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Sample records for test loop facilities

  1. Operation of the hot test loop facilities

    International Nuclear Information System (INIS)

    Cheong, Moon Ki; Park, Choon Kyeong; Won, Soon Yeon; Yang, Sun Kyu; Cheong, Jang Whan; Cheon, Se Young; Song, Chul Hwa; Jeon, Hyeong Kil; Chang, Suk Kyu; Jeong, Heung Jun; Cho, Young Ro; Kim, Bok Duk; Min, Kyeong Ho

    1994-12-01

    The objective of this project is to obtain the available experimental data and to develop the measuring techniques through taking full advantage of the facilities. The facilities operated by the thermal hydraulics department have been maintained and repaired in order to carry out the thermal hydraulics tests necessary for providing the available data. The performance tests for double grid type bottom end piece which was improved on the debris filtering effectivity were performed using the PWR-Hot Test Loop. The CANDU-Hot Test Loop was operated to carry out the pressure drop tests and strength tests of fuel. The Cold Test Loop was used to obtain the local velocity data in subchannel within fuel bundle and to understand the characteristic of pressure drop required for improving the nuclear fuel and to develop the advanced measuring techniques. RCS Loop, which is used to measure the CHF, is presently under design and construction. B and C Loop is designed and constructed to assess the automatic depressurization safety system behavior. 4 tabs., 79 figs., 7 refs. (Author) .new

  2. Operation of the nuclear fuel cycle test facilities -Operation of the hot test loop facilities

    International Nuclear Information System (INIS)

    Chun, S. Y.; Jeong, M. K.; Park, C. K.; Yang, S. K.; Won, S. Y.; Song, C. H.; Jeon, H. K.; Jeong, H. J.; Cho, S.; Min, K. H.; Jeong, J. H.

    1997-01-01

    A performance and reliability of a advanced nuclear fuel and reactor newly designed should be verified by performing the thermal hydraulics tests. In thermal hydraulics research team, the thermal hydraulics tests associated with the development of an advanced nuclear fuel and reactor haven been carried out with the test facilities, such as the Hot Test Loop operated under high temperature and pressure conditions, Cold Test Loop, RCS Loop and B and C Loop. The objective of this project is to obtain the available experimental data and to develop the advanced measuring techniques through taking full advantage of the facilities. The facilities operated by the thermal hydraulics research team have been maintained and repaired in order to carry out the thermal hydraulics tests necessary for providing the available data. The performance tests for the double grid type bottom end piece which was improved on the debris filtering effectivity were performed using the PWR-Hot Test Loop. The CANDU-Hot Test Loop was operated to carry out the pressure drop tests and strength tests of CANFLEX fuel. The Cold Test Loop was used to obtain the local velocity data in subchannel within HANARO fuel bundle and to study a thermal mixing characteristic of PWR fuel bundle. RCS thermal hydraulic loop was constructed and the experiments have been carried out to measure the critical heat flux. In B and C Loop, the performance tests for each component were carried out. (author). 19 tabs., 78 figs., 19 refs

  3. Operation of the nuclear fuel cycle test facilities -Operation of the hot test loop facilities

    Energy Technology Data Exchange (ETDEWEB)

    Chun, S. Y.; Jeong, M. K.; Park, C. K.; Yang, S. K.; Won, S. Y.; Song, C. H.; Jeon, H. K.; Jeong, H. J.; Cho, S.; Min, K. H.; Jeong, J. H.

    1997-01-01

    A performance and reliability of a advanced nuclear fuel and reactor newly designed should be verified by performing the thermal hydraulics tests. In thermal hydraulics research team, the thermal hydraulics tests associated with the development of an advanced nuclear fuel and reactor haven been carried out with the test facilities, such as the Hot Test Loop operated under high temperature and pressure conditions, Cold Test Loop, RCS Loop and B and C Loop. The objective of this project is to obtain the available experimental data and to develop the advanced measuring techniques through taking full advantage of the facilities. The facilities operated by the thermal hydraulics research team have been maintained and repaired in order to carry out the thermal hydraulics tests necessary for providing the available data. The performance tests for the double grid type bottom end piece which was improved on the debris filtering effectivity were performed using the PWR-Hot Test Loop. The CANDU-Hot Test Loop was operated to carry out the pressure drop tests and strength tests of CANFLEX fuel. The Cold Test Loop was used to obtain the local velocity data in subchannel within HANARO fuel bundle and to study a thermal mixing characteristic of PWR fuel bundle. RCS thermal hydraulic loop was constructed and the experiments have been carried out to measure the critical heat flux. In B and C Loop, the performance tests for each component were carried out. (author). 19 tabs., 78 figs., 19 refs.

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  5. Gas Test Loop Facilities Alternatives Assessment Report Rev 1

    International Nuclear Information System (INIS)

    William J. Skerjanc; William F. Skerjanc

    2005-01-01

    An important task in the Gas Test Loop (GTL) conceptual design was to determine the best facility to serve as host for this apparatus, which will allow fast-flux neutron testing in an existing nuclear facility. A survey was undertaken of domestic and foreign nuclear reactors and accelerator facilities to arrive at that determination. Two major research reactors in the U.S. were considered in detail, the Advanced Test Reactor (ATR) and the High Flux Isotope Reactor (HFIR), each with sufficient power to attain the required neutron fluxes. HFIR routinely operates near its design power limit of 100 MW. ATR has traditionally operated at less than half its design power limit of 250 MW. Both of these reactors should be available for at least the next 30 years. The other major U.S. research reactor, the Missouri University Research Reactor, does not have sufficient power to reach the required neutron flux nor do the smaller research reactors. Of the foreign reactors investigated, BOR-60 is perhaps the most attractive. Monju and BN 600 are power reactors for their respective electrical grids. Although the Joyo reactor is vigorously campaigning for customers, local laws regarding transport of radioactive material mean it would be very difficult to retrieve test articles from either Japanese reactor for post irradiation examination. PHENIX is scheduled to close in 2008 and is fully booked until then. FBTR is limited to domestic (Indian) users only. Data quality is often suspect in Russia. The only accelerator seriously considered was the Fuel and Material Test Station (FMTS) currently proposed for operation at Los Alamos National Laboratory. The neutron spectrum in FMTS is similar to that found in a fast reactor, but it has a pronounced high-energy tail that is atypical of fast fission reactor spectra. First irradiation in the FMTS is being contemplated for 2008. Detailed review of these facilities resulted in the recommendation that the ATR would be the best host for the GTL

  6. Experimental facilities for PEC reactor design central channel test loop: CPC-1 - thermal shocks loop: CEDI

    International Nuclear Information System (INIS)

    Calvaresi, C.; Moreschi, L.F.

    1983-01-01

    PEC (Prova Elementi di Combustibile: Fuel Elements Test) is an experimental fast sodium-cooled reactor with a power of 120 MWt. This reactor aims at studying the behaviour of fuel elements under thermal and neutron conditions comparable with those existing in fast power nuclear facilities. Given the particular structure of the core, the complex operations to be performed in the transfer cell and the strict operating conditions of the central channel, two experimental facilities, CPC-1 and CEDI, have been designed as a support to the construction of the reactor. CPC-1 is a 1:1 scale model of the channel, transfer-cell and loop unit of the channel, whereas CEDI is a sodium-cooled loop which enables to carry out tests of isothermal endurance and thermal shocks on the group of seven forced elements, by simulating the thermo-hydraulic and mechanical conditions existing in the reactor. In this paper some experimental test are briefy discussed and some facilities are listed, both for the CPC-1 and for the CEDI. (Auth.)

  7. The Phillips Laboratory capillary pumped loop test facility

    Science.gov (United States)

    Gluck, Donald F.; Kaylor, Marc C.

    1996-03-01

    An ammonia capillary pumped loop (CPL) test facility has been designed, fabricated, subject to acceptance tests, and assembled at Phillips Laboratory. Its intent is to support a wide range of Air Force programs, bringing CPL technology to flight readiness for operational systems. The facility provides a high degree of modularity and flexibility with several heating and cooling options, and capability for elevation (+/- 15 in.), tilt (+/-60°) and transport length variation. It has a 182 by 44 by 84 inch envelope, an expected heat load capability of 2500 W, and a temperature range of 0 to 50 °C. The evaporator section has two plates with four capillary pumps (CPs) each, with a starter pump on one plate. The CPs are 5/8 in., with TAG aluminum 6063-T6 casing and UHMW polyethylene wicks. The active lengths are 15 and 30 inch with both 10 and 15 micron wicks. The individual CPs have thermal and hydraulic isolation capability, and are removable. The transport section consists of stainless steel lines in a serpentine configuration, a 216 in3 free volume reservoir, and a mechanical pump. The vapor transport line contains a capillary device (which can be bypassed) for vapor blockage during startup. The condenser consists of two separately valved, parallel cold plates each with a downstream noncondensible gas trap. Cooling of up to 1500 W at -50 °C is provided by an FTS Systems chiller using Flourinert FC-72. An enclosure/exhaust system is provided for safety and emergency venting of ammonia. An ammonia charge station performs or supports the functions of proof pressure, flushing with ammonia, purging with gaseous nitrogen, evacuation of all or part of the CPL to 20 microns, and charging. Instrumentation consists of over 116 thermocouples, five of which are internal; one absolute and six differential pressure transducers; eleven watt transducers, and a reservoir load cell. The data acquisition system consists of a temperature scanner, Bernoulli drive, and two Macintosh

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

    International Nuclear Information System (INIS)

    Oakley, D.J.

    1978-01-01

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

  9. Safety report content and development for test loop facility on MARIA reactor

    International Nuclear Information System (INIS)

    Konechko, A.; Shumskij, A.M.; Mikul'ahin, V.E.

    1982-01-01

    A 600 kW test loop facility for investigatin.o safety problems is realized on MARIA reactor in Poland together with USSR organizations. Safety reports have been developed in two steps at the designstage. The 1st report being essentially a preliminary safety analysis was developed within the scope of the feasibility study. At the engineering design stage the preliminary test loop facility safety report had been prepared considering measures excluding the possibility of the MARIA reactor damage. The test loop facility safety report is fulfilled for normal, transient and emergency operation regimes. Separate safety basing for each group of experiments will be prepared. The report presents the test loop facility safety criteria coordinated by the nuclear safety comission. They contains the preliminary reports on the test loop facility safety. At the final stage of construction and at thecommitioning stage the start-up safety report will be developed which after required correction and adding up the putting into operation data will turn into operation safety report [ru

  10. PETER loop. Multifunctional test facility for thermal hydraulic investigations of PWR fuel elements

    International Nuclear Information System (INIS)

    Ganzmann, I.; Hille, D.; Staude, U.

    2009-01-01

    The reliable fuel element behavior during the complete fuel cycle is one of the fundamental prerequisites of a safe and efficient nuclear power plant operation. The fuel element behavior with respect to pressure drop and vibration impact cannot be simulated by means of fluid-structure interaction codes. Therefore it is necessary to perform tests using fuel element mock-ups (1:1). AREVA NP has constructed the test facility PETER (PWR fuel element tests in Erlangen) loop. The modular construction allows maximum flexibility for any type of fuel elements. Modern measuring instrumentation for flow, pressure and vibration characterization allows the analysis of cause and consequences of thermal hydraulic phenomena. PETER loop is the standard test facility for the qualification of dynamic fuel element behavior in flowing fluid and is used for failure mode analysis.

  11. Conceptual Design for a High-Temperature Gas Loop Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    James B. Kesseli

    2006-08-01

    This report documents an early-stage conceptual design for a high-temperature gas test loop. The objectives accomplished by the study include, (1) investigation of existing gas test loops to determine ther capabilities and how the proposed system might best complement them, (2) development of a preliminary test plan to help identify the performance characteristics required of the test unit, (3) development of test loop requirements, (4) development of a conceptual design including process flow sheet, mechanical layout, and equipment specifications and costs, and (5) development of a preliminary test loop safety plan.

  12. The Analysis of Loop Seal Purge Time for the KHNP Pressurizer Safety Valve Test Facility Using the GOTHIC Code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Ae; Kim, Chang Hyun; Kweon, Gab Joo; Park, Jong Woon [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of)

    2007-10-15

    The pressurizer safety valves (PSV) in Pressurized Water Reactors are required to provide the overpressure protection for the Reactor Coolant System (RCS) during the overpressure transients. Korea Hydro and Nuclear Power Company (KHNP) plans to build the PSV test facility for the purpose of providing the PSV pop-up characteristics and the loop seal dynamics for the new safety analysis. When the pressurizer safety valve is mounted in a loop seal configuration, the valve must initially pass the loop seal water prior to popping open on steam. The loop seal in the upstream of PSV prevents leakage of hydrogen gas or steam through the safety valve seat. This paper studies on the loop seal clearing dynamics using GOTHIC-7.2a code to verify the effects of loop seal purge time on the reactor coolant system overpressure.

  13. Reactor recirculation pump test loop

    International Nuclear Information System (INIS)

    Taka, Shusei; Kato, Hiroyuki

    1979-01-01

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

  14. 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

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

    International Nuclear Information System (INIS)

    Gerstner, Douglas M.

    2009-01-01

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

  16. Hardware-in-the-Loop Testing

    Data.gov (United States)

    Federal Laboratory Consortium — RTC has a suite of Hardware-in-the Loop facilities that include three operational facilities that provide performance assessment and production acceptance testing of...

  17. Pre-test analysis of a LBLOCA using the design data of the ATLAS facility, a reduced-height integral effect test loop for PWRs

    International Nuclear Information System (INIS)

    Hyun-Sik Park; Ki-Yong Choi; Dong-Jin Euh; Tae-Soon Kwon; Won-Pil Baek

    2005-01-01

    Full text of publication follows: The simulation capability of the KAERI integral effect test facility, ATLAS (Advanced Thermalhydraulic Test Loop for Accident Simulation), has been assessed for a large-break loss-of-coolant accident (LBLOCA) transient. The ATLAS facility is a 1/2 height-scaled, 1/144 area-scaled (1/288 in volume scale), and full-pressure test loop based on the design features of the APR1400, an evolutionary pressurized water reactor that has been developed by Korean industry. The APR1400 has four mechanically separated hydraulic trains for the emergency core cooling system (ECCS) with direct vessel injection (DVI). The APR1400 design features have brought about several new safety issues related to the LBLOCA including the steam-water interaction, ECC bypass, and boiling in the reactor vessel downcomer. The ATLAS facility will be used to investigate the multiple responses between the systems or between the components during various anticipated transients. The ATLAS facility has been designed according to a scaling method that is mainly based on the model suggested by Ishii and Kataoka. The ATLAS facility is being evaluated against the prototype plant APR1400 with the same control logics and accident scenarios using the best-estimated code, MARS. This paper briefly introduces the basic design features of the ATLAS facility and presents the results of pre-test analysis for a postulated LBLOCA of a cold leg. The LBLOCA analyses has been conducted to assess the validity of the applied scaling law and the similarity between the ATLAS facility and the APR1400. As the core simulator of the ATLAS facility has the 10% capability of the scaled full power, the blowdown phase can not be simulated, and the starting point of the accident scenario is around the end of blowdown. So it is an important problem to find the correct initial conditions. For the analyzed LBLOCA scenario, the ATLAS facility showed very similar thermal-hydraulic characteristics to the APR

  18. High-temperature helium-loop facility

    International Nuclear Information System (INIS)

    Tokarz, R.D.

    1981-09-01

    The high-temperature helium loop is a facility for materials testing in ultrapure helium gas at high temperatures. The closed loop system is capable of recirculating high-purity helium or helium with controlled impurities. The gas loop maximum operating conditions are as follows: 300 psi pressure, 500 lb/h flow rate, and 2100 0 F temperature. The two test sections can accept samples up to 3.5 in. diameter and 5 ft long. The gas loop is fully instrumented to continuously monitor all parameters of loop operation as well as helium impurities. The loop is fully automated to operate continuously and requires only a daily servicing by a qualified operator to replenish recorder charts and helium makeup gas. Because of its versatility and high degree of parameter control, the helium loop is applicable to many types of materials research. This report describes the test apparatus, operating parameters, peripheral systems, and instrumentation system. The experimental capabilities and test conand presents the results that have been obtained. The study has been conducted using a four-phase approach. The first phase develops the solution to the steady-state radon-diffusion equation in one-dimensieered barriers; disposal charge analysis; analysis of spent fuel policy implementation; spent f water. Field measurements and observations are reported for each site. Analytical data and field measurements are presented in tables and maps. Uranium concentrations in the sediments which were above detection limits ranged from 0.10 t 51.2 ppM. The mean of the logarithms of the uranium concentrations was 0.53. A group of high uranium concentrations occurs near the junctions of quadrangles AB, AC, BB, a 200 mK. In case 2), x-ray studies of isotopic phase separation in 3 He-- 4 He bcc solids were carried out by B. A. Fraass

  19. BWR recirculation loop discharge line break LOCA tests with break areas of 50 and 100% assuming HPCS failure at ROSA-III test facility

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Tasaka, Kanji; Yonomoto, Taisuke; Anoda, Yoshinari; Kumamaru, Hiroshige; Nakamura, Hideo; Murata, Hideo; Shiba, Masayoshi; Iriko, Masanori.

    1985-03-01

    This report presents the experimental results of RUN 962 and RUN 963 in ROSA-III program, which are 50 and 100 % break LOCA tests at the BWR recirculation pump discharge line, respectively. The ROSA-III test facility simulates a volumetrically scaled (1/424) BWR system and has four half-length electrically heated fuel bundles, two active recirculation loops, three types of ECCSs and steam and feedwater systems. The experimental data of RUN 962 and RUN 963 were compared with those of RUN 961, a 200 % discharge line break test to study the break area effects on the transient thermal hydraulic phenomena. The least flow areas at the jet pump drive nozzles and recirculation pump discharge nozzle in the broken recirculation loop limitted the discharge flows from the pressure vessel and the depressurization rate in the 100 and 200 % break tests, whereas the least flow area at break nozzle limitted the depressurization rate in the 50 % break test. The highest PCT was observed in the 50 % break test among the three tests. (author)

  20. The Self-Calibration Test of flowmeter installed in STELLA(Sodium Integral Effect Test Loop for Safety Simulation and Assessment) facility

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Minhwan; Jeong, Ji-Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The objective of this study is to describe the procedure of the self-calibration test for the flowmeters and to analyze the result of the test. In this work, the test procedure of the self-calibration of two flowmeters (FT-101, FT-102) installed in STELLA facility was described and the test result was analyzed. In regard to the long-term SFR development plan, a large-scale sodium thermal-hydraulic test project is being progressed by KAERI. This project is called STELLA (Sodium Integral Effect Test Loop for Safety Simulation and Assessment), and it is proceeding by adopting the QA (Quality Assurance) program. Due to the specificity of an experiment using sodium(Na) categorized as Class 3(pyrophoric material and water-prohibiting substance) by the Safety Control of Dangerous Substances Act, it is necessary to apply QA in consideration of the sodium experiment environment in certain parts. The one of them is about calibration of measuring instrument such as a flowmeter, thermocouple and pressure gauge. It is described in the QAP (Quality Assurance Procedures) of KAERI that calibration work should be conducted in accordance with self-calibration procedures in a special case where conventional calibration is not practicable. The calibration of two flowmeters (FT-101, FT-102) installed in STELLA facility is the typical example. As a result of test, it was confirmed that the flowmeters meet the pass criterion. Therefore, it was concluded that the flowmeters maintain instrument capacity a year ago.

  1. Technical specification of HANARO fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. Y

    1998-03-01

    The design and installation of the irradiation test facility for verification test of the fuel performance are very important in connection with maximization of the utilization of HANARO. HANARO fuel test loop was designed in accordance with the same code and standards of nuclear power plant because HANARO FTL will be operated the high pressure and temperature same as nuclear power plant operation conditions. The objective of this study is to confirm the operation limit, safety limit, operation condition and checking points of HANARO fuel test loop. This results will become guidances for the planning of irradiation testing and operation of HANARO fuel test loop. (author). 13 refs., 13 tabs., 8 figs.

  2. DeBeNe Test Facilities for Fast Breeder Development

    International Nuclear Information System (INIS)

    Storz, R.

    1980-10-01

    This report gives an overview and a short description of the test facilities constructed and operated within the collaboration for fast breeder development in Germany, Belgium and the Netherlands. The facilities are grouped into Sodium Loops (Large Facilities and Laboratory Loops), Special Equipment including Hot Cells and Reprocessing, Test Facilities without Sodium, Zero Power Facilities and In-pile Loops including Irradiation Facilities

  3. Thermal hydraulic behavior of a PWR under beyond-design-basis accident conditions: Conclusions from an experimental program in a 4-loop test facility (PKL)

    International Nuclear Information System (INIS)

    Umminger, K.J.; Kastner, W.; Mandl, R.M.; Weber, P.

    1993-01-01

    Within the scope of German reactor safety research, extensive experiments covering the behavior of nuclear power plants under accident conditions have been carried out in the PKL test facility which simulates a 4-loop, 1,300 MWe KWU-designed PWR. While the investigations dealing with design-basis accidents and with the efficiency of the emergency core cooling systems have been largely completed, the main interest nowadays concentrates on the investigation of beyond-design-basis accidents to demonstrate the safety margins of nuclear power plants and to investigate the contribution of the built-in safety features for a further reduction of the residual risk. The thermal hydraulic behavior of a PWR under these extreme accident conditions was experimentally investigated within the PKL III B test program. This paper presents the fundamental findings with some of the most important results being discussed in detail. Future plans are also outlined

  4. Fast Flux Test Facility

    International Nuclear Information System (INIS)

    Munn, W.I.

    1981-01-01

    The Fast Flux Test Facility (FFTF), located on the Hanford site a few miles north of Richland, Washington, is a major link in the chain of development required to sustain and advance Liquid Metal Fast Breeder Reactor (LMFBR) technology in the United States. This 400 MWt sodium cooled reactor is a three loop design, is operated by Westinghouse Hanford Company for the US Department of Energy, and is the largest research reactor of its kind in the world. The purpose of the facility is three-fold: (1) to provide a test bed for components, materials, and breeder reactor fuels which can significantly extend resource reserves; (2) to produce a complete body of base data for the use of liquid sodium in heat transfer systens; and (3) to demonstrate inherent safety characteristics of LMFBR designs

  5. Conceptural design of multipurpose sodium test loop

    International Nuclear Information System (INIS)

    Kim, W.C.; Lee, Y.W.; Nam, H.Y.; Chun, S.Y.; Kim, J.; Yuh, M.W.

    1982-01-01

    This report describes the conceptural design of the multipurpose sodium test loop (MSTL). This MSTL consists mainly of impurity control and measurement system, corrosion and masstransfer system and heat transfer system. Problems associated with liquid sodium coolant will be studied and operating experiences will be obtained by the use of this facility. This technology will be used to evaluate safety and reliability of large sodium facility in the future. The total cost excluding the cost of building construction is estimated to 175 thousand dollars. (Author)

  6. Comparison of three small-break loss-of-coolant accident tests with different break locations using the system-integrated modular advanced reactor-integral test loop facility to estimate the safety of the smart design

    OpenAIRE

    Hwang Bae; Dong Eok Kim; Sung-Uk Ryu; Sung-Jae Yi; Hyun-Sik Park

    2017-01-01

    Three small-break loss-of-coolant accident (SBLOCA) tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor), i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal–hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are s...

  7. UPTF loop seal tests and their RELAP simulation

    International Nuclear Information System (INIS)

    Tuomainen, M.; Tuunanen, J.

    1997-01-01

    In a pressurized water reactor the loop seals have an effect on the natural circulation. If a loop seal is filled with water it can cause a flow stagnation in the loop during two-phase natural circulation. Also the pressure loss over a filled loop seal is high, which lowers the water level in the core. Tests to investigate the loop seal behaviour were performed on a German Upper Plenum Test Facility (UPTF). The purpose of the tests was to study the amount of water in the loop seal under different steam flow rates. The tests were simulated with RELAP5/MOD3.2. With high steam flow rates the code had problems in simulating the amount of the water remaining in the pump elbow, but in general the agreement between the calculated results and the experimental data was good. (orig.)

  8. Engine Test Facility (ETF)

    Data.gov (United States)

    Federal Laboratory Consortium — The Air Force Arnold Engineering Development Center's Engine Test Facility (ETF) test cells are used for development and evaluation testing of propulsion systems for...

  9. Gas Test Loop Booster Fuel Hydraulic Testing

    International Nuclear Information System (INIS)

    Gas Test Loop Hydraulic Testing Staff

    2006-01-01

    The Gas Test Loop (GTL) project is for the design of an adaptation to the Advanced Test Reactor (ATR) to create a fast-flux test space where fuels and materials for advanced reactor concepts can undergo irradiation testing. Incident to that design, it was found necessary to make use of special booster fuel to enhance the neutron flux in the reactor lobe in which the Gas Test Loop will be installed. Because the booster fuel is of a different composition and configuration from standard ATR fuel, it is necessary to qualify the booster fuel for use in the ATR. Part of that qualification is the determination that required thermal hydraulic criteria will be met under routine operation and under selected accident scenarios. The Hydraulic Testing task in the GTL project facilitates that determination by measuring flow coefficients (pressure drops) over various regions of the booster fuel over a range of primary coolant flow rates. A high-fidelity model of the NW lobe of the ATR with associated flow baffle, in-pile-tube, and below-core flow channels was designed, constructed and located in the Idaho State University Thermal Fluids Laboratory. A circulation loop was designed and constructed by the university to provide reactor-relevant water flow rates to the test system. Models of the four booster fuel elements required for GTL operation were fabricated from aluminum (no uranium or means of heating) and placed in the flow channel. One of these was instrumented with Pitot tubes to measure flow velocities in the channels between the three booster fuel plates and between the innermost and outermost plates and the side walls of the flow annulus. Flow coefficients in the range of 4 to 6.5 were determined from the measurements made for the upper and middle parts of the booster fuel elements. The flow coefficient for the lower end of the booster fuel and the sub-core flow channel was lower at 2.3

  10. Gas Test Loop Booster Fuel Hydraulic Testing

    Energy Technology Data Exchange (ETDEWEB)

    Gas Test Loop Hydraulic Testing Staff

    2006-09-01

    The Gas Test Loop (GTL) project is for the design of an adaptation to the Advanced Test Reactor (ATR) to create a fast-flux test space where fuels and materials for advanced reactor concepts can undergo irradiation testing. Incident to that design, it was found necessary to make use of special booster fuel to enhance the neutron flux in the reactor lobe in which the Gas Test Loop will be installed. Because the booster fuel is of a different composition and configuration from standard ATR fuel, it is necessary to qualify the booster fuel for use in the ATR. Part of that qualification is the determination that required thermal hydraulic criteria will be met under routine operation and under selected accident scenarios. The Hydraulic Testing task in the GTL project facilitates that determination by measuring flow coefficients (pressure drops) over various regions of the booster fuel over a range of primary coolant flow rates. A high-fidelity model of the NW lobe of the ATR with associated flow baffle, in-pile-tube, and below-core flow channels was designed, constructed and located in the Idaho State University Thermal Fluids Laboratory. A circulation loop was designed and constructed by the university to provide reactor-relevant water flow rates to the test system. Models of the four booster fuel elements required for GTL operation were fabricated from aluminum (no uranium or means of heating) and placed in the flow channel. One of these was instrumented with Pitot tubes to measure flow velocities in the channels between the three booster fuel plates and between the innermost and outermost plates and the side walls of the flow annulus. Flow coefficients in the range of 4 to 6.5 were determined from the measurements made for the upper and middle parts of the booster fuel elements. The flow coefficient for the lower end of the booster fuel and the sub-core flow channel was lower at 2.3.

  11. Rocketball Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — This test facility offers the capability to emulate and measure guided missile radar cross-section without requiring flight tests of tactical missiles. This facility...

  12. Textiles Performance Testing Facilities

    Data.gov (United States)

    Federal Laboratory Consortium — The Textiles Performance Testing Facilities has the capabilities to perform all physical wet and dry performance testing, and visual and instrumental color analysis...

  13. GPS Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Global Positioning System (GPS) Test Facility Instrumentation Suite (GPSIS) provides great flexibility in testing receivers by providing operational control of...

  14. Comparison of three small-break loss-of-coolant accident tests with different break locations using the system-integrated modular advanced reactor-integral test loop facility to estimate the safety of the smart design

    Directory of Open Access Journals (Sweden)

    Hwang Bae

    2017-08-01

    Full Text Available Three small-break loss-of-coolant accident (SBLOCA tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor, i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal–hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

  15. Comparison of three small-break loss-of-coolant accident tests with different break locations using the system-integrated modular advanced reactor-integral test loop facility to estimate the safety of the smart design

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Hwang; Ryu, Sung Uk; Yi, Sung Jae; Park, Hyun Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Dong Eok [Dept. of Precision Mechanical Engineering, Kyungpook National University, Sangju (Korea, Republic of)

    2017-08-15

    Three small-break loss-of-coolant accident (SBLOCA) tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor), i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal–hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

  16. CLEAR test facility

    CERN Multimedia

    Ordan, Julien Marius

    2017-01-01

    A new user facility for accelerator R&D, the CERN Linear Electron Accelerator for Research (CLEAR), started operation in August 2017. CLEAR evolved from the former CLIC Test Facility 3 (CTF3) used by the Compact Linear Collider (CLIC). The new facility is able to host and test a broad range of ideas in the accelerator field.

  17. Ouellette Thermal Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Thermal Test Facility is a joint Army/Navy state-of-the-art facility (8,100 ft2) that was designed to:Evaluate and characterize the effect of flame and thermal...

  18. Fast Flux Test Facility primary sodium valves

    International Nuclear Information System (INIS)

    Rabe, G.B.; Ezra, B.C.

    1977-01-01

    The design and development of the valves used in the primary sodium coolant loop of the Fast Flux Test Facility is described. One tilting-disk check valve is used in the cold leg of the coolant loop. It is designed to limit flow reversal in the loop while maintaining a low pressure drop during forward flow. Two isolation valves are used in each coolant loop--one in the cold leg and one in the hot leg. They are of the motor-operated swinging-gate type. The design, analysis, and testing programs undertaken to develop and qualify these valves are described

  19. Smart Home Hardware-in-the-Loop Testing

    Energy Technology Data Exchange (ETDEWEB)

    Pratt, Annabelle

    2017-07-12

    This presentation provides a high-level overview of NREL's smart home hardware-in-the-loop testing. It was presented at the Fourth International Workshop on Grid Simulator Testing of Energy Systems and Wind Turbine Powertrains, held April 25-26, 2017, hosted by NREL and Clemson University at the Energy Systems Integration Facility in Golden, Colorado.

  20. Mark 1 Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Mark I Test Facility is a state-of-the-art space environment simulation test chamber for full-scale space systems testing. A $1.5M dollar upgrade in fiscal year...

  1. Structural Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — Provides a wide variety of testing equipment, fixtures and facilities to perform both unique aviation component testing as well as common types of materials testing...

  2. Pavement Testing Facility

    Data.gov (United States)

    Federal Laboratory Consortium — Comprehensive Environmental and Structural AnalysesThe ERDC Pavement Testing Facility, located on the ERDC Vicksburg campus, was originally constructed to provide an...

  3. The conversion of a room temperature NaK loop to a high temperature MHD facility for Li/V blanket testing

    International Nuclear Information System (INIS)

    Reed, C.B.; Haglund, R.C.; Miller, M.E.; Nasiatka, J.R.; Kirillov, I.R.; Ogorodnikov, A.P.; Preslitski, G.V.; Goloubovitch, G.P.; Xu, Zeng Yu

    1996-01-01

    The Vanadium/Lithium system has been the recent focus of ANL's Blanket Technology Pro-ram, and for the last several years, ANL's Liquid Metal Blanket activities have been carried out in direct support of the ITER (International Thermonuclear Experimental Reactor) breeding blanket task area. A key feasibility issue for the ITER Vanadium/Lithium breeding blanket is the Near the development of insulator coatings. Design calculations, Hua and Gohar, show that an electrically insulating layer is necessary to maintain an acceptably low magneto-hydrodynamic (MHD) pressure drop in the current ITER design. Consequently, the decision was made to convert Argonne's Liquid Metal EXperiment (ALEX) from a 200 degrees C NaK facility to a 350 degrees C lithium facility. The upgraded facility was designed to produce MHD pressure drop data, test section voltage distributions, and heat transfer data for mid-scale test sections and blanket mockups at Hartmann numbers (M) and interaction parameters (N) in the range of 10 3 to 10 5 in lithium at 350 degrees C. Following completion of the upgrade work, a short performance test was conducted, followed by two longer multiple-hour, MHD tests, all at 230 degrees C. The modified ALEX facility performed up to expectations in the testing. MHD pressure drop and test section voltage distributions were collected at Hartmann numbers of 1000

  4. ac power control in the Core Flow Test Loop

    International Nuclear Information System (INIS)

    McDonald, D.W.

    1980-01-01

    This work represents a status report on a development effort to design an ac power controller for the Core Flow Test Loop. The Core Flow Test Loop will be an engineering test facility which will simulate the thermal environment of a gas-cooled fast-breeder reactor. The problems and limitations of using sinusoidal ac power to simulate the power generated within a nuclear reactor are addressed. The transformer-thyristor configuration chosen for the Core Flow Test Loop power supply is presented. The initial considerations, design, and analysis of a closed-loop controller prototype are detailed. The design is then analyzed for improved performance possibilities and failure modes are investigated at length. A summary of the work completed to date and a proposed outline for continued development completes the report

  5. Ballistic Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Ballistic Test Facility is comprised of two outdoor and one indoor test ranges, which are all instrumented for data acquisition and analysis. Full-size aircraft...

  6. Corrosion Testing Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Corrosion Testing Facility is part of the Army Corrosion Office (ACO). It is a fully functional atmospheric exposure site, called the Corrosion Instrumented Test...

  7. Environmental Test Facility (ETF)

    Data.gov (United States)

    Federal Laboratory Consortium — The Environmental Test Facility (ETF) provides non-isolated shock testing for stand-alone equipment and full size cabinets under MIL-S-901D specifications. The ETF...

  8. Climatic Environmental Test Facilities

    Data.gov (United States)

    Federal Laboratory Consortium — RTTC has an extensive suite of facilities for supporting MIL-STD-810 testing, toinclude: Temperature/Altitude, Rapid Decompression, Low/High Temperature,Temperature...

  9. Wind Tunnel Testing Facilities

    Data.gov (United States)

    Federal Laboratory Consortium — NASA Ames Research Center is pleased to offer the services of our premier wind tunnel facilities that have a broad range of proven testing capabilities to customers...

  10. Accident analysis of HANARO fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. Y.; Chi, D. Y

    1998-03-01

    Steady state fuel test loop will be equipped in HANARO to obtain the development and betterment of advanced fuel and materials through the irradiation tests. The HANARO fuel test loop was designed to match the CANDU and PWR fuel operating conditions. The accident analysis was performed by RELAP5/MOD3 code based on FTL system designs and determined the detail engineering specification of in-pile test section and out-pile systems. The accident analysis results of FTL system could be used for the fuel and materials designer to plan the irradiation testing programs. (author). 23 refs., 20 tabs., 178 figs.

  11. Toroid magnet test facility

    CERN Multimedia

    2002-01-01

    Because of its exceptional size, it was not feasible to assemble and test the Barrel Toroid - made of eight coils - as an integrated toroid on the surface, prior to its final installation underground in LHC interaction point 1. It was therefore decided to test these eight coils individually in a dedicated test facility.

  12. Test and User Facilities | NREL

    Science.gov (United States)

    Test and User Facilities Test and User Facilities Our test and user facilities are available to | L | M | N | O | P | Q | R | S | T | U | V | W | X | Y | Z B Battery Thermal and Life Test Facility Biochemical Conversion Pilot Plant C Controllable Grid Interface Test System D Dynamometer Test Facilities

  13. Testing experience with fast flux test facility

    International Nuclear Information System (INIS)

    Noordhoff, B.H.; McGough, C.B.; Nolan, J.E.

    1975-01-01

    Early FFTF project planning emphasized partial and full-scale testing of major reactor and plant prototype components under expected environmental conditions, excluding radiation fields. Confirmation of component performance during FFTF service was considered essential before actual FFTF startup, to provide increased assurance against FFTF startup delays or operational difficulties and downtime. Several new sodium facilities were constructed, and confirmation tests on the prototype components are now in progress. Test conditions and results to date are reported for the primary pump, intermediate heat exchanger, sodium-to-air dump heat exchanger, large and small sodium valves, purification cold trap, in-vessel handling machine, instrument tree, core restraint, control rod system, low-level flux monitor, closed loop ex-vessel machine, refueling equipment, and selected maintenance equipment. The significance and contribution of these tests to the FFTF and Liquid Metal Fast Breeder Reactor (LMFBR) program are summarized. (U.S.)

  14. Liquid Lead-Bismuth Materials Test Loop

    International Nuclear Information System (INIS)

    Tcharnotskaia, Valentina; Ammerman, Curtt; Darling, Timothy; King, Joe; Li, Ning; Shaw, Don; Snodgrass, Leon; Woloshun, Keith

    2002-01-01

    We designed and built the Liquid Lead-Bismuth Materials Test Loop (MTL) to study the materials behavior in a flow of molten lead-bismuth eutectic (LBE). In this paper we present a description of the loop with main components and their functions. Stress distribution in the piping due to sustained, occasional and expansion loads is shown. The loop is designed so that a difference of 100 deg. C can be attained between the coldest and the hottest parts at a nominal flow rate of 8.84 GPM. Liquid LBE flow can be activated by a mechanical sump pump or by natural convection. In order to maintain a self-healing protective film on the surface of the stainless steel pipe, a certain concentration of oxygen has to be maintained in the liquid metal. We developed oxygen sensors and an oxygen control system to be implemented in the loop. The loop is outfitted with a variety of instruments that are controlled from a computer based data acquisition system. Initial experiments include preconditioning the loop, filling it up with LBE, running at uniform temperature and tuning the oxygen control system. We will present some preliminary results and discuss plans for the future tests. (authors)

  15. The engineering test facility

    International Nuclear Information System (INIS)

    Steiner, D.; Becraft, W.R.; Sager, P.H.

    1981-01-01

    The vehicle by which the fusion program would move into the engineering testing phase of fusion power development is designated the Engineering Test Facility (ETF). The ETF would provide a test-bed for reactor components in the fusion environment. In order to initiate preliminary planning for the ETF decision, the Office of Fusion Energy established the ETF Design Center activity to prepare the design of the ETF. This paper describes the design status of the ETF. (orig.)

  16. Engineering test facility

    International Nuclear Information System (INIS)

    Steiner, D.; Becraft, W.R.; Sager, P.H.

    1981-01-01

    The vehicle by which the fusion program would move into the engineering testing phase of fusion power development is designated the Engineering Test Facility (ETF). The ETF would provide a test-bed for reactor components in the fusion environment. In order to initiate preliminary planning for the ETF decision, the Office of Fusion Energy established the ETF Design Center activity to prepare the design of the ETF. This paper described the design status of the ETF

  17. Distributed Energy Resources Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — NREL's Distributed Energy Resources Test Facility (DERTF) is a working laboratory for interconnection and systems integration testing. This state-of-the-art facility...

  18. Thermal-hydraulic analyses for in-pile SCWR fuel qualification test loops and SCWR material loop

    Energy Technology Data Exchange (ETDEWEB)

    Vojacek, A.; Mazzini, G.; Zmitkova, J.; Ruzickova, M. [Research Centre Rez (Czech Republic)

    2014-07-01

    One of the R&D directions of Research Centre Rez is dedicated to the supercritical water-cooled reactor concept (SCWR). Among the developed experimental facilities and infrastructure in the framework of the SUSEN project (SUStainable ENergy) is construction and experimental operation of the supercritical water loop SCWL focusing on material tests. At the first phase, this SCWL loop is assembled and operated out-of-pile in the dedicated loop facilities hall. At this out-of-pile operation various operational conditions are tested and verified. After that, in the second phase, the SCWL loop will be situated in-pile, in the core of the research reactor LVR-15, operated at CVR. Furthermore, it is planned to carry out a test of a small scale fuel assembly within the SuperCritical Water Reactor Fuel Qualification Test (SCWR-FQT) loop, which is now being designed. This paper presents the results of the thermal-hydraulic analyses of SCWL loop out-of-pile operation using the RELAP5/MOD3.3. The thermal-hydraulic modeling and the performed analyses are focused on the SCWL loop model validation through a comparison of the calculation results with the experimental results obtained at various operation conditions. Further, the present paper focuses on the transient analyses for start-up and shut-down of the FQT loop, particularly to explore the ability of system codes ATHLET 3.0A to simulate the transient between subcritical conditions and supercritical conditions. (author)

  19. Feasibility study on the transient fuel test loop installation

    International Nuclear Information System (INIS)

    Kim, J. Y.; Lee, C. Y.

    1997-02-01

    The design and installation of the irradiation test facility for verification test of the fuel performance are very important in connection with maximization of the utilization of HANARO. The objective of this study is to investigate and analyze the test capsules and loops in research reactors of the other countries and to design preliminarily the eligible transient fuel test facility to be installed in HANARO. The principle subjects of this study are to analyze the contents, kinds and scopes of the irradiation test facilities for nuclear technology development. The guidances for the basic and detail design of the transient fuel test facility in the future are presented. The investigation and analysis of various kinds of test facilities that are now in operation at the research reactors of nuclear advanced countries are carried out. Based on the design data of HANARO the design materials for an eligible transient fuel test facility comprises two pacts : namely, in pile test fuel in reactor core site, and out of pile system regulates the experimental conditions in the in pile test section. Especially for power ramping and cycling selection of the eligible power variation equipment in HANARO is carried out. (author). 13 refs., 4 tabs., 46 figs

  20. Large coil test facility

    International Nuclear Information System (INIS)

    Nelms, L.W.; Thompson, P.B.

    1980-01-01

    Final design of the facility is nearing completion, and 20% of the construction has been accomplished. A large vacuum chamber, houses the test assembly which is coupled to appropriate cryogenic, electrical, instrumentation, diagnostc systems. Adequate assembly/disassembly areas, shop space, test control center, offices, and test support laboratories are located in the same building. Assembly and installation operations are accomplished with an overhead crane. The major subsystems are the vacuum system, the test stand assembly, the cryogenic system, the experimental electric power system, the instrumentation and control system, and the data aquisition system

  1. Detection Test for Leakage of CO2 into Sodium Loop

    International Nuclear Information System (INIS)

    Park, Sun Hee; Wi, Myung-Hwan; Min, Jae Hong

    2015-01-01

    This report is about the facility for the detection test for leakage of CO 2 into sodium loop. The facility for the detection test for leakage of CO 2 into sodium loop was introduced. The test will be carried out. Our experimental results are going to be expected to be used for approach methods to detect CO 2 leaking into sodium in heat exchangers. A sodium-and-carbon dioxide (Na-CO 2 ) heat exchanger is one of the key components for the supercritical CO 2 Brayton cycle power conversion system of sodium-cooled fast reactors (SFRs). A printed circuit heat exchanger (PCHE) is considered for the Na-CO 2 heat exchanger, which is known to have potential for reducing the volume occupied by the exchangers compared to traditional shell-and-tube heat exchangers. Among various issues about the Na- CO 2 exchanger, detection of CO 2 leaking into sodium in the heat exchanger is most important thing for its safe operation. It is known that reaction products from sodium and CO 2 such as sodium carbonate (Na 2 CO 3 ) and amorphous carbon are hardly soluble in sodium, which cause plug sodium channels. Detection technique for Na 2 CO 3 in sodium loop has not been developed yet. Therefore, detection of CO 2 and CO from reaction of sodium and CO 2 are proper to detect CO 2 leakage into sodium loop

  2. National Solar Thermal Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The National Solar Thermal Test Facility (NSTTF) is the only test facility in the United States of its type. This unique facility provides experimental engineering...

