed at the NASA Glenn Research Center. The rig is a scaled version of a direct-current brushless motor that would be located in the shroud of a thrust fan. This geometry is very attractive since the allowable speed of the armature is approximately the speed of the blade tips (Mach 1 or 1100 ft/s). The magnetic pressure generated in the motor acts over a large area and, thus, produces a large force or torque. This large force multiplied by the large velocity results in a high-power-density motor.

  9. Fuel rod-grid interaction wear: in-reactor tests (LWBR development program)

    International Nuclear Information System (INIS)

    Stackhouse, R.M.

    1979-11-01

    Wear of the Zircaloy cladding of LWBR irradiation test fuel rods, resulting from relative motion between rod and rod support contacts, is reported. Measured wear depths were small, 0.0 to 2.7 mils, but are important in fuel element behavior assessment because of the local loss of cladding thickness, as well as the effect on grid spring forces that laterally restrain the rods. An empirical wear analysis model, based on out-of-pile tests, is presented. The model was used to calculate the wear on the irradiation test fuel rods attributed to a combination of up-and-down motions resulting from power and pressure/temperature cycling of the test reactor, flow-induced vibrations, and assembly handling scratches. The calculated depths are generally deeper than the measured depths

  10. Fuel-rod response during the large-break LOCA Test LOC-6

    International Nuclear Information System (INIS)

    Vinjamuri, K.; Cook, B.A.; Hobbins, R.R.

    1981-01-01

    The large break Loss of Coolant Accident (LOCA) Test LOC-6 was conducted in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory by EG and G Idaho, Inc. The objectives of the PBF LOCA tests are to obtain in-pile cladding ballooning data under blowdown and reflood conditions and assess how well out-of-pile ballooning data represent in-pile fuel rod behavior. The primary objective of the LOC-6 test was to determine the effects of internal rod pressures and prior irradiation on the deformation behavior of fuel rods that reached cladding temperatures high in the alpha phase of zircaloy. Test LOC-6 was conducted with four rods of PWR 15 x 15 design with the exception of fuel stack length (89 cm) and enrichment (12.5 W% 235 U). Each rod was surrounded by an individual flow shroud

  11. Damage and failure of unirradiated and irradiated fuel rods tested under film boiling conditions

    International Nuclear Information System (INIS)

    Mehner, A.S.; Hobbins, R.R.; Seiffert, S.L.; MacDonald, P.E.; McCardell, R.K.

    1979-01-01

    Power-cooling-mismatch experiments are being conducted as part of the Thermal Fuels Behavior Program in the Power Burst Facility at the Idaho National Engineering Laboratory to evaluate the behavior of unirradiated and previously irradiated light water reactor fuel rods tested under stable film boiling conditions. The observed damage that occurs to the fuel rod cladding and the fuel as a result of film boiling operation is reported. Analyses performed as a part of the study on the effects of operating failed fuel rods in film boiling, and rod failure mechanisms due to cladding embrittlement and cladding melting upon being contacted by molten fuel are summarized

  12. Preliminary performance test of control rod position indicator for ballscrew type CEDM

    International Nuclear Information System (INIS)

    Yoo, J. Y.; Kim, J. H.; Hu, H.; Lee, J. S.; Kim, J. I.

    2003-01-01

    The reliability and accuracy of the information on control rod position are very important to the reactor safety and the design of the core protection system. A survey on the RSPT(Reed Switch Position Transmitter) type control rod position indication system and its actual implementation in the exiting nuclear power plants in Korea was performed first. The prototype of control rod position indicator having the high performance for the ballscrew type CEDM was developed on the basis of RSPT technology identified through the survey. The characteristics of control rod position indicator was defined and documented through design procedure and preliminary performance test

  13. The art of rigging

    CERN Document Server

    Biddlecombe, George

    1990-01-01

    The best manual ever produced on rigging a sailing ship, based on extensively revised and updated 1848 edition prepared by Biddlecombe, Master in the Royal Navy. Complete definition of terms, on-shore operations, process of rigging ships, reeving the running rigging and bending sails, rigging brigs, yachts and small vessels, more. 17 plates.

  14. Out-pile Test of Double Cladding Fuel Rod Mockups for a Nuclear Fuel Irradiation Test

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Jaemin; Park, Sungjae; Kang, Younghwan; Kim, Harkrho; Kim, Bonggoo; Kim, Youngki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-05-15

    An instrumented capsule for a nuclear fuel irradiation test has been developed to measure fuel characteristics, such as a fuel temperature, internal pressure of a fuel rod, a fuel pellet elongation and a neutron flux during an irradiation test at HANARO. In the future, nuclear fuel irradiation tests under a high temperature condition are expected from users. To prepare for this request, we have continued developing the technology for a high temperature nuclear fuel irradiation test at HANARO. The purpose of this paper is to verify the possibility that the temperature of a nuclear fuel can be controlled at a high temperature during an irradiation test. Therefore we designed and fabricated double cladding fuel rod mockups. And we performed out-pile tests using these mockups. The purposes of a out-pile test is to analyze an effect of a gap size, which is between an outer cladding and an inner cladding, on the temperature and the effect of a mixture ratio of helium gas and neon gas on the temperature. This paper presents the design and fabrication of double cladding fuel rod mockups and the results of the out-pile test.

  15. Experimental Analysis of Mast Lifting and Bending Forces on Vibration Patterns Before and After Pinion Reinstallation in an OH-58 Transmission Test Rig

    Science.gov (United States)

    Huff, Edward M.; Lewicki, David G.; Tumer, Irem Y.; Decker, Harry; Barszez, Eric; Zakrajsek, James J.; Norvig, Peter (Technical Monitor)

    2000-01-01

    As part of a collaborative research program between NASA Ames Research Center (ARC), NASA Glenn Research Center (GRC), and the US Army Laboratory, a series of experiments is being performed in GRC's 500 HP OH-58 Transmission Test Rig facility and ARC's AH-I Cobra and OH-58c helicopters. The findings reported in this paper were drawn from Phase-I of a two-phase test-rig experiment, and are focused on the vibration response of an undamaged pinion gear operating in the transmission test rig. To simulate actual flight conditions, the transmission system was run at three torque levels, as well as two mast lifting and two mast bending levels. The test rig was also subjected to disassembly and reassembly of the main pinion housing to simulate the effect of maintenance operations. An analysis of variance based on the total power of the spectral distribution indicates the relative effect of each experimental factor, including Wong interactions with torque. Reinstallation of the main pinion assembly is shown to introduce changes in the vibration signature, suggesting the possibility of a strong effect of maintenance on HUMS design and use. Based on these results, further research will be conducted to compare these vibration responses with actual OH58c helicopter transmission vibration patterns.

  16. Tensile and burst tests in support of the cadmium safety rod failure evaluation

    International Nuclear Information System (INIS)

    Thomas, J.K.

    1992-02-01

    The reactor safety rods may be subjected to high temperatures due to gamma heating after the core coolant level has dropped during the ECS phase of hypothetical LOCA event. Accordingly, an experimental safety rod testing subtask was established as part of a task to address the response of reactor core components to this accident. This report discusses confirmatory separate effects tests conducted to support the evaluation of failures observed in the safety rod thermal tests. As part of the failure evaluation, the potential for liquid metal embrittlement (LME) of the safety rod cladding by cadmium (Cd) -- aluminum (Al) solutions was examined. Based on the test conditions, literature data, and U-Bend tests, its was concluded that the SS304 safety rod cladding would not be subject to LME by liquid Cd-Al solutions under conditions relevant to the safety rod thermal tests or gamma heating accident. To confirm this conclusion, tensile tests on SS304 specimens were performed in both air and liquid Cd-Al solutions with the range of strain rates, temperatures, and loading conditions spanning the range relevant to the safety rod thermal tests and gamma heating accident

  17. Physics calculations for the RIA 1-3 irradiated rod test

    International Nuclear Information System (INIS)

    Young, T.E.

    1981-06-01

    The RIA 1-3 test would employ a square array of four pre-irradiated BWR rods to provide information on fuel failure modes and consequences of postulated Reactivity Initiated Accidents in power reactors. Calculations were done to: (1) predict R-O power distributions in the test rods for thermal-hydraulic and fuel-failure analysis; and (2) predict the steady-state and transient ratios of test fuel energy deposition to core energy deposition (Figures of Merit). Fission distributions for the test were computed with the RAFFL Monte Carlo code using an external neutron current source from a complete-reactor radial calculation with the SCAMP S/sub n/ code. Energies per fission for the rods were computed using the SINBAD buildup and depletion code, the GAMSOR gamma ray source code, and the QAD-BSA point-kernel shielding code. The calculated rod average-to-test average energy deposition ratios are 0.99, 0.99, and 0.97 for the rods irradiated to approximately 12 CWd/tu, and 1.04 for the rod irradiated to 4.8 GWd/tu. The maximum deviation of the power density of 1/12-rod azimuthal segments from the rod average is 4%. For an estimated control rod position of 0.591 m withdrawn the predicted radial average energy deposition at the axial peak in an average test rod is 1.71 (kW/m)/MW during preconditioning, and 1.84 (kJ/kg UO 2 ) MW.S during the burst. 16 figures, 7 tables

  18. What Really Rigs Up RIG-I?

    Science.gov (United States)

    Barik, Sailen

    2016-01-01

    RIG-I (retinoic acid-inducible gene 1) is an archetypal member of the cytoplasmic DEAD-box dsRNA helicase family (RIG-I-like receptors or RLRs), the members of which play essential roles in the innate immune response of the metazoan cell. RIG-I functions as a pattern recognition receptor that detects nonself RNA as a pathogen-associated molecular pattern (PAMP). However, the exact molecular nature of the viral RNAs that act as a RIG-I ligand has remained a mystery and a matter of debate. In this article, we offer a critical review of the actual viral RNAs that act as PAMPs to activate RIG-I, as seen from the perspective of a virologist, including a recent report that the viral Leader-read-through transcript is a novel and effective RIG-I ligand. © 2016 S. Karger AG, Basel.

  19. Experience using individually supplied heater rods in critical power testing of advanced BWR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Majed, M.; Morback, G.; Wiman, P. [ABB Atom AB, Vasteras (Sweden)] [and others

    1995-09-01

    The ABB Atom FRIGG loop located in Vasteras Sweden has during the last six years given a large experience of critical power measurements for BWR fuel designs using indirectly heated rods with individual power supply. The loop was built in the sixties and designed for maximum 100 bar pressure. Testing up to the mid eighties was performed with directly heated rods using a 9 MW, 80 kA power supply. Providing test data to develop critical power correlations for BWR fuel assemblies requires testing with many radial power distributions over the full range of hydraulic conditions. Indirectly heated rods give large advantages for the testing procedure, particularly convenient for variation of individual rod power. A test method being used at Stern Laboratories (formerly Westinghouse Canada) since the early sixties, allows one fuel assembly to simulate all required radial power distributions. This technique requires reliable indirectly heated rods with independently controlled power supplies and uses insulated electric fuel rod simulators with built-in instrumentation. The FRIGG loop was adapted to this system in 1987. A 4MW power supply with 10 individual units was then installed, and has since been used for testing 24 and 25 rod bundles simulating one subbundle of SVEA-96/100 type fuel assemblies. The experience with the system is very good, as being presented, and it is selected also for a planned upgrading of the facility to 15 MW.

  20. Heat transfer coefficient testing in nuclear fuel rod bundles with mixing vane grids

    International Nuclear Information System (INIS)

    Conner, Michael E.; Smith, L. David III; Holloway, Mary V.; Beasley, Donald E.

    2005-01-01

    An air heat transfer test facility was developed to test the heat transfer downstream of support grids in simulated PWR nuclear fuel rod bundles. The goal of this testing is to study the single-phase heat transfer coefficients downstream of grids with mixing vanes in a square-pitch rod bundle. The technique developed utilizes fully-heated grid spans and a specially designed thermocouple holder that can be moved axially down the rod bundle and aximuthally within a test rod. From this testing, the axial and aximuthally varying heat transfer coefficient can be determined. Different grid designs are tested and compared to determine the heat transfer enhancement associated with key grid features such as mixing vanes. (author)

  1. The Winfrith 9MW heat transfer rig

    International Nuclear Information System (INIS)

    Obertelli, J.D.

    1976-01-01

    The Winfrith 9MW Rig is used for studying heat transfer and flow resistance in a variety of test sections at system pressures up to 68 bar. The basic rig and its instrumentation are discussed together with the characteristics of the test section design. The rig has been used in studies involving the full scale simulation of Steam Generating Heavy Water (SGHW) fuel assemblies and the paper discusses the measurements made in this type of study. (author)

  2. Experimental data report for Test TS-2 reactivity initiated accident test in NSRR with pre-irradiated BWR fuel rod

    International Nuclear Information System (INIS)

    Nakamura, Takehiko; Yoshinaga, Makio; Sobajima, Makoto; Fujishiro, Toshio; Kobayashi, Shinsho; Yamahara, Takeshi; Sukegawa, Tomohide; Kikuchi, Teruo

    1993-02-01

    This report presents experimental data for Test TS-2 which was the second test in a series of Reactivity Initiated Accident (RIA) condition test using pre-irradiated BWR fuel rods, performed at the Nuclear Safety Research Reactor (NSRR) in February, 1990. Test fuel rod used in the Test TS-2 was a short sized BWR (7x7) type rod which was fabricated from a commercial rod irradiated at Tsuruga Unit 1 power reactor. The fuel had an initial enrichment of 2.79% and a burnup of 21.3Gwd/tU (bundle average). A pulse irradiation of the test fuel rod was performed under a cooling condition of stagnant water at atmospheric pressure and at ambient temperature which simulated a BWR's cold start-up RIA event. The energy deposition of the fuel rod in this test was evaluated to be 72±5cal/g·fuel (66±5cal/g·fuel in peak fuel enthalpy) and no fuel failure was observed. Descriptions on test conditions, test procedures, transient behavior of the test rod during the pulse irradiation, and, results of pre and post pulse irradiation examinations are described in this report. (author)

  3. Testing of LWR fuel rods to support criticality safety analysis of transport accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Purcell, P.C. [BNFL International Transport, Spent Fuel Services (United Kingdom); Dallongeville, M. [COGEMA Logistics (AREVA Group) (France)

    2004-07-01

    For the transport of low enriched materials, criticality safety may be demonstrated by applying pessimistic modelling assumptions that bound any realistic case. Where Light Water Reactor (LWR) fuel is being transported, enrichment levels are usually too high to permit this approach and more realistic data is needed. This requires a method by which the response of LWR fuel under impact accident conditions can be approximated or bounded. In 2000, BNFL and COGEMA LOGISTICS jointly commenced the Fuel Integrity Project (FIP) whose objective was to develop such methods. COGEMA LOGISTICS were well advanced with a method for determining the impact response of unirradiated fuel, but required further test data before acceptance by the Transport Regulators. The joint project team extensively discussed the required inputs to the FIP, from which it was agreed that BNFL would organise new tests on both unirradiated and irradiated fuel samples and COGEMA LOGISTICS would take major responsibility for evaluating the test results. Tests on unirradiated fuel rod samples involved both dynamic and quasi-static loading on fuel samples. PWR fuel rods loaded with uranium pellets were dropped vertically from 9m onto a rigid target and this was repeated on BWR fuel rods, similar tests on empty fuel rods were also conducted. Quasi-static tests were conducted on 530 mm long PWR and BWR fuel specimens under axial loading. Tests on irradiated fuel samples were conducted on high burn-up fuel rods of both PWR and BWR types. These were believed original to the FIP project and involved applying bending loads to simply supported pressurised rod specimens. In one test the fuel rod was heated to nearly 500oC during loading, all specimens were subject to axial impact before testing. Considerable experience of fuel rod testing and new data was gained from this test programme.

  4. Development of Welding and Instrumentation Technology for Nuclear Fuel Test Rod

    International Nuclear Information System (INIS)

    Joung, Chang Young; Ahn, Sung Ho; Heo, Sung Ho; Hong, Jin Tae; Kim, Ka Hye

    2013-01-01

    It is necessary to develop various types of welding, instrumentation and helium gas filling techniques that can conduct TIG spot welding exactly at a pin-hole of the end-cap on the nuclear fuel rod to fill up helium gas. The welding process is one of the most important among the instrumentation processes of the nuclear fuel test rod. To manufacture the nuclear fuel test rod, a precision welding system needs to be fabricated to develop various welding technologies of the fuel test rod jointing the various sensors and end-caps on a fuel cladding tube, which is charged with fuel pellets and component parts. We therefore designed and fabricated an orbital TIG welding system and a laser welding system. This paper describes not only some experiment results from weld tests for the parts of a nuclear fuel test rod, but also the contents for the instrumentation process of the dummy fuel test rod installed with the C-type T. C. A dummy nuclear fuel test rod was successfully fabricated with the welding and instrumentation technologies acquired with various tests. In the test results, the round welding has shown a good weldability at both the orbital TIG welding system and the fiber laser welding system. The spot welding to fill up helium gas has shown a good welding performance at a welding current of 30A, welding time of 0.4 sec and gap of 1 mm in a helium gas atmosphere. The soundness of the nuclear fuel test rod sealed by a mechanical sealing method was confirmed by helium leak tests and microstructural analyses

  5. Development of Welding and Instrumentation Technology for Nuclear Fuel Test Rod

    Energy Technology Data Exchange (ETDEWEB)

    Joung, Chang Young; Ahn, Sung Ho; Heo, Sung Ho; Hong, Jin Tae; Kim, Ka Hye [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    It is necessary to develop various types of welding, instrumentation and helium gas filling techniques that can conduct TIG spot welding exactly at a pin-hole of the end-cap on the nuclear fuel rod to fill up helium gas. The welding process is one of the most important among the instrumentation processes of the nuclear fuel test rod. To manufacture the nuclear fuel test rod, a precision welding system needs to be fabricated to develop various welding technologies of the fuel test rod jointing the various sensors and end-caps on a fuel cladding tube, which is charged with fuel pellets and component parts. We therefore designed and fabricated an orbital TIG welding system and a laser welding system. This paper describes not only some experiment results from weld tests for the parts of a nuclear fuel test rod, but also the contents for the instrumentation process of the dummy fuel test rod installed with the C-type T. C. A dummy nuclear fuel test rod was successfully fabricated with the welding and instrumentation technologies acquired with various tests. In the test results, the round welding has shown a good weldability at both the orbital TIG welding system and the fiber laser welding system. The spot welding to fill up helium gas has shown a good welding performance at a welding current of 30A, welding time of 0.4 sec and gap of 1 mm in a helium gas atmosphere. The soundness of the nuclear fuel test rod sealed by a mechanical sealing method was confirmed by helium leak tests and microstructural analyses.

  6. Testing of LWR fuel rods to support criticality safety analysis of transport accident conditions

    International Nuclear Information System (INIS)

    Purcell, P.C.; Dallongeville, M.

    2004-01-01

    For the transport of low enriched materials, criticality safety may be demonstrated by applying pessimistic modelling assumptions that bound any realistic case. Where Light Water Reactor (LWR) fuel is being transported, enrichment levels are usually too high to permit this approach and more realistic data is needed. This requires a method by which the response of LWR fuel under impact accident conditions can be approximated or bounded. In 2000, BNFL and COGEMA LOGISTICS jointly commenced the Fuel Integrity Project (FIP) whose objective was to develop such methods. COGEMA LOGISTICS were well advanced with a method for determining the impact response of unirradiated fuel, but required further test data before acceptance by the Transport Regulators. The joint project team extensively discussed the required inputs to the FIP, from which it was agreed that BNFL would organise new tests on both unirradiated and irradiated fuel samples and COGEMA LOGISTICS would take major responsibility for evaluating the test results. Tests on unirradiated fuel rod samples involved both dynamic and quasi-static loading on fuel samples. PWR fuel rods loaded with uranium pellets were dropped vertically from 9m onto a rigid target and this was repeated on BWR fuel rods, similar tests on empty fuel rods were also conducted. Quasi-static tests were conducted on 530 mm long PWR and BWR fuel specimens under axial loading. Tests on irradiated fuel samples were conducted on high burn-up fuel rods of both PWR and BWR types. These were believed original to the FIP project and involved applying bending loads to simply supported pressurised rod specimens. In one test the fuel rod was heated to nearly 500oC during loading, all specimens were subject to axial impact before testing. Considerable experience of fuel rod testing and new data was gained from this test programme

  7. Technical description of the NRC long-term whole-rod and crud performance test

    International Nuclear Information System (INIS)

    Einziger, R.E.; Fish, R.L.; Knecht, R.L.

    1982-09-01

    Westinghouse Hanford Company (WHC) and EG and G-Idaho are jointly conducting a long-term, low-temperature, spent-fuel, whole rod and crud behavior test to provide the Nuclear Regulatory Commission (NRC) with information to assist in the licensing of light water reactor (LWR) spent-fuel, dry storage facilities. Readily available fuel rods from an H.B. Robinson Unit 2 (PWR) fuel assembly and a Peach Bottom-II (BWR) fuel assembly were selected for use in the 50-month test. Both intact and defected rods will be tested in inert and oxidizing atmospheres. A 230 0 C test temperature was selected for the first 10-month run. Both nondestructive and destructive examinations are planned to characterize the fuel rod behavior during the 5-y test. Four interim examinations and a final examination will be conducted. Crud spallation behavior will be investigated by sampling the crud particulate from the test capsules at each of the four interim examinations and at the end of the test. The background to whole rod testing, description of rod breach mechanisms, and a detailed description of the test are presented in this document

  8. Design of a Portable Tire Test Rig and Vehicle Roll-Over Stability Control

    OpenAIRE

    Fox, Derek Martin

    2009-01-01

    Vehicle modeling and simulation have fast become the easiest and cheapest method for vehicle testing. No longer do multiple, intensive, physical tests need be performed to analyze the performance parameters that one wishes to validate. One component of the vehicle simulation that is crucial to the correctness of the result is the tire. Simulations that are run by a computer can be run many times faster than a real test could be performed, so the cost and complexity of the testing is reduced....

  9. Wear Characterization of Carbon Nanotubes Reinforced Acetal Spur, Helical, Bevel and Worm Gears Using a TS Universal Test Rig

    Science.gov (United States)

    Yousef, Samy; Osman, T. A.; Abdalla, Abdelrahman H.; Zohdy, Gamal A.

    2015-12-01

    Although the applications of nanotechnologies are increasing, there remains a significant barrier between nanotechnology and machine element applications. This work aims to remove this barrier by blending carbon nanotubes (CNT) with common types of acetal polymer gears (spur, helical, bevel and worm). This was done by using adhesive oil (paraffin) during injection molding to synthesize a flange and short bars containing 0.02% CNT by weight. The flanges and short bars were machined using hobbing and milling machines to produce nanocomposite polymer gears. Some defects that surfaced in previous work, such as the appearance of bubbles and unmelted pellets during the injection process, were avoided to produce an excellent dispersion of CNT in the acetal. The wear resistances of the gears were measured by using a TS universal test rig using constant parameters for all of the gears that were fabricated. The tests were run at a speed of 1420 rpm and a torque of 4 Nm. The results showed that the wear resistances of the CNT/acetal gears were increased due to the addition of CNT, especially the helical, bevel and worm gears.

  10. Examination of cadmium safety rod thermal test specimens and failure mechanism evaluation

    International Nuclear Information System (INIS)

    Thomas, J.K.; Peacock, H.B.; Iyer, N.C.

    1992-01-01

    The reactor safety rods may be subjected to high temperatures due to gamma heating after the core coolant level has dropped during the ECS phase of a hypothetical LOCA event. Accordingly, an experimental cadmium safety rod testing subtask was established as part of a task to address the response of reactor core components to this accident. Companion reports describe the experiments and a structural evaluation (finite element analysis) of the safety rod. This report deals primarily with the examination of the test specimens, evaluation of possible failure mechanisms, and confirmatory separate effects experiments. It is concluded that the failures observed in the cadmium safety rod thermal tests which occurred at low temperature (T 800 degrees C) with fast thermal ramp rates are concluded to be mechanical in nature without significant environmental degradation. Based on these tests, tasks were initiated to design and manufacture B 4 C safety rods to replace the cadmium safety rods. The B 4 C safety rods have been manufactured at this time and it is currently planned to charge them to the reactor in the near future. 60 refs

  11. Postirradiation examination results for the Irradiation Effects Test Series IE-ST-2, Rod IE-002

    International Nuclear Information System (INIS)

    Murdock, B.A.

    1977-12-01

    A postirradiation examination was conducted on a zircaloy-clad, UO 2 -fueled, pressurized water reactor (PWR) type rod which had been tested in the Power Burst Facility as part of the Irradiation Effects Test Series of the Thermal Fuels Behavior Program. The fuel rod, previously irradiated to a burnup of 15,800 MWd/t was subjected to a power ramp from 28 to 55 kW/m peak power at an average ramp rate of 4 kW/m/min. Posttest fuel restructuring and relocation, fission product redistribution, and fuel rod cladding deformation were evaluated and analyzed

  12. Seismic appraisal test of control rod drive mechanism of China experiment fast reactor

    International Nuclear Information System (INIS)

    Song Qing; Yang Hongyi; Jing Yueqing; Wen Jing; Liu Guijuan; Sun Lei

    2008-01-01

    The structure of the control rod drive mechanism in pool type sodium-cooled fast reactor is the characterized by long, thin, and geometric nonlinearity, and the seismic load is multiple activation. The anti-seismic evaluation is always paid great attention by the countries developing the technology worldwide. This article introduces the seismic appraisal test of the control rod drive mechanism of China Experimental Fast Reactor (CEFR) performed on a seismic platform which is vertical shaft style and multiple activation. The result of the test shows the structural integrity and the function of the control rod drive mechanism could meet the design requirements of the earthquake intensity. (authors)

  13. Post test investigation of the single rod tests ESSI 1-11 on temperature escalation in PWR fuel rod simulators due to the Zircaloy/steam reaction

    International Nuclear Information System (INIS)

    Hagen, S.; Kapulla, H.; Malauschek, H.; Katanishi, S.

    1987-03-01

    This KfK-report describes the posttest investigation of the single rod tests ESSI-1 to ESSI-11. The objective of these tests was to investigate the temperature escalation behaviour of Zircaloy clad PWR-fuel rods in steam. The investigation of the temperature escalation is part of the program of out-of-pile experiments (CORA) performed within the frame work of the PNS Severe Fuel Damage Program. The experimental arrangement consisted of fuel rod simulator (central tungsten heater, UO 2 ring pellets and Zircaloy cladding), Zircaloy shroud and fiber ceramic insulation. The introductory test ESSI-1 to ESSI-3 were scoping tests designed to obtain information on the temperature escalation of zircaloy in steam. ESSI-4 to ESSI-8 were run with increasing heating rates to investigate the influence of the oxide layer thickness at the start of the escalation. ESSI-9 to ESSI-11 were performed to investigate the influence of the insulation thickness on the escalation behaviour. In these tests we also learned that the gap between removed shroud and insulation has a remarkable influence due to heat removal by convection in the gap. After the test the fuel rod simulator was embedded into epoxy and cut by a diamond saw. The cross sections were photographed and investigated by metalograph microscope, SEM and EMP examinations. (orig./GL) [de

  14. Thermal stability and filterability of jet fuels containing PDR additives in small-scale tests and realistic rig simulations

    Energy Technology Data Exchange (ETDEWEB)

    Bauldreay, J.M.; Clark, R.H.; Heins, R.J. [Shell Research, Ltd., Chester (United Kingdom)

    1995-05-01

    Specification, small-scale and realistic fuel simulation tests have addressed concerns about the impact of pipeline drag reducer (PDR) flow modifying additives on jet fuel handling and performance. A typical PDR additive tended to block filters which were similar to those used in the specification Jet Fuel Thermal Oxidation Tester (JFTOT) and other thermal stability test apparatus. Blockages reduced flow rates and PDR concentrations downstream of the filters. Consequently two PDR additives (A&B) were tested in JFTOT apparatus without the usual in-line pre-filters as part of a Ministry of Defense (MoD) co-ordinated Round Robin exercise. Some fuel/PDR additive combinations caused decreases in JFTOT breakpoints. Effects were additive- (type, concentration and degree of shear) and fuel-dependent; most failures were caused by filter blockages and not by a failing lacquer rating. In further work at Thornton, the thermal stability characteristics of similar fuel/additive combinations have been examined in non-specification tests. In Flask Oxidation Tests, PDR additives caused no significant increase in the liquid phase oxidation rates of the fuels. Additives were tested in the Single Tube Heat Transfer Rig (STHTR) which duplicates many of the conditions of a heat exchanger element in an engine`s fuel supply system. B produced an average two-fold decrease in thermal stability in a Merox fuel; A had no significant effect. In hydrotreated fuel, B reduced the thermal stability up to five-fold. A had little effect below 205{degrees}C, while at higher temperatures there may have been a marginal improvement in thermal stability. Again, certain jet fuel/PDR combinations were seen to reduce thermal stability.

  15. Pressure suppression experiments in the PSS test rig of the GKSS

    International Nuclear Information System (INIS)

    Aust, E.

    1975-01-01

    A pressure suppression system has been developed for the advanced pressurized water reactor. Due to its compact layout, this system enables the reactor plant to be installed in the ship in a volume and weight saving manner. Because of significant differences in design and construction of this system as compared to similar systems for land based nuclear power plants, a test facility was built to experimentally demonstrate the effectiveness and the functioning of this system. The test facility will be described and a program of the major experimental tests will be given. Finally, some preliminary results of tests with air carry over in the wet well will be presented. (orig.) [de

  16. Hearing protector fit testing with off-shore oil-rig inspectors in Louisiana and Texas.

    Science.gov (United States)

    Murphy, William J; Themann, Christa L; Murata, Taichi K

    2016-11-01

    This field study aimed to assess the noise reduction of hearing protection for individual workers, demonstrate the effectiveness of training on the level of protection achieved, and measure the time required to implement hearing protector fit testing in the workplace. The National Institute for Occupational Safety and Health (NIOSH) conducted field studies in Louisiana and Texas to test the performance of HPD Well-Fit. Fit tests were performed on 126 inspectors and engineers working in the offshore oil industry. Workers were fit tested with the goal of achieving a 25-dB PAR. Less than half of the workers were achieving sufficient protection from their hearing protectors prior to NIOSH intervention and training; following re-fitting and re-training, over 85% of the workers achieved sufficient protection. Typical test times were 6-12 minutes. Fit testing of the workers' earplugs identified those workers who were and were not achieving the desired level of protection. Recommendations for other hearing protection solutions were made for workers who could not achieve the target PAR. The study demonstrates the need for individual hearing protector fit testing and addresses some of the barriers to implementation.

  17. Experimental Test Rig for Optimal Control of Flexible Space Robotic Arms

    Science.gov (United States)

    2016-12-01

    the test bed design. A single link arm with a torsional, helical spring at the base was finalized to investigate the effects of coupling due to...test bed design. A single link arm with a torsional, helical spring at the base was finalized to investigate the effects of coupling due to movement...Source: [4]. A challenge with space systems is that it costs a lot of money to put a satellite or spacecraft into space. Estimates to send one kilogram

  18. High power breakdown testing of a photonic band-gap accelerator structure with elliptical rods

    Directory of Open Access Journals (Sweden)

    Brian J. Munroe

    2013-01-01

    Full Text Available An improved single-cell photonic band-gap (PBG structure with an inner row of elliptical rods (PBG-E was tested with high power at a 60 Hz repetition rate at X-band (11.424 GHz, achieving a gradient of 128  MV/m at a breakdown probability of 3.6×10^{-3} per pulse per meter at a pulse length of 150 ns. The tested standing-wave structure was a single high-gradient cell with an inner row of elliptical rods and an outer row of round rods; the elliptical rods reduce the peak surface magnetic field by 20% and reduce the temperature rise of the rods during the pulse by several tens of degrees, while maintaining good damping and suppression of high order modes. When compared with a single-cell standing-wave undamped disk-loaded waveguide structure with the same iris geometry under test at the same conditions, the PBG-E structure yielded the same breakdown rate within measurement error. The PBG-E structure showed a greatly reduced breakdown rate compared with earlier tests of a PBG structure with round rods, presumably due to the reduced magnetic fields at the elliptical rods vs the fields at the round rods, as well as use of an improved testing methodology. A post-testing autopsy of the PBG-E structure showed some damage on the surfaces exposed to the highest surface magnetic and electric fields. Despite these changes in surface appearance, no significant change in the breakdown rate was observed in testing. These results demonstrate that PBG structures, when designed with reduced surface magnetic fields and operated to avoid extremely high pulsed heating, can operate at breakdown probabilities comparable to undamped disk-loaded waveguide structures and are thus viable for high-gradient accelerator applications.

  19. Understanding Rig Rates

    OpenAIRE

    Petter Osmundsen; Knut Einar Rosendahl; Terje Skjerpen

    2013-01-01

    We examine the largest cost component in offshore development projects, drilling rates, which have been high over the last years. To our knowledge, rig rates have not been analysed empirically before in the economic literature. By econometric analysis we examine the effects on Gulf of Mexico rig rates of gas and oil prices, rig capacity utilization, contract length and lead time, and rig specific characteristics. Having access to a unique data set containing contract information, we are able ...

  20. Advances in measuring techniques for turbine cooling test rigs - Status report

    Science.gov (United States)

    Pollack, F. G.

    1974-01-01

    Instrumentation development pertaining to turbine cooling research has resulted in the design and testing of several new systems. Pressure measurements on rotating components are being made with a rotating system incorporating ten miniature transducers and a slip-ring assembly. The system has been tested successfully up to speeds of 9000 rpm. An advanced system development combining pressure transducer and thermocouple signals is also underway. Thermocouple measurements on rotating components are transferred off the shaft by a 72-channel rotating data system. Thermocouple data channels are electronically processed on board and then removed from the shaft in the form of a digital serial train by one winding of a rotary transformer.

  1. Investigation of Spiral Bevel Gear Condition Indicator Validation via AC-29-2C Combining Test Rig Damage Progression Data with Fielded Rotorcraft Data

    Science.gov (United States)

    Dempsey, Paula J.

    2015-01-01

    This is the final of three reports published on the results of this project. In the first report, results were presented on nineteen tests performed in the NASA Glenn Spiral Bevel Gear Fatigue Test Rig on spiral bevel gear sets designed to simulate helicopter fielded failures. In the second report, fielded helicopter HUMS data from forty helicopters were processed with the same techniques that were applied to spiral bevel rig test data. Twenty of the forty helicopters experienced damage to the spiral bevel gears, while the other twenty helicopters had no known anomalies within the time frame of the datasets. In this report, results from the rig and helicopter data analysis will be compared for differences and similarities in condition indicator (CI) response. Observations and findings using sub-scale rig failure progression tests to validate helicopter gear condition indicators will be presented. In the helicopter, gear health monitoring data was measured when damage occurred and after the gear sets were replaced at two helicopter regimes. For the helicopters or tails, data was taken in the flat pitch ground 101 rotor speed (FPG101) regime. For nine tails, data was also taken at 120 knots true airspeed (120KTA) regime. In the test rig, gear sets were tested until damage initiated and progressed while gear health monitoring data and operational parameters were measured and tooth damage progression documented. For the rig tests, the gear speed was maintained at 3500RPM, a one hour run-in was performed at 4000 in-lb gear torque, than the torque was increased to 8000 in-lbs. The HUMS gear condition indicator data evaluated included Figure of Merit 4 (FM4), Root Mean Square (RMS) or Diagnostic Algorithm 1(DA1), + 3 Sideband Index (SI3) and + 1 Sideband Index (SI1). These were selected based on their sensitivity in detecting contact fatigue damage modes from analytical, experimental and historical helicopter data. For this report, the helicopter dataset was reduced to

  2. Advance Noise Control Fan II: Test Rig Fan Risk Management Study

    Science.gov (United States)

    Lucero, John

    2013-01-01

    Since 1995 the Advanced Noise Control Fan (ANCF) has significantly contributed to the advancement of the understanding of the physics of fan tonal noise generation. The 9'x15' WT has successfully tested multiple high speed fan designs over the last several decades. This advanced several tone noise reduction concepts to higher TRL and the validation of fan tone noise prediction codes.

  3. First interim examination of defected BWR and PWR rods tested in unlimited air at 2290C

    International Nuclear Information System (INIS)

    Einziger, R.E.; Cook, J.A.

    1983-01-01

    A five-year whole rod test was initiated to investigate the long-term stability of spent fuel rods under a variety of possible dry storage conditions. Both PWR and BWR rods were included in the test. The first interim examination was conducted after three months of testing to determine if there was any degradation in those defected rods stored in an unlimited air atmosphere. Visual observations, diametral measurements and radiographic smears were used to assess the degree of cladding deformation and particulate dispersal. The PWR rod showed no measurable change from the pre-test condition. The two original artificial defects had not changed in appearance and there was no diametral growth of the cladding. One of the defects in BWR rod showed significant deformation. There was approximately 10% cladding strain at the defect site and a small axial crack had formed. The fuel in the defect did not appear to be friable. The second defect showed no visible change and no cladding strain. Following examination, the test was continued at 230 0 C. Another interim examination is planned during the summer of 1983. This paper discusses the details and meaning of the data from the first interim examination

  4. Post-test examination of the VVER-1000 fuel rod bundle CORA-W2

    International Nuclear Information System (INIS)

    Hofmann, P.; Noack, V.; Burbach, J.; Metzger, H.; Schanz, G.; Hagen, S.; Sepold, L.

    1995-01-01

    The upper half of the bundle is completely oxidized, the lower half has kept the fuel rods relatively intact. The post-test examination results show the strong impact of the B 4 C absorber rod and the stainless steel grid spacers on the 'low-temperature' bundle damage initiation and progression. The B 4 C absorber rod completely disappeared in the upper half of the bundle. The multicomponent melts relocated and formed coolant channel blockages on solidification with a maximum extent of about 30% in the lower part of the bundle. At temperatures above the melting point of the ZrNb1 cladding extensive fuel dissolution occured. (orig./HP)

  5. Study on dynamic rod worth measurement method and its test verification

    International Nuclear Information System (INIS)

    Wu Lei; Liu Tongxian; Zhao Wenbo; Li Songling; Yu Yingrui

    2015-01-01

    An advanced rod worth measurement technique, the dynamic rod worth measurement method (DRWM) has been developed. Static Spatial Factors (SSF) and Dynamic Spatial Factor (DSF) were introduced to improve the inverse kinetics method. The three dimensional steady and transient simulations for the measurement process was carried out to calculate the modification factors. The rod worth measurement, test was performed on a research reactor to verify DRWM. The results showed that the DRWM method provided the improved accuracy and could be a replacement of the traditional methods. (authors)

  6. Thermohydraulic tests of 3x3-rod bundle maquette

    International Nuclear Information System (INIS)

    Ladeira, L.C.D.

    1986-10-01

    The results of a 3x3-rod bundle thermohydraulic research program, performed in the Thermohydraulics Laboratory of NUCLEBRAS' Nuclear Technology Development Center, are briefly described. This program included measurements of pressure drops in one and two-phase flows, heat transfer coefficients, mixing between interconnected subchannels in one-phase flow conditions and critical heat fluxes. The measurements covered the following parameter ranges: heat fluxes from zero to the critical values, pressure ranging from 1 to 15 ata, inlet temperature from 25 to 150 sup(0)C and flow rate from 20 to 300l/min. (author)

  7. Laboratory manual for salt mixing test in rod bundles

    International Nuclear Information System (INIS)

    Khan, H.U.R.; Chiu, C.; Todreas, N.

    1978-10-01

    This report is a Laboratory Manual dealing with the procedure employed during Salt Tracer Experiments, which are used for evaluating the hydraulic characteristics of a rod bundle. A description of the standard equipment used is given together with details of manufacture of non-standard items i.e., probes used for detecting the salt-concentration. Details of bundle construction have not been included as they are available in the references cited. An attempt has also been made to point out potential trouble areas and procedures

  8. Development of a Wind Turbine Test Rig and Rotor for Trailing Edge Flap Investigation: Static Flap Angles Case

    International Nuclear Information System (INIS)

    Abdelrahman, Ahmed; Johnson, David A

    2014-01-01

    One of the strategies used to improve performance and increase the life-span of wind turbines is active flow control. It involves the modification of the aerodynamic characteristics of a wind turbine blade by means of moveable aerodynamic control surfaces. Trailing edge flaps are relatively small moveable control surfaces placed at the trailing edge of a blade's airfoil that modify the lift of a blade or airfoil section. An instrumented wind turbine test rig and rotor were specifically developed to enable a wide-range of experiments to investigate the potential of trailing edge flaps as an active control technique. A modular blade based on the S833 airfoil was designed to allow accurate instrumentation and customizable settings. The blade is 1.7 meters long, had a constant 178mm chord and a 6° pitch. The modular aerodynamic parts were 3D printed using plastic PC-ABS material. The blade design point was within the range of wind velocities in the available large test facility. The wind facility is a large open jet wind tunnel with a maximum velocity of 11m/s in the test area. The capability of the developed system was demonstrated through an initial study of the effect of stationary trailing edge flaps on blade load and performance. The investigation focused on measuring the changes in flapwise bending moment and power production for different trailing edge flap spanwise locations and deflection angles. The relationship between the load reduction and deflection angle was linear as expected from theory and the highest reduction was caused by the flap furthest from the rotor center. Overall, the experimental setup proved to be effective in measuring small changes in flapwise bending moment within the wind turbine blade and will provide insight when (active) flap control is targeted

  9. A Foil Thrust Bearing Test Rig for Evaluation of High Temperature Performance and Durability

    Science.gov (United States)

    2008-04-01

    composed of similar elements used in journal bearings, but are designed to support a shaft axially. Often, discrete compliant pads are attached... shaft designed to mate with a test thrust runner. The runner is mounted to the shaft with four high strength bolts, and an interference fit ensures...attached to the drive is able to stop the spindle quickly through dynamic braking of the shaft rotational energy. This spindle arrangement has

  10. Firmware development and testing of the ATLAS Pixel Detector / IBL ROD card

    International Nuclear Information System (INIS)

    Gabrielli, A.; Balbi, G.; Falchieri, D.; Lama, L.; Travaglini, R.; Backhaus, M.; Bindi, M.; Chen, S.P.; Hauck, S.; Hsu, S.C.; Flick, T.; Wensing, M.; Kretz, M.; Kugel, A.

    2015-01-01

    The ATLAS Experiment is reworking and upgrading systems during the current LHC shut down. In particular, the Pixel detector has inserted an additional inner layer called the Insertable B-Layer (IBL). The Readout-Driver card (ROD), the Back-of-Crate card (BOC), and the S-Link together form the essential frontend data path of the IBL's off-detector DAQ system. The strategy for IBL ROD firmware development was three-fold: keeping as much of the Pixel ROD datapath firmware logic as possible, employing a complete new scheme of steering and calibration firmware, and designing the overall system to prepare for a future unified code version integrating IBL and Pixel layers. Essential features such as data formatting, frontend-specific error handling, and calibration are added to the ROD data path. An IBL DAQ test bench using a realistic front-end chip model was created to serve as an initial framework for full offline electronic system simulation. In this document, major firmware achievements concerning the IBL ROD data path implementation, test on the test bench and ROD prototypes, will be reported. Recent Pixel collaboration efforts focus on finalizing hardware and firmware tests for the IBL. The plan is to approach a complete IBL DAQ hardware-software installation by the end of 2014

  11. Visualization test facility of nuclear fuel rod emergency cooling system

    International Nuclear Information System (INIS)

    Candido, Marcos Antonio; Mesquita, Amir Zacarias; Rezende, Hugo Cesar; Santos, Andre Augusto Campagnole

    2013-01-01

    The nuclear reactors safety is determined according to their protection against the consequences that may result from postulated accidents. The Loss of Coolant Accident (LOCA) is one the most important design basis accidents (DBA). The failure may be due to rupture of the primary loop piping. Another accident postulated is due to lack of power in the pump motors in the primary circuit. In both cases the reactor shut down automatically due to the decrease of reactivity to maintain the fissions, and to the drop of control rods. In the event of an accident it is necessary to maintain the coolant flow to remove the fuel elements residual heat, which remains after shut down. This heat is a significant amount of the maximum thermal power generated in normal operation (about 7%). Recently this event has been quite prominent in the press due to the reactor accident in Fukushima nuclear power station. This paper presents the experimental facility under rebuilding at the Thermal Hydraulic Laboratory of the Nuclear Technology Development Center (CDTN) that has the objective of monitoring and visualization of the process of emergency cooling of a nuclear fuel rod simulator, heated by Joule effect. The system will help the comprehension of the heat transfer process during reflooding after a loss of coolant accident in the fuel of light water reactor core. (author)

  12. A full size test rig of dry and dry-wet towers

    International Nuclear Information System (INIS)

    Fesson, J.-P.

    1981-01-01

    In order to test the various systems submitted by French companies, with a view to their application to the 900 MW and 1300 MW nuclear units, the tower is divided into two parts, each permitting the evacuation of an identical thermal charge. The first part includes a cross-current wet zone in which the water flows vertically and the air horizontally, connected to a set of vertical dry batteries. The second part includes bands of packing along the counter-current system, alternating with horizontal dry exchangers [fr

  13. Compilation of three-dimensional coordinates and specific data of the instrumentation of the prestressed concrete pressure vessel/high temperature helium test rig

    International Nuclear Information System (INIS)

    Klausinger, D.