  3. Hot Hydrogen Test Facility

    International Nuclear Information System (INIS)

    W. David Swank

    2007-01-01

    The core in a nuclear thermal rocket will operate at high temperatures and in hydrogen. One of the important parameters in evaluating the performance of a nuclear thermal rocket is specific impulse, ISp. This quantity is proportional to the square root of the propellant's absolute temperature and inversely proportional to square root of its molecular weight. Therefore, high temperature hydrogen is a favored propellant of nuclear thermal rocket designers. Previous work has shown that one of the life-limiting phenomena for thermal rocket nuclear cores is mass loss of fuel to flowing hydrogen at high temperatures. The hot hydrogen test facility located at the Idaho National Lab (INL) is designed to test suitability of different core materials in 2500 C hydrogen flowing at 1500 liters per minute. The facility is intended to test non-uranium containing materials and therefore is particularly suited for testing potential cladding and coating materials. In this first installment the facility is described. Automated Data acquisition, flow and temperature control, vessel compatibility with various core geometries and overall capabilities are discussed

  4. French nuclear plant safeguard pump qualification testing: EPEC test loop

    International Nuclear Information System (INIS)

    Guesnon, H.

    1985-01-01

    This paper reviews the specifications to which nuclear power plant safeguard pumps must be qualified, and surveys the qualification methods and program used in France to verify operability of the pump assembly and major pump components. The EPEC test loop is described along with loop capabilities and acheivements up to now. This paper shows, through an example, the Medium Pressure Safety Injection Pump designed for service in 1300 MW nuclear power plants, and the interesting possibilities offered by qualification testing

  5. Universal Test Facility

    Science.gov (United States)

    Laughery, Mike

    A universal test facility (UTF) for Space Station Freedom is developed. In this context, universal means that the experimental rack design must be: automated, highly marketable, and able to perform diverse microgravity experiments according to NASA space station requirements. In order to fulfill these broad objectives, the facility's customers, and their respective requirements, are first defined. From these definitions, specific design goals and the scope of the first phase of this project are determined. An examination is first made into what types of research are most likely to make the UTF marketable. Based on our findings, the experiments for which the UTF would most likely be used included: protein crystal growth, hydroponics food growth, gas combustion, gallium arsenide crystal growth, microorganism development, and cell encapsulation. Therefore, the UTF is designed to fulfill all of the major requirements for the experiments listed above. The versatility of the design is achieved by taking advantage of the many overlapping requirements presented by these experiments.

  6. Mechanisms Engineering Test Loop - Phase 1 Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Kultgen, D. [Argonne National Lab. (ANL), Argonne, IL (United States); Grandy, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Hvasta, M. [Argonne National Lab. (ANL), Argonne, IL (United States); Lisowski, D. [Argonne National Lab. (ANL), Argonne, IL (United States); Toter, W. [Argonne National Lab. (ANL), Argonne, IL (United States); Borowski, A. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-09-01

    This report documents the current status of the Mechanisms Engineering Test Loop (METL) as of the end of FY2016. Currently, METL is in Phase I of its design and construction. Once operational, the METL facility will test small to intermediate-scale components and systems in order to develop advanced liquid metal technologies. Testing different components in METL is essential for the future of advanced fast reactors as it will provide invaluable performance data and reduce the risk of failures during plant operation.

  7. Optimal tests for electroweak loop effects

    International Nuclear Information System (INIS)

    Aoki, Kenichi; Aoyama, Hideaki; Harvard Univ., Cambridge, MA

    1986-01-01

    A statistical analysis is given for the experimental precision necessary for establishing loop effects in the electroweak theory. Cases with three observables, gauge boson masses and the Weinberg angle, is analyzed by an optimised test. An information on the Weinberg angle with even 5% error (+-.01 in sin 2 thetasub(W)) is shown to reduce the requirement for the measurements of gauge boson masses significantly. (orig.)

  8. TESLA Test Facility. Status

    International Nuclear Information System (INIS)

    Aune, B.

    1996-01-01

    The TESLA Test Facility (TTF), under construction at DESY by an international collaboration, is an R and D test bed for the superconducting option for future linear e+/e-colliders. It consists of an infrastructure to process and test the cavities and of a 500 MeV linac. The infrastructure has been installed and is fully operational. It includes a complex of clean rooms, an ultra-clean water plant, a chemical etching installation and an ultra-high vacuum furnace. The linac will consist of four cryo-modules, each containing eight 1 meter long nine-cell cavities operated at 1.3 GHz. The base accelerating field is 15 MV/m. A first injector will deliver a low charge per bunch beam, with the full average current (8 mA in pulses of 800 μs). A more powerful injector based on RF gun technology will ultimately deliver a beam with high charge and low emittance to allow measurements necessary to qualify the TESLA option and to demonstrate the possibility of operating a free electron laser based on the Self-Amplified-Spontaneous-Emission principle. Overview and status of the facility will be given. Plans for the future use of the linac are presented. (R.P.)

  9. CLIC Test Facility 3

    CERN Multimedia

    Kossyvakis, I; Faus-golfe, A

    2007-01-01

    The design of CLIC is based on a two-beam scheme, where short pulses of high power 30 GHz RF are extracted from a drive beam running parallel to the main beam. The 3rd generation CLIC Test Facility (CTF3) will demonstrate the generation of the drive beam with the appropriate time structure, the extraction of 30 GHz RF power from this beam, as well as acceleration of a probe beam with 30 GHz RF cavities. The project makes maximum use of existing equipment and infrastructure of the LPI complex, which became available after the closure of LEP.

  10. High Temperature Fluoride Salt Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    Aaron, Adam M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Cunningham, Richard Burns [Univ. of Tennessee, Knoxville, TN (United States); Fugate, David L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Holcomb, David Eugene [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Kisner, Roger A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Peretz, Fred J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wilson, Dane F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yoder, Jr, Graydon L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-12-01

    with 3 cm diameter graphite-based fuel pebbles slowly circulating up through the core. Molten salt coolant (FLiBe) at 700°C flows concurrently (at significantly higher velocity) with the pebbles and is used to remove heat generated in the reactor core (approximately 1280 W/pebble), and supply it to a power conversion system. Refueling equipment continuously sorts spent fuel pebbles and replaces spent or damaged pebbles with fresh fuel. By combining greater or fewer numbers of pebble channel assemblies, multiple reactor designs with varying power levels can be offered. The PB-AHTR design is discussed in detail in Reference [1] and is shown schematically in Fig. 1. Fig. 1. PB-AHTR concept (drawing taken from Peterson et al., Design and Development of the Modular PB-AHTR Proceedings of ICApp 08). Pebble behavior within the core is a key issue in proving the viability of this concept. This includes understanding the behavior of the pebbles thermally, hydraulically, and mechanically (quantifying pebble wear characteristics, flow channel wear, etc). The experiment being developed is an initial step in characterizing the pebble behavior under realistic PB-AHTR operating conditions. It focuses on thermal and hydraulic behavior of a static pebble bed using a convective salt loop to provide prototypic fluid conditions to the bed, and a unique inductive heating technique to provide prototypic heating in the pebbles. The facility design is sufficiently versatile to allow a variety of other experimentation to be performed in the future. The facility can accommodate testing of scaled reactor components or sub-components such as flow diodes, salt-to-salt heat exchangers, and improved pump designs as well as testing of refueling equipment, high temperature instrumentation, and other reactor core designs.

  11. Tritium Systems Test Facility

    International Nuclear Information System (INIS)

    Cafasso, F.A.; Maroni, V.A.; Smith, W.H.; Wilkes, W.R.; Wittenberg, L.J.

    1978-01-01

    This TSTF proposal has two principal objectives. The first objective is to provide by mid-FY 1981 a demonstration of the fuel cycle and tritium containment systems which could be used in a Tokamak Experimental Power Reactor for operation in the mid-1980's. The second objective is to provide a capability for further optimization of tritium fuel cycle and environmental control systems beyond that which is required for the EPR. The scale and flow rates in TSTF are close to those which have been projected for a prototype experimental power reactor (PEPR/ITR) and will permit reliable extrapolation to the conditions found in an EPR. The fuel concentrations will be the same as in an EPR. Demonstrations of individual components of the deuterium-tritium fuel cycle and of monitoring, accountability and containment systems and of a maintenance methodology will be achieved at various times in the FY 1979-80 time span. Subsequent to the individual component demonstrations--which will proceed from tests with hydrogen (and/or deuterium) through tracer levels of tritium to full operational concentrations--a complete test and demonstration of the integrated fuel processing and tritium containment facility will be performed. This will occur near the middle of FY 1981. Two options were considered for the TSTF: (1) The modification of an existing building and (2) the construction of a new facility

  12. Development of 3-Pin Fuel Test Loop and Utilization Technology

    International Nuclear Information System (INIS)

    Lee, Chung Young; Sim, B. S.; Lee, C. Y.

    2007-06-01

    The principal contents of this project are to design, fabricate and install the steady-state fuel test loop in HANARO for nuclear technology development. Procurement and, fabrication of main equipment, licensing and installation for fuel test loop have been performed. Following contents are described in the report. 1. Design - Design of the In-pile system and Out pile system 2. Fabrication and procurement of the equipment - Fabrication of the In-pile system and In-pool piping - Fabrication and procurement of the equipment of the out-pile system 3. Acquisition of the license - Preparation of the safety analysis report and acquisition of the license - Pre-service inspection of the facility 4. Installation and commissioning - Installation of the FTL - Development of the commissioning procedure

  13. SNS Target Test Facility for remote handling design and verification

    International Nuclear Information System (INIS)

    Spampinato, P.T.; Graves, V.B.; Schrock, S.L.

    1998-01-01

    The Target Test Facility will be a full-scale prototype of the Spallation Neutron Source Target Station. It will be used to demonstrate remote handling operations on various components of the mercury flow loop and for thermal/hydraulic testing. This paper describes the remote handling aspects of the Target Test Facility. Since the facility will contain approximately 1 cubic meter of mercury for the thermal/hydraulic tests, an enclosure will also be constructed that matches the actual Target Test Cell

  14. Gas Test Loop Functional and Technical Requirements

    International Nuclear Information System (INIS)

    Glen R. Longhurst; Soli T. Khericha; James L. Jones

    2004-01-01

    This document defines the technical and functional requirements for a gas test loop (GTL) to be constructed for the purpose of providing a high intensity fast-flux irradiation environment for developers of advanced concept nuclear reactors. This capability is needed to meet fuels and materials testing requirements of the designers of Generation IV (GEN IV) reactors and other programs within the purview of the Advanced Fuel Cycle Initiative (AFCI). Space nuclear power development programs may also benefit by the services the GTL will offer. The overall GTL technical objective is to provide developers with the means for investigating and qualifying fuels and materials needed for advanced reactor concepts. The testing environment includes a fast-flux neutron spectrum of sufficient intensity to perform accelerated irradiation testing. Appropriate irradiation temperature, gaseous environment, test volume, diagnostics, and access and handling features are also needed. This document serves to identify those requirements as well as generic requirements applicable to any system of this kind

  15. Gas Test Loop Functional and Technical Requirements

    Energy Technology Data Exchange (ETDEWEB)

    Glen R. Longhurst; Soli T. Khericha; James L. Jones

    2004-09-01

    This document defines the technical and functional requirements for a gas test loop (GTL) to be constructed for the purpose of providing a high intensity fast-flux irradiation environment for developers of advanced concept nuclear reactors. This capability is needed to meet fuels and materials testing requirements of the designers of Generation IV (GEN IV) reactors and other programs within the purview of the Advanced Fuel Cycle Initiative (AFCI). Space nuclear power development programs may also benefit by the services the GTL will offer. The overall GTL technical objective is to provide developers with the means for investigating and qualifying fuels and materials needed for advanced reactor concepts. The testing environment includes a fast-flux neutron spectrum of sufficient intensity to perform accelerated irradiation testing. Appropriate irradiation temperature, gaseous environment, test volume, diagnostics, and access and handling features are also needed. This document serves to identify those requirements as well as generic requirements applicable to any system of this kind.

  16. New in-pile water loop facility for IASCC studies at JMTR

    International Nuclear Information System (INIS)

    Tsukada, T.; Tsuji, H.; Nakajima, H.; Komori, Y.; Ito, H.

    2002-01-01

    Irradiation assisted stress corrosion cracking (IASCC) is caused by the synergistic effects of neutron and gamma radiation, residual and applied stresses and high temperature water environment on the structural materials of vessel internals. IASCC has been studied since the beginning of the 1980's and the phenomenological knowledge on IASCC is accrued extensively. However, mainly due to the experimental difficulties, data for the mechanistic understanding and prediction of failures of the specific in-vessel components are still insufficient and further well-controlled experiments are needed [1]. In recent years, efforts to perform the in-pile materials test for IASCC study have been made at some research reactors [2-4]. At JAERI, a high temperature water loop facility was designed to install at the Japan Materials Testing Reactor (JMTR) to carry out the in-core IASCC testing. This report describes an overview of design and specification of the loop facility. (authors)

  17. Hanford Tank Farms Waste Certification Flow Loop Test Plan

    Energy Technology Data Exchange (ETDEWEB)

    Bamberger, Judith A.; Meyer, Perry A.; Scott, Paul A.; Adkins, Harold E.; Wells, Beric E.; Blanchard, Jeremy; Denslow, Kayte M.; Greenwood, Margaret S.; Morgen, Gerald P.; Burns, Carolyn A.; Bontha, Jagannadha R.

    2010-01-01

    A future requirement of Hanford Tank Farm operations will involve transfer of wastes from double shell tanks to the Waste Treatment Plant. As the U.S. Department of Energy contractor for Tank Farm Operations, Washington River Protection Solutions anticipates the need to certify that waste transfers comply with contractual requirements. This test plan describes the approach for evaluating several instruments that have potential to detect the onset of flow stratification and critical suspension velocity. The testing will be conducted in an existing pipe loop in Pacific Northwest National Laboratory’s facility that is being modified to accommodate the testing of instruments over a range of simulated waste properties and flow conditions. The testing phases, test matrix and types of simulants needed and the range of testing conditions required to evaluate the instruments are described

  18. Eccentric Coil Test Facility (ECTF)

    International Nuclear Information System (INIS)

    Burn, P.B.; Walstrom, P.L.; Anderson, W.C.; Marguerat, E.F.

    1975-01-01

    The conceptual design of a facility for testing superconducting coils under some conditions peculiar to tokamak systems is given. A primary element of the proposed facility is a large 25 MJ background solenoid. Discussions of the mechanical structure, the stress distribution and the thermal stability for this coil are included. The systems for controlling the facility and diagnosing test coil behavior are also described

  19. Arc Heated Scramjet Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Arc Heated Scramjet Test Facility is an arc heated facility which simulates the true enthalpy of flight over the Mach number range of about 4.7 to 8 for free-jet...

  20. Summary of ALSEP Test Loop Solvent Irradiation Testing

    Energy Technology Data Exchange (ETDEWEB)

    Peterman, Dean Richard [Idaho National Lab. (INL), Idaho Falls, ID (United States); Olson, Lonnie Gene [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-08-01

    Separating the minor actinide elements (americium and curium) from the fission product lanthanides is an important step in closing the nuclear fuel cycle. Isolating the minor actinides will allow transmuting them to short lived or stable isotopes in fast reactors, thereby reducing the long-term hazard associated with these elements. The Actinide Lanthanide Separation Process (ALSEP) is being developed by the DOE-NE Material Recovery and Waste Form Development Campaign to accomplish this separation with a single process. To develop a fundamental understanding of the solvent degradation mechanisms for the ALSEP Process, testing was performed in the INL Radiolysis/Hydrolysis Test Loop for the extraction section of the ALSEP flowsheet. This work culminated in the completion of the level two milestone (M2FT-16IN030102021) "Complete ALSEP test loop solvent irradiation test.” This report summarizes the testing performed and the impact of radiation on the ALSEP Process performance as a function of dose.

  1. Development of Start-up and Shutdown Procedure for the HANARO Fuel Test Loop

    International Nuclear Information System (INIS)

    Park, S. K.; Sim, B. S.; Chi, D. Y.; Lee, J. M.; Lee, C. Y.; Ahn, S. H.

    2009-06-01

    A start-up and shutdown procedure for the HANARO fuel test loop has been developed. This is a facility for fuel and material irradiation tests. The facility provides experimental conditions similar to the normal operational pressures and temperatures of commercial PWR and CANDU plants. The normal operation modes of the HANARO fuel test loop are classified into loop shutdown, cold stand-by 1, cold stand-by 2, hot stand-by, and hot operation. The operation modes depend on the fission power of test fuels and the coolant temperature at the inlet of the in-pile test section. The HANARO must maintain a shutdown mode if the HANARO fuel test loop is loop shutdown, cold stand-by 1, cold stand-by 2, or hot stand-by. As the HANARO becomes power operation mode, the operation mode of the HANARO fuel test loop comes to hot operation from hot stand-by. The procedure for the HANARO fuel test loop consists of four main parts such as check of initial conditions, stat-up operation procedure, shutdown operation procedure, and check lists for operations. Several hot test operations ensure that the procedure is appropriate

  2. SDG and E - ERDA Geothermal Loop Experimental Facility. Bi-monthly report, May 1975-August 1976

    Energy Technology Data Exchange (ETDEWEB)

    Bishop, H.K.; Bricarello, J.R.; Campbell, J.A.; Lombard, G.L.; Mulliner, D.K.; Swanson, C.R.

    1976-09-01

    A description of the Geothermal Loop Experimental Facility (GLEF) its construction problems, and a discussion of start-up testing are included. A history and description of the operation and maintenance with the brine injection pump for the facility are presented. The GLEF was divided into five separate sections: steam and condensate system, brine system, purge water system, vent gas system, and cooling water system. An insight into the chemistry of each system is provided by analysis of samples taken. Scaling and corrosion effects of brine, steam, gas, and water in these systems are described in detail. (MHR)

  3. Molten Salt Test Loop (MSTL) system customer interface document.

    Energy Technology Data Exchange (ETDEWEB)

    Gill, David Dennis; Kolb, William J.; Briggs, Ronald D.

    2013-09-01

    The National Solar Thermal Test Facility at Sandia National Laboratories has a unique test capability called the Molten Salt Test Loop (MSTL) system. MSTL is a test capability that allows customers and researchers to test components in flowing, molten nitrate salt. The components tested can range from materials samples, to individual components such as flex hoses, ball joints, and valves, up to full solar collecting systems such as central receiver panels, parabolic troughs, or linear Fresnel systems. MSTL provides realistic conditions similar to a portion of a concentrating solar power facility. The facility currently uses 60/40 nitrate %E2%80%9Csolar salt%E2%80%9D and can circulate the salt at pressure up to 40 bar (600psi), temperature to 585%C2%B0C, and flow rate of 44-50kg/s(400-600GPM) depending on temperature. The purpose of this document is to provide a basis for customers to evaluate the applicability to their testing needs, and to provide an outline of expectations for conducting testing on MSTL. The document can serve as the basis for testing agreements including Work for Others (WFO) and Cooperative Research and Development Agreements (CRADA). While this document provides the basis for these agreements and describes some of the requirements for testing using MSTL and on the site at Sandia, the document is not sufficient by itself as a test agreement. The document, however, does provide customers with a uniform set of information to begin the test planning process.

  4. Electromagnetic Interface Testing Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Electromagnetic Interface Testing facilitysupports such testing asEmissions, Field Strength, Mode Stirring, EMP Pulser, 4 Probe Monitoring/Leveling System, and...

  5. RIA testing capability of the transient reactor test facility

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  6. Analysis of severe accidents on fast reactor test loop

    International Nuclear Information System (INIS)

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

    1975-01-01

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

  7. Characteristics of pressure control system on PWR/PHWR in pile loop facility

    International Nuclear Information System (INIS)

    Sarwani; Hendro, P.; Suwoto; Sutrisno

    1998-01-01

    PWR/PHWR in-pile loop facility is used for testing of fuel element bundle which is correspond to the condition of power reactor operation. So, this facility is designed at 150 bar of pressure and 350 o C of temperature. Pressure control system is one of the components of the facility and it is equipped with 6 electrical heaters (30 KW), water spray, pressure and temperature monitors. The characterization test of pressure control system has been carried out with operating of 2 electrical heaters (10 KW). The K eff calculation value is different 5.2% from pressure in the pressure control system can be increased to 160 bar within 27 hours. After the system pressure reached the nominal pressure, the pressure control system was in the steady state condition

  8. EMI Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — FUNCTION: Supports electromagnetic interference/radio frequency interference (EMI/RFI) testing of flight hardware. It is also used to support custom RF testing up to...

  9. Static Loads Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — FUNCTION: Provides the capability to perform large-scale structural loads testing on spacecraft and other structures. Results from these tests can be used to verify...

  10. Multi-loop PWR modeling and hardware-in-the-loop testing using ACSL

    International Nuclear Information System (INIS)

    Thomas, V.M.; Heibel, M.D.; Catullo, W.J.

    1989-01-01

    Westinghouse has developed an Advanced Digital Feedwater Control System (ADFCS) which is aimed at reducing feedwater related reactor trips through improved control performance for pressurized water reactor (PWR) power plants. To support control system setpoint studies and functional design efforts for the ADFCS, an ACSL based model of the nuclear steam supply system (NSSS) of a Westinghouse (PWR) was generated. Use of this plant model has been extended from system design to system testing through integration of the model into a Hardware-in-Loop test environment for the ADFCS. This integration includes appropriate interfacing between a Gould SEL 32/87 computer, upon which the plant model executes in real time, and the Westinghouse Distributed Processing family (WDPF) test hardware. A development program has been undertaken to expand the existing ACSL model to include capability to explicitly model multiple plant loops, steam generators, and corresponding feedwater systems. Furthermore, the program expands the ADFCS Hardware-in-Loop testing to include the multi-loop plant model. This paper provides an overview of the testing approach utilized for the ADFCS with focus on the role of Hardware-in-Loop testing. Background on the plant model, methodology and test environment is also provided. Finally, an overview is presented of the program to expand the model and associated Hardware-in-Loop test environment to handle multiple loops

  11. Solenoid Testing Facility

    Data.gov (United States)

    Federal Laboratory Consortium — Current Configuration: Accommodate a device under test up to 2.8 m diameter, 0.7 m height and 15,000 lbs. weight. Up to 10 g/s, 4.5 K helium flow. Up to 250 A test...

  12. The Integral Test Facility Karlstein

    Directory of Open Access Journals (Sweden)

    Stephan Leyer

    2012-01-01

    Full Text Available The Integral Test Facility Karlstein (INKA test facility was designed and erected to test the performance of the passive safety systems of KERENA, the new AREVA Boiling Water Reactor design. The experimental program included single component/system tests of the Emergency Condenser, the Containment Cooling Condenser and the Passive Core Flooding System. Integral system tests, including also the Passive Pressure Pulse Transmitter, will be performed to simulate transients and Loss of Coolant Accident scenarios at the test facility. The INKA test facility represents the KERENA Containment with a volume scaling of 1 : 24. Component heights and levels are in full scale. The reactor pressure vessel is simulated by the accumulator vessel of the large valve test facility of Karlstein—a vessel with a design pressure of 11 MPa and a storage capacity of 125 m3. The vessel is fed by a benson boiler with a maximum power supply of 22 MW. The INKA multi compartment pressure suppression Containment meets the requirements of modern and existing BWR designs. As a result of the large power supply at the facility, INKA is capable of simulating various accident scenarios, including a full train of passive systems, starting with the initiating event—for example pipe rupture.

  13. Airborne Test Bed Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Laboratory operates the main hangar on the Hanscom Air Force Base flight line. This very large building (~93,000sqft) accommodates the Laboratory's airborne test...

  14. 33-GVA interrupter test facility

    International Nuclear Information System (INIS)

    Parsons, W.M.; Honig, E.M.; Warren, R.W.

    1979-01-01

    The use of commercial ac circuit breakers for dc switching operations requires that they be evaluated to determine their dc limitations. Two 2.4-GVA facilities have been constructed and used for this purpose at LASL during the last several years. In response to the increased demand on switching technology, a 33-GVA facility has been constructed. Novel features incorporated into this facility include (1) separate capacitive and cryogenic inductive energy storage systems, (2) fiber-optic controls and optically-coupled data links, and (3) digital data acquisition systems. Facility details and planned tests on an experimental rod-array vacuum interrupter are presented

  15. Reverberant Acoustic Test Facility (RATF)

    Data.gov (United States)

    Federal Laboratory Consortium — The very large Reverberant Acoustic Test Facility (RATF) at the NASA Glenn Research Center (GRC), Plum Brook Station, is currently under construction and is due to...

  16. Elevated Fixed Platform Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Elevated Fixed Platform (EFP) is a helicopter recovery test facility located at Lakehurst, NJ. It consists of a 60 by 85 foot steel and concrete deck built atop...

  17. LLNL superconducting magnets test facility

    Energy Technology Data Exchange (ETDEWEB)

    Manahan, R; Martovetsky, N; Moller, J; Zbasnik, J

    1999-09-16

    The FENIX facility at Lawrence Livermore National Laboratory was upgraded and refurbished in 1996-1998 for testing CICC superconducting magnets. The FENIX facility was used for superconducting high current, short sample tests for fusion programs in the late 1980s--early 1990s. The new facility includes a 4-m diameter vacuum vessel, two refrigerators, a 40 kA, 42 V computer controlled power supply, a new switchyard with a dump resistor, a new helium distribution valve box, several sets of power leads, data acquisition system and other auxiliary systems, which provide a lot of flexibility in testing of a wide variety of superconducting magnets in a wide range of parameters. The detailed parameters and capabilities of this test facility and its systems are described in the paper.

  18. Freshwater Treatment and Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Freshwater Treatment and Test Facility, located at SANGB, has direct year-round access to water from Lake St. Clair and has a State of Michigan approved National...

  19. Mirror fusion test facility

    International Nuclear Information System (INIS)

    Post, R.F.

    1978-01-01

    The MFTF is a large new mirror facility under construction at Livermore for completion in 1981--82. It represents a scaleup, by a factor of 50 in plasma volume, a factor of 5 or more in ion energy, and a factor of 4 in magnetic field intensity over the Livermore 2XIIB experiment. Its magnet, employing superconducting NbTi windings, is of Yin-Yang form and will weigh 200 tons. MFTF will be driven by neutral beams of two levels of current and energy: 1000 amperes of 20 keV (accelerating potential) pulsed beams for plasma startup; 750 amperes of 80 keV beams of 0.5 second duration for temperature buildup and plasma sustainment. Two operating modes for MFTF are envisaged: The first is operation as a conventional mirror cell with n/sup tau/ approximately equal to 10 12 cm -3 sec, W/sub i/ = 50 keV, where the emphasis will be on studying the physics of mirror cells, particularly the issues of improved techniques of stabilization against ion cyclotron modes and of maximization of the electron temperature. The second possible mode is the further study of the Field Reversed Mirror idea, using high current neutral beams to sustain the field-reversed state. Anticipating success in the coming Livermore Tandem Mirror Experiment (TMX) MFTF has been oriented so that it could comprise one end cell of a scaled up TM experiment. Also, if MFTF were to succeed in achieving a FR state it could serve as an essentially full-sized physics prototype of one cell of a FRM fusion power plant

  20. Supercritical CO2 test loop operation and first test results

    International Nuclear Information System (INIS)

    Wright, Steven A.; Pickard, Paul S.

    2009-01-01

    The DOE Office of Nuclear Energy is investigating advanced Brayton cycles for use with next generation nuclear power plants. The focus of this work is on the supercritical CO 2 Brayton cycle which has the potential for high efficiency, and for reduced capital costs due to very compact turbomachinery. Sandia has fabricated and is operating a supercritical CO 2 (S-CO 2 ) test loop to investigate the key technology issues associated with this cycle. This loop is part of a multi-year phased development program to develop a megawatt (MW) class closed S-CO 2 Brayton cycle to demonstrate the applicability of this cycle for DOE Gen-IV program. The current loop has been configured as both a compression loop and as simple heated but unrecuperated Brayton cycle. A second split-flow or re-compression Brayton cycle is currently under development that will use approximately 1 MW of heat to run the Brayton cycle. Early configurations of this split-flow Brayton cycle will be operational later this fiscal year. The key issues for this cycle include the fundamental issues of compressor fluid performance and system control near the critical point, but also the supporting technology issues of bearings, sealing technologies, and rotor windage losses which are also essential to achieving efficiency and cost objectives. These tests are providing the first measurements and information on these key supercritical CO 2 power conversion systems questions. Important data for all these issues has been obtained. This report presents the major results of the testing by showing and comparing the measured compressor performance map with the predicted performance. The compression loop uses a ∼50 kWe motor driven compressor to spin a 37 mm OD compressor at design speeds up to 75,000 rpm with a pressure ratio of 1.8 and a flow rate of 3.53 kg/s for a compressor inlet condition of 305.3 K and 7690 kPa. The most recent configuration of this loop has added a small turbine and 260 kW of heater power is

  1. Fast flux test facility primary sodium check valve

    International Nuclear Information System (INIS)

    Rabe, G.B.; Nash, C.F.

    1975-01-01

    The design and development of a tilting-disc check valve for the primary sodium coolant loop of the Fast Flux Test Facility is described. The demanding design requirements specified for this system dictated a design with unique features. These features, along with the structural design and analysis requirements and the testing program used to develop and justify the design, are described

  2. Irradiation Facilities at the Advanced Test Reactor

    International Nuclear Information System (INIS)

    S. Blaine Grover

    2005-01-01

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

  3. W-1 Sodium Loop Safety Facility experiment centerline fuel thermocouple performance

    International Nuclear Information System (INIS)

    Meyers, S.C.; Henderson, J.M.

    1980-05-01

    The W-1 Sodium Loop Safety Facility (SLSF) experiment is the fifth in a series of experiments sponsored by the Department of Energy (DOE) as part of the National Fast Breeder Reactor (FBR) Safety Assurance Program. The experiments are being conducted under the direction of Argonne National Laboratory (ANL) and Hanford Engineering Development Laboratory (HEDL). The irradiation phase of the W-1 SLSF experiment was conducted between May 27 and July 20, 1979, and terminated with incipient fuel pin cladding failure during the final boiling transient. Experimental hardware and facility performed as designed, allowing completion of all planned tests and test objectives. This paper focuses on high temperature in-fuel thermocouples and discusses their development, fabrication, and performance in the W-1 experiment

  4. Xenon oscillation tests in four-loop PWR cores

    International Nuclear Information System (INIS)

    Aoki, Norihiko; Osaka, Kenichi; Shimada, Shoichiro; Tochihara, Hiroshi; Machii, Seigo

    1980-01-01

    The Kansai Electric Power Co.'s OHI Unit 1 and 2 are the first 4-loop PWRs in Japan which use 17 x 17 fuel assemblies and have essentially the same plant parameters. A 4-loop core has larger core radius and higher power density than those of 2- or 3-loop cores, and is less stable for Xe oscillation. It is therefore important to confirm that Xe oscillations in radial direction are sufficiently stable in a 4-loop core. Radial and axial Xe oscillation tests were performed during the startup physics tests of OHI Unit 1 and 2; Xe oscillation was induced by perturbation of control rods and the Xe effect on power distribution observed periodically. The test results show that the transverse Xe oscillation in the 4-loop core is sufficiently stable and that the agreement between the measurement and the calculated prediction is good. (author)

  5. Mirror Fusion Test Facility (MFTF)

    International Nuclear Information System (INIS)

    Thomassen, K.I.

    1978-01-01

    A large, new Mirror Fusion Test Facility is under construction at LLL. Begun in FY78 it will be completed at the end of FY78 at a cost of $94.2M. This facility gives the mirror program the flexibility to explore mirror confinement principles at a signficant scale and advances the technology of large reactor-like devices. The role of MFTF in the LLL program is described here

  6. (abstract) Cryogenic Telescope Test Facility

    Science.gov (United States)

    Luchik, T. S.; Chave, R. G.; Nash, A. E.

    1995-01-01

    An optical test Dewar is being constructed with the unique capability to test mirrors of diameter less than or equal to 1 m, f less than or equal to 6, at temperatures from 300 to 4.2 K with a ZYGO Mark IV interferometer. The design and performance of this facility will be presented.

  7. Review of application code and standards for mechanical and piping design of HANARO fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. Y.

    1998-02-01

    The design and installation of the irradiation test facility for verification test of the fuel performance are very important in connection with maximization of the utilization of HANARO. HANARO fuel test loop was designed in accordance with the same code and standards of nuclear power plant because HANARO FTL will be operated the high pressure and temperature same as nuclear power plant operation conditions. The objective of this study is to confirm the propriety of application code and standards for mechanical and piping of HANARO fuel test loop and to decide the technical specification of FTL systems. (author). 18 refs., 8 tabs., 6 figs.

  8. Tritium Systems Test Facility. Volume I

    International Nuclear Information System (INIS)

    Anderson, G.W.; Battleson, K.W.; Bauer, W.

    1976-10-01

    Sandia Laboratories proposes to build and operate a Tritium Systems Test Facility (TSTF) in its newly completed Tritium Research Laboratory at Livermore, California (see frontispiece). The facility will demonstrate at a scale factor of 1:200 the tritium fuel cycle systems for an Experimental Power Reactor (EPR). This scale for each of the TSTF subsystems--torus, pumping system, fuel purifier, isotope separator, and tritium store--will allow confident extrapolation to EPR dimensions. Coolant loop and reactor hall cleanup facilities are also reproduced, but to different scales. It is believed that all critical details of an EPR tritium system will be simulated correctly in the facility. Tritium systems necessary for interim devices such as the Ignition Test Reactor (ITR) or The Next Step (TNS) can also be simulated in TSTF at other scale values. The active tritium system will be completely enclosed in an inert atmosphere glove box which will be connected to the existing Gas Purification System (GPS) of the Tritium Research Laboratory. In effect, the GPS will become the scaled environmental control system which otherwise would have to be built especially for the TSTF

  9. Corrosion testing facilities in India

    International Nuclear Information System (INIS)

    Viswanathan, R.; Subramanian, Venu

    1981-01-01

    Major types of corrosion tests, establishment of specifications on corrosion testing and scope of their application in practice are briefly described. Important organizations in the world which publish specifications/standards are listed. Indian organizations which undertake corrosion testing and test facilities available at them are also listed. Finally in an appendix, a comprehensive list of specifications relevant to corrosion testing is given. It is arranged under the headings: environmental testing, humidity tests, salt spray/fog tests, immersion tests, specification corrosion phenomena, (tests) with respect to special corrosion media, (tests) with respect to specific corrosion prevention methods, and specific corrosion tests using electrical and electrochemical methods (principles). Each entry in the list furnishes information about: nature of the test, standard number, and its specific application. (M.G.B.)

  10. Engineering test facility design center

    International Nuclear Information System (INIS)

    Anon.

    1980-01-01

    The vehicle by which the fusion program would move into the engineering testing phase of fusion power development is designated the Engineering Test Facility (ETF). The ETF would provide a test bed for reactor components in the fusion environment. In order to initiate preliminary planning for the ETF decision, the Office of Fusion Energy established the ETF Design Center activity to prepare the design of the ETF. This section describes the status of this design

  11. Oak Ridge rf Test Facility

    International Nuclear Information System (INIS)

    Gardner, W.L.; Hoffman, D.J.; McCurdy, H.C.; McManamy, T.J.; Moeller, J.A.; Ryan, P.M.

    1985-01-01

    The rf Test Facility (RFTF) of Oak Ridge National Laboratory (ORNL) provides a national facility for the testing and evaluation of steady-state, high-power (approx.1.0-MW) ion cyclotron resonance heating (ICRH) systems and components. The facility consists of a vacuum vessel and two fully tested superconducting development magnets from the ELMO Bumpy Torus Proof-of-Principle (EBT-P) program. These are arranged as a simple mirror with a mirror ratio of 4.8. The axial centerline distance between magnet throat centers is 112 cm. The vacuum vessel cavity has a large port (74 by 163 cm) and a test volume adequate for testing prototypic launchers for Doublet III-D (DIII-D), Tore Supra, and the Tokamak Fusion Test Reactor (TFTR). Attached to the internal vessel walls are water-cooled panels for removing the injected rf power. The magnets are capable of generating a steady-state field of approx.3 T on axis in the magnet throats. Steady-state plasmas are generated in the facility by cyclotron resonance breakdown using a dedicated 200-kW, 28-GHz gyrotron. Available rf sources cover a frequency range of 2 to 200 MHz at 1.5 kW and 3 to 18 MHz at 200 kW, with several sources at intermediate parameters. Available in July 1986 will be a >1.0-MW, cw source spanning 40 to 80 MHz. 5 figs

  12. Instrumentation and measurement method for the ATLAS test facility

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Byong Jo; Chu, In Chul; Eu, Dong Jin; Kang, Kyong Ho; Kim, Yeon Sik; Song, Chul Hwa; Baek, Won Pil

    2007-03-15

    An integral effect test loop for pressurized water reactors (PWRs), the ATLAS is constructed by thermal-hydraulic safety research division in KAERI. The ATLAS facility has been designed to have the length scale of 1/2 and area scale of 1/144 compared with the reference plant, APR1400 which is a Korean evolution type nuclear reactors. A total 1300 instrumentations is equipped in the ATLAS test facility. In this report, the instrumentation of ATLAS test facility and related measurement methods were introduced.

  13. Commissioning of an Integral Effect Test Loop for SMART

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hyunsik; Bae, Hwang; Kim, Dongeok; Min, Kyoungho; Shin, Yongcheol; Ko, Yungjoo; Yi, Sungjae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    An integral-effect test loop for SMART, SMART-ITL (or FESTA), has been constructed at KAERI. Its height was preserved and its flow area and volume were scaled down to 1/49 compared with the prototype plant, SMART. The ratio of the hydraulic diameter is 1/7. The SMART is a 330 MW thermal power reactor, and its core exit temperature and PZR pressure are 323 .deg. C and 15 MPa during a normal working condition, respectively. The maximum power of the core heater in the SMART-ITL is 30% of the scaled full power. As shown in Fig. 1, the SMART-ITL consists of a primary system including a reactor pressure vessel with a pressurizer, four steam generators and four main coolant pumps, a secondary system, a safety system, and an auxiliary system. The SMART-ITL facility will be used to investigate the integral performance of the inter-connected components and possible thermal-hydraulic phenomena occurring in the SMART design, to validate its safety for various design basis events and broad transient scenarios, and to validate the related thermal-hydraulic models of the safety analysis codes. The scenarios include small-break loss-of coolant accident (SBLOCA) scenarios, complete loss of RCS flowrate (CLOF), steam generator tube rupture (SGTR), feedwater line break (FLB), and main steam line break (MSLB). The role of SMART-ITL will be extended to examine and verify the normal, abnormal, and emergency operating procedures required during the construction and export phases of SMART. After an extensive series of commissioning tests in 2012, the SMART-ITL facility is now in operation. In this paper, the major test results acquired during the commissioning tests will be discussed.

  14. UF/sub 6/ test loop for evaluation and implementation of international enrichment plant safeguards

    International Nuclear Information System (INIS)

    Cooley, J.N.; Fields, L.W.; Swindle, D.W. Jr.

    1987-01-01

    A functional test loop capable of simulating UF/sub 6/ flows, pressures, and pipe deposits characteristic of gas centrifuge enrichment plant piping has been designed and fabricated by the Enrichment Safeguards Program of Martin Marietta Energy Systems, Inc., for use by the International Atomic Energy Agency (IAEA) at its Safeguards Analytical Laboratory in Seibersdorf, Austria. The purpose of the test loop is twofold: (1) to enable the IAEA to evaluate and to calibrate enrichment safeguards measurement instrumentation to be used in limited frequency-unannounced access (LFUA) inspection strategy measurements at gas centrifuge enrichment plants and (2) to train IAEA inspectors in the use of such instrumentation. The test loop incorporates actual sections of cascade header pipes from the centrifuge enrichment plants subject to IAEA inspections. The test loop is described, applications for its use by the IAEA are detailed, and results from an initial demonstration session using the test loop are summarized. By giving the IAEA the in-house capability to evaluate LFUA inspection strategy approaches, to develop inspection procedures, to calibrate instrumentation, and to train inspectors, the UF/sub 6/ cascade header pipe test loop will contribute to the IAEA's success in implementing LFUA strategy inspections at gas centrifuge enrichment facilities subject to international safeguards inspections

  15. Summary of facility and operating experience on helium engineering demonstration loop (HENDEL)

    Energy Technology Data Exchange (ETDEWEB)

    Ouchi, Yoshihiro; Fujisaki, Katsuo; Kobayashi, Toshiaki; Kato, Michio; Ota, Yukimaru; Watanabe, Syuji; Kobayashi, Hideki; Mogi, Haruyoshi [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    1996-07-01

    The HENDEL is a test facility to perform full scale demonstration tests on the core internals and high temperature components for the High Temperature Engineering Test Reactor(HTTR). The main systems consist of Mother(M) and Adapter(A), fuel stack Test(T{sub 1}) and in-core structure Test(T{sub 2}) sections. The (M+A) section can supply high temperature helium gas to the test section. The M+A section completed in March 1982 has been operated for about 22900 hours till February 1995. The T{sub 1} and T{sub 2} sections, completed in March 1983 and June 1986, have been operated for about 19400 and 16700 hours, respectively. In this period, a large number of tests have been conducted to verify the performance and safety features of the HTTR components. The results obtained from these tests have been effectively applied to the detailed design, licensing procedures and construction of the HTTR. The operating experience of the HENDEL for more than 10 years also brought us establishment of the technique of operation of a large scale helium gas loop, handling of helium gas and maintenance of high temperature facilities. The technique will be available for the operation of the HTTR. This paper mainly describes the summary of plant facirities, operating experience and maintenance on the HENDEL. (author)

  16. The Brookhaven Accelerator Test Facility

    International Nuclear Information System (INIS)

    Batchelor, K.; Ben-Zvi, I.; Fernow, R.C.; Fischer, J.; Fisher, A.S.; Gallardo, J.; Jialin, Xie; Kirk, H.G.; Parsa, Z.; Palmer, R.B.; Rao, T.; Rogers, J.; Sheehan, J.; Tsang, T.Y.F.; Ulc, S.; Van Steenbergen, A.; Woodle, M.; Zhang, R.S.; McDonald, K.T.; Russell, D.P.; Jiang, Z.Y.; Pellegrini, C.; Wang, X.J.