    1977-04-01

    The positions of the thermoelements, strain gauges of various types, and of Gloetzl instruments installed by SGAE in the model vessel of the Common Project Prestressed Concrete Pressure Vessel/High Temperature Helium Test Rig are defined in cylindrical coordinates. The specific data of the instruments are given like configuration of multiple instruments; type, group and number of the instrument; number of cable and of channel; calibration factors; resistances of instruments and cables. (author)

  14. Experimental data report for Test TS-1 Reactivity Initiated Accident Test in NSRR with pre-irradiated BWR fuel rod

    International Nuclear Information System (INIS)

    Nakamura, Takehiko; Yoshinaga, Makio; Sobajima, Makoto; Fujishiro, Toshio; Horiki, Ohichiro; Yamahara, Takeshi; Ichihashi, Yoshinori; Kikuchi, Teruo

    1992-01-01

    This report presents experimental data for Test TS-1 which was the first in a series of tests, simulating Reactivity Initiated Accident (RIA) conditions using pre-irradiated BWR fuel rods, performed in the Nuclear Safety Research Reactor (NSRR) in October, 1989. Test fuel rod used in the Test TS-1 was a short-sized BWR (7 x 7) type rod which was fabricated from a commercial rod provided from Tsuruga Unit 1 power reactor. The fuel had an initial enrichment of 2.79 % and burnup of 21.3 GWd/t (bundle average). Pulse irradiation was performed at a condition of stagnant water cooling, atmospheric pressure and ambient temperature using a newly developed double container-type capsule. Energy deposition of the rod in this test was evaluated to be about 61 cal/g·fuel (55 cal/g·fuel in peak fuel enthalpy) and no fuel failure was observed. Descriptions on test conditions, test procedures, fuel burnup measurements, transient behavior of the test rod during pulse irradiation and results of post pulse irradiation examinations are contained in this report. (author)

  15. An Experimental Study on Dynamics of a Novel Dual-Belt Continuous Variable Transmission Based on a Newly Developed Test Rig

    Directory of Open Access Journals (Sweden)

    Pak Kin Wong

    2015-01-01

    Full Text Available A novel dual-belt Van Doorne’s continuous variable transmission (DBVCVT system, which is applicable to heavy-duty vehicles, has been previously proposed by the authors in order to improve the low torque capacity of traditional single-belt CVT. This DBVCVT is a novel design among continuously variable transmissions and is necessary to be prototyped for experimental study, and the analytical dynamic model for this DBVCVT also needs to be experimentally validated. So, this work originally fabricated a prototype of DBVCVT and integrates this prototype to a light-load hardware-in-the-loop test rig by replacing the engine and load equipment with the AC motor and magnetic powder dynamometer. Moreover, with the use of this newly developed test rig, this work implements the experimental study of this DBVCVT for the first time. The comparison of experimental and simulation results validates the previously proposed analytical model for DBVCVT, and some basic characteristics of the DBVCVT in terms of the reliability, speed ratio, and transmission efficiency are also experimentally studied. In all, this developed test rig with the analytical model lays the foundation for further study on this novel DBVCVT.

  16. Change in geometrical parameters of WWER high burnup fuel rods under operational conditions and transient testing

    International Nuclear Information System (INIS)

    Kanashov, B.; Amosov, S.; Lyadov, G.; Markov, D.; Ovchinnikov, V; Polenok, V.; Smirnov, A.; Sukhikh, A.; Bek, E.; Yenin, A.; Novikov, V.

    2001-01-01

    The paper discusses changes in fuel rods geometric parameters as result of operation conditions and burnups. The degree of geometry variability of fuel rods, cladding and column is one of the most important characteristics affecting fuel serviceability. On the other hand, changes in fuel rod geometric parameters influence fuel temperature, fission gas release, fuel-to-cladding stress strained state as well as the degree of interaction with FA skeleton elements and skeleton rigidity. Change in fuel-to-cladding gap is measured using compression technique. The axial distribution of fuel-to-cladding gap demonstrates the largest decrease of the gap in the region 500 to 2000 mm from the bottom of the fuel rod (WWER-440) and in the region of 500 to 3000 mm for WWER-1000. The cladding material creep in WWER fuel rods together with the radiation growth results in fuel rod cladding elongation. A set of transient tests for spent WWER-440 and WWER-1000 fuel rods carried out in SSC RIAR during a period 1995-1999, with the aim to estimate the changes in geometric parameters of FRs. The estimation of changes in outer diameter of cladding and fuel column and fuel-to-cladding gap are performed in transient conditions (changes in linear power range of 180 to 400 W/cm) for both WWER-440 and WWER-1000. WWER-440 fuel rods having the same burnup and close fuel-cladding contact before testing are subjected to considerable hoop cladding strain in testing up to 300 W/cm. But the hoop strain does not grow due to the structural changes in fuel column and decrease in central hole diameter occurred when the power is higher

  17. Cone dystrophy with "supernormal" rod ERG: psychophysical testing shows comparable rod and cone temporal sensitivity losses with no gain in rod function.

    Science.gov (United States)

    Stockman, Andrew; Henning, G Bruce; Michaelides, Michel; Moore, Anthony T; Webster, Andrew R; Cammack, Jocelyn; Ripamonti, Caterina

    2014-02-10

    We report a psychophysical investigation of 5 observers with the retinal disorder "cone dystrophy with supernormal rod ERG," caused by mutations in the gene KCNV2 that encodes a voltage-gated potassium channel found in rod and cone photoreceptors. We compared losses for rod- and for cone-mediated vision to further investigate the disorder and to assess whether the supernormal ERG is associated with any visual benefit. L-cone, S-cone, and rod temporal acuity (critical flicker fusion frequency) were measured as a function of target irradiance; L-cone temporal contrast sensitivity was measured as a function of temporal frequency. Temporal acuity measures revealed that losses for vision mediated by rods, S-cones, and L-cones are roughly equivalent. Further, the gain in rod function implied by the supernormal ERG provides no apparent benefit to near-threshold rod-mediated visual performance. The L-cone temporal contrast sensitivity function in affected observers was similar in shape to the mean normal function but only after the mean function was compressed by halving the logarithmic sensitivities. The name of this disorder is potentially misleading because the comparable losses found across rod and cone vision suggest that the disorder is a generalized cone-rod dystrophy. Temporal acuity and temporal contrast sensitivity measures are broadly consistent with the defect in the voltage-gated potassium channel producing a nonlinear distortion of the photoreceptor response but after otherwise normal transduction processes.

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  19. The use of eddy current testing for nuclear fuel rods cladding evaluation

    International Nuclear Information System (INIS)

    Silva Junior, Silverio F. da; Alencar, Donizete A.; Brito, Mucio Jose D. de

    2007-01-01

    Nuclear fuel rods cladding must be tested after their manufacture and during their operational life. This paper describes a study about the use of eddy current test method as a nondestructive tool for nuclear fuel rods cladding evaluation. The experiments were carried out using two different probes: an external probe and an internal probe. The main goal was to verify the sensitivity of the eddy current test system, to develop calibration and reference standards and to establish the main capabilities and limitations presented by this test method for this application. (author)

  20. Multi-rod burst test under a loss-of coolant accident condition, (4)

    International Nuclear Information System (INIS)

    Otomo, Takashi; Hashimoto, Masao; Kawasaki, Satoru; Furuta, Teruo; Uetsuka, Hiroshi

    1983-06-01

    Multi-rod burst test of No.7808 bundle was performed in steam to estimate quantitative coolant flow channel restriction caused by the ballooning of zircaloy claddings in a fuel assembly during a LOCA transient in LWRs. The test was conducted under the condition that the initial internal pressure in each rod was 35kg/cm 2 (RT) and the heating rate was 9 0 C/s in steam with flow rate of 0.4g/cm 2 .min. The following results were obtained; (1) Maximum and burst pressures in rods were in the range 45 to 48kg/cm 2 and 41 to 45kg/cm 2 , respectively. The burst temperature of cladding were estimated to be 850 to 880 0 C. (2) Axial portions of tubes with greater than 34% strain were observed in the range 0 to 40mm in most rod. The mean length was 19mm in the bundle. (3) The degree of maximum increase in cross-sectional area is 54.2% in the bundle(7 x 7) and 66.9% in the internal rods(5 x 5). (4) Maximum channel area restriction was 40.5% in the bundle(7 x 7) and 51.4% in the internal rods(5 x 5). (author)

  1. The out-of-pile test for internal pressure measurement of nuclear fuel rod using LVDT

    Energy Technology Data Exchange (ETDEWEB)

    Min, Sohn Jae; Kang, Y. H.; Kim, B. G. [and others

    2001-11-01

    As a part of the development of instrumentation technologies for the nuclear fuel irradiation test in HANARO, the internal pressure measurement technique of the nuclear fuel rod is being developed using LVDT. The objectives of this test were to understand the LVDT's characteristics and to study its application techniques for fuel irradiation technology. It will be required to analyze the acquired internal pressure of fuel rod during fuel irradiation test in HANARO. The out-of-pile test system for pressure measurement was developed, and the test with the LVDT at room temperature(19 .deg. C) were performed. A out-of-pile test were implemented in 1 kg/cm{sup 2} increment from 1 kg/cm{sup 2} to 30 kg/cm{sup 2} and repeated 6 times at each condition. The LVDT's sensitivities were obtained by following two ways, the one by test and the other by calculation from characteristics data. These two sensitivities were compared and analyzed. The calculation method for internal pressure of nuclear fuel rod at specified temperature was also established. This report describes the system configuration, the out-of-pile test procedures, and the results. The results of the out-of-pile test will be used to predict accurately the internal pressure of fuel rod during irradiation test. And, the well qualified out-of-pile tests are needed to understand the LVDT's detail characteristics for the detail design of the fuel irradiation capsule.

  2. The out-of-pile test for internal pressure measurement of nuclear fuel rod using LVDT

    Energy Technology Data Exchange (ETDEWEB)

    Son, J. M.; Kim, B. K.; Kim, D. S.; Joo, K. N.; Park, S. J.; Kang, Y. H.; Kim, Y. K.; Yeum, K. I. [KAERI, Taejon (Korea, Republic of)

    2002-05-01

    As a part of the development of instrumentation technologies for the nuclear fuel irradiation test in HANARO(High-flux Advanced Nuclear Application Reactor), the internal pressure measurement technique of the nuclear fuel rod is being developed using LVDT(Linear Variable Differential Transformer). The objectives of this test were to understand the LVDT's characteristics and to study its application techniques for fuel irradiation technology. It will be required to analyze the acquired internal pressure of fuel rod during fuel irradiation test in HANARO. Therefore, the out of pile test system for pressure measurement was developed, and the test with the LVDT at room temperature were performed. This test were implemented in 1 kg/cm{sup 2} increment from 1 kg/cm{sup 2} to 30 kg/cm{sup 2}, and repeated 6 times at same condition. The LVDT's sensitivities were obtained by following two ways, the one by test and the other by calculation from characteristics data. These two sensitivities were compared and analyzed. The calculation method for internal pressure of nuclear fuel rod at specified temperature was also established. The results of the out-of-pile test will be used to predict accurately the internal pressure of fuel rod during irradiation test. And, the well qualified out-of-pile tests are needed to understand the LVDT's detail characteristics at high temperature for the detail design of the fuel irradiation capsule.

  3. Fretting wear characteristic tests of X2-GEN midgrid for SMART under a FIV rod trace

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young Ho; Lee, Kang Hee; Kim, Jae Yong; Kim, Hyung Kyu [KAERI, Daejeon (Korea, Republic of)

    2011-12-15

    The KEPCO Nuclear Fuel Co. requested the fretting wear characteristic tests of a X2-GEN midgrid under a FIV rod trace at room temperature air. The following results were obtained for the fretting wear test. {center_dot} Fretting wear tests under a FIV rod trace Based on the result of the fretting wear tests of the X2-GEN and 17ACE7 1x1 mid-grid under a FIV rod trace, X2-GEN mid-grid showed a slightly severe wear volume rather than 17ACE7 spring. But, maximum wear depth shows an opposite behavior. This is due to spring shape effect. The fretting wear mechanisms at each mid-grid were influenced by each spring shape, that are depended on the different impacting behavior under a FIV rod motion. Up to 5x105 cycles, wear characteristics of each mid-grid shows a relatively similar wear rate. Consequently, it is necessary to further study for examining exact fretting wear behavior under a FIV rod tra

  4. SEFLEX - fuel rod simulator effects in flooding experiments. Pt. 2

    International Nuclear Information System (INIS)

    Ihle, P.; Rust, K.

    1986-03-01

    This report presents typical data and a limited heat transfer analysis from unblocked bundle reflood tests of an experimental thermal-hydraulic program. Full-length bundles of 5 x 5 fuel rod simulators having a gas-filled gap between the Zy cladding and the alumina pellets were tested in the test rig designed for the earlier Flooding Experiments with Blocked Arrays (FEBA-program). The 5 x 5 FEBA rod bundle tests were performed with gapless heater rods. These rods have a close thermal contact between the stainless steel cladding and the electric insulation material. A comparison of the SEFLEX data with the reference data of FEBA obtained under identical initial and reflood conditions shows the influence of different fuel rod simulators on the thermal-hydraulic behavior during forced feed bottom reflooding of unblocked and blocked arrays. Compared to bundles of gapless rods, bundles of rods with Zy claddings and a gas filled gap between claddings and pellets, which more closely represent the features that exist in an actual fuel rod geometry, produced higher quench front velocities, enhanced removal of stored heat in the rods, reduced peak cladding temperatures, increased grid spacer effects and absolutely unproblematic coolability of 90 percent blockages with bypass. The data offer the opportunity for further validation of computer codes to make realistic predictions of safety margins during a LOCA in a PWR. (orig./HP) [de

  5. SEFLEX fuel rod simulator effects in flooding experiments. Pt. 3

    International Nuclear Information System (INIS)

    Ihle, P.; Rust, K.

    1986-03-01

    This report presents typical data and a limited heat transfer analysis from blocked bundle reflood tests of an experimental thermal-hydraulic program. Full-length bundles of 5x5 fuel rod simulators having a gas-filled gap between the Zy cladding and the alumina pellets were tested in the test rig designed for the earlier Flooding Experiments with Blocked Arrays (FEBA-program). The 5x5 FEBA rod bundle tests were performed with gapless heater rods. These rods have a close thermal contact between the stainless steel cladding and the electric insulation material. A comparison of the SEFLEX data with the reference data of FEBA obtained under identical initial and reflood conditions shows the influence of different fuel rod simulators on the thermal-hydraulic behavior during forced feed bottom reflooding of unblocked and blocked arrays. Compared to bundles of gapless rods, bundles of rods with Zy claddings and a gas filled gap between claddings and pellets, which more closely represent the features that exist in an actual fuel rod geometry, produced higher quench front velocities, enhanced removal of stored heat in the rods, reduced peak cladding temperatures, increased grid spacer effects and absolutely unproblematic coolability of 90 percent blockages with bypass. The data offer the opportunity for further validation of computer codes to make realistic predictions of safety margins during a LOCA in a PWR. (orig./HP) [de

  6. New design head pendulum test rig for window airbag development; Neu entwickelte Kopfpendel-Testanlage fuer die Window-Airbag -Auslegung

    Energy Technology Data Exchange (ETDEWEB)

    Grundheber, C. [Adam Opel AG, Ruesselsheim (Germany); Lindstromberg, M. [Siemens Restraint Systems, Alzenau (Germany)

    2001-07-01

    Comparisons between vehicle crash tests and pendulum tests demonstrate good correlation. The head pendulum rig presents a faster, more economical alternative to the usual sled facilities. This is above all due to the simpler test set-up, which only considers the body parts relevant to the window airbag development, i.e. dummy head, neck and shoulder. The tuning of the system to new baseline values can also be carried out very quickly, by specific adjustments to the spring-damper arrangement. (orig.)

  7. NAPP liquid shutoff rod system design, development, testing and precommissioning feed back study

    International Nuclear Information System (INIS)

    Soni, K.L.; Kaushik, R.V.; Mahajan, S.V.

    1989-01-01

    The development testing of a liquid poison shutoff rod system has enabled the evolution of a proven and acceptable design of the secondary shutdown system for 235 MWe standardised Pressurised Heavy Water Reactors (PHWRs). The availability of a full scale test loop is also proving for checking and qualifying the various suggestions for online improvements in the system. (author)

  8. Modeling and Experimental Tests on the Hydraulically Driven Control Rod option for IRIS Reactor

    International Nuclear Information System (INIS)

    Cammi, Antonio; Ricotti, Marco E.; Vitulo, Alessia

    2004-01-01

    The adoption of Internal Control Rod Drive Mechanisms (ICRDMs) represents a valuable alternative to classical, external CRDMs based on electro-magnetic devices, as adopted in current PWRs. The advantages on the safety features of the reactor are apparent: inherent elimination of the Rod Ejection accidents and of possible concerns about the vessel head penetrations. A further positive feedback on the design is the reduction of the primary system overall dimensions. Within the frame of the ICRDM concepts, the Hydraulically Driven Control Rod solution is investigated as a possible option for the IRIS integral reactor. After a brief comparison of the solutions currently proposed for integral reactors, the configuration of the Hydraulic Control Rod device for IRIS, made up by an external movable piston and an internal fixed cylinder, is described. A description of the whole control system is reported as well. Particular attention is devoted to the Control Rod profile characterization, performed by means of a Computational Fluid Dynamics (CFD) analysis. The investigation of the system behavior has been carried out, including the dynamic equilibrium and its stability properties, the withdrawal and insertion step movement and the sensitivity study on command time periods. A suitable dynamic model has been set up for the mentioned purposes: the models corresponding to the various Control Rod system devices have been written in an Object-Oriented language (Modelica), thus allowing an easy implementation of such a system into the simulator for the whole reactor. Finally, a preliminary low pressure, low temperature, reduced length experimental facility has been built. Tests on HDCR stability and operational transients have been performed. The results are compared with the dynamic system model and CFD simulation model, showing good agreement between simulations and experimental data. During these preliminary tests, the control system performed correctly, allowing stable dynamic

  9. Development of techniques for joining fuel rod simulators to test assemblies

    International Nuclear Information System (INIS)

    Moorhead, A.J.; Reed, R.W.

    1980-01-01

    A unique tubular electrode carrier is described for gas tungsten-arc welding small-diameter nuclear fuel rod simulators to the tubesheet of a test assembly. Both the close-packed geometry of the array of simulators and the extension of coaxial electrical conductors from each simulator hindered access to the weld joint. Consequently, a conventional gas tungsten-arc torch could not be used. Two seven-rod assemblies that were mockups of the simulator-to-tubesheet joint area were welded and successfully tested. Modified versions of the electrode carrier for brazing electrical leads to the upper ends of the fuel pin simulators are also described. Satisfactory brazes have been made on both single-rod mockups and an array of 25 simulators by using the modified electrode carrier and a filler metal with a composition of 71.5 Ag-28 Cu-0.5 Ni

  10. Hydraulically driven control rod concept for integral reactors: fluid dynamic simulation and preliminary test

    International Nuclear Information System (INIS)

    Ricotti, M.E.; Cammi, A.; Lombardi, C.; Passoni, M.; Rizzo, C.; Carelli, M.; Colombo, E.

    2003-01-01

    The paper deals with the preliminary study of the Hydraulically Driven Control Rod concept, tailored for PWR control rods (spider type) with hydraulic drive mechanism completely immersed in the primary water. A specific solution suitable for advanced versions of the IRIS integral reactor is under investigation. The configuration of the Hydraulic Control Rod device, made up by an external movable piston and an internal fixed cylinder, is described. After a brief description of the whole control system, particular attention is devoted to the Control Rod characterization via Computational Fluid Dynamics (CFD) analysis. The investigation of the system behavior, including dynamic equilibrium and stability properties, has been carried out. Finally, preliminary tests were performed in a low pressure, low temperature, reduced length experimental facility. The results are compared with the dynamic control model and CFD simulation model, showing good agreement between simulations and experimental data. During these preliminary tests, the control system performs correctly, allowing stable dynamic equilibrium positions for the Control Rod and stable behavior during withdrawal and insertion steps. (author)

  11. A New Design of the Universal Test Rig to Measure the Wear Characterizations of Polymer Acetal Gears (Spur, Helical, Bevel, and Worm

    Directory of Open Access Journals (Sweden)

    Samy Yousef

    2015-01-01

    Full Text Available This work aims to study the wear characterization of common types of acetal polymer gears (spur, helical, bevel, and worm using a new TS universal test rig, in order to obtain reliable results and as a reference when compared with acetal nanocomposite gears later. The TS universal test rig consists of three different units that are connected by a main driver shaft and a pair of constantly meshing metal spur gears, which transfer power to the bevel and worm test units. The first unit is used to test the bevel gears, the second unit is used to test the spur and helical gears, and the third unit is used to test the worm gears. The loading mechanism is similarly designed to block the brake mechanism. Hobbing and milling machines were used to machine an injection-moulded polymer flanges and produce the tested gears. All gear pairs, except the worm gear, have identical gear ratios. The experiments were performed at speed 1420 rpm and the torque was 4 Nm. The results showed that the wear rates (in the form of weight loss of spur gears were consistent with the previous results and the other gear types had larger wear rates.

  12. Thermal analysis of lithium cooled natural circulation loop module for fuel rod testing in the Fast Flux Test Facility

    International Nuclear Information System (INIS)

    Eyler, L.L.; Kim, D.; Stover, R.L.; Beaver, T.R.

    1987-01-01

    Maximum heat removal capability of a lithium cooled natural circulation fuel rod test module design is determined. Loop geometry is optimized within limitations of design specifications for nominal operation temperatures, materials, and test module environment. Results provide test module operation limits and range of potential uncertainties. 3 refs., 12 figs

  13. Post-test examination of the VVER-1000 fuel rod bundle CORA-W2

    Energy Technology Data Exchange (ETDEWEB)

    Hofmann, P.; Noack, V.; Burbach, J.; Metzger, H.; Schanz, G.; Hagen, S.; Sepold, L.

    1995-08-01

    The upper half of the bundle is completely oxidized, the lower half has kept the fuel rods relatively intact. The post-test examination results show the strong impact of the B{sub 4}C absorber rod and the stainless steel grid spacers on the `low-temperature` bundle damage initiation and progression. The B{sub 4}C absorber rod completely disappeared in the upper half of the bundle. The multicomponent melts relocated and formed coolant channel blockages on solidification with a maximum extent of about 30% in the lower part of the bundle. At temperatures above the melting point of the ZrNb1 cladding extensive fuel dissolution occured. (orig./HP)

  14. CODEX-B4C experiment. Core degradation test with boron carbide control rod

    International Nuclear Information System (INIS)

    Hozer, Z.; Nagy, I.; Windberg, P.; Balasko, M.; Matus, L.; Prokopiev, O.; Pinter, A.; Horvath, M.; Gyenes, Gy.; Czitrovszky, A.; Nagy, A.; Jani, P.

    2003-11-01

    The CODEX-B4C bundle test has been successfully performed on 25 th May 2001 in the framework of the COLOSS project of the EU 5 th FWP. The high temperature degradation of a VVER-1000 type bundle with B 4 C control rod was investigated with electrically heated fuel rods. The experiment was carried out according to a scenario selected in favour of methane formation. Degradation of control rod and fuel bundle took place at temperatures ∼2000 deg C, cooling down of the bundle was performed in steam atmosphere. The gas composition measurement indicated no methane production during the experiment. High release of aerosols was detected in the high temperature oxidation phase. The on-line measured data are collected into a database and are available for code validation and development. (author)

  15. CODEX-B4C experiment. Core degradation test with boron carbide control rod

    Energy Technology Data Exchange (ETDEWEB)

    Hozer, Z; Nagy, I; Windberg, P; Balasko, M; Matus, L; Prokopiev, O; Pinter, A; Horvath, M; Gyenes, Gy [KFKI Atomic Energy Research Institute, Budapest (Hungary); Czitrovszky, A; Nagy, A; Jani, P [Research Institute for Solid State Physics and Optics, Budapest (Hungary)

    2003-11-01

    The CODEX-B4C bundle test has been successfully performed on 25{sup th} May 2001 in the framework of the COLOSS project of the EU 5{sup th} FWP. The high temperature degradation of a VVER-1000 type bundle with B{sub 4}C control rod was investigated with electrically heated fuel rods. The experiment was carried out according to a scenario selected in favour of methane formation. Degradation of control rod and fuel bundle took place at temperatures {approx}2000 deg C, cooling down of the bundle was performed in steam atmosphere. The gas composition measurement indicated no methane production during the experiment. High release of aerosols was detected in the high temperature oxidation phase. The on-line measured data are collected into a database and are available for code validation and development. (author)

  16. AgInCd control rod failure in the QUENCH-13 bundle test

    International Nuclear Information System (INIS)

    Sepold, L.; Lind, T.; Csordas, A. Pinter; Stegmaier, U.; Steinbrueck, M.; Stuckert, J.

    2009-01-01

    The QUENCH off-pile experiments performed at the Karlsruhe Research Center are to investigate the high-temperature behavior of Light Water Reactor (LWR) core materials under transient conditions and in particular the hydrogen source term resulting from the water injection into an uncovered LWR core. The typical LWR-type QUENCH test bundle, which is electrically heated, consists of 21 fuel rod simulators with a total length of approximately 2.5 m. The Zircaloy-4 rod claddings and the grid spacers are identical to those used in Pressurized Water Reactors (PWR) whereas the fuel is represented by ZrO 2 pellets. In the QUENCH-13 experiment the single unheated fuel rod simulator in the center of the test bundle was replaced by a PWR-type control rod. The QUENCH-13 experiment consisting of pre-oxidation, transient, and quench water injection at the bottom of the test section investigated the effect of an AgInCd/stainless steel/Zircaloy-4 control rod assembly on early-phase bundle degradation and on reflood behavior. Furthermore, in the frame of the EU 6th Framework Network of Excellence SARNET, release and transport of aerosols of a failed absorber rod were to be studied in QUENCH-13, which was accomplished with help of aerosol measurements performed by PSI-Switzerland and AEKI-Hungary. Control rod failure was initiated by eutectic interaction of steel cladding and Zircaloy-4 guide tube and was indicated at about 1415 K by axial peak absorber and bundle temperature responses and additionally by the on-line aerosol monitoring system. Significant releases of aerosols and melt relocation from the control rod were observed at an axial peak bundle temperature of 1650 K. At a maximum bundle temperature of 1820 K reflood from the bottom was initiated with cold water at a flooding rate of 52 g/s. There was no noticeable temperature escalation during quenching. This corresponds to the small amount of about 1 g in hydrogen production during the quench phase (compared to 42 g of H 2

  17. Mechanical tests of the bolt of the gripper latch on the control rod cluster

    International Nuclear Information System (INIS)

    Lemaire, E.; Couet, D.; Molinie, D.; Grandjean, Y.; Radat, M.P.; Guttmann, D.

    1998-01-01

    Failure analysis and mechanical testing indicate that control rod drive mechanisms malfunctioning by 1995-96 is due to rupture by fatigue of a bolt inside the stationary gripper assembly. Fatigue is enhanced by free working following surface adaptation and unscrewing of the assembly. These results are taken into account for the choice of a new anti-rotation device. (authors)

  18. Ultrasonic Nondestructive Evaluation of Pultruded Rod Stitched Efficient Unitized Structure (PRSEUS) During Large-Scale Load Testing and Rod Push-Out Testing

    Science.gov (United States)

    Johnston, Patrick H.; Juarez, Peter D.

    2016-01-01

    The Pultruded Rod Stitched Efficient Unitized Structure (PRSEUS) is a structural concept developed by the Boeing Company to address the complex structural design aspects associated with a pressurized hybrid wing body (HWB) aircraft configuration. The HWB has long been a focus of NASA's environmentally responsible aviation (ERA) project, following a building block approach to structures development, culminating with the testing of a nearly full-scale multi-bay box (MBB), representing a segment of the pressurized, non-circular fuselage portion of the HWB. PRSEUS is an integral structural concept wherein skins, frames, stringers and tear straps made of variable number of layers of dry warp-knit carbon-fiber stacks are stitched together, then resin-infused and cured in an out-of-autoclave process. The PRSEUS concept has the potential for reducing the weight and cost and increasing the structural efficiency of transport aircraft structures. A key feature of PRSEUS is the damage-arresting nature of the stitches, which enables the use of fail-safe design principles. During the load testing of the MBB, ultrasonic nondestructive evaluation (NDE) was used to monitor several sites of intentional barely-visible impact damage (BVID) as well as to survey the areas surrounding the failure cracks after final loading to catastrophic failure. The damage-arresting ability of PRSEUS was confirmed by the results of NDE. In parallel with the large-scale structural testing of the MBB, mechanical tests were conducted of the PRSEUS rod-to-overwrap bonds, as measured by pushing the rod axially from a short length of stringer.

  19. Full scale mock-up tests for rod bundle thermal-hydraulics in Japan

    International Nuclear Information System (INIS)

    Sugawara, S.

    1995-01-01

    This poster describes tests aimed at development and validation of principal design methodology of rod bundle thermal-hydraulics correlations. The works are based on domestic data base using the full-scale mock-up test facilities. The scope of the tests comprises DNB heat flux, transient DNB heat flux, post DNB heat transfer, pressure drop and void distribution. The works have been performed under collaboration among electric facilities, NPP vendors, universities, governmental corporations. 1 tab., 14 figs

  20. Experiments on the quench behavior of fuel rods

    International Nuclear Information System (INIS)

    Hofmann, P.; Noack, V.; Burbach, J.; Metzger, H.

    1995-01-01

    Because of the importance of the observed reflood phenomena for safety of current and future LWRs, the Forschungszentrum Karlsruhe (FZKA) started a program to investigate the mechanisms of quench-induced oxidation of Zircaloy. A small scale test-rig was designed and built in which it is possible to quench single Zircaloy rods by water and steam. The report describes the status of this work in May 1995. Some experimental results are presented. (orig./HP)

  1. Experiments on the quench behavior of fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Hofmann, P.; Noack, V.; Burbach, J.; Metzger, H.

    1995-08-01

    Because of the importance of the observed reflood phenomena for safety of current and future LWRs, the Forschungszentrum Karlsruhe (FZKA) started a program to investigate the mechanisms of quench-induced oxidation of Zircaloy. A small scale test-rig was designed and built in which it is possible to quench single Zircaloy rods by water and steam. The report describes the status of this work in May 1995. Some experimental results are presented. (orig./HP)

  2. Joint test rig for testing and calibrating of different methods of two-phase mass flow measurement

    International Nuclear Information System (INIS)

    Reimann, J.; Demski, A.; Hahn, H.; Harten, U.; John, H.; Megerle, A.; Mueller, S.; Pawlak, L.; Wanner, E.

    1977-01-01

    The steam-water loop was completed by building in two throttling valves upstream of the mixing chamber. By producing steam by throttling the total mass flow may be increased up to 35% compared to the former method of operating the loop. Furthermore, throttling stabilizes the single phase mass flow measurement. The data aquisition system and computation of the reference values has been finished. The computer program contains the equations of state of steam/water and the calibration curves for all signal transducers. The 5 beam γ-densitometer has been finished mechanically and supplied with the electronics. First calibration tests are fully satisfactory. The instrumentation of the air-water loop completed. At low quality the mass fluxes are increased by a factor of 5 compared with the steam-water-loop. The regime of dispersed bubble flow is fully reached in the test section. To detect flow regimes air-water as well as in steam-water flow, a local impedance probe was used. In addition, the phase distribution across the channel could be detected by traversing the probe. The boundaries of the air-water flow regimes detected by the probe are in good correspondance with other investigations. For the first time, such experiments have been carried out in horizontal steam-water flow. The results indicate that the region of slug flow becomes smaller with increasing pressure. (orig./RW) [de

  3. Procedure for vibration test of the fuel rod supported by spacer grids

    International Nuclear Information System (INIS)

    Choi, Myoung Hwan; Kang, Heung Seok; Yoon, Kyung Ho; Kim, Hyung Kyu; Song, Kee Nam

    2002-07-01

    One of the methods that are used to compare and verify the supporting performance of the spacer grids developed is the vibration characteristic test. In this report there are two aims. One is of the understand of the experimental method and procedure performing the modal testing using I-DEAS TDAS module. The other is the investigation of the vibration behaviors of a dummy fuel rod supported by 8 optimized H type spacer grids. This report describes the method and procedure of modal testing to obtain the vibration characteristics such as amplitudes, natural frequencies and mode shapes of the fuel rod using a shaker, a non-contact gap sensor and an accelerometer. This report provides a test procedure in detail so that anyone can be easily understood and use the I-DEAS TDAS program. The I-DEAS TDAS program related to the modal testing has several tasks including the Modal analysis, Signal Processing et al.. This report includes model preparation to prepare the geometrical model, Signal Processing (Sine/Standard measurement) to acquire the signal, Modal analysis to obtain the frequencies and mode shapes, Correlation to analyze the relation between the test and FE analysis and Post Processing tasks. In addition, this report contains the actual test and analysis data of a dummy fuel rod in length 3847mm supported by 8 optimized H type spacer grids

  4. Investigation of friction in rectangular Nitrile-Butadiene Rubber (NBR) hydraulic rod seals for defence applications

    Energy Technology Data Exchange (ETDEWEB)

    Bhaumik, Shankar; Guruprasad, S.; Bhandari, P. [R and DE , Dighi (India); Kumaraswamy, A. [Defence Institute of Advanced Technology, Girinagar (India)

    2015-11-15

    Contact based FE simulations have been carried out to estimate the contact pressure distribution at seal/rod interface at sealed oil pressures of 10, 20 and 30 MPa and constant rod velocity of 0.12 m/s. Oil film thickness at the interface was then computed analytically at various combinations of oil pressures and rod velocities. Seal contact pressure and oil film thickness data along with surface roughness, intermolecular interaction between seal/rod interfaces has been perused to estimate the friction in Nitrile-Butadiene Rubber (NBR) rectangular hydraulic rod seals using theoretical models such as Inverse hydrodynamic lubrication (IHL), Greenwood-Williamson (GW) and Wassink's models. The friction at seal/rod interface was also measured experimentally using a specially designed test rig. The comparison of theoretical and experimental data revealed that, friction computed from GW and Wassink's models had good agreement with the experimental results.

  5. Recent High Heat Flux Tests on W-Rod-Armored Mockups

    International Nuclear Information System (INIS)

    Nygren, Richard E.; Youchison, Dennis L.; McDonald, Jimmie M.; Lutz, Thomas J.; Miszkiel, Mark E.

    2000-01-01

    In the authors initial high heat flux tests on small mockups armored with W rods, done in the small electron beam facility (EBTS) at Sandia National Laboratories, the mockups exhibited excellent thermal performance. However, to reach high heat fluxes, they reduced the heated area to only a portion (approximately25%) of the sample. They have now begun tests in their larger electron beam facility, EB 1200, where the available power (1.2 MW) is more than enough to heat the entire surface area of the small mockups. The initial results indicate that, at a given power, the surface temperatures of rods in the EB 1200 tests is somewhat higher than was observed in the EBTS tests. Also, it appears that one mockup (PW-10) has higher surface temperatures than other mockups with similar height (10mm) W rods, and that the previously reported values of absorbed heat flux on this mockup were too high. In the tests in EB 1200 of a second mockup, PW-4, absorbed heat fluxes of approximately22MW/m 2 were reached but the corresponding surface temperatures were somewhat higher than in EBTS. A further conclusion is that the simple 1-D model initially used in evaluating some of the results from the EBTS testing was not adequate, and 3-D thermal modeling will be needed to interpret the results

  6. FY15 Status Report: CIRFT Testing of Spent Nuclear Fuel Rods from Boiler Water Reactor Limerick

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wang, Hong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jiang, Hao [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-06-01

    The objective of this project is to perform a systematic study of used nuclear fuel (UNF, also known as spent nuclear fuel [SNF]) integrity under simulated transportation environments using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL) in August 2013. Under Nuclear Regulatory Commission (NRC) sponsorship, ORNL completed four benchmark tests, four static tests, and twelve dynamic or cycle tests on H. B. Robinson (HBR) high burn-up (HBU) fuel. The clad of the HBR fuels was made of Zircaloy-4. Testing was continued in fiscal year (FY) 2014 using Department of Energy (DOE) funds. The additional CIRFT was conducted on three HBR rods (R3, R4, and R5) in which two specimens failed and one specimen was tested to over 2.23 10⁷ cycles without failing. The data analysis on all the HBR UNF rods demonstrated that it is necessary to characterize the fatigue life of the UNF rods in terms of (1) the curvature amplitude and (2) the maximum absolute of curvature extremes. The maximum extremes are significant because they signify the maximum of tensile stress for the outer fiber of the bending rod. CIRFT testing has also addressed a large variation in hydrogen content on the HBR rods. While the load amplitude is the dominant factor that controls the fatigue life of bending rods, the hydrogen content also has an important effect on the lifetime attained at each load range tested. In FY 15, ten SNF rod segments from BWR Limerick were tested using ORNL CIRFT, with one under static and nine dynamic loading conditions. Under static unidirectional loading, a moment of 85 N·m was obtained at maximum curvature 4.0 m⁻¹. The specimen did not show any sign of failure in three repeated loading cycles to almost same maximum curvature. Ten cyclic tests were conducted with amplitude varying from 15.2 to 7.1 N·m. Failure was observed in nine of the tested rod specimens. The cycles to failure were

  7. Results of the first nuclear blowdown test on single fuel rods (LOC-11 Series in PBF)

    International Nuclear Information System (INIS)

    Larson, J.R.; Evans, D.R.; McCardell, R.K.

    1978-01-01

    This paper presents results of the first nuclear blowdown tests (LOC-11A, LOC-11B, LOC-11C) ever conducted. The Loss-of-Coolant Accident (LOCA) Test Series is being conducted in the Power Burst Facility (PBF) reactor at the Idaho National Engineering Laboratory, near Idaho Falls, Idaho, for the Nuclear Regulatory Commission. The objective of the LOC-11 tests was to obtain data on the behavior of pressurized and unpressurized rods when exposed to a blowdown similar to that expected in a pressurized water reactor (PWR) during a hypothesized double-ended cold-leg break. The data are being used for the development and verification of analytical models that are used to predict coolant and fuel rod pressure during a LOCA in a PWR

  8. Demonstration test of the spent fuel rod cutting process with tube cutter mechanism

    International Nuclear Information System (INIS)

    Lee, Jong Youl; Jung, Jae Hoo; Hong, Dong Hee; Yoon, Ji Sup; Lee, Eun Pyo

    2001-03-01

    In this paper, the verification by computer graphics technology for the spent fuel rod cutting devise which belongs to the spent fuel disassembly processes, the performance tests of the real device, and the demonstration tests with tube cutter mechanism are described. The graphical design system is used throughout the design stages from conceptual design to motion analysis like collision detection. By using this system, the device and the process are optimized. The performance test of the real device and the demonstration test using the tube cutter mechanism in the hot cell are carried out. From these results, the spent fuel rod cutting device is improved based on the considerations of circularity of the rod cross-section, debris generation, and fire risk etc. Also, this device is improved to be operated automatically via remote control system considering later use in closed environment like Hot-cell (radioactive area) and the modulization in the structure of this device makes maintenance easy. The result of the performance test and the demonstration in this report is expected to contribute to the optimization of the pre-treatment processes for the reuse of the spent fuel like DUPIC process and the final disposal

  9. A rigged market

    International Nuclear Information System (INIS)

    Thomas, M.

    2000-01-01

    The mobile rig market remains a unique sector of the global upstream oil and gas industry. Big oil is continuing to emerge blinking from the darkness of its recent cash-starved existence to bask in the glory of a resurgent oil price. But the rig sector is once again lagging behind the pace being set by operators as they open up their wallets for new or delayed exploration and production projects. This paper gives statistics on worldwide count and contracts

  10. Drilling rig mast

    Energy Technology Data Exchange (ETDEWEB)

    Bulgakov, E.S.; Barashkov, V.A.; Lebedev, A.I.; Panin, N.M.; Sirotkin, N.V.

    1981-01-07

    A drilling rig mast is proposed that contains a portal with a carrier shaft hinged to it and struts with stays. In order to decrease the time expended in the assembly and dessembly of the drilling rig, the portal is constructed from mobile and immobile parts that are connected together by a ball pivot; the immobile section of the portal has a T-shaped recess for directing the mobile section.

  11. Development and testing of improved polyimide actuator rod seals at higher temperatures for use in advanced aircraft hydraulic systems

    Science.gov (United States)

    Robinson, E. D.; Waterman, A. W.; Nelson, W. G.

    1972-01-01

    Polyimide second stage rod seals were evaluated to determine their suitability for application in advanced aircraft systems. The configurations of the seals are described. The conditions of the life cycle tests are provided. It was determined that external rod seal leakage was within prescribed limits and that the seals showed no signs of structural degradation.

  12. Control rod position and temperature coefficients in HTTR power-rise tests. Interim report

    International Nuclear Information System (INIS)

    Fujimoto, Nozomu; Nojiri, Naoki; Takada, Eiji; Saito, Kenji; Kobayashi, Shoichi; Sawahata, Hiroaki; Kokusen, Sigeru

    2001-03-01

    Power-rise tests of the High Temperature Engineering Test Reactor (HTTR) have been carried out aiming to achieve 100% power. So far, 50% of power operation and many tests have been carried out. In the HTTR, temperature change in core is so large to achieve the outlet coolant temperature of 950degC. To improve the calculation accuracy of the HTTR reactor physics characteristics, control rod positions at criticality and temperature coefficients were measured at each step to achieve 50% power level. The calculations were carried out using Monte Carlo code and diffusion theory with temperature distributions in the core obtained by reciprocal calculation of thermo-hydraulic code and diffusion theory. Control rod positions and temperature coefficients were calculated by diffusion theory and Monte Carlo method. The test results were compared to calculation results. The control rod positions at criticality showed good agreement with calculation results by Monte Carlo method with error of 50 mm. The control position at criticality at 100% was predicted around 2900mm. Temperature coefficients showed good agreement with calculation results by diffusion theory. The improvement of calculation will be carried out comparing the measured results up to 100% power level. (author)

  13. Fuel rod failure during film boiling (PCM-1 test in the PBF)

    International Nuclear Information System (INIS)

    Domenico, W.F.; Stanley, C.J.; Mehner, A.S.

    1978-01-01

    The Power-Cooling-Mismatch (PCM) Test, PCM-1 was conducted in the Power Burst Facility (PFB) in March of 1978. The PCM Test Series is being conducted at the Idaho National Engineering Laboratory by EG and G Idaho, Inc., under contract to the USNRC and is designed to characterize the behavior of nuclear fuel rods operating under conditions of high power or low coolant flow or both leading to departure from nucleate boiling. The PCM-1 test was performed to provide in-pile data for a ''worst case'' PCM incident. The objective of this experiment was to study the behavior of a single pressurized water reactor (PWR) fuel rod subjected to a high-power and low flow environment which would result in cladding failure at full power. The ''worst case'' conditions established for the experiment consisted of a rod peak power of 78.7 kW/m and a coolant mass flux of 1356 kg/s.m 2 . Fuel temperatures at the stipulated operating conditions were such that a significant volume of molten fuel was present when failure occurred which produced a high probability of molten fuel-coolant interaction (MFCI) with the possibility of a vapor explosion

  14. Overview of Japanese control rods development program

    International Nuclear Information System (INIS)

    Koyama, M.

    1984-01-01

    The Japanese control rods development program was established based on the fast breeder reactor program. Therefore, PNC's efforts have been made mainly for the development of analysis, design and fabrication technologies for ''JOYO'' and ''MONJU'' control rods. Laboratory studies were performed to obtain the information for absorber materials. The design and fabrication of the sealed and vented type control rod pins were completed, and water loop tests and in-sodium tests were carried out. Irradiation behavior of enriched B 4 C pellets with low and high density in DFR was examined. Japan's experimental fast reactor, JOYO, has been operated at the rated power of 50MWt and 75MWt since April 1977 when the MK-I core (breeder core) attained initial criticality. Post irradiation examinations on control rod, removed from the reactor, were carried out and their performance behavior were evaluated. In the MK-II core, a control rods monitoring program has been in investigation. Absorber Materials Irradiation Rigs (AMIR) are scheduled to be loaded and irradiated in the JOYO MK-II core from 1984. (author)

  15. FEM simulation of friction testing method based on combined forward rod-backward can extrusion

    DEFF Research Database (Denmark)

    Nakamura, T; Bay, Niels; Zhang, Z. L

    1997-01-01

    A new friction testing method by combined forward rod-backward can extrusion is proposed in order to evaluate frictional characteristics of lubricants in forging processes. By this method the friction coefficient mu and the friction factor m can be estimated along the container wall and the conical...... curves are obtained by rigid-plastic FEM simulations in a combined forward rod-backward can extrusion process for a reduction in area R-b = 25, 50 and 70 percent in the backward can extrusion. It is confirmed that the friction factor m(p) on the punch nose in the backward cart extrusion has almost...... in a mechanical press with aluminium alloy A6061 as the workpiece material and different kinds of lubricants. They confirm the analysis resulting in reasonable values for the friction coefficient and the friction factor....

  16. Development of oxygen sensing technology in an irradiated fuel rod. Characteristic test of oxygen sensor

    International Nuclear Information System (INIS)

    Saito, Junichi; Hoshiya, Taiji; Sakurai, Fumio; Sakai, Haruyuki

    1996-03-01

    At the Department of JMTR (Japan Materials Test Reactor), the re-instrumentation technologies to a high burnup fuel rod irradiated in an LWR have been developed to study irradiation behavior of the fuel during power transient. It has been progressed developing a chemical sensor as one of the re-instrumentation technologies. This report summarizes the results of characteristic tests of an oxygen sensor made of Yttria Stabilized Zirconia (YSZ) as a solid electrolyte. Several kinds of experiments were carried out to evaluate the electromotive force (emf) performance, stability and lifetime of the oxygen sensor with Ni/NiO, Cr/Cr 2 O 3 and Fe/FeO, respectively as a reference electrode. From the experimental data, it is suggested that the reference electrode of Ni/NiO reveals the most appropriate characteristic of the sensor to measure the partial oxygen pressure in a fuel rod. It is the final goal of this development to clarify the change of oxygen chemical potential in a fuel rod during power transient. (author)

  17. POST-IRRADIATION ANALYSES OF U-MO DISPERSION FUEL RODS OF KOMO TESTS AT HANARO

    Directory of Open Access Journals (Sweden)

    H.J. RYU

    2013-12-01

    Full Text Available Since 2001, a series of five irradiation test campaigns for atomized U-Mo dispersion fuel rods, KOMO-1, -2, -3, -4, and -5, has been conducted at HANARO (Korea in order to develop high performance low enriched uranium dispersion fuel for research reactors. The KOMO irradiation tests provided valuable information on the irradiation behavior of U-Mo fuel that results from the distinct fuel design and irradiation conditions of the rod fuel for HANARO. Full size U-Mo dispersion fuel rods of 4–5 g-U/cm3 were irradiated at a maximum linear power of approximately 105 kW/m up to 85% of the initial U-235 depletion burnup without breakaway swelling or fuel cladding failure. Electron probe microanalyses of the irradiated samples showed localized distribution of the silicon that was added in the matrix during fuel fabrication and confirmed its beneficial effect on interaction layer growth during irradiation. The modifications of U-Mo fuel particles by the addition of a ternary alloying element (Ti or Zr, additional protective coatings (silicide or nitride, and the use of larger fuel particles resulted in significantly reduced interaction layers between fuel particles and Al.