    1990-01-01

    The Accelerator Test Facility (ATF), presently under construction at Brookhaven National laboratory, is described. It consists of a 50-MeV electron beam synchronizable to a high-peak power CO 2 laser. The interaction of electrons with the laser field will be probed, with some emphasis on exploring laser-based acceleration techniques. 5 refs., 2 figs

  17. Construction of the two-phase critical flow test facility

    International Nuclear Information System (INIS)

    Chung, C. H.; Chang, S. K.; Park, H. S.; Min, K. H.; Choi, N. H.; Kim, C. H.; Lee, S. H.; Kim, H. C.; Chang, M. H.

    2002-03-01

    The two-phase critical test loop facility has been constructed in the KAERI engineering laboratory for the simulation of small break loss of coolant accident entrained with non-condensible gas of SMART. The test facility can operate at 12 MPa of pressure and 0 to 60 C of sub-cooling with 0.5 kg/s of non- condensible gas injection into break flow, and simulate up to 20 mm of pipe break. Main components of the test facility were arranged such that the pressure vessel containing coolant, a test section simulating break and a suppression tank inter-connected with pipings were installed vertically. As quick opening valve opens, high pressure/temperature coolant flows through the test section forming critical two-phase flow into the suppression tank. The pressure vessel was connected to two high pressure N2 gas tanks through a control valve to control pressure in the pressure vessel. Another N2 gas tank was also connected to the test section for the non-condensible gas injection. The test facility operation was performed on computers supported with PLC systems installed in the control room, and test data such as temperature, break flow rate, pressure drop across test section, gas injection flow rate were all together gathered in the data acquisition system for further data analysis. This test facility was classified as a safety related high pressure gas facility in law. Thus the loop design documentation was reviewed, and inspected during construction of the test loop by the regulatory body. And the regulatory body issued permission for the operation of the test facility

  18. Performance of the CNEN MHD Blow-Down Loop Facility

    Energy Technology Data Exchange (ETDEWEB)

    Bertolini, E.; Brown, R.; Gasparotto, M.; Gay, P.; Toschi, R. [Laboratorio Conversione Diretta, CNEN, Frascati (Italy)

    1968-11-15

    The CNEN facility has been designed, manufactured and used for alkali-seeded noble gas MHD energy conversion research, as the major experimental effort during the first five-year CNEN Research Programme on MHD. The main specifications and the general arrangement with information on preliminary commissioning tests of some components were given at the Salzburg Symposium. Since then the facility has been successfully commissioned and from March 1967 has been working on MHD experiments. Efforts were made to reduce any adverse effects on the experimental MHD results that were due to inherent limitations of an experimental apparatus (particularly under open-circuit conditions). Great emphasis was placed on problems of caesium vaporization and the mixing with helium, the purity level of the mixture, measurements and the control system. The insulation of the plasma from ground was carefully treated, increasing the ratio between insulator resistance and typical plasma resistance as much as possible. Fluidynamic tests at room and high temperatures have shown that stability in the gas parameters (temperature, pressure and mass flow) can be maintained within few per cent for tens of seconds after a transient, giving a behaviour similar to a continuously running system. The high- temperature, alumina pebble-bed heater has successfully operated, bringing the helium-caesium mixtures up to 2000 Degree-Sign K and up to 4 atm abs pressure, and undergoing seven thermal cycles, for a total of more than 2000 hours operation at top temperature. Preheated generator ducts using alumina as insulator and tantalum for electrodes performed satisfactorily, very much attention having been given in the design to reduction of thermal shocks and to obviating possible paths for caesium leakage and short-circuiting of electrode leads. The pulsed liquid nitrogen precooled magnet has been run for about 50 pulses at high field ( Asymptotically-Equal-To 4.5 tesla) with an operating time of about 10

  19. An Investigation of Loop Seal Clearings in ATLAS SBLOCA Tests

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeonsik; Cho, Seok; Kang, Kyoungho; Park, Hyunsik; Min, Kyeongho; Choi, Namhyeon; Park, Jonggook; Kim, Bokdeuk; Choi, Kiyong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    In most of the SBLOCA cases, the pressure of the upper-head region will increase mainly owing to the accumulated steam and water inventory in the upper-plenum. This build-up pressure acts as a suppression force to the core water level, and resultantly the core water level will decrease possibly up to and/or below the top of the active core region. Simultaneously, the downcomer water level will increase owing to the evacuated water inventory from the lower part of the core region. This unbalanced hydro-static pressure between the core and downcomer region acts as a potential pushing force to the reactor coolant pump (RCP) side intermediate leg. The potential pushing force will be increased with time to overcome the hydro-static head in the upflow intermediate leg. The unbalanced hydro-static pressure can finally be dissolved with the occurrence of the loop seal clearing. A minimum core collapsed water level, located below the elevation of the loop seal bottom leg in the ATLAS tests, is taken at this time. Since the loop seal bottom leg is located below the core top for typical PWR plants such as an APR1400, the water level depression may uncover the core upper regions until the core water level recovers with the progress of the clearing of the loop seal upflow leg. At this moment, the core temperature may increase to a peak cladding temperature (PCT) owing to an excessive core uncovery by the minimum core collapsed water level. Therefore, the loop seal clearing phenomenon is very important with respect to the PCT occurrence, which is one of the most important parameters to insure the safety of the reactor system. The loop seal clearing behavior seems to be closely related to the break location and break size. Usually, a loop seal in the break loop is cleared first, and the number of loop seal clearings is dependent on the break size. The larger the break size, the more the loop seals that are cleared. An investigation of LSC in the SBLOCA for DVI line and CL breaks

  20. TLTA/6431, Two-Loop-Test-Apparatus, BWR/6 Simulator, Small-Break LOCA

    International Nuclear Information System (INIS)

    1992-01-01

    1 - Description of test facility: The Two-Loop-Test-Apparatus (TLTA) is a 1:624 volume scaled BWR/6 simulator. It was the predecessor of the better-scaled FIST facility. The facility is capable of full BWR system pressure and has a simulated core with a full size 8 x 8, full power single bundle of indirect electrically heated rods. All major BWR systems are simulated including lower plenum, guide tube, core region (bundle and bypass), upper plenum, steam separator, steam dome, annular downcomer, recirculation loops and ECC injection systems. The fundamental scaling consideration was to achieve real-time response. A number of the scaling compromises present in TLTA were corrected in the FIST configuration. These compromises include a number of regional volumes and component elevations. 2 - Description of test: 64.45 sqcm small break LOCA with activation of the full emergency core cooling system, but without activation of the automatic decompression system

  1. Sodium Loop Safety Facility W-2 experiment fuel pin rupture detection system. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Hoffman, M.A.; Kirchner, T.L.; Meyers, S.C.

    1980-05-01

    The objective of the Sodium Loop Safety Facility (SLSF) W-2 experiment is to characterize the combined effects of a preconditioned full-length fuel column and slow transient overpower (TOP) conditions on breeder reactor (BR) fuel pin cladding failures. The W-2 experiment will meet this objective by providing data in two technological areas: (1) time and location of cladding failure, and (2) early post-failure test fuel behavior. The test involves a seven pin, prototypic full-length fast test reactor (FTR) fuel pin bundle which will be subjected to a simulated unprotected 5 cents/s reactivity transient overpower event. The outer six pins will provide the necessary prototypic thermal-hydraulic environment for the center pin.

  2. Sodium Loop Safety Facility W-2 experiment fuel pin rupture detection system

    International Nuclear Information System (INIS)

    Hoffman, M.A.; Kirchner, T.L.; Meyers, S.C.

    1980-05-01

    The objective of the Sodium Loop Safety Facility (SLSF) W-2 experiment is to characterize the combined effects of a preconditioned full-length fuel column and slow transient overpower (TOP) conditions on breeder reactor (BR) fuel pin cladding failures. The W-2 experiment will meet this objective by providing data in two technological areas: (1) time and location of cladding failure, and (2) early post-failure test fuel behavior. The test involves a seven pin, prototypic full-length fast test reactor (FTR) fuel pin bundle which will be subjected to a simulated unprotected 5 cents/s reactivity transient overpower event. The outer six pins will provide the necessary prototypic thermal-hydraulic environment for the center pin

  3. Fusion Materials Irradiation Test Facility

    International Nuclear Information System (INIS)

    Kemp, E.L.; Trego, A.L.

    1979-01-01

    A Fusion Materials Irradiation Test Facility is being designed to be constructed at Hanford, Washington, The system is designed to produce about 10 15 n/cm-s in a volume of approx. 10 cc and 10 14 n/cm-s in a volume of 500 cc. The lithium and target systems are being developed and designed by HEDL while the 35-MeV, 100-mA cw accelerator is being designed by LASL. The accelerator components will be fabricated by US industry. The total estimated cost of the FMIT is $105 million. The facility is scheduled to begin operation in September 1984

  4. Aircraft Test & Evaluation Facility (Hush House)

    Data.gov (United States)

    Federal Laboratory Consortium — The Aircraft Test and Evaluation Facility (ATEF), or Hush House, is a noise-abated ground test sub-facility. The facility's controlled environment provides 24-hour...

  5. Millimeter-wave Instrumentation Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Millimeter-wave Instrumentation Test Facility conducts basic research in propagation phenomena, remote sensing, and target signatures. The facility has a breadth...

  6. Fuels and materials testing capabilities in Fast Flux Test Facility

    International Nuclear Information System (INIS)

    Baker, R.B.; Chastain, S.A.; Culley, G.E.; Ethridge, J.L.; Lovell, A.J.; Newland, D.J.; Pember, L.A.; Puigh, R.J.; Waltar, A.E.

    1989-01-01

    The Fast Flux Test Facility (FFTF) reactor, which started operating in 1982, is a 400 MWt sodium-cooled fast neutron reactor located in Hanford, Washington State, and operated by Westinghouse Hanford Co. under contract with U.S. Department of Energy. The reactor has a wide variety of functions for irradiation tests and special tests, and its major purpose is the irradiation of fuel and material for liquid metal reactor, nuclear reactor and space reactor projects. The review first describes major technical specifications and current conditions of the FFTF reactor. Then the plan for irradiation testing is outlined focusing on general features, fuel pin/assembly irradiation tests, and absorber irradiation tests. Assemblies for special tests include the material open test assembly (MOTA), fuel open test assembly (FOTA), closed loop in-reactor assembly (CLIRA), and other special fuel assemblies. An interim examination and maintenance cell (FFTF/IEM cell) and other hot cells are used for nondestructive/destructive tests and physical/mechanical properties test of material after irradiation. (N.K.)

  7. ITER primary cryopump test facility

    International Nuclear Information System (INIS)

    Petersohn, N.; Mack, A.; Boissin, J.C.; Murdoc, D.

    1998-01-01

    A cryopump as ITER primary vacuum pump is being developed at FZK under the European fusion technology programme. The ITER vacuum system comprises of 16 cryopumps operating in a cyclic mode which fulfills the vacuum requirements in all ITER operation modes. Prior to the construction of a prototype cryopump, the concept is tested on a reduced scale model pump. To test the model pump, the TIMO facility is being built at FZK in which the model pump operation under ITER environmental conditions, except for tritium exposure, neutron irradiation and magnetic fields, can be simulated. The TIMO facility mainly consists of a test vessel for ITER divertor duct simulation, a 600 W refrigerator system supplying helium in the 5 K stage and a 30 kW helium supply system for the 80 K stage. The model pump test programme will be performed with regard to the pumping performance and cryogenic operation of the pump. The results of the model pump testing will lead to the design of the full scale ITER cryopump. (orig.)

  8. Natural circulation in an integral CANDU test facility

    International Nuclear Information System (INIS)

    Ingham, P.J.; Sanderson, T.V.; Luxat, J.C.; Melnyk, A.J.

    2000-01-01

    Over 70 single- and two-phase natural circulation experiments have been completed in the RD-14M facility, an integral CANDU thermalhydraulic test loop. This paper describes the RD-14M facility and provides an overview of the impact of key parameters on the results of natural circulation experiments. Particular emphasis will be on phenomena which led to heat up at high system inventories in a small subset of experiments. Clarification of misunderstandings in a recently published comparison of the effectiveness of natural circulation flows in RD-14M to integral facilities simulating other reactor geometries will also be provided. (author)

  9. Data acquisition for the Sodium Loop Safety Facility experiment P4

    International Nuclear Information System (INIS)

    Baldwin, R.D.; Kraimer, M.R.; Wilson, R.E.; Gilbert, D.M.

    1982-01-01

    Data acquisition for the Sodium Loop Safety Facility (SLSF) experiment P4 used three computers for the continuous collection of data and two computers for the routing and displaying of data. Four of these computer systems were located at the Engineering Test Reactor (ETR) site, in Idaho, to access sensor signals from the analog to digital interfaces. The fifth system was located at Argonne National Laboratory (ANL), in Illinois, and was used mainly for display and storage of data. All display computers were connected together using the DECNET software package. The transmission of data was managed over a dedicated phone line using 9600 baud long distance modems. A stand-alone high speed data acquisition system was also used to record data during planned reactor transients

  10. Production Facility Prototype Blower 1000 Hour Test Results II

    Energy Technology Data Exchange (ETDEWEB)

    Wass, Alexander Joseph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Woloshun, Keith Albert [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dale, Gregory E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dalmas, Dale Allen [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Romero, Frank Patrick [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2018-01-08

    Long duration tests of the Aerzen GM 12.4 roots style blower in a closed loop configuration provides valuable data and lessons learned for long-term operation at the Mo-99 production facility. The blower was operated in a closed loop configuration with the flow conditions anticipated in plant operation with a Mo-100 target inline. The additional thermal energy generated from beam heating of the Mo-100 disks were not included in these tests. Five 1000 hour tests have been completed since the first test was performed in January of 2016. All five 1000 hour tests have proven successful in exposing preventable issues related to oil and helium leaks. All blower tests to this date have resulted in stable blower performance and consistency. A summary of the results for each test, including a review of the first and second tests, are included in this report.

  11. Mirror Fusion Test Facility magnet

    International Nuclear Information System (INIS)

    Henning, C.H.; Hodges, A.J.; Van Sant, J.H.; Hinkle, R.E.; Horvath, J.A.; Hintz, R.E.; Dalder, E.; Baldi, R.; Tatro, R.

    1979-01-01

    The Mirror Fusion Test Facility (MFTF) is the largest of the mirror program experiments for magnetic fusion energy. It seeks to combine and extend the near-classical plasma confinement achieved in 2XIIB with the most advanced neutral-beam and magnet technologies. The product of ion density and confinement time will be improved more than an order of magnitude, while the superconducting magnet weight will be extrapolated from the 15 tons in Baseball II to 375 tons in MFTF. Recent reactor studies show that the MFTF will traverse much of the distance in magnet technology towards the reactor regime. Design specifics of the magnet are given

  12. Coupled hydrodynamic-structural analysis of an integral flowing sodium test loop in the TREAT reactor

    International Nuclear Information System (INIS)

    Zeuch, W.R.; A-Moneim, M.T.

    1979-01-01

    A hydrodynamic-structural response analysis of the Mark-IICB loop was performed for the TREAT (Transient Reactor Test Facility) test AX-1. Test AX-1 is intended to provide information concerning the potential for a vapor explosion in an advanced-fueled LMFBR. The test will be conducted in TREAT with unirradiated uranium-carbide fuel pins in the Mark-IICB integral flowing sodium loop. Our analysis addressed the ability of the experimental hardware to maintain its containment integrity during the reference accident postulated for the test. Based on a thermal-hydraulics analysis and assumptions for fuel-coolant interaction in the test section, a pressure pulse of 144 MPa maximum pressure and pulse width of 1.32 ms has been calculated as the reference accident. The response of the test loop to the pressure transient was obtained with the ICEPEL and STRAW codes. Modelling of the test section was completed with STRAW and the remainder of the loop was modelled by ICEPEL

  13. Dismantling of the 50 MW steam generator test facility

    International Nuclear Information System (INIS)

    Nakai, S.; Onojima, T.; Yamamoto, S.; Akai, M.; Isozaki, T.; Gunji, M.; Yatabe, T.

    1997-01-01

    We have been dismantling the 50MW Steam Generator Test Facility (50MWSGTF). The objectives of the dismantling are reuse of sodium components to a planned large scale thermal hydraulics sodium test facility and the material examination of component that have been operated for long time in sodium. The facility consisted of primary sodium loop with sodium heater by gas burner as heat source instead of reactor, secondary sodium loop with auxiliary cooling system (ACS) and water/steam system with steam temperature and pressure reducer instead of turbine. It simulated the 1 loop of the Monju cooling system. The rated power of the facility was 50MWt and it was about 1/5 of the Monju power plant. Several sodium removal methods are applied. As for the components to be dismantled such as piping, intermediate heat exchanger (IHX), air cooled heat exchangers (AC), sodium is removed by steam with nitrogen gas in the air or sodium is burned in the air. As for steam generators which material tests are planned, sodium is removed by steam injection with nitrogen gas to the steam generator. The steam generator vessel is filled with nitrogen and no air in the steam generator during sodium removal. As for sodium pumps, pump internal structure is pulled out from the casing and installed into the tank. After the installation, sodium is removed by the same method of steam generator. As for relatively small reuse components such as sodium valves, electromagnet flow meters (EMFs) etc., sodium is removed by alcohol process. (author)

  14. Acoustic detection of boiling in the Sodium Loop Safety Facility in-reactor experiment P1

    International Nuclear Information System (INIS)

    Carey, W.M.; Anderson, T.T.; Bobis, J.P.

    1976-06-01

    Acoustic data were obtained from two high-temperature lithium niobate microphones on the loop background noise and transient pressure pulses during the Sodium Loop Safety Facility (SLSF) P1 in-reactor experiment. This experiment simulated an LMFBR loss-of-piping-integrity (LOPI) transient on a nineteen element, end-of-life, enriched-UO 2 fuel assembly. The microphones were exposed to liquid sodium at a distance 4.85 meters above the reactor core at temperatures between 315 0 and 590 0 C. The distance and location of the microphones in the P1 Test Train provided an attenuative transmission path which was undesirable for optimum acoustic detection of sodium boiling and fuel failure. The data gathered on the loop background noise was observed to be dominated by pump and electrical noise at frequencies below 1.5 KHz and appeared to be dominated by flow induced local turbulence noise at higher frequencies. During the period of time that the sodium in the fuel assembly was at its saturation temperature 943 0 C (1730 0 F), as indicated by the wire wrap thermocouples, several discrete pulses were observed with peak-to-peak pressure between 3.3 kPa and 7.9 kPa and center frequencies between 360 and 550 Hz. The pulses occurred at two separate gradually increasing repetition rates. These observations appear to be consistent with the result of an impulsive forcing function interacting with a band passed Helmholtz resonator. These data are consistent with the hypothesis that sodium boiling occurred in the P1 fuel assembly, resulting in the formation of individual voids that collapsed upon reaching the subcooled sodium. These data provide pertinent information regarding the feasibility of sodium boiling detection and may provide additional insight into the dynamics of the void behavior

  15. Design criteria of out-pile system of HANARO fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. Y.

    1997-07-01

    The objective of HANARO aims at the development and localization of nuclear technologies through the engineering tests. Thus it is very important the design and installation of the irradiation test facilities to be installed at the irradiation hole for verification test of the fuel performance are in connection with maximization of the utilization of HANARO. The principle subjects of this study are to presend and informed the detail design criteria and technical specification of out-pile system of HANARO fuel test loop for the developing of the fuel and reactor material. This results will become guidance for the planning of the irradiation testing using the HANARO fuel test loop. (author). 16 refs., 31 tabs., 9 figs.

  16. Test of Flow Characteristics in Tubular Fuel Assembly I - Establishment of test loop and measurement validation test

    International Nuclear Information System (INIS)

    Park, Jong Hark; Chae, H. T.; Park, C.; Kim, H.

    2005-12-01

    Tubular type fuel has been developed as one of candidates for Advanced HANARO Reactor(AHR). It is necessary to test the flow characteristics such as velocity in each flow channels and pressure drop of tubular type fuel. A hydraulic test-loop to examine the hydraulic characteristics for a tubular type fuel has been designed and constructed. It consists of three parts; a) piping-loop including pump and motor, magnetic flow meter and valves etc, b) test-section part where a simulated tubular type fuel is located, and 3) data acquisition system to get reading signals from sensors or instruments. In this report, considerations during the design and installation of the facility and the selection of data acquisition sensors and instruments are described in detail. Before doing the experiment to measure the flow velocities in flow channels, a preliminary tests have been done for measuring the coolant velocities using pitot-tube and for validating the measurement accuracy as well. Local velocities of the radial direction in circular tubes are measured at regular intervals of 60 degrees by three pitot-tubes. Flow rate inside the circular flow channel can be obtained by integrating the velocity distribution in radial direction. The measured flow rate was compared to that of magnetic flow meter. According to the results, two values had a good agreement, which means that the measurement of coolant velocity by using pitot-tube and the flow rate measured by the magnetic flow meter are reliable. Uncertainty analysis showed that the error of velocity measurement by pitot-tube is less than ±2.21%. The hydraulic test-loop also can be adapted to others such as HANARO 18 and 36 fuel, in-pile system of FTL(Fuel Test Loop), etc

  17. Counter-part Test and Code Analysis of the Integral Test Loop, SNUF

    International Nuclear Information System (INIS)

    Park, Goon Cherl; Bae, B. U.; Lee, K. H.; Cho, Y. J.

    2007-02-01

    The thermal-hydraulic phenomena of Direct Vessel Injection (DVI) line Small-Break Loss-of-Coolant Accident (SBLOCA) in pressurized water reactor, APR1400, were investigated. The reduced-height and reduced-pressure integral test loop, SNUF (Seoul National University Facility), was constructed with scaling down the prototype. For the appropriate test conditions in the experiment of SNUF, the energy scaling methodology was suggested as scaling the coolant mass inventory and thermal power for the reduced-pressure condition. From the MARS code analysis, the energy scaling methodology was confirmed to show the reasonable transient when ideally scaled-down SNUF model was compared to the prototype model. In the experiments according to the conditions determined by energy scaling methodology, the phenomenon of downcomer seal clearing had a dominant role in decrease of the system pressure and increase of the coolant level of core. The experimental results was utilized to validate the calculation capability of MARS

  18. CLOSED LOOP AOCS TESTING OF AN AUTONOMOUS STAR TRACKER

    DEFF Research Database (Denmark)

    Jørgensen, John Leif

    1999-01-01

    not even a high quality star pattern generator may be able to pass the outlier rejection filtering of the ASC thus efficiently precluding artificial stimuli during AIT tests. In order to circumvent this impasse, the ASC has a series of build-in features enabling simple, yet comprehensive, closed loop...

  19. A one-loop test of string duality

    International Nuclear Information System (INIS)

    Vafa, C.

    1995-01-01

    We test Type IIA-heterotic string duality in six dimensions by showing that the sigma model anomaly of the heterotic string is generated by a combination of a tree level and a string one-loop correction on the Type IIA side. (orig.)

  20. Scoping erosion flow loop test results in support of Hanford WTP

    International Nuclear Information System (INIS)

    Duignan, M.; Imrich, K.; Fowley, M.; Restivo, M.; Reigel, M.

    2015-01-01

    The Waste Treatment and Immobilization Plant (WTP) will process Hanford Site tank waste by converting the waste into a stable glass form. Before the tank waste can be vitrified, the baseline plan is to process the waste through the Pretreatment (PT) Facility where it will be mixed in various process vessels using Pulse Jet Mixers (PJM) and transferred to the High Level Waste (HLW) or Low Activity Waste (LAW) vitrification facilities. The Department of Energy (DOE) and Defense Nuclear Facility Safety Board (DNFSB), as well as independent review groups, have raised concerns regarding the design basis for piping erosion in the PT Facility. Due to the complex nature of slurry erosion/corrosion wear and the unique conditions that exist within the PT Facility, additional testing has been recommended by these entities. Pipe loop testing is necessary to analyze the potential for localized wear at elbows and bends, close the outstanding PT and HLW erosion/corrosion technical issues, and underpin BNI's design basis for a 40-year operational life for black cell piping and vessels. SRNL is consulting with the DOE Office of River Protection (ORP) to resolve technical concerns related to piping erosion/corrosion (wear) design basis for PT. SRNL was tasked by ORP to start designing, building, and testing a flow loop to obtain long-term total-wear rate data using bounding simulant chemistry, operating conditions, and prototypical materials. The initial test involved a scoping paint loop to locate experimentally the potential high-wear locations. This information will provide a basis for the placement of the many sensitive wear measurement instruments in the appropriate locations so that the principal flow-loop test has the best chance to estimate long-term erosion and corrosion. It is important to note that the scoping paint loop test only utilized a bounding erosion simulant for this test. A full chemical simulant needs to be added for the complete test flow loop. The

  1. Survey of solar thermal test facilities

    Energy Technology Data Exchange (ETDEWEB)

    Masterson, K.

    1979-08-01

    The facilities that are presently available for testing solar thermal energy collection and conversion systems are briefly described. Facilities that are known to meet ASHRAE standard 93-77 for testing flat-plate collectors are listed. The DOE programs and test needs for distributed concentrating collectors are identified. Existing and planned facilities that meet these needs are described and continued support for most of them is recommended. The needs and facilities that are suitable for testing components of central receiver systems, several of which are located overseas, are identified. The central contact point for obtaining additional details and test procedures for these facilities is the Solar Thermal Test Facilities Users' Association in Albuquerque, N.M. The appendices contain data sheets and tables which give additional details on the technical capabilities of each facility. Also included is the 1975 Aerospace Corporation report on test facilities that is frequently referenced in the present work.

  2. Drop test facility available to private industry

    International Nuclear Information System (INIS)

    Shappert, L.B.; Box, W.D.

    1983-01-01

    In 1978, a virtually unyielding drop test impact pad was constructed at Oak Ridge National Laboratory's (ORNL's) Tower Shielding Facility (TSF) for the testing of heavy shipping containers designed for transporting radioactive materials. Because of the facility's unique capability for drop-testing large, massive shipping packages, it has been identified as a facility which can be made available for non-DOE users

  3. Results from tests of TFL Hydragard sampling loop

    International Nuclear Information System (INIS)

    Steimke, J.L.

    1995-03-01

    When the Defense Waste Processing Facility (DWPF) is operational, processed radioactive sludge will be transferred in batches to the Slurry Mix Evaporator (SME), where glass frit will be added and the contents concentrated by boiling. Batches of the slurry mixture are transferred from the SME to the Melter Feed Tank (MFT). Hydragard reg-sign sampling systems are used on the SME and the MFT for collecting slurry samples in vials for chemical analysis. An accurate replica of the Hydragard sampling system was built and tested in the thermal Fluids Laboratory (TFL) to determine the hydragard accuracy. It was determined that the original Hydragard valve frequently drew a non-representative sample stream through the sample vial that ranged from frit enriched to frit depleted. The Hydragard valve was modified by moving the plunger and its seat backwards so that the outer surface of the plunger was flush with the inside diameter of the transfer line when the valve was open. The slurry flowing through the vial accurately represented the composition of the slurry in the reservoir for two types of slurries, different dilution factors, a range of transfer flows and a range of vial flows. It was then found that the 15 ml of slurry left in the vial when the Hydragard valve was closed, which is what will be analyzed at DWPF, had a lower ratio of frit to sludge as characterized by the lithium to iron ratio than the slurry flowing through it. The reason for these differences is not understood at this time but it is recommended that additional experimentation be performed with the TFL Hydragard loop to determine the cause

  4. Model experiments on simulation of the WWER water-chemical conditions at loop facilities of the MIR reactor

    International Nuclear Information System (INIS)

    Benderskaya, O.S.; Zotov, E.A.; Kuprienko, V.A.; Ovchinnikov, V.A.

    1999-01-01

    The experiments on simulation of the WWER type reactors water-chemical conditions have been started at the State Scientific Center RIAR. These experiments are being conducted at the multi-loop research MIR reactor at the PVK-2 loop facility. The dosage stand was created. It allows introduction of boric acid, potassium and lithium hydroxides, ammonia solutions and gaseous hydrogen. Corrosion tests of the Russian E-635 and E-110 alloys are being conducted at the PVK-2 loop under the WWER water-chemical conditions. If necessary, fuel elements are periodically extracted from the reactor to perform visual examination, to measure their length, diameter, to remove the deposits from the claddings, to measure the burnup and to distribute the fission products over the fuel element by gamma-spectrometry. The chemical analytical 'on line' equipment produced by the ORBISPHERE Laboratory (Switzerland) will be commissioned in the nearest future to measure concentration of the dissolved hydrogen and oxygen as well as pH and specific conductivity. The objective of the report is to familiarize the participants of the IAEA Technical Committee with the capabilities of performing the model water-chemical experiments under the MIR reactor loop facility conditions. (author)

  5. Engineering test facility design definition

    Science.gov (United States)

    Bercaw, R. W.; Seikel, G. R.

    1980-01-01

    The Engineering Test Facility (ETF) is the major focus of the Department of Energy (DOE) Magnetohydrodynamics (MHD) Program to facilitate commercialization and to demonstrate the commercial operability of MHD/steam electric power. The ETF will be a fully integrated commercial prototype MHD power plant with a nominal output of 200 MW sub e. Performance of this plant is expected to meet or surpass existing utility standards for fuel, maintenance, and operating costs; plant availability; load following; safety; and durability. It is expected to meet all applicable environmental regulations. The current design concept conforming to the general definition, the basis for its selection, and the process which will be followed in further defining and updating the conceptual design.

  6. The Brookhaven Accelerator Test Facility

    International Nuclear Information System (INIS)

    Batchelor, K.; Chou, T.S.; Fernow, R.C.; Fischer, J.; Gallardo, J.; Kirk, H.G.; Koul, R.; Palmer, R.B.; Pellegrini, C.; Sheehan, J.; Srinivasan-Rao, T.; Ulc, S.; Woodle, M.; Bigio, I.; Kurnit, N.; McDonald, K.T.

    1989-01-01

    The Brookhaven Accelerator Test Facility ATF will consist of a 50-100 MeV/c electron linac and a 100 GW CO 2 laser system. A high brightness RF-gun operating at 2,856 MHz is to be used as the injector into the linac. The RF-gun contains a Nd:Yag-laser-driven photocathode capable of producing a stream of six ps electron pulses separated by 12.5 ns. The maximum charge in a micropulse will be one nano-Coulomb. The CO 2 laser pulse length will be a few picoseconds and will be synchronized with the electron pulse. The first experimental beam is expected in Fall 89. The design electron beam parameters are given and possible initial experiments are discussed. 9 refs., 1 fig., 3 tabs

  7. The Brookhaven Accelerator Test Facility

    International Nuclear Information System (INIS)

    Batchelor, K.; Chou, T.S.; Fernow, R.C.

    1988-01-01

    The Brookhaven Accelerator Test Facility (ATF) will consist of a 50--100 MeV/c electron linac and a 100 GW CO 2 laser system. A high brightness RF-gun operating at 2856 MHz is to be used as the injector into the linac. The RF-gun contains a Nd:Yag-laser-driven photocathode capable of producing a stream of six ps electron pulses separated by 12.5 ns. The maximum charge in a micropulse will be one nano-Coulomb. The CO 2 laser pulse length will be a few picoseconds and will be synchronized with the electron pulse. The first experimental beam is expected in Fall 89. The design electron beam parameters are given and possible initial experiments are discussed. 9 refs., 1 fig., 3 tabs

  8. Seismic proving test of BWR primary loop recirculation system

    International Nuclear Information System (INIS)

    Sato, H.; Shigeta, M.; Karasawa, Y.

    1987-01-01

    The seismic proving test of BWR Primary Loop Recirculation system is the second test to use the large-scale, high-performance vibration table of Tadotsu Engineering Laboratory. The purpose of this test is to prove the seismic reliability of the primary loop recirculation system (PLR), one of the most important safety components in the BWR nuclear plants, and also to confirm the adequacy of seismic analysis method used in the current seismic design. To achieve the purpose, the test was conducted under conditions and scale as near as possible to actual systems. The strength proving test was carried out with the test model mounted on the vibration table in consideration of basic design earthquake ground motions and other conditions to confirm the soundness of structure and the strength against earthquakes. Detailed analysis and analytic evaluation of the data obtained from the test was conducted to confirm the adequacy of the seismic analysis method and earthquake response analysis method used in the current seismic design. Then, on the basis of the results obtained, the seismic safety and reliability of BWR primary loop recirculation of the actual plants was fully evaluated

  9. PACTEL and PWR PACTEL Test Facilities for Versatile LWR Applications

    Directory of Open Access Journals (Sweden)

    Virpi Kouhia

    2012-01-01

    Full Text Available This paper describes construction and experimental research activities with two test facilities, PACTEL and PWR PACTEL. The PACTEL facility, comprising of reactor pressure vessel parts, three loops with horizontal steam generators, a pressurizer, and emergency core cooling systems, was designed to model the thermal-hydraulic behaviour of VVER-440-type reactors. The facility has been utilized in miscellaneous applications and experiments, for example, in the OECD International Standard Problem ISP-33. PACTEL has been upgraded and modified on a case-by-case basis. The latest facility configuration, the PWR PACTEL facility, was constructed for research activities associated with the EPR-type reactor. A significant design basis is to utilize certain parts of PACTEL, and at the same time, to focus on a proper construction of two new loops and vertical steam generators with an extensive instrumentation. The PWR PACTEL benchmark exercise was launched in 2010 with a small break loss-of-coolant accident test as the chosen transient. Both facilities, PACTEL and PWR PACTEL, are maintained fully operational side by side.

  10. PACTEL and PWR PACTEL Test Facilities for Versatile LWR Applications

    International Nuclear Information System (INIS)

    Virpi Kouhia, V.; Purhonen, H.; Riikonen, V.; Puustinen, M.; Kyrki-Rajamaki, R.; Vihavainen, J.

    2012-01-01

    This paper describes construction and experimental research activities with two test facilities, PACTEL and PWR PACTEL. The PACTEL facility, comprising of reactor pressure vessel parts, three loops with horizontal steam generators, a pressurizer, and emergency core cooling systems, was designed to model the thermal-hydraulic behaviour of VVER-440-type reactors. The facility has been utilized in miscellaneous applications and experiments, for example, in the OECD International Standard Problem ISP-33. PACTEL has been upgraded and modified on a case-by-case basis. The latest facility configuration, the PWR PACTEL facility, was constructed for research activities associated with the EPR-type reactor. A significant design basis is to utilize certain parts of PACTEL, and at the same time, to focus on a proper construction of two new loops and vertical steam generators with an extensive instrumentation. The PWR PACTEL benchmark exercise was launched in 2010 with a small break loss-of-coolant accident test as the chosen transient. Both facilities, PACTEL and PWR PACTEL, are maintained fully operational side by side.

  11. Clemson University Wind Turbine Drivetrain Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    Tuten, James Maner [Clemson Univ., SC (United States); Haque, Imtiaz [Clemson Univ., SC (United States); Rigas, Nikolaos [Clemson Univ., SC (United States)

    2016-03-30

    two dynamometer test rigs is the first significant achievement for the project. These test rigs embody a new manner of test due to the system configuration and completely new design with a free floating loading hub in the LAU. This project provided the catalyst for the advancement to this new test rig configuration that has been adopted by every significant wind turbine test rig constructed since the inception of this project. There are currently two different vendors supplying these new systems. Catalyzing this new design is the second major success of the project. With the increased market penetration of wind energy over the past decade, many regions and countries have developed specific electrical grid specifications and performance codes for large wind farms to ensure operational reliability and stability. These grid codes provide requirements for interconnection with the grid during low or high voltage phenomena, typically encountered during and after system fault events. Given the installed infrastructure of the Wind Turbine Drivetrain Testing Facility (WTDTF), a natural expansion of facility capability was to include the necessary equipment for performing fault ride-through evaluations of wind turbines to the Low Voltage Ride Through (LVRT) codes. Once the decision was made to expand the scope of the original grant into fault ride-through testing, it was clear that there are several markets, not just wind, which could benefit from this type of test and that simple fault ride-through testing could be extended into a broader scope of electrical testing capabilities. It was at this point that Clemson University was awarded a second grant to build a 15 MW Hardware-In-the-Loop (HIL) Grid Simulator in order to establish world class electrical testing capabilities to compliment the mechanical testing at the WTDTF. This third significant achievement resulted in the 15 MW HIL Grid Simulator as the corner stone of the Duke Energy eGRID Center and is collocated with the

  12. Effect of steam generator configuration in a loss of the RHR during mid-loop operation at PKL facility

    Energy Technology Data Exchange (ETDEWEB)

    Villanueva, J. F.; Carlos, S.; Martorell, S.; Sanchez, F. [Dpto. Ingenieria Quimica Y Nuclear, Universitat Politecnica de Valencia, Camino Vera s/n, 46022 Valencia (Spain)

    2012-07-01

    The loss of the residual heat removal system in mid-loop conditions may occur with a non-negligible contribution to the plant risk, so the analysis of the accidental sequences and the actions to mitigate the accident are of great interest in shutdown conditions. In order to plan the appropriate measures to mitigate the accident is necessary to understand the thermal-hydraulic processes following the loss of the residual heat removal system during shutdown. Thus, transients of this kind have been simulated using best-estimate codes in different integral test facilities and compared with experimental data obtained in different facilities. In PKL (Primaerkreislauf-Versuchsanlage, primary coolant loop test facility) test facility different series of experiments have been undertaken to analyze the plant response in shutdown. In this context, the E3 and F2 series consist of analyzing the loss of the residual heat removal system with a reduced inventory in the primary system. In particular, the experiments were developed to investigate the influence of the steam generators secondary side configuration on the plant response, what involves the consideration of different number of steam generators filled with water and ready for activation, on the heat transfer mechanisms inside the steam generators U-tubes. This work presents the results of such experiments calculated using, RELAP5/Mod 3.3. (authors)

  13. Natural circulation in a scaled PWR integral test facility

    International Nuclear Information System (INIS)

    Kiang, R.L.; Jeuck, P.R. III

    1987-01-01

    Natural circulation is an important mechanism for cooling a nuclear power plant under abnormal operating conditions. To study natural circulation, we modeled a type of pressurized water reactor (PWR) that incorporates once-through steam generators. We conducted tests of single-phase natural circulations, two-phase natural circulations, and a boiler condenser mode. Because of complex geometry, the natural circulations observed in this facility exhibit some phenomena not commonly seen in a simple thermosyphon loop

  14. Electromagnetic Interference (EMI) and TEMPEST Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — Electromagnetic Interference (EMI), Electromagnetic Compatibility (EMC) and TEMPEST testing are conducted at EPG's Blacktail Canyon Test Facility in one of its two...

  15. Fuel elements and fuel element materials. Experimental facilities for fission products lift-off tests

    International Nuclear Information System (INIS)

    Blanchard, R.J.; Veyrat, J.F.

    1978-01-01

    One of the hypothetical accidents on the HTGR primary cooling circuits is the failure of a circuit resulting in a depressurization in the primary loops of the reactor. There is a risk of release of fission products in relation to the size of the failure. Experimental facilities for HTGR tests were developed: an in pile helium loop Comedie and an out of pile helium loop

  16. MTR loop at the MPR-GA. Siwabessy reactor of Serpong Indonesia for testing of LEU fuel

    International Nuclear Information System (INIS)

    Arbie, B.; Sunaryadi, D.; Supadi, S.

    1991-01-01

    The main objective of the MTR-Loop is for testing the specimens of MTR fuel element uprated conditions with respect to the normal conditions of the reactor fuel elements. It is intended to verify the suitability of the fuel elements for operation in a research reactor under preset temperature and pressure conditions. The most important part of the MTR loop is the test section. The fuel elements to be tested are positioned in the test section. For heat removal there is a cooling water flowing through the test section. On this paper the description of the MTR-Loop is described. Installation of the MTR-Loop will be performed in the middle of 1990. In order to facilitate the investigation of fuel behaviour and performance of the new fuel elements the supporting facilities are also already available in the RSG-GAS. (orig.)

  17. Detail design of test loop for FIV in fuel bundle and preliminary test

    Energy Technology Data Exchange (ETDEWEB)

    Sim, Woo Gunl; Lee, Wan Young; Kim, Sung Won [Hannam University, Taejeon (Korea)

    2002-04-01

    It is urgent to develop the analytical model for the structural/mechanical integrity of fuel rod. In general, it is not easy to develop a pure analytical model. Occasionally, experimental results have been utilized for the model.Because of this reason, it is required to design proper test loop. Using the optimized test loop, With the optimized test loop, the dynamic behaviour of the rod will be evaluated and the critical flow velocity, which the rod loses the stability in, will be measured for the design of the rod. To verify the integrity of the fuel rod, it is required to evaluate the dynamic behaviour and the critical flow velocity with the test loop. The test results will be utilized to the design of the rod. Generally, the rod has a ground vibration due to turbulence in wide range of flow velocity and the amplitude of vibration becomes larger by the resonance, in a range of the velocity where occurs vortex. The rod loses stability in critical flow velocity caused by fluid-elastic instability. For the purpose of the present work to perform the conceptional design of the test loop, it is necessary (1) to understand the mechanism of the flow-induced vibration and the related experimental coefficients, (2) to evaluate the existing test loops for improving the loop with design parameters and (3) to decide the design specifications of the major equipments of the loop. 35 refs., 14 figs., 4 tabs. (Author)

  18. Tests in the ATLE loop on the PIUS design

    International Nuclear Information System (INIS)

    Bredolt, U.; Babala, D.; Kemppainen, J.