  18. The mobile rig industry

    International Nuclear Information System (INIS)

    Karlsen, J.K.

    1992-08-01

    This study is part of the project ''A competitive Norway'', based on the theories and methods presented in the book ''The competitive advantage of nations'', by Michael E. Porter. The rig market may be segmented according to the type of service, the geographical market and the type of equipment. The focus of the report is exploration and appraisal drilling using jackup rigs and semi-submersible rigs in the Northwestern European market. Market shares of Norwegian, other European and US suppliers in the jackup and semi-submerisible market segments on the Norwegian continental shelf and the entire Northwestern market are presented. The main driving force behind the rig demand is the price of oil, but technological trends and changes in relative importance of the various geographical markets are also important. The industry is fairly fragmented on the supplier side, while the oil companies as customers have substantial bargaining power. There are high exit barriers because of the high capital intensity of the business. Combined with a highly volatile demand, this forces the industry through cycles of low capacity utilization and depressed rig day rates. 8 refs., 19 figs., 2 tabs

  19. Data report of a tight-lattice rod bundle thermal-hydraulic tests (1). Base case test using 37-rod bundle simulated water-cooled breeder reactor (Contract research)

    International Nuclear Information System (INIS)

    Kureta, Masatoshi; Tamai, Hidesada; Liu, Wei; Akimoto, Hajime; Sato, Takashi; Watanabe, Hironori; Ohnuki, Akira

    2006-03-01

    Japan Atomic Energy Agency has been performing tight-lattice rod bundle thermal-hydraulic tests to realize essential technologies for the technological and engineering feasibility of super high burn-up water-cooled breeder reactor featured by a high breeding ratio and super high burn-up by reducing the core water volume in water-cooled reactor. The tests are performing to make clear the fundamental subjects related to the boiling transition (BT) (Subjects: BT criteria under a highly tight-lattice rod bundle, effects of gap-width between rods and of rod-bowing) using 37-rod bundles (Base case test section (1.3mm gap-width), Two parameter effect test sections (Gap-width effect one (1.0mm) and Rod-bowing one)). In the present report, we summarize the test results from the base case test section. The thermal-hydraulic characteristics using the large scale test section were obtained for the critical power, the pressure drop and the wall heat transfer under a wide range of pressure, flow rate, etc. including normal operational conditions of the designed reactor. Effects of local peaking factor on the critical power were also obtained. (author)

  20. Data report of tight-lattice rod bundle thermal-hydraulic tests (2). Gap-width effect test using 37-rod bundle simulated water-cooled breeder reactor (Contract research)

    International Nuclear Information System (INIS)

    Tamai, Hidesada; Kureta, Masatoshi; Liu, Wei; Akimoto, Hajime; Sato, Takashi; Watanabe, Hironori; Ohnuki, Akira

    2006-11-01

    Japan Atomic Energy Agency has been performing tight-lattice rod bundle thermal-hydraulic tests to realize essential technologies for the technological and engineering feasibility of super high burn-up water-cooled breeder reactor featured by a high breeding ratio and super high burn-up by reducing the core water volume in water-cooled reactor. The tests are performing to make clear the fundamental subjects related to the boiling transition (BT) (Subjects: BT criteria under a highly tight-lattice rod bundle, effects of gap-width between rods and of rod-bowing) using 37-rod bundles (Base case test section (1.3mm gap-width), Two parameter effect test sections (Gap-width effect one (1.0mm) and Rod-bowing one)). In the present report, we summarize the test results from the gap-width effect test section. The thermal-hydraulic characteristics were obtained for the critical power under the steady-state and transient conditions, the pressure drop and the wall heat transfer within a wide range of pressure, flow rate, etc. including normal operational conditions of the designed reactor. Then the gap-width effects were also obtained from the comparison between the results using the base case test section and the gap-width effect one. (author)

  1. Methods for acquiring data in power ramping experiments with WWER fuel rods at high burnup

    Energy Technology Data Exchange (ETDEWEB)

    Bobrov, S N; Grachev, A F; Ovchinnikov, V A; Poliakov, I S; Matveev, N P [Research Inst. of Atomic Reactors, Dimitrovgrad (Russian Federation); Novikov, V V [Institute of Inorganic Materials, Moscow (Russian Federation)

    1997-08-01

    A programme on in-pile test which involve fuel burnup up to 60 MWd/kg and up to 12 fuel rods in the experimental rig is considered. Testing methods with reference to the MIR-M1 reactor are reported. Power ramping regime can be realized either by an increase of the total reactor capacity or by displacement of the nearest to the experimental cell control rods or by combination of these two ways. A total thermal capacity of the fuel rod cluster is determined by means of the thermal balance technique. The thermal capacity of each separate fuel rod can be estimated from the distribution of their relative activity within the accuracy range 5-10%. The important condition for this procedure is to keep the initial distribution of the fuel rod heating during power ramping. Means of instrumentation are described. They are standard detectors of loop facilities and transducers installed both in the irradiation rigs and fuel rods. Different ways of processing data on the fuel rod loss of integrity are reported. When the time of fuel rod loss of tightness is placed in correspondence with its capacity, processing can be made either on the maximum fuel rod heat load or on that at crack location. The information acquired in the experiments on the burnup values, heat rating distribution, kinetics of fission product gas emission, fuel rod elongation, fuel rod diameter changes, crack availability and fission products migration is used for the development and verification of calculation codes. (author). 1 ref., 4 figs, 1 tab.

  2. TRIP-ID: A tool for a smart and interactive identification of Magic Formula tyre model parameters from experimental data acquired on track or test rig

    Science.gov (United States)

    Farroni, Flavio; Lamberti, Raffaele; Mancinelli, Nicolò; Timpone, Francesco

    2018-03-01

    Tyres play a key role in ground vehicles' dynamics because they are responsible for traction, braking and cornering. A proper tyre-road interaction model is essential for a useful and reliable vehicle dynamics model. In the last two decades Pacejka's Magic Formula (MF) has become a standard in simulation field. This paper presents a Tool, called TRIP-ID (Tyre Road Interaction Parameters IDentification), developed to characterize and to identify with a high grade of accuracy and reliability MF micro-parameters from experimental data deriving from telemetry or from test rig. The tool guides interactively the user through the identification process on the basis of strong diagnostic considerations about the experimental data made evident by the tool itself. A motorsport application of the tool is shown as a case study.

  3. Development of Mechanical Sealing and Laser Welding Technology to Instrument Thermocouple for Nuclear Fuel Test Rod

    International Nuclear Information System (INIS)

    Joung, Chang-Young; Ahn, Sung-Ho; Hong, Jin-Tae; Kim, Ka-Hye; Huh, Sung-Ho

    2015-01-01

    Zircaloy-4 of the nuclear fuel test rod, AISI 316L of the mechanical sealing parts, and the MI (mineral insulated) cable at a thermocouple instrumentation are hetero-metals, and are difficult to weld to dissimilar materials. Therefore, a mechanical sealing method to instrument the thermocouple should be conducted using two kinds of sealing process as follows: One is a mechanical sealing process using Swagelok, which is composed of sealing components that consists of an end-cap, a seal tube, a compression ring and a Swagelok nut. The other is a laser welding process used to join a seal tube, and an MI cable, which are made of the same material. The mechanical sealing process should be sealed up with the mechanical contact compressed by the strength forced between a seal tube and an end-cap, and the laser welding process should be conducted to have no defects on the sealing area between a seal tube and an MI cable. Therefore, the mechanical sealing and laser welding techniques need to be developed to accurately measure the centerline temperature of the nuclear fuel test rod in an experimental reactor. The mechanical sealing and laser welding tests were conducted to develop the thermocouple instrumentation techniques for the nuclear fuel test rod. The optimum torque value of a Swagelok nut to seal the mechanical sealing part between the end-cap and seal tube was established through various torque tests using a torque wrench. The optimum laser welding conditions to seal the welding part between a seal tube and an MI cable were obtained through various welding tests using a laser welding system

  4. Development of Mechanical Sealing and Laser Welding Technology to Instrument Thermocouple for Nuclear Fuel Test Rod

    Energy Technology Data Exchange (ETDEWEB)

    Joung, Chang-Young; Ahn, Sung-Ho; Hong, Jin-Tae; Kim, Ka-Hye; Huh, Sung-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Zircaloy-4 of the nuclear fuel test rod, AISI 316L of the mechanical sealing parts, and the MI (mineral insulated) cable at a thermocouple instrumentation are hetero-metals, and are difficult to weld to dissimilar materials. Therefore, a mechanical sealing method to instrument the thermocouple should be conducted using two kinds of sealing process as follows: One is a mechanical sealing process using Swagelok, which is composed of sealing components that consists of an end-cap, a seal tube, a compression ring and a Swagelok nut. The other is a laser welding process used to join a seal tube, and an MI cable, which are made of the same material. The mechanical sealing process should be sealed up with the mechanical contact compressed by the strength forced between a seal tube and an end-cap, and the laser welding process should be conducted to have no defects on the sealing area between a seal tube and an MI cable. Therefore, the mechanical sealing and laser welding techniques need to be developed to accurately measure the centerline temperature of the nuclear fuel test rod in an experimental reactor. The mechanical sealing and laser welding tests were conducted to develop the thermocouple instrumentation techniques for the nuclear fuel test rod. The optimum torque value of a Swagelok nut to seal the mechanical sealing part between the end-cap and seal tube was established through various torque tests using a torque wrench. The optimum laser welding conditions to seal the welding part between a seal tube and an MI cable were obtained through various welding tests using a laser welding system.

  5. Assessment of the Structural Integrity of a Prototypical Instrumented IFMIF High Flux Test Module Rig by Fully 3D X-Ray Microtomography

    International Nuclear Information System (INIS)

    Tiseanu, I.; Craciunescu, T.; Mandache, B.N.; Simon, M.; Heinzel, V.; Stratmanns, E.; Simakov, S.P.; Leichtle, D.

    2006-01-01

    An inspection procedure to asses the mechanical integrity of IFMIF (International Fusion Materials Irradiation Facility) capsules and rigs during the irradiation campaign is necessary. Due to its penetration ability and contrast mechanism, the X-ray micro-tomography is the only known tool that could meet these requirements. In the High Flux Test Module (HFTM) of IFMIF miniaturized specimens are densely packed in capsules. The capsules which wear electric heaters and thermocouples are housed in rigs. To assure a well defined thermal contact the heater wires have to be attached to the capsules by brazing them into grooves. The examination of the quality of the braze material layer is of crucial interest in order to assure the best heat coupling of the heater wires to the capsule. A high density of the heaters is necessary to maintain the required temperature and, in addition NaK filling of narrow channels is employed for improving the 3D-heat transfer between the irradiation specimens and the capsule wall. Fully 3D tomographic inspections of a prototypical HFTM instrumented capsule, developed and manufactures at FZK, were conducted. In order to identify the optimum irradiation parameters and scanning configuration we carried out a comparative NDT analysis on two micro-tomography facilities, our compact, high magnification installation at NILPRP and two high-end industrial tomography facilities with higher X-ray energy and intensity at HWM. At optimum inspection parameters of a microfocus X-ray source (U=220 kV and I=300 μA) the geometry resolution was about 30-50 microns for characteristic dimension of the sample of 50 mm. Voids of 30 microns diameter and cracks of about 20 microns width can be detected. The absolute error of geometrical measurements should be sufficient for the assessment of the structural integrity of the irradiation capsule and for the geometry description within the thermal-hydraulic modeling. Space resolution could be further improved if one

  6. LWR fuel rod testing facilities in high flux reactor (HFT) Petten for investigation of power cycling and ramping behaviour

    Energy Technology Data Exchange (ETDEWEB)

    Markgraf, J; Perry, D; Oudaert, J [Commission of the European Communities, Joint Reserach Centre, Petten Establishment, Petten (Netherlands)

    1983-06-01

    LWR fuel rod irradiation experiments are being performed in HFR's Pool Side Facility (PSF) by means of pressurized boiling water capsules (BWFC). For more than 6 years the major application of these devices has been in performing irradiation programs to investigate the power ramp behaviour of PWR and BWR rods which have been pre-irradiated in power reactors. Irradiation devices with various types of monitoring instrumentation are available, e.g. for fuel rod length, fuel stack length, fuel rod internal pressure and fuel rod central temperature measurements. The application scope covers PWR and BWR fuel rods, with burn-up levels up to 45 MWd/kg(U), max. linear heat generation rates up to 700 W/cm and simulation of constant power change rates between 0.05 and 1000 W/cm.min. The paper describes the different designs of LWR fuel rod testing facilities and associated non-destructive testing devices in use at the HFR Petten. It also addresses the new test capabilities that will become available after exchange of the HFR vessel in 1983. Furthermore it shows some typical results. (author)

  7. Experimental data report for test TS-3 Reactivity Initiated Accident test in the NSRR with pre-irradiated BWR fuel rod

    International Nuclear Information System (INIS)

    Nakamura, Takehiko; Yoshinaga, Makio; Fujishiro, Toshio; Kobayashi, Shinsho; Yamahara, Takeshi; Sukegawa, Tomohide; Kikuchi, Teruo; Sobajima, Makoto.

    1993-09-01

    This report presents experimental data for Test TS-3 which was the third test in a series of Reactivity Initiated Accident (RIA) tests using pre-irradiated BWR fuel rods, performed in the Nuclear Safety Research Reactor (NSRR) in September, 1990. Test fuel rod used in the Test TS-3 was a short-sized BWR (7 x 7) type rod which was re-fabricated from a commercial rod irradiated in the Tsuruga Unit 1 power reactor of Japan Atomic Power Co. The fuel had an initial enrichment of 2.79 % and a burnup of 26 Gwd/tU. A pulse irradiation of the test fuel rod was performed under a cooling condition of stagnant water at atmospheric pressure and at ambient temperature which simulated a BWR's cold start-up RIA event. The energy deposition of the fuel rod in this test was evaluated to be 94 ± 4 cal/g · fuel (88 ± 4 cal/g · fuel in peak fuel enthalpy) and no fuel failure was observed. Descriptions on test conditions, test procedures, transient behavior of the test rod during the pulse irradiation, and results of pre-pulse and post-pulse irradiation examinations are described in this report. (author)

  8. TR-PIV Performance Test for a Flow Field Measurement in a Single Rod Test Section

    International Nuclear Information System (INIS)

    Park, Ju Yong; Shin, Chang Hwan; Lee, Chi Young; Oh, Dong Seok; In, Wang Kee

    2011-01-01

    For large enhancement of performance of Pressurized Water Reactor(PWR), dual-cooled fuel is being developed in Korea Atomic Energy Research Institute(KAERI). This nuclear fuel is a ring shape fuel which is different from conventional cylindrical nuclear fuel and cooling water flows both inner and outer channel. For this fuel, it widens the surface area. But it is bigger outer diameter of fuel rods. So, interval between fuel rods narrows. This because of outer channel flow is unstable. So, measurement of turbulence flow and perturbation that influence in heat transfer elevation is important.. To understand heat transfer characteristics by turbulence, measurement of flow perturbation element is necessary. To measure these turbulence characteristics, hot wire anemometer is widely used. However, it has many disadvantages such as low durability of prove, and big probe size. For these reasons, TR-PIV(Time-Resolved Particle Image Velocimetry) system is employed for better flow measurement in our research institute. TR-PIV system is consisted of laser system and high-speed camera that have high frequency. So, was judged that can measurement complicated turbulence flow and perturbation. In this paper, introduce TR-PIV system, and with results acquiring in single rod flow through this system, and wish to introduce about after this practical use plan

  9. Drop performance test of conceptually designed control rod assembly for prototype generation IV sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young Kyu; Lee, Jae Han; Kim, Hoe Woong; KIm, Sung Kyun; Kim, Jong Bum [Sodium-cooled Fast Reactor NSSS Design Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2017-06-15

    The control rod assembly controls reactor power by adjusting its position during normal operation and shuts down chain reactions by its free drop under scram conditions. Therefore, the drop performance of the control rod assembly is important for the safety of a nuclear reactor. In this study, the drop performance of the conceptually designed control rod assembly for the prototype generation IV sodium-cooled fast reactor that is being developed at the Korea Atomic Energy Research Institute as a next-generation nuclear reactor was experimentally investigated. For the performance test, the test facility and test procedure were established first, and several free drop performance tests of the control rod assembly under different flow rate conditions were then carried out. Moreover, performance tests under several types and magnitudes of seismic loading conditions were also conducted to investigate the effects of seismic loading on the drop performance of the control rod assembly. The drop time of the conceptually designed control rod assembly for 0% of the tentatively designed flow rate was measured to be 1.527 seconds, and this agrees well with the analytically calculated drop time. It was also observed that the effect of seismic loading on the drop time was not significant.

  10. Response of unirradiated and irradiated PWR fuel rods tested under power-cooling-mismatch conditions

    International Nuclear Information System (INIS)

    MacDonald, P.E.; Quapp, W.J.; Martinson, Z.R.; McCardell, R.K.; Mehner, A.S.

    1978-01-01

    This report summarizes the results from the single-rod power-cooling-mismatch (PCM) and irradiation effects (IE) tests conducted to date in the Power Burst Facility (PBF) at the U.S. DOE Idaho National Engineering Laboratory. This work was performed for the U.S. NRC under contact to the Department of Energy. These tests are part of the NRC Fuel Behavior Program, which is designed to provide data for the development and verification of analytical fuel behavior models that are used to predict fuel response to abnormal or postulated accident conditions in commercial LWRs. The mechanical, chemical and thermal response of both previously unirradiated and previously irradiated LWR-type fuel rods tested under power-cooling-mismatch condition is discussed. A brief description of the test designs is presented. The results of the PCM thermal-hydraulic studies are summarized. Primary emphasis is placed on the behavior of the fuel and cladding during and after stable film boiling. (orig.) [de

  11. Essays on bid rigging

    NARCIS (Netherlands)

    Seres, Gyula

    2016-01-01

    Manipulating prices in auctions raises antitrust concerns. Collusion lowers the revenue of the auctioneer and creates information rents. Bid rigging is a prevalent phenomenon and the affected market is enormous. Public procurement amounts to between 10 and 25 percent of national GDP in

  12. Hydraulic lifter for an underwater drilling rig

    Energy Technology Data Exchange (ETDEWEB)

    Garan' ko, Yu L

    1981-01-15

    A hydraulic lifter is suggested for an underwater drilling rig. It includes a base, hydraulic cylinders for lifting the drilling pipes connected to the clamp holder and hydraulic distributor. In order to simplify the design of the device, the base is made with a hollow chamber connected to the rod cavities and through the hydraulic distributor to the cavities of the hydraulic cylinders for lifting the drilling pipes. The hydraulic distributor is connected to the hydrosphere through the supply valve with control in time or by remote control. The base is equipped with reverse valves whose outlets are on the support surface of the base.

  13. Advancing rig design: latest rig technologies improving efficiency and safety

    Energy Technology Data Exchange (ETDEWEB)

    Greenaway, R.

    1997-12-01

    Recent advances in drilling rig technologies that improve the ways for finding oil and natural gas, and are also solving some safety and transportation problems, have been reviewed. The coiled tubing drilling rig developed by joint venture TransOcean Ensign Drilling Technology was one of the innovations described. It is able to run a three-and-a-quarter inch coiled tubing, the only system capable of doing this in a land-based application. Tesco Corporation`s new casing drilling rig, which is expected to lower the cost of moving the rig, and Brinkerhoff Drilling`s new generation modular (NGM)-rig, claimed to be the most mobile rig in North America, are other new developments worthy of note. Tesco`s casing drilling rig has the potential to reduce drilling costs by as much as 30 to 40 per cent, while the NGM-rig could reduce rig mobilization time by 50 to 80 per cent, and the number of wells drilled by the same rig could increase by 20 per cent, due to the NGM-rig`s versatility and flexibility.

  14. Investigation of TIG welding characteristics with a dual cooled rod for the fuel irradiation test

    International Nuclear Information System (INIS)

    Kim, Soo Sung; Kim, Hyung Kyu

    2008-01-01

    To establish the fabrication process, and for satisfying the requirements of the irradiation test, an TIG(Tungsten Inert Gas) welding machine for the dual cooled rods specimens was developed, and the preliminary welding experiments were performed to optimize the welding process conditions. Cladding tubes of 15.9 and 9 mm for the outer and inner diameters, respectively with a 0.57 mm thickness and end caps were used for the specimens. This paper describes the experimental results of the TIG welds and the micrograph examinations of the TIG welded specimens corresponding to various welding conditions for the dual cooled fuel irradiation test. The investigations revealed that the present TIG process satisfied the requirements for the fuel irradiation test in the HANARO research reactor

  15. Sensor for measurement of fuel rod gas pressure during loss-of-fluid-tests

    International Nuclear Information System (INIS)

    Billeter, T.R.

    1979-05-01

    Qualification tests have been conducted of a measurement system for determining the pressure of certain fuel rods in the loss-of-fluid-test (LOFT) reactor. Because of physical size (0.35-in. OD by 5.5-in length) and operational characteristics, an eddy current device was selected as the most promising measurement transducer for the application. The sensor must operate at pressure up to 17.2 MPa (2500 psig) and at temperatures up to 800 0 F. During the reactor transient caused by loss of coolant flow, sensor temperature and applied pressure will vary rapidly and significantly. Consequently, qualification tests included subjection of the sensor to rapid depressurization, temperature transients, and blowdowns in an autoclave, as well as to calibrations and various slow temperature cycles

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

    International Nuclear Information System (INIS)

    Rajan Babu, V.; Veerasamy, R.; Patri, Sudheer; Ignatius Sundar Raj, S.; Kumar Krovvidi, S.C.S.P.; Dash, S.K.; Meikandamurthy, C.; Rajan, K.K.; Puthiyavinayagam, P.; Chellapandi, P.; Vaidyanathan, G.; Chetal, S.C.

    2010-01-01

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

  17. Laboratory simulation of rod-to-rod mechanical interactions during postulated loss-of-coolant accidents in a PWR involving cladding oxidation

    International Nuclear Information System (INIS)

    Hindle, E.D.; Haste, T.J.; Harrison, W.R.

    1987-01-01

    Creep deformation of Zircaloy cladding in postulated PWR loss-of-coolant accidents may lead to rod-to-rod mechanical interactions. Tests have been performed in the electrically heated FOURSQUARE rig at 750 0 C and 850 0 C in steam to investigate this effect. Conservatisms inherent in a simple 'square with rounded corners' coolant channel blockage model have been quantified; about 5-10% flow area may remain even at strains which in ideal circumstances would give total blockage. Reduction of average burst strains produced by an oxide layer (up to 13 μm) has been demonstrated, resulting from strain concentration at oxide cracks. (author)

  18. Development of experimental method for self-wastage behavior in sodium-water reaction. Development of test rig (SWAT-2R) and study for experimental procedure

    International Nuclear Information System (INIS)

    Abe, Yuta; Shimoyama, Kazuhito; Kurihara, Akikazu

    2014-07-01

    In case of water leak from a penetrated crack on a tube of steam generator in the sodium cooled fast reactor (SFR), self-wastage, that increases the size of leak, may take place by corrosion related to chemical reaction between sodium and water. If the self-wastage continues in a certain period of time, the intact tube bundle may be damaged as a result of enlarged leak. For the safety evaluation of the accident, JAEA has been developing the analytical method of self-wastage using the multi-dimensional sodium-water reaction code. Experiments conducted so far used mainly crack-type test pieces. However, reproducibility was limited and it was difficult to evaluate individual effects of the phenomena in detail. This report describes the development of new experimental rig (SWAT-2R). SWAT-2R enables to examine corrosion effecting factors that were ambiguous in the previous studies. The report includes description of development of micro-leak test piece, examination of experimental procedure. The results will provide fundamental data for validation of the self-wastage analytical method. (author)

  19. A new generation drilling rig: hydraulically powered and computer controlled

    Energy Technology Data Exchange (ETDEWEB)

    Laurent, M.; Angman, P.; Oveson, D. [Tesco Corp., Calgary, AB, (Canada)

    1999-11-01

    Development, testing and operation of a new generation of hydraulically powered and computer controlled drilling rig that incorporates a number of features that enhance functionality and productivity, is described. The rig features modular construction, a large heated common drilling machinery room, permanently-mounted draw works which, along with the permanently installed top drive, significantly reduces rig-up/rig-down time. Also featured are closed and open hydraulic systems and a unique hydraulic distribution manifold. All functions are controlled through a programmable logic controller (PLC), providing almost unlimited interlocks and calculations to increase rig safety and efficiency. Simplified diagnostic routines, remote monitoring and troubleshooting are also part of the system. To date, two rigs are in operation. Performance of both rigs has been rated as `very good`. Little or no operational problems have been experienced; downtime has averaged 0.61 per cent since August 1998 when the the first of the two rigs went into operation. The most important future application for this rig is for use with the casing drilling process which eliminates the need for drill pipe and tripping. It also reduces the drilling time lost due to unscheduled events such as reaming, fishing and taking kicks while tripping. 1 tab., 6 figs.

  20. Thyc, a 3D thermal-hydraulic code for rod bundles. Recent developments and validation tests

    International Nuclear Information System (INIS)

    Caremoli, C.; Rascle, P.; Aubry, S.; Olive, J.

    1993-09-01

    PWR or LMFBR cores or fuel assemblies, PWR steam generators, condensers, tubular heat exchangers, are basic components of a nuclear power plant involving two-phase flows in tube or rod bundles. A deep knowledge of the detailed flow patterns on the shell side is necessary to evaluate DNB margins in reactor cores, singularity effects (grids, wire spacers, support plates, baffles), corrosion on steam generator tube sheet, bypass effects and vibration risks. For that purpose, Electricite de France has developed, since 1986, a general purpose code named THYC (Thermal HYdraulic Code) designed to study three-dimensional single and two phase flows in rod or tube bundles (pressurized water reactor cores, steam generators, condensers, heat exchangers). It considers the three-dimensional domain to contain two kinds of components: fluid and solids. The THYC model is obtained by space-time averaging of the instantaneous equations (mass, momentum and energy) of each phase over control volumes including fluid and solids. This paper briefly presents the physical model and the numerical method used in THYC. Then, validation tests (comparison with experiments) and applications (coupling with three-dimensional neutronics code and DNB predictions) are presented. They emphasize the last developments and new capabilities of the code. (authors). 10 figs., 3 tabs., 21 refs

  1. EPRI/DOE High-Burnup Fuel Sister Rod Test Plan Simplification and Visualization

    Energy Technology Data Exchange (ETDEWEB)

    Saltzstein, Sylvia J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sorenson, Ken B. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hanson, B. D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Shimskey, R. W. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Klymyshyn, N. A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Webster, R. A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Jensen, P. J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); MacFarlan, P. J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Billone, Mike [Argonne National Lab. (ANL), Argonne, IL (United States); Scaglione, John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Montgomery, Rose [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bevard, Bruce [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-09-15

    The EPRI/DOE High-Burnup Confirmatory Data Project (herein called the “Demo”) is a multi-year, multi-entity test with the purpose of providing quantitative and qualitative data to show if high-burnup fuel mechanical properties change in dry storage over a ten-year period. The Demo involves obtaining 32 assemblies of high-burnup PWR fuel of common cladding alloys from the North Anna Nuclear Power Plant, loading them in an NRC-licensed TN-32B cask, drying them according to standard plant procedures, and then storing them on the North Anna dry storage pad for ten years. After the ten-year storage time, the cask will be opened and the mechanical properties of the rods will be tested and analyzed.

  2. The "Rod and Fran Test": relationship priming influences cognitive-perceptual performance.

    Science.gov (United States)

    Baldwin, Mark W; Bagust, Jeff; Docherty, Sharon; Browman, Alexander S; Jackson, Joshua C

    2014-01-01

    We theorized that interpersonal relationships can provide structures for experience. In particular, we tested whether primes of same-sex versus mixed-sex relationships could foster cognitive-perceptual processing styles known to be associated with independence versus interdependence respectively. Seventy-two participants visualized either a same-sex or other-sex relationship partner and then performed two measures of cognitive-perceptual style. On a computerized Rod and Frame Test, individuals were more field-dependent after visualizing a mixed-sex versus same-sex relationship partner. On a measure involving perceptions of group behavior, participants demonstrated more holistic/contextually based perception after being primed with a female versus male relationship partner. These findings support the hypothesis that activated cognitive structures representing interpersonal relationships can shape individuals' cognitive-perceptual performance.

  3. Feasibility evaluation of x-ray imaging for measurement of fuel rod bowing in CFTL test bundles

    International Nuclear Information System (INIS)

    Baker, S.P.

    1980-06-01

    The Core Flow Test Loop (CFTL) is a high temperature, high pressure, out-of-reactor helium-circulating system. It is designed for detailed study of the thermomechanical performance, at prototypic steady-state and transient operating conditions, of electrically heated rods that simulate segments of core assemblies in the Gas-Cooled Fast Breeder reactor demonstration plant. Results are presented of a feasibility evaluation of x-ray imaging for making measurements of the displacement (bowing) of fuel rods in CFTL test bundles containing electrically heated rods. A mock-up of a representative CFTL test section consisting of a test bundle and associated piping was fabricated to assist in this evaluation

  4. Assessment of the structural integrity of a prototypical instrumented IFMIF high flux test module rig by fully 3D X-ray microtomography

    Energy Technology Data Exchange (ETDEWEB)

    Tiseanu, Ion [National Institute for Laser, Plasma and Radiation Physics, Plasma Physics and Nuclear Fusion Laboratory NILPRP, P.O. Box MG-36, R-77125 Bucharest-Magurele (Romania)], E-mail: tiseanu@infim.ro; Simon, Martin [Hans Waelischmiller GmbH (HWM), Schiessstattweg 16, D-88677 Markdorf (Germany); Craciunescu, Teddy; Mandache, Bogdan N. [National Institute for Laser, Plasma and Radiation Physics, Plasma Physics and Nuclear Fusion Laboratory NILPRP, P.O. Box MG-36, R-77125 Bucharest-Magurele (Romania); Heinzel, Volker; Stratmanns, Erwin; Simakov, Stanislaw P.; Leichtle, Dieter [Forschungszentrum Karlsruhe (FZK), Institut fuer Reaktorsicherheit IRS, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2007-10-15

    An inspection procedure to assess the mechanical integrity of the International Fusion Materials Irradiation Facility (IFMIF) capsules and rigs during the irradiation campaign is necessary. Due to its penetration ability and contrast mechanism, the X-ray microtomography is the only known tool that could meet these requirements. In the high flux test module (HFTM) of IFMIF miniaturized specimens are densely packed in capsules. The capsules, which wear electric heaters and thermocouples, are housed in rigs. To assure a well-defined thermal contact the heater wires have to be attached to the capsules by brazing them into grooves. The examination of the quality of the braze material layer is of crucial interest in order to assure the best heat coupling of the heater wires to the capsule. A high density of the heaters is necessary to maintain the required temperature and, in addition NaK filling of narrow channels is employed for improving the 3D-heat transfer between the irradiation specimens and the capsule wall. Fully 3D tomographic inspections of a prototypical HFTM instrumented capsule, developed and manufactured at FZK, were conducted. In order to identify the optimum irradiation parameters and scanning configuration we carried out a comparative NDT analysis on two microtomography facilities: a compact, high magnification installation at NILPRP and a high-end industrial tomography facility with higher X-ray energy and intensity at HWM. At optimum inspection parameters of a directional microfocus X-ray source (U = 220 kV and I = 300 {mu}A) the geometry resolution was about 30 microns for characteristic dimension of the sample of 50 mm. Voids of 30 microns diameter and cracks of about 20 microns width can be detected. The absolute error of geometrical measurements is sufficient for the assessment of the structural integrity of the irradiation capsule and for the geometry description within the thermal-hydraulic modeling. The space resolution and the overall

  5. Assessment of the structural integrity of a prototypical instrumented IFMIF high flux test module rig by fully 3D X-ray microtomography

    International Nuclear Information System (INIS)

    Tiseanu, Ion; Simon, Martin; Craciunescu, Teddy; Mandache, Bogdan N.; Heinzel, Volker; Stratmanns, Erwin; Simakov, Stanislaw P.; Leichtle, Dieter

    2007-01-01

    An inspection procedure to assess the mechanical integrity of the International Fusion Materials Irradiation Facility (IFMIF) capsules and rigs during the irradiation campaign is necessary. Due to its penetration ability and contrast mechanism, the X-ray microtomography is the only known tool that could meet these requirements. In the high flux test module (HFTM) of IFMIF miniaturized specimens are densely packed in capsules. The capsules, which wear electric heaters and thermocouples, are housed in rigs. To assure a well-defined thermal contact the heater wires have to be attached to the capsules by brazing them into grooves. The examination of the quality of the braze material layer is of crucial interest in order to assure the best heat coupling of the heater wires to the capsule. A high density of the heaters is necessary to maintain the required temperature and, in addition NaK filling of narrow channels is employed for improving the 3D-heat transfer between the irradiation specimens and the capsule wall. Fully 3D tomographic inspections of a prototypical HFTM instrumented capsule, developed and manufactured at FZK, were conducted. In order to identify the optimum irradiation parameters and scanning configuration we carried out a comparative NDT analysis on two microtomography facilities: a compact, high magnification installation at NILPRP and a high-end industrial tomography facility with higher X-ray energy and intensity at HWM. At optimum inspection parameters of a directional microfocus X-ray source (U = 220 kV and I = 300 μA) the geometry resolution was about 30 microns for characteristic dimension of the sample of 50 mm. Voids of 30 microns diameter and cracks of about 20 microns width can be detected. The absolute error of geometrical measurements is sufficient for the assessment of the structural integrity of the irradiation capsule and for the geometry description within the thermal-hydraulic modeling. The space resolution and the overall

  6. Modernisation of a test rig for determination of vehicle shock absorber characteristics by considering vehicle suspension elements and unsprung masses

    Science.gov (United States)

    Maniowski, M.; Para, S.; Knapczyk, M.

    2016-09-01

    This paper presents a modernization approach of a standard test bench for determination of damping characteristics of automotive shock absorbers. It is known that the real-life work conditions of wheel-suspension dampers are not easy to reproduce in laboratory conditions, for example considering a high frequency damper response or a noise emission. The proposed test bench consists of many elements from a real vehicle suspension. Namely, an original tyre-wheel with additional unsprung mass, a suspension spring, an elastic top mount, damper bushings and a simplified wheel guiding mechanism. Each component was tested separately in order to identify its mechanical characteristics. The measured data serve as input parameters for a numerical simulation of the test bench behaviour by using a vibratory model with 3 degrees of freedom. Study on the simulation results and the measurements are needed for further development of the proposed test bench.

  7. Comparative performance analysis of ice plant test rig with TiO2-R-134a nano refrigerant and evaporative cooled condenser

    Directory of Open Access Journals (Sweden)

    Amrat Kumar Dhamneya

    2018-03-01

    Full Text Available The nanoparticle is used in chillers for increasing system performance. The increasing concentration of nanoparticles (TiO2 in refrigerant increases the performances of the system due decreasing compressor work done and enhance heat transfer rate. For hot and dry climate condition, performances of air-cooled condenser minimize, and C. O. P. decreases extensively in chillers due to heat transfer rate decreases in the condenser. In the condenser, nano-refrigerants are not cool at the desired level, and the system was faulty. These drawbacks of the nano-particles mixed refrigerator have promoted the research and improving heat rejection rate in the condenser. In this article, vapour compression refrigeration system coupled with evaporative cooling pad, and nano-refrigerant, for improving the performance of the system in hot & dry weather is proposed and compared experimentally. Combined evaporative cooling system and ice plant test rig have been proposed for the appropriate heat rejection offered in the condenser due to a faulty system run at high pressure. The experimental investigations revealed that the performance characteristics of the evaporatively-cooled condenser are significantly enhanced. Maximum C.O.P. increases by about 51% in the hot and dry climate condition than the normal system.

  8. Design of an Adaptive Power Regulation Mechanism and a Nozzle for a Hydroelectric Power Plant Turbine Test Rig

    Science.gov (United States)

    Mert, Burak; Aytac, Zeynep; Tascioglu, Yigit; Celebioglu, Kutay; Aradag, Selin; ETU Hydro Research Center Team

    2014-11-01

    This study deals with the design of a power regulation mechanism for a Hydroelectric Power Plant (HEPP) model turbine test system which is designed to test Francis type hydroturbines up to 2 MW power with varying head and flow(discharge) values. Unlike the tailor made regulation mechanisms of full-sized, functional HEPPs; the design for the test system must be easily adapted to various turbines that are to be tested. In order to achieve this adaptability, a dynamic simulation model is constructed in MATLAB/Simulink SimMechanics. This model acquires geometric data and hydraulic loading data of the regulation system from Autodesk Inventor CAD models and Computational Fluid Dynamics (CFD) analysis respectively. The dynamic model is explained and case studies of two different HEPPs are performed for validation. CFD aided design of the turbine guide vanes, which is used as input for the dynamic model, is also presented. This research is financially supported by Turkish Ministry of Development.

  9. Analytical support for the B4C control rod test QUENCH-07

    International Nuclear Information System (INIS)

    Homann, C.; Hering, W.; Fernandez Benitez, J.A.; Ortega Bernardo, M.

    2003-04-01

    Degradation of B 4 C absorber rods during a beyond design accident in a nuclear power reactor may be a safety concern. Among others, the integral test QUENCH-07 is performed in the FZK QUENCH facility and supported by analytical work within the Euratom Fifth Framework Programme on Nuclear Fission Safety to get a more profound database. Since the test differed substantially from previous QUENCH tests, much more work had to be done for pretest calculations than usual to guarantee the safety of the facility and to derive the test protocol. Several institutions shared in this work with different computer code systems, as used for nuclear reactor safety analyses. Due to this effort, problems could be identified and solved, leading to several modifications of the originally planned test conduct, until a feasible test protocol could be derived and recommended. All calculations showed the same trends. Especially the high temperatures and hence the small safety margin for the facility were a concern. In this report, contributions of various authors, engaged in this work, are presented. The test QUENCH-07 and the related computational support by the engaged institutions were co-financed by the European Community under the Euratom Fifth Framework Programme on Nuclear Fission Safety 1998 - 2002 (COLOSS Project, contract No. FIKS-CT-1999-00002). (orig.)

  10. Analytical support for the B{sub 4}C control rod test QUENCH-07

    Energy Technology Data Exchange (ETDEWEB)

    Homann, C.; Hering, W. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Reaktorsicherheit]|[Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Programm Nukleare Sicherheitsforschung; Birchley, J. [Paul Scherrer Inst. (Switzerland); Fernandez Benitez, J.A.; Ortega Bernardo, M. [Univ. Politecnica de Madrid (Spain)

    2003-04-01

    Degradation of B{sub 4}C absorber rods during a beyond design accident in a nuclear power reactor may be a safety concern. Among others, the integral test QUENCH-07 is performed in the FZK QUENCH facility and supported by analytical work within the Euratom Fifth Framework Programme on Nuclear Fission Safety to get a more profound database. Since the test differed substantially from previous QUENCH tests, much more work had to be done for pretest calculations than usual to guarantee the safety of the facility and to derive the test protocol. Several institutions shared in this work with different computer code systems, as used for nuclear reactor safety analyses. Due to this effort, problems could be identified and solved, leading to several modifications of the originally planned test conduct, until a feasible test protocol could be derived and recommended. All calculations showed the same trends. Especially the high temperatures and hence the small safety margin for the facility were a concern. In this report, contributions of various authors, engaged in this work, are presented. The test QUENCH-07 and the related computational support by the engaged institutions were co-financed by the European Community under the Euratom Fifth Framework Programme on Nuclear Fission Safety 1998 - 2002 (COLOSS Project, contract No. FIKS-CT-1999-00002). (orig.)

  11. Field Demonstraton of Existing Microhole Coiled Tubing Rig (MCTR) Technology

    Energy Technology Data Exchange (ETDEWEB)

    Kent Perry; Samih Batarseh; Sheriff Gowelly; Thomas Hayes

    2006-05-09

    The performance of an advanced Microhole Coiled Tubing Rig (MCTR) has been measured in the field during the drilling of 25 test wells in the Niobrara formation of Western Kansas and Eastern Colorado. The coiled tubing (CT) rig designed, built and operated by Advanced Drilling Technologies (ADT), was documented in its performance by GTI staff in the course of drilling wells ranging in depth from 500 to nearly 3,000 feet. Access to well sites in the Niobrara for documenting CT rig performance was provided by Rosewood Resources of Arlington, VA. The ADT CT rig was selected for field performance evaluation because it is one of the most advanced commercial CT rig designs that demonstrate a high degree of process integration and ease of set-up and operation. Employing an information collection protocol, data was collected from the ADT CT rig during 25 drilling events that encompassed a wide range of depths and drilling conditions in the Niobrara. Information collected included time-function data, selected parametric information indicating CT rig operational conditions, staffing levels, and field observations of the CT rig in each phase of operation, from rig up to rig down. The data obtained in this field evaluation indicates that the ADT CT rig exhibited excellent performance in the drilling and completion of more than 25 wells in the Niobrara under varied drilling depths and formation conditions. In the majority of the 25 project well drilling events, ROP values ranged between 300 and 620 feet per hour. For all but the lowest 2 wells, ROP values averaged approximately 400 feet per hour, representing an excellent drilling capability. Most wells of depths between 500 and 2,000 feet were drilled at a total functional rig time of less than 16 hours; for wells as deep at 2,500 to 3,000 feet, the total rig time for the CT unit is usually well under one day. About 40-55 percent of the functional rig time is divided evenly between drilling and casing/cementing. The balance of

  12. Tidal fields in general relativity: D'Alembert's principle and the test rigid rod

    International Nuclear Information System (INIS)

    Faulkner, J.; Flannery, B.P.

    1978-01-01

    To the general relativist, tidal forces are a manifestation of the Riemann tensor; the relativist therefore uses the Riemann tensor to calculate the effects of such forces. In contrast, we show that the intorduction of gravitational ''probes'' (or ''test rigid rods'') and the adoption of a view-point closely allied to d'Alembert's principle, give an enormous simplification in cases of interest. No component of the Riemann tensor need to be calculated as such. In the corotating orbital case (or Roche problem) the calculation of the relevant distortional field becomes trivial. As a by-product of this investigation, there emerges an illuminating strong field generalization of de Sitter's weak field precession for slowly spinning gyroscopes

  13. Fuel integrity project: analysis of light water reactor fuel rods test results

    Energy Technology Data Exchange (ETDEWEB)

    Dallongeville, M.; Werle, J. [COGEMA Logistics (AREVA Group) (France); McCreesh, G. [BNFL Nuclear Sciences and Technology Services (United Kingdom)

    2004-07-01

    BNFL Nuclear Sciences and Technology Services and COGEMA LOGISTICS started in the year 2000 a joint project known as FIP (Fuel Integrity Project) with the aim of developing realistic methods by which the response of LWR fuel under impact accident conditions could be evaluated. To this end BNFL organised tests on both unirradiated and irradiated fuel pin samples and COGEMA LOGISTICS took responsibility for evaluating the test results. Interpretation of test results included simple mechanical analysis as well as simulation by Finite Element Analysis. The first tests that were available for analysis were an irradiated 3 point bending commissioning trial and a lateral irradiated hull compression test, both simulating the loading during a 9 m lateral regulatory drop. The bending test span corresponded roughly to a fuel pin intergrid distance. The outcome of the test was a failure starting at about 35 mm lateral deflection and a few percent of total deformation. Calculations were carried out using the ANSYS code employing a shell and brick model. The hull lateral compaction test corresponds to a conservative compression by neighbouring pins at the upper end of the fuel pin. In this pin region there are no pellets inside. The cladding broke initially into two and later into four parts, all of which were rather similar. Initial calculations were carried out with LS-DYNA3D models. The models used were optimised in meshing, boundary conditions and material properties. The calculation results compared rather well with the test data, in particular for the detailed ANSYS approach of the 3 point bending test, and allowed good estimations of stresses and deformations under mechanical loading as well as the derivation of material rupture criteria. All this contributed to the development of realistic numerical analysis methods for the evaluation of LWR fuel rod behaviour under both normal and accident transport conditions. This paper describes the results of the 3 point bending

  14. Fuel integrity project: analysis of light water reactor fuel rods test results

    International Nuclear Information System (INIS)

    Dallongeville, M.; Werle, J.; McCreesh, G.

    2004-01-01

    BNFL Nuclear Sciences and Technology Services and COGEMA LOGISTICS started in the year 2000 a joint project known as FIP (Fuel Integrity Project) with the aim of developing realistic methods by which the response of LWR fuel under impact accident conditions could be evaluated. To this end BNFL organised tests on both unirradiated and irradiated fuel pin samples and COGEMA LOGISTICS took responsibility for evaluating the test results. Interpretation of test results included simple mechanical analysis as well as simulation by Finite Element Analysis. The first tests that were available for analysis were an irradiated 3 point bending commissioning trial and a lateral irradiated hull compression test, both simulating the loading during a 9 m lateral regulatory drop. The bending test span corresponded roughly to a fuel pin intergrid distance. The outcome of the test was a failure starting at about 35 mm lateral deflection and a few percent of total deformation. Calculations were carried out using the ANSYS code employing a shell and brick model. The hull lateral compaction test corresponds to a conservative compression by neighbouring pins at the upper end of the fuel pin. In this pin region there are no pellets inside. The cladding broke initially into two and later into four parts, all of which were rather similar. Initial calculations were carried out with LS-DYNA3D models. The models used were optimised in meshing, boundary conditions and material properties. The calculation results compared rather well with the test data, in particular for the detailed ANSYS approach of the 3 point bending test, and allowed good estimations of stresses and deformations under mechanical loading as well as the derivation of material rupture criteria. All this contributed to the development of realistic numerical analysis methods for the evaluation of LWR fuel rod behaviour under both normal and accident transport conditions. This paper describes the results of the 3 point bending

  15. The testing report of the development for the lithium grains and lithium rod automatic machine

    International Nuclear Information System (INIS)

    Qian Zongkui; Kong Xianghong; Huang Yong

    2008-06-01

    With the development of lithium industry, the lithium grains and lithium rod, as additive or catalyzer, having a big comparatively acreage and a strong activated feature, have a broad application. The lithium grains and lithium rod belong to the kind of final machining materials. The principle of the lithium grains and lithium rod that how to take shape through the procedures of extrusion, cutting, anti-conglutination, threshing and so on are analysed, A sort of lithium grains and lithium rod automatic machine is developed. (authors)

  16. Calculation study of nonequilibrium post-CHF heat transfer in rod bundle test using modified RELAP5/MOD2

    International Nuclear Information System (INIS)

    Hassan, Y.A.

    1987-01-01

    To date there is only very limited data for non-equilibrium convective film boiling in rod bundle geometries. A recent nine (3 x 3) rod bundle post-critical-flux (CHF) test from the Lehigh University test facility was simulated using RELAP5/MOD2, to assess its capabilities in predicting the overall convective mechanisms in post-CHF heat transfer in rod bundle geometries. The code calculations were compared with experimental data. The code predicted low vapor superheats and void fraction oscillations. A new interfacial heat transfer between the droplet/steam resulted in a reasonable prediction of vapor superheats. A revised dispersed flow film boiling correlation which accounts for the enhancement of steam convective cooling by droplet-induced turbulence was incorporated in the code. Comparison with the data showed a fair agreement

  17. Results of VVER fuel rods tests in the MIR.M1 reactor under power cycling conditions

    International Nuclear Information System (INIS)

    Burukin, A.; Izhutov, A.; Ovchinnikov, V.; Kalygin, V.; Markov, D.; Pimenov, Y.; Novikov, V.; Medvedev, A.; Nesterov, B.

    2011-01-01

    The paper presents the main results of the 50 ... 60 MWd/kgU burnup VVER fuel rods tests performed in the MIR.M1 reactor loop facilities under power cycling. The non-destructive PIE results are presented as well. A series of experiments was performed, including overall measurement of fuel rod parameters test, in one of which 300 cycles were done. Irradiation under power cycling conditions and PIE of high-burnup VVER fuel rods showed the following: 1) all fuel rods claddings preserved their integrity under irradiation at linear heat rate (LHR) higher than the NPP operating one; 2) experimental data were obtained on the axial and radial cladding strain and fission gas release (FGR) from 50 ... 60 MWd/kgU burnup VVER-440 and VVER-1000 fuel rods as well as on the kinetics of the change in these parameters and fuel temperature under the power cycling; 3) non-destructive PIE results are in a satisfactory correlation with the data obtained by means of in-pile measurement gages during irradiation. (authors)

  18. Multiscale characterization of White Etching Cracks (WEC) in a 100Cr6 bearing from a thrust bearing test rig

    DEFF Research Database (Denmark)

    Danielsen, Hilmar Kjartansson; Guzmán, F. Gutiérrez; Dahl, Kristian Vinter

    2017-01-01

    A common cause for premature bearing failures in wind turbine gearboxes are the so-called White Etching Cracks (WEC). These undirected, three-dimensional cracks are bordered by regions of altered microstructure and ultimately lead to a cracking or spalling of the raceway. An accelerated WEC test...... significant grain refinement. Atom probe tomography showed the microstructure in the undamaged zone has a plate-like martensitic structure with carbides, while no carbides were detected in the WEA where the microstructure consisted of equiaxed 10 nm grains. A three dimensional characterisation of WEC network...

  19. Prototypical Rod Construction Demonstration Project. Phase 3, Final report: Volume 1, Cold checkout test report, Book 3

    Energy Technology Data Exchange (ETDEWEB)

    1993-05-01

    The objective of Phase 3 of the Prototypical Rod consolidation Demonstration Project (PRCDP) was to procure, fabricate, assemble, and test the Prototypical Rod consolidation System as described in the NUS Phase 2 Final Design Report. This effort required providing the materials, components, and fabricated parts which makes up all of the system equipment. In addition, it included the assembly, installation, and setup of this equipment at the Cold Test Facility. During the Phase 3 effort the system was tested on a component, subsystem, and system level. This volume 1, discusses the PRCDP Phase 3 Test Program that was conducted by the HALLIBURTON NUS Environmental Corporation under contract AC07-86ID12651 with the United States Department of Energy. This document, Volume 1, Book 3 discusses the following topics: Downender Test Results and Analysis Report; NFBC Canister Upender Test Results and Analysis Report; Fuel Assembly Handling Fixture Test Results and Analysis Report; and Fuel Canister Upender Test Results and Analysis Report.

  20. Tensile and impact testing of an HFBR [High Flux Beam Reactor] control rod follower

    International Nuclear Information System (INIS)

    Czajkowski, C.J.; Schuster, M.H.; Roberts, T.C.; Milian, L.W.

    1989-08-01

    The Materials Technology Group of the Department of Nuclear Energy (DNE) at Brookhaven National Laboratory (BNL) undertook a program to machine and test specimens from a control rod follower from the High Flux Beam Reactor (HFBR). Tensile and Charpy impact specimens were machined and tested from non-irradiated aluminum alloys in addition to irradiated 6061-T6 from the HFBR. The tensile test results on irradiated material showed a two-fold increase in tensile strength to a maximum of 100.6 ksi. The impact resistance of the irradiated material showed a six-fold decrease in values (3 in-lb average) compared to similar non-irradiated material. Fracture toughness (K I ) specimens were tested on an unirradiated compositionally and dimensionally similar (to HFBR follower) 6061 T-6 material with K max values of 24.8 ± 1.0 Ksi√in (average) being obtained. The report concludes that the specimens produced during the program yielded reproducible and believable results and that proper quality assurance was provided throughout the program. 9 figs., 6 tabs

  1. Gas cooled fast reactor control rod drive mechanism deceleration unit. Test program

    International Nuclear Information System (INIS)

    Wagner, T.H.

    1981-10-01

    This report presents the results of the airtesting portion of the proof-of-principle testing of a Control Rod Scram Deceleration Device developed for use in the Gas Cooled Fast Reactor (GCFR). The device utilizes a grooved flywheel to decelerate the translating assembly (T/A). Two cam followers on the translating assembly travel in the flywheel grooves and transfer the energy of the T/A to the flywheel. The grooves in the flywheel are straight for most of the flywheel length. Near the bottom of the T/A stroke the grooves are spiraled in a decreasing slope helix so that the cam followers accelerate the flywheel as they transfer the energy of the falling T/A. To expedite proof-of-principle testing, some of the materials used in the fabrication of certain test article components were not prototypic. With these exceptions the concept appears to be acceptable. The initial test of 300 scrams was completed with only one failure and the failure was that of a non-prototypic cam follower outer sleeve material

  2. Experimental test on aluminium rod submitted to a laminar water flow

    International Nuclear Information System (INIS)

    Britto Aghina, L.O. de; Cruz, J.R.B.

    1986-06-01

    The result obtained from a experiment with an aluminium rod submitted to a laminar water flow is compared to the result predicted by empirical correlations used in the vibration analysis of the RPR reactor fuel rods. (L.C.J.A.)

  3. Experimental system description for air-water CCFL tests of the 161-rod FLECHT-SEASET test vessel upper plenum

    International Nuclear Information System (INIS)

    Fogdall, S.P.; Anderson, J.L.

    1983-01-01

    A series of countercurrent flow limiting (CCFL) experiments has been performed by EG and G Idaho, Inc. in the Steam-Air-Water (SAW) test facility at the Idaho National Engineering Laboratory on behalf of the US Nuclear Regulatory Commission (NRC). Tests were performed in a mockup of the vessel for the 161-Rod Systems Effects Test (SET) facility of the FLECHT-SEASET program, conducted by the Westinghouse Electric Corporation. Westinghouse and the NRC will use the test results to provide a CCFL correlation to predict the flooding behavior in the upper plenum of the SET vessel. This paper presents a description of the experimental system and the test conduct, including data validation and uncertainty analysis. The test objectives centered on experimentally obtaining coefficients in the Wallis correlation for flooding with the specific vessel geometry. The test conditions and vessel configuration are described and the design of the test loop, instrumentation, and data acquisition are discussed. The establishment of a test point and the resultant data are described

  4. Analysis of Radial Plutonium Isotope Distribution in Irradiated Test MOX Fuel Rods

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Jae Yong; Lee, Byung Ho; Koo, Yang Hyun; Kim, Han Soo

    2009-01-15

    After Rod 3 and 6 (KAERI MOX) were irradiated in the Halden reactor, their post-irradiation examinations are being carried out now. In this report, PLUTON code was implemented to analyze Rod 3 and 6 (KAERI MOX). In the both rods, the ratio of a maximum burnup to an average burnup in the radial distribution was 1.3 and the contents of {sup 239}Pu tended to increase as the radial position approached the periphery of the fuel pellet. The detailed radial distribution of {sup 239}Pu and {sup 240}Pu, however, were somewhat different. To find the reason for this difference, the contents of Pu isotopes were investigated as the burnup increased. The content of {sup 239}Pu decreased with the burnup. The content of {sup 240}Pu increased with the burnup by the 20 GWd/tM but decreased over the 20 GWd/tM. The local burnup of Rod 3 is higher than that of Rod 6 due to the hole penetrated through the fuel rod. The content of {sup 239}Pu decreased more rapidly than that of {sup 240}Pu in the Rod 6 with the increased burnup. It resulted in a radial distribution of {sup 239}Pu and {sup 240}Pu similar to Rod 3. The ratio of Xe to Kr is a parameter to find where the fissions occur in the nuclear fuel. In both Rod 3 and 6, it was 18.3 in the whole fuel rod cross section, which showed that the fissions occurred in the plutonium.

  5. Nondestructive testing of PWR type fuel rods by eddy currents and metrology in the OSIRIS reactor pool

    International Nuclear Information System (INIS)

    Faure, M.; Marchand, L.

    1985-02-01

    The Saclay Reactor Department has developed a nondestructive test bench, now installed above channel 1 of the OSIRIS reactor. As part of investigations into the dynamics of PWR fuel degradation, a number of fuel rods underwent metrological and eddy current inspection, after irradiation [fr

  6. Critical heat flux tests for self-spaced square finned 7 fuel rod bundle

    International Nuclear Information System (INIS)

    Moon, Sang Ki; Chun, Se Young; Choi, Ki Young; Park, Jong Kuk; Hwang, Dae Hyun; Zee, Sung Quun; Kim, Keung Koo

    2001-09-01

    Now, KAERI is developing a new advanced reactor aimed at achieving highly enhanced safety and reliability, and improved economics. SSF (Self-Spaced Square Finned) fuel rod bundle is considered as a suitable one for the new advanced reactor. The SSF fuel rods have rectangular shapes and four fins at the corners, and are arranged in triangular geometry. While the SSF fuel rod bundle is considered to have enhanced cooling efficiency, the correlations used for commercial PWR might be able to be applied. The application results of some conventional correlations show that the SSF fuel rod bundle show an enhanced CHF performance about 10 to 40 %. When some conventional CHF correlations are applied to CHF data with a similar geometry to the SSF fuel rod bundle, conventional CHF correlations including a correlation developed in Russia are judged not to be suitable for the development of SSF fuel rod bundle and for the use in a safety analysis code. From CHF experiments for SSF 7 fuel rod bundle performed in KAERI, the following results are obtained: the CHF increases with increasing mass flux, and the CHF increasing rate decreases at high mass flux conditions. The exit quality decreases with increasing mass flux. The overall effect of the mass flux on the CHF and exit quality coincides with previous understanding. Compared to the CHF data of IPPE with the same system pressure and inlet temperature, the CHF data of KAERI show the similar values. Thus, the reliability of IPPE CHF data can be confirmed indirectly

  7. Control rod

    International Nuclear Information System (INIS)

    Kawakami, Kazuo; Shimoshige, Takanori; Nishimura, Akira

    1979-01-01

    Purpose: A control rod has been developed, which provided a plurality of through-holes in the vicinity of the sheath fitting position, in order to flatten burn-up, of fuel rods in positions confronting a control rod. Thereby to facilitate the manufacture of the control rods and prevent fuel rod failures. Constitution: A plurality of through-holes are formed in the vicinity of the sheath fitting position of a central support rod to which a sheath for the control rod is fitted. These through-holes are arranged in the axial direction of the central support rod. Accordingly, burn-up of fuel rods confronting the control rods can be reduced by through-holes and fuel rod failures can be prevented. (Yoshino, Y.)

  8. Design concept and testing of an in-bundle gamma densitometer for subchannel void fraction measurements in the THTF electrically heated rod bundle

    International Nuclear Information System (INIS)

    Felde, D.K.

    1982-04-01

    A design concept is presented for an in-bundle gamma densitometer system for measurement of subchannel average fluid density and void fraction in rod or tube bundles. This report describes (1) the application of the design concept to the Thermal-Hydraulic Test Facility (THTF) electrically heated rod bundle; and (2) results from tests conducted in the THTF

  9. Production and testing of flexible welding flux rods, used for protecting briquetting press molds from wear

    Energy Technology Data Exchange (ETDEWEB)

    Loescher, B.; Czerwinski, M.; Dittrich, V.

    1985-11-01

    Production, properties and trial application are discussed for the Feroplast ZIS 218 welding powder rod, developed for automated surface armouring of brown coal briquetting press moulds by arc welding. The welding rod has a diameter of 8 mm and can be bent to a radius of less than 150 mm for reeling. The welding rod is produced by mixing 9% plasticizer (Miravithen and polyisobutylene according to GDR patent 203 269) to the steel welding powder. Weldability of the rod proved to be favourable; there was no emission of toxic fumes during welding. Microscopic studies of the welded surface coating showed that welding with 650A achieved the best coat pore structure. At the Schwarze Pumpe Gasworks the trial service life of various briquet press moulds, reinforced with Ferroplast ZIS 218, proved to be not shorter than that of moulds reinforced with the conventional ZIS powder welding method. 1 reference.

  10. Temperature analysis of the control rods at the scram shutdown of the HTTR. Evaluation by using measurement data at scram test of HTTR

    Energy Technology Data Exchange (ETDEWEB)

    Takada, Eiji; Fujimoto, Nozomu; Nakagawa, Shigeaki [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Matsuda, Atsuko [Toshiba Co., Tokyo (Japan)

    2003-03-01

    In the High Temperature Engineering Test Reactor (HTTR), since the primary coolant temperature become 950 degrees centigrade at the high temperature test operation, the special alloy Alloy800H is used for cladding tubes and spines of the control rods to endure the high temperature. The temperature limitation of control rod is 900 degrees centigrade according to the strength data of Alloy800H. The scram shutdown by loss of off-site electric power at the high temperature test operation was assumed as an transient of the temperature of the control rods cladding might exceed 900 degrees centigrade. In this report, the temperature of the control rods is analyzed by using the measurement data of the rise-to-power test. From the result of this analysis, it was confirmed that the control rod temperature does not exceed the limit even at the transient of the loss of off-site electric power from the high temperature test operation. (author)

  11. External attachment of titanium sheathed thermocouples to zirconium nuclear fuel rods for the loss-of-fluid-test (LOFT) Reactor

    International Nuclear Information System (INIS)

    Welty, R.K.

    1980-01-01

    A welding process to attach titanium sheathed thermocouples to the outside of the zircaloy clad fuel rods has been developed. A laser beam was selected as the optimum welding process because of the extremely high energy input per unit volume that can be achieved allowing local fusion of a small area irrespective of the difference in material thickness to be joined. Irradiation tests showed no degradation of thermocouples or weld structure. Fast thermal cycle and heater rod blowdown reflood tests were made to subject the weldments to high temperatures, high pressure steam, and fast water quench cycles. From the behavior of these tests, it was concluded that the attachment welds would survive a series of reactor safety tests. 2 refs

  12. Pre-test prediction and post-test analysis of PWR fuel rod ballooning in the MT-3 in-pile LOCA simulation experiment in the NRU reactor

    International Nuclear Information System (INIS)

    Donaldson, A.T.; Horwood, R.A.; Healey, T.

    1983-01-01

    The USNRC and the UKAEA have jointly funded a series of in-pile LOCA simulation experiments in the Canadian NRU reactor in order to secure further information on the thermal hydraulic and clad deformation response of PWR fuel rod bundles. Test MT-3 in the series was performed using reflood rate and rod internal pressure conditions specified by the UK nuclear industry. The parameters were selected to ensure the development of a near-isothermal clad temperature history during which zircaloy was required to balloon and rupture near the alpha-alpha/beta phase transition. Specification of the reflood rate conditions was assisted by the performance of a precursor test on an unpressurised rod bundle and by complementary application of appropriate thermal hydraulic analyses. Identification of the rod internal pressure needed to cause ballooning and rupture was achieved using a creep deformation model, BALLOON, in conjunction with the clad thermal history defined by the prior thermal hydraulic test. This paper presents the basis of the BALLOON analysis and describes its application in calculating the fill gas pressure for rods MT-3, their axial ballooning profile and the clad temperature at peak radial strain elevations. (author)

  13. Hoisting and Rigging (Formerly Hoisting and Rigging Manual)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-06-01

    This standard is intended as a reference document to be used by supervisors, line managers, safety personnel, equipment operators, and any other personnel responsible for safety of hoisting and rigging operations at DOE sites. It quotes or paraphrases the US OSHA and ANSI requirements. It also encompasses, under one cover,hoisting and rigging requirements, codes, standards, and regulations, eliminating the need to maintain extensive (and often incomplete) libraries of hoisting and rigging standards throughout DOE. The standard occasionally goes beyond the minimum general industry standards established by OSHA and ANSI, and also delineates the more stringent requirements necessary to accomplish the complex, diversified, critical, and often hazardous hoisting and rigging work found with the DOE complex.

  14. Temperature escalation in PWR fuel rod simulators due to the zircaloy/steam reaction: Tests ESSI-1,2,3

    International Nuclear Information System (INIS)

    Hagen, S.; Malauschek, H.; Wallenfels, K.P.; Peck, S.O.

    1983-08-01

    This report discusses the test conduct, results, and posttest appearance of three scoping tests (ESSI-1,2,3) investigating temperature escalation in zircaloy clad fuel rods. The experiments are part of an out-of-pile program using electrically heated fuel rod simulators to investigate PWR fuel element behavior up to temperatures of 2000 0 C. These experiments are part of the PNS Severe Fuel Damage Program. The temperature escalation is caused by the exothermal zircaloy/steam reaction, whose reaction rate increases exponentially with the temperature. The tests were performed using different initial oxide layers as a major parameter, obtained by varying the heatup rates and steam exposure times. (orig./RW) [de

  15. Firmware development and testing of the ATLAS Pixel Detector / IBL ROD card

    CERN Document Server

    Gabrielli, Alessandro; The ATLAS collaboration; Balbi, Gabriele; Bindi, Marcello; Chen, Shaw-pin; Falchieri, Davide; Flick, Tobias; Hauck, Scott Alan; Hsu, Shih-Chieh; Kretz, Moritz; Kugel, Andreas; Lama, Luca; Travaglini, Riccardo; Wensing, Marius; ATLAS Pixel Collaboration

    2015-01-01

    The ATLAS Experiment is reworking and upgrading systems during the current LHC shut down. In particular, the Pixel detector has inserted an additional inner layer called Insertable B-Layer (IBL). The Readout-Driver card (ROD), the Back-of-Crate card (BOC), and the S-Link together form the essential frontend data path of the IBL’s off-detector DAQ system. The strategy for IBL ROD firmware development was three-fold: keeping as much of the Pixel ROD datapath firmware logic as possible, employing a complete new scheme of steering and calibration firmware and designing the overall system to prepare for a future unified code version integrating IBL and Pixel layers. Essential features such as data formatting, frontend-specific error handling, and calibration are added to the ROD data path. An IBL DAQ testbench using realistic frontend chip model was created to serve as an initial framework for full offline electronic system simulation. In this document, major firmware achievements concerning the IBL ROD data pat...

  16. A drilling rig tower

    Energy Technology Data Exchange (ETDEWEB)

    Mironov, A.A.; Barashkov, V.A.; Bulgakov, E.S.; Kuldoshin, I.P.; Lebedev, A.I.; Papin, N.M.; Rebrik, B.M.; Sirotkin, N.V.

    1981-05-23

    Presentation is made of a drilling rig tower, comprising a gantry, a support shaft with a bracing strut and drawings out, and turn buckles. In order to increase the reliability of the tower in operation, to decrease the over all dimensions in a transport position, and to decrease the amount of time taken to transfer the tower from an operational position into a transportable one, and vice versa, the tower is equipped with a rotary frame made in the form of a triangular prism, whose lateral edges are connected by hinges: the first one with the lower part of the support shaft, the second with the gantry, and the third one to the upper part of the support shaft by means of the drawings out. The large boundary of the rotary frame is connected by a hinge to the support shaft by means of a bracing strut, which is equipped with a slide block connected to it by a hinge, and the rotary frame has a guide for the slide block reinforced to it on the large boundary. Besides this, the lateral edge of the rotary frame is connected to the gantry by means of turn buckles.

  17. Experimental investigation of cooling by top spray and bottom flooding of a simulated 64 rod bundle for a BWR. Pt. 2. Main experiment with modified test section

    International Nuclear Information System (INIS)

    Nilsson, L.; Gustafson, L.; Harju, R.

    1978-06-01

    The cooling of an electrically heated, full scale 64-rod bundle has been investigated under simulated emergency core cooling conditions. Emphasis was laid on measurements of rod cladding and canister temperatures. By means of difference pressure measurements the levels in bundle, by-pass and downcomer could be estimated and thus the effective reflooding velocity. The test section was modified compared to the pre-tests, in order to improve system effects simulation. A new rod bundle was installed including a hollow, water, rod and 63 indirectly heated rods. Parameter effects of coolant mass flow rate and distribution, initial cladding temperature, pressure and power were studied. The effect of the way the test section was vented was also investigated and turned out to be very significant. (author)

  18. Control rods

    International Nuclear Information System (INIS)

    Maruyama, Hiromi.

    1984-01-01

    Purpose: To realize effective utilization, cost reduction and weight reduction in neutron absorbing materials. Constitution: Residual amount of neutron absorbing material is averaged between the top end region and other regions of a control rod upon reaching to the control rod working life, by using a single kind of neutron absorbing material and increasing the amount of the neutron absorber material at the top end region of the control rod as compared with that in the other regions. Further, in a case of a control rod having control rod blades such as in a cross-like control rod, the amount of the neutron absorbing material is decreased in the middle portion than in the both end portions of the control rod blade along the transversal direction of the rod, so that the residual amount of the neutron absorbing material is balanced between the central region and both end regions upon reaching the working life of the control rod. (Yoshihara, H.)

  19. Development of carbon/carbon composite control rod for HTTR. 1. Preparation of elements and their fracture tests

    International Nuclear Information System (INIS)

    Eto, Motokuni; Ishiyama, Shintaro; Ugachi, Hirokazu

    1996-08-01

    For the High Temperature Engineering Test Reactor(HTTR) the control rod sleeve is made of Alloy 800H for which a particular process is imposed when the reactor needs to be scrammed. The less restricted operation of the reactor would be attained if there would be the control rod more resistant to high temperature and neutron irradiation. This report summarizes the results which have been obtained as of March 1996 in the course of the development of the C/C composite control rod. Materials used were pitch- or PAN-based fiber-reinforced 2-dimensional carbon composites, from which preforms of the elements of a control rod were fabricated. The preforms were carbonized at 1000degC after being impregnated with pitch. Then they were graphitized at 3000degC, followed by a purification treatment with halogen. The elements included the pellet holder, lace truck and pin. The pin was fabricated by the fiber laminating technique. A control rod is to consist of pellet holders which are connected by the lace trucks with pins. Various strength tests were carried out on these elements. An irradiation of the elements made of PAN-based material was performed in JRR-3 at 900±50degC to a neutron fluence of 1x10 25 n/m 2 (E>29fJ). As for the strength tests on the elements, there were some differences between PAN- and pitch-based composites: In general, elements made of PAN-based composite showed the more plastic behavior before they fractured, whereas those of pitch-based material behaved in the more brittle manner. Fracture tests of the irradiated elements showed that fracture load and fracture displacement enough for assuring the integrity of the control rod structure were maintained even after the irradiation. It was also found that if the applied load was parallel to the fiber felt plane both fracture load and strain increased, whereas the load increase and strain decrease were observed for the applied load against the plane. (J.P.N.)

  20. Development of carbon/carbon composite control rod for HTTR. 1. Preparation of elements and their fracture tests

    Energy Technology Data Exchange (ETDEWEB)

    Eto, Motokuni; Ishiyama, Shintaro; Ugachi, Hirokazu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1996-08-01

    For the High Temperature Engineering Test Reactor(HTTR) the control rod sleeve is made of Alloy 800H for which a particular process is imposed when the reactor needs to be scrammed. The less restricted operation of the reactor would be attained if there would be the control rod more resistant to high temperature and neutron irradiation. This report summarizes the results which have been obtained as of March 1996 in the course of the development of the C/C composite control rod. Materials used were pitch- or PAN-based fiber-reinforced 2-dimensional carbon composites, from which preforms of the elements of a control rod were fabricated. The preforms were carbonized at 1000degC after being impregnated with pitch. Then they were graphitized at 3000degC, followed by a purification treatment with halogen. The elements included the pellet holder, lace truck and pin. The pin was fabricated by the fiber laminating technique. A control rod is to consist of pellet holders which are connected by the lace trucks with pins. Various strength tests were carried out on these elements. An irradiation of the elements made of PAN-based material was performed in JRR-3 at 900{+-}50degC to a neutron fluence of 1x10{sup 25} n/m{sup 2} (E>29fJ). As for the strength tests on the elements, there were some differences between PAN- and pitch-based composites: In general, elements made of PAN-based composite showed the more plastic behavior before they fractured, whereas those of pitch-based material behaved in the more brittle manner. Fracture tests of the irradiated elements showed that fracture load and fracture displacement enough for assuring the integrity of the control rod structure were maintained even after the irradiation. It was also found that if the applied load was parallel to the fiber felt plane both fracture load and strain increased, whereas the load increase and strain decrease were observed for the applied load against the plane. (J.P.N.)

  1. Equipment available for automating rig operations

    International Nuclear Information System (INIS)

    McNair, W.L.

    1990-01-01

    Several manufacturers are producing automated rig equipment, from complete systems to individual functions for existing drilling rigs. Significant improvements in well site time, costs of operations, and improved drilling performance have led drilling contractors to install this equipment on their rigs. This paper details some of the equipment available for automating rigs

  2. Experimental rigs for MHD studies

    International Nuclear Information System (INIS)

    Venkataramani, N.; Jayakumar, R.; Iyer, D.R.; Dixit, N.S.

    1976-01-01

    An MHD experimental rig is a miniature MHD installation consisting of basic equipments necessary for specific investigations. Some of the experimental rigs used in the investigations being carried out at the Bhabha Atomic Research Centre, Bombay (India) are dealt with. The experiments included diagnostics and evaluation of materials in seeded combustion plasmas and argon plasmas. The design specifications, schematics and some of the results of the investigations are also mentioned. (author)

  3. Temperature escalation in PWR fuel rod simulator bundles due to the Zircaloy/steam reaction: Test ESBU-2A

    International Nuclear Information System (INIS)

    Hagen, S.; Kapulla, H.; Malauschek, H.; Wallenfels, K.P.; Peck, S.O.

    1984-07-01

    This report describes the test conduct and results of the bundle test ESBU-2A, which was run to investigate the temperature escalation of zircaloy clad fuel rods. This investigation of temperature escalation is part of a series of out-of-pile experiments, performed within the framework of the PNS Severe Fuel Damage Program. The test bundle was of a 3 x 3 array of fuel rod simulators with a 0.4 m heated length. The fuel rod simulators were electrically heated and consisted of tungsten heaters, UO 2 annular pellets, and zircaloy cladding. A nominal steam flow of 0.7 g/s was inlet to the bundle. The bundle was surrounded by a zircaloy shroud which was insulated with ZrO 2 fiber ceramic wrap. The initial heatup rate of the bundle was 0.4 0 C/s. The temperature escalation began at the 255 mm elevation after 1200 0 C had been reached. At this elevation, the measured peak temperature was limited to 1500 0 C. It was concluded from different thermocouple results, that induced by this first escalation melt was formed in the lower part of the bundle. Consequently, the escalation in the lower part must be much higher, at least up to the melting temperature of zircaloy. Due to the failure in the steam production system, steam starvation in the upper region may explain the beginning of the escalation at the 255 mm elevation. The maximum temperature reached was 2175 0 C on the center rod at the end of the test. The unregularities in the steam supply may be the reason for less oxidation than expected. (orig./GL) [de

  4. An operational 150 kV microfocus rod anode X-ray system for nondestructive testing

    International Nuclear Information System (INIS)

    Fontijn, E.A.

    1978-01-01

    This paper describes an operational state of the art 150 kV microfocus rod anode X-ray system having ultra-high radiographic resolution capabilities. A cocal spot size of 0.050 mm is provided. Heretofore unattainable long rod anode lengths coupled with very small diameters are now possible using mini-magnetic lens technology. Over-all rod anode diameters as small as 9 mm with useful lengths of 1 m or more are possible, permitting panoramic inspections where previously only lower resolution radioisotope radiographic techniques were possible. Radiographic sensitivity of better than 1% has been reported with film-focal-distances on the order of 8 mm through 3 mm of steel. The system has been successfully applied to steam generator and heat exchanger tube-to-tubesheet weldments in both Europe and the USA. Other application areas include marine and aircraft jet engine inspection and numerous other applications where high reliability requirements indicate the use of a ultra-sensitive radiographic technique as is routinely demonstrated with the 150 kV Microfocus Rod Anode X-ray System. (orig.) [de

  5. Guided-Wave Testing of Trunnion Rods at Greenup Lock and Dam, Kentucky

    Science.gov (United States)

    2014-04-01

    fatigue cracks which go through opening and closing periods during their progression, microcracks in post-tension trunnion anchor rods are believed to be...produce undesired interference effects. A new 1.5 in. diameter, medium-damped lead zirconate titanate ( PZT ), ceramic-based crystal (i.e., an Accuscan

  6. Evaluation of the ability of rod drop tests to verify the stability margins in FTR

    International Nuclear Information System (INIS)

    Harris, R.A.; Sevenich, R.A.

    1976-01-01

    Predictions of the stability characteristics of FTR indicate that the reactor can be easily controlled even under the worst possible conditions. Nevertheless, experimental verification and monitoring of these characteristics will be performed during operation of the reactor. An initial evaluation of rod drop experiments which could possibly provide this verification is presented

  7. Power ramp performance of some 15 x 15 PWR test fuel rods tested in the STUDSVIK SUPER-RAMP and SUPER-RAMP extension projects

    International Nuclear Information System (INIS)

    Djurle, S.

    2000-01-01

    This paper presents results obtained from the STUDSVIK SUPER-RAMP (SR) and SUPER-RAMP EXTENSION (SRX) projects. As parts of these projects test fuel rods of the same PWR type were base irradiated in the Obrigheim power reactor and power ramp tested in the STUDSVIK R2 reactor. Some of the rods were ramped using an inlet coolant water temperature 50 deg. C below the normal one. Fabricated data on the test fuel rods are presented as well as data on the base irradiation, interim examination, conditioning irradiation, power ramp irradiation and results of the post irradiation examination. The data on the change of diameter at ridges due to power ramping have shown that a lower clad temperature during ramping leads to smaller deformations. Most likely this may be explained as due to a smaller creep rate in the cladding at the lower temperature, resulting in a more severe stress situation. The combination of low cladding temperature, high ramp terminal level and the presence of a stress corrosion agent may have caused the failure of one of the test rods. (author)

  8. Temperature escalation in PWR fuel rod simulator bundles due to the zircaloy/steam reaction: Test ESBU-1

    International Nuclear Information System (INIS)

    Hagen, S.; Malauschek, H.; Peck, S.O.; Wallenfels, K.P.

    1983-12-01

    This report describes the test conduct and results of the bundle test ESBU-1. The test objective was the investigation of temperature escalation of zircaloy clad fuel rods. The investigation of the temperature escalation is part of a program of out-of-pile experiments, performed within the framework of the PNS Several Fuel Damage Program. The bundle was composed of a 3x3 array of fuel rod simulators surrounded by a zircaloy shroud which was insulated with a ZrO 2 fiber ceramic wrap. The fuel rod simulators comprised a tungsten heater, UO 2 annular pellets, and zircaloy cladding over a 0.4 m heated length. A steam flow of 1 g/s was inlet to the bundle. The most pronounced temperature escalation was found on the central rod. The initial heatup rate of 2 0 C/s at 1100 0 C increased to approximately 6 0 C/s. The maximum temperature reached was 2250 0 C. The following fast temperature decrease was caused by runoff of molten zircaloy. Molten zircaloy swept down the thin cladding oxide layer formed during heatup. The melt dissolved the surface of the UO 2 pellets and refroze as a coherent lump in the lower part of the bundle. The remaining pellets fragmented during cooldown and formed a powdery layer on the refrozen lump. The lump was sectioned posttest at several elevations: Dissolution of UO 2 by the molten zircaloy, interaction between the melt and previously oxidized zircaloy, and oxidation of the melt had occurred. (orig.) [de

  9. The Third Dryout Fuel Behaviour Test Series in IFA-613

    International Nuclear Information System (INIS)

    Ianiri, Raffaella

    1998-02-01

    The objective of the dryout experiment with the instrumented fuel assembly IFA-613 is to provide information on the consequences induced on fuel by short terms dry outs having characteristics similar to those anticipated to occur from pump trips in a Boiling Water Reactor (BWR). For the third experiment it was planned to test one fresh and two pre-irradiated segments. Unfortunately one of the channels, Channel A developed a leakage and was not suitable for testing anymore. The rig was loaded with only two rods: one fresh PWR rod with a design similar to the fresh rod in IFA-613.1 and one pre-irradiated PWR segment (N1310 with a burn-up of 29 MWd/kgU). Both rods were equipped with a clad extensometer and two clad surface thermocouples (upper and lower position). The rig was loaded during the December 1997 shutdown and the dryout tests were performed on 16th January 1998. Both rods experienced temperature excursions with a target peak clad temperature (PCT) of 650 o C. According to the measured cladding temperatures, the time above the target temperature was about 4-5 s for both rods. The lower thermocouple did not indicate dryout at any occasion. The rig was unloaded immediately after the testing. (author)

  10. Jet Exit Rig Six Component Force Balance

    Science.gov (United States)

    Castner, Raymond; Wolter, John; Woike, Mark; Booth, Dennis

    2012-01-01

    A new six axis air balance was delivered to the NASA Glenn Research Center. This air balance has an axial force capability of 800 pounds, primary airflow of 10 pounds per second, and a secondary airflow of 3 pounds per second. Its primary use was for the NASA Glenn Jet Exit Rig, a wind tunnel model used to test both low-speed, and high-speed nozzle concepts in a wind tunnel. This report outlines the installation of the balance in the Jet Exit Rig, and the results from an ASME calibration nozzle with an exit area of 8 square-inches. The results demonstrated the stability of the force balance for axial measurements and the repeatability of measurements better than 0.20 percent.

  11. Control rod

    International Nuclear Information System (INIS)

    Igarashi, Takao; Sugawara, Satoshi; Yoshimoto, Yuichiro; Saito, Shozo; Fukumoto, Takashi.

    1987-01-01

    Purpose: To reduce the weight and thereby obtain satisfactory operationability of control rods by combining absorbing nuclear chain type neutron absorbers and conventional type neutron absorbers in the axial direction of blades. Constitution: Neutron absorber rods and long life type neutron absorber rods are disposed in a tie rod and a sheath. The neutron absorber rod comprises a poison tube made of stainless steels and packed with B 4 C powder. The long life type neutron absorber rod is prepared by packing B-10 enriched boron carbide powder into a hafnium metal rod, hafnium pipe, europium and stainless made poison tube. Since the long life type absorber rod uses HF as the absorbing nuclear chain type neutron absorber, it absorbs neutrons to form new neutron absorbers to increase the nuclear life. (Yoshino, Y.)

  12. Flow mixing inside a control-rod guide tube – Experimental tests and CFD simulations

    International Nuclear Information System (INIS)

    Angele, Kristian; Odemark, Ylva; Cehlin, Mathias; Hemström, Bengt; Högström, Carl-Maikel; Henriksson, Mats; Tinoco, Hernan; Lindqvist, Hans

    2011-01-01

    This paper covers a combined experimental and computational effort carried out at Vattenfall Research and Development AB in order to study the thermal mixing in the annular region between a top tube and a control-rod stem. The low frequency thermal fluctuations in this region can result in problems with thermal fatigue and have caused cracks in the control-rod stems of several nuclear reactors (). The flow in the vertical annular region formed by the top tube and the control-rod stem is characterized by the mixing of hot bypass flow with cold crud-removal flow. The crud-removal flow is flowing upwards along the control-rod stem, and the warmer bypass flow is entering through eight horizontal holes positioned in the lower part of the guide tube and four holes in the upper part of the top tube, forming jets. Two full-scale models of a control rod, including the control-rod stem and the guide tube, were constructed. The first model, designed to work at atmospheric conditions, was made of Plexiglass, in order to be able to visualize the mixing process, whereas the second one was made of steel to allow for a higher temperature difference between the two flows, and the heating of the top tube. CFD simulations of the case at atmospheric conditions were also carried out. Both the experiments and the simulations showed that the mixing region between the cold crud-removal flow and the warm bypass flow is dominated by large flow structures coming from above. The process is characterized by low frequency, high amplitude temperature fluctuations. The process is basically hydrodynamic, caused by the downward transport of flow structures originated at the upper bypass inlets. The damping thermal effects through buoyancy is of secondary importance, as also the scaling analysis shows, however a slight damping of the temperature fluctuations can be seen due to natural convection due to a pre-heating of the cold crud-removal flow. The comparison between numerical and experimental

  13. The high temperature out-of-pile test of LVDT for internal pressure measurement of nuclear fuel rod

    Energy Technology Data Exchange (ETDEWEB)

    Son, J. M.; Kim, B. K.; Kim, D. S.; Yoon, K. B.; Sin, Y. T.; Park, S. J.; Kang, Y. H. [KAERI, Taejon (Korea, Republic of)

    2002-10-01

    As a part of the development of instrumentation technologies for the nuclear fuel irradiation test in HANARO(High-flux Advanced Nuclear Application Reactor), the internal pressure measurement technique of the nuclear fuel rod is being developed using LVDT(Linear Variable Differential Transformer). As the results of out-of-pile test at room temperature, it was concluded that the well qualified out-of-pile tests were needed to understand the LVDT's detail characteristics at high temperature for the detail design of the fuel irradiation capsule, because LVDT is very sensitive to variation of temperature. Therefore, the high temperature out-of-pile test system for pressure measurement was developed, and this test was performed under the temperature condition between room temperature and 300 .deg. C increasing the pressure from 0 bar to 30 bar. The LVDT's high temperature characteristics and temperature sensitivity of LVDT were analyzed through this experiment. Based on the result of this test, the method for the application of LVDT at high temperature was introduced. It is known that the results will be used to predict accurately the internal pressure of fuel rod during irradiation test.

  14. Rig`s electricity to power top drive drilling system

    Energy Technology Data Exchange (ETDEWEB)

    Liderth, D.

    1996-05-01

    Permanent magnet brushless electric motors to supply torque to more space-efficient top drive drilling assemblies was the solution designed by Kaman Electromagnetic Corporation, working hand-in-hand with Calgary-based Tesco Drilling Technology, to remedy problems created by the bulkiness of standard hydraulic top drive systems. The biggest advantage of using electric over hydraulic top drive systems is the ability to tap into the rig`s existing power source, which lowers both cost and effort. A better power to weight ratio and reduced maintenance requirements are other desirable advantages.

  15. Computers make rig life extension an option

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-10-01

    The worldwide semisubmersible drilling rig fleet is approaching retirement. But replacement is not an attractive option even though dayrates are reaching record highs. In 1991, Schlumberger Sedco Forex managers decided that an alternative might exist if regulators and insurers could be convinced to extend rig life expectancy through restoration. Sedco Forex chose their No. 704 semisubmersible, an 18-year North Sea veteran, to test their process. The first step was to determine what required restoration, meaning fatigue life analysis of each weld on the huge vessel. If inspected, the task would be unacceptably time-consuming and of questionable accuracy. Instead a suite of computer programs modeled the stress seen by each weld, statistically estimated the sea states seen by the rig throughout its North Sea service and calibrated a beam-element model on which to run their computer simulations. The elastic stiffness of the structure and detailed stress analysis of each weld was performed with ANSYS, a commercially available finite-element analysis program. The use of computer codes to evaluate service life extension is described.

  16. CFD simulation and validation of turbulent mixing in a rod bundle with vaned spacer grids based on LDV test

    International Nuclear Information System (INIS)

    Chen Xi; Li Songwei; Li Zhongchun; Du Sijia; Zhang Yu; Peng Huanhuan

    2017-01-01

    Spacer grids with mixing vanes are generally used in fuel assemblies of Pressurized Water Reactor (PWR), because that mixing vanes could enhance the lateral turbulent mixing in subchannels. Thus, heat exchangements are more efficient, and the value of departure from nucleate boiling (DNB) is greatly increased. Actually turbulent mixing is composed of two kinds of flows: swirling flow inside the subchannel and cross flow between subchannels. Swirling flow could induce mixing between hot water near the rod and cold water in the center of the subchannel, and may accelerate deviation of the bubbles from the rod surface. Besides, crossing flow help to mixing water between hot subchannels and cold subchannels, which impact relatively large flow area. As a result, how to accurately capture and how to predict the complicated mixing phenomenon are of great concernments. Recently many experimental studies has been conducted to provide detailed turbulent mixing in rod bundle, among which Laser Doppler Velocimetry method is widely used. With great development of Computational Fluid Dynamics, CFD has been validated as an analysis method for nuclear engineering, especially for single phase calculation. This paper presents the CFD simulation and validation of the turbulent mixing induced by spacer grid with mixing vanes in rod bundles. Experiment data used for validation came from 5 x 5 rod bundle test with LDV technology, which is organized by Science and Technology on Reactor System Design Technology Laboratory. A 5 x 5 rod bundle with two spacer grids were used. Each rod has dimension of 9.5 mm in outer diameter and distance between rods is 12.6 mm. Two axial bulk velocities were conducted at 3.0 m/s for high Reynolds number and 1.0 m/s for low Reynolds number. Working pressure was 1.0 bar, and temperature was about 25degC. Two different distances from the downstream of the mixing spacer grid and one from upstream were acquired. Mean axial velocities and turbulent intensities

  17. Experimental study and FEM simulation of the simple shear test of cylindrical rods

    Science.gov (United States)

    Wirti, Pedro H. B.; Costa, André L. M.; Misiolek, Wojciech Z.; Valberg, Henry S.

    2018-05-01

    In the presented work an experimental simple shear device for cutting cylindrical rods was used to obtain force-displacement data for a low-carbon steel. In addition, and FEM 3D-simulation was applied to obtain internal shear stress and strain maps for this material. The experimental longitudinal grid patterns and force-displacement curve were compared with numerical simulation results. Many aspects of the elastic and plastic deformations were described. It was found that bending reduces the shear yield stress of the rod material. Shearing starts on top and bottom die-workpiece contact lines evolving in an arc-shaped area. Due to this geometry, stress concentrates on the surface of the rod until the level of damage reaches the critical value and the fracture starts here. The volume of material in the plastic zone subjected to shearing stress has a very complex shape and is function of a dimensionless geometrical parameter. Expressions to calculate the true shear stress τ and strain γ from the experimental force-displacement data were proposed. The equations' constants are determined by fitting the experimental curve with the stress τ and strain γ simulation point tracked data.

  18. Firmware development and testing of the ATLAS Pixel Detector / IBL ROD card

    CERN Document Server

    Balbi, G; The ATLAS collaboration; Gabrielli, A; Lama, L; Travaglini, R; Backhaus, M; Bindi, M; Chen, S-P; Flick, T; Kretz, M; Kugel, A; Wensing, M

    2014-01-01

    The ATLAS Experiment is reworking and upgrading systems during the current LHC shut down. In particular, the Pixel detector has inserted an additional inner layer called Insertable B-Layer (IBL). The Readout-Driver card (ROD), the Back-of-Crate card (BOC), and the S-Link together form the essential frontend data path of the IBL’s off-detector DAQ system. The strategy for IBLROD firmware development was three-fold: keeping as much of the PixelROD datapath firmware logic as possible, employing a complete new scheme of steering and calibration firmware and designing the overall system to prepare for a future unified code version integrating IBL and Pixel layers. Essential features such as data formatting, frontend-specific error handling, and calibration are added to the ROD data path. An IBLDAQ testbench using realistic frontend chip model was created to serve as an initial framework for full offline electronic system simulation. In this document, major firmware achievements concerning the IBLROD data path im...