    1992-01-01

    This paper describes experimental demonstration of the self-protective features of Process Inherent Ultimate Safety (PIUS) design in a large scale test loop in ABB Atoms engineering laboratories. The loop employs real time simulation of core power as a function of coolant conditions in an electrically heated fuel assembly model. System responses to various severe transients were studied. Comparisons were made with predictions of the RIGEL code, which has been developed specifically for study of PIUS type reactors. A comparison between test results and calculated results was made for main state variables such as pressure, temperatures, concentrations, heat fluxes and mass flow rates. The tests have demonstrated the self-protective thermal-hydraulics of pressurized water reactor primary systems designed according to the PIUS principle and verified the capability of the RIGEL code to predict their behavior during severe accidents and in normal operation transients

  19. Demonstration poloidal coil test facility

    International Nuclear Information System (INIS)

    Sato, Masahiko; Kawano, Katumi; Tada, Eisuke

    1989-01-01

    A new compact cryogenic cold compressor was developed by Japan Atomic Energy Research Institute (JAERI) in collaboration with Isikawajima-Harima Heavy Industries Co., Ltd. (IHI) in order to produce the supercritical helium below 4.2 K for Demonstration Poloidal Coils (DPC) which are forced-flow cooled type superconducting pulse coils. This compressor is one of key components for DPC test facility. The cold compressor reduces pressure in liquid helium bath, which contains liquid helium of around 3,000 l, down to 0.5 atm efficiently. Consequently, supercritical helium down to 3.5 K is produced and supplied to the DPC coils. A centrifugal compressor with dynamic gas bearing is selected as a compressor mechanism to realize high adiabatic efficiency and large flow rate. In this performance tests, the compressor was operated for 220 h at saturated condition from 0.5 to 1.0 atm without any failure. High adiabatic efficiency (more than 60 %) is achieved with wide flow range (25-65 g/s) and the design value is fully satisfied. The compressor can rotate up to 80,000 rpm at maximum then the coil supply temperature of supercritical helium is 3.5 K. (author)

  20. Liquefied Gaseous Fuels Spill Test Facility

    International Nuclear Information System (INIS)

    1993-02-01

    The US Department of Energy's liquefied Gaseous Fuels Spill Test Facility is a research and demonstration facility available on a user-fee basis to private and public sector test and training sponsors concerned with safety aspects of hazardous chemicals. Though initially designed to accommodate large liquefied natural gas releases, the Spill Test Facility (STF) can also accommodate hazardous materials training and safety-related testing of most chemicals in commercial use. The STF is located at DOE's Nevada Test Site near Mercury, Nevada, USA. Utilization of the Spill Test Facility provides a unique opportunity for industry and other users to conduct hazardous materials testing and training. The Spill Test Facility is the only facility of its kind for either large- or small-scale testing of hazardous and toxic fluids including wind tunnel testing under controlled conditions. It is ideally suited for test sponsors to develop verified data on prevention, mitigation, clean-up, and environmental effects of toxic and hazardous gaseous liquids. The facility site also supports structured training for hazardous spills, mitigation, and clean-up. Since 1986, the Spill Test Facility has been utilized for releases to evaluate the patterns of dispersion, mitigation techniques, and combustion characteristics of select materials. Use of the facility can also aid users in developing emergency planning under US P.L 99-499, the Superfund Amendments and Reauthorization Act of 1986 (SARA) and other regulations. The Spill Test Facility Program is managed by the US Department of Energy (DOE), Office of Fossil Energy (FE) with the support and assistance of other divisions of US DOE and the US Government. DOE/FE serves as facilitator and business manager for the Spill Test Facility and site. This brief document is designed to acquaint a potential user of the Spill Test Facility with an outline of the procedures and policies associated with the use of the facility

  1. Successful start for new CLIC test facility

    CERN Document Server

    2004-01-01

    A new test facility is being built to study key feasibility issues for a possible future linear collider called CLIC. Commissioning of the first part of the facility began in June 2003 and nominal beam parameters have been achieved already.

  2. High temperature, high pressure gas loop - the Component Flow Test Loop (CFTL)

    International Nuclear Information System (INIS)

    Gat, U.; Sanders, J.P.; Young, H.C.

    1984-01-01

    The high-pressure, high-temperature, gas-circulating Component Flow Test Loop located at Oak Ridge National Laboratory was designed and constructed utilizing Section III of the ASME Boiler and Pressure Vessel Code. The quality assurance program for operating and testing is also based on applicable ASME standards. Power to a total of 5 MW is available to the test section, and an air-cooled heat exchanger rated at 4.4 MW serves as heat sink. The three gas-bearing, completely enclosed gas circulators provide a maximum flow of 0.47 m 3 /s at pressures to 10.7 MPa. The control system allows for fast transients in pressure, power, temperature, and flow; it also supports prolonged unattended steady-state operation. The data acquisition system can access and process 10,000 data points per second. High-temperature gas-cooled reactor components are being tested

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

    International Nuclear Information System (INIS)

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

    2007-02-01

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

  4. LOCA simulation tests in the RD-12 loop with multiple heat channels

    International Nuclear Information System (INIS)

    Ardron, K.H.; McGee, G.R.; Hawley, E.H.

    1985-11-01

    A series of tests has been performed in the RD-12 loop to study the bahaviour of a CANDU-type, primary heat transport system (PHTS) during the blowdown and injection phases of a loss-of-coolant accident (LOCA). Specifically, the tests were used to investigate flow stagnation and refilling of the core following a LOCA. RD-12 is a pressurized water loop with the basic geometry of a CANDU reactor PHTS, but at approximately 1/125 volume scale. The loop consists of U-tube steam generators, pumps, headers, feeders, and heated channels arranged in the symmetrical figure-of-eight configuration of the CANDU PHTS. In the LOCA simulation tests, the loop contained four horizontal heated channels, each containing a seven-element assembly of indirectly heated, fuel-rod simulators. The channels were nominally identical, and were arranged in parallel pairs between the headers in each half-circuit. Tests were carried out using various restricting orifices to represent pipe breaks of different sizes. The break sizes were specifically chosen such that stagnation conditions in the heated channels would be likely to occur. In some tests, the primary pumps were programmed to run down over a 100-s period to simulate a LOCA with simultaneous loss of pump power. Test results showed that, for certain break sizes, periods of low flow occurred in the channels in one half of the loop, leading to flow stratification and sheath temperature excursions. This report reviews the results of two of the tests, and discusses possible mechanisms that may have led to the low channel flow conditions observed in some cases. Plans for future experiments in the larger scale RD-14 facility are outlined. 5 refs

  5. Disassembly and removal of sodium instrumentation test loop

    International Nuclear Information System (INIS)

    Ishikawa, Okinobu; Onojima, Takamitu; Nagai, Keiichi

    2000-07-01

    In 1999, the Sodium Instrumentation Test Loop was disassembled and removed. This report describes the tasks and experiences obtained in removing sodium from a storage tank, disassembling, and cleansing components and related activities. Overall the disassembly, handling and cleansing tasks proceeded as planned and the activities were carried out efficiently and safely. Documentation of the process is meant to establish not only a procedure, but also a guideline for future similar tasks. (author)

  6. Cryogenic test facility at VECC, Kolkata

    International Nuclear Information System (INIS)

    Sarkar, Amit; Bhunia, Uttam; Pradhan, J.; Sur, A.; Bhandari, R.K.; Ranganathan, R.

    2003-01-01

    In view of proposed K-500 superconducting cyclotron project, cryogenic test facility has been set up at the centre. The facility can broadly be categorized into two- a small scale test facility and a large scale test facility. This facility has been utilized for the calibration of liquid helium level probe, cryogenic temperature probe, and I-B plot for a 7 T superconducting magnet. Spiral-shaped superconducting short sample with specific dimension and specially designed stainless steel sample holder has already been developed for the electrical characterisation. The 1/5 th model superconducting coil along with its quench detection circuit and dump resistor has also been developed

  7. The test section of the COSIMA blowdown test facility

    International Nuclear Information System (INIS)

    Bruederle, F.; Hain, K.

    1980-08-01

    The test section of the COSIMA blowdown test facility has been designed as a geometric analogy of the core of a pressurized water reactor for a shortened single fuel rod simulator. Its design and instrumentation together with the whole loop allow to simulate out of pile and trace by measurements the energy and hydraulic conditions arising in a blowdown. Special attention is being given in this report to one particular design problem: the number of load cycles up to incipient cracking of the test section as a pressure vessel containing hot water at high pressures and subjected to extreme rates of temperature variation in excess of 300 K/min. The methods of calculating cyclic loads as specified in the German Technical Rules for Boilers (TRD) have been supplemented in such a way that the number of load cycles up to incipient cracking may now be determined not only by the mean wall temperature, which is difficult to measure, but equally also well by the outer wall temperature, which is easy to measure precisely. (orig.) [de

  8. Analysis of ATLAS LTC-04R Test for Loop Seal Reformation Phenomena using RELAP5

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Sang-Gyu; Kim, Dae-Hun; Kim, Han-Gon [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    The loop seal reformation issue was selected to be the analysis topic of the DSP-04 based on the technical discussion between the participants and the operating agencies (KAERI and KINS) and domestic experts meetings. After that, KAERI performed LTC-04R test which is 4 inch top-slot cold-leg break test using ATLAS facility in December 27, 2015. KHNP CRI, as a participant of the DSP-04, performed the blind calculation and open calculation using RELAP5/Mod3.3 patch 3. This paper deals with the results of open calculation for ATLAS LTC-04R test. The results of several sensitivity analyses such as the critical flow modeling sensitivity and break flow system modeling sensitivity will be discussed. Several possible factors in the loop seal reformation behavior are examined in the sensitivity analysis. Heat loss modeling, fine break system modeling, fine loop seal nodalization and off-take modeling are not significant factor in the loop seal reformation. Still critical flow model and discharge coefficient are dominant factors. Based on the ATLAS LTC-04R, Ransom-Trapp model shows better prediction in the break flow than the Henry-Fauske model.

  9. The design of in-pile test section for fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Park, K. N.; Lee, J. M.; Shim, B. S.; Zee, D. Y.; Park, S. H.; Ahn, S. H.; Lee, J. Y.; Kim, Y. J. [KAERI, Taejon (Korea, Republic of)

    2004-07-01

    As an equipment for nuclear fuel's general performance irradiation test in HANARO, Fuel Test Loop(FTL) has been developed that can irradiate the pin to the maximum number of 3 at the core irradiation hole(IR1 hole) by considering for it's utility and user's irradiation requirement. 3-Pin FTL consists of In-Pile Test Section (IPS) and Out-of-Pile System (OPS). IPS consists for IPS Vessel assembly, In-Pool Piping, IPS Support, In-Pool Piping Support etc. Design that such IPS considers interference item consisted to do not bear in existing facilities by one. IVA that is connected to the OPS are controlled and regulated by means of system pressure, system temperature and the water quality. IPS Vessel assembly is consisted of outer pressure vessel, inner pressure vessel, IPS head, inner assembly and test fuel carrier. After 3-Pin FTL development which is expected to be finished by the 2006, FTL will be used for the irradiation test of the new PWR-type fuel and can maximize the usage of HANARO.

  10. Performance test of ex-core high temperature and high pressure water loop test equipment (Contract research)

    International Nuclear Information System (INIS)

    Nakano, Hiroko; Uehara, Toshiaki; Takeuchi, Tomoaki; Shibata, Hiroshi; Nakamura, Jinichi; Matsui, Yoshinori; Tsuchiya, Kunihiko

    2016-03-01

    In Japan Atomic Energy Agency, we started research and development so as to monitor the situations in the Nuclear Plant Facilities during a severe accident, such as a radiation-resistant monitoring camera, a radiation-resistant transmission system for conveying the in-core information, and a heat-resistant signal cable. As a part of developments of the heat-resistant signal cable, we prepared ex-core high-temperature and high-pressure water loop test equipment, which can simulate the conditions of BWRs and PWRs, for evaluating reliability and properties of sheath materials of the cable. This equipment consists of autoclave, water conditioning tank, high-pressure metering pump, preheater, heat exchanger and water purification equipment, etc. This report describes the basic design and the performance test results of ex-core high-temperature and high-pressure water loop test equipment. (author)

  11. Argonne's new Wakefield Test Facility

    International Nuclear Information System (INIS)

    Simpson, J.D.

    1992-01-01

    The first phase of a high current, short bunch length electron beam research facility, the AWA, is near completion at Argonne. At the heart of the facility is a photocathode based electron gun and accelerating sections designed to deliver 20 MeV pulses with up to 100 nC per pulse and with pulse lengths of approximately 15 ps (fw). Using a technique similar to that originated at Argonne's AATF facility, a separate weak probe pulse can be generated and used to diagnose wake effects produced by the intense pulses. Initial planned experiments include studies of plasma wakefields and dielectric wakefield devices, and expect to demonstrate large, useful accelerating gradients (> 100 MeV/m). Later phases of the facility will increase the drive bunch energy to more than 100 MeV to enable acceleration experiments up to the GeV range. Specifications, design details, and commissioning progress are presented

  12. Importance of tests in nuclear facilities

    International Nuclear Information System (INIS)

    Guillemard, B.

    1985-10-01

    In nuclear facilities, safety related systems and equipments are subject, along their whole service-life, to numerous tests. This paper analyses the role of tests in the successive stages of design, construction, exploitation of a nuclear facility. It examines several aspects of test quality control: definition of needs, test planning, intrinsic quality of each test, control of interfaces (test are both the end and the starting point of many actions concerned by quality) and the application [fr

  13. Fast Flux Test Facility (FFTF) maintenance provisions

    International Nuclear Information System (INIS)

    Marshall, J.L.

    1981-05-01

    The Fast Flux Test Facility (FFTF) was designed with maintainability as a primary parameter, and facilities and provisions were designed into the plant to accommodate the maintenance function. This paper describes the FFTF and its systems. Special maintenance equipment and facilities for performing maintenance on radioactive components are discussed. Maintenance provisions designed into the plant to enhance maintainability are also described

  14. Controlled Archaeological Test Site (CATS) Facility

    Data.gov (United States)

    Federal Laboratory Consortium — CATS facility is at the Construction Engineering Research Laboratory (CERL), Champaign, IL. This 1-acre test site includes a variety of subsurface features carefully...

  15. Customer interface document for the Molten Salt Test Loop (MSTL) system.

    Energy Technology Data Exchange (ETDEWEB)

    Pettit, Kathleen; Kolb, William J.; Gill, David Dennis; Briggs, Ronald D.

    2012-03-01

    The National Solar Thermal Test Facility at Sandia National Laboratories has a unique test capability called the Molten Salt Test Loop (MSTL) system. MSTL is a test capability that allows customers and researchers to test components in flowing, molten nitrate salt. The components tested can range from materials samples, to individual components such as flex hoses, ball joints, and valves, up to full solar collecting systems such as central receiver panels, parabolic troughs, or linear Fresnel systems. MSTL provides realistic conditions similar to a portion of a concentrating solar power facility. The facility currently uses 60/40 nitrate 'solar salt' and can circulate the salt at pressure up to 600psi, temperature to 585 C, and flow rate of 400-600GPM depending on temperature. The purpose of this document is to provide a basis for customers to evaluate the applicability to their testing needs, and to provide an outline of expectations for conducting testing on MSTL. The document can serve as the basis for testing agreements including Work for Others (WFO) and Cooperative Research and Development Agreements (CRADA). While this document provides the basis for these agreements and describes some of the requirements for testing using MSTL and on the site at Sandia, the document is not sufficient by itself as a test agreement. The document, however, does provide customers with a uniform set of information to begin the test planning process.

  16. Loop corrections and a new test of inflation

    CERN Document Server

    Tasinato, Gianmassimo; Nurmi, Sami; Wands, David

    2013-01-01

    Inflation is the leading paradigm for explaining the origin of primordial density perturbations and the observed temperature fluctuations of the cosmic microwave background. However many open questions remain, in particular whether one or more scalar fields were present during inflation and how they contributed to the primordial density perturbation. We propose a new observational test of whether multiple fields, or only one (not necessarily the inflaton) generated the perturbations. We show that our test, relating the bispectrum and trispectrum, is protected against loop corrections at all orders, unlike previous relations.

  17. MES lead bismuth forced circulation loop and test results

    International Nuclear Information System (INIS)

    Ono, Mikinori; Mine, Tatsuya; Kitano, Teruaki; Kamata, Kin-ya

    2003-01-01

    Liquid lead-bismuth is a promising material as future reactor coolant or intensive neutron source material for accelerator driven system (ADS). Mitsui Engineering and Shipbuilding Co., Ltd. (MES) completed lead-bismuth coolant (LBC) forced circulation loop in May 2001 and acquired engineering data on economizer, electro magnetic pump, electro magnetic flow meter and so on. For quality control of LBC, oxygen sensor and filtering element are developing using some hydrogen and moisture mixed gases. Structural materials corrosion test for accelerator driver system (ADS) will start soon. And thermal hydraulic test for ADS will start in tree years. (author)

  18. Ten years operating experience at the Fast Flux Test Facility: A decade of excellence

    International Nuclear Information System (INIS)

    Swaim, D.J.; Waldo, J.B.; Farabee, O.A.

    1991-07-01

    The Fast Flux Test Facility is a 400 MW(t) fast reactor cooled by three sodium loops. The Fast Flux Test Facility is managed by the Westinghouse Hanford Company for the US Department of Energy. The Fast Flux Test Facility was designed and constructed to provide irradiation testing of fuels and materials for the US Department of Energy Liquid Metal Reactor research program. Facility activities have increased to include fusion power materials testing, passive safety testing, isotope production, and international collaboration. 5 figs

  19. Design criteria and fabrication in-pile test section of HANARO fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. Y.

    1997-10-01

    Safety state fuel test loop will be equipped in HANARO to obtain the development and betterments of advanced fuel and materials through the irradiation tests. The objective of this study is to determine the design criteria and technical specification of in-pile test section and to specify the manufacturing requirements of in-pile test section. HANARO fuel test loop was designed to meet the CANDU and PWR fuel testing and in-pile section will be manufactured and installed in HANARO. The design criteria and technical specification of in-pile test section could be used the fuel and materials design with for irradiation testing IPS of HANARO fuel test loop. This results will become guidances for the planning and programming of irradiation testing. (author). 12 refs., tabs., figs.

  20. Design and study of Engineering Test Facility - Helium Circulator

    International Nuclear Information System (INIS)

    Jiang Huijing; Ye Ping; Zhao Gang; Geng Yinan; Wang Jie

    2015-01-01

    Helium circulator is one of the key equipment of High-temperature Gas-cooled Reactor Pebble-bed Module (HTR-PM). In order to simulate most normal and accident operating conditions of helium circulator in HTR-PM, a full scale, rated flow rate and power, engineering test loop, which was called Engineering Test Facility - Helium Circulator (ETF-HC), was designed and established. Two prototypes of helium circulator, which was supported by Active Magnetic Bearing (AMB) or sealed by dry gas seals, would be tested on ETF-HC. Therefore, special interchangeable design was under consideration. ETF-HC was constructed compactly, which consisted of eleven sub-systems. In order to reduce the flow resistance of the circuit, special ducts, elbows, valves and flowmeters were selected. Two stages of heat exchange loops were designed and a helium - high pressure pure water heat exchanger was applied to ensure water wouldn't be vaporized while simulating accident conditions. Commissioning tests were carried out and operation results showed that ETF-HC meets the requirement of helium circulator operation. On this test facility, different kinds of experiments were supposed to be held, including mechanical and aerodynamic performance tests, durability tests and so on. These tests would provide the features and performance of helium circulator and verify its feasibility, availability and reliability. (author)

  1. Systematic Unit Testing in a Read-eval-print Loop

    DEFF Research Database (Denmark)

    Nørmark, Kurt

    2010-01-01

    .  The process of collecting the expressions and their results imposes only little extra work on the programmer.  The use of the tool provides for creation of test repositories, and it is intended to catalyze a much more systematic approach to unit testing in a read-eval-print loop.  In the paper we also discuss...... how to use a test repository for other purposes than testing.  As a concrete contribution we show how to use test cases as examples in library interface documentation.  It is hypothesized---but not yet validated---that the tool will motivate the Lisp programmer to take the transition from casual...

  2. Subsonic Constant-Area MHD Generator Experiments with the CNEN Blow-Down Loop Facility

    Energy Technology Data Exchange (ETDEWEB)

    Bertolini, E.; Gasparotto, M.; Gay, P.; Toschi, R. [Laboratorio Conversione Diretta, CNEN, Frascati (Italy)

    1968-11-15

    The design of the facility, described at the Salzburg Symposium, was somewhat modified following the results of the commissioning tests; the changes were mainly concerned with the thermal insulation, duct materials and caesium recovery system. The facility went into full operation in March 1967 and since then two series of MHD experiments, a total of twenty-six runs, have been performed. During the MHD runs the facility has been working mostly under the following operating conditions: stagnation temperature 1500 to 1800 Degree-Sign K; stagnation pressure-1 to 3 atm. abs.; mass How 50 to 150 g/sec; seeding 2 to 5 at.%- ; magnetic field 0 to 45 k G; Mach number 0.4 to 0.8; Hall parameter up to 6. The main purpose of the experiments was to study the performance of relatively small generators (cross-section 3 x 5 cm{sup 2}, length 8-20 cm) both when the non-equilibrium ionization is expected to be negligible and when it should be, in a very idealized model, relevant. As a first step, efforts were made to ascertain whether any of the unsatisfactory results reported in Salzburg, both for equilibrium and non-equilibrium generators, stemmed not from the basic functioning principle of an MHD small-scale generator but rather from some inadequacy of the experimental apparatus. Therefore particular attention was paid to: ceasium vaporization and mixing with helium; plasma insulation from ground; electrical insulation from ground and from each other of those electrically conductive parts of the facility which may, during the functioning, come into contact with the plasma; temperature control of the duct; purity level; duct materials; measurement system and control. In the equilibrium regime the Faraday field measured is very close to the ideal value and it reaches 80 V/cm (400 volts between electrodes); the Hall field still remains below the ideal value uB{beta}L (50% at {beta} = 3). The maximum Hall field was about 35 V/cm for a corresponding voltage of 600 V. Preionization

  3. Large scale sodium interactions. Part 1. Test facility design

    International Nuclear Information System (INIS)

    King, D.L.; Smaardyk, J.E.; Sallach, R.A.

    1977-01-01

    During the design of the test facility for large scale sodium interaction testing, an attempt was made to keep the system as simple and yet versatile as possible; therefore, a once through design was employed as opposed to any type of conventional sodium ''loop.'' The initial series of tests conducted at the facility call for rapidly dropping from 20 kg to 225 kg of sodium at temperatures from 825 0 K to 1125 0 K into concrete crucibles. The basic system layout is described. A commercial drum heater is used to melt the sodium which is in 55 gallon drums and then a slight argon pressurization is used to force the liquid sodium through a metallic filter and into a dump tank. Then the sodium dump tank is heated to the desired temperature. A diaphragm is mechanically ruptured and the sodium is dumped into a crucible that is housed inside a large steel test chamber

  4. Facility for generating crew waste water product for ECLSS testing

    Science.gov (United States)

    Buitekant, Alan; Roberts, Barry C.

    1990-01-01

    An End-use Equipment Facility (EEF) has been constructed which is used to simulate water interfaces between the Space Station Freedom Environmental Control and Life Support Systems (ECLSS) and man systems. The EEF is used to generate waste water to be treated by ECLSS water recovery systems. The EEF will also be used to close the water recovery loop by allowing test subjects to use recovered hygiene and potable water during several phases of testing. This paper describes the design and basic operation of the EEF.

  5. Helium Loop for the HCPB Test Blanket Module

    International Nuclear Information System (INIS)

    Neuberger, H.; Boccaccini, L.V.; Ghidersa, B. E.; Jin, X.; Meyder, R.

    2006-01-01

    In the frame of the activities of the EU Breeder Blanket Programme and of the Test Blanket Working Group, the Helium loop for the Helium Cooled Pebble Bed Test Blanket Module (HCPB-TBM) in ITER has been investigated with regard to the layout definition, selection of components, control, dimensioning and integration. This paper presents the status of development. The loop design for the HCPB-TBM in ITER will mainly base on the experience gained from Helium Loop Karlsruhe (HELOKA) which is currently developed at the FZK for experiments under ITER relevant conditions. The ITER loop will be equipped with similar components like HELOKA and will mainly consist of a circulator with variable speed drive, a recuperator, an electric heater, a cooler, a dust filter and auxilary components e.g. pipework and valves. A Coolant Purification System (CPS) and a Pressure Control System (PCS) are foreseen to meet the requirements on coolant conditioning. To prepare a TBM for a new experimental campaign, a succession of operational states like '' cold maintenance '', '' baking '' and '' cold standby '' is required. Before a pulse operation, a '' hot stand-by '' state should be achieved providing the TBM with inlet coolant at nominal conditions. This operation modus is continued in the dwell time waiting for the successive pulse. A '' tritium out-gassing '' will be also required after several TBM-campaigns to remove the inventory rest of T in the beds for measurement purpose. The dynamic circuit behaviour during pulses, transition between different operational states as well as the behaviour in accident situations are investigated with RELAP. The main components of the loop will be accommodated inside the Tokamak Cooling Water System(TCWS)- vault from where the pipes require connection to the TBM which is attached to port 16 of the vacuum vessel. Therefore pipes across the ITER- building of about 110 m in length (each) are required. Additional equipment is also located in the port cell

  6. Solar cooling in the hardware-in-the-loop test; Solare Kuehlung im Hardware-in-the-Loop-Test

    Energy Technology Data Exchange (ETDEWEB)

    Lohmann, Sandra; Radosavljevic, Rada; Goebel, Johannes; Gottschald, Jonas; Adam, Mario [Fachhochschule Duesseldorf (Germany). Erneuerbare Energien und Energieeffizienz E2

    2012-07-01

    The first part of the BMBF-funded research project 'Solar cooling in the hardware-in-the-loop test' (SoCool HIL) deals with the simulation of a solar refrigeration system using the simulation environment Matlab / Simulink with the toolboxes Stateflow and Carnot. Dynamic annual simulations and DoE supported parameter variations were used to select meaningful system configurations, control strategies and dimensioning of components. The second part of this project deals with hardware-in-the-loop tests using the 17.5 kW absorption chiller of the company Yazaki Europe Limited (Hertfordshire, United Kingdom). For this, the chiller is operated on a test bench in order to emulate the behavior of other system components (solar circuit with heat storage, recooling, buildings and cooling distribution / transfer). The chiller is controlled by a simulation of the system using MATLAB / Simulink / Carnot. Based on the knowledge on the real dynamic performance of the chiller the simulation model of the chiller can then be validated. Further tests are used to optimize the control of the chiller to the current cooling load. In addition, some changes in system configurations (for example cold backup) are tested with the real machine. The results of these tests and the findings on the dynamic performance of the chiller are presented.

  7. Detection Test for Leakage of CO{sub 2} into Sodium Loop

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sun Hee; Wi, Myung-Hwan; Min, Jae Hong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    This report is about the facility for the detection test for leakage of CO{sub 2} into sodium loop. The facility for the detection test for leakage of CO{sub 2} into sodium loop was introduced. The test will be carried out. Our experimental results are going to be expected to be used for approach methods to detect CO{sub 2} leaking into sodium in heat exchangers. A sodium-and-carbon dioxide (Na-CO{sub 2}) heat exchanger is one of the key components for the supercritical CO{sub 2} Brayton cycle power conversion system of sodium-cooled fast reactors (SFRs). A printed circuit heat exchanger (PCHE) is considered for the Na-CO{sub 2} heat exchanger, which is known to have potential for reducing the volume occupied by the exchangers compared to traditional shell-and-tube heat exchangers. Among various issues about the Na- CO{sub 2} exchanger, detection of CO{sub 2} leaking into sodium in the heat exchanger is most important thing for its safe operation. It is known that reaction products from sodium and CO{sub 2} such as sodium carbonate (Na{sub 2}CO{sub 3}) and amorphous carbon are hardly soluble in sodium, which cause plug sodium channels. Detection technique for Na{sub 2}CO{sub 3} in sodium loop has not been developed yet. Therefore, detection of CO{sub 2} and CO from reaction of sodium and CO{sub 2} are proper to detect CO{sub 2} leakage into sodium loop.

  8. Engineered Barrier Test Facility status report, 1984

    International Nuclear Information System (INIS)

    Phillips, S.J.; Adams, M.R.; Gilbert, T.W.; Meinhardt, C.C.; Mitchell, R.M.; Waugh, W.J.

    1985-02-01

    This report provides a general summary of activities completed to date at the Hanford Engineered Barrier Test Facility. This facility is used to test and compare construction practices and performance of alternative designs of engineered barrier cover systems. These cover systems are being evaluated for potential use for isolation and confinement of buried waste disposal structures

  9. High-pressure test loop design and application

    International Nuclear Information System (INIS)

    Burnette, R.D.; Graves, J.N.; Blair, P.G.; Baldwin, N.L.

    1980-07-01

    A high-pressure test loop (HPTL) has been constructed for the purpose of performing a number of chemistry experiments at simulated HTGR conditions of temperature, pressure, flow, and impurity content. The HPTL can be used to develop, modify, and verify computer codes for a variety of chemical processes involving gas phase transport in the reactor. Processes such as graphite oxidation, fission product transport, fuel reactions, purification systems, and dust entrainment can be studied at high pressure, which would largely eliminate difficulties in correlating existing laboratory data and reactor conditions

  10. SRF cavity testing using a FPGA Self Excited Loop

    CERN Document Server

    Ben-Zvi, Ilan

    2018-01-01

    This document provides a detailed description of procedures for very-high precision calibration and testing of superconducting RF cavities using digital Low-Level RF (LLRF) electronics based on Field Programmable Gate Arrays (FPGA). The use of a Self-Excited Loop with an innovative procedure for fast turn-on allows the measurement of the forward, reflected and transmitted power from a single port of the directional coupler in front of the cavity, thus eliminating certain measurement errors. Various procedures for measuring the quality factor as a function of cavity fields are described, including a single RF pulse technique. Errors are estimated for the measurements.

  11. Energy Systems Test Area (ESTA). Power Systems Test Facilities

    Science.gov (United States)

    Situ, Cindy H.

    2010-01-01

    This viewgraph presentation provides a detailed description of the Johnson Space Center's Power Systems Facility located in the Energy Systems Test Area (ESTA). Facilities and the resources used to support power and battery systems testing are also shown. The contents include: 1) Power Testing; 2) Power Test Equipment Capabilities Summary; 3) Source/Load; 4) Battery Facilities; 5) Battery Test Equipment Capabilities Summary; 6) Battery Testing; 7) Performance Test Equipment; 8) Battery Test Environments; 9) Battery Abuse Chambers; 10) Battery Abuse Capabilities; and 11) Battery Test Area Resources.

  12. Preliminary considerations of an intense slow positron facility based on a 78Kr loop in the high flux isotopes reactor

    International Nuclear Information System (INIS)

    Hulett, L.D. Jr.; Donohue, D.L.; Peretz, F.J.; Montgomery, B.H.; Hayter, J.B.

    1990-01-01

    Suggestions have been made to the National Steering Committee for the Advanced Neutron Source (ANS) by Mills that provisions be made to install a high intensity slow positron facility, based on a 78 Kr loop, that would be available to the general community of scientists interested in this field. The flux of thermal neutrons calculated for the ANS is E + 15 sec -1 m -2 , which Mills has estimated will produce 5 mm beam of slow positrons having a current of about 1 E + 12 sec -1 . The intensity of such a beam will be a least 3 orders of magnitude greater than those presently available. The construction of the ANS is not anticipated to be complete until the year 2000. In order to properly plan the design of the ANS, strong considerations are being given to a proof-of-principle experiment, using the presently available High Flux Isotopes Reactor, to test the 78 Kr loop technique. The positron current from the HFIR facility is expected to be about 1 E + 10 sec -1 , which is 2 orders of magnitude greater than any other available. If the experiment succeeds, a very valuable facility will be established, and important formation will be generated on how the ANS should be designed. 3 refs., 1 fig

  13. Two-phase flow patterns recognition and parameters estimation through natural circulation test loop image analysis

    Energy Technology Data Exchange (ETDEWEB)

    Mesquita, R.N.; Libardi, R.M.P.; Masotti, P.H.F.; Sabundjian, G.; Andrade, D.A.; Umbehaun, P.E.; Torres, W.M.; Conti, T.N.; Macedo, L.A. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil). Nuclear Engineering Center], e-mail: rnavarro@ipen.br

    2009-07-01

    Visualization of natural circulation test loop cycles is used to study two-phase flow patterns associated with phase transients and static instabilities of flow. Experimental studies on natural circulation flow were originally related to accidents and transient simulations relative to nuclear reactor systems with light water refrigeration. In this regime, fluid circulation is mainly caused by a driving force ('thermal head') which arises from density differences due to temperature gradient. Natural circulation phenomenon has been important to provide residual heat removal in cases of 'loss of pump power' or plant shutdown in nuclear power plant accidents. The new generation of compact nuclear reactors includes natural circulation of their refrigerant fluid as a security mechanism in their projects. Two-phase flow patterns have been studied for many decades, and the related instabilities have been object of special attention recently. Experimental facility is an all glass-made cylindrical tubes loop which contains about twelve demineralized water liters, a heat source by an electrical resistor immersion heater controlled by a Variac, and a helicoidal heat exchanger working as cold source. Data is obtained through thermo-pairs distributed over the loop and CCD cameras. Artificial intelligence based algorithms are used to improve (bubble) border detection and patterns recognition, in order to estimate and characterize, phase transitions patterns and correlate them with the periodic static instability (chugging) cycle observed in this circuit. Most of initial results show good agreement with previous numerical studies in this same facility. (author)

  14. Two-phase flow patterns recognition and parameters estimation through natural circulation test loop image analysis

    International Nuclear Information System (INIS)

    Mesquita, R.N.; Libardi, R.M.P.; Masotti, P.H.F.; Sabundjian, G.; Andrade, D.A.; Umbehaun, P.E.; Torres, W.M.; Conti, T.N.; Macedo, L.A.

    2009-01-01

    Visualization of natural circulation test loop cycles is used to study two-phase flow patterns associated with phase transients and static instabilities of flow. Experimental studies on natural circulation flow were originally related to accidents and transient simulations relative to nuclear reactor systems with light water refrigeration. In this regime, fluid circulation is mainly caused by a driving force ('thermal head') which arises from density differences due to temperature gradient. Natural circulation phenomenon has been important to provide residual heat removal in cases of 'loss of pump power' or plant shutdown in nuclear power plant accidents. The new generation of compact nuclear reactors includes natural circulation of their refrigerant fluid as a security mechanism in their projects. Two-phase flow patterns have been studied for many decades, and the related instabilities have been object of special attention recently. Experimental facility is an all glass-made cylindrical tubes loop which contains about twelve demineralized water liters, a heat source by an electrical resistor immersion heater controlled by a Variac, and a helicoidal heat exchanger working as cold source. Data is obtained through thermo-pairs distributed over the loop and CCD cameras. Artificial intelligence based algorithms are used to improve (bubble) border detection and patterns recognition, in order to estimate and characterize, phase transitions patterns and correlate them with the periodic static instability (chugging) cycle observed in this circuit. Most of initial results show good agreement with previous numerical studies in this same facility. (author)

  15. I and C functional test facility user guide

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Ki Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-07-01

    The objective of I and C functional test facility (FTF) is to validate newly developed digital control and protection algorithm, alarm reduction algorithm and the function of operator support system and so on. Test facility is divided into three major parts; software, hardware and graphic user interface. Software consists of mathematical modeling which simulates 3 loop pressurizer water reactor, 993 MWe Westinghouse plant and supervisory module which interpret user instructions and data interface program. FTF is implemented in HP747I workstation using FORTRAN77 and ``C`` language under UNIX operating system. This User Guide provides file structure, instructions and program modification method and provides initial data, malfunction list, process variables list and simulation diagram as an appendix to test developed prototype. 12 figs. (Author).

  16. I and C functional test facility user guide

    International Nuclear Information System (INIS)

    Kwon, Ki Chun

    1996-07-01

    The objective of I and C functional test facility (FTF) is to validate newly developed digital control and protection algorithm, alarm reduction algorithm and the function of operator support system and so on. Test facility is divided into three major parts; software, hardware and graphic user interface. Software consists of mathematical modeling which simulates 3 loop pressurizer water reactor, 993 MWe Westinghouse plant and supervisory module which interpret user instructions and data interface program. FTF is implemented in HP747I workstation using FORTRAN77 and ''C'' language under UNIX operating system. This User Guide provides file structure, instructions and program modification method and provides initial data, malfunction list, process variables list and simulation diagram as an appendix to test developed prototype. 12 figs. (Author)

  17. EPRI flow-loop/in situ test program for motor-operated valves

    International Nuclear Information System (INIS)

    Hosler, J.F.; Dorfman, L.S.

    1994-01-01

    The Electric Power Research Institute is undertaking a comprehensive research program to develop and validate methods for predicting the performance of common motor-operated gate, global, and butterfly valves. To assess motor-operated valve (MOV) performance characteristics and provide a basis for methods validation, full-scale testing was conducted on 62 MOVs. Tests were performed in four flow-loop facilities and in nine nuclear units. Forty-seven gate, five globe, and 10 butterfly valves were tested under a wide range of flow and differential pressure conditions. The paper describes the test program scope, test configurations, instrumentation and data acquisition, testing approach, and data analysis methods. Key results are summarized

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-12-15

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

  19. CryoModule Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — CMTFis able to test complete SRF cryomodules at cryogenic operating temperatures and with RF Power. CMTF will house the PIP-II Injector Experiment allowing test of...

  20. Naval Aerodynamics Test Facility (NATF)

    Data.gov (United States)

    Federal Laboratory Consortium — The NATF specializes in Aerodynamics testing of scaled and fullsized Naval models, research into flow physics found on US Navy planes and ships, aerosol testing and...

  1. Preliminary Design of the Liquid Lead Corrosion Test Loop

    International Nuclear Information System (INIS)

    Cho, Chung Ho; Cha, Jae Eun; Cho, Choon Ho; Song, Tae Yung; Kim, Hee Reyoung

    2005-01-01

    Recently, Lead-Bismuth Eutectic (LBE) or Lead has newly attracted considerable attraction as a coolant to get the more inherent safety. Above all, LBE is preferred as the coolant and target material for an Accelerator-Driven System (ADS) due to its high production rate of neutrons, effective heat removal, and good radiation damage properties. But, the LBE or Lead as a coolant has a challenging problem that the LBE or Lead is more corrosive to the construction materials and fuel cladding material than the sodium because the solubility of Ni, Cr and Fe is high. After all, the LBE or Lead corrosion has been considered as an important design limit factor of ADS and Liquid Metal cooled Fast Reactors (LMFR). The Korea Atomic Energy Research Institute (KAERI) has been developing an ADS called HYPER. HYPER is designed to transmute Transuranics (TRU), Tc-99 and I-129 coming from Pressurized Water Reactors (PWRs) and uses an LBE as a coolant and target material. Also, an experimental apparatuses for the compatibility of fuel cladding and structural material with the LBE or Lead are being under the construction or design. The main objective of the present paper is introduction of Lead corrosion test loop which will be built the upside of the LBE corrosion test loop by the end of October of 2005

  2. Efficiency of the pre-heater against flow rate on primary the beta test loop

    International Nuclear Information System (INIS)

    Edy Sumarno; Kiswanta; Bambang Heru; Ainur R; Joko P

    2013-01-01

    Calculation of efficiency of the pre-heater has been carried out against the flow rate on primary the BETA Test Loop. BETA test loop (UUB) is a facilities of experiments to study the thermal hydraulic phenomenon, especially for thermal hydraulic post-LOCA (Lost of Coolant Accident). Sequences removal on the BETA Test Loop contained a pre-heater that serves as a getter heat from the primary side to the secondary side, determination of efficiency is to compare the incoming heat energy with the energy taken out by a secondary fluid. Characterization is intended to determine the performance of a pre-heater, then used as tool for analysis, and as a reference design experiments. Calculation of efficiency methods performed by operating the pre-heater with fluid flow rate variation on the primary side. Calculation of efficiency on the results obtained that the efficiency change with every change of flow rate, the flow rate is 71.26% on 163.50 ml/s and 60.65% on 850.90 ml/s. Efficiency value can be even greater if the pre-heater tank is wrapped with thermal insulation so there is no heat leakage. (author)

  3. Construction and commissioning test report of the CEDM test facility

    Energy Technology Data Exchange (ETDEWEB)

    Chung, C. H.; Kim, J. T.; Park, W. M.; Youn, Y. J.; Jun, H. G.; Choi, N. H.; Park, J. K.; Song, C. H.; Lee, S. H.; Park, J. K

    2001-02-01

    The test facility for performance verification of the control element drive mechanism (CEDM) of next generation power plant was installed at the site of KAERI. The CEDM was featured a mechanism consisting of complicated mechanical parts and electromagnetic control system. Thus, a new CEDM design should go through performance verification tests prior to it's application in a reactor. The test facility can simulate the reactor operating conditions such as temperature, pressure and water quality and is equipped with a test chamber to accomodate a CEDM as installed in the power plant. This test facility can be used for the following tests; endurance test, coil cooling test, power measurement and reactivity rod drop test. The commissioning tests for the test facility were performed up to the CEDM test conditions of 320 C and 150 bar, and required water chemistry was obtained by operating the on-line water treatment system.

  4. Construction and commissioning test report of the CEDM test facility

    International Nuclear Information System (INIS)

    Chung, C. H.; Kim, J. T.; Park, W. M.; Youn, Y. J.; Jun, H. G.; Choi, N. H.; Park, J. K.; Song, C. H.; Lee, S. H.; Park, J. K.

    2001-02-01

    The test facility for performance verification of the control element drive mechanism (CEDM) of next generation power plant was installed at the site of KAERI. The CEDM was featured a mechanism consisting of complicated mechanical parts and electromagnetic control system. Thus, a new CEDM design should go through performance verification tests prior to it's application in a reactor. The test facility can simulate the reactor operating conditions such as temperature, pressure and water quality and is equipped with a test chamber to accomodate a CEDM as installed in the power plant. This test facility can be used for the following tests; endurance test, coil cooling test, power measurement and reactivity rod drop test. The commissioning tests for the test facility were performed up to the CEDM test conditions of 320 C and 150 bar, and required water chemistry was obtained by operating the on-line water treatment system

  5. Buffet test in the National Transonic Facility

    Science.gov (United States)

    Young, Clarence P., Jr.; Hergert, Dennis W.; Butler, Thomas W.; Herring, Fred M.

    1992-01-01

    A buffet test of a commercial transport model was accomplished in the National Transonic Facility at the NASA Langley Research Center. This aeroelastic test was unprecedented for this wind tunnel and posed a high risk to the facility. This paper presents the test results from a structural dynamics and aeroelastic response point of view and describes the activities required for the safety analysis and risk assessment. The test was conducted in the same manner as a flutter test and employed onboard dynamic instrumentation, real time dynamic data monitoring, automatic, and manual tunnel interlock systems for protecting the model. The procedures and test techniques employed for this test are expected to serve as the basis for future aeroelastic testing in the National Transonic Facility. This test program was a cooperative effort between the Boeing Commercial Airplane Company and the NASA Langley Research Center.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-04-01

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

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

    International Nuclear Information System (INIS)

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

    2008-01-01

    A test facility, referred to as the High Temperature Gas-Cooled Reactor Component Test Facility or CTF, will be sited at Idaho National Laboratory for the purposes of supporting development of high temperature gas thermal-hydraulic technologies (helium, helium-Nitrogen, CO2, etc.) as applied in heat transport and heat transfer applications in High Temperature Gas-Cooled Reactors. Such applications include, but are not limited to: primary coolant; secondary coolant; intermediate, secondary, and tertiary heat transfer; and demonstration of processes requiring high temperatures such as hydrogen production. The facility will initially support completion of the Next Generation Nuclear Plant. It will secondarily be open for use by the full range of suppliers, end-users, facilitators, government laboratories, and others in the domestic and international community supporting the development and application of High Temperature Gas-Cooled Reactor technology. This pre-conceptual facility configuration study, which forms the basis for a cost estimate to support CTF scoping and planning, accomplishes the following objectives: (1) Identifies pre-conceptual design requirements; (2) Develops test loop equipment schematics and layout; (3) Identifies space allocations for each of the facility functions, as required; (4) Develops a pre-conceptual site layout including transportation, parking and support structures, and railway systems; (5) Identifies pre-conceptual utility and support system needs; and (6) Establishes pre-conceptual electrical one-line drawings and schedule for development of power needs

  8. Battery Post-Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — Post-test diagnostics of aged batteries can provide additional information regarding the cause of performance degradation, which, previously, could be only inferred...

  9. Ballast Water Treatment Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — FUNCTION: Provides functionality for the full-scale testing and controlled simulation of ship ballasting operations for assessment of aquatic nuisance species (ANS)...

  10. Testing lifting systems in nuclear facilities

    International Nuclear Information System (INIS)

    Kling, H.; Laug, R.

    1984-01-01

    Lifting systems in nuclear facilities must be inspected at regular intervals after having undergone their first acceptance test. These inspections are frequently carried out by service firms which not only employ the skilled personnel required for such jobs but also make available the necessary test equipment. The inspections in particular include a number of sophisticated load tests for which test load systems have been developed to allow lifting systems to be tested so that reactor specific boundary conditions are taken into account. In view of the large number of facilities to be inspected, the test load system is a modular system. (orig.) [de

  11. E-4 Test Facility Design Status

    Science.gov (United States)

    Ryan, Harry; Canady, Randy; Sewell, Dale; Rahman, Shamim; Gilbrech, Rick

    2001-01-01

    Combined-cycle propulsion technology is a strong candidate for meeting NASA space transportation goals. Extensive ground testing of integrated air-breathing/rocket system (e.g., components, subsystems and engine systems) across all propulsion operational modes (e.g., ramjet, scramjet) will be needed to demonstrate this propulsion technology. Ground testing will occur at various test centers based on each center's expertise. Testing at the NASA John C. Stennis Space Center will be primarily concentrated on combined-cycle power pack and engine systems at sea level conditions at a dedicated test facility, E-4. This paper highlights the status of the SSC E-4 test Facility design.

  12. The PANDA facility and first test results

    International Nuclear Information System (INIS)

    Dreier, J.; Huggenberger, M.; Aubert, C.; Bandurski, T.; Fischer, O.; Healzer, J.; Lomperski, S.; Strassberger, H.J.; Varadi, G.; Yadigaroglu, G.

    1996-01-01

    The PANDA test facility at the Paul Scherrer Institute is used to study the long-term performance of the Simplified Boiling Water Reactor's passive containment cooling system. The PANDA tests demonstrate performance on a larger scale than previous tests and examine the effects of any non-uniform spatial distributions of steam and non-condensable gases in the system. The facility is in 1:1 vertical scale and 1:25 scale for volume, power etc. Extensive facility characterization tests and steady-state passive containment condenser performance tests are presented. The results of the base case test of a series of transient system behaviour tests are reviewed. The first PANDA tests exhibited reproducibility, and indicated that the Simplified Boiling Water Reactor's containment is likely to be favorably responsive and highly robust to changes in the thermal transport patterns. (orig.) [de

  13. Fast flux test facility hazards assessment

    International Nuclear Information System (INIS)

    Sutton, L.N.

    1994-01-01

    This document establishes the technical basis in support of Emergency Planning Activities for the Fast Flux Test Facility on the Hanford Site. The document represents an acceptable interpretation of the implementing guidance document for DOE Order 5500.3A. Through this document, the technical basis for the development of facility specific Emergency Action Levels and the Emergency Planning Zone is demonstrated

  14. Development of Design Concept and Applied Technology for RCP Performance Test Facility

    International Nuclear Information System (INIS)

    Park, Sang Jin; Lee, Jung Ho; Yoon, Seok Ho

    2010-02-01

    Performance test facility for RCP (reactor coolant pump) is essential to verify the performance and reliability of RCP before installation in the nuclear power plant. The development of RCP for new-type reactor and the performance verification of hydraulic revolving body also needs the RCP test facility. The design concept of test loop and the technology of flow rate measurement are investigated in this research

  15. An experimental investigation of 1% SBLOCA on PSB-VVER test facility

    Energy Technology Data Exchange (ETDEWEB)

    Lipatov, I.A.; Dremin, G.I.; Galtchanskaia, S.A.; Gorbunov, Yu.S. [Electrogorsk Research and Engineering Center, EREC, Electrogorsk (Russian Federation); Elkin, I.V. [Russian Research Centre Kurchatov Inst., Moscow (Russian Federation)

    2001-07-01

    The paper presents the results of the three tests carried out in the PSB-VVER large-scale integral test facility. The PSB-VVER test facility is a four loop, full pressure scaled down model bearing structural similarities to the primary system of the NRP with VVER-1000 Russian design reactor. Volume-power scale is 1/300 while elevation scale is 1/1. (orig.)

  16. Test facilities for HTR, (2)

    International Nuclear Information System (INIS)

    Ishizuka, Hiroshi; Hayakawa, Hitoshi; Miki, Toshiya.

    1981-01-01

    The core of the multi-purpose high temperature gas-cooled experimental reactor is a circular column as a whole, in which the columns of stacked graphite blocks of hexagonal prism are arranged. The blocks in a column are doweled so as not to move horizontally, but adjacent columns vibrate while colliding mutually at the time of an earthquake because there is a gap between them. For the purpose to know the vibrating characteristics of a column surrounded by gap, Fuji Electric Co., Ltd., carried out the experiment. The tested column, the testing setup and the test result are reported. The distribution of flow rate in the core must be clarified, and the design data must be established early for confirming the feasibility of core design. The core structure tester was installed in Japan Atomic Energy Research Institute. The 1/2.75 scale model of the reactor bed was used, and the sealing performance of the block assemblies was tested. The sealing tester is related also to the distribution of flow rate in the core, and the basic performance of seal elements and the cross flow in fuel blocks were tested. The one-column tester and the seal element/two-column tester, the piping unit and the blower filter unit compose this tester. (Kako, I.)

  17. 400 Area/Fast Flux Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The 400 Area at Hanford is home primarily to the Fast Flux Test Facility (FFTF), a DOE-owned, formerly operating, 400-megawatt (thermal) liquid-metal (sodium)-cooled...

  18. Strategic closed-loop facility location problem with carbon market trading

    DEFF Research Database (Denmark)

    Diabat, A.; Abdallah, T.; Al-Refaie, A.

    2013-01-01

    and recovery of products in a closed-loop configuration. Remanufacturing is the basis of profit-oriented reverse logistics in which recovered products are restored to a marketable condition in order to be resold to the primary or secondary market. In this paper, we introduce a multiechelon multicommodity...... facility location problem with a trading price of carbon emissions and a cost of procurement. The company might either incur costs if the carbon cap, normally assigned by regulatory agencies, is lower than the total emissions, or gain profit if the carbon cap is higher than the total emissions. A numerical...

  19. Sultan - forced flow, high field test facility

    International Nuclear Information System (INIS)

    Horvath, I.; Vecsey, G.; Weymuth, P.; Zellweger, J.

    1981-01-01

    Three European laboratories: CNEN (Frascati, I) ECN (Petten, NL) and SIN (Villigen, CH) decided to coordinate their development efforts and to install a common high field forced flow test facility at Villigen Switzerland. The test facility SULTAN (Supraleiter Testanlage) is presently under construction. As a first step, an 8T/1m bore solenoid with cryogenic periphery will be ready in 1981. The cryogenic system, data acquisition system and power supplies which are contributed by SIN are described. Experimental feasibilities, including cooling, and instrumentation are reviewed. Progress of components and facility construction is described. Planned extension of the background field up to 12T by insert coils is outlined. 5 refs

  20. Advanced Control Test Operation (ACTO) facility

    International Nuclear Information System (INIS)

    Ball, S.J.

    1987-01-01

    The Advanced Control Test Operation (ACTO) project, sponsored by the US Department of Energy (DOE), is being developed to enable the latest modern technology, automation, and advanced control methods to be incorporated into nuclear power plants. The facility is proposed as a national multi-user center for advanced control development and testing to be completed in 1991. The facility will support a wide variety of reactor concepts, and will be used by researchers from Oak Ridge National Laboratory (ORNL), plus scientists and engineers from industry, other national laboratories, universities, and utilities. ACTO will also include telecommunication facilities for remote users

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  2. Massachusetts Large Blade Test Facility Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Rahul Yarala; Rob Priore

    2011-09-02

    Project Objective: The Massachusetts Clean Energy Center (CEC) will design, construct, and ultimately have responsibility for the operation of the Large Wind Turbine Blade Test Facility, which is an advanced blade testing facility capable of testing wind turbine blades up to at least 90 meters in length on three test stands. Background: Wind turbine blade testing is required to meet international design standards, and is a critical factor in maintaining high levels of reliability and mitigating the technical and financial risk of deploying massproduced wind turbine models. Testing is also needed to identify specific blade design issues that may contribute to reduced wind turbine reliability and performance. Testing is also required to optimize aerodynamics, structural performance, encourage new technologies and materials development making wind even more competitive. The objective of this project is to accelerate the design and construction of a large wind blade testing facility capable of testing blades with minimum queue times at a reasonable cost. This testing facility will encourage and provide the opportunity for the U.S wind industry to conduct more rigorous testing of blades to improve wind turbine reliability.

  3. Repetitively pulsed material testing facility

    International Nuclear Information System (INIS)

    Zucker, O.; Bostick, W.; Gullickson, R; Long, J.; Luce, J.; Sahlin, H.

    1975-01-01

    A continuously operated, 1 pps, dense-plasma-focus device capable of delivering a minimum of 10 15 neutrons per pulse for material testing purposes is described. Moderate scaling from existing results is sufficient to provide 2 x 10 13 n/cm 2 .s to a suitable target. The average power consumption, which has become a major issue as a result of the energy crisis, is analyzed with respect to other plasma devices and is shown to be highly favorable. A novel approach to the capacitor bank and switch design allowing repetitive operation is discussed. (U.S.)

  4. Detector development and test facility

    International Nuclear Information System (INIS)

    Reeder, D.D.

    1993-01-01

    Following the ideas presented in the proposal to the DoE, we have begun to acquire the equipment needed to design, develop construct and test the electronic and mechanical features of detectors used in High Energy Physics Experiments. A guiding principle for the effort is to achieve integrated electronic and mechanical designs which meet the demanding specifications of the modern hadron collider environment yet minimize costs. This requires state of the art simulation of signal processing as well as detailed calculations of heat transfer and finite element analysis of structural integrity

  5. Repetitively pulsed material testing facility

    International Nuclear Information System (INIS)

    Zucker, O.; Bostick, W.; Gullickson, R.; Long, J.; Luce, J.; Sahlin, H.

    1975-01-01

    A continuously operated, 1 pps, dense-plasma-focus device capable of delivering a minimum of 10 15 neutrons per pulse for material testing purposes is described. Moderate scaling from existing results is sufficient to provide 2 x 10 13 n/cm 2 . s to a suitable target. The average power consumption, which has become a major issue as a result of the energy crisis, is analyzed with respect to other plasma devices and is shown to be highly favorable. A novel approach to the capacitor bank and switch design allowing repetitive operation is discussed

  6. Fusion Materials Irradiation Test Facility: experimental capabilities and test matrix

    International Nuclear Information System (INIS)

    Opperman, E.K.

    1982-01-01

    This report describes the experimental capabilities of the Fusion Materials Irradiation Test Facility (FMIT) and reference material specimen test matrices. The description of the experimental capabilities and the test matrices has been updated to match the current single test cell facility ad assessed experimenter needs. Sufficient detail has been provided so that the user can plan irradiation experiments and conceptual hardware. The types of experiments, irradiation environment and support services that will be available in FMIT are discussed

  7. Directory of transport packaging test facilities

    International Nuclear Information System (INIS)

    1983-08-01

    Radioactive materials are transported in packagings or containers which have to withstand certain tests depending on whether they are Type A or Type B packagings. In answer to a request by the International Atomic Energy Agency, 13 Member States have provided information on the test facilities and services existing in their country which can be made available for use by other states by arrangement for testing different kinds of packagings. The directory gives the technical information on the facilities, the services, the tests that can be done and in some cases even the financial arrangement is included

  8. SRL incinerator components test facility

    International Nuclear Information System (INIS)

    Freed, E.J.

    1982-08-01

    A full-scale (5 kg waste/hour) controlled-air incinerator, the ICTF, is presently being tested with simulated waste as part of a program to develop technology for incineration of Savannah River Plant solid transuranic wastes. This unit is designed specifically to incinerate relatively small quantities of solid combustible waste that are contaminated up to 10 5 times the present nominal 10 nCi/g threshold value for such isotopes as 238 Pu, 239 Pu, 242 Cm, and 252 Cf. Automatic incinerator operation and control has been incorporated into the design, simulating the future plant design which minimizes operator radiation exposure. Over 3000 kg of nonradioactive wastes characteristic of plutonium finishing operations have been incinerated at throughputs exceeding 5 kg/hr. Safety and reliability were the major design objectives. In addition to the incinerator tests, technical data were gathered on two different off-gas systems: a wet system composed of three scrubbers in series, and a dry system employing sintered metal filters

  9. Automation of electromagnetic compatability (EMC) test facilities

    Science.gov (United States)

    Harrison, C. A.

    1986-01-01

    Efforts to automate electromagnetic compatibility (EMC) test facilities at Marshall Space Flight Center are discussed. The present facility is used to accomplish a battery of nine standard tests (with limited variations) deigned to certify EMC of Shuttle payload equipment. Prior to this project, some EMC tests were partially automated, but others were performed manually. Software was developed to integrate all testing by means of a desk-top computer-controller. Near real-time data reduction and onboard graphics capabilities permit immediate assessment of test results. Provisions for disk storage of test data permit computer production of the test engineer's certification report. Software flexibility permits variation in the tests procedure, the ability to examine more closely those frequency bands which indicate compatibility problems, and the capability to incorporate additional test procedures.

  10. Ice condenser testing facility and plans

    International Nuclear Information System (INIS)

    Kannberg, L.D.; Ross, B.A.; Eschbach, E.J.; Ligotke, M.W.

    1987-01-01

    A facility is being constructed to experimentally validate the ICEDF computer code. The code was developed to estimate the extent of fission product retention in the ice compartments of pressurized water reactor ice condenser containment systems during severe accidents. The design and construction of the facility is based on a test design that addresses the validation needs of the code for conditions typical of those expected to occur during severe pressurized water reactor accidents. Detailed facility design has followed completion of a test design (i.e., assembled test cases each involving a different set of aerosol and thermohydraulic flow conditions). The test design was developed with the aid of statistical test design software and was scrutinized for applicability with the aid of ICEDF simulations. The test facility will incorporate a small section of a prototypic ice condenser (e.g., a cross section comprising the equivalent of four 1-ft-diameter ice baskets to their full prototypic height of 48 ft). The development of the test design, the detailed facility design, and the construction progress are described in this paper

  11. Engineering design of IFMIF/EVEDA lithium test loop. Electro-magnetic pump and pressure drop

    International Nuclear Information System (INIS)

    Kondo, Hiroo; Furukawa, Tomohiro; Hirakawa, Yasushi; Iuchi, Hiroshi; Kanemura, Takuji; Ida, Mizuho; Watanabe, Kazuyoshi; Wakai, Eiichi; Nakamura, Kazuyuki; Horiike, H.; Yamaoka, N.; Matsushita, I.

    2011-01-01

    The Engineering Validation and Engineering Design Activities (EVEDA) for the International Fusion Materials Irradiation Facility (IFMIF) is proceeding as one of the ITER Broader Approach (ITER-BA). A Li circulation loop for testing hydraulic stability of the Li target (high speed free-surface flow of liquid Li as a beam target) and Li purification traps are under construction in the Japan Atomic Energy Agency as a major Japanese activities in the EVEDA. This paper presents specification of an electro-magnetic pump (EMP) for the EVEDA Li Test Loop (ELTL) and evaluation of the pressure drop in the main loop of the ELTL. The EMP circulates the liquid Li at a large flow rate up to 0.05 m 3 /s (3000 l/min) under a vacuum cover gas (Ar) pressure of 10 -3 Pa, thus the evaluation of cavitation generation is a crucial issue. The EMP used in the ELTL consists of two EMPs aligned in series through a U-tube whose size of one EMP is 0.8 m square and 2.6 m in length. The calculation of the pressure drop in the main Li loop to the EMP is approx. 25 kPa at the design maximum flow rate of 0.05 m 3 /s. On the other hand the height from the EMP to a Li tank to supply Li to the EMP is designed to be 9.72 m, and secures a static pressure and the cavitation number of 18 kPa and 3.4 respectively at the maximum flow rate in a vacuum condition. As a result, it is confirmed to prevent cavitation at the inlet of the EMP in this design. (author)

  12. ICH antenna development on the ORNL RF Test Facility

    International Nuclear Information System (INIS)

    Gardner, W.L.; Bigelow, T.S.; Haste, G.R.; Hoffman, D.J.; Livesey, R.L.

    1987-01-01

    A compact resonant loop antenna is installed on the ORNL Radio Frequency Test Facility (RFTF). Facility characteristics include a steady-state magnetic field of ∼ 0.5 T at the antenna, microwave-generated plasmas with n e ∼ 10 12 cm -3 and T e ∼ 8 eV, and 100 kW of 25-MHz rf power. The antenna is tunable from ∼22--75 MHz, is designed to handle ≥1 MW of rf power, and can be moved 5 cm with respect to the port flange. Antenna characteristics reported and discussed include the effect of magnetic field on rf voltage breakdown at the capacitor, the effects of magnetic field and plasma on rf voltage breakdown between the radiating element and the Faraday shield, the effects of graphite on Faraday shield losses, and the efficiency of coupling to the plasma. 2 refs., 4 figs

  13. Recommissioning the K-1600 seismic test facility

    International Nuclear Information System (INIS)

    Wynn, C.C.; Brewer, D.W.

    1991-01-01

    The Center for Natural Phenomena Engineering (CNPE) was established under the technical direction of Dr. James E. Beavers with a mandate to assess, by analyses and testing, the seismic capacity of building structures that house sensitive processes at the Oak Ridge Y-12 Plant. This mandate resulted in a need to recommission the K-1600 Seismic Test Facility (STF) at the Oak Ridge K-25 Site, which had been shutdown for 6 years. This paper documents the history of the facility and gives some salient construction, operation, and performance details of its 8-ton, 20-foot center of gravity payload biaxial seismic simulator. A log of activities involved in the restart of this valuable resource is included as Table 1. Some of the problems and solutions associated with recommissioning the facility under a relatively limited budget are included. The unique attributes of the shake table are discussed. The original mission and performance requirements are compared to current expanded mission and performance capabilities. Potential upgrades to further improve the capabilities of the test facility as an adjunct to the CNPE are considered. Additional uses for the facility are proposed, including seismic qualification testing of devices unique to enrichment technologies and associated hazardous waste treatment and disposal processes. In summary, the STF restart in conjunction with CNPE has added a vital, and unique facility to the list of current national resources utilized for earthquake engineering research and development

  14. Construction of solid waste form test facility

    International Nuclear Information System (INIS)

    Park, Hyun Whee; Lee, Kang Moo; Koo, Jun Mo; Jung, In Ha; Lee, Jong Ryeul; Kim, Sung Whan; Bae, Sang Min; Cho, Kang Whon; Sung, Suk Jong

    1989-02-01

    The Solid Waste Form Test Facility (SWFTF) is now construction at DAEDUCK in Korea. In SWFTF, the characteristics of solidified waste products as radiological homogeneity, mechanical and thermal property, water resistance and lechability will be tested and evaluated to meet conditions for long-term storage or final disposal of wastes. The construction of solid waste form test facility has been started with finishing its design of a building and equipments in Sep. 1984, and now building construction is completed. Radioactive gas treatment system, extinguishers, cooling and heating system for the facility, electrical equipments, Master/Slave manipulator, power manipulator, lead glass and C.C.T.V. has also been installed. SWFTF will be established in the beginning of 1990's. At this report, radiation shielding door, nondestructive test of the wall, instrumentation system for the utility supply system and cell lighting system are described. (Author)

  15. Design of a hydrogen test facility

    International Nuclear Information System (INIS)

    Morgan, M.J.; Beam, J.E.; Sehmbey, M.S.; Pais, M.R.; Chow, L.C.; Hahn, O.J.

    1992-01-01

    The Air Force has sponsored a program at the University of Kentucky which will lead to a better understanding of the thermal and fluid instabilities during blowdown of supercritical fluids at cryogenic temperatures. An integral part of that program is the design and construction of that hydrogen test facility. This facility will be capable of providing supercritical hydrogen at 30 bars and 35 K at a maximum flow rate of 0.1 kg/s for 90 seconds. Also presented here is an extension of this facility to accommodate the use of supercritical helium

  16. Status of superconducting RF test facility (STF)

    International Nuclear Information System (INIS)

    Hayano, Hitoshi

    2005-01-01

    A superconducting technology was recommended for the main linac design of the International Linear Collider (ILC) by the International Technology Recommendation Panel (ITRP). The basis for this design has been developed and tested at DESY, and R and D is progressing at many laboratories around the world including DESY, Orsay, KEK, FNAL, SLAC, Cornell, and JLAB. In order to promote Asian SC-technology for ILC, construction of a test facility in KEK was discussed and decided. The role and status of the superconducting RF test facility (STF) is reported in this paper. (author)

  17. A negative ion source test facility

    Energy Technology Data Exchange (ETDEWEB)

    Melanson, S.; Dehnel, M., E-mail: morgan@d-pace.com; Potkins, D.; Theroux, J.; Hollinger, C.; Martin, J.; Stewart, T.; Jackle, P.; Withington, S. [D-Pace, Inc., P.O. Box 201, Nelson, British Columbia V1L 5P9 (Canada); Philpott, C.; Williams, P.; Brown, S.; Jones, T.; Coad, B. [Buckley Systems Ltd., 6 Bowden Road, Mount Wellington, Auckland 1060 (New Zealand)

    2016-02-15

    Progress is being made in the development of an Ion Source Test Facility (ISTF) by D-Pace Inc. in collaboration with Buckley Systems Ltd. in Auckland, NZ. The first phase of the ISTF is to be commissioned in October 2015 with the second phase being commissioned in March 2016. The facility will primarily be used for the development and the commercialization of ion sources. It will also be used to characterize and further develop various D-Pace Inc. beam diagnostic devices.

  18. Manual for operation of the multipurpose thermalhydraulic test facility TOPFLOW (Transient Two Phase Flow Test Facility)

    International Nuclear Information System (INIS)

    Beyer, M.; Carl, H.; Schuetz, H.; Pietruske, H.; Lenk, S.

    2004-07-01

    The Forschungszentrum Rossendorf (FZR) e. V. is constructing a new large-scale test facility, TOPFLOW, for thermalhydraulic single effect tests. The acronym stands for transient two phase flow test facility. It will mainly be used for the investigation of generic and applied steady state and transient two phase flow phenomena and the development and validation of models of computational fluid dynamic (CFD) codes. The manual of the test facility must always be available for the staff in the control room and is restricted condition during operation of personnel and also reconstruction of the facility. (orig./GL)

  19. Startup of large coil test facility

    International Nuclear Information System (INIS)

    Haubenreich, P.N.; Bohanan, R.E.; Fietz, W.A.; Luton, J.N.; May, J.R.

    1984-01-01

    The Large Coil Test Facility (LCTF) is being used to test superconducting toroidal field coils about one-third the size of those for INTOR. Data were obtained on performance of refrigerator, helium distribution, power supplies, controls, and data acquisition systems and on the acoustic emission, voltages, currents, and mechanical strains during charging and discharging the coils. (author)

  20. Closing the brain-to-brain loop in laboratory testing.

    Science.gov (United States)

    Plebani, Mario; Lippi, Giuseppe

    2011-07-01

    Abstract The delivery of laboratory services has been described 40 years ago and defined with the foremost concept of "brain-to-brain turnaround time loop". This concept consists of several processes, including the final step which is the action undertaken on the patient based on laboratory information. Unfortunately, the need for systematic feedback to improve the value of laboratory services has been poorly understood and, even more risky, poorly applied in daily laboratory practice. Currently, major problems arise from the unavailability of consensually accepted quality specifications for the extra-analytical phase of laboratory testing. This, in turn, does not allow clinical laboratories to calculate a budget for the "patient-related total error". The definition and use of the term "total error" refers only to the analytical phase, and should be better defined as "total analytical error" to avoid any confusion and misinterpretation. According to the hierarchical approach to classify strategies to set analytical quality specifications, the "assessment of the effect of analytical performance on specific clinical decision-making" is comprehensively at the top and therefore should be applied as much as possible to address analytical efforts towards effective goals. In addition, an increasing number of laboratories worldwide are adopting risk management strategies such as FMEA, FRACAS, LEAN and Six Sigma since these techniques allow the identification of the most critical steps in the total testing process, and to reduce the patient-related risk of error. As a matter of fact, an increasing number of laboratory professionals recognize the importance of understanding and monitoring any step in the total testing process, including the appropriateness of the test request as well as the appropriate interpretation and utilization of test results.

  1. SULTAN test facility: Summary of recent results

    International Nuclear Information System (INIS)

    Stepanov, Boris; Bruzzone, Pierluigi; Sedlak, Kamil; Croari, Giancarlo

    2013-01-01

    The test campaigns of the ITER conductors in the SULTAN test facility re-started in December 2011 after three months break. The main focus of the activities is about the qualification tests of the Central Solenoid (CS) conductors, with three different samples for a total six variations of strand suppliers and cable layouts. In 2012, five Toroidal Field (TF) conductor samples have also been tested as part of the supplier and process qualification phase of the European, Korean, Chinese and Russian Federation Agencies. A summary of the test results for all the ITER samples tested in the last period is presented, including an updated statistics of the broad transition, the performance degradation and the impact of layout variations. The role of SULTAN test facility during the ITER construction is reviewed, and the load of work for the next three years is anticipated

  2. Power Burst Facility Severe Fuel Damage test series

    International Nuclear Information System (INIS)

    Buescher, B.J.; Osetek, D.J.; Ploger, S.A.

    1982-01-01

    The Severe Fuel Damage (SFD) tests planned for the Power Burst Facility (PBF) are described. Bundles containing 32 zircaloy-clad, PWR-type fuel rods will be subjected to severe overheating transients in a high-pressure, superheated-steam environment. Cladding temperatures are expected to reach 2400 0 K, resulting in cladding ballooning and rupture, severe cladding oxidation, cladding melting, fuel dissolution, fuel rod fragmentation, and possibly, rubble bed formation. An experiment effluent collection system is being installed and the PBF fission product monitoring system is being upgraded to meet the special requirements of the SFD tests. Scoping calculations were performed to evaluate performance of the SFD test design and to establish operational requirements for the PBF loop

  3. Characterizing experiments of the PPOOLEX test facility

    Energy Technology Data Exchange (ETDEWEB)

    Puustinen, M.; Laine, J. (Lappeenranta Univ. of Technology, Nuclear Safety Research Unit (Finland))

    2008-07-15

    This report summarizes the results of the characterizing test series in 2007 with the scaled down PPOOLEX facility designed and constructed at Lappeenranta University of Technology. Air and steam/air mixture was blown into the dry well compartment and from there through a DN200 blowdown pipe to the condensation pool (wet well). Altogether eight air and four steam/air mixture experiments, each consisting of several blows (tests), were carried out. The main purpose of the experiment series was to study the general behavior of the facility and the performance of basic instrumentation. Proper operation of automation, control and safety systems was also tested. The test facility is a closed stainless steel vessel divided into two compartments, dry well and wet well. The facility is equipped with high frequency measurements for capturing different aspects of the investigated phenomena. The general behavior of the PPOOLEX facility differs significantly from that of the previous POOLEX facility because of the closed two-compartment structure of the test vessel. Heat-up by several tens of degrees due to compression in both compartments was the most obvious evidence of this. Temperatures also stratified. Condensation oscillations and chugging phenomenon were encountered in those tests where the fraction of non-condensables had time to decrease significantly. A radical change from smooth condensation behavior to oscillating one occurred quite abruptly when the air fraction of the blowdown pipe flow dropped close to zero. The experiments again demonstrated the strong diminishing effect that noncondensable gases have on dynamic unsteady loadings experienced by submerged pool structures. BWR containment like behavior related to the beginning of a postulated steam line break accident was observed in the PPOOLEX test facility during the steam/air mixture experiments. The most important task of the research project, to produce experimental data for code simulation purposes, can be

  4. Test facilities for future linear colliders

    International Nuclear Information System (INIS)

    Ruth, R.D.

    1995-12-01

    During the past several years there has been a tremendous amount of progress on Linear Collider technology world wide. This research has led to the construction of the test facilities described in this report. Some of the facilities will be complete as early as the end of 1996, while others will be finishing up around the end 1997. Even now there are extensive tests ongoing for the enabling technologies for all of the test facilities. At the same time the Linear Collider designs are quite mature now and the SLC is providing the key experience base that can only come from a working collider. All this taken together indicates that the technology and accelerator physics will be ready for a future Linear Collider project to begin in the last half of the 1990s

  5. Low Level RF System for Jefferson Lab Cryomodule Test Facility

    International Nuclear Information System (INIS)

    Tomasz Plawski; Trent Allison; Jean Delayen; J. Hovater; Thomas Powers

    2003-01-01

    The Jefferson Lab Cryomodule Test Facility (CMTF) has been upgraded to test and commission SNS and CEBAF Energy Upgrade cryomodules. Part of the upgrade was to modernize the superconducting cavity instrumentation and control. We have designed a VXI based RF control system exclusively for the production testing of superconducting cavities. The RF system can be configured to work either in Phase Locked Loop (PLL) or Self Excited Loop (SEL) mode. It can be used to drive either SNS 805 MHz or CEBAF Energy Upgrade 1497 MHz superconducting cavities and can be operated in pulsed or continuous wave (CW) mode. The base design consists of RF-analog and digital sections. The RF-analog section includes a Voltage Control Oscillator (VCO), phase detector, IandQ modulator and ''low phase shift'' limiter. The digital section controls the analog section and includes ADC, FPGA, and DAC . We will discuss the design of the RF system and how it relates to the support of cavity testing

  6. Kauai Test Facility hazards assessment document

    Energy Technology Data Exchange (ETDEWEB)

    Swihart, A

    1995-05-01

    The Department of Energy Order 55003A requires facility-specific hazards assessment be prepared, maintained, and used for emergency planning purposes. This hazards assessment document describes the chemical and radiological hazards associated with the Kauai Test Facility, Barking Sands, Kauai, Hawaii. The Kauai Test Facility`s chemical and radiological inventories were screened according to potential airborne impact to onsite and offsite individuals. The air dispersion model, ALOHA, estimated pollutant concentrations downwind from the source of a release, taking into consideration the toxicological and physical characteristics of the release site, the atmospheric conditions, and the circumstances of the release. The greatest distance to the Early Severe Health Effects threshold is 4.2 kilometers. The highest emergency classification is a General Emergency at the {open_quotes}Main Complex{close_quotes} and a Site Area Emergency at the Kokole Point Launch Site. The Emergency Planning Zone for the {open_quotes}Main Complex{close_quotes} is 5 kilometers. The Emergency Planning Zone for the Kokole Point Launch Site is the Pacific Missile Range Facility`s site boundary.

  7. Plasma-Materials Interactions Test Facility

    International Nuclear Information System (INIS)

    Uckan, T.

    1986-11-01

    The Plasma-Materials Interactions Test Facility (PMITF), recently designed and constructed at Oak Ridge National Laboratory (ORNL), is an electron cyclotron resonance microwave plasma system with densities around 10 11 cm -3 and electron temperatures of 10-20 eV. The device consists of a mirror cell with high-field-side microwave injection and a heating power of up to 0.8 kW(cw) at 2.45 GHz. The facility will be used for studies of plasma-materials interactions and of particle physics in pump limiters and for development and testing of plasma edge diagnostics

  8. Real-Time Hardware-in-the-Loop Laboratory Testing for Multisensor Sense and Avoid Systems

    Directory of Open Access Journals (Sweden)

    Giancarmine Fasano

    2013-01-01

    Full Text Available This paper focuses on a hardware-in-the-loop facility aimed at real-time testing of architectures and algorithms of multisensor sense and avoid systems. It was developed within a research project aimed at flight demonstration of autonomous non-cooperative collision avoidance for Unmanned Aircraft Systems. In this framework, an optionally piloted Very Light Aircraft was used as experimental platform. The flight system is based on multiple-sensor data integration and it includes a Ka-band radar, four electro-optical sensors, and two dedicated processing units. The laboratory test system was developed with the primary aim of prototype validation before multi-sensor tracking and collision avoidance flight tests. System concept, hardware/software components, and operating modes are described in the paper. The facility has been built with a modular approach including both flight hardware and simulated systems and can work on the basis of experimentally tested or synthetically generated scenarios. Indeed, hybrid operating modes are also foreseen which enable performance assessment also in the case of alternative sensing architectures and flight scenarios that are hardly reproducible during flight tests. Real-time multisensor tracking results based on flight data are reported, which demonstrate reliability of the laboratory simulation while also showing the effectiveness of radar/electro-optical fusion in a non-cooperative collision avoidance architecture.

  9. Startup of Large Coil Test Facility

    International Nuclear Information System (INIS)

    Haubenreich, P.N.; Bohanan, R.E.; Fietz, W.A.; Luton, J.N.; May, J.R.

    1984-01-01

    The Large Coil Test Facility (LCTF) is being used to test superconducting toroidal field coils about one-third the size of those for INTOR. Eventually, six different coils from four countries will be tested. Operations began in 1983 with acceptance testing of the helium refrigerator/liquefier system. Comprehensive shakedown of the facility and tests with the first three coils (from Japan, the United States, and Switzerland) were successfully accomplished in the summer of 1984. Currents up to 10,200 A and fields up to 6.4 T were reached. Data were obtained on performance of refrigerator, helium distribution, power supplies, controls, and data acquisition systems and on the acoustic emission, voltages, currents, and mechanical strains during charging and discharging the coils

  10. Startup of Large Coil Test Facility

    International Nuclear Information System (INIS)

    Haubenreich, P.N.; Bohanan, R.E.; Fietz, W.A.; Luton, J.N.; May, J.R.

    1985-01-01

    The Large Coil Test Facility (LCTF) is being used to test superconducting toroidal field coils about one-third the size of those for INTOR. Eventually, six different coils from four countries will be tested. Operations began in 1983 with acceptance testing of the helium refrigerator/liquefier system. Comprehensive shakedown of the facility and tests with the first three coils (from Japan, the United States, and Switzerland) were successfully accomplished in the summer of 1984. Currents up to 10,200 A and fields up to 6.4 T were reached. Data were obtained on performance of refrigerator, helium distribution, power supplies, controls, and data acquisition systems and on the acoustic emission, voltages, currents, and mechanical strains during charging and discharging the coils

  11. Considerations on a PAHR test facility

    International Nuclear Information System (INIS)

    Boenisch, G.; Groetzbach, G.; Heinzel, V.; Kleefeld, K.; Kuechle, M.; Mueller, R.A.; Royl, P.; Schramm, K.; Smidt, D.; Werle, H.

    1976-01-01

    On the basis of a hypothetical core disruptive accident (HCDA) analysis the phenomena of the post accident phase are first identified which require experimental investigations and can only be studied in pile. Then the experimental requests for both debris bed and molten fuel pool studies are specified and grouped into three categories. For two of the categories the requests can be satisfied with loop experiments in thermal reactors. For the third category a 70 cm diameter test bed is needed and here the proposal is to use a flat core fast reactor with the test bed located below the core heated by axial leakage neutrons. Finally a conceptual design for such a reactor is presented where the test bed is loaded into an ex-vessel device and is removable on a carriage to a hot cell building. Maintenance and safety problems are briefly discussed and alternative solutions are mentioned

  12. Safety test facilities - status, needs, future directions

    International Nuclear Information System (INIS)

    Heusener, G.; Cogne, F.

    1979-08-01

    A survey is given of the in-pile programs which are presently or in the near future being performed in the DeBeNe-area and in France. Only those in-pile programs are considered which are dealing with severe accidents that might lead to disruption of major parts of the core. By comparing the needs with the goals of the present programs points are identified which are not sufficiently well covered up till now. The future procedure is described: the existing facilities will be used to the largest possible extent. Whenever it is necessary, upgrading and improvement will be foreseen. Studies of a Test Facility allowing the transient testing of large pin bundles should be continued. The construction of such a facility in Europe in the near future however seems premature

  13. The ACE-DTU Planar Near-Field Ground Penetrating Radar Antenna Test Facility

    DEFF Research Database (Denmark)

    Lenler-Eriksen, Hans-Rudolph; Meincke, Peter

    2004-01-01

    The ACE-DTU planar near-field ground penetrating radar (GPR) antenna test facility is used to measure the plane-wave transmitting spectrum of a GPR loop antenna close to the air-soil interface by means of a probe buried in soil. Probe correction is implemented using knowledge about the complex...

  14. The BNL Accelerator Test Facility control system

    International Nuclear Information System (INIS)

    Malone, R.; Bottke, I.; Fernow, R.; Ben-Zvi, I.

    1993-01-01

    Described is the VAX/CAMAC-based control system for Brookhaven National Laboratory's Accelerator Test Facility, a laser/linac research complex. Details of hardware and software configurations are presented along with experiences of using Vsystem, a commercial control system package

  15. FFTF [Fast Flux Test Facility] management

    International Nuclear Information System (INIS)

    Bennett, C.L.

    1986-11-01

    Fuel Management at the Fast Flux Test Facility (FFTF) involves more than just the usual ex-core and in-core management of standard fuel and non-fuel components between storage locations and within the core since it is primarily an irradiation test facility. This mission involves testing an ever increasing variety of fueled and non-fueled experiments, each having unique requirements on the reactor core as well as having its own individual impact on the reload design. This paper describes the fuel management process used by the Westinghouse Hanford Company Core Engineering group that has led to the successful reload design of nine operating cycles and the irradiation of over 120 tests

  16. The NRU blowdown test facility commissioning program

    Energy Technology Data Exchange (ETDEWEB)

    Walsworth, J A; Zanatta, R J; Yamazaki, A R; Semeniuk, D D; Wong, W; Dickson, L W; Ferris, C E; Burton, D H [Atomic Energy of Canada Ltd., Chalk River, ON (Canada). Chalk River Nuclear Labs.

    1990-12-31

    A major experimental program has been established at the Chalk River Nuclear Laboratories (CRL) that will provide essential data on the thermal and mechanical behaviour of nuclear fuel under abnormal reactor operating conditions and on the transient release, transport and deposition of fission product activity from severely degraded fuel. A number of severe fuel damage (SFD) experiments will be conducted within the Blowdown Test Facility (BTF) at CRL. A series of experiments are being conducted to commission this new facility prior to the SFD program. This paper describes the features and the commissioning program for the BTF. A development and testing program is described for critical components used on the reactor test section. In-reactor commissioning with a fuel assembly simulator commenced in 1989 June and preliminary results are given. The paper also outlines plans for future all-effects, in-reactor tests of CANDU-designed fuel. (author). 11 refs., 3 tabs., 7 figs.

  17. Natural Circulation High Pressure Loop Dynamics Around Operating Point, Tests and Modelling With Retran 02

    International Nuclear Information System (INIS)

    Masriera, N.A; Doval, A.S; Mazufri, C.M

    2000-01-01

    The Natural Circulation High Pressure Loop (CAPCN) reproduces in scale all the one-dimensional thermal-hydraulic phenomena occurring in the primary loop of CAREM-25 reactor.It plays an important role in the qualification process of calculating computer codes.This facility demanded to develop several technological solutions in order to achieve the measuring and control quality required by that process.This engineering and experimental development allowed completing the first stage of dynamic tests during 1998.The trends of recorded data were systematically evaluated in terms of the deviations of main variables in response to different perturbations.By this analysis a group of eight transients was selected, providing a Minimum Representative Set (MRS) of dynamic tests, allowing the evaluation of all dynamic phenomena.Each of these transients was simulated with RETRAN-02, using a spreadsheet to facilitate the consistent elaboration and modification of input files.Comparing measured data and computer simulations, it may be concluded that it is possible to reproduce the dynamic response of all the transients with a level of approximation quite homogeneous and generally acceptable.It is possible to identify the detailed physical models that fit better the dynamic phenomena, and which of the limitations of RETRAN code are more relevant

  18. Conceptual studies of plasma engineering test facility

    International Nuclear Information System (INIS)

    Hiraoka, Toru; Tazima, Teruhiko; Sugihara, Masayoshi; Kasai, Masao; Shinya, Kichiro

    1979-04-01

    Conceptual studies have been made of a Plasma Engineering Test Facility, which is to be constructed following JT-60 prior to the experimental power reactor. The physical aim of this machine is to examine self-ignition conditions. This machine possesses all essential technologies for reactor plasma, i.e. superconducting magnet, remote maintenance, shielding, blanket test modules, tritium handling. Emphasis in the conceptual studies was on structural consistency of the machine and whether the machine would be constructed practically. (author)

  19. ACIGA's high optical power test facility

    International Nuclear Information System (INIS)

    Ju, L; Aoun, M; Barriga, P

    2004-01-01

    Advanced laser interferometer detectors utilizing more than 100 W of laser power and with ∼10 6 W circulating laser power present many technological problems. The Australian Consortium for Interferometric Gravitational Astronomy (ACIGA) is developing a high power research facility in Gingin, north of Perth, Western Australia, which will test techniques for the next generation interferometers. In particular it will test thermal lensing compensation and control strategies for optical cavities in which optical spring effects and parametric instabilities may present major difficulties

  20. FAST FLUX TEST FACILITY DRIVER FUEL MEETING

    Energy Technology Data Exchange (ETDEWEB)

    None,

    1966-06-01

    The Pacific Northwest Laboratory has convened this meeting to enlist the best talents of our laboratories and industry in soliciting factual, technical information pertinent to the Pacific Northwest's Laboratory's evaluation of the potential fuel systems for the Fast Flux Test Facility. The particular factors emphasized for these fuel systems are those associated with safety, ability to meet testing objectives, and economics. The proceedings includes twenty-three presentations, along with a transcript of the discussion following each, as well as a summary discussion.

  1. Test facility TIMO for testing the ITER model cryopump

    International Nuclear Information System (INIS)

    Haas, H.; Day, C.; Mack, A.; Methe, S.; Boissin, J.C.; Schummer, P.; Murdoch, D.K.