  19. Assessment of CCFL model of RELAP5/MOD3 against simple vertical tubes and rod bundle tests

    International Nuclear Information System (INIS)

    Cho, Sung Jae; Arne, Nam Sung; Chung, Bub Dong; Kim, Hho Jung

    1991-01-01

    The CCFL model used in RELAP5/MOD3 version 5m5 has been assessed against simple vertical tubes and rod bundle tests performed at a facility of Korea Atomic Energy Research Institute. The effect of changes in tube diameter and nodalization of tube section were investigated. The roles of interfacial drags on the flooding characteristics are discussed. Difference between the calculation and the experiment are also discussed. A comparison between model assessment results and the test data showed that the calculated value lay well on the experimental flooding curve specified by user, but the pressure jump before onset of flooding was not calculated

  20. A study of natural circulation cooling using a flow visualization rig

    International Nuclear Information System (INIS)

    Bowman, W.C.; Ferch, R.L.; Omar, A.M.

    1985-01-01

    A flow visualization rig has been built at Monserco Limited to provide visual insight into the thermalhydraulic phenomena which occur during single phase and two phase thermosyphoning in a figure-of-eight heat transport loop. Tests performed with the rig have provided design information for the scaling and instrumentation of a high pressure rig being investigated for simulating CANDU reactor conditions during natural circulation cooling. A videotape was produced, for viewing at this presentation, to show important thermalhydraulic features of the thermosyphoning process. The rig is a standard figure-of-eight loop with two steam generators and three heated channels per pass. An elevated surge tank open to atmosphere was used for pressure control. Two variable speed pumps provided forced circulation for warming up the rig, and for establishing the desired initial conditions for testing. Test rig power could be varied between 0 and 15 kW

  1. Assessment of the linear power level in fuel rods irradiated in the CALLISTO loop in the high flux materials testing reactor BR2

    International Nuclear Information System (INIS)

    Malambu, E.; Raedt, Ch. de; Weber, M.

    1999-01-01

    The pressurized light-water-cooled testing facility CALLISTO was designed to test the behaviour of advanced fuel rods (UO 2 or MOX, possibly with burnable poisons) under conditions representative of actual LWRs up to high burn-up rates. The accurate determination of the fission powers in each of the nine rods, and hence of the burn-up values, is carried out according to a rather elaborate procedure. (author)

  2. Acceleration Test Method for Failure Prediction of the End Cap Contact Region of Sodium Cooled Fast Reactor Fuel Rod

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyung-Kyu; Lee, Young-Ho; Lee, Hyun-Seung; Lee, Kang-Hee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2017-05-15

    This paper reports the results of an acceleration test to predict the contact-induced failure that could occur at the cylinder-to-hole joint for the fuel rod of a sodium-cooled fast reactor (SFR). To incorporate the fuel life of the SFR currently under development at KAERI (around 35,000 h), the acceleration test method of reliability engineering was adopted in this work. A finite element method was used to evaluate the flow-induced vibration frequency and amplitude for the test parameter values. Five specimens were tested. The failure criterion during the life of the SFR fuel was applied. The S-N curve of the HT-9, the material of concern, was used to obtain the acceleration factor. As a result, a test time of 16.5 h was obtained for each specimen. It was concluded that the B{sub 0.004} life would be guaranteed for the SFR fuel rods with 99% confidence if no failure was observed at any of the contact surfaces of the five specimens.

  3. Breakdown voltage at the electric terminals of GCFR-core flow test loop fuel rod simulators in helium and air

    International Nuclear Information System (INIS)

    Huntley, W.R.; Conley, T.B.

    1979-12-01

    Tests were performed to determine the ac and dc breakdown voltage at the terminal ends of a fuel rod simulator (FRS) in helium and air atmospheres. The tests were performed at low pressures (1 to 2 atm) and at temperatures from 20 to 350 0 C (68 to 660 0 F). The area of concern was the 0.64-mm (0.025-in.) gap between the coaxial conductor of the FRS and the sheaths of the four internal thermocouples as they exit the FRS. The tests were prformed to ensure a sufficient safety margin during Core Flow Test Loop (CFTL) operations that require potentials up to 350 V ac at the FRS terminals. The primary conclusion from the test results is that the CFTL cannot be operated safely if the terminal ends of the FRSs are surrounded by a helium atmosphere but can be operated safely in air

  4. Measurement Report for the Four-Rod LHC Crab Cavity. Cold Tests held in July 2014

    CERN Document Server

    Navarro Tapia, Maria; Calaga, Rama; Hernandez Chahin, Karim Gibran; Junginger, Tobias; Macpherson, Alick; Torres-Sanchez, Roberto; CERN. Geneva. ATS Department

    2015-01-01

    The performance of the four-rod cavity prototype considered for the HL-LHC upgrade has already been assessed at CERN at cryogenic temperatures three times in the last two years [1, 2, 3]. In this report, the results of the latest measurements, carried out in July 2014, are shown. These measurements were to check the improvement of the cavity performance due to the change of the input and pick-up antennas. An estimation of the residual resistance of the Niobium was also performed.

  5. Design and test of the borosilicate glass burnable poison rod for Qinshan nuclear power plant core

    International Nuclear Information System (INIS)

    Huang Jinhua; Sun Hanhong

    1988-08-01

    Material for the burnable poison of Qinshan Nuclear Power Plant core is GG-17 borosilicate glass. The chemical composition and physico-chemical properties of GG-17 is very close to Pyrex-7740 glass used by Westinghouse. It is expected from the results of the experiments that the borosilicate glass burnable poison rod can be successfully used in Qinshan Nuclear Power Plant due to good physical, mechanical, corrosion-resistant and irradiaton properties for both GG-17 glass and cold-worked stainless steel cladding. Change of material for burnable poison from boron-bearing stainless steel to borosilicate glass will bring about much more economic benefit to Qinshan Naclear Power Plant

  6. Sea testing and optimisation of power production on a scale 1:4.5 test rig of the offshore wave energy converter wave dragon. Summary of final technical report

    Energy Technology Data Exchange (ETDEWEB)

    2006-06-15

    The 4-11 MW Wave Dragon is a slack moored device that can be deployed in large parks wherever a sufficient wave climate and a water depth of more than 20 m is found--typically this is the case in the North Sea and in the Atlantic, offering significant economic and environmental benefits for the EU. The primary objective of the project was to establish the scientific knowledge base needed for deploying a full-scale prototype of the overtopping wave energy converter Wave Dragon. This has been obtained through long-term field-testing on a test rig with all systems installed. The scale 1:4.5 prototype has an installed power of 20 kW corresponding to 4 MW in full-scale with full-turbine deployment and is grid connected. The scale 1:4.5 prototype has been designed based on the conclusions from a previous EU Craft project. The basic test rig construction is provided through a project sponsored by the Danish Energy Authority. The test site is in protected waters in Nissum Bredning, Denmark, where the wave climate resembles North Sea conditions (scale 1:4.5) which in accordance with model law resembles a power scale of 1:200. The test results after more than 20,000 hours of operation cover: Long-term field testing of turbine operation, control strategy testing and optimisation, power monitoring and evaluation, stress and strain measurements and analysis, and mooring and cable systems analysis. The model tools developed in the previous EU Craft project have been validated and slightly modified based on the measured data. A Life Cycle Analysis and Finite Element Modelling have been performed. A report on market analysis, economic risk assessment and job creation potential has also been carried out. The project has established the necessary scientific and technical knowledge base for engaging in the establishment of a full-scale prototype in exposed waters. This includes the existence of a well-established design basis and documentation of technical viability through long

  7. Replacement rod

    International Nuclear Information System (INIS)

    Hatfield, S.C.

    1989-01-01

    This patent describes in an elongated replacement rod for use with fuel assemblies of the type having two end fittings connected by guide tubes with a plurality of rod and guide tube cell defining spacer grids containing rod support features and mixing vanes. The grids secured to the guide tubes in register between the end fittings at spaced intervals. The fuel rod comprising: an asymmetrically beveled tip; a shank portion having a straight centerline; and a permanently diverging portion between the tip and the shank portion

  8. Test Facility Construction for Flow Visualization on Mixing Flow inside Subchannels of PWR Rod Bundle

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seok; Jeon, Byong-Guk; Youn, Young-Jung; Choi, Hae-Seob; Euh, Dong-Jin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Flow inside rod bundles has a similarity with flow in porous media. To ensure thermal performance of a nuclear reactor, detailed information of the heat transfer and turbulent mixing flow phenomena taking place within the subchannels is required. The subchannel analysis is one of the key thermal-hydraulic calculations in the safety analysis of the nuclear reactor core. At present, subchannel computer codes are employed to simulate fuel elements of nuclear reactor cores and predict the performance of cores under normal operating and hypothetical accident conditions. The ability of these subchannels codes to predict both the flow and enthalpy distribution in fuel assemblies is very important in the design of nuclear reactors. Recently, according to the modern tend of the safety analysis for the nuclear reactor, a new component scale analysis code, named CUPID, and has been developed in KAERI. The CUPID code is based on a two-fluid and three-field model, and both the open and porous media approaches are incorporated. The PRIUS experiment has addressed many key topics related to flow behaviour in a rod bundle. These issues are related to the flow conditions inside a nuclear fuel element during normal operation of the plant or in accident scenarios. From the second half of 2016, flow visualization will be performed by using a high speed camera and image analysis technique, from which detailed information for the two-dimensional movement of single phase flow is quantified.

  9. PNNL Hoisting and Rigging Manual

    Energy Technology Data Exchange (ETDEWEB)

    Haynie, Todd O.; Fullmer, Michael W.

    2008-12-29

    This manual describes the safe and cost effective operation, inspection, maintenance, and repair requirements for cranes, hoists, fork trucks, slings, rigging hardware, and hoisting equipment. It is intended to be a user's guide to requirements, codes, laws, regulations, standards, and practices that apply to Pacific Northwest National Laboratory (PNNL) and its subcontractors.

  10. Replacement team of mining drilling rigs

    OpenAIRE

    Hamodi, Hussan; Lundberg, Jan

    2014-01-01

    This paper presents a practical model to calculate the optimal replacement time (ORT) of drilling rigs used in underground mining. As a case study, cost data for drilling rig were collected over four years from a Swedish mine. The cost data include acquisition, operating, maintenance and downtime costs when using a redundant rig. A discount rate is used to determine the value of these costs over time. The study develops an optimisation model to identify the ORT of a mining drilling rig which ...

  11. Rigged Hilbert spaces for chaotic dynamical systems

    International Nuclear Information System (INIS)

    Suchanecki, Z.; Antoniou, I.; Bandtlow, O.F.

    1996-01-01

    We consider the problem of rigging for the Koopman operators of the Renyi and the baker maps. We show that the rigged Hilbert space for the Renyi maps has some of the properties of a strict inductive limit and give a detailed description of the rigged Hilbert space for the baker maps. copyright 1996 American Institute of Physics

  12. Hoisting and rigging manual: Uncontrolled document

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1991-05-01

    This document is a draft copy of a Hoisting and Rigging Manual for the Department of Energy. The manual is divided into ten chapters. The chapter titles follow: terminology and definitions; operator training and qualification; overhead and gantry cranes; mobile cranes; forklift trucks; hoists; hooks; wire rope, slings, and rigging accessories; construction hoisting and rigging equipment requirements; references.

  13. NHR dynamic analysis of control rod and fuel assembly of test model

    International Nuclear Information System (INIS)

    Wang Jiachun; Cai Laizhong

    2001-01-01

    The basic purpose is to analyze the dynamic response of the structure, with the seismic excitation, which is the important components of 200 MW Heating Reactor, including the control rod, fuel assembly, zirconium alloy boxes and the relevant parts. The author presents the simplification and building of the model. By comparing the effects under different constraint conditions, the final analyzed model is determined after the preliminary analysis. Then the model is calculated to obtain the frequencies of the model, the analysis of the response spectrum and the time series data under some seismic excitations. From the outcome what is received above, the influence of the basic frequency is discussed. And the displacement and acceleration responses of different sample points are obtained and analyzed to predict the safety of the reactor

  14. Water rod

    International Nuclear Information System (INIS)

    Kashiwai, Shin-ichi; Yokomizo, Osamu; Orii, Akihito.

    1992-01-01

    In a reactor core of a BWR type reactor, the area of a flow channel in a lower portion of a downcoming pipe for downwardly releasing steams present at the top portion in a water rod is increased. Further, a third coolant flow channel (an inner water rod) is disposed in an uprising having an exit opened near the inlet of the water rod and an inlet opened at the outside near the top portion of the water and having an increase flow channel area in the upper portion. The downcoming pipe in the water rod is filled with steams, and the void ratio is increased by so much as the flow channel area of the downcoming pipe is increased. Since the pressure difference between the inlet and the exit of the inner water rod is greater than the pressure difference between the inlet and the exit of the water rod, most of water flown into the inner water rod is discharged out of the exit in the form of water as it is. Since the area of the flow channel is increased in the portion of the inner water rod, void efficiency in the upper portion of the reactor core is decreased by so much. Since the void ratio is thus increased in the lower portion and the void efficiency is decreased in the upper portion of the reactor core, axial void distribution can be flattened. (N.H.)

  15. Some insights into the role of axial gas flow in fuel rod behaviour during the LOCA based on Halden tests and calculations with the FALCON-PSI code

    International Nuclear Information System (INIS)

    Khvostov, G.; Wiesenack, W.; Zimmermann, M.A.; Ledergerber, G.

    2011-01-01

    Highlights: → A model for the dynamics of axial gas redistribution in fuel rods during the LOCA is developed and coupled to the FALCON fuel behaviour code. → The first verification of the model is carried out using the data of the selected Halden LOCA tests. → According to calculation, the short rods used in the Halden tests show a small effect of the delayed gas redistribution during the clad ballooning. → The predicted effect is significant in the full length rods, eventually resulting in a considerable delay of the predicted moment of cladding rupture. → The predicted delay of cladding burst may be large enough to eventually affect the efficiency of the emergency core cooling system. - Abstract: A model for axial gas flow in a fuel rod during the LOCA is integrated into the FRELAX model that deals with the thermal behaviour and fuel relocation in the fuel rods of the Halden LOCA test series. The first verification was carried out using the experimental data for the inner pressure during the gas outflow after cladding rupture in tests 3, 4 and 5. Furthermore, the modified FRELAX model is implicitly coupled to the FALCON fuel behaviour code. The analysis with the new methodology shows that the dynamics of axial gas-flow along the rod and through the cladding rupture can have a strong influence on the fuel rod behaviour. Specifically, a delayed axial gas redistribution during the heat-up phase of the LOCA can result in a drop of local pressure in the ballooned area, which is eventually able to affect the cladding burst. The results of the new model seem to be useful when analysing some of the Halden LOCA tests (showing considerable fuel relocation) and selected cases of LOCA in full-length fuel rods. While the short rods used in the Halden tests only show a very small effect of the delayed gas redistribution during the clad ballooning, such an effect is predicted to be significant in the full-scale rods - with a power peak located sufficiently away from

  16. Instrumented Taylor anvil-on-rod impact tests for validating applicability of standard strength models to transient deformation states

    Science.gov (United States)

    Eakins, D. E.; Thadhani, N. N.

    2006-10-01

    Instrumented Taylor anvil-on-rod impact tests have been conducted on oxygen-free electronic copper to validate the accuracy of current strength models for predicting transient states during dynamic deformation events. The experiments coupled the use of high-speed digital photography to record the transient deformation states and laser interferometry to monitor the sample back (free surface) velocity as a measure of the elastic/plastic wave propagation through the sample length. Numerical continuum dynamics simulations of the impact and plastic wave propagation employing the Johnson-Cook [Proceedings of the Seventh International Symposium on Ballistics, 1983, The Netherlands (Am. Def. Prep. Assoc. (ADPA)), pp. 541-547], Zerilli-Armstrong [J. Appl. Phys. C1, 1816 (1987)], and Steinberg-Guinan [J. Appl. Phys. 51, 1498 (1980)] constitutive equations were used to generate transient deformation profiles and the free surface velocity traces. While these simulations showed good correlation with the measured free surface velocity traces and the final deformed sample shape, varying degrees of deviations were observed between the photographed and calculated specimen profiles at intermediate deformation states. The results illustrate the usefulness of the instrumented Taylor anvil-on-rod impact technique for validating constitutive equations that can describe the path-dependent deformation response and can therefore predict the transient and final deformation states.

  17. The Winfrith horizontal impact rig

    International Nuclear Information System (INIS)

    Barr, P.

    1985-12-01

    The Horizontal Impact Rig has been designed to allow studies of the impact of radioactive material transport containers and their associated transport vehicles and impact limiters, using large scale models, and to allow physically large missiles to be projected for studying the impact behaviour of metal and concrete structures. It provides an adequately rigid support structure for impact experiments with targets of large dimensions. Details of its design, instrumentation, performance prediction and construction are given. (U.K.)

  18. Fuel rod D07/B15 from Ringhals 2 PWR: Source material for corrosion/leach tests in groundwater. Fuel rod/pellet characterization program. Pt. 1

    International Nuclear Information System (INIS)

    Forsyth, R.

    1987-03-01

    A joint SKB/STUDSVIK experimental program to determine the corrosion rates and to establish the corrosion mechanisms of spent UO 2 fuel in groundwater under both oxidizing and reducing conditions is in progress in the Hot Cell Laboratory of Studsvik Energiteknik AB. High burnup fuel of both BWR and PWR type are studied. Characterization of the spent fuel at both rod and pellet level is an important part of the experimental program. Experiments on PWR fuel have been concentrated so far on specimens from one rod, manufacturer's number 03688, which had occupied position B15 in assembly D07. This assembly had been irradiated for 5 cycles in the Ringhals 2 reactor between 1977 and 1983. The calculated assembly burnup was 41.3 MWd/kg U. The present report is a collection of separate reports describing those items in the characterization program which have been performed so far. No overall summary of the experimental results is given here, and the report should be viewed as a collection of reference data. (orig.)

  19. Reliability assessment of shut-off rod drive mechanism for TAPP - 3 and 4 and critical facility through life cycle testing

    International Nuclear Information System (INIS)

    Singh, Manjit; Badodkar, D.N.; Singh, N.K.; Dalal, N.S.; Mishra, M.K.; Veda Vyas, G.; Kothari, C.B.; Rao, V.V.S.S.; Saraf, R.K.

    2006-01-01

    Shut-off rod drive mechanism forms a safety critical system of a nuclear reactor. It is the space constraints for the given reactor layout, which makes design of shut-off rod drive mechanism (SRDM) a custom built design. Design of SRDM adopts fail-safe, replaceability and the simplicity criterion ensuring very high reliability of its operation. Shut-off rod drive mechanism for TAPP-3 and 4 and 'Critical Facility' have been recently designed and developed at Division of Remote Handling and Robotics (DRHR), BARC. These are designed with a number of advanced features and these are significantly different than those used in Dhruva and 220 MWe PHWRs. Design of SRDM is qualified through proto typing and life cycle testing on a full-scale test set-up. This paper gives details of qualification and life cycle test data for prototype SRDM for TAPP-3 and 4 and 'Critical Facility' and reliability assessment. (author)

  20. Destructive examination of 3-cycle LWR fuel rods from Turkey Point Unit 3 for the Climax-Spent Fuel Test

    International Nuclear Information System (INIS)

    Atkin, S.D.

    1981-06-01

    The destructive examination results of five light water reactor rods from the Turkey Point Unit 3 reactor are presented. The examinations included fission gas collection and analyses, burnup and hydrogen analyses, and a metallographic evaluation of the fuel, cladding, oxide, and hydrides. The rods exhibited a low fission gas release with all other results appearing representative for pressurized water reactor fuel rods with similar burnups (28 GWd/MTU) and operating histories

  1. Control rod

    International Nuclear Information System (INIS)

    Igarashi, Takao; Yoshimoto, Yuichiro; Sugawara, Satoshi; Fukumoto, Takashi; Endo, Zen-ichiro; Saito, Shozo; Shinpo, Katsutoshi; Nishimura, Akira; Ozawa, Michihiro

    1988-01-01

    Purpose: To provide a sufficient shutdown margin upon reactor shutdown, prevent sheath deformation without decreasing neutron absorbents and prevent impact shocks exerted to structural materials. Constitution: The control rod of the present invention comprises a neutron absorption region, a sheath deformation means attached to the side wall and means for restricting and supporting axial movement of the neutron absorbent rod. Then, the amount of absorptive nuclei chained absorbents in the lower region is reduced than that in the upper region. In this way, effective neutron absorbing performance can be obtained relative to the neutron importance distribution during reactor shutdown. In addition, since the operationability is improved by reducing the weight of the control rod and the absorptive nuclei chained neutron abosrbers are used, mechanical nuclear life of the control rod can be increased. Thus, it is possible to prevent the outward deformation of the sheath, as well as prevent collision between the neutron absorber rod and the structural material on the side of inserting the control rod generated upon reactor scram by a simple structure. (Kamimura, M.)

  2. Temperature escalation in PWR fuel rod simulator bundles due to the zircaloy/steam reaction: Post test investigations of bundle test ESBU-2A

    International Nuclear Information System (INIS)

    Hagen, S.; Kapulla, H.; Malauschek, H.; Wallenfels, K.P.; Buescher, B.

    1986-11-01

    This KfK report describes the post test investigation of bundle experiment ESBU-2a. ESBU-2a was the second of two bundle tests on the temperature escalation of zircaloy clad fuel rods. The investigation of the temperature escalation is part of the program of out-of-pile experiments performed within the frame work of the PNS-Severe Fuel Damage program. The bundle was composed of a 3x3 fuel rod array of our fuel rod simulators (central tungsten heater, UO 2 -ring pellet and zircaloy cladding). The length was 0.4 meter. The bundle was heated to a maximum temperature of 2175 0 C. Molten cladding which dissolved part of the UO 2 pellets and slumped away from the already oxidized cladding formed a lump in the lower part of the bundle. After the test the bundle was embedded in epoxy and sectioned with a diamand saw, in the region of the refrozen melt. The cross sections were investigated by metallographic examination. The refrozen (U,Zr,O) melt consists variously of three phases with increasing oxygen content (metallic α-Zry, metallic (U,Zr) alloy and a (U,Zr)O 2 mixed oxide), two phases (α-Zry, (U,Zr)O 2 mixed oxide), or one phase ((U,Zr)O 2 mixed oxide). The cross sections show the increasing oxidation of the cladding with increasing elevation (temperature). A strong azimuthal dependency of the oxidation is found. In regions where the initial oxidized cladding is contacted by the melt one can recognize the interaction between the metallic melt and ZrO 2 of the cladding. Oxygen is taken away from the ZrO 2 . If the melt is in direct contact with steam a relatively well defined oxide layer is formed. (orig.) [de

  3. Development of device for grid spring fatigue and a cell-based fuel rod fretting wear tests

    International Nuclear Information System (INIS)

    Kim, Hyung Kyu; Yoon, Kyung Ho; Kang, Heung Seok; Song, Kee Nam

    2001-05-01

    As an activity of experimental research on the cause and the remedy of LWR fuel fretting failure, developed is test equipment for fatigue of grid spring and cell-based fuel rod fretting wear test. The equipment enables to perform the fretting wear test in the case of gap existence between spring and cladding, which has not been possible by the previously developed one (KAERI/TR-1570/2000). It can also provide fatigue test capability with the frequency of more than 10 Hz. Used are a servo-motor, an eccentric cylinder and lever mechanism for driving system as was similarly used for the previous equipment. In fretting wear test, up to 2 span-length of a fuel cladding tube can be accommodated. For fatigue test, on the other hand, a device for clamping the spring fixture is installed additionally. As a feature of the present equipment, the gap or the contacting force between a spring and a tube can be adjusted during the fretting wear test, while an initial spring force can be simulated for the fatigue test. Tests will be conducted in air at room temperature. In this report, every part of the equipment is explained with photographs, which will provide an easy understanding. Test procedure such as specimen installation, sequence of operation and program handling is also given. As a performance test of the present equipment, displacement range is measured when the hinge of the lever locates at its maximum and minimum positions. This will be used as basic information when additional eccentric cylinder is necessary for different displacement ranges

  4. Application of advanced model of radiative heat transfer in a rod geometry to QUENCH and PARAMETER tests

    International Nuclear Information System (INIS)

    Vasiliev, A.D.; Kobelev, G.V.; Astafieva, V.O.

    2007-01-01

    accident. Special attention is paid to deriving of exact analytical values of view factors and mean beam lengths (which are a good tool in radiative heat transfer concerning gas media) for a number of standard geometries. Generalized Hottel's method of strings is used for rods of finite lengths. Monte-Carlo method is used for validation of new model in application to standard geometries. The developed model is successfully applied for modeling of PARAMETER-SF1 and QUENCH-06 tests, which use the triangular and square rod assembly respectively. (author)

  5. Hydraulic fracture conductivity: effects of rod-shaped proppant from lattice-Boltzmann simulations and lab tests

    Science.gov (United States)

    Osiptsov, Andrei A.

    2017-06-01

    The goal of this study is to evaluate the conductivity of random close packings of non-spherical, rod-shaped proppant particles under the closure stress using numerical simulation and lab tests, with application to the conductivity of hydraulic fractures created in subterranean formation to stimulate production from oil and gas reservoirs. Numerical simulations of a steady viscous flow through proppant packs are carried out using the lattice Boltzmann method for the Darcy flow regime. The particle packings were generated numerically using the sequential deposition method. The simulations are conducted for packings of spheres, ellipsoids, cylinders, and mixtures of spheres with cylinders at various volumetric concentrations. It is demonstrated that cylinders provide the highest permeability among the proppants studied. The dependence of the nondimensional permeability (scaled by the equivalent particle radius squared) on porosity obtained numerically is well approximated by the power-law function: K /Rv2 = 0.204ϕ4.58 in a wide range of porosity: 0.3 ≤ ϕ ≤ 0.7. Lattice-Boltzmann simulations are cross-verified against finite-volume simulations using Navier-Stokes equations for inertial flow regime. Correlations for the normalized beta-factor as a function of porosity and normalized permeability are presented as well. These formulae are in a good agreement with the experimental measurements (including packings of rod-shaped particles) and existing laboratory data, available in the porosity range 0.3 ≤ ϕ ≤ 0.5. Comparison with correlations by other authors is also given.

  6. Instrumentation of fuel safety test rods of the PWR system in the Phebus reactor

    International Nuclear Information System (INIS)

    Schley, Robert; Leveque, J.P.; Aujollet, J.M.; Dutraive, Pierre; Colome, Jean; Bouly, J.C.

    1979-01-01

    The tests were performed in an experimental cell centred in the core of the PHEBUS water reactor of 50 MW. The CEA make two types of apparatus for testing the safety of PWR fuel. One is for testing a single fuel stick and the other a bunch of 25 sticks. The instrumentation described enables the main parameters of the test to be known: temperatures of the fuel - central temperature of the UO 2 - cladding surface temperatures; temperature of the cooling circuits - thermal balance - temperatures of the structures, etc.; coolant pressure; internal pressure of the fuel sticks; direction and flow rate of the fluid. This instrumentation and the technological problems to be overcome are described and the results of the first tests carried out are given [fr

  7. Hydraulic burst tests at elevated temperatures on Zircaloy cladding from fuel rods irradiated in the Winfrith SGHWR

    International Nuclear Information System (INIS)

    Garlick, A.; Hindmarch, P.

    1980-09-01

    Closed-end hydraulic burst tests have been carried out at 613K on lengths of cladding cut from fuel rods that had been irradiated in the SGHWR to 25 n/m 2 . The effects of reactor exposure on the mechanical properties of the Zircaloy cladding, initially in the stress-relieved and fully recrystallised conditions, have been evaluated from measurements of the 0.2% proof stress, the ultimate burst stress, the total circumferential elongation and the reduction in wall thickness at fracture. It is shown that after irradiation, the measured strength properties of stress-relieved cladding remained higher than for that in the fully recrystallised condition, although the large differences observed before irradiation were considerably reduced. The irradiation-induced increase in proof stress measured during these tests was compared with US results from uniaxial tensile tests and, after correcting for the effect of stress-ratio, it is concluded that close agreement exists between the two sets of data for Zircaloy in the fully recrystallised condition. In contrast, the agreement for stress-relieved Zircaloy is less good, although the maximum increase in proof stress after high neutron doses for this material is similar for data from the two sources. After irradiation, the ductility of fully recrystallised Zircaloy remained higher than that of stress-relieved material and there was no evidence to suggest that a serious loss of ductility had occurred for Zircaloy in either condition of heat-treatment as a result of reactor exposure. (author)

  8. Burnout experiments with 6 x 6, 8 x 8 and 7 x 7 rod bundle test sections using freon as model fluid

    International Nuclear Information System (INIS)

    Fulfs, H.; Katsaounis, A.; Minden, C.v.

    1976-01-01

    This paper reports on burnout experiments at staedy state condition using Freon12 as model fluid. The experiments were carried out with three test sections with 6 x 6, 8 x 8 and 7 x 7 rod bundles. The axial flux distribution of the rods is either constant or reactor like. The transformed measured points using STEVENS and BOURE scaling factors to equivalent water conditions respectively, were compared to the burnout correlation W3 using the reactor layout program DYNAMIT. The DYNAMIT code is a thermohydraulic lay-out reactor program without consideration of mixing flow between the subchannels. (orig.) [de

  9. The Carbapenem Inactivation Method (CIM), a Simple and Low-Cost Alternative for the Carba NP Test to Assess Phenotypic Carbapenemase Activity in Gram-Negative Rods

    NARCIS (Netherlands)

    Zwaluw, K. van der; Haan, A. de; Pluister, G.N.; Bootsma, H.J.; Neeling, A.J. de; Schouls, L.M.

    2015-01-01

    A new phenotypic test, called the Carbapenem Inactivation Method (CIM), was developed to detect carbapenemase activity in Gram-negative rods within eight hours. This method showed high concordance with results obtained by PCR to detect genes coding for the carbapenemases KPC, NDM, OXA-48, VIM, IMP

  10. Overburden Stress Normalization and Rod Length Corrections for the Standard Penetration Test (SPT)

    OpenAIRE

    Deger, Tonguc Tolga

    2014-01-01

    The Standard Penetration Test (SPT) has been a staple of geotechnical engineering practice for more than 70 years. Empirical correlations based on in situ SPT data provide an important basis for assessment of a broad range of engineering parameters, and for empirically based analysis and design methods spanning a significant number of areas of geotechnical practice. Despite this longstanding record of usage, the test itself is relatively poorly standardized with regard to the allowable variab...

  11. CONTROL ROD

    Science.gov (United States)

    Walker, D.E.; Matras, S.

    1963-04-30

    This patent shows a method of making a fuel or control rod for a nuclear reactor. Fuel or control material is placed within a tube and plugs of porous metal wool are inserted at both ends. The metal wool is then compacted and the tube compressed around it as by swaging, thereby making the plugs liquid- impervious but gas-pervious. (AEC)

  12. Fuel rods

    International Nuclear Information System (INIS)

    Hattori, Shinji; Kajiwara, Koichi.

    1980-01-01

    Purpose: To ensure the safety for the fuel rod failures by adapting plenum springs to function when small forces such as during transportation of fuel rods is exerted and not to function the resilient force when a relatively great force is exerted. Constitution: Between an upper end plug and a plenum spring in a fuel rod, is disposed an insertion member to the lower portion of which is mounted a pin. This pin is kept upright and causes the plenum spring to function resiliently to the pellets against the loads due to accelerations and mechanical vibrations exerted during transportation of the fuel rods. While on the other hand, if a compression force of a relatively high level is exerted to the plenum spring during reactor operation, the pin of the insertion member is buckled and the insertion member is inserted to the inside of the plenum spring, whereby the pellets are allowed to expand freely and the failures in the fuel elements can be prevented. (Moriyama, K.)

  13. Rodding Surgery

    Science.gov (United States)

    ... Physical activity prior to surgery,  Length of the operation; anesthesia issues,  Reason for the choice of rod,  Time in the hospital,  Length of recovery time at home,  Pain management including control of muscle spasms,  The rehabilitation plan. ...

  14. Control rods

    International Nuclear Information System (INIS)

    Koga, Isao; Masuoka, Ryuzo.

    1979-01-01

    Purpose: To prevent fuel element failures during power conditioning by removing liquid absorbents in poison tubes of control rods in a fast power up step and extracting control rods to slightly increase power in a medium power up step. Constitution: A plurality of poison tubes are disposed in a coaxial or plate-like arrangement and divided into a region capable of compensating the reactivity from the initial state at low temperature to 40% power operation and a region capable of compensating the reactivity in the power up above 40% power operation. Soluble poisons are used as absorbers in the poison tubes corresponding to above 40% power operation and they are adapted to be removed independently from the driving of control rods. The poison tubes filled with the soluble absorbers are responsible for the changes in the reactivity from the initial state at low temperature to the medium power region and the reactivity control is conducted by the elimination of liquid absorbers from the poison tubes. In the succeeding slight power up region above the medium power, power up is proceeding by extracting the control rods having remaining poison tubes filled with solid or liquid absorbers. (Seki, T.)

  15. 29 CFR 1918.54 - Rigging gear.

    Science.gov (United States)

    2010-07-01

    ... 29 Labor 7 2010-07-01 2010-07-01 false Rigging gear. 1918.54 Section 1918.54 Labor Regulations...) SAFETY AND HEALTH REGULATIONS FOR LONGSHORING Vessel's Cargo Handling Gear § 1918.54 Rigging gear. (a... other alternate device shall be provided to allow trimming of the gear and to prevent employees from...

  16. ORNL rod-bundle heat-transfer test data. Volume 3. Thermal-hydraulic test facility experimental data report for test 3.06.6B - transient film boiling in upflow

    International Nuclear Information System (INIS)

    Mullins, C.B.; Felde, D.K.; Sutton, A.G.; Gould, S.S.; Morris, D.G.; Robinson, J.J.

    1982-05-01

    Reduced instrument responses are presented for Thermal-Hyraulic Test Facility (THTF) Test 3.06.6B. This test was conducted by members of the Oak Ridge National Laboratory Pressurized-Water-Reactor (PWR) Blowdown Heat Transfer (BDHT) Separate-Effects Program on August 29, 1980. The objective of the program was to investigate heat transfer phenomena believed to occur in PWR's during accidents, including small and large break loss-of-coolant accidents. Test 3.06.6B was conducted to obtain transient film boiling data in rod bundle geometry under reactor accident-type conditions. The primary purpose of this report is to make the reduced instrument responses for THTF Test 3.06.6B available. Included in the report are uncertainties in the instrument responses, calculated mass flows, and calculated rod powers

  17. Simulation of the fuel rod bundle test QUENCH-03 using the system codes ASTEC and ATHLET-CD

    International Nuclear Information System (INIS)

    Kruse, P.; Koch, M.K.

    2011-01-01

    The QUENCH-03 test was performed on the 21. of January 1999 at FZK (Forschungszentrum Karlsruhe) to investigate the behaviour on reflood of PWR (Pressurized Water Reactor) fuel rods with little oxidation. This paper presents the results of the simulation of QUENCH-03 performed with the version V1.3 of the integral code ASTEC (Accident Source Term Evaluation Code) which is being developed by IRSN (France) in cooperation with GRS (Germany) and with the program version 2.1A of the mechanistic code ATHLET-CD (Analysis of Thermal-hydraulics of Leaks and Transients - Core Degradation) which is under development by GRS. At first the QUENCH test facility and the QUENCH test program in general are described. The test conduct of the test QUENCH-03 follows as well as a description of the used codes ASTEC and ATHLET-CD with the associated modeling of the test section. The results of this calculation show that during the heat-up and transient phase both codes can calculate bundle and shroud temperatures as well as the hydrogen production in good approximation to the experimental data. During the quench phase and up to the end of the test only the oxidation model PRATER of ASTEC simulates the hydrogen production very well, the other oxidation models of ASTEC cannot calculate to some extent the measured amount of hydrogen. ATHLET-CD underestimates the integral amount at the end of the test. In the ASTEC calculations the temperatures during the quench phase show qualitatively good results, only time delays on some elevations of the bundle could be noticed. ATHLET-CD reproduces the thermal behaviour up to the first temperature escalation very well, after that the temperatures are partly over-estimated. The time delay recognized in the ASTEC calculations are seen as well. The results of the integral code ASTEC emphasize that the calculation of QUENCH-03 is possible and leading to good results concerning hydrogen release and corresponding temperatures. Because the QUENCH-03 test was

  18. One big rig, two valuable functions

    Energy Technology Data Exchange (ETDEWEB)

    Anon

    2004-11-01

    A hybrid coil tubing and conventional workover rig, tailor-made for conditions on Alaska's remote North Slope is described. The dual function rig, owned by BP Exploration, towers 142 feet above the barren Arctic tundra, and weighs between 1.5 and 2 million pounds, rests on eight enormous wheels that stand 11.5 feet tall and 3.5 feet wide, and is supported by 64 smaller tires in between. The rig includes the hybrid coiled tubing rig and a conventional workover rig; it exerts less than 100 pounds per square inch of pressure on the tender Arctic surface as it moves forward at a top speed of two miles per hour. It is considered by its developers as the next-step change in providing cost-effective access to reserves in the large, mature and remote oilfields such as those of Alaska's Prudhoe Bay. The rig is the product of cooperation between Schlumberger expertise in coiled tubing drilling and Nordic-Calista's know-how of jointed pipe operations and operating rigs in an Arctic environment. It is the first time in Prudhoe Bay, and probably in the world, that a coiled tubing unit was installed on a rig to do coiled-tubing sidetracks, i.e to drill a secondary wellbore away from the original wellbore. Since the first unit was commissioned in 1996, the rig has drilled 280 wells. Rig No. 2, much improved and commissioned in 2002, drilled about 30 wells to date. Unlike Rig No, 1, Rig No. 2 can change reels without a crane, and it has a hydraulic pipe skate that moves jointed pipe to and from the rig floor without human aid. The expectation is that using this rig it will be possible to do micro-hole exploration drilling on the North Slope (i.e. drilling a small surface hole with jointed pipe and then switch to coiled tubing), substantially cutting the cost of exploration.

  19. Full-fluence tests of experimental thermosetting fuel rods for the high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Bullock, R.E.

    1981-01-01

    The irradiation performance of injected thermosetting fuel rods is compared to that of standard pitch-temperature gas-cooled reactor requirements. The primary objective of the experiments reported here was to obtain additional irradiation data at higher fluences for resin-based rods with intermediate binder char contents within the 15 to 30 wt% ''window of acceptability'' that had been previously established. 12 refs

  20. Development of carbon/carbon composite control rod for HTTR. 2. Concept, specifications and mechanical test of materials

    International Nuclear Information System (INIS)

    Eto, Motokuni; Ishiyama, Shintaro; Fukaya, Kiyoshi; Saito, Tamotsu; Ishihara, Masahiro; Hanawa, Satoshi.

    1998-01-01

    A concept and specifications of carbon/carbon composite (C/C) control rod were proposed, aiming at the application of the material to the HTTR. The outer diameter and length of the control rod were kept as the same as those of the present control rod, i.e., 113 mm and 3094 mm, respectively. According to the concept, the rod consists of ten units which are connected in series using bolts. Then, the stresses generated by dead loads in the control rod elements were estimated and compared with the design strengths which were derived from the results of measurements of tensile, compressive, bending and shear strengths of two candidate materials, AC250 (Across Co.) and CX-270 (Toyo Tanso Co.). Design strength was preliminarily determined as one-third or one-fifth of the mean strength. Ratio of the design strength to generated stress for the AC250 (2D) was : Tensile stress in the outer sleeve tube, 66, tensile and shear stresses in the M16 bolt, 8.8 and 8.5, shear stress in the plug support bolt M8, 2.43. These results are believed to indicate the mechanical integrity of the control rod structure. Data available on the candidate materials were also compiled in the Appendix. (author)

  1. Freon Rig design for performing to heat transfer experiments for nuclear reactors fuel bundles

    International Nuclear Information System (INIS)

    Flores, L.F.V.

    1981-01-01

    The main features of a Freon Rig design for performing to heat transfer experiments for PWR and BWR fuel bundles, are presented. The project is based on a Freon Rig pressurized at 30 bar with a flow rate up to 80 m 3 /h. The maximum power fed to test sections is of about 420 KW D.C. The rig was designed to use scaling techniques wich would enable a fluid of low latente heat to be used in place of water, thereby reducing the cost of testes. (Author) [pt

  2. ORNL rod-bundle heat-transfer test data. Volume 7. Thermal-Hydraulic Test Facility experimental data report for test series 3.07.9 - steady-state film boiling in upflow

    International Nuclear Information System (INIS)

    Mullins, C.B.; Felde, D.K.; Sutton, A.G.; Gould, S.S.; Morris, D.G.; Robinson, J.J.

    1982-05-01

    Thermal-Hydraulic Test Facility (THTF) test series 3.07.9 was conducted by members of the Oak Ridge National Laboratory Pressurized-Water Reactor (ORNL-PWR) Blowdown Heat Transfer (BDHT) Separate-Effects Program on September 11, September 18, and October 1, 1980. The objective of the program is to investigate heat transfer phenomena believed to occur in PWRs during accidents, including small- and large-break loss-of-coolant accidents. Test series 3.07.9 was designed to provide steady-state film boiling data in rod bundle geometry under reactor accident-type conditions. This report presents the reduced instrument responses for THTF test series 3.07.9. Also included are uncertainties in the instrument responses, calculated mass flows, and calculated rod powers

  3. Control rod

    International Nuclear Information System (INIS)

    Takeda, Toshikazu; Inoue, Kotaro.

    1979-01-01

    Purpose: To flatten the power distribution in the reactor core without impairing neutron economy by disposing pins containing elements of lower atomic number in the central region of a shroud and loading pins containing depleted uranium in the periphery region thereof. Constitution: The shroud has a layer of pins containing depleted uranium in the peripheral region and a layer of pins containing elements of lower atomic number such as beryllium in the central region. Heat removal from those pins containing depleted uranium and elements of lower atomic number (neutron moderator) is effected by sodium flow outside of the cladding material. The control rod operation is conducted by inserting or extracting the central portion (pins containing elements of lower atomic number such as beryllium) inside of the stainless pipe. Upon extraction of the control rod, the moderator in the central region is removed whereby high speed neutrons are no more deccelerated and the absorption rate to the depleted uranium is decreased. This can flatten the power distribution in the reactore core with the disposition of a plurality of control rods at a better neutron economy as compared with the use of neutron absorber such as boron. (Seki, T.)

  4. Control rod

    International Nuclear Information System (INIS)

    Fukumoto, Takashi; Hirakawa, Hiromasa; Kawashima, Norio; Goto, Yasuyuki.

    1994-01-01

    Neutron absorbers are contained in a tubular member comprising, integrally a tubular portion and four corners disposed at the outer circumference of the tubular portion at every 90deg, to provide a neutron absorbing tube. A plurality of neutron absorbing tubes are arranged in parallel in the lateral direction, and adjacent corners are joined, into a blade to constitute a control rod. Such a control rod has a great structural strength, simple in the structure and relatively light in weight and can contain a great amount of neutron absorbers. Upon formation of the control rod by arranging the blades in a cross-like shape, at least a portion thereof is constituted with short neutron absorbing tubes shorter than the entire length of the blade, and gaps are formed at positions in adjacent in the axial direction. With such a constitution, there is no worry that a wing end of the blade collides against or be abraded with a fuel channel box or a fuel support. Even if fuel channels are vibrated upon scram of the reactor, such as occurrence of earthquakes, it can be inserted to the reactor easily. (N.H.)

  5. 40th annual Reed rig census

    International Nuclear Information System (INIS)

    Fitts, R.L.; Stokes, T.A.

    1992-01-01

    This paper reports that declines characterize the 1992 rig census-in the number of available drilling rigs, in the number of active rigs, in rig utilization rate, in the number of rig owners and in industry optimism. The number of rotary rigs available for U.S. drilling fell by 255 units (11.3%) during the past 12 months, an attrition rate almost four times greater than in 1991. But despite the high attrition, only 59.7% of remaining rigs were working during the time the census was taken. Results of the 1992 census bring emphasis to an industry trend that became apparent in early 1991. The major oil companies, and many independents, continued their exodus form the U.S., and the remaining independents, which were hurt by low natural gas prices and unfavorable tax treatment of intangible drilling costs, were not able to pick u the drilling slack. Consequently, the past year has been disastrous for many U.S. drilling contractors, and the outlook for this industry segment remains bleak

  6. Coupled thermal-hydraulic and neutronic simulations of Phenix control rod withdrawal tests with SIMMER-IV

    International Nuclear Information System (INIS)

    Kriventsev, Vladimir; Gabrielli, Fabrizio; Rineiski, Andrei

    2014-01-01

    The “end-of-life” tests performed in the Phenix reactor before its final shutdown in 2009, in particular the Control Rod (CR) withdrawal experiments provide an excellent opportunity for the validation and verification of the reactor physics computer codes and modeling approaches. SIMMER-IV, a modern three-dimensional reactor safety code, has been recently employed at Karlsruhe Institute of Technology (KIT) for simulating Phenix experiments in the framework of a benchmark exercise organized under the IAEA project. In this paper, we report and discuss main results obtained with SIMMER-IV at KIT. Particular attention is devoted to the coupling features of thermal-hydraulics and neutronics and their mutual influences. The reactor reactivity, power and neutron flux distributions calculated with SIMMER-IV are in good agreement both with experimental results and with calculations with advanced neutronics codes, such as ERANOS, while the CR reactivity worth is overestimated due to neglecting heterogeneity effects. Because of its multi-physics capabilities SIMMER also calculates the temperature distributions which are in a good agreement with the experimental test results. In this work we describe the improvements in SIMMER neutronics model by employing a correction that is based on the results of cell calculations performed with ERANOS. The study confirms that the 3D SIMMER-IV code can accurately predict major fast reactor neutronics and thermal hydraulic parameters, provided that a special treatment is employed for CR modeling. The results of calculations are analyzed in frames of SIMMER-IV validation and verification assessment. (author)

  7. Characterisation of a refurbished 1½ stage turbine test rig for flowfield mapping behind blading with non-axisymmetric contoured endwalls

    CSIR Research Space (South Africa)

    Snedden, Glen C

    2007-09-01

    Full Text Available such that they should provide for a 5° rotor blade incidence change either side of the design point at the hub. Figures 11 to 13 give the results of this first series of tests. Once again the power output is below the design point by some 24% and the stage... are captured in figures 14 to 16. Once again the results indicate similar disparities between design and actual results as well as between annular and contoured turbine designs as the first technique. Finally all the results are collated into two...