    2001-01-01

    Within the framework of the European Fusion Technology Programme, FZK is involved in the research and development process for a vacuum pump system of a future fusion reactor. As a result of these activities, the concept and the necessary requirements for the primary vacuum system of the ITER fusion reactor were defined. Continuing that development process, FZK has been preparing the test facility TIMO (Test facility for ITER Model pump) since 1996. This test facility provides for testing a cryopump all needed infrastructure as for example a process gas supply including a metering system, a test vessel, the cryogenic supply for the different temperature levels and a gas analysing system. For manufacturing the ITER model pump an order was given to the company L' Air Liquide in the form of a NET contract. (author)

  2. Test facility TIMO for testing the ITER model cryopump

    International Nuclear Information System (INIS)

    Haas, H.; Day, C.; Mack, A.; Methe, S.; Boissin, J.C.; Schummer, P.; Murdoch, D.K.

    1999-01-01

    Within the framework of the European Fusion Technology Programme, FZK is involved in the research and development process for a vacuum pump system of a future fusion reactor. As a result of these activities, the concept and the necessary requirements for the primary vacuum system of the ITER fusion reactor were defined. Continuing that development process, FZK has been preparing the test facility TIMO (Test facility for ITER Model pump) since 1996. This test facility provides for testing a cryopump all needed infrastructure as for example a process gas supply including a metering system, a test vessel, the cryogenic supply for the different temperature levels and a gas analysing system. For manufacturing the ITER model pump an order was given to the company L'Air Liquide in the form of a NET contract. (author)

  3. Hot helium flow test facility summary report

    International Nuclear Information System (INIS)

    1980-06-01

    This report summarizes the results of a study conducted to assess the feasibility and cost of modifying an existing circulator test facility (CTF) at General Atomic Company (GA). The CTF originally was built to test the Delmarva Power and Light Co. steam-driven circulator. This circulator, as modified, could provide a source of hot, pressurized helium for high-temperature gas-cooled reactor (HTGR) and gas-cooled fast breeder reactor (GCFR) component testing. To achieve this purpose, a high-temperature impeller would be installed on the existing machine. The projected range of tests which could be conducted for the project is also presented, along with corresponding cost considerations

  4. Buildings, fields of activity, testing facilities

    International Nuclear Information System (INIS)

    1974-01-01

    Since 1969 the activities of the Materialpruefungsanstalt Stuttgart (MPA) have grown quickly as planned, especially in the field of reactor safety research, which made it necessary to increase the staff to approximately 165 members, to supplement the machines and equipment and to extend the fields of activities occasioning a further departmental reorganization. At present the MPA has the following departments: 1. Teaching (materials testing, materials science and strength of materials) 2. Materials and Welding Technology 3. Materials Science and General Materials Testing with Tribology 4. Design and Strength 5. Creep and Fatigue Testing 6. Central Facilities 7. Vessel and Component Testing. (orig./RW) [de

  5. The accomplishments of lithium target and test facility validation activities in the IFMIF/EVEDA phase

    Science.gov (United States)

    Arbeiter, Frederik; Baluc, Nadine; Favuzza, Paolo; Gröschel, Friedrich; Heidinger, Roland; Ibarra, Angel; Knaster, Juan; Kanemura, Takuji; Kondo, Hiroo; Massaut, Vincent; Saverio Nitti, Francesco; Miccichè, Gioacchino; O'hira, Shigeru; Rapisarda, David; Sugimoto, Masayoshi; Wakai, Eiichi; Yokomine, Takehiko

    2018-01-01

    As part of the engineering validation and engineering design activities (EVEDA) phase for the international fusion materials irradiation facility IFMIF, major elements of a lithium target facility and the test facility were designed, prototyped and validated. For the lithium target facility, the EVEDA lithium test loop was built at JAEA and used to test the stability (waves and long term) of the lithium flow in the target, work out the startup procedures, and test lithium purification and analysis. It was confirmed by experiments in the Lifus 6 plant at ENEA that lithium corrosion on ferritic martensitic steels is acceptably low. Furthermore, complex remote handling procedures for the remote maintenance of the target in the test cell environment were successfully practiced. For the test facility, two variants of a high flux test module were prototyped and tested in helium loops, demonstrating their good capabilities of maintaining the material specimens at the desired temperature with a low temperature spread. Irradiation tests were performed for heated specimen capsules and irradiation instrumentation in the BR2 reactor at SCK-CEN. The small specimen test technique, essential for obtaining material test results with limited irradiation volume, was advanced by evaluating specimen shape and test technique influences.

  6. Capability of the RELAP5 code to simulate natural circulation behaviour in test facilities

    International Nuclear Information System (INIS)

    Mangal, Amit; Jain, Vikas; Nayak, A.K.

    2011-01-01

    In the present study, one of the extensively used best estimate code RELAP5 has been used for simulation of steady state, transient and stability behavior of natural circulation based experimental facilities, such as the High-Pressure Natural Circulation Loop (HPNCL) and the Parallel Channel Loop (PCL) installed and operating at BARC. The test data have been generated for a range of pressure, power and subcooling conditions. The computer code RELAP5/MOD3.2 was applied to predict the transient natural circulation characteristics under single-phase and two-phase conditions, thresholds of flow instability, amplitude and frequency of flow oscillations for different operating conditions of the loops. This paper presents the effect of nodalisation in prediction of natural circulation behavior in test facilities and a comparison of experimental data in with that of code predictions. The errors associated with the predictions are also characterized

  7. The WR-1 corrosion test facility

    International Nuclear Information System (INIS)

    Murphy, E.V.; Simmons, G.R.

    1978-07-01

    This report describes a new Corrosion Test Facility which has recently been installed in the WR-1 organic-cooled research reactor. The irradiation facility is a single insert, installed in a reactor site, which can deliver a fast neutron flux density of 2.65 x 10 17 neutrons/(m 2 .s) to specimens under irradiation. A self-contained controlled-chemistry cooling water circuit removes the gamma- and neutron-heat generated in the insert and specimens. Specimen temperatures typically vary from 245 deg C to 280 deg C across the insert core region. (author)

  8. Several new thermo-hydraulic test facilities in NPIC

    International Nuclear Information System (INIS)

    Ye Shurong; Sun Yufa; Ji Fuyun; Zong Guifang; Guo Zhongchuan

    1997-01-01

    Several new thermo-hydraulic test facilities are under construction in Nuclear Power Institute of Chinese (NPIC) at Chengdu. These facilities include: 1. Nuclear Power Component Comprehensive Test Facility. 2. Reactor Hydraulic Modeling Test Facility. 3. Control Rod Drive Line Hydraulic Test Facility. 4. Large Scale Thermo-Hydraulic Test Facility. The construction of these facilities will make huge progress in the research and development capability of nuclear power technology in CHINA. The author will present a brief description of the design parameters flowchart and test program of these facilities

  9. UF6 test loop for evaluation and implementation of international enrichment plant safeguards

    International Nuclear Information System (INIS)

    Cooley, J.N.; Fields, L.W.; Swindle, D.W. Jr.

    1987-06-01

    A functional test loop capable of simulating UF 6 flows, pressures, and pipe deposits characteristic of gas centrifuge enrichment plant piping has been designed and fabricated by the Enrichment Safeguards Program of Martin Marietta Energy Systems, Inc., for use by International Atomic Energy Agency (IAEA) at its Safeguards Analytical Laboratory in Seibersdorf, Austria. Purpose of the test loop is twofold: (1) to enable the IAEA to evaluate and to calibrate enrichment safeguards measurement instrumentation to be used in limited frequency-unannounced access (LFUA) inspection strategy measurements at gas centrifuge enrichment plants and (2) to train IAEA inspectors in the use of such instrumentation. The test loop incorporates actual sections of cascade header pipes from the centrifuge enrichment plants subject to IAEA inspections. The test loop is described, applications for its use by the IAEA are detailed, and results from an initial demonstration session using the test loop are summarized

  10. Application of the X-in-the-Loop Testing Method in the FCV Hybrid Degree Test

    Directory of Open Access Journals (Sweden)

    Haiyu Gao

    2018-02-01

    Full Text Available With the development of fuel cell vehicle technology, an effective testing method that can be applied to develop and verify the fuel cell vehicle powertrain system is urgently required. This paper presents the X-in-the-Loop (XiL testing method in the fuel cell vehicle (FCV hybrid degree test to resolve the first and key issues for the powertrain system design, and the test process and scenarios were designed. The hybrid degree is redefined into the static hybrid degree for system architecture design and the dynamic hybrid degree for vehicle control strategy design, and an integrated testing platform was introduced and a testing application was implemented by following the designed testing flowchart with two loops. Experimental validations show that the sizing of the FCE (Fuel Cell Engine, battery pack, and traction motor with the powertrain architecture can be determined, the control strategy can be evaluated seamlessly, and a systematic powertrain testing solution can be achieved through the whole development process. This research has developed a new testing platform and proposed a novel testing method on the fuel cell vehicle powertrain system, which will be a contribution to fuel cell vehicle technology and its industrialization.

  11. Mirror Fusion Test Facility magnet system

    International Nuclear Information System (INIS)

    VanSant, J.H.; Kozman, T.A.; Bulmer, R.H.; Ng, D.S.

    1981-01-01

    In 1979, R.H. Bulmer of Lawrence Livermore National Laboratory (LLNL) discussed a proposed tandem-mirror magnet system for the Mirror Fusion Test Facility (MFTF) at the 8th symposium on Engineering Problems in Fusion Research. Since then, Congress has voted funds for expanding LLNL's MFTF to a tandem-mirror facility (designated MFTF-B). The new facility, scheduled for completion by 1985, will seek to achieve two goals: (1) Energy break-even capability (Q or the ratio of fusion energy to plasma heating energy = 1) of mirror fusion, (2) Engineering feasibility of reactor-scale machines. Briefly stated, 22 superconducting magnets contained in a 11-m-diam by 65-m-long vacuum vessel will confine a fusion plasma fueled by 80 axial streaming-plasma guns and over 40 radial neutral beams. We have already completed a preliminary design of this magnet system

  12. Cold moderator test facilities working group

    International Nuclear Information System (INIS)

    Bauer, Guenter S.; Lucas, A. T.

    1997-09-01

    The working group meeting was chaired by Bauer and Lucas.Testing is a vital part of any cold source development project. This applies to specific physics concept verification, benchmarking in conjunction with computer modeling and engineering testing to confirm the functional viability of a proposed system. Irradiation testing of materials will always be needed to continuously extend a comprehensive and reliable information database. An ever increasing worldwide effort to enhance the performance of reactor and accelerator based neutron sources, coupled with the complexity and rising cost of building new generation facilities, gives a new dimension to cold source development and testing programs. A stronger focus is now being placed on the fine-tuning of cold source design to maximize its effectiveness in fully exploiting the facility. In this context, pulsed spallation neutron sources pose an extra challenge due to requirements regarding pulse width and shape which result from a large variety of different instrument concepts. The working group reviewed these requirements in terms of their consequences on the needs for testing equipment and compiled a list of existing and proposed facilities suitable to carry out the necessary development work.

  13. HTS power lead testing at the Fermilab magnet test facility

    Energy Technology Data Exchange (ETDEWEB)

    Rabehl, R.; Carcagno, R.; Feher, S.; Huang, Y.; Orris, D.; Pischalnikov, Y.; Sylvester, C.; Tartaglia, M.; /Fermilab

    2005-08-01

    The Fermilab Magnet Test Facility has tested high-temperature superconductor (HTS) power leads for cryogenic feed boxes to be placed at the Large Hadron Collider (LHC) interaction regions and at the new BTeV C0 interaction region of the Fermilab Tevatron. A new test facility was designed and operated, successfully testing 20 pairs of HTS power leads for the LHC and 2 pairs of HTS power leads for the BTeV experiment. This paper describes the design and operation of the cryogenics, process controls, data acquisition, and quench management systems. Results from the facility commissioning are included, as is the performance of a new insulation method to prevent frost accumulation on the warm ends of the power leads.

  14. HTS power lead testing at the Fermilab magnet test facility

    International Nuclear Information System (INIS)

    Rabehl, R.; Carcagno, R.; Feher, S.; Huang, Y.; Orris, D.; Pischalnikov, Y.; Sylvester, C.; Tartaglia, M.

    2005-01-01

    The Fermilab Magnet Test Facility has tested high-temperature superconductor (HTS) power leads for cryogenic feed boxes to be placed at the Large Hadron Collider (LHC) interaction regions and at the new BTeV CO interaction region of the Fermilab Tevatron. A new test facility was designed and operated, successfully testing 20 pairs of HTS power leads for the LHC and 2 pairs of HTS power leads for the BTeV experiment. This paper describes the design and operation of the cryogenics, process controls, data acquisition, and quench management systems. Results from the facility commissioning are included, as is the performance of a new insulation method to prevent frost accumulation on the warm ends of the power leads

  15. An Experience of Thermowell Design in RCP Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Y. S.; Kim, B. D.; Youn, Y. J.; Jeon, W. J.; Kim, S.; Bae, B. U.; Cho, Y. J.; Choi, H. S.; Park, J. K; Cho, S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Flow rates for the test should vary in the range of 90% to 130% of rated flowrate under prototypic operational conditions, as shown in Table 1. Generally for the flow control, a combination of a control valve and an orifice was used in previous RCP test facilities. From the commissioning startup of the RCP test facility, it was found the combination of valve and orifice induced quite a large vibration for the RCP. As a solution to minimize the vibration and to facilitate the flowrate control, one of KAERI's staff suggested a variable restriction orifice (VRO), which controls most of the required flowrates except highest flowrates, as shown in Fig. 2. For the highest flowrates, e.g., around run-out flowrate (130%), control valves in bypass lines were also used to achieve required flowrates. From a performance test, it was found the VRO is very effective measures to control flowrates in the RCP test facility. During the commissioning startup operation, one of thermowells located at the upstream of the RCP was cracked due to high speed coolant velocity, which was - fortunately - found under a leakage test before running the RCP test loop. The cracked thermowell, whose tapered-shank was detached from the weld collar after uninstalling, is shown in Fig. 3. As can be seen the figure, most of the cross-section at the root of the thermowell shank was cracked. In this paper, an investigation of the integrity of thermowells in the RCP test facility was performed according to the current code and overall aspects on the thermowell designs were also discussed. An RCP test facility has been constructed in KAERI. During the commissioning startup operation, one of thermowells was cracked due to high speed coolant velocity. To complete the startup operation, a modified design of thermowells was proposed and all the original thermowells were replaced by the modified ones. From evaluation of the original and modified designs of thermowells according to the recent PTC code, the

  16. The Advanced Test Reactor Irradiation Facilities and Capabilities

    International Nuclear Information System (INIS)

    S. Blaine Grover; Raymond V. Furstenau

    2007-01-01

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

  17. A test matrix sequencer for research test facility automation

    Science.gov (United States)

    Mccartney, Timothy P.; Emery, Edward F.

    1990-01-01

    The hardware and software configuration of a Test Matrix Sequencer, a general purpose test matrix profiler that was developed for research test facility automation at the NASA Lewis Research Center, is described. The system provides set points to controllers and contact closures to data systems during the course of a test. The Test Matrix Sequencer consists of a microprocessor controlled system which is operated from a personal computer. The software program, which is the main element of the overall system is interactive and menu driven with pop-up windows and help screens. Analog and digital input/output channels can be controlled from a personal computer using the software program. The Test Matrix Sequencer provides more efficient use of aeronautics test facilities by automating repetitive tasks that were once done manually.

  18. Design of Test Loops for Forced Convection Heat Transfer Studies at Supercritical State

    Science.gov (United States)

    Balouch, Masih N.

    Worldwide research is being conducted to improve the efficiency of nuclear power plants by using supercritical water (SCW) as the working fluid. One such SCW reactor considered for future development is the CANDU-Supercritical Water Reactor (CANDU-SCWR). For safe and accurate design of the CANDU-SCWR, a detailed knowledge of forced-convection heat transfer in SCW is required. For this purpose, two supercritical fluid loops, i.e. a SCW loop and an R-134a loop are developed at Carleton University. The SCW loop is designed to operate at pressures as high as 28 MPa, temperatures up to 600 °C and mass fluxes of up to 3000 kg/m2s. The R-134a loop is designed to operate at pressures as high as 6 MPa, temperatures up to 140 °C and mass fluxes in the range of 500-6000 kg/m2s. The test loops designs allow for up to 300 kW of heating power to be imparted to the fluid. Both test loops are of the closed-loop design, where flow circulation is achieved by a centrifugal pump in the SCW loop and three parallel-connected gear pumps in the R-134a loop, respectively. The test loops are pressurized using a high-pressure nitrogen cylinder and accumulator assembly, which allows independent control of the pressure, while simultaneously dampening pump induced pressure fluctuations. Heat exchangers located upstream of the pumps control the fluid temperature in the test loops. Strategically located measuring instrumentation provides information on the flow rate, pressure and temperature in the test loops. The test loops have been designed to accommodate a variety of test-section geometries, ranging from a straight circular tube to a seven-rod bundle, achieving heat fluxes up to 2.5 MW/m2 depending on the test-section geometry. The design of both test loops allows for easy reconfiguration of the test-section orientation relative to the gravitational direction. All the test sections are of the directly-heated design, where electric current passing through the pressure retaining walls of the

  19. Performance of smokeless gasoline fire test facility

    International Nuclear Information System (INIS)

    Griffin, J.F.; Watkins, R.A.

    1978-01-01

    Packaging for radioactive materials must perform satisfactorily when subjected to temperatures simulating an accident involving a fire. The new thermal test facility has proved to be a reliable method for satisfactorily performing the required test. The flame provides sufficient heat to assure that the test is valid, and the temperature can be controlled satisfactorily. Also, the air and water mist systems virtually eliminate any smoke and thereby exceed the local EPA requirements. The combination of the two systems provides an inexpensive, low maintenance technique for elimination of the smoke plume

  20. Kauai Test Facility hazards assessment document

    International Nuclear Information System (INIS)

    Swihart, A.

    1995-05-01

    The Department of Energy Order 55003A requires facility-specific hazards assessment be prepared, maintained, and used for emergency planning purposes. This hazards assessment document describes the chemical and radiological hazards associated with the Kauai Test Facility, Barking Sands, Kauai, Hawaii. The Kauai Test Facility's chemical and radiological inventories were screened according to potential airborne impact to onsite and offsite individuals. The air dispersion model, ALOHA, estimated pollutant concentrations downwind from the source of a release, taking into consideration the toxicological and physical characteristics of the release site, the atmospheric conditions, and the circumstances of the release. The greatest distance to the Early Severe Health Effects threshold is 4.2 kilometers. The highest emergency classification is a General Emergency at the open-quotes Main Complexclose quotes and a Site Area Emergency at the Kokole Point Launch Site. The Emergency Planning Zone for the open-quotes Main Complexclose quotes is 5 kilometers. The Emergency Planning Zone for the Kokole Point Launch Site is the Pacific Missile Range Facility's site boundary

  1. Dynamic modelling and hardware-in-the-loop testing of PEMFC

    Energy Technology Data Exchange (ETDEWEB)

    Vath, Andreas; Soehn, Matthias; Nicoloso, Norbert; Hartkopf, Thomas [Technische Universitaet Darmstadt/Institut fuer Elektrische Energie wand lung, Landgraf-Georg-Str. 4, D-64283 Darmstadt (Germany); Lemes, Zijad; Maencher, Hubert [MAGNUM Automatisierungstechnik GmbH, Bunsenstr. 22, D-64293 Darmstadt (Germany)

    2006-07-03

    Modelling and hardware-in-the-loop (HIL) testing of fuel cell components and entire systems open new ways for the design and advance development of FCs. In this work proton exchange membrane fuel cells (PEMFC) are dynamically modelled within MATLAB-Simulink at various operation conditions in order to establish a comprehensive description of their dynamic behaviour as well as to explore the modelling facility as a diagnostic tool. Set-up of a hardware-in-the-loop (HIL) system enables real time interaction between the selected hardware and the model. The transport of hydrogen, nitrogen, oxygen, water vapour and liquid water in the gas diffusion and catalyst layers of the stack are incorporated into the model according to their physical and electrochemical characteristics. Other processes investigated include, e.g., the membrane resistance as a function of the water content during fast load changes. Cells are modelled three-dimensionally and dynamically. In case of system simulations a one-dimensional model is preferred to reduce computation time. The model has been verified by experiments with a water-cooled stack. (author)

  2. E-ELT M1 test facility

    Science.gov (United States)

    Dimmler, M.; Marrero, J.; Leveque, S.; Barriga, P.; Sedghi, B.; Mueller, M.

    2012-09-01

    During the advanced design phase of the European Extremely Large Telescope (E-ELT) several critical components have been prototyped. During the last year some of them have been tested in dedicated test stands. In particular, a representative section of the E-ELT primary mirror has been assembled with 2 active and 2 passive segments. This test stand is equipped with complete prototype segment subunits, i.e. including support mechanisms, glass segments, edge sensors, position actuators as well as additional metrology for monitoring. The purpose is to test various procedures such as calibration, alignment and handling and to study control strategies. In addition the achievable component and subsystem performances are evaluated, and interface issues are identified. In this paper an overview of the activities related to the E-ELT M1 Test Facility will be given. Experiences and test results are presented.

  3. Upgraded Features of Newly Constructed Fuel Assembly Mechanical Characterization Test Facility in KAERI

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kang Hee; Kang, Heung Seok; Yoon, Kyung Ho; Kim, Hyung Kyu; Lee, Young Ho; Kim, Soo Ho; Yang, Jae Ho [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    Fuel assembly mechanical characterization test facility (FAMeCT) in KAERI is newly constructed with upgraded functional features such as increased loading capacity, under-water vibration testing and severe earthquake simulation for extended fuel design guideline. The facility building is compactly designed in the scale of 3rd floor building and has regions for assembly-wise mechanical test equipment, dynamic load (seismic) simulating test system, small scale hydraulic loop and component wise test equipment. Figure 1 shows schematic regional layout of the facility building. Mechanical test platform and system is designed to increase loading capacity for axial compression test. Structural stability of the support system of new upper core plate simulator is validated through a limit case functional test. Fuel assembly mechanical characterization test facility in KAERI is newly constructed and upgraded with advanced functional features such as uprated loading capacity, under-water vibration testing and severe earthquake simulation for extended fuel design guideline. This paper briefly introduce the test facility construction and scope of the facility and is focused on the upgraded design features of the facility. Authors hope to facilitate the facility more in the future and collaborate with the industry.

  4. Experimental data report for transient flow calibration facility tests IIB101, IIB102 and IIB201

    International Nuclear Information System (INIS)

    Wambach, J.L.

    1980-01-01

    Thermal-hydraulic response data are presented for the transient performance tests of a pitot tube rake (IIB201) and a modular drag disc-turbine transducer (DTT) rake (IIB101, IIB102). The tests were conducted in a system which provided full scale simulation of the pressure vessel and broken loop hot leg piping of the Loss of Fluid Test Facility (LOFT). A load cell system was used to provide a reference mass flow rate measurement

  5. FY 1995 progress report on the ANS thermal-hydraulic test loop operation and results

    Energy Technology Data Exchange (ETDEWEB)

    Siman-Tov, M.; Felde, D.K.; Farquharson, G.; McDuffee, J.L.; McFee, M.T.; Ruggles, A.E.; Wendel, M.W.; Yoder, G.L.

    1997-07-01

    The Thermal-Hydraulic Test Loop (THTL) is an experimental facility constructed to support the development of the Advanced Neutron Source Reactor (ANSR) at Oak Ridge National Laboratory (ORNL). The THTL facility was designed and built to provide known thermal-hydraulic (T/H) conditions for a simulated full-length coolant subchannel of the ANS reactor core, thus facilitating experimental determination of FE and CHF thermal limits under expected ANSR T/H conditions. Special consideration was given to allow operation of the system in a stiff mode (constant flow) and in a soft mode (constant pressure drop) for proper implementation of true FE and DNB experiments. The facility is also designed to examine other T/H phenomena, including onset of incipient boiling (IB), single-phase heat transfer coefficients and friction factors, and two-phase heat transfer and pressure drop characteristics. Tests will also be conducted that are representative of decay heat levels at both high pressure and low pressure as well as other quasi-equilibrium conditions encountered during transient scenarios. A total of 22 FE tests and 2 CHF tests were performed during FY 1994 and FY 1995 with water flowing vertically upward. Comparison of these data as well as extensive data from other investigators led to a proposed modification to the Saha and Zuber correlation for onset of significant void (OSV), applied to FE prediction. The modification takes into account a demonstrated dependence of the OSV or FE thermal limits on subcooling levels, especially in the low subcooling regime.

  6. FY 1995 progress report on the ANS thermal-hydraulic test loop operation and results

    International Nuclear Information System (INIS)

    Siman-Tov, M.; Felde, D.K.; Farquharson, G.; McDuffee, J.L.; McFee, M.T.; Ruggles, A.E.; Wendel, M.W.; Yoder, G.L.

    1997-07-01

    The Thermal-Hydraulic Test Loop (THTL) is an experimental facility constructed to support the development of the Advanced Neutron Source Reactor (ANSR) at Oak Ridge National Laboratory (ORNL). The THTL facility was designed and built to provide known thermal-hydraulic (T/H) conditions for a simulated full-length coolant subchannel of the ANS reactor core, thus facilitating experimental determination of FE and CHF thermal limits under expected ANSR T/H conditions. Special consideration was given to allow operation of the system in a stiff mode (constant flow) and in a soft mode (constant pressure drop) for proper implementation of true FE and DNB experiments. The facility is also designed to examine other T/H phenomena, including onset of incipient boiling (IB), single-phase heat transfer coefficients and friction factors, and two-phase heat transfer and pressure drop characteristics. Tests will also be conducted that are representative of decay heat levels at both high pressure and low pressure as well as other quasi-equilibrium conditions encountered during transient scenarios. A total of 22 FE tests and 2 CHF tests were performed during FY 1994 and FY 1995 with water flowing vertically upward. Comparison of these data as well as extensive data from other investigators led to a proposed modification to the Saha and Zuber correlation for onset of significant void (OSV), applied to FE prediction. The modification takes into account a demonstrated dependence of the OSV or FE thermal limits on subcooling levels, especially in the low subcooling regime

  7. Large lithium loop experience

    International Nuclear Information System (INIS)

    Kolowith, R.; Owen, T.J.; Berg, J.D.; Atwood, J.M.

    1981-10-01

    An engineering design and operating experience of a large, isothermal, lithium-coolant test loop are presented. This liquid metal coolant loop is called the Experimental Lithium System (ELS) and has operated safely and reliably for over 6500 hours through September 1981. The loop is used for full-scale testing of components for the Fusion Materials Irradiation Test (FMIT) Facility. Main system parameters include coolant temperatures to 430 0 C and flow to 0.038 m 3 /s (600 gal/min). Performance of the main pump, vacuum system, and control system is discussed. Unique test capabilities of the ELS are also discussed

  8. Safety assessment for the rf Test Facility

    International Nuclear Information System (INIS)

    Nagy, A.; Beane, F.

    1984-08-01

    The Radio Frequency Test Facility (RFTF) is a part of the Magnetic Fusion Program's rf Heating Experiments. The goal of the Magnetic Fusion Program (MFP) is to develop and demonstrate the practical application of fusion. RFTF is an experimental device which will provide an essential link in the research effort aiming at the realization of fusion power. This report was compiled as a summary of the analysis done to ensure the safe operation of RFTF

  9. The Brookhaven National Laboratory Accelerator Test Facility

    International Nuclear Information System (INIS)

    Batchelor, K.

    1992-01-01

    The Brookhaven National Laboratory Accelerator Test Facility comprises a 50 MeV traveling wave electron linear accelerator utilizing a high gradient, photo-excited, raidofrequency electron gun as an injector and an experimental area for study of new acceleration methods or advanced radiation sources using free electron lasers. Early operation of the linear accelerator system including calculated and measured beam parameters are presented together with the experimental program for accelerator physics and free electron laser studies

  10. Underground large scale test facility for rocks

    International Nuclear Information System (INIS)

    Sundaram, P.N.

    1981-01-01

    This brief note discusses two advantages of locating the facility for testing rock specimens of large dimensions in an underground space. Such an environment can be made to contribute part of the enormous axial load and stiffness requirements needed to get complete stress-strain behavior. The high pressure vessel may also be located below the floor level since the lateral confinement afforded by the rock mass may help to reduce the thickness of the vessel

  11. Operating experience of steam generator test facility

    International Nuclear Information System (INIS)

    Sureshkumar, V.A.; Madhusoodhanan, G.; Noushad, I.B.; Ellappan, T.R.; Nashine, B.K.; Sylvia, J.I.; Rajan, K.K.; Kalyanasundaram, P.; Vaidyanathan, G.

    2006-01-01

    Steam Generator (SG) is the vital component of a Fast Reactor. It houses both water at high pressure and sodium at low pressure separated by a tube wall. Any damage to this barrier initiates sodium water reaction that could badly affect the plant availability. Steam Generator Test Facility (SGTF) has been set up in Indira Gandhi Centre for Atomic Research (IGCAR) to test sodium heated once through steam generator of 19 tubes similar to the PFBR SG dimension and operating conditions. The facility is also planned as a test bed to assess improved designs of the auxiliary equipments used in Fast Breeder Reactors (FBR). The maximum power of the facility is 5.7 MWt. This rating is arrived at based on techno economic consideration. This paper covers the performance of various equipments in the system such as Electro magnetic pumps, Centrifugal sodium pump, in-sodium hydrogen meters, immersion heaters, and instrumentation and control systems. Experience in the system operation, minor modifications, overall safety performance, and highlights of the experiments carried out etc. are also brought out. (author)

  12. Project description: ORNL PWR blowdown heat transfer separate-effects program, Thermal-Hydraulic Test Facility (THTF)

    International Nuclear Information System (INIS)

    1976-02-01

    The ORNL Pressurized-Water Reactor Blowdown Heat Transfer (PWR-BDHT) Program is an experimental separate-effects study of the relations among the principal variables that can alter the rate of blowdown, the presence of flow reversal and rereversal, time delay to critical heat flux, the rate at which dryout progresses, and similar time-related functions that are important to LOCA analysis. Primary test results will be obtained from the Thermal-Hydraulic Test Facility (THTF), a large nonnuclear pressurized-water loop that incorporates a 49-rod electrically heated bundle. Supporting experiments will be carried out in two additional test loops - the Forced Convection Test Facility (FCTF), a small high-pressure facility in which single heater rods can be tested in annular geometry; and an air-water loop which is used to evaluate two-phase flow-measuring instrumentation

  13. 40 CFR 792.43 - Test system care facilities.

    Science.gov (United States)

    2010-07-01

    .... (a) A testing facility shall have a sufficient number of animal rooms or other test system areas, as... different tests. (b) A testing facility shall have a number of animal rooms or other test system areas... waste and refuse or for safe sanitary storage of waste before removal from the testing facility...

  14. Review of design criteria and safety analysis of safety class electric building for fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. Y.

    1998-02-01

    Steady state fuel test loop will be equipped in HANARO to obtain the development and betterment of advanced fuel and materials through the irradiation tests. HANARO fuel test loop was designed for CANDU and PWR fuel testing. Safety related system of Fuel Test Loop such as emergency cooling water system, component cooling water system, safety ventilation system, high energy line break mitigation system and remote control room was required 1E class electric supply to meet the safety operation in accordance with related code. Therefore, FTL electric building was designed to construction and install the related equipment based on seismic category I. The objective of this study is to review the design criteria and analysis the safety function of safety class electric building for fuel test loop, and this results will become guidance for the irradiation testing in future. (author). 10 refs., 6 tabs., 30 figs.

  15. 305 Building Cold Test Facility Management Plan

    International Nuclear Information System (INIS)

    Whitehurst, R.

    1994-01-01

    This document provides direction for the conduct of business in Building 305 for cold testing tools and equipment. The Cold Test Facility represents a small portion of the overall building, and as such, the work instructions already implemented in the 305 Building will be utilized. Specific to the Cold Test there are three phases for the tools and equipment as follows: 1. Development and feature tests of sludge/fuel characterization equipment, fuel containerization equipment, and sludge containerization equipment to be used in K-Basin. 2. Functional and acceptance tests of all like equipment to be installed and operated in K-Basin. 3. Training and qualification of K-Basin Operators on equipment to be installed and operated in the Basin

  16. Simulation Facilities and Test Beds for Galileo

    Science.gov (United States)

    Schlarmann, Bernhard Kl.; Leonard, Arian

    2002-01-01

    Galileo is the European satellite navigation system, financed by the European Space Agency (ESA) and the European Commission (EC). The Galileo System, currently under definition phase, will offer seamless global coverage, providing state-of-the-art positioning and timing services. Galileo services will include a standard service targeted at mass market users, an augmented integrity service, providing integrity warnings when fault occur and Public Regulated Services (ensuring a continuity of service for the public users). Other services are under consideration (SAR and integrated communications). Galileo will be interoperable with GPS, and will be complemented by local elements that will enhance the services for specific local users. In the frame of the Galileo definition phase, several system design and simulation facilities and test beds have been defined and developed for the coming phases of the project, respectively they are currently under development. These are mainly the following tools: Galileo Mission Analysis Simulator to design the Space Segment, especially to support constellation design, deployment and replacement. Galileo Service Volume Simulator to analyse the global performance requirements based on a coverage analysis for different service levels and degrades modes. Galileo System Simulation Facility is a sophisticated end-to-end simulation tool to assess the navigation performances for a complete variety of users under different operating conditions and different modes. Galileo Signal Validation Facility to evaluate signal and message structures for Galileo. Galileo System Test Bed (Version 1) to assess and refine the Orbit Determination &Time Synchronisation and Integrity algorithms, through experiments relying on GPS space infrastructure. This paper presents an overview on the so called "G-Facilities" and describes the use of the different system design tools during the project life cycle in order to design the system with respect to

  17. 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

  18. Alpha Fuels Environmental Test Facility impact gun

    International Nuclear Information System (INIS)

    Anderson, C.G.

    1978-01-01

    The Alpha Fuels Environmental Test Facility (AFETF) impact gun is a unique tool for impact testing 238 PuO 2 -fueled heat sources of up to 178-mm dia at velocities to 300 m/s. An environmentally-sealed vacuum chamber at the muzzle of the gun allows preheating of the projectile to 1,000 0 C. Immediately prior to impact, the heat source projectile is completely sealed in a vacuum-tight catching container to prevent escape of its radioactive contents should rupture occur. The impact velocity delivered by this gas-powered gun can be regulated to within +-2%

  19. Production circulator fabrication and testing for core flow test loop. Final report, Phase III

    Energy Technology Data Exchange (ETDEWEB)

    1981-05-01

    The performance testing of two production helium circulators utilizing gas film lubrication is described. These two centrifugal-type circulators plus an identical circulator prototype will be arranged in series to provide the helium flow requirements for the Core Flow Test Loop which is part of the Gas-Cooled Fast Breeder Reactor Program (GCFR) at the Oak Ridge National Laboratory. This report presents the results of the Phase III performance and supplemental tests, which were carried out by MTI during the period of December 18, 1980 through March 19, 1981. Specific test procedures are outlined and described, as are individual tests for measuring the performance of the circulators. Test data and run descriptions are presented.

  20. Production circulator fabrication and testing for core flow test loop. Final report, Phase III

    International Nuclear Information System (INIS)

    1981-05-01

    The performance testing of two production helium circulators utilizing gas film lubrication is described. These two centrifugal-type circulators plus an identical circulator prototype will be arranged in series to provide the helium flow requirements for the Core Flow Test Loop which is part of the Gas-Cooled Fast Breeder Reactor Program (GCFR) at the Oak Ridge National Laboratory. This report presents the results of the Phase III performance and supplemental tests, which were carried out by MTI during the period of December 18, 1980 through March 19, 1981. Specific test procedures are outlined and described, as are individual tests for measuring the performance of the circulators. Test data and run descriptions are presented

  1. 40 CFR 160.43 - Test system care facilities.

    Science.gov (United States)

    2010-07-01

    ... testing facility shall have a number of animal rooms or other test system areas separate from those... housed, facilities shall exist for the collection and disposal of all animal waste and refuse or for safe sanitary storage of waste before removal from the testing facility. Disposal facilities shall be so...

  2. Description and Operational Experiences of the Engineering Test Facility - Helium Technology (ETF-HT)

    International Nuclear Information System (INIS)

    Zhang Zuoyi; Yang Mingde; Bo Hanliang; Duan Riqqiang; Zhu Hongye

    2014-01-01

    This paper presents the configuration of the Engineering Test Facility - Helium Technology (ETF-HT) and the information of its key components and subsystems, which is located in the Changping campus of Tsinghua University. The ETF-HT facility began to be constructed in Jan. 2009. The main objective of the facility is to test and verify the thermo-hydraulic performance of one full-sized modular unit of HTR-PM helically coiled SG assembly. In the ETF-HT facility, electricity energy is used to heat the loop helium, centrifugal blower is used to circulate the helium medium, and the heat sink is one would-tested SG module. Up to now, except for the tested SG module, preheater and hot gas duct under way of construction, the other components has been installed in situ. Via the temporary connection of the installed components, the preliminary operation of the loop has been carried out to test its performances as can be done, which include the loop leak tightness, blower pneumatic performance and electrical heater at partial thermal load. (author)

  3. Status on the construction of the fuel irradiation test facility

    International Nuclear Information System (INIS)

    Park, Kook Nam; Sim, Bong Shick; Lee, Chung Young; Yoo, Seong Yeon

    2005-01-01

    As a facility to examine general performance of nuclear fuel under irradiation condition in HANARO, Fuel Test Loop(FTL) has been developed which can accommodate 3 fuel pins at the core irradiation hole(IR1 hole) taking consideration user's test requirement. 3-Pin FTL consists of In-Pile Test Section (IPS) and Out-of- Pile System (OPS). Test condition in IPS such as pressure, temperature and the water quality, can be controlled by OPS. 3-Pin FTL Conceptual design was set up in 2001 and had completed detail design including a design requirement and basic Piping and Instrument Diagram (P and ID) in 2004. The safety analysis report was prepared and submitted in early 2005 to the regulatory body(KINS) for review and approval of FTL. In 2005, the development team is going to purchase and manufacture hardware and make a contract for construction work. In 2006, the development team is going to install an FTL system performance test shall be done as a part of commissioning. After a 3-Pin FTL development which is expected to be finished by the 2007, FTL will be used for the irradiation test of the new PWR-type fuel and the usage of HANARO will be enhanced

  4. A two-loop test of M(atrix) theory

    International Nuclear Information System (INIS)

    Becker, K.

    1997-01-01

    We consider the scattering of two Dirichlet zero-branes in M(atrix) theory. Using the formulation of M(atrix) theory in terms of ten-dimensional super Yang-Mills theory dimensionally reduced to (0+1) dimensions, we obtain the effective (velocity-dependent) potential describing these particles. At one loop we obtain the well-known result for the leading order of the effective potential V eff ∝v 4 /r 7 , where v and r are the relative velocity and distance between the two zero-branes, respectively. A calculation of the effective potential at two loops shows that no renormalizations of the v 4 term of the effective potential occur at this order. (orig.)

  5. Vitrification Facility integrated system performance testing report

    International Nuclear Information System (INIS)

    Elliott, D.

    1997-01-01

    This report provides a summary of component and system performance testing associated with the Vitrification Facility (VF) following construction turnover. The VF at the West Valley Demonstration Project (WVDP) was designed to convert stored radioactive waste into a stable glass form for eventual disposal in a federal repository. Following an initial Functional and Checkout Testing of Systems (FACTS) Program and subsequent conversion of test stand equipment into the final VF, a testing program was executed to demonstrate successful performance of the components, subsystems, and systems that make up the vitrification process. Systems were started up and brought on line as construction was completed, until integrated system operation could be demonstrated to produce borosilicate glass using nonradioactive waste simulant. Integrated system testing and operation culminated with a successful Operational Readiness Review (ORR) and Department of Energy (DOE) approval to initiate vitrification of high-level waste (HLW) on June 19, 1996. Performance and integrated operational test runs conducted during the test program provided a means for critical examination, observation, and evaluation of the vitrification system. Test data taken for each Test Instruction Procedure (TIP) was used to evaluate component performance against system design and acceptance criteria, while test observations were used to correct, modify, or improve system operation. This process was critical in establishing operating conditions for the entire vitrification process

  6. Advanced Test Reactor National Scientific User Facility

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  7. Digital tape unit test facility software

    Science.gov (United States)

    Jackson, J. T.

    1971-01-01

    Two computer programs are described which are used for the collection and analysis of data from the digital tape unit test facility (DTUTF). The data are the recorded results of skew tests made on magnetic digital tapes which are used on computers as input/output media. The results of each tape test are keypunched onto an 80 column computer card. The format of the card is checked and the card image is stored on a master summary tape via the DTUTF card checking and tape updating system. The master summary tape containing the results of all the tape tests is then used for analysis as input to the DTUTF histogram generating system which produces a histogram of skew vs. date for selected data, followed by some statistical analysis of the data.

  8. TFTR neutral-beam test facility

    International Nuclear Information System (INIS)

    Turitzin, N.M.; Newman, R.A.

    1981-11-01

    TFTR Neutral Beam System will have thirteen discharge ion sources, each with its own power supply. Twelve of these will be utilized for supplemental heating of the TFTR tokamak plasma, while the thirteenth will be dedicated to an off-machine test chamber for source development and/or conditioning. A test installation for one source was set up using prototype equipment to discover and correct possible deficiencies, and to properly coordinate the equipment. This test facility represents the first opportunity for assembling an integrated system of hardware supplied by diverse vendors, each of whom designed and built his equipment to performance specifications. For the installation and coordination of the different portions of the total system, particular attention was given to personnel safety and safe equipment operation. This paper discusses various system components, their characteristics, interconnection and control. Results of the recently initiated test phase will be reported at a later date

  9. Advanced Test Reactor National Scientific User Facility

    Energy Technology Data Exchange (ETDEWEB)

    Frances M. Marshall; Jeff Benson; Mary Catherine Thelen

    2011-08-01

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

  10. Experimental Study on the Natural Circulation Characteristics in the Primary Loop of the SMART Reactor by using the VISTA Facility

    International Nuclear Information System (INIS)

    Park, Hyun-Sik; Choi, Ki-Yong; Cho, Seok; Yi, Sung-Jae; Park, Choon-Kyung; Chung, Moon-Ki

    2007-01-01

    The SMART uses a two-phase natural circulation in the PRHRS loop to remove the heat from the steam generators to the PRHRS heat exchangers, while a single phase natural circulation occurs in the primary loop to transfer the decay heat from the core to the steam generator. Natural circulation operation with a power range of 20 ∼ 25% was considered for SMART and nowadays the possibility of increasing the power level during the natural circulation operation is being investigated. Previously Park et al. performed several experiments by using the VISTA facility on the thermal-hydraulic characteristics of the PRHRS for the SMART-P, which includes a single-phase natural circulation in the primary loop. From the analysis with the TASS-SMR code it was shown that the reference temperature for the primary steam generator inlet temperature should be increased in order to compensate for the decreased core flow. To investigate the possibility of an increase of the power and reference temperature, it is necessary to get experimental data to characterize the natural circulation phenomena in the primary loop of the SMART. In this paper, the characteristics of natural circulation in the primary loop are experimentally investigated during various operational conditions by using the VISTA facility

  11. The Great Plains Wind Power Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    Schroeder, John [Texas Tech Univ., Lubbock, TX (United States)

    2014-01-30

    This multi-year, multi-faceted project was focused on the continued development of a nationally-recognized facility for the testing, characterization, and improvement of grid-connected wind turbines, integrated wind-water desalination systems, and related educational and outreach topics. The project involved numerous faculty and graduate students from various engineering departments, as well as others from the departments of Geosciences (in particular the Atmospheric Science Group) and Economics. It was organized through the National Wind Institute (NWI), which serves as an intellectual hub for interdisciplinary and transdisciplinary research, commercialization and education related to wind science, wind energy, wind engineering and wind hazard mitigation at Texas Tech University (TTU). Largely executed by an academic based team, the project resulted in approximately 38 peer-reviewed publications, 99 conference presentations, the development/expansion of several experimental facilities, and two provisional patents.

  12. Facility effluent monitoring plan for the fast flux test facility

    International Nuclear Information System (INIS)

    Nickels, J.M.; Dahl, N.R.

    1992-11-01

    A facility effluent monitoring plan is required by the US Department of Energy in US Department of Energy Order 5400.1 for any operations that involve hazardous materials and radioactive substances that could affect employee or public safety or the environment. A Facility Effluent Monitoring Plan determination was performed during calendar year 1991 and the evaluation requires the need for a facility effluent monitoring plan. This facility effluent monitoring plan assesses effluent monitoring systems and evaluates whether they are adequate to ensure the public health and safety as specified in applicable federal, state, and local requirements

  13. Results from the CLIC Test Facility

    CERN Document Server

    Braun, H; Bossart, Rudolf; Chautard, F; Corsini, R; Delahaye, J P; Godot, J C; Hutchins, S; Kamber, I; Madsen, J H B; Rinolfi, Louis; Rossat, G; Schreiber, S; Suberlucq, Guy; Thorndahl, L; Wilson, Ian H; Wuensch, Walter

    1996-01-01

    In order to study the principle of the Compact Linear Collider (CLIC) based on the Two Beam Acceleration (TBA) scheme at high frequency, a CLIC Test Facility (CTF) has been set-up at CERN. After four years of successful running, the experimental programme is now fully completed and all its objectives reached, particularly the generation of a high intensity drive beam with short bunches by a photo-injector, the production of 30 GHz RF power and the acceleration of a probe beam by 30 GHz structures. A summary of the CTF results and their impact on linear collider design is given. This covers 30 GHz high power testing, study of intense, short single bunches; as well as RF-Gun, photocathode and beam diagnostic developments. A second phase of the test facility (CTF2) is presently being installed to demonstrate the feasibility of the TBA scheme by constructing a fully engineered, 10 m long, test section very similar to the CLIC drive and main linacs, producing up to 480 MW of peak RF power at 30 GHz and acceleratin...

  14. Test facility for rewetting experiments at CDTN

    International Nuclear Information System (INIS)

    Rezende, Hugo C.; Mesquita, Amir Z.; Ladeira, Luiz C.D.; Santos, Andre A.C.

    2015-01-01

    One of the most important subjects in nuclear reactor safety analysis is the reactor core rewetting after a Loss-of-Coolant Accident (LOCA) in a Light Water Reactor LWR. Several codes for the prediction of the rewetting evolution are under development based on experimental results. In a Pressurized Water Reactor (PWR) the reflooding phase of a LOCA is when the fuel rods are rewetted from the bottom of the core to its top after having been totally uncovered and dried out. Out-of-pile reflooding experiments performed with electrical heated fuel rod simulators show different quench behavior depending the rods geometry. A test facility for rewetting experiments (ITR - Instalacao de Testes de Remolhamento) has been constructed at the Thermal Hydraulics Laboratory of the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), with the objective of performing investigations on basic phenomena that occur during the reflood phase of a LOCA in a PWR, using tubular and annular test sections. This paper presents the design aspects of the facility, and the current stage of the works. The mechanical aspects of the installation as its instrumentation are described. Two typical tests are presented and results compered with theoretical calculations using computer code. (author)

  15. Thermionic system evaluated test (TSET) facility description

    Science.gov (United States)

    Fairchild, Jerry F.; Koonmen, James P.; Thome, Frank V.

    1992-01-01

    A consortium of US agencies are involved in the Thermionic System Evaluation Test (TSET) which is being supported by the Strategic Defense Initiative Organization (SDIO). The project is a ground test of an unfueled Soviet TOPAZ-II in-core thermionic space reactor powered by electrical heat. It is part of the United States' national thermionic space nuclear power program. It will be tested in Albuquerque, New Mexico at the New Mexico Engineering Research Institute complex by the Phillips Laboratoty, Sandia National Laboratories, Los Alamos National Laboratory, and the University of New Mexico. One of TSET's many objectives is to demonstrate that the US can operate and test a complete space nuclear power system, in the electrical heater configuration, at a low cost. Great efforts have been made to help reduce facility costs during the first phase of this project. These costs include structural, mechanical, and electrical modifications to the existing facility as well as the installation of additional emergency systems to mitigate the effects of utility power losses and alkali metal fires.

  16. Test facility for rewetting experiments at CDTN

    Energy Technology Data Exchange (ETDEWEB)

    Rezende, Hugo C.; Mesquita, Amir Z.; Ladeira, Luiz C.D.; Santos, Andre A.C., E-mail: hcr@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (SETRE/CDTN/CNEN-MG), Belo Horizonte, MG (Brazil). Servico de Tecnologia de Reatores

    2015-07-01

    One of the most important subjects in nuclear reactor safety analysis is the reactor core rewetting after a Loss-of-Coolant Accident (LOCA) in a Light Water Reactor LWR. Several codes for the prediction of the rewetting evolution are under development based on experimental results. In a Pressurized Water Reactor (PWR) the reflooding phase of a LOCA is when the fuel rods are rewetted from the bottom of the core to its top after having been totally uncovered and dried out. Out-of-pile reflooding experiments performed with electrical heated fuel rod simulators show different quench behavior depending the rods geometry. A test facility for rewetting experiments (ITR - Instalacao de Testes de Remolhamento) has been constructed at the Thermal Hydraulics Laboratory of the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), with the objective of performing investigations on basic phenomena that occur during the reflood phase of a LOCA in a PWR, using tubular and annular test sections. This paper presents the design aspects of the facility, and the current stage of the works. The mechanical aspects of the installation as its instrumentation are described. Two typical tests are presented and results compered with theoretical calculations using computer code. (author)

  17. Facility for endurance tests of thermal insulations

    International Nuclear Information System (INIS)

    Mauersberger, R.

    1984-01-01

    In the following report the design and construction of an experimental facility for endurance tests of thermal insulations is presented. It's name in abbreviation is 'ADI' standing for the German words A nlage zum Dauertest von Isolierungen . This test facility was build by HRB in order to investigate the performance of thermal insulation systems of hot gas ducts for the process heat-reactor-project. The tests are intended to simulate the conditions of reactor operation. They include short-time experiments for selection of insulation-concepts and in a second step long-time experiments as performance tests. During these tests are measured the effective heat conductivity the local heat losses the temperature profiles of the insulation, of the fixing elements and along the wall of the duct. The design-data required to perform all these tasks are shown in the first picture: The gas-atmosphere must be Helium in tests like in reactor with regard to the special thermal and hydraulic properties of Helium and to the influence of Helium on mechanic friction and wear. The hot gas temperature in the PNP-reactor will be 950 deg. C and should be equal in the experiments. The temperature on the cold side of the insulation has to be adjustable from 50 deg. C up to 300 deg. C. The Helium pressure in the hot gas ducts of a HTR-plant is about 42 bar. The ADI was laid out for 70 bar to cover the hole range of interest. A Helium mass flow has to stream through the insulated test duct in order to realize equal temperatures on the hot side of the insulation. A flow rate of 4,5 kg/s is sufficient for this requirement. The axial pressure gradient along the insulation must be the same as in the reactor, because this has an essential influence on the heat losses. This pressure gradient is about 40 Pa/m

  18. Reflooding phenomena of German PWR estimated from CCTF [Cylindrical Core Test Facility], SCTF [Slab Core Test Facility] and UPTF [Upper Plenum Test Facility] results

    International Nuclear Information System (INIS)

    Murao, Y.; Iguchi, T.; Sugimoto, J.

    1988-09-01

    The reflooding behavior in a PWR with a combined injection type ECCS was studied by comparing the test results from Cylindrical Core Test Facility (CCTF), Slab Core Test Facility (SCTF) and Upper Plenum Test Facility (UPTF). Core thermal-hydraulics is discussed mainly based on SCTF test data. In addition, the water accumulation behavior in hot legs and the break-through characteristics at tie plate are discussed

  19. Cylindrical core reflood test facility (CCTF) and slab core reflood test facility (SCTF) for Japan Atomic Energy Research Institute (JAERI)

    International Nuclear Information System (INIS)

    1981-01-01

    IHI has designed and constructed the CCTF at JAERI to be used in the safety analysis research on the loss of coolant accident in a PWR plant. This test facility is planned so that reflood phenomenon in the PWR plant (a phenomenon is that the bared and overheated core is reflooded by the emergency core cooling system when the coolant loss accident occurred) is simulated under various test conditions. The CCTF is the largest-scale test plant in the world, composed of approximately 2000 simulated fuel rods (electric heaters), 1 simulated pressure vessel, 4 primary cooling loops, 2 simulated steam generators, emergency core cooling system, and so on. The test conditions are controlled, and the test steps are sequentially progressed by the computing system, and test data are collected by the data acquisition system. Furthermore, IHI is now designing and constructing the SCTF in accordance with the JAERI research plan. The SCTF is similar to the CCTF in scale. Main feature of the SCTF is the form of the simulated core and the simulated pressure vessel, which is of slab construction to be representative of the radial section of the PWR reactor. Reliable and various data for safety analysis are expected by the CCTF and the SCTF. (author)

  20. Development of Integral Effect Test Facility P and ID and Technical Specification for SMART Fluid System

    International Nuclear Information System (INIS)

    Lee, Sang Il; Jung, Y. H.; Yang, H. J.; Song, S. Y.; Han, O. J.; Lee, B. J.; Kim, Y. A.; Lim, J. H.; Park, K. W.; Kim, N. G.

    2010-01-01

    SMART integral test loop is the thermal hydraulic test facility with a high pressure and temperature for simulating the major systems of the prototype reactor, SMART-330. The objective of this project is to conduct the basic design for constructing SMART ITL. The major results of this project include a series of design documents, technical specifications and P and ID. The results can be used as the fundamental materials for making the detailed design which is essential for manufacturing and installing SMART ITL

  1. Testing of cobalt-free alloys for valve applications using a special test loop

    International Nuclear Information System (INIS)

    Benhamou, C.

    1992-01-01

    Considering that use of cobalt alloys should be avoided as far as possible in PWR components, a programme aimed at establishing the performance of cobalt-free alloys has been performed for valve applications, where cobalt alloys are mainly used. Referring to past work, two types of cobalt-free alloys were selected: Ni-Cr-B-Si and Ni-Cr-Fe alloys. Cobalt-free valves' behaviour has been evaluated comparatively with cobalt valves by implementation of a programme in a special PWR test loop. At the issue of the loop test programme, which included endurance, thermal shock and erosion tests, cobalt-free alloys candidate to replace cobalt alloys are proposed in relation with valve type (globe valve and swing check valve). The following was established: (i) Colmonoy 4-26 (Ni-Cr-B-Si alloy) and Cenium Z20 (Ni-Cr-Fe alloy) deposited by plasma arc process were found suitable for use in 3inch swing check valves; (ii) for integral parts acting as guide rings, Nitronic 60 and Cesium Z20/698 were tested successfully; (iii) for small-bore components such as 2inch globe valves, no solution can yet be proposed; introduction of cobalt-free alloys is dependent on the development of automatic advanced arc surfacing techniques applied to small-bore components

  2. An automated test facility for neutronic amplifiers

    International Nuclear Information System (INIS)

    Beattie, W.J.

    1997-01-01

    Neutronic amplifiers are used at the Chalk River Laboratory in applications such as neutron flux monitoring and reactor control systems. Routine preventive maintenance of control and safety systems included annual calibration and characterization of the neutronic amplifiers. An investigation into the traditional methods of annual routine maintenance of amplifiers concluded that frequency and phase response measurements in particular were labour intensive and subject to non-repeatable errors. A decision was made to upgrade testing methods and facilities by using programmable test equipment under the control of a computer. In order to verify the results of the routine measurements, expressions for the transfer functions were derived from the circuit diagrams. Frequency and phase responses were then calculated and plotted thus providing a bench-mark to which the test results can be compared. (author)

  3. Heating facility for blanket and performance test

    Energy Technology Data Exchange (ETDEWEB)

    Furuya, Kazuyuki; Kuroda, Toshimasa; Enoeda, Mikio; Sato, Satoshi; Hatano, Toshihisa; Takatsu, Hideyuki; Ohara, Yoshihiro [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Hara, Shigemitsu

    1999-03-01

    A design and a fabrication of heating test facility for a mock-up of the blanket module to be installed in International Thermonuclear Experimental Reactor (ITER) have been conducted to evaluate/demonstrate its heat removal performance and structural soundness under cyclic heat loads. To simulate surface heat flux to the blanket module, infrared heating method is adopted so as to heat large surface area uniformly. The infrared heater is used in vacuum environment (10{sup -4} Torr{approx}), and the lamps are cooled by air flowing through an annulus between the lamp and a cover tube made of quartz glass. Elastomer O rings (available to be used up to {approx}300degC) and used for vacuum seal at outer surface of the cover tube. To prevent excessive heating of the O ring, the end part of the cover tube is specially designed including the tube shape, flow path of air and gold coating on the surface of the cover tube to protect the O ring against thermal radiation from glowing tungsten filament. To examine the performance of the facility, steady state and cyclic operation of the infrared heater were conducted using a small-scaled shielding blanket mock-up as a test specimen. The important results are as follows: (1) Heat flux at the surface of the small-scaled mock-up measured by a calorimeter was {approx}0.2 MW/m{sup 2}. (2) A comparison of thermal analysis results and measured temperature responses showed that the small-scaled mock-up had good heat removal performance. (3) Steady state operation and cyclic operation with step response between the rated and zero powers of the infrared heater were successfully performed, and it was confirmed that this heating facility was well-prepared and available for the thermal cyclic test of a blanket module. (author)

  4. Large coil test facility conceptual design report

    International Nuclear Information System (INIS)

    Nelms, L.W.; Thompson, P.B.; Mann, T.L.

    1978-02-01

    In the development of a superconducting toroidal field (TF) magnet for The Next Step (TNS) tokamak reactor, several different TF coils, about half TNS size, will be built and tested to permit selection of a design and fabrication procedure for full-scale TNS coils. A conceptual design has been completed for a facility to test D-shaped TF coils, 2.5 x 3.5-m bore, operating at 4-6 K, cooled either by boiling helium or by forced-flow supercritical helium. Up to six coils can be accommodated in a toroidal array housed in a single vacuum tank. The principal components and systems in the facility are an 11-m vacuum tank, a test stand providing structural support and service connections for the coils, a liquid nitrogen system, a system providing helium both as saturated liquid and at supercritical pressure, coils to produce a pulsed vertical field at any selected test coil position, coil power supplies, process instrumentation and control, coil diagnostics, and a data acquisition and handling system. The test stand structure is composed of a central bucking post, a base structure, and two horizontal torque rings. The coils are bolted to the bucking post, which transmits all gravity loads to the base structure. The torque ring structure, consisting of beams between adjacent coils, acts with the bucking structure to react all the magnetic loads that occur when the coils are energized. Liquid helium is used to cool the test stand structure to 5 K to minimize heat conduction to the coils. Liquid nitrogen is used to precool gaseous helium during system cooldown and to provide thermal radiation shielding

  5. FMIT - the fusion materials irradiation test facility

    International Nuclear Information System (INIS)

    Liska, D.J.

    1980-01-01

    A joint effort by the Hanford Engineering Development Laboratory (HEDL) and Los Alamos Scientific Laboratory (LASL) has produced a preliminary design for a Fusion Materials Irradiation Test Facility (FMIT) that uses a high-power linear accelerator to fire a deuteron beam into a high-speed jet of molten lithium. The result is a continuous energy spectrum of neutrons with a 14-MeV average energy which can irradiate material samples to projected end-of-life levels in about 3 years, with a total accumulated fluence of 10 21 to 10 22 n/cm 2

  6. Lewis Research Center space station electric power system test facilities

    Science.gov (United States)

    Birchenough, Arthur G.; Martin, Donald F.

    1988-01-01

    NASA Lewis Research Center facilities were developed to support testing of the Space Station Electric Power System. The capabilities and plans for these facilities are described. The three facilities which are required in the Phase C/D testing, the Power Systems Facility, the Space Power Facility, and the EPS Simulation Lab, are described in detail. The responsibilities of NASA Lewis and outside groups in conducting tests are also discussed.

  7. Testing of Local Velocity Transducer Used at Sodium Thermal Hydraulic Test Facilities

    International Nuclear Information System (INIS)

    Kim, Tae Joon; Eoh, Jae Hyuk; Hwang, In Koo; Jeong, Ji Young; Kim, Jong Man; Lee, Yong Bum; Kim, Yeong Il

    2012-01-01

    KAERI (Korea Atomic Energy Research Institute) will perform a test for a thermal hydraulic simulation with STELLA-1 for a Component Performance Test Sodium Loop in the year 2012, and subsequently it will construct for STELLA-2 for a Sodium Thermalhydraulic Experimental Facility in the year 2016. The STELLA-2 consists of a scaled reactor vessel with a core of electric heaters, four IHXs, two PHTS pumps, two DHXs, and two AHXs. In STELLA-2, several kinds of flow measurements exists. In this paper, the local velocity transducer as a prototype tested in IPPE (in Russia), was manufactured as a prototype by a shop in KAERI. This local velocity transducer will be used to measure the flow rate in a pool

  8. Experience with the instrumentation tests in large sodium test facilities

    International Nuclear Information System (INIS)

    Lauhoff, Th.; Ruppert, E.; Stehle, H.; Vinzens, K.

    1976-01-01

    A facility is described for fast breeder core components (AKB) to test specially instrumented fuel dummies and blanket elements, and also absorber elements under simulated normal and extreme reactor conditions. In addition to endurance testing of a special sodium and high temperature sub-assembly, instrumentation is provided to investigate thermohydraulic and vibrational behaviour of core elements. During tests of > 3000 h at temperatures above 820 K the main sub-assembly characteristics, e.g. pressure drop, leakage flow, vibration and noise spectra can be reproduced. The use of eddy current flow meters, strain gauges, magnetostrictive noise sensors, pressure transducers, thermocouples, and acoustic surveillance devices, are described. (U.K.)

  9. Diagnostics of high-speed liquid lithium jet for IFMIF/EVEDA lithium test loop

    International Nuclear Information System (INIS)

    Kanemura, Takuji; Kondo, Hiroo; Furukawa, Tomohiro; Sugiura, Hirokazu; Horiike, Hiroshi; Yamaoka, Nobuo; Ida, Mizuho; Nakamura, Kazuyuki; Matsushita, Izuru

    2011-01-01

    Regarding R and Ds on the International Fusion Materials Irradiation Facility (IFMIF), hydraulic stability of the liquid Li jet simulating the IFMIF Li target is planned to be validated using EVEDA Li Test Loop (ELTL). IFMIF is an accelerator-based deuteron-lithium (Li) neutron source for research and development of fusion reactor materials. The stable Li target is required in IFMIF to maintain the quality of the neutron fluence and integrity of the Li target itself. This paper presents diagnostics of the Li jet to be implemented in validation tests of the jet stability in ELTL, and those specifications and methodologies are introduced. In the tests, the following physical parameters need to be measured; thickness of the jet; surface structure (height, length/width and frequency of free-surface waves); local flow velocity at the free surface; and Li evaporation rate. With regard to measurement of jet thickness and the surface wave height, a contact-type liquid level sensor is to be used. As for measurement of wave velocity and visual understanding of detailed free-surface structure, a high-speed video camera is to be leveraged. With respect to Li evaporation measurement, weight change of specimens installed near the free surface and frequency change of a crystal quartz are utilized. (author)

  10. Consolidated Incineration Facility metals partitioning test

    International Nuclear Information System (INIS)

    Burns, D.B.

    1993-01-01

    Test burns were conducted at Energy and Environmental Research Corporation's rotary kiln simulator, the Solid Waste Incineration Test Facility, using surrogate CIF wastes spiked with hazardous metals and organics. The primary objective for this test program was measuring heavy metals partition between the kiln bottom ash, scrubber blowdown solution, and incinerator stack gas. Also, these secondary waste streams were characterized to determine waste treatment requirements prior to final disposal. These tests were designed to investigate the effect of several parameters on metals partitioning: incineration temperature; waste chloride concentration; waste form (solid or liquid); and chloride concentration in the scrubber water. Tests were conducted at three kiln operating temperatures. Three waste simulants were burned, two solid waste mixtures (paper, plastic, latex, and one with and one without PVC), and a liquid waste mixture (containing benzene and chlorobenzene). Toxic organic and metal compounds were spiked into the simulated wastes to evaluate their fate under various combustion conditions. Kiln offgases were sampled for volatile organic compounds (VOC), semi-volatile organic compounds (SVOC), polychlorinated dibenz[p]dioxins and polychlorinated dibenzofurans (PCDD/PCDF), metals, particulate loading and size distribution, HCl, and combustion products. Stack gas sampling was performed to determine additional treatment requirements prior to final waste disposal. Significant test results are summarized below

  11. Results from Carbon Dioxide Washout Testing Using a Suited Manikin Test Apparatus with a Space Suit Ventilation Test Loop

    Science.gov (United States)

    Chullen, Cinda; Conger, Bruce; McMillin, Summer; Vonau, Walt; Kanne, Bryan; Korona, Adam; Swickrath, Mike

    2016-01-01

    NASA is developing an advanced portable life support system (PLSS) to meet the needs of a new NASA advanced space suit. The PLSS is one of the most critical aspects of the space suit providing the necessary oxygen, ventilation, and thermal protection for an astronaut performing a spacewalk. The ventilation subsystem in the PLSS must provide sufficient carbon dioxide (CO2) removal and ensure that the CO2 is washed away from the oronasal region of the astronaut. CO2 washout is a term used to describe the mechanism by which CO2 levels are controlled within the helmet to limit the concentration of CO2 inhaled by the astronaut. Accumulation of CO2 in the helmet or throughout the ventilation loop could cause the suited astronaut to experience hypercapnia (excessive carbon dioxide in the blood). A suited manikin test apparatus (SMTA) integrated with a space suit ventilation test loop was designed, developed, and assembled at NASA in order to experimentally validate adequate CO2 removal throughout the PLSS ventilation subsystem and to quantify CO2 washout performance under various conditions. The test results from this integrated system will be used to validate analytical models and augment human testing. This paper presents the system integration of the PLSS ventilation test loop with the SMTA including the newly developed regenerative Rapid Cycle Amine component used for CO2 removal and tidal breathing capability to emulate the human. The testing and analytical results of the integrated system are presented along with future work.

  12. Conceptual Design of Forced Convection Molten Salt Heat Transfer Testing Loop

    Energy Technology Data Exchange (ETDEWEB)

    Manohar S. Sohal; Piyush Sabharwall; Pattrick Calderoni; Alan K. Wertsching; S. Brandon Grover

    2010-09-01

    This report develops a proposal to design and construct a forced convection test loop. A detailed test plan will then be conducted to obtain data on heat transfer, thermodynamic, and corrosion characteristics of the molten salts and fluid-solid interaction. In particular, this report outlines an experimental research and development test plan. The most important initial requirement for heat transfer test of molten salt systems is the establishment of reference coolant materials to use in the experiments. An earlier report produced within the same project highlighted how thermophysical properties of the materials that directly impact the heat transfer behavior are strongly correlated to the composition and impurities concentration of the melt. It is therefore essential to establish laboratory techniques that can measure the melt composition, and to develop purification methods that would allow the production of large quantities of coolant with the desired purity. A companion report describes the options available to reach such objectives. In particular, that report outlines an experimental research and development test plan that would include following steps: •Molten Salts: The candidate molten salts for investigation will be selected. •Materials of Construction: Materials of construction for the test loop, heat exchangers, and fluid-solid corrosion tests in the test loop will also be selected. •Scaling Analysis: Scaling analysis to design the test loop will be performed. •Test Plan: A comprehensive test plan to include all the tests that are being planned in the short and long term time frame will be developed. •Design the Test Loop: The forced convection test loop will be designed including extensive mechanical design, instrument selection, data acquisition system, safety requirements, and related precautionary measures. •Fabricate the Test Loop. •Perform the Tests. •Uncertainty Analysis: As a part of the data collection, uncertainty analysis will

  13. Pressure and Temperature of the Room 1 for the Pipe Break Accidents of the 3-Pin Fuel Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    Park, S. K.; Chi, D. Y.; Sim, B. S.; Park, K. N.; Ahn, S. H.; Lee, J. M.; Lee, C. Y.; Kim, H. R

    2005-08-15

    This report deals with the prediction of the pressure and temperature of the room 1 for the pipe break accidents of the 3-pin fuel test loop. The 3-pin fuel test loop is an experimental facility for nuclear fuel tests at the operation conditions similar to those of PWR and CANDU power plants. Because the most processing systems of the 3-pin fuel test loop are placed in the room 1. The structural integrity of the room 1 should be evaluated for the postulated accident conditions. Therefore the pressures and temperatures of the room 1 needed for the structural integrity evaluation have been calculated by using MARS code. The pressures and temperatures of the room 1 have been calculated in various conditions such as the thermal hydraulic operation parameters, the locations of pipe break, and the thermal properties of the room 1 wall. It is assumed that the pipe break accident occurs in the letdown operation without regeneration, because the mass and energy release to the room 1 is expected to be the largest. As a result of the calculations the maximum pressure and temperature are predicted to be 208kPa and 369.2K(96.0 .deg. C) in case the heat transfer is considered in the room 1 wall. However the pressure and temperature are asymptotically 243kPa and 378.1K(104.9 .deg. C) assuming that the heat transfer does not occur in the room 1 wall.

  14. Mirror fusion test facility plasma diagnostics system

    International Nuclear Information System (INIS)

    Thomas, S.R. Jr.; Coffield, F.E.; Davis, G.E.; Felker, B.

    1979-01-01

    During the past 25 years, experiments with several magnetic mirror machines were performed as part of the Magnetic Fusion Energy (MFE) Program at LLL. The latest MFE experiment, the Mirror Fusion Test Facility (MFTF), builds on the advances of earlier machines in initiating, stabilizing, heating, and sustaining plasmas formed with deuterium. The goals of this machine are to increase ion and electron temperatures and show a corresponding increase in containment time, to test theoretical scaling laws of plasma instabilities with increased physical dimensions, and to sustain high-beta plasmas for times that are long compared to the energy containment time. This paper describes the diagnostic system being developed to characterize these plasma parameters

  15. Sensitivity analysis on the interfacial drag in SPACE code to simulate UPTF separate effect test about loop seal clearance phenomenon

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sukho; Lim, Sanggyu; You, Gukjong; Park, Youngsheop [Korea Hydro and Nuclear Power Company, Ltd., Daejeon (Korea, Republic of)

    2014-05-15

    The nuclear thermal hydraulic system code known as SPACE (Safety and Performance Analysis CodE) was developed and its V and V (Verification and Validation) have been conducted using well-known SETs (Separate Effect Tests) and IETs (Integral Effect Tests). At the same time, the SBLOCA (Small Break Loss of Coolant Accident) methodology in accordance with Appendix K of 10CFR50 for the APR1400 (Advanced Power Reactor 1400) was developed and applied to regulatory body for licensing in 2013. Especially, the SBLOCA methodology developed using SPACE v2.14 code adopts inherent test matrix independent of V and V test to show its conservatism for important phenomena. In this paper, the predictability of SPACE code for UPTF (Upper Plenum Test Facility) test simulating loop seal clearance of SBLOCA important phenomena and the related sensitivity analysis are introduced.

  16. An integral effect test facility of the SMART, SMART ITL

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hyun Sik; Moon, Sang Ki; Kim, Yeon Sik; Cho, Seok; Choi, Ki Yong; Bae, Hwang; Kim, Dong Eok; Choi, Nam Hyun; Min, Kyoung Ho; Ko, Yung Joo; Shin, Yong Cheol; Park, Rae Joon; Lee, Won Jae; Song, Chul Hwa; Yi, Sung Jae [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    SMART (System integrated Modular Advanced ReacTor) is a 330 MWth integral pressurized water reactor (iPWR) developed by KAERI and had obtained standard design approval (SDA) from Korean regulatory authority on July 2012. In this SMART design main components including a pressurizer, reactor coolant pumps and steam generators are installed in a reactor pressure vessel without any large connecting pipes. As the LBLOCA scenario is inherently excluded, its safety systems could be simplified only to ensure the safety during the SBLOCA scenarios and the other system transients. An integral effect test loop for the SMART (SMART ITL), or called as FESTA, had been designed to simulate the integral thermal hydraulic behavior of the SMART. The objectives of the SMART ITL are to investigate and understand the integral performance of reactor systems and components and the thermal hydraulic phenomena occurred in the system during normal, abnormal and emergency conditions, and to verify the system safety during various design basis events of the SMART. The integral effect test data will also be used to validate the related thermal hydraulic models of the safety analysis code such as TASS/SMR S, which is used for performance and accident analysis of the SMART design. This paper introduces the scaling analysis and scientific design of the integral test facility of the SMART, SMART ITL and its scaling analysis results.

  17. Thermal performance test of hot gas ducts of helium engineering demonstration loop (HENDEL)

    International Nuclear Information System (INIS)

    Hishida, Makoto; Kunitomi, Kazuhiko; Ioka, Ikuo; Umenishi, Koji; Kondo, Yasuo; Tanaka, Toshiyuki; Shimomura, Hiroaki

    1984-01-01

    A hot gas duct provided with internal thermal insulation is supposed to be used for an experimental very high-temperature gas-cooled reactor (VHTR) which has been developed by the Japan Atomic Energy Research Institute (JAERI). This type of hot gas duct has not been used so far in industrial facilities, and only a couple of tests on such a large-scale model of hot gas duct have been conducted. The present test was to investigate the thermal performance of the hot gas ducts which are installed as parts of a helium engineering demonstration loop (HENDEL) of JAERI. Uniform temperature and heat flux distributions at the surface of the duct were observed, the experimental correlation being obtained for the effective thermal conductivity of the internal thermal insulation layer. The measured temperature distribution of the pressure tube was in good agreement with the calculation by a TRUMP heat transfer computer code. The temperature distribution of the inner tube of VHTR hot gas duct was evaluated, and no hot spot was detected. These results would be very valuable for the design and development of VHTR. (author)

  18. 40 CFR 160.31 - Testing facility management.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 23 2010-07-01 2010-07-01 false Testing facility management. 160.31... GOOD LABORATORY PRACTICE STANDARDS Organization and Personnel § 160.31 Testing facility management. For each study, testing facility management shall: (a) Designate a study director as described in § 160.33...

  19. 40 CFR 792.31 - Testing facility management.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 31 2010-07-01 2010-07-01 true Testing facility management. 792.31... facility management. For each study, testing facility management shall: (a) Designate a study director as... appropriately tested for identity, strength, purity, stability, and uniformity, as applicable. (e) Assure that...

  20. In-pile experiments and test facilities proposed for fast reactor safety

    International Nuclear Information System (INIS)

    Grolmes, M.A.; Avery, R.; Goldman, A.J.; Fauske, H.K.; Marchaterre, J.F.; Rose, D.; Wright, A.E.

    1976-01-01

    The role of in-pile experiments in support of the resolution of fast breeder reactor safety and licensing issues has been re-examined, with emphasis on key safety issues. Experiment needs have been related to the specific characteristics of these safety issues and to realistic requirements for additional test facility capabilities which can be achieved and utilized within the next ten years. It is found that those safety issues related to the energetics of core disruptive accidents have the largest impact on new facility requirements. However, utilization of existing facilities with modifications can provide for a continuing increase in experiment capability and experiment results on a timely bases. Emphasis has been placed upon maximum utilization of existing facilities and minimum requirements for new facilities. This evaluation has concluded that a new Safety Test Facility, STF, along with major modifications to the EBR II facility, improvement in TREAT capabilities, the existing Sodium Loop Safety Facility and corresponding Support Facilities provide the essential elements of the Safety Research Experiment Facilities (SAREF) required for resolution of key issues

  1. A probabilistic safety assessment of in-pile test loop in HWRR

    International Nuclear Information System (INIS)

    Cao Xuewu; Li Zhaohuan

    1991-07-01

    The PSA methodology has been applied to the in-pile test loop which is installed in the Heavy Water Research Reactor (HWRR). This loop is designed and operated for fuel assembly testing of the Qinshan PWR plant. This analysis is to assess the safety and to evaluate the design of this operating loop. The procedure and models are similar to a PSA on nuclear power plant. The major contents in the analysis consist of the familiarization of the object, the investigation and selection of accident initiators, setting events and fault trees, data collections, quantitative calculations, qualitative and result analyses and final conclusion. This analysis is only limited to the initiators of in-pile loop itself and possible errors made by operators during normal operation. The accident occurence is less than 10 -4 a -1 which may be recommended as an acceptance risk for safety operation of an in-pile test loop. Finally, suggestions have been raised to improve the design of test loop, especially in reducing operation errors by local operators

  2. Test results of reliable and very high capillary multi-evaporators / condenser loop

    Energy Technology Data Exchange (ETDEWEB)

    Van Oost, S; Dubois, M; Bekaert, G [Societe Anonyme Belge de Construction Aeronautique - SABCA (Belgium)

    1997-12-31

    The paper present the results of various SABCA activities in the field of two-phase heat transport system. These results have been based on a critical review and analysis of the existing two-phase loop and of the future loop needs in space applications. The research and the development of a high capillary wick (capillary pressure up to 38 000 Pa) are described. These activities have led towards the development of a reliable high performance capillary loop concept (HPCPL), which is discussed in details. Several loop configurations mono/multi-evaporators have been ground tested. The presented results of various tests clearly show the viability of this concept for future applications. Proposed flight demonstrations as well as potential applications conclude this paper. (authors) 7 refs.

  3. Test results of reliable and very high capillary multi-evaporators / condenser loop

    Energy Technology Data Exchange (ETDEWEB)

    Van Oost, S.; Dubois, M.; Bekaert, G. [Societe Anonyme Belge de Construction Aeronautique - SABCA (Belgium)

    1996-12-31

    The paper present the results of various SABCA activities in the field of two-phase heat transport system. These results have been based on a critical review and analysis of the existing two-phase loop and of the future loop needs in space applications. The research and the development of a high capillary wick (capillary pressure up to 38 000 Pa) are described. These activities have led towards the development of a reliable high performance capillary loop concept (HPCPL), which is discussed in details. Several loop configurations mono/multi-evaporators have been ground tested. The presented results of various tests clearly show the viability of this concept for future applications. Proposed flight demonstrations as well as potential applications conclude this paper. (authors) 7 refs.

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  5. FY 1993 progress report on the ANS thermal-hydraulic test loop operation and results

    Energy Technology Data Exchange (ETDEWEB)

    Siman-Tov, M.; Felde, D.K.; Farquharson, G. [and others

    1994-07-01

    The Thermal-Hydraulic Test Loop (THTL) is an experimental facility constructed to support the development of the Advanced Neutron Source Reactor (ANSR) at Oak Ridge National Laboratory (ORNL). Highly subcooled heavy-water coolant flows vertically upward at a very high mass flux of almost 27 MG/m{sup 2}-s. In a parallel fuel plate configuration as in the ANSR, the flow is subject to a potential excursive static-flow instability that can very rapidly lead to flow starvation and departure from nucleate boiling (DNB) in the ``hot channel``. The current correlations and experimental data bases for flow excursion (FE) and critical heat flux (CHF) seldom evaluate the specific combination of ANSR operating parameters. The THTL facility was designed and built to provide known thermal-hydraulic (T/H) conditions for a simulated full-length coolant subchannel of the ANS reactor core, thus facilitating experimental determination of FE and CHF thermal limits under expected ANSR T/H conditions. A series of FE tests with water flowing vertically upward was completed over a nominal heat flux range of 6 to 17 MW/m{sup 2}, a mass flux range of 8 to 28 Mg/m{sup 2}-s, an exit pressure range of 1.4 to 2.1 MPa, and an inlet temperature range of 40 to 50 C. FE experiments were also conducted using as ``soft`` a system as possible to secure a true FE phenomena (actual secondary burnout). True DNB experiments under similar conditions were also conducted. To the author`s knowledge, no other FE data have been reported in the literature to date that dover such a combination of conditions of high mass flux, high heat flux, and moderately high pressure.

  6. FY 1993 progress report on the ANS thermal-hydraulic test loop operation and results

    International Nuclear Information System (INIS)

    Siman-Tov, M.; Felde, D.K.; Farquharson, G.

    1994-07-01

    The Thermal-Hydraulic Test Loop (THTL) is an experimental facility constructed to support the development of the Advanced Neutron Source Reactor (ANSR) at Oak Ridge National Laboratory (ORNL). Highly subcooled heavy-water coolant flows vertically upward at a very high mass flux of almost 27 MG/m 2 -s. In a parallel fuel plate configuration as in the ANSR, the flow is subject to a potential excursive static-flow instability that can very rapidly lead to flow starvation and departure from nucleate boiling (DNB) in the ''hot channel''. The current correlations and experimental data bases for flow excursion (FE) and critical heat flux (CHF) seldom evaluate the specific combination of ANSR operating parameters. The THTL facility was designed and built to provide known thermal-hydraulic (T/H) conditions for a simulated full-length coolant subchannel of the ANS reactor core, thus facilitating experimental determination of FE and CHF thermal limits under expected ANSR T/H conditions. A series of FE tests with water flowing vertically upward was completed over a nominal heat flux range of 6 to 17 MW/m 2 , a mass flux range of 8 to 28 Mg/m 2 -s, an exit pressure range of 1.4 to 2.1 MPa, and an inlet temperature range of 40 to 50 C. FE experiments were also conducted using as ''soft'' a system as possible to secure a true FE phenomena (actual secondary burnout). True DNB experiments under similar conditions were also conducted. To the author's knowledge, no other FE data have been reported in the literature to date that dover such a combination of conditions of high mass flux, high heat flux, and moderately high pressure

  7. Fast Flux Test Facility core system

    International Nuclear Information System (INIS)

    Ethridge, J.L.; Baker, R.B.; Leggett, R.D.; Pitner, A.L.; Waltar, A.E.

    1990-11-01

    A review of Liquid Metal Reactor (LMR) core system accomplishments provides an excellent road map through the maze of issues that faced reactor designers 10 years ago. At that time relatively large uncertainties were associated with fuel pin and fuel assembly performance, irradiation of structural materials, and performance of absorber assemblies. The extensive core systems irradiation program at the US Department of Energy's Fast Flux Test Facility (FFTF) has addressed each of these principal issues. As a result of the progress made, the attention of long-range LMR planners and designers can shift away from improving core systems and focus on reducing capital costs to ensure the LMR can compete economically in the 21st century with other nuclear reactor concepts. 3 refs., 6 figs., 1 tab

  8. The CERN linear collider test facility (CTF)

    International Nuclear Information System (INIS)

    Baconnier, Y.; Battisti, S.; Bossart, R.; Delahaye, J.P.; Geissler, K.K.; Godot, J.C.; Huebner, K.; Madsen, J.H.B.; Potier, J.P.; Riche, A.J.; Sladen, J.; Suberlucq, G.; Wilson, I.; Wuensch, W.

    1992-01-01

    The CTF (Collider Test Facility) was brought into service last year. The 3 GHz gun produced a beam of 3 MeV/c which was accelerated to 40 MeV/c. This beam, passing a prototype CLIC (linear collider) structure, generated a sizeable amount of 30 GHz power. This paper describes the results and experience with the gun driven by a 8 ns long laser pulse and its CsI photo cathode, the beam behaviour, the beam diagnostics in particular with the bunch measurements by Cerenkov or transition radiation light and streak camera, the photo cathode research, and the beam dynamics studies on space charge effects. (Author)4 figs., tab., 6 refs

  9. Design, Evaluation and Test Technology Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The mission of this facility, which is composed of numerous specialized facilities, is to provide capabilities to simulate a wide range of environments for component...

  10. Plan for 3-D full-scale earthquake testing facility

    International Nuclear Information System (INIS)

    Ohtani, K.