  8. Advanced control strategies for a drill rig

    International Nuclear Information System (INIS)

    Banerjee, A.; Hiller, M.; Fink, B.

    1996-01-01

    The construction of tunnels is usually undertaken using a combination of blasting and drilling to achieve rock excavation. Easy handling and high accuracy, and thus greater efficiency, in drilling rigs is an essential ingredient of successful competition in the market place. This article describes a cartesian control concept used for a twin boom drill rig. This simplifies the handling of a drilling boom, reduces the duration of a working cycle and increases security. A remote control system has been added to the drill rig to support the operator working in complicated environments. (UK)

  9. Body Language Advanced 3D Character Rigging

    CERN Document Server

    Allen, Eric; Fong, Jared; Sidwell, Adam G

    2011-01-01

    Whether you're a professional Character TD or just like to create 3D characters, this detailed guide reveals the techniques you need to create sophisticated 3D character rigs that range from basic to breathtaking. Packed with step-by-step instructions and full-color illustrations, Body Language walks you through rigging techniques for all the body parts to help you create realistic and believable movements in every character you design. You'll learn advanced rigging concepts that involve MEL scripting and advanced deformation techniques and even how to set up a character pipeline.

  10. A 3D numerical study of LO2/GH2 supercritical combustion in the ONERA-Mascotte Test-rig configuration

    Science.gov (United States)

    Benmansour, Abdelkrim; Liazid, Abdelkrim; Logerais, Pierre-Olivier; Durastanti, Jean-Félix

    2016-02-01

    Cryogenic propellants LOx/H2 are used at very high pressure in rocket engine combustion. The description of the combustion process in such application is very complex due essentially to the supercritical regime. Ideal gas law becomes invalid. In order to try to capture the average characteristics of this combustion process, numerical computations are performed using a model based on a one-phase multi-component approach. Such work requires fluid properties and a correct definition of the mixture behavior generally described by cubic equations of state with appropriated thermodynamic relations validated against the NIST data. In this study we consider an alternative way to get the effect of real gas by testing the volume-weighted-mixing-law with association of the component transport properties using directly the NIST library data fitting including the supercritical regime range. The numerical simulations are carried out using 3D RANS approach associated with two tested turbulence models, the standard k-Epsilon model and the realizable k-Epsilon one. The combustion model is also associated with two chemical reaction mechanisms. The first one is a one-step generic chemical reaction and the second one is a two-step chemical reaction. The obtained results like temperature profiles, recirculation zones, visible flame lengths and distributions of OH species are discussed.

  11. Development of a hardware-in-the-loop-test rig to verify the reliability of oil burner pumps. Application by the use of biocide in domestic heating oil; Entwicklung eines Hardware-in-the-loop Pruefstands zum Nachweis der Betriebssicherheit von Oelbrennerpumpen. Anwendungen bei Einsatz von Biozidadditiven

    Energy Technology Data Exchange (ETDEWEB)

    Rheinberg, Oliver van; Lukito, Jayadi; Liska, Martin [Oel-Waerme-Institut gGmbH (OWI), Aachen-Herzogenrath (Germany)

    2009-09-15

    Within this project, a hardware-in-the-loop test rig has been developed to investigate the influence of different fuels on the reliability of oil burner pumps. The test rig is constructed with commercial burner components. One test rig consists of four pump cycles, where the fuel recirculates for max. 2000 h. Low powered electric motors of 90 Watts have been used deliberately, so that the apparatus is more sensitive to failure due to an increase in pump load. A practise relevant intermittent operating mode has been implemented for the simulation of real operation characteristics. The measured variable and evaluation parameters are start-up torque, intake pressure, fuel pump pressure and temperature. Operation failures of oil burner pumps in the field, due to an over-additisation of biocides, have been observed. These failures could be reproducibly simulated on the pump test stands. The results of the project are a redefinition of limits of biocide concentration and the development of new biocides, which are suitable for use in domestic heating oil with a content of up to 20 % Fatty-Acid-Methyl-Ester. (orig.)

  12. Wythenshawe boiler rig. Thirty years of support to the UK nuclear power industry

    International Nuclear Information System (INIS)

    Rudge, Andy; Woolsey, Ian S.; Moore, Andrew

    2010-01-01

    The Wythenshawe Boiler Rig in Manchester, UK, recently celebrated thirty years of operation in support of the UK nuclear power industry. The Boiler Rig, owned by EDF Energy and operated on EDF Energy's behalf by Serco plc, is a full scale once-through boiler test facility for the investigation of chemistry and corrosion related topics. This paper presents an overview of the design and operation of the Boiler Rig together with some of the technical highlights from its thirty years of operation, many of which have relevance to power plant operations beyond those plants for which the work was performed. (orig.)

  13. Thermal hydraulic test of advanced fuel bundle with spectral shift rod (SSR) for BWR. Effect of thermal hydraulic parameters on steady state characteristics

    International Nuclear Information System (INIS)

    Kondo, Takao; Kitou, Kazuaki; Chaki, Masao; Ohga, Yukiharu; Makigami, Takeshi

    2011-01-01

    Japanese national project of next generation light water reactor (LWR) development started in 2008. Under this project, spectral shift rod (SSR) is being developed. SSR, which replaces conventional water rod (WR) of boiling water reactor (BWR) fuel bundle, was invented to enhance the BWR's merit, spectral shift effect for uranium saving. In SSR, water boils by neutron and gamma-ray direct heating and water level is formed as a boundary of the upper steam region and the lower water region. This SSR water level can be controlled by core flow rate, which amplifies the change of average core void fraction, resulting in the amplified spectral shift effect. This paper presents the steady state test results of the base geometry case in SSR thermal hydraulic test, which was conducted under the national project of next generation LWR. In the test, thermal hydraulic parameters, such as flow rate, pressure, inlet subcooling and heater rod power are changed to evaluate these effects on SSR water level and other SSR characteristics. In the test results, SSR water level rose as flow rate rose, which showed controllability of SSR water level by flow rate. The sensitivities of other thermal hydraulic parameters on SSR water level were also evaluated. The obtained data of parameter's sensitivities is various enough for the further analytical evaluation. The fluctuation of SSR water level was also measured to be small enough. As a result, it was confirmed that SSR's steady state performance was as planned and that SSR design concept is feasible. (author)

  14. Inspection of steel poles; ultrasonic testing of anchor ground rods and cathodic reactions : Corrosion detection : an emerging problem in buried steel structures

    Energy Technology Data Exchange (ETDEWEB)

    Pandey, A.K.; Randle, R.E.; Stewart, A.H. [EDM International Inc., Fort Collins, CO (United States)

    2002-07-01

    A typical inspection of steel utility poles routinely overlooks what is below ground, such as anchor rods, stub angles in lattice towers, and direct embedded steel poles. Stub angles are lap or butt spliced to the tower leg and extend several feet below ground line. A case study concerning stub angles (Oberst 1998) is discussed. An inspection of steel poles erected in 1929 revealed that 40 per cent of legs had complete loss of galvanizing, 10 per cent of legs had greater than 10 per cent loss of cross-section, and 2 per cent of legs had greater than 80 per cent loss of cross-section. All corrosion was found within one foot of ground line. A relatively new concept is direct embedded steel poles. An emerging problem concerns tree induced anchor rod corrosion. A corrosion technique for anchor rods was developed and has been commercially available for the past three years. Its effectiveness was verified at the Montana Power Company 500 kV Colstrip Project, where 3 anchor failures were detected in 1995 due to corrosion wastage. The rods are classified as being in good condition up to 10 per cent loss of cross-section, moderate corrosion for losses between 10 and 25 per cent, and excessive corrosion for losses greater than 25 per cent. The results obtained at the Montana Power Company indicated the technique was 98 per cent accurate. The authors discuss the capabilities and limitations of the technique. It was also applied for the Anchor Rod Inspection Project of the Georgia Power Company (GPC). The technique is evaluated in the laboratory, then optimized. Field prototypes are developed, followed by an evaluation at different test sites. figs.

  15. Development of a real-time fuel cell stack modelling solution with integrated test rig interface for the generic fuel cell modelling environment (GenFC) software

    Energy Technology Data Exchange (ETDEWEB)

    Fraser, S.D.; Monsberger, M.; Hacker, V. [Graz Univ. of Technology, Graz (Austria). Christian Doppler Laboratory for Fuel Cell Systems; Gubner, A.; Reimer, U. [Forschungszentrum Julich, Julich (Germany)

    2007-07-01

    Since the late 1980s, numerous FC models have been developed by scientists and engineers worldwide to design, control and optimize fuel cells (FCs) and fuel cell (FC) power systems. However, state-of-the-art FC models have only a small range of applications within the versatile field of FC modelling. As fuel cell technology approaches commercialization, the scientific community is faced with the challenge of providing robust fuel cell models that are compatible with established processes in industrial product development. One such process, known as Hardware in the Loop (HiL), requires real-time modelling capability. HiL is used for developing and testing hardware components by adding the complexity of the related dynamic systems with mathematical representations. Sensors and actuators are used to interface simulated and actual hardware components. As such, real-time fuel cell models are among the key elements in the development of the Generic Fuel Cell Modelling Environment (GenFC) software. Six European partners are developing GenFC under the support of the Sixth European Framework Programme for Research and Technological Development (FP6). GenFC is meant to increase the use of fuel cell modelling for systems design and to enable cost- and time-efficient virtual experiments for optimizing operating parameters. This paper presented an overview of the GenFC software and the GenFC HiL functionality. It was concluded that GenFC is going to be an extendable software tool providing FC modelling techniques and solutions to a wide range of different FC modelling applications. By combining the flexibility of the GenFC software with this HiL-specific functionality, GenFC is going to promote the use of FC model-based HiL technology in FC system development. 9 figs.

  16. Sucker rods

    Energy Technology Data Exchange (ETDEWEB)

    Rylov, B M; Kostur, I N; Shcheigiy, B I; Sukhanov, V S

    1983-01-01

    As an addendum to A.s. USSR patent No 769087, this particular sucker rod utilizes a differential piston spring that has been attached outside the body of the auxiliary pump. The pump cylinder is attached to the intake line of the main pump. The lower part of the auxiliary pump is equipped with vertical slits, while the differential piston is equipped with a perforated pusher and support under the spring; it can also be shifted as necessary with respect to the vertical slits.

  17. Nuclear fuel rod end plug weld inspection

    International Nuclear Information System (INIS)

    Parker, M. A.; Patrick, S. S.; Rice, G. F.

    1985-01-01

    Apparatus and method for testing TIG (tungsten inert gas) welds of end plugs on a sealed nuclear reactor fuel rod. An X-ray fluorescent spectrograph testing unit detects tungsten inclusion weld defects in the top end plug's seal weld. Separate ultrasonic weld inspection system testing units test the top end plug's seal and girth welds and test the bottom end plug's girth weld for penetration, porosity and wall thinning defects. The nuclear fuel rod is automatically moved into and out from each testing unit and is automatically transported between the testing units by rod handling devices. A controller supervises the operation of the testing units and the rod handling devices

  18. 39th annual Reed rig census

    International Nuclear Information System (INIS)

    Crowhurst, M.E.; Fitts, R.L.

    1991-01-01

    This paper reports on cutbacks in U.S. exploration and development drilling during the first half of 1991 which squeezed most of the optimism out of the drilling industry. Just how rough the year has been is underscored by the results of this year's rig census. The number of rotary rigs available for U.S. drilling declined by only 69 units (3%) during the past 12 months. But despite those withdrawals from competition, only 66% of the remaining rigs were working at the time the census was taken. Results of the 1991 census contrasted sharply with the stability and optimism that seemed apparent a year ago when 72% of the available rig fleet met the census definition of active. At that time, the mini-boom in horizontal drilling coupled with tax-credit- driven gas drilling led to a relatively high rig utilization rate and suggested that rig supply and demand might be close to an economically acceptable balance. However, it quickly became apparent in early 1991 that industry optimism was unjustified. Horizontal drilling began to drop and the lowest natural gas prices in 12 years triggered rapid declines in gas drilling. Although oil prices have been relatively stable and above $18 per bbl since January 1989, most major operators have concluded that a better return on investment can be had outside the U.S. and have drastically cut their domestic drilling budgets. These factors, combined with softened energy demand from the worldwide recession, further slowed U.S. drilling. The long awaited balance between rig supply and demand has seemingly slipped away. The 1991 Reed rig census describes an industry facing several more rough years. Details of this year's census include: The available U.S. fleet now stands at 2,251 rigs, down by 69 from the 2,320-unit total in 1990, and the lowest since 1976. Rigs meeting the census definition of active numbered 1,485, down 192 (11.4%) from the 1,677 active rigs counted a year earlier

  19. Learning Inverse Rig Mappings by Nonlinear Regression.

    Science.gov (United States)

    Holden, Daniel; Saito, Jun; Komura, Taku

    2017-03-01

    We present a framework to design inverse rig-functions-functions that map low level representations of a character's pose such as joint positions or surface geometry to the representation used by animators called the animation rig. Animators design scenes using an animation rig, a framework widely adopted in animation production which allows animators to design character poses and geometry via intuitive parameters and interfaces. Yet most state-of-the-art computer animation techniques control characters through raw, low level representations such as joint angles, joint positions, or vertex coordinates. This difference often stops the adoption of state-of-the-art techniques in animation production. Our framework solves this issue by learning a mapping between the low level representations of the pose and the animation rig. We use nonlinear regression techniques, learning from example animation sequences designed by the animators. When new motions are provided in the skeleton space, the learned mapping is used to estimate the rig controls that reproduce such a motion. We introduce two nonlinear functions for producing such a mapping: Gaussian process regression and feedforward neural networks. The appropriate solution depends on the nature of the rig and the amount of data available for training. We show our framework applied to various examples including articulated biped characters, quadruped characters, facial animation rigs, and deformable characters. With our system, animators have the freedom to apply any motion synthesis algorithm to arbitrary rigging and animation pipelines for immediate editing. This greatly improves the productivity of 3D animation, while retaining the flexibility and creativity of artistic input.

  20. The evaluation of secondary system oxygen-scavenging chemicals using a water-circulating rig

    Energy Technology Data Exchange (ETDEWEB)

    Collins, M.W. [Nuclear Dept., HMS Sultan (United Kingdom)

    2002-07-01

    To assess the efficiency, mode of action and possible by-products of chemical dosing agents, e.g. oxygen scavengers, a circulating water rig was constructed. The rig uses a demineralized water supply as a source of make-up water to fill a recirculating loop of approx. 10 litres volume. The rig pipework is made of polythene with standard off-the shelf pipe fittings and connectors. The following parameters can be measured within the rig: pH and conductivity measured by in-line monitor, dissolved oxygen level, temperature. The system has already been used for some preliminary testing. The following oxygen scavengers have been used for tests: ascorbic acid (vitamin C), N,N-diethyl-hydroxylamine (DEHA), Hydroquinone, hydrazine hydrate and anhydrous sodium sulfite. (authors)

  1. The evaluation of secondary system oxygen-scavenging chemicals using a water-circulating rig

    International Nuclear Information System (INIS)

    Collins, M.W.

    2002-01-01

    To assess the efficiency, mode of action and possible by-products of chemical dosing agents, e.g. oxygen scavengers, a circulating water rig was constructed. The rig uses a demineralized water supply as a source of make-up water to fill a recirculating loop of approx. 10 litres volume. The rig pipework is made of polythene with standard off-the shelf pipe fittings and connectors. The following parameters can be measured within the rig: pH and conductivity measured by in-line monitor, dissolved oxygen level, temperature. The system has already been used for some preliminary testing. The following oxygen scavengers have been used for tests: ascorbic acid (vitamin C), N,N-diethyl-hydroxylamine (DEHA), Hydroquinone, hydrazine hydrate and anhydrous sodium sulfite. (authors)

  2. Fuel rods

    International Nuclear Information System (INIS)

    Adachi, Hajime; Ueda, Makoto

    1985-01-01

    Purpose: To provide a structure capable of measuring, in a non-destructive manner, the releasing amount of nuclear gaseous fission products from spent fuels easily and at a high accuracy. Constitution: In order to confirm the integrity and the design feasibility of a nuclear fuel rod, it is important to accurately determine the amount of gaseous nuclear fission products released from nuclear pellets. In a structure where a plurality of fuel pellets are charged in a fuel cladding tube and retained by an inconel spring, a hollow and no-sealed type spacer tube made of zirconium or the alloy thereof, for example, not containing iron, cobalt, nickel or manganese is formed between the spring and the upper end plug. In the fuel rod of such a structure, by disposing a gamma ray collimator and a gamma ray detector on the extension of the spacer pipe, the gamma rays from the gaseous nuclear fission products accumulated in the spacer pipe can be detected while avoiding the interference with the induction radioactivity from inconel. (Kamimura, M.)

  3. Experimental study of the reflooding of a constricted tube in the REFLEX rig

    International Nuclear Information System (INIS)

    Denham, M.K.; Elliott, D.F.; Britton-Jones, K.A.

    1982-08-01

    The Winfrith experimental programme in support of the PWR is focussed on fuel thermal and hydraulic performance under hypothetical accident conditions, and includes studies of reflooding heat transfer of single tubes and fuel rod clusters under simulated accident conditions, aimed at improving understanding of the processes involved and providing data for code development and validation. The work described is part of a study of the possible effects of clad ballooning on ECCS effectiveness. During a large loss of coolant accident the primary circuit will depressurise and the core will overheat. The Zircaloy fuel cladding may swell, partially blocking the coolant passages by the formation of local ''balloons''. An experiment was carried out in the REFLEX single tube reflooding rig, to study, in a simple geometry, the effect of the partial blockage of the tube on the fluid flow and heat transfer during reflooding. The blockage consisted of a tapering entrance with a flow area 60 percent less than the unconstricted tube, and a tapering exit. The flow could be viewed through windows. 66 refloods were carried out over a pressure range of 1 to 4 bar. Results of these tests are presented. (U.K.)

  4. NSRR experiment with un-irradiated uranium-zirconium hydride fuel. Design, fabrication process and inspection data of test fuel rod

    Energy Technology Data Exchange (ETDEWEB)

    Sasajima, Hideo; Fuketa, Toyoshi; Ishijima, Kiyomi; Kuroha, Hiroshi; Ikeda, Yoshikazu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Aizawa, Keiichi

    1998-08-01

    An experiment plan is progressing in the Nuclear Safety Research Reactor (NSRR) to perform pulse-irradiation with uranium-zirconium hydride (U-ZrH{sub x}) fuel. This fuel is widely used in the training research and isotope production reactor of GA (TRIGA). The objectives of the experiment are to determine the fuel rod failure threshold and to investigate fuel behavior under simulated reactivity initiated accident (RIA) conditions. This report summarizes design, fabrication process and inspection data of the test fuel rods before pulse-irradiation. The experiment with U-ZrH{sub x} fuel will realize precise safety evaluation, and improve the TRIGA reactor performance. The data to be obtained in this program will also contribute development of next-generation TRIGA reactor and its safety evaluation. (author)

  5. The Testing of Fuel Rod Models with Zr1Nb Alloy Cladding in Water Vapor at Temperature of Hypothetical Accident Situation in WWER-1000 Type Reactors

    International Nuclear Information System (INIS)

    Krasnorutsky, V.S.; Petel'guzov, I.A.; Gritsina, V.M.; Rodak, A.G.; Belash, N.N.; Yakovlev, V.K.

    2006-01-01

    In the article happen to results of testing the fuel rod models, their welded joints, changing the mechanical characteristics of shells of models from experimental parties of pipes from Zr1Nb alloy (Zr+1 mass%Nb) at heating of models, pervaded helium before pressures, using in earned one's living fuel rods (2,2 MPa), before the temperature 770 degree C and above occurs an overblown fuels, but at temperature 820...830 degree C shells can be broken at the expense of pressure of warming gas. Swept away reduction plasticity and embrittlement shells after the heating under temperature of 900...1200 degree C and cooling before room temperature pipes-shells from Zr1Nb alloy and from the staff alloy E110

  6. Experimental benchmark data for PWR rod bundle with spacer-grids

    International Nuclear Information System (INIS)

    Dominguez-Ontiveros, Elvis E.; Hassan, Yassin A.; Conner, Michael E.; Karoutas, Zeses

    2012-01-01

    In numerical simulations of fuel rod bundle flow fields, the unsteady Navier–Stokes equations have to be solved in order to determine the time (phase) dependent characteristics of the flow. In order to validate the simulations results, detailed comparison with experimental data must be done. Experiments investigating complex flows in rod bundles with spacer grids that have mixing devices (such as flow mixing vanes) have mostly been performed using single-point measurements. In order to obtain more details and insight on the discrepancies between experimental and numerical data as well as to obtain a global understanding of the causes of these discrepancies, comparisons of the distributions of complete phase-averaged velocity and turbulence fields for various locations near spacer-grids should be performed. The experimental technique Particle Image Velocimetry (PIV) is capable of providing such benchmark data. This paper describes an experimental database obtained using two-dimensional Time Resolved Particle Image Velocimetry (TR-PIV) measurements within a 5 × 5 PWR rod bundle with spacer-grids that have flow mixing vanes. One of the unique characteristic of this set-up is the use of the Matched Index of Refraction technique employed in this investigation to allow complete optical access to the rod bundle. This unique feature allows flow visualization and measurement within the bundle without rod obstruction. This approach also allows the use of high temporal and spatial non-intrusive dynamic measurement techniques namely TR-PIV to investigate the flow evolution below and immediately above the spacer. The experimental data presented in this paper includes explanation of the various cases tested such as test rig dimensions, measurement zones, the test equipment and the boundary conditions in order to provide appropriate data for comparison with Computational Fluid Dynamics (CFD) simulations. Turbulence parameters of the obtained data are presented in order to gain

  7. Intra-abdominal recurrence of colorectal cancer detected by radioimmunoguided surgery (RIGS system)

    International Nuclear Information System (INIS)

    Sardi, A.; Workman, M.; Mojzisik, C.; Hinkle, G.; Nieroda, C.; Martin, E.W. Jr.

    1989-01-01

    Since 1986, 32 patients with metastatic colorectal cancer have undergone second-look radioimmunoguided surgery (RIGS system). The primary tumor was located in the right and transverse colon in 11 patients, left and sigmoid colon in 16, and rectum in five. The carcinoembryonic antigen level was elevated in 30 patients (94%); all patients underwent a computed tomographic scan of the abdomen and pelvis. The overall sensitivity of the computed tomographic scan was 41% (abdomen other than liver, 27%; liver, 58%; and pelvis, 22%). The RIGS system identified recurrent tumor in 81% of the patients. The most common site of metastasis was the liver (41%), independent of the primary location. Local/regional recurrences alone accounted for 40% of all recurrences. In six patients (18%), recurrent tumor was found only with the RIGS system. The RIGS system is more dependable in localizing clinically obscure metastases than other methods, and carcinoembryonic antigen testing remains the most accurate preoperative method to indicate suspected recurrences

  8. Technical specification for IR rig manufacture

    Energy Technology Data Exchange (ETDEWEB)

    Han Hyon Soo; Cho, W. K.; Kim, S. D.; Park, U. J.; Hong, S. B.; Yoo, K. M

    2000-10-01

    IR Rig is one of the equipments are required in HANARO core for a radioisotope target. The various conditions like high radiation, high heat, rapid flow and vibration may cause swelling, Brittleness and acceleration of corrosion in HANARO core. These specific problems can be prevented and the safety of such equipment are prerequisite as well as durableness and surveillance. Therefore, the selection of material has to be made on the basis of small cross-section area, low energy emission by the gamma ray due to the absorption of neutron and short half life. The body is consist of aluminum and Inconel-750 was used for the internal spring(coil) which is known to be durable. The whole production process including the purchase of accessory, mechanical processing, welding and assembly was carried out according to the standard procedure to meet the requirement. A design, manufacture, utilization of reactor core and the other relevant uses were fit to class ''T'' to certify the whole process as general. And design, fabrication, analytical test, materials and accessory were carried out based on the ASME, ASTM, ANSI, AWS, JIS and KS standard.

  9. Irradiation test HT-31: high-temperature irradiation behavior of LASL-made extruded fuel rods and LASL-made coated particles

    International Nuclear Information System (INIS)

    Wagner, P.; Reiswig, R.D.; Hollabaugh, C.M.; White, R.W.; Davidson, K.V.; Schell, D.H.

    1977-04-01

    Three LASL-made extruded graphite and coated particle fuel rods have been irradiated in the Oak Ridge National Laboratory High Fluence Isotope Reactor test HT-31. Test conditions were about 9 x 10 21 nvt(E > .18 MeV) at 1250 0 C. The graphite matrix showed little or no effect of the irradiation. LASL-made ZrC containing coated particles with ZrC coats and ZrC-doped pyrolytic carbon coats showed no observable effects of the irradiation

  10. Assessment of the prediction capability of the TRANSURANUS fuel performance code on the basis of power ramp tested LWR fuel rods

    International Nuclear Information System (INIS)

    Pastore, G.; Botazzoli, P.; Di Marcello, V.; Luzzi, L.

    2009-01-01

    The present work is aimed at assessing the prediction capability of the TRANSURANUS code for the performance analysis of LWR fuel rods under power ramp conditions. The analysis refers to all the power ramp tested fuel rods belonging to the Studsvik PWR Super-Ramp and BWR Inter-Ramp Irradiation Projects, and is focused on some integral quantities (i.e., burn-up, fission gas release, cladding creep-down and failure due to pellet cladding interaction) through a systematic comparison between the code predictions and the experimental data. To this end, a suitable setup of the code is established on the basis of previous works. Besides, with reference to literature indications, a sensitivity study is carried out, which considers the 'ITU model' for fission gas burst release and modifications in the treatment of the fuel solid swelling and the cladding stress corrosion cracking. The performed analyses allow to individuate some issues, which could be useful for the future development of the code. Keywords: Light Water Reactors, Fuel Rod Performance, Power Ramps, Fission Gas Burst Release, Fuel Swelling, Pellet Cladding Interaction, Stress Corrosion Cracking

  11. Computer gaming comes to service rig training

    Energy Technology Data Exchange (ETDEWEB)

    Mowers, J.

    2007-05-15

    This article addressed the challenge of providing service rig workers with a good understanding of the tasks and risks involved in the job before they even step out into the field. The product, SimuLynx was presented. SimuLynx is based on video and gaming technology to immerse the user in the service rig work environment with other crew members. The user tries to perform the different steps of a junior floorhand's job while a coach gives directions. The article discussed how the system works. For example, when faced with a task, the user chooses from several options. The coach informs the virtual junior floorhand if the decision was right or wrong. He will also give warnings and let the user realize the consequences of a wrong action. The benefits of the system were also presented. For example, instead of 30 days of on-the-job training, an employee may only need several days after going through the program. Other benefits that were reviewed included reduced accident risk-levels for new workers; higher rig efficiency; and, lower training costs. In addition, a potential recruit can decide if the job is right for him before signing up for an expensive course or starting off with a service rig company. As well, the rig company can decide if someone is suitable before hiring that person. 3 figs.

  12. Fuel rods

    International Nuclear Information System (INIS)

    Fukushima, Kimichika.

    1984-01-01

    Purpose: To reduce the size of the reactor core upper mechanisms and the reactor container, as well as decrease the nuclear power plant construction costs in reactors using liquid metals as the coolants. Constitution: Isotope capturing devices comprising a plurality of pipes are disposed to the gas plenum portion of a nuclear fuel rod main body at the most downstream end in the flowing direction of the coolants. Each of the capturing devices is made of nickel, nickel alloys, stainless steel applied with nickel plating on the surface, nickel alloys applied with nickel plating on the surface or the like. Thus, radioactive nuclides incorporated in the coolants are surely captured by the capturing devices disposed at the most downstream end of the nuclear fuel main body as the coolants flow along the nuclear fuel main body. Accordingly, since discharging of radioactive nuclides to the intermediate fuel exchange system can be prevented, the maintenance or reparing work for the system can be facilitated. (Moriyama, K.)

  13. Control rods

    International Nuclear Information System (INIS)

    Hirukawa, Koji.

    1979-01-01

    Purpose: To ensure the fuel safety by constituting a control rod with a plurality of poison bodies suspended in a cross-like section and shorter length poison bodies made movable and engageable in the gap between each of the above poison bodies thereby maintaining the function of the shorter length poison constant. Constitution: Cross-like supports are secured to the upper and lower parts of a driving shaft journaled in a sheath and poison bodies composed of neutron absorber poisons of a large thermal neutron absorption cross section and neutron absorber poison tubes for containing them are suspended from the supports. A movable cross-like support is mounted slidably at its base to the lower part of the driving shaft and poison bodies shorter than the above poison bodies and composed of neutron absorber poisons having a greater absorption cross section at the neutron energy region higher than thermal neutron region and neutron poison tubes for containing them are suspended to the movable support at the position capable of engaging in the gap between each of the poison bodies. (Kawakami, Y.)

  14. A constructive presentation of rigged Hilbert spaces

    International Nuclear Information System (INIS)

    Celeghini, Enrico

    2015-01-01

    We construct a rigged Hilbert space for the square integrable functions on the line L2(R) adding to the generators of the Weyl-Heisenberg algebra a new discrete operator, related to the degree of the Hermite polynomials. All together, continuous and discrete operators, constitute the generators of the projective algebra io(2). L 2 (R) and the vector space of the line R are shown to be isomorphic representations of such an algebra and, as both these representations are irreducible, all operators defined on the rigged Hilbert spaces L 2 (R) or R are shown to belong to the universal enveloping algebra of io(2). The procedure can be extended to orthogonal and pseudo-orthogonal spaces of arbitrary dimension by tensorialization.Circumventing all formal problems the paper proposes a kind of toy model, well defined from a mathematical point of view, of rigged Hilbert spaces where, in contrast with the Hilbert spaces, operators with different cardinality are allowed. (paper)

  15. Study in flow rig by using radiotracer

    International Nuclear Information System (INIS)

    Widatalla, R. K.

    2012-06-01

    Application of radioisotope technology have proved itself to be effective techniques for troubleshooting and optimizing industrial process in petrochemical industry In this study gamma scanning technique has been employed for better understanding of malfunctions by using the flow rig system. The scanning were carried out using 9 9mT C gamma radiation source with activity of 1 mCi and quantity of 5 ml to measure the flow rate for the water flow rig The experiment was repeated by reducing the data interval time to get more precise result. The investigations were also carried out using 5 ml of 9 9mT C with activity of 0.3 mCi for measuring the Residence Time Distribution (RTD) inside the flow rig tank which enables calculating the effective volume for the operating tank and its dead volume. The results proved that the technique is sensitive, reliable and can be adopted to investigate industrial reactors. (Author)

  16. 5 X 5 rod bundle flow field measurements downstream a PWR spacer grid

    Energy Technology Data Exchange (ETDEWEB)

    Castro, Higor F.P.; Silva, Vitor V A.; Santos, André A.C.; Veloso, Maria A.F., E-mail: higorfabiano@gmail.com, E-mail: mdora@nuclear.ufmg.br, E-mail: vitors@cdtn.br, E-mail: aacs@cdtn.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil); Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2017-07-01

    The spacer grids are structures present in nuclear fuel assembly of Pressurized Water Reactors (PWR). They play an important structural role and also assist in heat removal through the assembly by promoting increased turbulence of the flow. Understanding the flow dynamics downstream the spacer grid is paramount for fuel element design and analysis. This paper presents water flow velocity profiles measurements downstream a spacer grid in a 5 x 5 rod bundle test rig with the objective of highlighting important fluid dynamic behavior near the grid and supplying data for CFD simulation validation. These velocity profiles were obtained at two different heights downstream the spacer grid using a LDV (Laser Doppler Velocimetry) through the top of test rig. The turbulence intensities and patterns of the swirl and cross flow were evaluated. The tests were conducted for Reynolds numbers ranging from 1.8 x 10{sup 4} to 5.4 x 10{sup 4}. This experimental research was carried out in thermo-hydraulics laboratory of Nuclear Technology Development Center – CDTN. Results show great repeatability and low uncertainties (< 1.24 %). Details of the flow field show how the mixture and turbulence induced by the spacer grid quickly decays downstream the spacer grid. It is shown that the developed methodology can supply high resolution low uncertainty results that can be used for validation of CFD simulations. (author)

  17. 5 X 5 rod bundle flow field measurements downstream a PWR spacer grid

    International Nuclear Information System (INIS)

    Castro, Higor F.P.; Silva, Vitor V A.; Santos, André A.C.; Veloso, Maria A.F.

    2017-01-01

    The spacer grids are structures present in nuclear fuel assembly of Pressurized Water Reactors (PWR). They play an important structural role and also assist in heat removal through the assembly by promoting increased turbulence of the flow. Understanding the flow dynamics downstream the spacer grid is paramount for fuel element design and analysis. This paper presents water flow velocity profiles measurements downstream a spacer grid in a 5 x 5 rod bundle test rig with the objective of highlighting important fluid dynamic behavior near the grid and supplying data for CFD simulation validation. These velocity profiles were obtained at two different heights downstream the spacer grid using a LDV (Laser Doppler Velocimetry) through the top of test rig. The turbulence intensities and patterns of the swirl and cross flow were evaluated. The tests were conducted for Reynolds numbers ranging from 1.8 x 10"4 to 5.4 x 10"4. This experimental research was carried out in thermo-hydraulics laboratory of Nuclear Technology Development Center – CDTN. Results show great repeatability and low uncertainties (< 1.24 %). Details of the flow field show how the mixture and turbulence induced by the spacer grid quickly decays downstream the spacer grid. It is shown that the developed methodology can supply high resolution low uncertainty results that can be used for validation of CFD simulations. (author)

  18. Design and Checkout of a High Speed Research Nozzle Evaluation Rig

    Science.gov (United States)

    Castner, Raymond S.; Wolter, John D.

    1997-01-01

    The High Flow Jet Exit Rig (HFJER) was designed to provide simulated mixed flow turbojet engine exhaust for one- seventh scale models of advanced High Speed Research test nozzles. The new rig was designed to be used at NASA Lewis Research Center in the Nozzle Acoustic Test Rig and the 8x6 Supersonic Wind Tunnel. Capabilities were also designed to collect nozzle thrust measurement, aerodynamic measurements, and acoustic measurements when installed at the Nozzle Acoustic Test Rig. Simulated engine exhaust can be supplied from a high pressure air source at 33 pounds of air per second at 530 degrees Rankine and nozzle pressure ratios of 4.0. In addition, a combustion unit was designed from a J-58 aircraft engine burner to provide 20 pounds of air per second at 2000 degrees Rankine, also at nozzle pressure ratios of 4.0. These airflow capacities were designed to test High Speed Research nozzles with exhaust areas from eighteen square inches to twenty-two square inches. Nozzle inlet flow measurement is available through pressure and temperature sensors installed in the rig. Research instrumentation on High Speed Research nozzles is available with a maximum of 200 individual pressure and 100 individual temperature measurements. Checkout testing was performed in May 1997 with a 22 square inch ASME long radius flow nozzle. Checkout test results will be summarized and compared to the stated design goals.

  19. Prototypical Rod Consolidation Demonstration Project

    International Nuclear Information System (INIS)

    1993-05-01

    The objective of Phase 3 of the Prototypical Rod consolidation Demonstration Project (PRCDP) was to procure, fabricate, assemble, and test the Prototypical Rod consolidation System as described in the NUS Phase 2 Final Design Report. This effort required providing the materials, components, and fabricated parts which makes up all of the system equipment. In addition, it included the assembly, installation, and setup of this equipment at the Cold Test Facility. During the Phase 3 effort the system was tested on a component, subsystem, and system level. This volume 1, discusses the PRCDP Phase 3 Test Program that was conducted by the HALLIBURTON NUS Environmental Corporation under contract AC07-86ID12651 with the United States Department of Energy. This document, Volume 1, Book 2 discusses the following topics: Fuel Rod Extraction System Test Results and Analysis Reports and Clamping Table Test Results and Analysis Reports

  20. Control rod drives

    International Nuclear Information System (INIS)

    Nakamura, Akira.

    1984-01-01

    Purpose: To enable to monitor the coupling state between a control rod and a control rod drive. Constitution: After the completion of a control rod withdrawal, a coolant pressure is applied to a control rod drive being adjusted so as to raise only the control rod drive and, in a case where the coupling between the control rod drive and the control rod is detached, the former is elevated till it contacts the control rod and then stopped. The actual stopping position is detected by an actual position detection circuit and compared with a predetermined position stored in a predetermined position detection circuit. If both of the positions are not aligned with each other, it is judged by a judging circuit that the control rod and the control rod drives are not combined. (Sekiya, K.)

  1. Parametric study of the behaviour of a pre irradiated BWR fuel rod under conditions of LOCA simulated in the halden in pile test system with the FALCON code

    Energy Technology Data Exchange (ETDEWEB)

    Khvostov, G.; Zimmermann, M. A. [Laboratory for Reactor Physics and Systems Behaviour, Paul Scherrer Institut, Villigen (Switzerland); Ledergerber, G. [Kernkraftwerk Leibstadt AG, Leibstadt (Switzerland); Kolstad, E. [Institute for Energy Technology - OECD Halden Reactor Project, Halden (Norway); Montgomery, R. O. [Anatech Corporation, San Diego (United States)

    2008-10-15

    A new LOCA test at Halden was planned as the first experiment within the Halden LOCA program addressing the behaviour of commercially irradiated BWR fuel of medium burn up with burst of the cladding expected to occur at a temperature of about 1050.deg.C, which is essentially higher than in the preceding experiments. The specific measures to be adopted have been suggested based upon a parametric study using the FALCON fuel behaviour code and aimed at an optimized design of the test fuel rod for the given high target cladding temperature of 1150 .deg. C (peak local). The analysis has shown a reasonable agreement with the fundamental experimental findings, such as correlations of NUREG 0630, as well as consistency with the data from Halden LOCA testing available so far. Thus, a general conclusion is drawn about the applicability of the methodology developed at PSI to the analysis of LWR fuel rod behaviour during LOCA, in consideration of the effects of fuel burn up.

  2. Deep water challenges for drilling rig design

    Energy Technology Data Exchange (ETDEWEB)

    Roth, M [Transocean Sedco Forex, Houston, TX (United States)

    2001-07-01

    Drilling rigs designed for deep water must meet specific design considerations for harsh environments. The early lessons for rig design came from experiences in the North Sea. Rig efficiency and safety considerations must include structural integrity, isolated/redundant ballast controls, triple redundant DP systems, enclosed heated work spaces, and automated equipment such as bridge cranes, pipe handling gear, offline capabilities, subsea tree handling, and computerized drill floors. All components must be designed to harmonize man and machine. Some challenges which are unique to Eastern Canada include frequent storms and fog, cold temperature, icebergs, rig ice, and difficult logistics. This power point presentation described station keeping and mooring issues in terms of dynamic positioning issues. The environmental influence on riser management during forced disconnects was also described. Design issues for connected deep water risers must insure elastic stability, and control deflected shape. The design must also keep stresses within acceptable limits. Codes and standards for stress limits, flex joints and tension were also presented. tabs., figs.

  3. Customization creates more efficient, cleaner rigs

    Energy Technology Data Exchange (ETDEWEB)

    Budd, G.

    2002-08-01

    Technological advances in drilling equipment are essential to improving efficiency in the oilpatch; getting the technological upper hand on the competition is no less important for drilling equipment manufacturers than for actors in other sectors of the industry. While off-the-shelf uniformity that reduces unit cost has been the trend in fabricating field gas compression modules, custom manufacturing has become very popular in the rig manufacturing sector. Examples from Crown Energy Technologies and Tesco Corporation, both of Calgary, Aecon Industrial's Edmonton operations, PCL Industrial Construction Ltd of Nisku, and Toromont Process Systems of Houston and Calgary are described to illustrate the widespread demand for customized drilling rigs, including the growing preference for electric drives. Top drive systems, as opposed to rotary drives also have become very popular; six out of ten rigs are sold with electric top drives today compared with fewer than 10 rigs a decade ago. At the same time, Tesco has recently signed a deal with Conoco Inc to construct three revolutionary drilling rigs using Tesco's proprietary Casing Drilling Technology, which uses standard oilfield casing instead of drill pipe, allowing operators to simultaneously drill, case and evaluate oil and gas wells. Aecon and PCL Industrial Construction have had much demand for customized spools and modules particularly from the oil sands industry, while Toromont Process Systems is also expanding its Calgary facilities to meet the demand for its dual gas compression equipment used by power stations for gas compression and co-generation, natural gas refrigeration and specialty gas processing.

  4. Numerical Tests for the Problem of U-Pu Fuel Burnup in Fuel Rod and Polycell Models Using the MCNP Code

    Science.gov (United States)

    Muratov, V. G.; Lopatkin, A. V.

    An important aspect in the verification of the engineering techniques used in the safety analysis of MOX-fuelled reactors, is the preparation of test calculations to determine nuclide composition variations under irradiation and analysis of burnup problem errors resulting from various factors, such as, for instance, the effect of nuclear data uncertainties on nuclide concentration calculations. So far, no universally recognized tests have been devised. A calculation technique has been developed for solving the problem using the up-to-date calculation tools and the latest versions of nuclear libraries. Initially, in 1997, a code was drawn up in an effort under ISTC Project No. 116 to calculate the burnup in one VVER-1000 fuel rod, using the MCNP Code. Later on, the authors developed a computation technique which allows calculating fuel burnup in models of a fuel rod, or a fuel assembly, or the whole reactor. It became possible to apply it to fuel burnup in all types of nuclear reactors and subcritical blankets.

  5. Testing of improved polyimide actuator rod seals at high temperature and under vacuum conditions for use in advanced aircraft hydraulic systems

    Science.gov (United States)

    Sellereite, B. K.; Waterman, A. W.; Nelson, W. G.

    1974-01-01

    Polyimide second-stage rod seals were evaluated to determine their suitability for applications in space station environments. The 6.35-cm (2.5-in.)K-section seal was verified for thermal cycling operation between room temperature and 478 K (400 F) and for operation in a 133 micron PA(0.000001 mm Hg) vacuum environment. The test seal completed the scheduled 96 thermal cycles and 1438 hr in vacuum with external rod seal leakage well within the maximum allowable of two drops per 25 actuation cycles. At program completion, the seals showed no signs of structural degradation. Posttest inspection showed the seals retained a snug fit against the shaft and housing walls, indicating additional wear life capability. Evaluation of a molecular flow section during vacuum testing, to inhibit fluid loss through vaporization, showed it to be beneficial with MIL-H-5606, a petroleum-base fluid, in comparison with MIL-H-83282, a synthetic hydrocarbon-base fluid.

  6. Prototypical Rod Consolidation Demonstration Project

    International Nuclear Information System (INIS)

    1993-05-01

    The objective of Phase III of the Prototypical Rod Consolidation Demonstration Project (PRCDP) was to procure, fabricate, assemble, and test the Prototypical Rod Consolidation System as described in the NUS Phase II Final Design Report. This effort required providing the materials, components, and fabricated parts which makes up all of the system equipment. In addition, it included the assembly, installation, and setup of this equipment at the Cold Test Facility. During the Phase III effort the system was tested on a component, subsystem, and system level. Volume IV provides the Operating and Maintenance Manual for the Prototypical Rod Consolidation System that was installed at the Cold Test Facility. This document, Book 1 of Volume IV, discusses: Process overview functional descriptions; Control system descriptions; Support system descriptions; Maintenance system descriptions; and Process equipment descriptions

  7. Prototypical Rod Consolidation Demonstration Project

    International Nuclear Information System (INIS)

    1993-05-01

    The objective of Phase III of the Prototypical Rod Consolidation Demonstration Project (PRCDP) was to procure, fabricate, assemble, and test the Prototypical Rod Consolidation System as described in the NUS Phase II Final Design Report. This effort required providing the materials, components, and fabricated parts which makes up all of the system equipment. In addition, it included the assembly, installation, and setup of this equipment at the Cold Test Facility. During the Phase III effort the system was tested on a component, subsystem, and system level. Volume IV provides the Operating and Maintenance Manual for the Prototypical Rod Consolidation System that was installed at the Cold Test Facility. This document, Book 4 of Volume IV, discusses: Off-normal operating and recovery procedures; Emergency response procedures; Troubleshooting procedures; and Preventive maintenance procedures

  8. Wear of control rod cluster assemblies and of instrumentation thimbles: first results obtained with the vibrateau wear simulator

    International Nuclear Information System (INIS)

    Zbinden, M.; Hersant, D.

    1993-07-01

    Several REP components are affected by a particular sort of damage called impact/sliding wear. This kind of wear, originating from flow induced vibrations, affects loosely supported tubular structures. The main involved components are: - the RCCAs claddings and the guides tubes, - the instrumentation thimbles, - the fuel rods claddings, - the SG tubes. The R and D Division is concerned with studies aiming to understand and to master the phenomena leading to this wear. The MTC Branch is charged of the study of the wear itself. Tests are carried out on wear rigs to understand and to model wear mechanisms. The following work is related to the two first wear tests campaigns on the VIBRATEAU wear simulator: - a reproducibility test series in order to assess the spreading of the experimental results, - a comparative test series on surface treatments used to improve the components war resistance. (authors). 7 figs., 2 tabs., 4 refs

  9. Control rod assembly

    International Nuclear Information System (INIS)

    Takahashi, Akio.

    1982-01-01

    Purpose: To enable reliable insertion and drops of control rods, as well as insure a sufficient flow rate of coolants flowing through the control rods for attaining satisfactory cooling thereof to enable relexation of thermal stress resulted to rectifying mechanisms or the likes. Constitution: To the outer circumference of a control rod contained vertically movably within a control rod guide tube, resistive members are retractably provided in such a way as to project to close the gap between outer circumference of the control rod and the inner surface of the control rod guide tube upon engagement of a gripper of control rod drives, and retract upon release of the engagement of the gripper. Thus, since the resistive members project to provide a greater resistance to the coolants flowing between them and the control rod guide tube in the normal operation where the gripper is engaged to drive the control rod by the control rod drives, a major part of the coolant flowing into the control rod guide tube flows into the control rod. This enables to cool the control rod effectively and make the temperature distribution uniform for the coolant flowing from the upper end of the control rod guide tube to thereby attain the relaxation of the thermal stress resulted in the rectifying mechanisms or the likes. (Moriyama, K.)

  10. Control rod displacement

    International Nuclear Information System (INIS)

    Nakazato, S.

    1987-01-01

    This patent describes a nuclear reactor including a core, cylindrical control rods, a single support means supporting the control rods from their upper ends in spaced apart positions and movable for displacing the control rods in their longitudinal direction between a first end position in which the control rods are fully inserted into the core and a second end position in which the control rods are retracted from the core, and guide means contacting discrete regions of the outer surface of each control rod at least when the control rods are in the vicinity of the second end position. The control rods are supported by the support means for longitudinal movement without rotation into and out of the core relative to the guide means to thereby cause the outer surface of the control rods to experience wear as a result of sliding contact with the guide means. The support means are so arranged with respect to the core and the guide means that it is incapable of rotation relative to the guide means. The improvement comprises displacement means being operatively coupled to a respective one of the control rods for periodically rotating the control rod in a single angular direction through an angle selected to change the locations on the outer surfaces of the control rods at which the control rods are contacted by the guide means during subsequent longitudinal movement of the control rods

  11. Irradiation test OF-2: high-temperature irradiation behavior of LASL-made fuel rods and LASL-made coated particles

    International Nuclear Information System (INIS)

    Wagner, P.; Reiswig, R.D.; Hollabaugh, C.M.; White, R.W.; O'Rourke, J.A.; Davidson, K.V.; Schell, D.H.

    1977-10-01

    Three LASL-made, substoichiometric ZrC-coated particles with inert kernels, and two high-density molded graphite fuel rods that contained LASL-made, ZrC-coated fissile particles were irradiated in the Oak Ridge Research Reactor test OF-2. The severest test conditions were 8.36 x 10 21 nvt (E greater than 0.18 MeV) at 1350 0 C. The graphite matrix showed no effect of the irradiation. There was no interaction between the matrix and any of the particle coats. The loose ZrC coated particles with inert kernels showed no irradiation effects. The graded ZrC-C coats on the fissile particles were cracked. It is postulated that the cracking is associated with the low LTI deposition rate and is not related to the ZrC

  12. Mach 0.3 Burner Rig Facility at the NASA Glenn Materials Research Laboratory

    Science.gov (United States)

    Fox, Dennis S.; Miller, Robert A.; Zhu, Dongming; Perez, Michael; Cuy, Michael D.; Robinson, R. Craig

    2011-01-01

    This Technical Memorandum presents the current capabilities of the state-of-the-art Mach 0.3 Burner Rig Facility. It is used for materials research including oxidation, corrosion, erosion and impact. Consisting of seven computer controlled jet-fueled combustors in individual test cells, these relatively small rigs burn just 2 to 3 gal of jet fuel per hour. The rigs are used as an efficient means of subjecting potential aircraft engine/airframe advanced materials to the high temperatures, high velocities and thermal cycling closely approximating actual operating environments. Materials of various geometries and compositions can be evaluated at temperatures from 700 to 2400 F. Tests are conducted not only on bare superalloys and ceramics, but also to study the behavior and durability of protective coatings applied to those materials.

  13. Seismic scrammability of HTTR control rods

    International Nuclear Information System (INIS)

    Nishiguchi, I.; Iyoku, T.; Ito, N.; Watanabe, Y.; Araki, T.; Katagiri, S.

    1990-01-01

    Scrammability tests on HTTR (High-Temperature Engineering Test Reactor) control rods under seismic conditions have been carried out and seismic conditions influences on scram time as well as functional integrity were examined. A control rod drive located in a stand-pipe at the top of a reactor vessel, raises and lowers a pair of control rods by suspension cables. Each flexible control rod consists of 10 neutron absorber sections held together by a metal spine passing through the center. It falls into a hole in graphite blocks due to gravity at scram. In the tests, a full scale control rod drive and a pair of control rods were employed with a column of graphite blocks in which holes for rods were formed. Blocks misalignment and contact with the hole surface during earthquakes were considered as major causes of disturbance in scram time. Therefore, the following parameters were set up in the tests: excitation direction, combination or horizontal and vertical excitation, acceleration, frequency and block to block gaps. Main results obtained from tests are as follow. 1) Every scram time obtained under the design conditions was within 6 seconds. On the contrary, the scram times were 5.2 seconds when there were no vibration. Therefore, it was concluded that the seismic effects on scram time were not significant. 2) Scram time became longer with increase in both acceleration and horizontal excitation frequency, and control rods fell very smoothly without any jerkiness. This suggests that collision between control rods and hole surface is the main disturbing factor of falling motion. 3) Mechanical and functional integrity of control rod drive mechanism, control rods and graphite blocks was confirmed after 140 seismic scrammability tests. (author). 10 figs, 1 tab

  14. Comparison of energy measurements in the standard penetration test using the cathead and rope method. Phases I and II. Final report

    International Nuclear Information System (INIS)

    Kovacs, W.D.; Salomone, L.A.; Yokel, F.Y.

    1983-11-01

    Studies conducted on the Standard Penetration Test (SPT) and its present use in engineering practice show that a wide variation in the use of different pieces of SPT equipment, procedures and personnel results in a range of energy measured in the drill rods from 30 to 85 percent of the standard SPT energy. The potential energy and kinetic energy of the hammer were measured prior to impact, and the energy passing through the drill rods was calculated from a force-time measurement in the rods. It was found that safety (type) hammers tend to allow more kinetic energy to pass through the hammer-anvil system than donut (type) hammers. The energy passing through the drill rods was calculated by using a digital processing oscilloscope and an SPT Calibrator. Lessons learned in evaluating the energy measurement by these two methods are discussed. The combined effect of the drill rig used, the operator and his procedures, and the SPT equipment should be considered when energy is to be evaluated. The variation of average energy ratio within various drill rig models was found to be about as large as that among drill rig models. It was therefore impossible to make a statistically significant estimate of the reference energy which is representative of the average energy delivered in the US practice. 29 references, 24 figures, 20 tables

  15. Control rod drive mechanism

    International Nuclear Information System (INIS)

    Nakamura, Akira.

    1981-01-01

    Purpose: To ensure the scram operation of a control rod by the reliable detection for the position of control rods. Constitution: A permanent magnet is provided to the lower portion of a connecting rod in engagement with a control rod and a tube having a plurality of lead switches arranged axially therein in a predetermined pitch is disposed outside of the control rod drives. When the control rod moves upwardly in the scram operation, the lead switches are closed successively upon passage of the permanent magnet to operate the electrical circuit provided by way of each of the lead switches. Thus, the position for the control rod during the scram can reliably be determined and the scram characteristic of the control rod can be recognized. (Furukawa, Y.)

  16. Prototypical Rod Consolidation Demonstration Project

    International Nuclear Information System (INIS)

    1993-05-01

    The objective of Phase 3 of the Prototypical Rod consolidation Demonstration Project (PRCDP) was to procure, fabricate, assemble, and test the Prototypical Rod consolidation System as described in the NUS Phase 2 Final Design Report. This effort required providing the materials, components, and fabricated parts which makes up all of the system equipment. In addition, it included the assembly, installation, and setup of this equipment at the Cold Test Facility. During the Phase 3 effort the system was tested on a component, subsystem, and system level. This volume 1, discusses the PRCDP Phase 3 Test Program that was conducted by the HALLIBURTON NUS Environmental Corporation under contract AC07-86ID12651 with the United States Department of Energy. This document, Volume 1, Book 1 discusses the following topics: the background of the project; test program description; summary of tests and test results; problem evaluation; functional requirements confirmation; recommendations; and completed test documentation for tests performed in Phase 3

  17. Development of a control rod drive

    International Nuclear Information System (INIS)

    1991-01-01

    In the period under review, the computer codes required for transients calculation have been completed, as well as the programs for modelling and testing the hot-gas temperature control by means of combined core rod and reflector rod operation. The specification of requirements to be fulfilled by the rod drive computer and the neutron flux measuring system has been done relying essentially on the data obtained by the transients calculations performed and the resulting informations on operating conditions. The work for optimization of the core rod drive with regard to rod driving speeds and the 'three-point switch' with hysteresis for controlled, automatic core rod operation has been concentrating on the case of specified, normal operation of the reactor. (orig./DG) [de

  18. Benchmark Analyses on the Control Rod Withdrawal Tests Performed During the PHÉNIX End-of-Life Experiments. Report of a Coordinated Research Project 2008–2011

    International Nuclear Information System (INIS)

    2014-06-01

    The IAEA supports Member State activities in advanced fast reactor technology development by providing a major fulcrum for information exchange and collaborative research programmes. The IAEA’s activities in this field are mainly carried out within the framework of the Technical Working Group on Fast Reactors (TWG-FR), which assists in the implementation of corresponding IAEA activities and ensures that all technical activities are in line with the expressed needs of Member States. In the broad range of activities, the IAEA proposes and establishes coordinated research projects (CRPs) aimed at improving Member States’ capabilities in fast reactor design and analysis. An important opportunity to conduct collaborative research activities was provided by the experimental campaign run by the French Alternative Energies and Atomic Energy Commission (CEA, Commissariat à l’énergie atomique et aux énergies alternatives) at the PHÉNIX, a prototype sodium cooled fast reactor. Before the definitive shutdown in 2009, end-of-life tests were conducted to gather additional experience on the operation of sodium cooled reactors. Thanks to the CEA opening the experiments to international cooperation, the IAEA decided in 2007 to launch the CRP entitled Control Rod Withdrawal and Sodium Natural Circulation Tests Performed during the PHÉNIX End-of-Life Experiments. The CRP, together with institutes from seven States, contributed to improving capabilities in sodium cooled fast reactor simulation through code verification and validation, with particular emphasis on temperature and power distribution calculations and the analysis of sodium natural circulation phenomena. The objective of this publication is to document the results and main achievements of the benchmark analyses on the control rod withdrawal test performed within the framework of the PHÉNIX end-of-life experimental campaign

  19. Benchmark Analyses on the Control Rod Withdrawal Tests Performed During the PHÉNIX End-of-Life Experiments. Report of a Coordinated Research Project 2008–2011

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2014-06-15

    The IAEA supports Member State activities in advanced fast reactor technology development by providing a major fulcrum for information exchange and collaborative research programmes. The IAEA’s activities in this field are mainly carried out within the framework of the Technical Working Group on Fast Reactors (TWG-FR), which assists in the implementation of corresponding IAEA activities and ensures that all technical activities are in line with the expressed needs of Member States. In the broad range of activities, the IAEA proposes and establishes coordinated research projects (CRPs) aimed at improving Member States’ capabilities in fast reactor design and analysis. An important opportunity to conduct collaborative research activities was provided by the experimental campaign run by the French Alternative Energies and Atomic Energy Commission (CEA, Commissariat à l’énergie atomique et aux énergies alternatives) at the PHÉNIX, a prototype sodium cooled fast reactor. Before the definitive shutdown in 2009, end-of-life tests were conducted to gather additional experience on the operation of sodium cooled reactors. Thanks to the CEA opening the experiments to international cooperation, the IAEA decided in 2007 to launch the CRP entitled Control Rod Withdrawal and Sodium Natural Circulation Tests Performed during the PHÉNIX End-of-Life Experiments. The CRP, together with institutes from seven States, contributed to improving capabilities in sodium cooled fast reactor simulation through code verification and validation, with particular emphasis on temperature and power distribution calculations and the analysis of sodium natural circulation phenomena. The objective of this publication is to document the results and main achievements of the benchmark analyses on the control rod withdrawal test performed within the framework of the PHÉNIX end-of-life experimental campaign.

  20. When War Rigs the Vote

    DEFF Research Database (Denmark)

    Hansen, Bertel Teilfeldt

    2014-01-01

    This paper investigates the effect of intrastate conflict on electoral manipulation. Using a rationalist bargaining model, it produces a hypothesis stating that actors in post-conflict elections will have increased incentive to reallocate seats through manipulation. To test this causal claim a new...... % seat threshold critical for obtaining absolute majority, the intensity of intrastate conflict before each election exhibits a large, positive jump right at the cut-off. This is interpreted as evidence of conflict having a substantial, manipulation-inducing effect on the largest parties in parliament...... – in the aftermath of war they tend to tamper with election results in order to gain absolute majority....

  1. Prototypical Rod Construction Demonstration Project

    International Nuclear Information System (INIS)

    1993-05-01

    The objective of Phase 3 of the Prototypical Rod consolidation Demonstration Project (PRCDP) was to procure, fabricate, assemble, and test the Prototypical Rod consolidation System as described in the NUS Phase 2 Final Design Report. This effort required providing the materials, components, and fabricated parts which makes up all of the system equipment. In addition, it included the assembly, installation, and setup of this equipment at the Cold Test Facility. During the Phase 3 effort the system was tested on a component, subsystem, and system level. This volume 1, discusses the PRCDP Phase 3 Test Program that was conducted by the HALLIBURTON NUS Environmental Corporation under contract AC07-86ID12651 with the United States Department of Energy. This document, Volume 1, Book 3 discusses the following topics: Downender Test Results and Analysis Report; NFBC Canister Upender Test Results and Analysis Report; Fuel Assembly Handling Fixture Test Results and Analysis Report; and Fuel Canister Upender Test Results and Analysis Report

  2. Mechanical properties of bioresorbable self-reinforced posterior cervical rods.

    Science.gov (United States)

    Savage, Katherine; Sardar, Zeeshan M; Pohjonen, Timo; Sidhu, Gursukhman S; Eachus, Benjamin D; Vaccaro, Alexander

    2014-04-01

    A biomechanical study. To test the mechanical and physical properties of self-reinforced copolymer bioresorbable posterior cervical rods and compare their mechanical properties to commonly used Irene titanium alloy rods. Bioresorbable instrumentation is becoming increasingly common in surgical spine procedures. Compared with metallic implants, bioresorbable implants are gradually reabsorbed as the bone heals, transferring the load from the instrumentation to bone, eliminating the need for hardware removal. In addition, bioresorbable implants produce less stress shielding due to a more physiological modulus of elasticity. Three types of rods were used: (1) 5.5 mm copolymer rods and (2) 3.5 mm and (3) 5.5 mm titanium alloy rods. Four tests were used on each rod: (1) 3-point bending test, (2) 4-point bending test, (3) shear test, and (4) differential scanning calorimeter test. The outcomes were recorded: Young modulus (E), stiffness, maximum load, deflection at maximum load, load at 1.0% strain of the rod's outer surface, and maximum bending stress. The Young modulus (E) for the copolymer rods (mean range, 6.4-6.8 GPa) was significantly lower than the 3.5 mm titanium rods (106 GPa) and the 5.5 mm titanium rods (95 GPa). The stiffness of the copolymer rods (mean range, 16.6-21.4 N/mm) was also significantly lower than the 3.5 mm titanium alloy rods (43.6 N/mm) and the 5.5 mm titanium alloy rods (239.6 N/mm). The mean maximum shear load of the copolymer rods was 2735 N and they had significantly lower mean maximum loads than the titanium rods. Copolymer rods have adequate shear resistance, but less load resistance and stiffness compared with titanium rods. Their stiffness is closer to that of bone, causing less stress shielding and better gradual dynamic loading. Their use in semirigid posterior stabilization of the cervical spine may be considered.

  3. Control rod drive

    International Nuclear Information System (INIS)

    Hawke, B.C.

    1986-01-01

    A reactor core, one or more control rods, and a control rod drive are described for selectively inserting and withdrawing the one or more control rods into and from the reactor core, which consists of: a support structure secured beneath the reactor core; control rod positioning means supported by the support structure for movably supporting the control rod for movement between a lower position wherein the control rod is located substantially beneath the reactor core and an upper position wherein at least an upper portion of the control rod extends into the reactor core; transmission means; primary drive means connected with the control rod positioning means by the transmission means for positioning the control rod under normal operating conditions; emergency drive means for moving the control rod from the lower position to the upper position under emergency conditions, the emergency drive means including a weight movable between an upper and a lower position, means for movably supporting the weight, and means for transmitting gravitational force exerted on the weight to the control rod positioning means to move the control rod upwardly when the weight is pulled downwardly by gravity; the transmission means connecting the control rod positioning means with the emergency drive means so that the primary drive means effects movement of the weight and the control rod in opposite directions under normal conditions, thus providing counterbalancing to reduce the force required for upward movement of the control rod under normal conditions; and restraint means for restraining the fall of the weight under normal operating conditions and disengaging the primary drive means to release the weight under emergency conditions

  4. Injecting rabies immunoglobulin (RIG) into wounds only: A significant saving of lives and costly RIG.

    Science.gov (United States)

    Bharti, Omesh Kumar; Madhusudana, Shampur Narayan; Wilde, Henry

    2017-04-03

    An increasing number of dog bite victims were being presented to public hospitals in Himachal Pradesh in 2014 amidst virtual non availability of any rabies immunoglobulin (RIG). Only a small quantity of equine rabies immunoglobulin (eRIG) was available from the government owned Central Research Institute (CRI) Kasauli. This available eRIG was used in 269 patients as an emergency response and only for local infiltration of severe bite wounds by suspected rabid dogs. This was followed by rabies vaccination, using the WHO approved intra-dermal Thai Red Cross Society vaccination schedule. A subgroup of 26 patients were later identified who had been severely bitten by laboratory confirmed rabid dogs. They were followed for more than one year and all were found to be alive.

  5. Fuel rod leak detector

    International Nuclear Information System (INIS)

    Womack, R.E.

    1978-01-01

    A typical embodiment of the invention detects leaking fuel rods by means of a radiation detector that measures the concentration of xenon-133 ( 133 Xe) within each individual rod. A collimated detector that provides signals related to the energy of incident radiation is aligned with one of the ends of a fuel rod. A statistically significant sample of the gamma radiation (γ-rays) that characterize 133 Xe is accumulated through the detector. The data so accumulated indicates the presence of a concentration of 133 Xe appropriate to a sound fuel rod, or a significantly different concentration that reflects a leaking fuel rod

  6. Prototypical Rod Consolidation Demonstration Project

    International Nuclear Information System (INIS)

    1993-05-01

    The objective of Phase 3 of the Prototypical Rod consolidation Demonstration Project (PRCDP) was to procure, fabricate, assemble, and test the Prototypical Rod consolidation System as described in the NUS Phase 2 Final Design Report. This effort required providing the materials, components, and fabricated parts which makes up all of the system equipment. In addition, it included the assembly, installation, and setup of this equipment at the Cold Test Facility. During the Phase 3 effort the system was tested on a component, subsystem, and system level. This volume 1, discusses the PRCDP Phase 3 Test Program that was conducted by the HALLIBURTON NUS Environmental Corporation under contract AC07-86ID12651 with the United States Department of Energy. This document, Volume 1, Book 9 discusses the following topics: Integrated System Normal Operations Test Results and Analysis Report; Integrated System Off-Normal Operations Test Results and Analysis Report; and Integrated System Maintenance Operations Test Results and Analysis Report

  7. Stress-life relation of the rolling-contact fatigue spin rig

    Science.gov (United States)

    Butler, Robert H; Carter, Thomas L

    1957-01-01

    The rolling-contact fatigue spin rig was used to test groups of SAE 52100 9.16-inch-diameter balls lubricated with a mineral oil at 600,000-, 675,000-, and 750,000-psi maximum Hertz stress. Cylinders of AISI M-1 vacuum and commercial melts and MV-1 (AISI M-50) were used as race specimens. Stress-life exponents produced agree closely with values accepted in industry. The type of failure obtained in the spin rig was similar to the subsurface fatigue spells found in bearings.

  8. Refabricated and instrumented fuel rods

    International Nuclear Information System (INIS)

    Silberstein, K.

    2005-01-01

    Nuclear Fuel for power reactors capabilities evaluation is strongly based on the intimate knowledge of its behaviour under irradiation. This knowledge can be acquired from refabricated and instrumented fuel rods irradiated at different levels in commercial reactors. This paper presents the development and qualification of a new technique called RECTO related to a double-instrumented rod re-fabrication process developed by CEA/LECA hot laboratory facility at CADARACHE. The technique development includes manufacturing of the properly dimensioned cavity in the fuel pellet stack to house the thermocouple and the use of a newly designed pressure transducer. An analytic irradiation of such a double-instrumented fuel rod will be performed in OSIRIS test reactor starting October 2004. (Author)

  9. Improvement of Measurement Accuracy of Coolant Flow in a Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Jintae; Kim, Jong-Bum; Joung, Chang-Young; Ahn, Sung-Ho; Heo, Sung-Ho; Jang, Seoyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In this study, to improve the measurement accuracy of coolant flow in a coolant flow simulator, elimination of external noise are enhanced by adding ground pattern in the control panel and earth around signal cables. In addition, a heating unit is added to strengthen the fluctuation signal by heating the coolant because the source of signals are heat energy. Experimental results using the improved system shows good agreement with the reference flow rate. The measurement error is reduced dramatically compared with the previous measurement accuracy and it will help to analyze the performance of nuclear fuels. For further works, out of pile test will be carried out by fabricating a test rig mockup and inspect the feasibility of the developed system. To verify the performance of a newly developed nuclear fuel, irradiation test needs to be carried out in the research reactor and measure the irradiation behavior such as fuel temperature, fission gas release, neutron dose, coolant temperature, and coolant flow rate. In particular, the heat generation rate of nuclear fuels can be measured indirectly by measuring temperature variation of coolant which passes by the fuel rod and its flow rate. However, it is very difficult to measure the flow rate of coolant at the fuel rod owing to the narrow gap between components of the test rig. In nuclear fields, noise analysis using thermocouples in the test rig has been applied to measure the flow velocity of coolant which circulates through the test loop.

  10. Technology trends, energy prices affect worldwide rig activity

    International Nuclear Information System (INIS)

    Rappold, K.

    1995-01-01

    The major worldwide offshore rig markets have improved slightly this year, while the onshore markets generally lagged slightly. Offshore rig utilization rates have remained strong worldwide, with some areas reaching nearly 100%. Total worldwide offshore rig (jack ups, semisubmersible, drillships, submersibles, and barges) utilization was about 86%. Offshore drilling activity is driven primarily by oil and natural gas price expectations. Natural gas prices tend to drive North American offshore drilling activity, including the shallow waters in the Gulf of Mexico. International offshore drilling activity and deepwater projects in the Gulf of Mexico are more closely tied to oil prices. The paper discusses US rig count, directional drilling activity, jack up rig demand, semisubmersibles demand, rig replacement costs, and new construction

  11. Evaluation of wheel/rail contact mechanics : roller rig concept design review.

    Science.gov (United States)

    2014-07-01

    A need exists for a new test rig design with advanced sensing technologies that will allow the railroad industry and regulatory : agencies to better understand the wheel-rail contact dynamics and mechanics, especially as it pertains to high-speed rai...

  12. RESULTS OF TESTS OF LIGHTNING-RODS WITH UNIVERSAL CLAMPS BY THE APERIODIC IMPULSES OF CURRENT OF ARTIFICIAL LIGHTNING WITH THE PEAK-TEMPORAL PARAMETERS RATIONED ON FOREIGN STANDARDS

    Directory of Open Access Journals (Sweden)

    M.I. Baranov

    2015-06-01

    Full Text Available Purpose. Test in obedience to the requirements of row of operating foreign standards of round metallic lightning-rods with the flat metallic universal clamps of the special type on firmness to direct action of аperiodic impulses of current of temporal form 10/350 μs by amplitude of 50 кА (N− class and 100 кА (H− class. Methodology. The order of leadthrough of these tests is certain the followings normative documents: International IEC 62305-1: 2010, Russian national GOST R IEC 62305-1-2010 and German national DIN EN 50164-1:2008 Standards. Results. Conducted on a powerful high-voltage pulsed current of artificial linear lightning with the peak-temporal parameters and admittances of test rationed on the indicated foreign standards rationed that all of the lightning-rods tested in collection with universal clamps, isolating holders and ceramic elements of roof of technical building were survive electrodynamics and electrothermal action of in-use single short blow of an artificial storm digit. Originality. First in domestic practice the similar model tests of lightning-rods are conducted with universal clamps, executed from different explorer materials, on firmness to flowing to on by it the indicated large impulsive currents of artificial lightning. Practical value. Real firmness to lightning of round copper and zincked steel lightning-rods is certain with the flat copper, zincked steel and non-rusting steel universal clamps of the special execution.

  13. Resonances, scattering theory and rigged Hilbert spaces

    International Nuclear Information System (INIS)

    Parravicini, G.; Gorini, V.; Sudarshan, E.C.G.

    1979-01-01

    The problem of decaying states and resonances is examined within the framework of scattering theory in a rigged Hilbert space formalism. The stationary free, in, and out eigenvectors of formal scattering theory, which have a rigorous setting in rigged Hilbert space, are considered to be analytic functions of the energy eigenvalue. The value of these analytic functions at any point of regularity, real or complex, is an eigenvector with eigenvalue equal to the position of the point. The poles of the eigenvector families give origin to other eigenvectors of the Hamiltonian; the singularities of the out eigenvector family are the same as those of the continued S matrix, so that resonances are seen as eigenvectors of the Hamiltonian with eigenvalue equal to their location in the complex energy plane. Cauchy theorem then provides for expansions in terms of complete sets of eigenvectors with complex eigenvalues of the Hamiltonian. Applying such expansions to the survival amplitude of a decaying state, one finds that resonances give discrete contributions with purely exponential time behavior; the background is of course present, but explicitly separated. The resolvent of the Hamiltonian, restricted to the nuclear space appearing in the rigged Hilbert space, can be continued across the absolutely continuous spectrum; the singularities of the continuation are the same as those of the out eigenvectors. The free, in and out eigenvectors with complex eigenvalues and those corresponding to resonances can be approximated by physical vectors in the Hilbert space, as plane waves can. The need for having some further physical information in addition to the specification of the total Hamiltonian is apparent in the proposed framework. The formalism is applied to the Lee-Friedrichs model. 48 references

  14. Innovative technology for a cost-effective land rig

    Energy Technology Data Exchange (ETDEWEB)

    Mehra, S.; Bryce, T.

    1996-05-01

    Sedco Forex has recently completed a new land drilling rig, currently deployed in Gabon, that integrates well construction activities with multiskilling to create cost savings across the board in drilling operations. Historically, operators have produced a comprehensive tender package specifying strictly the type and size of individual rig components and the number of personnel required to drill. In this case, the drilling contractor provides a fit-for-purpose rig, consistent with field location, well profile, operator`s priorities, and local constraints.

  15. Innovative technology for a cost-effective land rig

    International Nuclear Information System (INIS)

    Mehra, S.; Bryce, T.

    1996-01-01

    Sedco Forex has recently completed a new land drilling rig, currently deployed in Gabon, that integrates well construction activities with multiskilling to create cost savings across the board in drilling operations. Historically, operators have produced a comprehensive tender package specifying strictly the type and size of individual rig components and the number of personnel required to drill. In this case, the drilling contractor provides a fit-for-purpose rig, consistent with field location, well profile, operator's priorities, and local constraints

  16. Fission reactor control rod

    International Nuclear Information System (INIS)

    Irie, Tomoo.

    1991-01-01

    The present invention concerns a control rod in a PWR type reactor. A control rod has an inner cladding tube and an outer cladding tube disposed coaxially, and a water draining hole is formed at the inside of the inner cladding tube. Neutron absorbers are filled in an annular gap between the outer cladding tube and the inner cladding tube. The water draining hole opens at the lower end thereof to the top end of the control rod and at the upper end thereof to the side of the upper end plug of the control rod. If the control rod is dropped to a control rod guide thimble for reactor scram, coolants from the control rod guide thimble are flown from the lower end of the water draining hole and discharged from the upper end passing through the water draining hole. In this way, water from the control rod guide thimble is removed easily when the control rod is dropped. Further, the discharging amount of water itself is reduced by the provision of the water draining hole. Accordingly, sufficient control rod dropping speed can be attained. (I.N.)

  17. Unstable quantum states and rigged Hilbert spaces

    International Nuclear Information System (INIS)

    Gorini, V.; Parravicini, G.

    1978-10-01

    Rigged Hilbert space techniques are applied to the quantum mechanical treatment of unstable states in nonrelativistic scattering theory. A method is discussed which is based on representations of decay amplitudes in terms of expansions over complete sets of generalized eigenvectors of the interacting Hamiltonian, corresponding to complex eigenvalues. These expansions contain both a discrete and a continuum contribution. The former corresponds to eigenvalues located at the second sheet poles of the S matrix, and yields the exponential terms in the survival amplitude. The latter arises from generalized eigenvectors associated to complex eigenvalues on background contours in the complex plane, and gives the corrections to the exponential law. 27 references

  18. Prototypical Rod Consolidation Demonstration Project

    International Nuclear Information System (INIS)

    1993-05-01

    The objective of Phase 3 of the Prototypical Rod consolidation Demonstration Project (PRCDP) was to procure, fabricate, assemble, and test the Prototypical Rod consolidation System as described in the NUS Phase 2 Final Design Report. This effort required providing the materials, components, and fabricated parts which makes up all of the system equipment. In addition, it included the assembly, installation, and setup of this equipment at the Cold Test Facility. During the Phase 3 effort the system was tested on a component, subsystem, and system level. This volume 1, discusses the PRCDP Phase 3 Test Program that was conducted by the HALLIBURTON NUS Environmental Corporation under contract AC07-86ID12651 with the United States Department of Energy. This document, Volume 1, Book 4 discusses the following topics: Rod Compaction/Loading System Test Results and Analysis Report; Waste Collection System Test Results and Analysis Report; Waste Container Transfer Fixture Test Results and Analysis Report; Staging and Cutting Table Test Results and Analysis Report; and Upper Cutting System Test Results and Analysis Report

  19. Prototypical Rod Consolidation Demonstration Project

    International Nuclear Information System (INIS)

    1993-05-01

    The objective of Phase 3 of the Prototypical Rod consolidation Demonstration Project (PRCDP) was to procure, fabricate, assemble, and test the Prototypical Rod consolidation System as described in the NUS Phase 2 Final Design Report. This effort required providing the materials, components, and fabricated parts which makes up all of the system equipment. In addition, it included the assembly, installation, and setup of this equipment at the Cold Test Facility. During the Phase 3 effort the system was tested on a component, subsystem, and system level. This volume 1, discusses the PRCDP Phase 3 Test Program that was conducted by the HALLIBURTON NUS Environmental Corporation under contract AC07-86ID12651 with the United States Department of Energy. This document, Volume 1, Book 8 discusses Control System SOT Tests Results and Analysis Report. This is a continuation of Book 7

  20. The role of the rigged Hilbert space in quantum mechanics

    International Nuclear Information System (INIS)

    Madrid, Rafael de la

    2005-01-01

    There is compelling evidence that, when a continuous spectrum is present, the natural mathematical setting for quantum mechanics is the rigged Hilbert space rather than just the Hilbert space. In particular, Dirac's braket formalism is fully implemented by the rigged Hilbert space rather than just by the Hilbert space. In this paper, we provide a pedestrian introduction to the role the rigged Hilbert space plays in quantum mechanics, by way of a simple, exactly solvable example. The procedure will be constructive and based on a recent publication. We also provide a thorough discussion on the physical significance of the rigged Hilbert space

  1. Eigenfunction expansions and scattering theory in rigged Hilbert spaces

    Energy Technology Data Exchange (ETDEWEB)

    Gomez-Cubillo, F [Dpt. de Analisis Matematico, Universidad de Valladolid. Facultad de Ciencias, 47011 Valladolid (Spain)], E-mail: fgcubill@am.uva.es

    2008-08-15

    The work reviews some mathematical aspects of spectral properties, eigenfunction expansions and scattering theory in rigged Hilbert spaces, laying emphasis on Lippmann-Schwinger equations and Schroedinger operators.

  2. Fuel rod failure due to marked diametral expansion and fuel rod collapse occurred in the HBWR power ramp experiment

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki

    1985-12-01

    In the power ramp experiment with the BWR type light water loop at the HBWR, the two pre-irradiated fuel rods caused an unexpected pellet-cladding interaction (PCI). One occurred in the fuel rod with small gap of 0.10 mm, which was pre-irradiated up to the burn-up of 14 MWd/kgU. At high power, the diameter of the rod was increased markedly without accompanying significant axial elongation. The other occurred in the rod with a large gap of 0.23 mm, which was pre-irradiated up to the burn-up of 8 MWd/kgU. The diameter of the rod collapsed during a diameter measurement at the maximum power level. The causes of those were investigated in the present study by evaluating in-core data obtained from equipped instruments in the experiment. It was revealed from the investigation that these behaviours were attributed to the local reduction of the coolant flow occurred in the region of a transformer in the ramp rig. The fuel cladding material is seemed to become softened due to temperature increase caused by the local reduction of the coolant flow, and collapsed by the coolant pressure, either locally or wholly depending on the rod diametral gap existed. (author)

  3. Design, irradiation, and post-irradiation examination of the UC and (U,Pu)C fuel rods of the test groups Mol-11/K1 and Mol-11/K2

    International Nuclear Information System (INIS)

    Freund, D.; Elbel, H.; Steiner, H.

    1976-06-01

    The test groups K1 and K2 of the irradiation experiment Mol-11 are reported. Design, irradiation, and post-irradiation examination of the fuel rods irradiated are described. Mol-11/K1 consisted of one fuel rod with UC of 94% T.D. and helium bonding. This test group was intended to prove the high power irradiation capsule in pile. Mol-11/K2 consists of three fuel rods in total. One of these is presently still in the reactor. In this test group mixed carbide fuel of 83% T.D. and 15% Pu content under helium bonding is irradiated. The fuel rod K2-2 was provided with a capillary tube for the continuous measurement of fission gas pressure built up. 1.4988 stainless steel was chosen as cladding material. The final burnup lies between 35 and 70 MWd/kg M. Post-irradiation examination of the two test groups covers a theoretical analysis of the irradiation behaviour. (orig./GSCH) [de

  4. Safety rod driving device

    International Nuclear Information System (INIS)

    Murakami, Kiyonobu; Kurosaki, Akira.

    1988-01-01

    Purpose: To rapidly insert safety rods for a criticality experiment device into a reactor core container to stop the criticality reaction thereby prevent reactivity accidents. Constitution: A cylinder device having a safety rod as a cylinder rod attached with a piston at one end is constituted. The piston is elevated by pressurized air and attracted and fixed by an electromagnet which is a stationary device disposed at the upper portion of the cylinder. If the current supply to the electromagnet is disconnected, the safety rod constituting the cylinder rod is fallen together with the piston to the lower portion of the cylinder. Since the cylinder rod driving device has neither electrical motor nor driving screw as in the conventional device, necessary space can be reduced and the weight is decreased. In addition, since the inside of the nuclear reactor can easily be shielded completely from the external atmosphere, leakage of radioactive materials can be prevented. (Horiuchi, T.)

  5. Detection of failed fuel rods in shrouded BWR fuel assemblies

    International Nuclear Information System (INIS)

    Baero, G.; Boehm, W.; Goor, B.; Donnelly, T.

    1988-01-01

    A manipulator and an ultrasonic testing (UT) technique were developed to identify defective fuel rods in shrouded BWR fuel assemblies. The manipulator drives a UT probe axially through the bottom tie plate into the water channels between the fuel rods. The rotating UT probe locates defective fuel rods by ingressed water which attenuates the UT-signal. (author)

  6. Non-linear friction in reciprocating hydraulic rod seals: Simulation and measurement

    International Nuclear Information System (INIS)

    Bullock, A K; Tilley, D G; Johnston, D N; Bowen, C R; Keogh, P S

    2009-01-01

    Non-linear seal friction can impede the performance of hydraulic actuation systems designed for high precision positioning with favourable dynamic response. Methods for predicting seal friction are required to help develop sealing systems for this type of application. Recent simulation techniques have claimed progress, although have yet to be validated experimentally. A conventional reciprocating rod seal is analysed using established elastohydrodynamic theory and the mixed lubrication Greenwood-Williamson-average Reynolds model. A test rig was used to assess the accuracy of the simulation results for both instroke and outstroke. Inverse hydrodynamic theory is shown to predict a U 0.5 power law between rod speed and friction. Comparison with experimental data shows the theory to be qualitatively inaccurate and to predict friction levels an order of magnitude lower than those measured. It was not possible to model the regions very close to the inlet and outlet due to the high pressure gradients at the edges of the contact. The mixed lubrication model produces friction levels within the correct order of magnitude, although incorrectly predicts higher friction during instroke than outstroke. Previous experiments have reported higher friction during instroke than outstroke for rectangular seals, suggesting that the mixed lubrication model used could possibly be suitable for symmetric seals, although not for seal tribology in general.

  7. Rod bundle burnout data and correlation comparisons

    International Nuclear Information System (INIS)

    Yoder, G.L.; Morris, D.G.; Mullins, C.B.

    1985-01-01

    Rod bundle burnout data from 30 steady-state and 3 transient tests were obtained from experiments performed in the Thermal Hydraulic Test Facility at the Oak Ridge National Laboratory. The tests covered a parameter range relevant to intact core reactor accidents ranging from large break to small break loss-ofcoolant conditions. Instrumentation within the 64-rod test section indicated that burnout occurred over an axial range within the bundle. The distance from the point where the first dry rod was detected to the point where all rods were dry was up to 60 cm in some of the tests. The burnout data should prove useful in developing new correlations for use in reactor thermalhydraulic codes. Evaluation of several existing critical heat flux correlations using the data show that three correlations, the Barnett, Bowring, and Katto correlations, perform similarly and correlate the data better than the Biasi correlation

  8. Status of rod consolidation

    International Nuclear Information System (INIS)

    Bailey, W.J.

    1985-04-01

    Two of the factors that need to be taken into account with rod consolidation are (1) the effects on rods from their removal from the fuel assembly and (2) the effects on rods as a result of the consolidation process. Potential components of both factors are described in the report. Discussed under (1) are scratches on the fuel rod surfaces, rod breakage, crud, extended burnup, and possible cladding embrittlement due to hydrogen injection at BWRs. Discussed under (2) are the increased water temperature (less than 10 0 C) because of closer packing of the rods, formation of crevices between rods in the close-packed mode, contact with dissimilar metals, and the potential for rapid heating of fuel rods following the loss of water from a spent fuel storage pool. Another factor that plays an important role in rod consolidation is the cost of disposal of the nonfuel-bearing components of the fuel assembly. Also, the dose rate from the components - especially Inconel spacer grids - can affect the handling procedures. Several licensing issues that exist are described. A list of recommendations is provided. 98 refs., 5 figs., 5 tabs

  9. Preliminary Single-Phase Mixing Test using Wire Mesh System in a wire-wrapped 37-rod Bundle

    International Nuclear Information System (INIS)

    Bae, Hwang; Kim, Hyungmo; Lee, Dong Won; Choi, Hae Seob; Choi, Sun Rock; Chang, Seokkyu; Kim, Seok; Euh, Dongjin; Lee, Hyeongyeon

    2014-01-01

    In this paper, preliminary tests of the wire-mesh sensor are introduced before measuring of mixing coefficient in the wire-wrapped 37-pin fuel assembly for a sodium-cooled fast reactor. Through this preliminary test, it was confirmed that city water can be used as a tracer for demineralized water as a base. A simple test was performed to evaluate the characteristics of a wire mesh with of a short pipe shape. The conductivity of de-mineralized water and city water is linearly increased for the limited temperature ranges as the temperature is increased. The reliability of the wire mesh sensor was estimated based on the averages and standard deviations of the plane image using the cross points. A wire mesh sensor is suitable to apply to a single-phase flow measurement for a mixture with de-mineralized water and city water. A wire mesh sensor and system have been traditionally used to measure the void fraction of a two-phase flow field with gas and liquid. Recently, Ylonen et al. successfully designed and commissioned a measurement system for a single-phase flow using a wire mesh sensor

  10. Deposition stress effects on thermal barrier coating burner rig life

    Science.gov (United States)

    Watson, J. W.; Levine, S. R.

    1984-01-01

    A study of the effect of plasma spray processing parameters on the life of a two layer thermal barrier coating was conducted. The ceramic layer was plasma sprayed at plasma arc currents of 900 and 600 amps onto uncooled tubes, cooled tubes, and solid bars of Waspalloy in a lathe with 1 or 8 passes of the plasma gun. These processing changes affected the residual stress state of the coating. When the specimens were tested in a Mach 0.3 cyclic burner rig at 1130 deg C, a wide range of coating lives resulted. Processing factors which reduced the residual stress state in the coating, such as reduced plasma temperature and increased heat dissipation, significantly increased coating life.

  11. Burner rig alkali salt corrosion of several high temperature alloys

    Science.gov (United States)

    Deadmore, D. L.; Lowell, C. E.

    1977-01-01

    The hot corrosion of five alloys was studied in cyclic tests in a Mach 0.3 burner rig into whose combustion chamber various aqueous salt solutions were injected. Three nickel-based alloys, a cobalt-base alloy, and an iron-base alloy were studied at temperatures of 700, 800, 900, and 1000 C with various salt concentrations and compositions. The relative resistance of the alloys to hot corrosion attack was found to vary with temperature and both concentration and composition of the injected salt solution. Results indicate that the corrosion of these alloys is a function of both the presence of salt condensed as a liquid on the surface and of the composition of the gas phases present.

  12. Physical models and codes for prediction of activity release from defective fuel rods under operation conditions and in leakage tests during refuelling

    International Nuclear Information System (INIS)

    Likhanskii, V.; Evdokimov, I.; Khoruzhii, O.; Sorokin, A.; Novikov, V.

    2003-01-01

    It is appropriate to use the dependences, based on physical models, in the design-analytical codes for improving of reliability of defective fuel rod detection and for determination of defect characteristics by activity measuring in the primary coolant. In the paper the results on development of some physical models and integral mechanistic codes, assigned for prediction of defective fuel rod behaviour are presented. The analysis of mass transfer and mass exchange between fuel rod and coolant showed that the rates of these processes depends on many factors, such as coolant turbulent flow, pressure, effective hydraulic diameter of defect, fuel rod geometric parameters. The models, which describe these dependences, have been created. The models of thermomechanical fuel behaviour, stable gaseous FP release were modified and new computer code RTOP-CA was created thereupon for description of defective fuel rod behaviour and activity release into the primary coolant. The model of fuel oxidation in in-pile conditions, which includes radiolysis and RTOP-LT after validation of physical models are planned to be used for prediction of defective fuel rods behaviour

  13. Control rod shutdown system

    International Nuclear Information System (INIS)

    Miyamoto, Yoshiyuki; Higashigawa, Yuichi.

    1996-01-01

    The present invention provides a control rod terminating system in a BWR type nuclear power plant, which stops an induction electric motor as rapidly as possible to terminate the control rods. Namely, the control rod stopping system controls reactor power by inserting/withdrawing control rods into a reactor by driving them by the induction electric motor. The system is provided with a control device for controlling the control rods and a control device for controlling the braking device. The control device outputs a braking operation signal for actuating the braking device during operation of the control rods to stop the operation of the control rods. Further, the braking device has at least two kinds of breaks, namely, a first and a second brakes. The two kinds of brakes are actuated by receiving the brake operation signals at different timings. The brake device is used also for keeping the control rods after the stopping. Even if a stopping torque of each of the breaks is small, different two kinds of brakes are operated at different timings thereby capable of obtaining a large stopping torque as a total. (I.S.)

  14. Control rod drives

    International Nuclear Information System (INIS)

    Futatsugi, Masao.

    1980-01-01

    Purpose: To secure the reactor operation safety by the provision of a fluid pressure detecting section for control rod driving fluid and a control rod interlock at the midway of the flow pass for supplying driving fluid to the control rod drives. Constitution: Between a driving line and a direction control valve are provided a pressure detecting portion, an alarm generating device, and a control rod inhibition interlock. The driving fluid from a driving fluid source is discharged by way of a pump and a manual valve into the reactor in which the control rods and reactor fuels are contained. In addition, when the direction control valve is switched and the control rods are inserted and extracted by the control rod drives, the pressure in the driving line is always detected by the pressure detection section, whereby if abnormal pressure is resulted, the alarm generating device is actuated to warn the abnormality and the control rod inhibition interlock is actuated to lock the direction control valve thereby secure the safety operation of the reactor. (Seki, T.)

  15. Why Rods and Cocci

    Indian Academy of Sciences (India)

    Bacteria exhibit a wide variety of shapes but the commonly studied species of bacteria are generally either spherical in shape which are called cocci (singular coccus) or have a cylindrical shape and are called rods or bacilli (singular bacillus). In reality rods and cocci are the ends of a continuum. Sonle of the cocci are.

  16. Control rod drives

    International Nuclear Information System (INIS)

    Oonuki, Koji.

    1981-01-01

    Purpose: To increase the driving speed of control rods at rapid insertion with an elongate control rod and an extension pipe while ensuring sufficient buffering performance in a short buffering distance, by providing a plurality of buffers to an extension pipe between a control rod drive source and a control rod in LMFBR type reactor. Constitution: First, second and third buffers are respectively provided to an acceleration piston, an extension pipe and a control rod respectively and the insertion positions for each of the buffers are displaced orderly from above to below. Upon disconnection of energizing current for an electromagnet, the acceleration piston, the extension pipe and the control rod are rapidly inserted in one body. The first, second and third buffers are respectively actuated at each of their falling strokes upon rapid insertion respectively, and the acceleration piston, the extension pipe and the control rod receive the deceleration effect in the order correspondingly. Although the compression force is applied to the control rod only near the stroke end, it does not cause deformation. (Kawakami, Y.)

  17. Water-level representation by men and women as a function of rod-and-frame test proficiency and visual and postural information.

    Science.gov (United States)

    Robert, M; Ohlmann, T

    1994-01-01

    In the water-level task, it has been repeatedly shown that, compared with men, women more often fail to represent the surface of a liquid as horizontal regardless of the tilt of the container. An attempt was made to reduce this robust gender gap through the manipulation of relevant upright references conveyed both by the position of the stimuli and the posture of the subject. It was reasoned that bringing the women to focus on such gravitational references through postural adjustment might help their performance equal that of men, thus shedding some light on the nature of the difficulty they experience in the standard setting. A lesser effect was anticipated among men. However, the results showed that, even after controlling for proficiency in the correlated visuospatial situation of the rod-and-frame test, the performance of men always surpassed that of women. Irrespective of gender, water-level representation on vertical sheets was unaffected by the subject's posture, whereas it improved when horizontal sheets were coupled with the most unstable posture. Whereas the persistence of the yet-unaccounted-for gender difference was underscored, the contributions of visual and postural cues issued at arm and full-body levels were discussed.

  18. Gulf of Mexico rig activity up, international lags

    International Nuclear Information System (INIS)

    Rappold, K.

    1994-01-01

    Demand for jack up and semisubmersible rigs has improved in the Gulf of Mexico following a decline in activity earlier this year. International drilling activity, however, has shown slight declines in several regions. Relatively firm natural gas prices have helped buoy rig activity in North America. Rig day rates have not followed suit, mainly because of the influx of rigs from weaker international markets. Day rates in the US may not increase until international activity picks up and the world-wide drilling market tightens. Oil prices have hit almost $20/bbl, mainly because of the recent oil worker' strike in Nigeria and good demand. Natural gas prices in the US have hovered around $2.00/MMBTU, and many industry analysts expect gas prices to remain strong over the next few years. This paper gives data on drilling rig counts and crude oil and gas prices in the Gulf of Mexico and onshore

  19. Failed fuel rod detector

    Energy Technology Data Exchange (ETDEWEB)

    Uchida, Katsuya; Matsuda, Yasuhiko

    1984-05-02

    The purpose of the project is to enable failed fuel rod detection simply with no requirement for dismantling the fuel assembly. A gamma-ray detection section is arranged so as to attend on the optional fuel rods in the fuel assembly. The fuel assembly is adapted such that a gamma-ray shielding plate is detachably inserted into optional gaps of the fuel rods or, alternatively, the fuel assembly can detachably be inserted to the gamma-ray shielding plate. In this way, amount of gaseous fission products accumulated in all of the plenum portions in the fuel rods as the object of the measurement can be determined without dismantling the fuel assembly. Accordingly, by comparing the amounts of the gaseous fission products, the failed fuel rod can be detected.

  20. Ergonomic exposure on a drilling rig

    DEFF Research Database (Denmark)

    Jensen, Carsten; Jensen, Chris

    . In a relatively old study on American drilling rigs it was indicated that lower back problems was a frequent cause of absence (Clemmer et al. 1991). Most of the incidents causing lower back injuries were associated with heavy lifting or pushing/pulling objects by roustabouts, floorhands, derrickmen and welders......, but for some of the most frequent problems, such as musculoskeletal problems, it is difficult to determine whether the causes are work‐related or not. As manual handling (lifting, pushing, etc.) in awkward body postures increase the risk of developing musculoskeletal disorders, it should be expected that work......‐related health problems contribute to sickness absence in the offshore industry, if these working postures are common. However, also work‐related psychosocial factors, personal factors and other factors may contribute to the development of lower back disorders, which often have a multifactorial background...

  1. Investigating Knowledge Transfer Mechanisms for Oil Rigs

    DEFF Research Database (Denmark)

    Vianello, Giovanna; Ahmed, Saeema

    2009-01-01

    It is widely recognized, both in industry and academia, that clear strategies in knowledge transfer positively influence the success of a firm. A firm should support the transfer of knowledge by standardizing communication channels within and across departments, based upon personalization......, codification or a combination of these two strategies. The characteristics of the business influence the choice of communication channels used for knowledge transfer. This paper presents a case study exploring the transfer of knowledge within and across projects, specifically the transfer of service knowledge...... in the case of complex machinery. The strategies used for knowledge transfer were analysed and compared with the expected transfer mechanisms, similarities and differences were investigated and are described. A family of four identical rigs for offshore drilling was the selected case. The transfer...

  2. Endothelial RIG-I activation impairs endothelial function

    Energy Technology Data Exchange (ETDEWEB)

    Asdonk, Tobias, E-mail: tobias.asdonk@ukb.uni-bonn.de [Department of Medicine/Cardiology, University of Bonn, Sigmund-Freud-Str. 25, 53105 Bonn (Germany); Motz, Inga; Werner, Nikos [Department of Medicine/Cardiology, University of Bonn, Sigmund-Freud-Str. 25, 53105 Bonn (Germany); Coch, Christoph; Barchet, Winfried; Hartmann, Gunther [Institute for Clinical Chemistry and Clinical Pharmacology, University of Bonn, Sigmund-Freud-Str. 25, 53105 Bonn (Germany); Nickenig, Georg; Zimmer, Sebastian [Department of Medicine/Cardiology, University of Bonn, Sigmund-Freud-Str. 25, 53105 Bonn (Germany)

    2012-03-30

    Highlights: Black-Right-Pointing-Pointer RIG-I activation impairs endothelial function in vivo. Black-Right-Pointing-Pointer RIG-I activation alters HCAEC biology in vitro. Black-Right-Pointing-Pointer EPC function is affected by RIG-I stimulation in vitro. -- Abstract: Background: Endothelial dysfunction is a crucial part of the chronic inflammatory atherosclerotic process and is mediated by innate and acquired immune mechanisms. Recent studies suggest that pattern recognition receptors (PRR) specialized in immunorecognition of nucleic acids may play an important role in endothelial biology in a proatherogenic manner. Here, we analyzed the impact of endothelial retinoic acid inducible gene I (RIG-I) activation upon vascular endothelial biology. Methods and results: Wild type mice were injected intravenously with 32.5 {mu}g of the RIG-ligand 3pRNA (RNA with triphosphate at the 5 Prime end) or polyA control every other day for 7 days. In 3pRNA-treated mice, endothelium-depended vasodilation was significantly impaired, vascular oxidative stress significantly increased and circulating endothelial microparticle (EMP) numbers significantly elevated compared to controls. To gain further insight in RIG-I dependent endothelial biology, cultured human coronary endothelial cells (HCAEC) and endothelial progenitor cells (EPC) were stimulated in vitro with 3pRNA. Both cells types express RIG-I and react with receptor upregulation upon stimulation. Reactive oxygen species (ROS) formation is enhanced in both cell types, whereas apoptosis and proliferation is not significantly affected in HCAEC. Importantly, HCAEC release significant amounts of proinflammatory cytokines in response to RIG-I stimulation. Conclusion: This study shows that activation of the cytoplasmatic nucleic acid receptor RIG-I leads to endothelial dysfunction. RIG-I induced endothelial damage could therefore be an important pathway in atherogenesis.

  3. Endothelial RIG-I activation impairs endothelial function

    International Nuclear Information System (INIS)

    Asdonk, Tobias; Motz, Inga; Werner, Nikos; Coch, Christoph; Barchet, Winfried; Hartmann, Gunther; Nickenig, Georg; Zimmer, Sebastian

    2012-01-01

    Highlights: ► RIG-I activation impairs endothelial function in vivo. ► RIG-I activation alters HCAEC biology in vitro. ► EPC function is affected by RIG-I stimulation in vitro. -- Abstract: Background: Endothelial dysfunction is a crucial part of the chronic inflammatory atherosclerotic process and is mediated by innate and acquired immune mechanisms. Recent studies suggest that pattern recognition receptors (PRR) specialized in immunorecognition of nucleic acids may play an important role in endothelial biology in a proatherogenic manner. Here, we analyzed the impact of endothelial retinoic acid inducible gene I (RIG-I) activation upon vascular endothelial biology. Methods and results: Wild type mice were injected intravenously with 32.5 μg of the RIG-ligand 3pRNA (RNA with triphosphate at the 5′end) or polyA control every other day for 7 days. In 3pRNA-treated mice, endothelium-depended vasodilation was significantly impaired, vascular oxidative stress significantly increased and circulating endothelial microparticle (EMP) numbers significantly elevated compared to controls. To gain further insight in RIG-I dependent endothelial biology, cultured human coronary endothelial cells (HCAEC) and endothelial progenitor cells (EPC) were stimulated in vitro with 3pRNA. Both cells types express RIG-I and react with receptor upregulation upon stimulation. Reactive oxygen species (ROS) formation is enhanced in both cell types, whereas apoptosis and proliferation is not significantly affected in HCAEC. Importantly, HCAEC release significant amounts of proinflammatory cytokines in response to RIG-I stimulation. Conclusion: This study shows that activation of the cytoplasmatic nucleic acid receptor RIG-I leads to endothelial dysfunction. RIG-I induced endothelial damage could therefore be an important pathway in atherogenesis.

  4. A Built for Purpose Micro-Hole Coiled Tubing Rig (MCTR)

    Energy Technology Data Exchange (ETDEWEB)

    Bart Patton

    2007-09-30

    This report will serve as the final report on the work performed from the contract period October 2005 thru April 2007. The project 'A Built for Purpose Microhole Coiled Tubing Rig (MCTR)' purpose was to upgrade an existing state-of-the-art Coiled Tubing Drilling Rig to a Microhole Coiled Tubing Rig (MCTR) capable of meeting the specifications and tasks of the Department of Energy. The individual tasks outlined to meet the Department of Energy's specifications are: (1) Concept and development of lubricator and tool deployment system; (2) Concept and development of process control and data acquisition; (3) Concept and development of safety and efficiency improvements; and (4) Final unit integration and testing. The end result of the MCTR upgrade has produced a unit capable of meeting the following requirements: (1) Capable of handling 1-inch through 2-3/8-inch coiled tubing (Currently dressed for 2-3/8-inch coiled tubing and capable of running up to 3-1/2-inch coiled tubing); (2) Capable of drilling and casing surface, intermediate, production and liner hole intervals; (3) Capable of drilling with coiled tubing and has all controls and installation piping for a top drive; (4) Rig is capable of running 7-5/8-inch range 2 casing; and (5) Capable of drilling 5,000 ft true vertical depth (TVD) and 6,000 ft true measured depth (TMD).

  5. Modelling and simulation of dynamic wheel-rail interaction using a roller rig

    International Nuclear Information System (INIS)

    Anyakwo, A; Pislaru, C; Ball, A; Gu, F

    2012-01-01

    The interaction between the wheel and rail greatly influences the dynamic response of railway vehicles on the track. A roller rig facility can be used to study and monitor real time parameters that influence wheel-rail interaction such as wear, adhesion, friction and corrugation without actual field tests being carried out. This paper presents the development of the mathematical models for full scale roller rig and 1/5 scale roller rig and the wear prediction model based on KTH wear function. The simulated critical speed for the 1/5 scale roller rig is about one-fifth of the critical speed for the full scale model so the simulated results compare well with the theory related to wheel-rail contact and dynamics. Also the differences between the simulated rolling radii for the full scale model with and without wear function are analysed. This paper presents the initial stage of a large scale research project where the influence of wear on the wheel-rail performance will be studied in more depth.

  6. Results of water chemistry control in the in-pile ''Callisto'' loop (an experimental PWR rig installed in the BR2 reactor)

    International Nuclear Information System (INIS)

    Weber, M.; Benoit, P.; Dekeyser, J.; Verwimp, A.

    1994-01-01

    Since June 1992, a new experimental facility, called CALLISTO, is being irradiated in the BR2 materials testing reactor at Mol, Belgium. The main objective of the present test campaign is to study the behaviour of advanced fuel to high burn-up rates in a realistic PWR environment. Three in-pile sections, containing each 9 fuel rods, are loaded inside the reactor vessel and are connected to a common out-of-pile pressurized water circulation loop (ref.1). The later is branched-off into a purification circuit (feed-bleed concept) and further equipped with safety and auxiliary systems. To cope with the test programme, the equipments are designed so that the guidelines of a PWR primary water chemistry can be followed (ref.2). Real steady-state conditions cannot be observed because the typical BR2 cycle (3 weeks running/3 weeks shut-down) is much shorter and because the rig is cooled down during each reactor shut-down. The purpose of this poster is to provide results of chemical parameters recorded during the cycling behaviour of the CALLISTO primary water. (authors). 4 figs., 1 tab., 2 refs

  7. Development of a fuel-rod simulator and small-diameter thermocouples for high-temperature, high-heat-flux tests in the Gas-Cooled Fast Reactor Core Flow Test Loop

    International Nuclear Information System (INIS)

    McCulloch, R.W.; MacPherson, R.E.

    1983-03-01

    The Core Flow Test Loop was constructed to perform many of the safety, core design, and mechanical interaction tests in support of the Gas-Cooled Fast Reactor (GCFR) using electrically heated fuel rod simulators (FRSs). Operation includes many off-normal or postulated accident sequences including transient, high-power, and high-temperature operation. The FRS was developed to survive: (1) hundreds of hours of operation at 200 W/cm 2 , 1000 0 C cladding temperature, and (2) 40 h at 40 W/cm 2 , 1200 0 C cladding temperature. Six 0.5-mm type K sheathed thermocouples were placed inside the FRS cladding to measure steady-state and transient temperatures through clad melting at 1370 0 C

  8. Development of a fuel-rod simulator and small-diameter thermocouples for high-temperature, high-heat-flux tests in the Gas-Cooled Fast Reactor Core Flow Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    McCulloch, R.W.; MacPherson, R.E.

    1983-03-01

    The Core Flow Test Loop was constructed to perform many of the safety, core design, and mechanical interaction tests in support of the Gas-Cooled Fast Reactor (GCFR) using electrically heated fuel rod simulators (FRSs). Operation includes many off-normal or postulated accident sequences including transient, high-power, and high-temperature operation. The FRS was developed to survive: (1) hundreds of hours of operation at 200 W/cm/sup 2/, 1000/sup 0/C cladding temperature, and (2) 40 h at 40 W/cm/sup 2/, 1200/sup 0/C cladding temperature. Six 0.5-mm type K sheathed thermocouples were placed inside the FRS cladding to measure steady-state and transient temperatures through clad melting at 1370/sup 0/C.

  9. Fuel rod behaviour during transients

    International Nuclear Information System (INIS)

    Bilsby, C.F.; Haste, T.J.; Garlick, A.; Cameron, R.F.

    1982-04-01

    The clad deformation code CANSWELL-2 is described. This is used, either as a stand-alone code or within MABEL-2, to predict and analyse the results of LOCA simulations in the Halden and NRU reactors and in the KfK and PROPAT rigs. Experimental evidence on fuel behaviour in RIA, PCM and ATWS events is presented with inclusion of certain FRAP-T5 results. Published calculations from the accident codes FRAP-T4 and FRAP-T5 are compared with experimental results in simulated loss of coolant tests in the Power Burst Facility. The limitations of this code in its treatment of RIA, PCM and ATWS events are considered. (U.K.)

  10. Control rod drives

    International Nuclear Information System (INIS)

    Hayakawa, Hiroyasu.

    1979-01-01

    Purpose: To enable rapid control in a simple circuit by providing a motor control device having an electric capacity capable of simultaneously driving all of the control rods rapidly only in the inserting direction as well as a motor controlling device capable of fine control for the insertion and extraction at usual operation. Constitution: The control rod drives comprise a first motor control device capable of finely controlling the control rods both in inserting and extracting directions, a second motor control device capable of rapidly driving the control rods only in the inserting direction, and a first motor switching circuit and a second motor switching circuit switched by switches. Upon issue of a rapid insertion instruction for the control rods, the second motor switching circuit is closed by the switch and the second motor control circuit and driving motors are connected. Thus, each of the control rod driving motors is driven at a high speed in the inserting direction to rapidly insert all of the control rods. (Yoshino, Y.)

  11. Multiple fuel rod gripper

    International Nuclear Information System (INIS)

    Shields, E.P.

    1987-01-01

    An apparatus is described for gripping an array of rods comprising: (a) gripping members grippingly engageable with the rods, each of which has a hollow portion terminating in an open end for receiving the end of one of the rods; (b) a closing means for causing the hollow portion of each of the gripping members to apply substantially the same gripping force onto the end of its respective rod, including (i) a locking plate having a plurality of tapered holes registrable with the array of rods, wherein the exterior of each of the gripping members is tapered and nested within one of the tapered holes, (ii) a withdrawing means having a hydraulic plunger operatively connected to each of the gripping members for applying a substantially identical withdrawing force on each of the gripping members, whereby the hollow portion of each of the gripping members applies substantially the same gripping force on its respective rod, and (c) means for detecting whether each of the gripping members has grippingly engaged its respective rod

  12. Fuel rod technology

    International Nuclear Information System (INIS)

    Bezold, H.; Romeiser, H.J.

    1979-07-01

    By extensive mechanization and automation of the fuel rod production, also at increasing production numbers, an efficient production shall be secured, simultaneously corresponding to the high quality standard of the fuel rods. The works done up to now concentrated on the lay out of a rough concept for a mechanized production course. Detail-studies were made for the problems of fuel rod humidity, filling and resistance welding. Further promotion of this project and thus further report will be stopped, since the main point of these works is the production technique. (orig.) [de

  13. Estimation of irradiated control rod worth

    International Nuclear Information System (INIS)

    Varvayanni, M.; Catsaros, N.; Antonopoulos-Domis, M.

    2009-01-01

    When depleted control rods are planned to be used in new core configurations, their worth has to be accurately predicted in order to deduce key design and safety parameters such as the available shutdown margin. In this work a methodology is suggested for the derivation of the distributed absorbing capacity of a depleted rod, useful in the case that the level of detail that is known about the irradiation history of the control rod does not allow an accurate calculation of the absorber's burnup. The suggested methodology is based on measurements of the rod's worth carried out in the former core configuration and on corresponding calculations based on the original (before first irradiation) absorber concentration. The methodology is formulated for the general case of the multi-group theory; it is successfully tested for the one-group approximation, for a depleted control rod of the Greek Research Reactor, containing five neutron absorbers. The computations reproduce satisfactorily the irradiated rod worth measurements, practically eliminating the discrepancy of the total rod worth, compared to the computations based on the nominal absorber densities.

  14. Control Rod Malfunction at the NRAD Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Thomas L. Maddock

    2010-05-01

    The neutron Radiography Reactor (NRAD) is a training, research, and isotope (TRIGA) reactor located at the INL. The reactor is normally shut down by the insertion of three control rods that drop into the core when power is removed from electromagnets. During a routine shutdown, indicator lights on the console showed that one of the control rods was not inserted. It was initially thought that the indicator lights were in error because of a limit switch that was out of adjustment. Through further testing, it was determined that the control rod did not drop when the scram switch was initially pressed. The control rod anomaly led to a six month shutdown of the reactor and an in depth investigation of the reactor protective system. The investigation looked into: scram switch operation, console modifications, and control rod drive mechanisms. A number of latent issues were discovered and corrected during the investigation. The cause of the control rod malfunction was found to be a buildup of corrosion in the control rod drive mechanism. The investigation resulted in modifications to equipment, changes to both operation and maintenance procedures, and additional training. No reoccurrences of the problem have been observed since corrective actions were implemented.

  15. French LMFBR's control rods experience and development

    International Nuclear Information System (INIS)

    Arnaud, G.; Guigon, A.; Verset, L.

    1983-06-01

    Since the last ten years, the French program has been, first of all, directed to the setting up, and then the development of, at once, the Phenix control rods, and next, the Super-Phenix ones. The vented pin design, with porous plug and sodium bonding, which allows the choices of large diameters, has been taken, since the Rapsodie experience was decisive. The absorber material is sintered, 10 B enriched, boron carbide. The can is made of 316 type stainless steel, stabilised, or not, with titanium. The experience gained in Phenix up to now is important, and deals with about six loads of control rods. Results confirm the validity of the design of the absorber pins. Some difficulties has been encountered for the guiding devices, due to the swelling of the steel. They have required design and material improvements. Such difficulties are discarded by a new design of the bearing, for the Super-Phenix control rods. The other parts of these rods, from the Primary Shut-Down System, are strictly derived from Phenix. The design of the rods from the Secondary Shut-Down System is rather different, but it's not the case for the design of the absorber pins: in many a way, they are derived from Phenix pins and from Rapsodie control rods. Both types of rods irradiation tests are in progress in Phenix [fr

  16. Flow distribution and pressure loss in subchannels of a wire-wrapped 37-pin rod bundle for sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Seok Kyu; Euh, Dong Jin; Choi, Hae Seob; Kim, Hyung Mo; Choi, Sun Rock; Lee, Hyeong Yeon [Thermal-Hydraulic Safety Research Department, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-04-15

    A hexagonally arrayed 37-pin wire-wrapped rod bundle has been chosen to provide the experimental data of the pressure loss and flow rate in subchannels for validating subchannel analysis codes for the sodium-cooled fast reactor core thermal/hydraulic design. The iso-kinetic sampling method has been adopted to measure the flow rate at subchannels, and newly designed sampling probes which preserve the flow area of subchannels have been devised. Experimental tests have been performed at 20-115% of the nominal flow rate and 60 degrees C (equivalent to Re ∼ 37,100) at the inlet of the test rig. The pressure loss data in three measured subchannels were almost identical regardless of the subchannel locations. The flow rate at each type of subchannel was identified and the flow split factors were evaluated from the measured data. The predicted correlations and the computational fluid dynamics results agreed reasonably with the experimental data.

  17. Control rod ejection analysis during a depressurization accident and the development of a rod-ejection-preventing device

    International Nuclear Information System (INIS)

    Mitake, S.; Itoh, K.; Fukushima, H.; Inoue, T.

    1982-01-01

    The control rods used for the experimental VHTR are suspended in the core by means of flexible steel cables and it is conceivable that an accidental rod ejection could occur due to a depressurization accident. The computer code AFLADE was developed in order to analyze the possibility of accidental rod ejection, and several studies were performed. The parametric study results showed that the adopted design condition for the VHTR core will not cause a rod ejection accident. In parallel with these accident analyses, a rod-ejection-preventing device was developed in preparation for a hypothetical accident, and its function was verified by the component tests

  18. Functional characterizations of RIG-I to GCRV and viral/bacterial PAMPs in grass carp Ctenopharyngodon idella.

    Directory of Open Access Journals (Sweden)

    Lijun Chen

    Full Text Available BACKGROUND: RIG-I (retinoic acid inducible gene-I is one of the key cytosolic pattern recognition receptors (PRRs for detecting nucleotide pathogen associated molecular patterns (PAMPs and mediating the induction of type I interferon and inflammatory cytokines in innate immune response. Though the mechanism is well characterized in mammals, the study of the accurate function of RIG-I in teleosts is still in its infancy. METHODOLOGY/PRINCIPAL FINDINGS: To clarify the functional characterizations of RIG-I in grass carp Ctenopharyngodon idella (CiRIG-I, six representative overexpression plasmids were constructed and transfected into C. idella kidney (CIK cell lines to obtain stably expressing recombinant proteins, respectively. A virus titer test and 96-well plate staining assay showed that all constructs exhibited the antiviral activity somewhat. The quantitative real-time RT-PCR (qRT-PCR demonstrated that mRNA expressions of CiIPS-1, CiIFN-I and CiMx2 were regulated by not only virus (GCRV or viral PAMP (poly(IC challenge but also bacterial PAMPs (LPS and PGN stimulation in the steadily transfected cells. The results showed that the full-length CiRIG-I played a key role in RLR pathway. The repressor domain (RD exerted an inhibitory function of the signaling channel under all utilized challenges. Caspase activation and recruitment domains (CARDs showed a positive role in GCRV and poly(I:C challenge. Helicase motifs were crucial for the signaling pathway upon LPS and PGN stimulation. Interestingly, ΔCARDs (CARDs deleted showed positive modulation in RIG-I signal transduction. CONCLUSIONS/SIGNIFICANCE: The results provided some novel insights into RIG-I sensing with a strikingly broad regulation in teleosts, responding not only to the dsRNA virus or synthetic dsRNA but also bacterial PAMPs.

  19. Prototypical Rod Consolidation Demonstration Project

    International Nuclear Information System (INIS)

    1993-05-01

    The objective of Phase 3 of the Prototypical Rod consolidation Demonstration Project (PRCDP) was to procure, fabricate, assemble, and test the Prototypical Rod consolidation System as described in the NUS Phase 2 Final Design Report. This effort required providing the materials, components, and fabricated parts which makes up all of the system equipment. In addition, it included the assembly, installation, and setup of this equipment at the Cold Test Facility. During the Phase 3 effort the system was tested on a component, subsystem, and system level. This volume 1, discusses the PRCDP Phase 3 Test Program that was conducted by the HALLIBURTON NUS Environmental Corporation under contract AC07-86ID12651 with the United States Department of Energy. This document, Volume 1, Book 5 discusses the following topics: Lower Cutting System Test Results and Analysis Report; NFBC Loading System Test Results and Analysis Report; Robotic Bridge Transporter Test Results and Analysis Report; RM-10A Remotec Manipulator Test Results and Analysis Report; and Manipulator Transporter Test Results and Analysis Report

  20. Burnable poison rod

    International Nuclear Information System (INIS)

    Natsume, Tomohiro.

    1988-01-01

    Purpose: To decrease the effect of water elimination and the effect of burn-up residue boron, thereby reduce the effect of burnable poison rods as the neutron poisons at the final stage of reactor core lifetime. Constitution: In a burnable poison rod according to the present invention, a hollow burnable poison material is filled in an external fuel can, an inner fuel can mounted with a carbon rod is inserted to the hollow portion of the burnable poison material and helium gases are charged in the outer fuel can. In such a burnable poison rod, the reactivity worths after the burning are reduced to one-half as compared with the conventional case. Accordingly, since the effect of the burnable poison as the neutron poisons is reduced at the final stage of the reactor core of lifetime, the excess reactivity of the reactor core is increased. (Horiuchi, T.)

  1. Control rod driving mechanism

    International Nuclear Information System (INIS)

    Ooshima, Yoshio.

    1983-01-01

    Purpose: To perform reliable scram operation, even if abnormality should occur in a system instructing scram operation in FBR type reactors. Constitution: An aluminum alloy member to be melt at a predetermined temperature (about 600sup(o)C) is disposed to a connection part between a control rod and a driving mechanism, whereby the control rod is detached from the driving mechanism and gravitationally fallen to the reactor core. (Ikeda, J.)

  2. Control rod drives

    International Nuclear Information System (INIS)

    Hayakawa, Hiroyasu; Kawamura, Atsuo.

    1979-01-01

    Purpose: To reduce pellet-clad mechanical interactions, as well as improve the fuel safety. Constitution: In the rod drive of a bwr type reactor, an electric motor operated upon intermittent input such as of pulse signals is connected to a control rod. A resolver for converting the rotational angle of the motor to electric signals is connected to the rotational shaft of the motor and the phase difference between the output signal from the resolver and a reference signal is adapted to detect by a comparator. Based on the detection result, the controller is actuated to control a motor for control rod drive so that fine control for the movement of the control rod is made possible. This can reduce the moving distance of the control rod, decrease the thermal stress applied to the control rod and decrease the pellet clad mechanical interaction failures due to thermal expansion between the cladding tube and the pellets caused by abrupt changes in the generated power. (Furukawa, Y.)

  3. Control rod drive

    International Nuclear Information System (INIS)

    Okutani, Tetsuro.

    1988-01-01

    Purpose: To provide a simple and economical control rod drive using a control circuit requiring no pulse circuit. Constitution: Control rods in a BWR type reactor are driven by hydraulic pressure and inserted or withdrawn in the direction of applying the hydraulic pressure. The direction of the hydraulic pressure is controlled by a direction control valve. Since the driving for the control rod is extremely important in view of the operation, a self diagnosis function is disposed for rapid inspection of possible abnormality. In the present invention, two driving contacts are disposed each by one between the both ends of a solenoid valve of the direction control valve for driving the control rod and the driving power source, and diagnosis is conducted by alternately operating them. Therefore, since it is only necessary that the control circuit issues a driving instruction only to one of the two driving contacts, the pulse circuit is no more required. Further, since the control rod driving is conducted upon alignment of the two driving instructions, the reliability of the control rod drive can be improved. (Horiuchi, T.)

  4. Method of inserting fuel rod

    International Nuclear Information System (INIS)

    Kamimoto, Shuji; Imoo, Makoto; Tsuchida, Kenji.

    1991-01-01

    The present invention concerns a method of inserting a fuel rod upon automatic assembling, automatic dismantling and reassembling of a fuel assembly in a light water moderated reactor, as well as a device and components used therefor. That is, a fuel rod is inserted reliably to an aimed point of insertion by surrounding the periphery of the fuel rod to be inserted with guide rods, and thereby suppressing the movement of the fuel rod during insertion. Alternatively, a fuel rod is inserted reliably to a point of insertion by inserting guide rods at the periphery of the point of insertion for the fuel rod to be inserted thereby surrounding the point of insertion with the guide rods or fuel rods. By utilizing fuel rods already present in the fuel assembly as the guide rods described above, the fuel rod can be inserted reliably to the point of insertion with no additional devices. Dummy fuel rods are previously inserted in a fuel assembly which are then utilized as the above-mentioned guide rods to accurately insert the fuel rod to the point of insertion. (I.S.)

  5. Technical measurement of small fission gas inventory in fuel rod with laser puncturing system

    International Nuclear Information System (INIS)

    Kim, Hee Moon; Kim, Sung Ryul; Lee, Byoung Oon; Yang, Yong Sik; Baek, Sang Ryul; Song, Ung Sup

    2012-01-01

    The fission gas release cause degradation of fuel rod. It influences fuel temperature and internal pressure due to low thermal conductivity. Therefore, fission gas released to internal void of fuel rod must be measured with burnup. To measure amount of fission gas, fuel rod must be punctured by a steel needle in a closed chamber. Ideal gas law(PV=nRT) is applied to obtain atomic concentration(mole). Steel needle type is good for large amount of fission gas such as commercial spent fuel rod. But, some cases with small fuel rig in research reactor for R/D program are not available to use needle type because of large chamber volume. The laser puncturing technique was developed to solve measurement of small amount of fission gas. This system was very rare equipment in other countries. Fine pressure gage and strong vacuum system were installed, and the chamber volume was reduced at least. Fiber laser was used for easy operation

  6. Control of two-phase erosion corrosion with the amine 5-aminopentanol: rig and plant trials

    International Nuclear Information System (INIS)

    Lewis, G.G.; Greene, J.C.; Tyldesley, J.D.; Wetton, E.A.M.; Fountain, M.J.

    1994-01-01

    Control of two-phase erosion corrosion in the once through mild steel boilers of the gas cooled nuclear power station at Wylfa was achieved by using the amine 2-amino, 2 methylpropan-1-ol (AMP). In a search to find a more cost effective amine, 5-aminopentanol (5-AP) emerged, from a laboratory based programme to determine basicity and volatility, as the most promising candidate. The effectiveness of 5-AP in controlling erosion corrosion was demonstrated in a rig test, carried out on a full scale replica of a Wylfa boiler tube. Following on from the rig test, a plant trial at Wylfa PS demonstrated 5-AP's superior thermal stability (compared to AMP). It also provided confirmation that the laboratory generated data on basicity and volatility was applicable to plant and hence also the accuracy of the figures for predicted amine usage. (orig.)

  7. Analysis and modification of a single-mesh gear fatigue rig for use in diagnostic studies

    Science.gov (United States)

    Zakrajsek, James J.; Townsend, Dennis P.; Oswald, Fred B.; Decker, Harry J.

    1992-01-01

    A single-mesh gear fatigue rig was analyzed and modified for use in gear mesh diagnostic research. The fatigue rig allowed unwanted vibration to mask the test-gear vibration signal, making it difficult to perform diagnostic studies. Several possible sources and factors contributing to the unwanted components of the vibration signal were investigated. Sensor mounting location was found to have a major effect on the content of the vibration signal. In the presence of unwanted vibration sources, modal amplification made unwanted components strong. A sensor location was found that provided a flatter frequency response. This resulted in a more useful vibration signal. A major network was performed on the fatigue rig to reduce the influence of the most probable sources of the noise in the vibration signal. The slave gears were machined to reduce weight and increase tooth loading. The housing and the shafts were modified to reduce imbalance, looseness, and misalignment in the rotating components. These changes resulted in an improved vibration signal, with the test-gear mesh frequency now the dominant component in the signal. Also, with the unwanted sources eliminated, the sensor mounting location giving the most robust representation of the test-gear meshing energy was found to be at a point close to the test gears in the load zone of the bearings.

  8. Detection device for control rod scram

    International Nuclear Information System (INIS)

    Ishiyama, Satoshi.

    1989-01-01

    The device of the present invention comprises a control rod dropping separately from a control rod driving mechanism main body, a following tube falling separately accompanying therewith and a guide tube for guiding the dropping of the control rod and the following tube. Further, rare earth permanent magnets are embedded with the pole being axially oriented in the following tube and bobbins each mounted with an inner flange made of high magnetic permeability material are disposed to the guide tube. Coils are wound in the bobbin. In this control rod scram detection device, since magnetic fluxes can effectively be supplied to the coils, it is possible to obtain stable and highly reliable scram detection signals. Further, since the coils and the bobbins can be manufactured separately from the guide tube, their assemblies can be tested independently from the guide tube. (K.M.)

  9. Endoscopic PIV measurements in a low pressure turbine rig

    Energy Technology Data Exchange (ETDEWEB)

    Kegalj, Martin; Schiffer, Heinz-Peter [Technische Universitaet Darmstadt (Germany). Department of Gas Turbines and Aerospace Propulsion

    2009-10-15

    Particle-Image-Velocimetry (PIV) is a useful way to acquire information about the flow in turbomachinery. Several premises have to be fulfilled to achieve high-quality data, for example, optical access, low vibrations and low reflections. However, not all test facilities comply with these requirements. If there is no optical access to the test area, measurements cannot be performed. The use of borescopic optics is a possible solution to this issue, as the access required is very small. Several different techniques can be used to measure the three components of the velocity vector, one of which is Stereo-PIV. These techniques require either large optical access from several viewing angles or highly complex setups. Orthogonal light sheet orientations in combination with borescopic optics using Planar-PIV can deliver sufficient information about the flow. This study will show the feasibility of such an approach in an enclosed test area, such as the interblade space in a Low-Pressure-Turbine-Rig. The results from PIV will be compared with data collected with conventional techniques, such as the Five-Hole-Probe and the 2-component Hot-Wire-Anemometry. An analysis of time- and phase-averaged data will be performed. (orig.)

  10. Development of the automatic control rod operation system for JOYO. Verification of automatic control rod operation guide system

    International Nuclear Information System (INIS)

    Terakado, Tsuguo; Suzuki, Shinya; Kawai, Masashi; Aoki, Hiroshi; Ohkubo, Toshiyuki

    1999-10-01

    The automatic control rod operation system was developed to control the JOYO reactor power automatically in all operation modes(critical approach, cooling system heat up, power ascent, power descent), development began in 1989. Prior to applying the system, verification tests of the automatic control rod operation guide system was conducted during 32nd duty cycles of JOYO' from Dec. 1997 to Feb. 1998. The automatic control rod operation guide system consists of the control rod operation guide function and the plant operation guide function. The control rod operation guide function provides information on control rod movement and position, while the plant operation guide function provide guidance for plant operations corresponding to reactor power changes(power ascent or power descent). Control rod insertion or withdrawing are predicted by fuzzy algorithms. (J.P.N.)

  11. CONCENTRIC TUBE-FOULING RIG FOR INVESTIGATION OF FOULING DEPOSIT FORMATION FROM PASTEURISER OF VISCOUS FOOD LIQUID

    Directory of Open Access Journals (Sweden)

    N. I. KHALID

    2013-02-01

    Full Text Available This paper reports the work on developing concentric tube-fouling rig, a new fouling deposit monitoring device. This device can detect and quantify the level of fouling deposit formation. It can also functioning as sampler for fouling deposit study, which can be attached at any food processing equipment. The design is initiated with conceptual design. The rig is designed with inner diameter of 7 cm and with tube length of 37 cm. A spiral insert with 34.5 cm length and with 5.4 cm diameter is fitted inside the tube to ensure the fluid flows around the tube. In this work, the rig is attached to the lab-scale concentric tube-pasteurizer to test its effectiveness and to collect a fouling sample after pasteurization of pink guava puree. Temperature changes are recorded during the pasteurization and the data is used to plot the heat transfer profile. Thickness of the fouling deposit is also measured. The trends for thickness, heat resistance profile and heat transfer profile for concentric tube-fouling rig matched the trends obtained from lab-scale concentric tube-pasteurizer very well. The findings from this work have shown a good potential of this rig however there is a limitation with spiral insert, which is discussed in this paper.

  12. Statistikeren skiftede spor som 49-årig

    DEFF Research Database (Denmark)

    Sølund, Sune; Rootzén, Helle

    2010-01-01

    En uddannelse til coach har ændret Helle Rototzens liv. Som 49-årig forlod hun et forskerliv på deltid til fordel for en karriere som DTU's eneste kvindelige institutdirektør.......En uddannelse til coach har ændret Helle Rototzens liv. Som 49-årig forlod hun et forskerliv på deltid til fordel for en karriere som DTU's eneste kvindelige institutdirektør....

  13. Control rod velocity limiter

    International Nuclear Information System (INIS)

    Cearley, J.E.; Carruth, J.C.; Dixon, R.C.; Spencer, S.S.; Zuloaga, J.A. Jr.

    1986-01-01

    This patent describes a velocity control arrangement for a reciprocable, vertically oriented control rod for use in a nuclear reactor in a fluid medium, the control rod including a drive hub secured to and extending from one end therefrom. The control device comprises: a toroidally shaped control member spaced from and coaxially positioned around the hub and secured thereto by a plurality of spaced radial webs thereby providing an annular passage for fluid intermediate the hub and the toroidal member spaced therefrom in coaxial position. The side of the control member toward the control rod has a smooth generally conical surface. The side of the control member away from the control rod is formed with a concave surface constituting a single annular groove. The device also comprises inner and outer annular vanes radially spaced from one another and spaced from the side of the control member away from the control rod and positioned coaxially around and spaced from the hub and secured thereto by spaced radial webs thereby providing an annular passage for fluid intermediate the hub and the vanes. The vanes are angled toward the control member, the outer edge of the inner vane being closer to the control member and the inner edge of the outer vane being closer to the control member. When the control rod moves in the fluid in the direction toward the drive hub the vanes direct a flow of fluid turbulence which provides greater resistance to movement of the control rod in the direction toward the drive hub than in the other direction

  14. Experimental studies of the effect of rod spacing on burnout in a simulated rod bundle

    International Nuclear Information System (INIS)

    Lee, D.H.; Little, R.B.

    1962-08-01

    Tests on a dumb-bell shaped flow passage simulating the gap between rods in a fuel element indicated that burnout was not significantly affected by inter-rod gap in the range 0.032'' to 0.22''. Test conditions were: 960 p.s.i.a., 2 x 10 6 1b/ft 2 hr mass velocity, and 10% mean exit quality with vertical upflow of water. (author)

  15. 21. century drilling rigs -- Tesco introduces new modular design

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    1998-10-01

    Development of a modular, hydraulic, self-elevating drilling rig, dubbed the `21. century drilling rig` was announced by the Tesco Corporation. The rig equipment is housed in 8 by 20 by 8.5 feet high sea containers that can be handled by a 20-ton oilfield picker. These containers, weighing about 15,000 to 20,000 pounds on average, eliminate the need for heavy and bulky standard oilfield skid buildings, besides avoiding costly over-width and over-weight permits. The containers can be easily shipped around the world at a fraction of the cost of shipping standard oilfield skid buildings. Time for shipping on land is comparable to conventional rigs, but with the added advantage of smaller and lighter loads, promising fewer transportation problems during spring breakup. Tesco also designed and built an 85-foot long, triple-axle, 24-wheel catwalk trailer to transport the top drive, drawworks and double telescoping mast as one unit. Another novel characteristic of this unit is that the hydraulic system is capable of selectable distribution of power to the main functions such as the top drive, drawworks, or mud pump, similar to the electric SCR rig. The rig also features a computerized control system managed by programmable logic controllers. The split crown and split block to facilitate wireline work, are other innovative features worthy of note.

  16. Control rod housing alignment

    International Nuclear Information System (INIS)

    Dixon, R.C.; Deaver, G.A.; Punches, J.R.; Singleton, G.E.; Erbes, J.G.; Offer, H.P.

    1990-01-01

    This patent describes a process for measuring the vertical alignment between a hole in a core plate and the top of a corresponding control rod drive housing within a boiling water reactor. It comprises: providing an alignment apparatus. The alignment apparatus including a lower end for fitting to the top of the control rod drive housing; an upper end for fitting to the aperture in the core plate, and a leveling means attached to the alignment apparatus to read out the difference in angularity with respect to gravity, and alignment pin registering means for registering to the alignment pin on the core plate; lowering the alignment device on a depending support through a lattice position in the top guide through the hole in the core plate down into registered contact with the top of the control rod drive housing; registering the upper end to the sides of the hole in the core plate; registering the alignment pin registering means to an alignment pin on the core plate to impart to the alignment device the required angularity; and reading out the angle of the control rod drive housing with respect to the hole in the core plate through the leveling devices whereby the angularity of the top of the control rod drive housing with respect to the hole in the core plate can be determined

  17. LOFT advanced fuel rod instrumentation development

    International Nuclear Information System (INIS)

    Billeter, T.R.; Brown, R.L.; Chan, A.I.Y.; Day, C.K.; Meyers, S.C.; Sheen, E.M.; Stringer, J.L.

    1978-01-01

    Advanced fuel rod instrumentation for the Loss of Fluid Test (LOFT) reactor is being developed by the Hanford Engineering Development Laboratory for the Nuclear Regulatory Commission. This effort calls for development of sensors to measure fuel rod axial motion, fuel centerline temperature (to 2200 0 C), fuel rod plenum gas pressure (to 2500 psig), and plenum gas temperature (to 1500 0 F). A parallel test and evaluation of several modified commercial sensors was undertaken and will result in commercial availability of the final qualified sensors. Necessary test facilities were prepared for the development and evaluation effort. Tests to date indicate a three coil Linear Variable Differential Transformer (LVDT), operated from temperature compensating signal source and processing electronics, will meet the desired requirements

  18. Process development and fabrication for sphere-pac fuel rods

    International Nuclear Information System (INIS)

    Welty, R.K.; Campbell, M.H.

    1981-06-01

    Uranium fuel rods containing sphere-pac fuel have been fabricated for in-reactor tests and demonstrations. A process for the development, qualification, and fabrication of acceptable sphere-pac fuel rods is described. Special equipment to control fuel contamination with moisture or air and the equipment layout needed for rod fabrication is described and tests for assuring the uniformity of the fuel column are discussed. Fuel retainers required for sphere-pac fuel column stability and instrumentation to measure fuel column smear density are described. Results of sphere-pac fuel rod fabrication campaigns are reviewed and recommended improvements for high throughput production are noted

  19. Design, manufacture and construction of a compressor test rig and the start of experimental operation of a low speed axial compressor at Dresden. Final report; Konstruktion, Fertigung und Aufbau eines Verdichterpruefstandes und Aufnahme des Versuchsbetriebes an einem Niedergeschwindigkeits-Axialverdichter in Dresden. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Bernstein, W.; Bernhard, H.; Biesinger, T.; Boos, P.; Moeckel, H.; Sauer, H.

    1996-12-01

    In this report, the design, manufacture and construction of the low speed compressor, the build-up of the compressor test rig and the measurement technique used are described. The first measured results obtained after setting to work and the start of experimental operation on the rotational symmetry at the compressor inlet and outlet and of a flow field behind the rotor and stator of the third stage are described. The operating period of 540 hours to the end of the subject shows faultfree operation of the experimental plant. (orig./AKF) [Deutsch] Im vorliegenden Bericht werden Konstruktion, Fertigung und Aufbau des Niedergeschwindigkeitsverdichters, der Aufbau des Verdichterpruefstandes und die verwendete Messtechnik beschrieben. Die nach der Inbetriebnahme und Aufnahme des Versuchsbetriebes erhaltenen ersten Messergebnisse zur Rotationssymmetrie am Ein- und Austritt des Verdichters und von einem Stroemungsfeld hinter dem Rotor und Stator der dritten Stufe werden geschildert. Die zum Abschluss des Themas erreichte Betriebszeit von 540 Stunden weist auf einen stoerungsfreien Betrieb der Versuchsanlage hin. (orig./AKF)

  20. Control rod drives

    International Nuclear Information System (INIS)

    Asano, Hiromitsu.

    1979-01-01

    Purpose: To drive control rods at an optimum safety speed corresponding to the reactor core output. Constitution: The reactor power is detected by a neutron detector and the output signal is applied to a process computer. The process computer issues a signal representing the reactor core output, which is converted through a function generator into a signal representing the safety speed of control rods. The converted signal is further supplied to a V/F converter and converted into a pulse signal. The pulse signal is inputted to a step motor driving circuit, which actuates a step motor to operate the control rods always at a safety speed corresponding to the reactor core power. (Furukawa, Y.)