    2001-01-01

    Based on the lessons learnt from the Great Hanshin-Awaji Earthquake, National Research Institute for Earth Science and Disaster Prevention plan to construct the 3-D Full-Scale Earthquake Testing Facility. This will be the world's largest and strongest shaking table facility. This paper describes the outline of the project for this facility. This facility will be completed in early 2005. (author)

  11. Reflux condensation behavior in SBLOCA tests of ATLAS facility

    International Nuclear Information System (INIS)

    Kim, Yeon-Sik; Park, Hyun-Sik; Cho, Seok; Choi, Ki-Yong; Kang, Kyoung-Ho

    2017-01-01

    Highlights: • Behavior of a reflux condensation heat transfer was investigated for SBLOCA tests. • Behavior of the reflux condensate in HL, SG inlet plenum, and U-tubes were evaluated. • Concept of a steam moisturizing phenomenon was introduced and discussed. • Test data and MARS calculations were compared and discussed on the reflux condensate. - Abstract: The behavior of the reflux condensation heat transfer in a hot side steam generator (SG) U-tubes during a cold leg (CL) pipe and a direct vessel injection (DVI) line break in small break loss-of-coolant accident (SBLOCA) tests of the ATLAS facility was investigated including MARS code calculations. Among the SBLOCA tests, a 6″-CL pipe and 50%-DVI line break SBLOCA test were selected to investigate the behavior of the reflux condensation. A reflux condensation heat transfer seemed to occur from the time the SG U-tubes were half-empty to near the loop seal clearing (LSC). It was found that a transition regime existed between the reflux condensation heat transfer and reverse heat transfer. The remaining reflux condensate in SG U-tubes owing to the counter-current flow limit (CCFL) phenomenon and a separating effect of liquid carry-over and/or entrainment with steam moisturizing seemed to affect the thermal-hydraulic behavior of the transition regime. It was also found that the steam flowrate of the loop pipings and SG U-tubes seemed to have a strong effect on the duration time of the transition regime, e.g., a larger steam flowrate results in a longer duration. From a comparison of the reflux condensation behavior between the ATLAS tests and MARS code calculations, overall qualitative agreements were found between the two cases. The largest discrepancies were found in the SG inlet plenum water level between the two cases, and the authors suggest that the combination effects of the remaining reflux condensate in SG U-tubes and a separating effect of liquid carry-over and/or entrainment with steam

  12. Technical and economic feasibility study for the reactivation of the integral test facility of IPEN/CNEN Nuclear Engineering Center

    Energy Technology Data Exchange (ETDEWEB)

    Biaty, Flávia P.; Rocha, Marcelo da S.; Oliveira, Otávio L. de, E-mail: flavia.biaty@usp.br, E-mail: msrocha@ipen.br, E-mail: otavioluis@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil)

    2017-07-01

    The Integral Test Facility of Nuclear Engineering Center (CEN/IPEN/CNEN-SP), known as 'Loop 70', is a semi-industrial thermal-hydraulic test facility and can operate as a BWR (Boiling Water Reactor) or a PWR (Pressurizing Water Reactor) mode. Designed and built in the 1980's, it is currently disabled. The experimental circuits ('test loop') are facilities that reproduce the thermohydraulic and fluid dynamic conditions that occur inside a reactor and are used to simulate the practical reality which it is not possible to be obtained through mathematical models. In this context, this research project aims the development of a Business Plan to analyze the technical and economic feasibility related to the reactivation of the facility. This methodology (adapted to the government sector) is a decision-making tool that will offer a wide perspective of the project, set the guidelines and actions that will define the future of the facility and provide a general rule to make investments on it. This paper presents the historic aspects to better understand the Loop 70's current situation. It also presents information about similar facilities around the world, services that can be offered (thermal-hydraulics parameters measurements, equipment qualification and transient analysis due accident situations), results of the strategic analysis (SWOT) performed, specific goals for each critical success or failure factor of the facility, financial aspects related to the reactivation and an overview of the facility's perspectives. (author)

  13. Technical and economic feasibility study for the reactivation of the integral test facility of IPEN/CNEN Nuclear Engineering Center

    International Nuclear Information System (INIS)

    Biaty, Flávia P.; Rocha, Marcelo da S.; Oliveira, Otávio L. de

    2017-01-01

    The Integral Test Facility of Nuclear Engineering Center (CEN/IPEN/CNEN-SP), known as 'Loop 70', is a semi-industrial thermal-hydraulic test facility and can operate as a BWR (Boiling Water Reactor) or a PWR (Pressurizing Water Reactor) mode. Designed and built in the 1980's, it is currently disabled. The experimental circuits ('test loop') are facilities that reproduce the thermohydraulic and fluid dynamic conditions that occur inside a reactor and are used to simulate the practical reality which it is not possible to be obtained through mathematical models. In this context, this research project aims the development of a Business Plan to analyze the technical and economic feasibility related to the reactivation of the facility. This methodology (adapted to the government sector) is a decision-making tool that will offer a wide perspective of the project, set the guidelines and actions that will define the future of the facility and provide a general rule to make investments on it. This paper presents the historic aspects to better understand the Loop 70's current situation. It also presents information about similar facilities around the world, services that can be offered (thermal-hydraulics parameters measurements, equipment qualification and transient analysis due accident situations), results of the strategic analysis (SWOT) performed, specific goals for each critical success or failure factor of the facility, financial aspects related to the reactivation and an overview of the facility's perspectives. (author)

  14. NRI experimental facility for the testing of irradiation assisted stress corrosion cracking

    International Nuclear Information System (INIS)

    Ruscak, M.; Chvatal, P.; Zamboch, M.

    1998-01-01

    IASCC influencing reactor internals of both BWR and PWR reactors is a complex phenomenon covering influences of material structure, neutron fluence, neutron flux, chemistry of environment, gamma radiation and mechanical stress. To evaluate such degradation, tests should be performed under conditions similar to those in real structure. Nuclear Research Institute has built several experimental facilities in order to be able to test IASCC degradation of materials. Basically, reactor water loops, both PWR and BWR, could be used to model environmental conditions including gamma and neutron irradiation. Pre-irradiation can be done in irradiation channels under well controlled temperature conditions. During the experiment, in-pile conditions can be compared with those out of pile. It enables to clarify pure influence of irradiation. For testing of irradiated specimens, hot cell facility has been developed for slow strain rate tests. The paper will show all above mentioned facilities as well as some of the results observed with them. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    1979-10-01

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

  16. Integration of Hardware-in-the-loop Facilities Over the Internet

    Science.gov (United States)

    2009-04-15

    This briefing discusses a hardware in loop vehicle simulator in Warren, Michigan that provides the driver with realistic power response from the Power and Energy Systems Integration Lab over the internet.

  17. Summary of TRUEX Radiolysis Testing Using the INL Radiolysis Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    Dean R. Peterman; Lonnie G. Olson; Rocklan G. McDowell; Gracy Elias; Jack D. Law

    2012-03-01

    The INL radiolysis and hydrolysis test loop has been used to evaluate the effects of hydrolytic and radiolytic degradation upon the efficacy of the TRUEX flowsheet for the recovery of trivalent actinides and lanthanides from acidic solution. Repeated irradiation and subsequent re-conditioning cycles did result in a significant decrease in the concentration of the TBP and CMPO extractants in the TRUEX solvent and a corresponding decrease in americium and europium extraction distributions. However, the build-up of solvent degradation products upon {gamma}-irradiation, had little impact upon the efficiency of the stripping section of the TRUEX flowsheet. Operation of the TRUEX flowsheet would require careful monitoring to ensure extraction distributions are maintained at acceptable levels.

  18. Simulation of the fusion materials irradiation test facility lithium and heat transport systems for abnormal events study

    International Nuclear Information System (INIS)

    Carlson, W.F.; Elyashar, N.N.

    1981-01-01

    A digital computer model of Fusion Materials Irradiation Test Facility's heat transport system has been developed. The model utilizes a set of coupled differential equations to simulate the dynamic behavior of the primary and secondary heat transport loop systems. The model has been used to investigate the stability of the proposed control schemes for lithium temperature and flow rate and for an extensive study of equipment failures and malfunction analysis. It was determined that certain equipment failures and malfunctions in the primary loop require a response from the control system within less than one second of the occurrence of the failure. The effects of equipment failures in the secondary loop were found to be less dramatic than the equivalent failures in the primary loop. The failures in the secondary loop generally required control action in time frames of the order of minutes

  19. EFFLUENT TREATMENT FACILITY PEROXIDE DESTRUCTION CATALYST TESTING

    International Nuclear Information System (INIS)

    HALGREN DL

    2008-01-01

    The 200 Area Effluent Treatment Facility (ETF) main treatment train includes the peroxide destruction module (PDM) where the hydrogen peroxide residual from the upstream ultraviolet light/hydrogen peroxide oxidation unit is destroyed. Removal of the residual peroxide is necessary to protect downstream membranes from the strong oxidizer. The main component of the PDM is two reaction vessels utilizing granular activated carbon (GAC) as the reaction media. The PDM experienced a number of operability problems, including frequent plugging, and has not been utilized since the ETF changed to groundwater as the predominant feed. The unit seemed to be underperforming in regards to peroxide removal during the early periods of operation as well. It is anticipated that a functional PDM will be required for wastewater from the vitrification plant and other future streams. An alternate media or methodology needs to be identified to replace the GAC in the PDMs. This series of bench scale tests is to develop information to support an engineering study on the options for replacement of the existing GAC method for peroxide destruction at the ETF. A number of different catalysts will be compared as well as other potential methods such as strong reducing agents. The testing should lead to general conclusions on the viability of different catalysts and identify candidates for further study and evaluation

  20. Corium melt researches at VESTA test facility

    Directory of Open Access Journals (Sweden)

    Hwan Yeol Kim

    2017-10-01

    Full Text Available VESTA (Verification of Ex-vessel corium STAbilization and VESTA-S (-small test facilities were constructed at the Korea Atomic Energy Research Institute in 2010 to perform various corium melt experiments. Since then, several tests have been performed for the verification of an ex-vessel core catcher design for the EU-APR1400. Ablation tests of an impinging ZrO2 melt jet on a sacrificial material were performed to investigate the ablation characteristics. ZrO2 melt in an amount of 65–70 kg was discharged onto a sacrificial material through a well-designed nozzle, after which the ablation depths were measured. Interaction tests between the metallic melt and sacrificial material were performed to investigate the interaction kinetics of the sacrificial material. Two types of melt were used: one is a metallic corium melt with Fe 46%, U 31%, Zr 16%, and Cr 7% (maximum possible content of U and Zr for C-40, and the other is a stainless steel (SUS304 melt. Metallic melt in an amount of 1.5–2.0 kg was delivered onto the sacrificial material, and the ablation depths were measured. Penetration tube failure tests were performed for an APR1400 equipped with 61 in-core instrumentation penetration nozzles and extended tubes at the reactor lower vessel. ZrO2 melt was generated in a melting crucible and delivered down into an interaction crucible where the test specimen is installed. To evaluate the tube ejection mechanism, temperature distributions of the reactor bottom head and in-core instrumentation penetration were measured by a series of thermocouples embedded along the specimen. In addition, lower vessel failure tests for the Fukushima Daiichi nuclear power plant are being performed. As a first step, the configuration of the molten core in the plant was investigated by a melting and solidification experiment. Approximately 5 kg of a mixture, whose composition in terms of weight is UO2 60%, Zr 10%, ZrO2 15%, SUS304 14%, and B4C 1%, was melted in a

  1. Validity and Utilization of the Out-Pile Testing Facilities at HANARO

    International Nuclear Information System (INIS)

    Choo, Kee-Nam; Cho, Man-Soon; Yang, Sung-Woo; Shin, Yoon-Taek; Park, Seng-Jae; Jun, Byung-Hyuk; Kim, Myong-Seop

    2016-01-01

    Various neutron irradiation facilities such as rabbit irradiation facilities, loop facilities and the capsule irradiation facilities for irradiation tests of nuclear materials, fuels and radioisotope products have been developed at HANARO. Among these irradiation facilities, the capsule is the most useful device for coping with the various test requirements at HANARO. To support the national research and development programs on nuclear reactors and the nuclear fuel cycle technology in Korea, new irradiation capsules have been developed and actively utilized for the irradiation tests requested by numerous users. The environmental conditions for these reactors are generally beyond present day reactor technology, especially regarding the higher neutron fluence and higher operating temperature. To effectively support the national R and Ds relevant to the future nuclear systems, the development of advanced irradiation technologies concerning higher neutron fluence and irradiation temperature are being preferentially developed at HANARO. The utilization of the out-pile testing facilities to satisfy the criteria of safety evaluation for a new device installed in the core of HANARO was summarized. In addition, the validity of the out-pile testing facilities was evaluated and proved to be effective for verifying the integrity of irradiation capsule

  2. Validity and Utilization of the Out-Pile Testing Facilities at HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Choo, Kee-Nam; Cho, Man-Soon; Yang, Sung-Woo; Shin, Yoon-Taek; Park, Seng-Jae; Jun, Byung-Hyuk; Kim, Myong-Seop [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Various neutron irradiation facilities such as rabbit irradiation facilities, loop facilities and the capsule irradiation facilities for irradiation tests of nuclear materials, fuels and radioisotope products have been developed at HANARO. Among these irradiation facilities, the capsule is the most useful device for coping with the various test requirements at HANARO. To support the national research and development programs on nuclear reactors and the nuclear fuel cycle technology in Korea, new irradiation capsules have been developed and actively utilized for the irradiation tests requested by numerous users. The environmental conditions for these reactors are generally beyond present day reactor technology, especially regarding the higher neutron fluence and higher operating temperature. To effectively support the national R and Ds relevant to the future nuclear systems, the development of advanced irradiation technologies concerning higher neutron fluence and irradiation temperature are being preferentially developed at HANARO. The utilization of the out-pile testing facilities to satisfy the criteria of safety evaluation for a new device installed in the core of HANARO was summarized. In addition, the validity of the out-pile testing facilities was evaluated and proved to be effective for verifying the integrity of irradiation capsule.

  3. Tritium Systems Test Facility. Volume II. Appendixes

    International Nuclear Information System (INIS)

    Anderson, G.W.; Battleson, K.W.; Bauer, W.

    1976-10-01

    This document includes the following appendices: (1) vacuum pumping, (2) tritium migration into the power cycle, (3) separation of hydrogen isotopes, (4) tritium research laboratory, (5) TSTF containment and cleanup, (6) instrumentation and control, (7) gas heating in torus, and (8) TSTF fuel loop operating procedures

  4. Construction and testing of the Mirror Fusion Test Facility magnets

    International Nuclear Information System (INIS)

    Kozman, T.; Shimer, D.; VanSant, J.; Zbasnik, J.

    1986-08-01

    This paper describes the construction and testing of the Mirror Fusion Test Facility superconducting magnet set. Construction of the first Yin Yang magnet was started in 1978. And although this particular magnet was later modified, the final construction of these magnets was not completed until 1985. When completed these 42 magnets weighed over 1200 tonnes and had a maximum stored energy of approximately 1200 MJ at full field. Together with power supplies, controls and liquid nitrogen radiation shields the cost of the fabrication of this system was over $100M. General Dynamics/Convair Division was responsible for the system design and the fabrication of 20 of the magnets. This contract was the largest single procurement action at the Lawrence Livermore National Laboratory. During the PACE acceptance tests, the 26 major magnets were operated at full field for more than 24 hours while other MFTF subsystems were tested. From all of the data, the magnets operated to the performance specifications. For physics operation in the future, additional helium and nitrogen leak checking and repair will be necessary. In this report we will discuss the operation and testing of the MFTF Magnet System, the world's largest superconducting magnet set built to date. The topics covered include a schedule of the major events, summary of the fabrication work, summary of the installation work, summary of testing and test results, and lessons learned

  5. Modeling a forced to natural convection boiling test with the program LOOP-W

    International Nuclear Information System (INIS)

    Carbajo, J.J.

    1984-01-01

    Extensive testing has been conducted in the Simulant Boiling Flow Visualization (SBFV) loop in which water is boiled in a vertical transparent tube by circulating hot glycerine in an annulus surrounding the tube. Tests ranged from nonboiling forced convection to oscillatory boiling natural convection. The program LOOP-W has been developed to analyze these tests. This program is a multi-leg, one-dimensional, two-phase equilibrium model with slip between the phases. In this study, a specific test, performed at low power where non-boiling forced convection was changed to boiling natural convection and then to non-boiling again, has been modeled with the program LOOP-W

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

    Science.gov (United States)

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

    2006-01-01

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

  7. Status of the ELISE test facility

    International Nuclear Information System (INIS)

    Franzen, P.; Wünderlich, D.; Riedl, R.; Nocentini, R.; Fantz, U.; Fröschle, M.; Heinemann, B.; Martens, C.; Kraus, W.; Ruf, B.; Bonomo, F.; Pimazzoni, A.

    2015-01-01

    The test facility ELISE, equipped with a large radio frequency (RF) driven ion source (1×0.9 m2) of half the size of the ion source for the ITER neutral beam injection (NBI) system, is operational since beginning of 2013. The first experimental campaign was dedicated to a thorough qualification of the test facility and its diagnostic tools at low RF power (80 kW in total, i.e. 20 kW per driver) in volume operation, i.e. operation without cesium, where the negative hydrogen ion production is done in the plasma volume only. This paper reports on the main results of the second and third experimental campaigns, where Cs was inserted in the ion source for an enhancement of the negative ion production by the surface process. The second experimental campaign was done still with low RF power, both for hydrogen and deuterium, with pulse lengths of up to 500 s. The results of this campaign are rather encouraging, especially in hydrogen, where large current densities with respect to the low RF power could be achieved at a ratio of co-extracted electrons to extracted ions of 0.5-0.6 at the relevant source pressure of 0.3 Pa. Similar large extracted ion currents could be achieved also in deuterium, but with larger amounts of co-extracted electrons. The required ratio of co-extracted electrons to extracted ions of one could be achieved only in short pulses. The third experimental campaign aimed then for approaching the required ITER NBI parameters with respect to the ion and electron extracted currents, both for hydrogen and deuterium, by increasing the RF power with short pulses, i.e. beam-on times of up to 10 s and RF-on time up to 20 s. Current densities near the ITER NBI requirements could be achieved in hydrogen at a ratio of co-extracted electrons to extracted ions of 0.5-0.6 at the relevant source pressure of 0.3 Pa. As it was the case for the low RF operation, the required filter field was significantly lower than expected from the experience with the small

  8. Gingin High Optical Power Test Facility

    International Nuclear Information System (INIS)

    Zhao, C; Blair, D G; Barrigo, P

    2006-01-01

    The Australian Consortium for Gravitational Wave Astronomy (ACIGA) in collaboration with LIGO is developing a high optical power research facility at the AIGO site, Gingin, Western Australia. Research at the facility will provide solutions to the problems that advanced gravitational wave detectors will encounter with extremely high optical power. The problems include thermal lensing and parametric instabilities. This article will present the status of the facility and the plan for the future experiments

  9. Portland, Oregon Test Data Set Arterial Loop Detector Data

    Data.gov (United States)

    Department of Transportation — This set of data files was acquired under USDOT FHWA cooperative agreement DTFH61-11-H-00025 as one of the four test data sets acquired by the USDOT Data Capture and...

  10. Portland, Oregon Test Data Set Freeway Loop Detector Data

    Data.gov (United States)

    Department of Transportation — This set of data files was acquired under USDOT FHWA cooperative agreement DTFH61-11-H-00025 as one of the four test data sets acquired by the USDOT Data Capture and...

  11. Electronics and Telemetry Engineering and Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Electronics Laboratory is a fully equipped facility providing the capability to support electronic product development from highly complex weapon system sensors,...

  12. Validity and Reliability of Orthodontic Loops between Mechanical Testing and Computer Simulation: An Finite Element Method Study

    Directory of Open Access Journals (Sweden)

    Gaurav Sepolia

    2014-01-01

    Full Text Available The magnitude and direction of orthodontic force is one of the essential concerns of orthodontic tooth movements. Excessive force may cause root resorption and mobility of the tooth, whereas low force level may results in prolonged treatment. The addition of loops allows the clinician to more accurately achieve the desired results. Aims and objectives: The purpose of the study was to evaluate the validity and reliability of orthodontic loops between mechanical testing and computer simulation. Materials and methods: Different types of loops were taken and divided into four groups: The Teardrop loop, Opus loop, L loop and T loop. These were artificially activated for multiple lengths and studied using the FEM. Results: The Teardrop loop showed the highest force level, and there is no significant difference between mechanical testing and computer simulation.

  13. ORNL instrumentation performance for Slab Core Test Facility (SCTF)-Core I Reflood Test Facility

    International Nuclear Information System (INIS)

    Hardy, J.E.; Hess, R.A.; Hylton, J.O.

    1983-11-01

    Instrumentation was developed for making measurements in experimental refill-reflood test facilities. These unique instrumentation systems were designed to survive the severe environmental conditions that exist during a simulated pressurized water reactor loss-of-coolant accident (LOCA). Measurement of in-vessel fluid phenomena such as two-phase flow velocity and void fraction and film thickness and film velocity are required for better understanding of reactor behavior during LOCAs. The Advanced Instrumentation for Reflood Studies (AIRS) Program fabricated and delivered instrumentation systems and data reduction software algorithms that allowed the above measurements to be made. Data produced by AIRS sensors during three experimental runs in the Japanese Slab Core Test Facility are presented. Although many of the sensors failed before any useful data could be obtained, the remaining probes gave encouraging and useful results. These results are the first of their kind produced during simulated refill-reflood stage of a LOCA near actual thermohydrodynamic conditions

  14. New facility for testing LHC HTS power leads

    CERN Document Server

    Rabehl, Roger Jon; Fehér, S; Huang, Y; Orris, D; Pischalnikov, Y; Sylvester, C D; Tartaglia, M

    2005-01-01

    A new facility for testing HTS power leads at the Fermilab Magnet Test Facility has been designed and operated. The facility has successfully tested 19 pairs of HTS power leads, which are to be integrated into the Large Hadron Collider Interaction Region cryogenic feed boxes. This paper describes the design and operation of the cryogenics, process controls, data acquisition, and quench management systems. HTS power lead test results from the commissioning phase of the project are also presented.

  15. ORNL 150 keV neutral beam test facility

    International Nuclear Information System (INIS)

    Gardner, W.L.; Kim, J.; Menon, M.M.; Schilling, G.

    1977-01-01

    The 150 keV neutral beam test facility provides for the testing and development of neutral beam injectors and beam systems of the class that will be needed for the Tokamak Fusion Test Reactor (TFTR) and The Next Step (TNS). The test facility can simulate a complete beam line injection system and can provide a wide range of experimental operating conditions. Herein is offered a general description of the facility's capabilities and a discussion of present system performance

  16. FBR structural material test facility in flowing sodium environment

    International Nuclear Information System (INIS)

    Shanmugasundaram, M.; Kumar, Hemant; Ravi, S.

    2016-01-01

    In Fast Breeder Reactor (FBR), components such as Control and Safety Rod Drive Mechanism (CSRDM), Diverse Safety Rod Drive Mechanism (DSRDM), Transfer arm and primary sodium pumps etc., are experiencing friction and wear between the moving parts in contact with liquid sodium at high temperature. Hence, it is essential to evaluate the friction and wear behaviour to validate the design of components. In addition, the above core structural reactor components such as core cover plate, control plugs etc., undergoes thermal striping which is random thermal cycling induced by flow stream resulting from the mixing of non isothermal jets near that component. This leads to development of surface cracks and assist in crack growth which in turn may lead to failure of the structural component. Further, high temperature components are often subjected to low cycle fatigue due to temperature gradient induced cyclic thermal stresses caused by start-ups, shutdowns and transients. Also steady state operation at elevated temperature introduces creep and the combination of creep and fatigue leads to creep-fatigue interactions. Therefore, resistance to low cycle fatigue, creep and creep-fatigue are important considerations in the design of FBR components. Liquid sodium is used as coolant and hence the study of the above properties in dynamic sodium are equally important. In view of the above, facility for materials testing in sodium (INSOT) has been constructed and in operation for conducting the experiments such as tribology, thermal stripping, low cycle fatigue, creep and creep-fatigue interaction etc. The salient features of the operation and maintenance of creep and fatigue loops of INSOT facility are discussed in detail. (author)

  17. Passive safety testing at the Fast Flux Test Facility

    International Nuclear Information System (INIS)

    Lucoff, D.M.

    1989-01-01

    During 1986, the Fast Flux Test Facility (FFTF) conducted several tests designed to improve the understanding of the passive safety characteristics of an oxide-fueled liquid-metal reactor (LMR). Static and dynamic tests were performed over a broad range of power, flow, and temperature conditions that extended beyond those for normal operation. Key results of these tests are presented. Stable operation at low power with natural circulation cooling was demonstrated. A passive safety enhancement feature, the gas expansion module (GEM) was developed specifically to offset the large amount of cooldown reactivity that needs to be controlled in an oxide-fueled LMR undergoing an unprotected loss-of-flow accident. Nine GEMs were built and successfully tested in FFTF. With the reactor at 50% power (200 MW (thermal)), the main coolant pumps were turned off and the normal control rod scram response was inhibited. The GEMs and inherent core reactivity feedback mechanisms took the core subcritical with a modest peak coolant temperature transient that reached 85 degrees C above the pretransient value and always maintained a >400 degrees C margin to the sodium boiling point (910 degrees C)

  18. Design of a fusion engineering test facility

    International Nuclear Information System (INIS)

    Sager, P.H.

    1980-01-01

    The fusion Engineering Test Facility (ETF) is being designed to provide for engineering testing capability in a program leading to the demonstration of fusion as a viable energy option. It will combine power-reactor-type components and subsystems into an integrated tokamak system and provide a test bed to test blanket modules in a fusion environment. Because of the uncertainties in impurity control two basic designs are being developed: a design with a bundle divertor (Design 1) and one with a poloidal divertor (Design 2). The two designs are similar where possible, the latter having somewhat larger toroidal field (TF) coils to accommodate removal of the larger torus sectors required for the single-null poloidal divertor. Both designs have a major radius of 5.4 m, a minor radius of 1.3 m, and a D-shaped plasma with an elongation of 1.6. Ten TF coils are incorporated in both designs, producing a toroidal field of 5.5 T on-axis. The ohmic heating and equilibrium field (EF) coils supply sufficient volt-seconds to produce a flat-top burn of 100 s and a duty cycle of 135 s, including a start of 12 s, a burn termination of 10 s, and a pumpdown of 13 s. The total fusion power during burn is 750 MW, giving a neutron wall loading of 1.5 MW/m 2 . In Design 1 of the poloidal field (PF) coils except the fast-response EF coils are located outside the FT coils and are superconducting. The fast-response coils are located inside the TF coil bore near the torus and are normal conducting so that they can be easily replaced.In Design 2 all of the PF coils are located outside the TF coils and are superconducting. Ignition is achieved with 60 MW of neutral beam injection at 150 keV. Five megawatts of radio frequency heating (electron cyclotron resonance heating) is used to assist in the startup and limit the breakdown requirement to 25 V

  19. Simulation of atmosphere stratification in the HDR test facility with the CONTAIN code

    International Nuclear Information System (INIS)

    Skerlavaj, A.; Mavko, B.; Kljenak, I.

    2001-01-01

    The test E11.2 'Hydrogen distribution in loop flow geometry', which was performed in the Heissdampf Reaktor containment test facility in Germany, was simulated with the CONTAIN computer code. The predicted pressure history and thermal stratification are in relatively good agreement with the measurements. The compositional stratification within the containment was qualitatively well predicted, although the degree of the stratification in the dome area was slightly underestimated. The analysis of simulation results enabled a better understanding of the physical phenomena during the test.(author)

  20. SRF cavity testing using a FPGA Self Excited Loop

    Energy Technology Data Exchange (ETDEWEB)

    Ben-Zvi, I. [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2017-08-30

    Various authors have previously studied the theory and practice of cavity testing, notably an extensive treatment by Powers [1] and Padamsee [2]. The advent of the digital Low Level RF (LLRF) electronics based on Field Programmable Logic Arrays (FPGA) provides various improvements over the rather complex systems used in the past as well as enabling new measurement techniques.In this document we reintroduce a technique that seems to have fallen out of practice in recent times, that is obtaining the coupling constant β through measurements from just one port, the reflected power port, of the directional coupler placed in front of the cavity.

  1. Thermal hydraulic studies in steam generator test facility

    International Nuclear Information System (INIS)

    Vinod, V.; Suresh Kumar, V.A.; Noushad, I.B.; Ellappan, T.R.; Rajan, K.K.; Rajan, M.; Vaidyanathan, G.

    2005-01-01

    Full text of publication follows: A 500 MWe fast breeder reactor is being constructed at Kalpakkam, India. This is a sodium cooled reactor with two primary and two secondary sodium loops with total 8 steam generators. The typical advantage of fast breeder plants is the high operating temperature of steam cycles and the high plant efficiency. To produce this high pressure and high temperature steam, once through straight tube vertical sodium heated steam generators are used. The steam is generated from the heat produced in the reactor core and being transported through primary and secondary sodium circuits. The steam generator is a 25 m high middle supported steam generator with expansion bend and 23 m heat transfer length. Steam Generator Test Facility (SGTF) constructed at Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam aims at performing various tests on a 5.5 MWt steam generator. This vertically simulated test article is similar in all respects to the proposed 157 MWt steam generator module for the Prototype Fast Breeder Reactor (PFBR), with reduced number of tubes. Heat transfer performance tests are done with this 19 tube steam generator at various load conditions. Sodium circuit for the SGTF is equipped with oil fired heater as heat source and centrifugal sodium pump, to pump sodium at 105 m 3 /hr flow rate. Other typical components like sodium to air heat exchanger, sodium purification system and hydrogen leak detection system is also present in the sodium circuit. High pressure steam produced in the steam generator is dumped in a condenser and recycled. Important tests planned in SGTF are the heat transfer performance test, stability test, endurance test and performance test of steam generator under various transients. The controlled operation of steam generator will be studied with possible control schemes. A steady state simulation of the steam generator is done with a mathematical model. This paper gives the details of heat transfer

  2. Establishment and operation of a photovoltaic cell test facility

    Energy Technology Data Exchange (ETDEWEB)

    Pearsall, N.M.; Forbes, I.

    1999-07-01

    This report describes the setting up of a test facility at the University of Northumbria. Details of the equipment specification and procurement are given, and the commissioning and initial operation of the facility, and the measurement procedures for I-V characteristics, spectral response measurements, optical scanning and test charges are outlined. The business plan for the test facility is discussed, and operating experience is reviewed in terms of publicity, services provided, and collaboration.

  3. Argonne to open new facility for advanced vehicle testing

    CERN Multimedia

    2002-01-01

    Argonne National Laboratory will open it's Advanced Powertrain Research Facility on Friday, Nov. 15. The facility is North America's only public testing facility for engines, fuel cells, electric drives and energy storage. State-of-the-art performance and emissions measurement equipment is available to support model development and technology validation (1 page).

  4. 10 CFR 26.125 - Licensee testing facility personnel.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 1 2010-01-01 2010-01-01 false Licensee testing facility personnel. 26.125 Section 26.125 Energy NUCLEAR REGULATORY COMMISSION FITNESS FOR DUTY PROGRAMS Licensee Testing Facilities § 26.125... reports, if any; results of tests that establish employee competency for the position he or she holds...

  5. 21 CFR 58.31 - Testing facility management.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 1 2010-04-01 2010-04-01 false Testing facility management. 58.31 Section 58.31... management. For each nonclinical laboratory study, testing facility management shall: (a) Designate a study... appropriately tested for identity, strength, purity, stability, and uniformity, as applicable. (e) Assure that...

  6. Upgrade of the Cryogenic CERN RF Test Facility

    CERN Document Server

    Pirotte, O; Brunner, O; Inglese, V; Koettig, T; Maesen, P; Vullierme, B

    2014-01-01

    With the large number of superconducting radiofrequency (RF) cryomodules to be tested for the former LEP and the present LHC accelerator a RF test facility was erected early in the 1990’s in the largest cryogenic test facility at CERN located at Point 18. This facility consisted of four vertical test stands for single cavities and originally one and then two horizontal test benches for RF cryomodules operating at 4.5 K in saturated helium. CERN is presently working on the upgrade of its accelerator infrastructure, which requires new superconducting cavities operating below 2 K in saturated superfluid helium. Consequently, the RF test facility has been renewed in order to allow efficient cavity and cryomodule tests in superfluid helium and to improve its thermal performances. The new RF test facility is described and its performances are presented.

  7. Upgrade of the cryogenic CERN RF test facility

    International Nuclear Information System (INIS)

    Pirotte, O.; Benda, V.; Brunner, O.; Inglese, V.; Maesen, P.; Vullierme, B.; Koettig, T.

    2014-01-01

    With the large number of superconducting radiofrequency (RF) cryomodules to be tested for the former LEP and the present LHC accelerator a RF test facility was erected early in the 1990’s in the largest cryogenic test facility at CERN located at Point 18. This facility consisted of four vertical test stands for single cavities and originally one and then two horizontal test benches for RF cryomodules operating at 4.5 K in saturated helium. CERN is presently working on the upgrade of its accelerator infrastructure, which requires new superconducting cavities operating below 2 K in saturated superfluid helium. Consequently, the RF test facility has been renewed in order to allow efficient cavity and cryomodule tests in superfluid helium and to improve its thermal performances. The new RF test facility is described and its performances are presented

  8. Hydrogen Infrastructure Testing and Research Facility Video (Text Version)

    Science.gov (United States)

    grid integration, continuous code improvement, fuel cell vehicle operation, and renewable hydrogen Systems Integration Facility or ESIF. Research projects including H2FIRST, component testing, hydrogen

  9. Analysis on working pressure selection of ACME integral test facility

    International Nuclear Information System (INIS)

    Chen Lian; Chang Huajian; Li Yuquan; Ye Zishen; Qin Benke

    2011-01-01

    An integral effects test facility, advanced core cooling mechanism experiment facility (ACME) was designed to verify the performance of the passive safety system and validate its safety analysis codes of a pressurized water reactor power plant. Three test facilities for AP1000 design were introduced and review was given. The problems resulted from the different working pressures of its test facilities were analyzed. Then a detailed description was presented on the working pressure selection of ACME facility as well as its characteristics. And the approach of establishing desired testing initial condition was discussed. The selected 9.3 MPa working pressure covered almost all important passive safety system enables the ACME to simulate the LOCAs with the same pressure and property similitude as the prototype. It's expected that the ACME design would be an advanced core cooling integral test facility design. (authors)

  10. DOE LeRC photovoltaic systems test facility

    Science.gov (United States)

    Cull, R. C.; Forestieri, A. F.

    1978-01-01

    The facility was designed and built and is being operated as a national facility to serve the needs of the entire DOE National Photovoltaic Program. The object of the facility is to provide a place where photovoltaic systems may be assembled and electrically configured, without specific physical configuration, for operation and testing to evaluate their performance and characteristics. The facility as a breadboard system allows investigation of operational characteristics and checkout of components, subsystems and systems before they are mounted in field experiments or demonstrations. The facility as currently configured consist of 10 kW of solar arrays built from modules, two inverter test stations, a battery storage system, interface with local load and the utility grid, and instrumentation and control necessary to make a flexible operating facility. Expansion to 30 kW is planned for 1978. Test results and operating experience are summaried to show the variety of work that can be done with this facility.

  11. Chemical decontamination for decommissioning purposes. (Vigorous decontamination tests of steel samples in a special test loop)

    International Nuclear Information System (INIS)

    Bregani, F.; Pascali, R.; Rizzi, R.

    1984-01-01

    The aim of the research activities described was to develop vigorous decontamination techniques for decommissioning purposes, taking into account the cost of treatment of the radwaste, to achieve possibly unrestricted release of the treated components, and to obtain know-how for in situ hard decontamination. The decontamination procedures for strong decontamination have been optimized in static and dynamic tests (DECO-loop). The best values have been found for: (i) hydrochloric acid: 4 to 5% vol. at low temperature, 0.7 to 1% vol. at high temperature (80 0 C); (ii) hydrofluoric plus nitric acid: 1.5% vol. HF + 5% vol. HNO 3 at low temperature; 0.3 to 0.5% vol. HF + 2.5 to 5% vol. HNO 3 at high temperature. High flow rates are not necessary, but a good re-circulation of the solution is needed. The final contamination levels, after total oxide removal, are in accordance with limits indicated for unrestricted release of materials in some countries. The arising of the secondary waste is estimated. Decontamination of a 10 m 2 surface would typically produce 0.5 to 3.0 kg of dry waste, corresponding to 1.6 to 10 kg of concrete conditioned waste

  12. Initial liquid metal magnetohydrodynamic thin film flow experiments in the MeGA-loop facility at UCLA

    International Nuclear Information System (INIS)

    Morley, N.B.; Gaizer, A.A.; Tillack, M.S.; Abdou, M.A.

    1995-01-01

    Free surface thin film flows of liquid metal were investigated experimentally in the presence of a coplanar magnetic field. This investigation was performed in a new magnetohydrodynamic (MHD) flow facility, the MeGA-loop, utilizing a low melting temperature lead-bismuth alloy as the working metal. Owing to the relatively low magnetic field produced by the present field coil system, the ordinary hydrodynamic and low MHD interaction regimes only were investigated. At the high flow speeds necessary for self cooling, the importance of a well designed and constructed channel becomes obvious. Partial MHD drag, increasing the film height, is observed as Haβ 2 becomes greater than unity. MHD laminarization of the turbulent film flows is observed when Haβ/Re>0.002, but fully laminar flow was not reached. Suggestions for facility upgrades to achieve greater MHD interaction are presented in the context of these initial results. (orig.)

  13. Optimal closed-loop identification test design for internal model control

    NARCIS (Netherlands)

    Zhu, Y.; Bosch, van den P.P.J.

    2000-01-01

    In this work, optimal closed-loop test design for control is studied. Simple design formulas are derived based on the asymptotic theory of Ljung. The control scheme used is internal model control (IMC) and the design constraint is the power of the process output or that of the reference signal. The

  14. Improvements in flight table dynamic transparency for hardware-in-the-loop facilities

    Science.gov (United States)

    DeMore, Louis A.; Mackin, Rob; Swamp, Michael; Rusterholtz, Roger

    2000-07-01

    Flight tables are a 'necessary evil' in the Hardware-In-The- Loop (HWIL) simulation. Adding the actual or prototypic flight hardware to the loop, in order to increase the realism of the simulation, forces us to add motion simulation to the process. Flight table motion bases bring unwanted dynamics, non- linearities, transport delays, etc to an already difficult problem sometimes requiring the simulation engineer to compromise the results. We desire that the flight tables be 'dynamically transparent' to the simulation scenario. This paper presents a State Variable Feedback (SVF) control system architecture with feed-forward techniques that improves the flight table's dynamic transparency by significantly reducing the table's low frequency phase lag. We offer some actual results with existing flight tables that demonstrate the improved transparency. These results come from a demonstration conducted on a flight table in the KHILS laboratory at Eglin AFB and during a refurbishment of a flight table for the Boeing Company of St. Charles, Missouri.

  15. TRAC-PF1 code verification with data from the OTIS test facility

    International Nuclear Information System (INIS)

    Childerson, M.T.; Fujita, R.K.

    1985-01-01

    A computer code (TRAC-PF1/MOD1) developed for predicting transient thermal and hydraulic integral nuclear steam supply system (NSSS) response was benchmarked. Post-small break loss-of-coolant accident (LOCA) data from a scaled, experimental facility, designated the One-Through Integral System (OTIS), were obtained for the Babcock and Wilcox NSSS and compared to TRAC predictions. The OTIS tests provided a challenging small break LOCA data set for TRAC verification. The major phases of a small break LOCA observed in the OTIS tests included pressurizer draining and loop saturation, intermittent reactor coolant system circulation, boiler-condenser mode, and the initial stages of refill. The TRAC code was successful in predicting OTIS loop conditions (system pressures and temperatures) after modification of the steam generator model. In particular, the code predicted both pool and auxiliary-feedwater initiated boiler-condenser mode heat transfer

  16. Testing FlexRay ECUs with a hardware-in-the-loop simulator; Test von FlexRay-Steuergeraeten am Hardware-in-the-Loop Simulator

    Energy Technology Data Exchange (ETDEWEB)

    Stroop, J.; Koehl, S. [dSPACE GmbH, Paderborn (Germany); Peller, M.; Riedesser, P. [BMW AG, Muenchen (Germany)

    2005-07-01

    To master the data communication of complex and safety relevant systems within future vehicles, the BMW Group prepares the application of FlexRay. The accompanying development process plays an important role for the quality, stability and reliability of those systems. Hardware-in-the-loop simulation and test stands are indispensable constituents and they are an integral part of the validation process. The following contribution describes the technology that is used within the BMW Group in more detail, especially in terms of communication networks with FlexRay. (orig.)

  17. Facility of Laboratories for Sustainable Habitation - an Initial Design of a Closed-Loop Environment

    OpenAIRE

    Quantius, Dominik; Schubert, Daniel; Maiwald, Volker; Parìs Lopéz, Rosa; Hauslage, Jens; Seboldt, Wolfgang; Doule, Ondrej; Schlacht, Irene Lia; Ransom, Stephen

    2012-01-01

    An effective and self-sustainable artificial habitat design is essential for human spaceflight and expansion of mankind into orbit or towards other celestial bodies. Besides the necessity to create an artificial habitat for the extreme environments of space, development of a self-sustainable habitat can also enable more effective exploration of extreme environments on Earth. One major application of the habitat’s closed-loop capabilities can also be in enabling ecological habitation of human ...

  18. Test Requirements and Conceptual Design for a Potassium Test Loop to Support an Advanced Potassium Rankine Cycle Power Conversion Systems

    Energy Technology Data Exchange (ETDEWEB)

    Yoder, JR.G.L.

    2006-03-08

    Parameters for continuing the design and specification of an experimental potassium test loop are identified in this report. Design and construction of a potassium test loop is part of the Phase II effort of the project ''Technology Development Program for an Advanced Potassium Rankine Power Conversion System''. This program is supported by the National Aeronautics and Space Administration. Design features for the potassium test loop and its instrumentation system, specific test articles, and engineered barriers for ensuring worker safety and protection of the environment are described along with safety and environmental protection requirements to be used during the design process. Information presented in the first portion of this report formed the basis to initiate the design phase of the program; however, the report is a living document that can be changed as necessary during the design process, reflecting modifications as additional design details are developed. Some portions of the report have parameters identified as ''to be determined'' (TBD), reflecting the early stage of the overall process. In cases where specific design values are presently unknown, the report attempts to document the quantities that remain to be defined in order to complete the design of the potassium test loop and supporting equipment.

  19. Measurement of Li target thickness in the EVEDA Li Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    Kanemura, Takuji, E-mail: kanemura.takuji@jaea.go.jp [Japan Atomic Energy Agency, 4002 Narita, O-arai, Higashi-Ibaraki-gun, Ibaraki 311-1393 (Japan); Kondo, Hiroo; Furukawa, Tomohiro; Hirakawa, Yasushi [Japan Atomic Energy Agency, 4002 Narita, O-arai, Higashi-Ibaraki-gun, Ibaraki 311-1393 (Japan); Hoashi, Eiji; Yoshihashi, Sachiko; Horiike, Hiroshi [Osaka University, 2-1 Yamada-oka, Suita, Osaka 565-0871 (Japan); Wakai, Eiichi [Japan Atomic Energy Agency, 4002 Narita, O-arai, Higashi-Ibaraki-gun, Ibaraki 311-1393 (Japan)

    2015-10-15

    Highlights: • The objective is to validate stability of the IFMIF liquid Li target flowing at 15 m/s. • Design requirement of target thickness fluctuation is ±1 mm. • Mean and maximum wave amplitude are 0.26 and 1.46 mm, respectively. • Average thickness can be well predicted with developed analytical model. • Li target was adequately stable and satisfied design requirement. - Abstract: A high-speed (nominal: 15 m/s, range: 10–16 m/s) liquid lithium wall jet is planned to serve as the target for two 40 MeV and 125 mA deuteron beams in the International Fusion Materials Irradiation Facility (IFMIF). The design requirement of target thickness stability is 25 ± 1 mm under a vacuum of 10{sup −3} Pa. This paper presents the results of the target thickness measurement conducted in the EVEDA Li Test Loop under a wide range of conditions including the IFMIF condition (target speed of 10, 15, and 20 m/s; vacuum pressure of 10{sup −3} Pa; and Li temperature of 250 °C). For measurement, we use a laser probe method that we developed in advance; this method generates statistical measurements method using a laser distance meter. The measurement results obtained under the IFMIF nominal condition (15 m/s, 10{sup −3} Pa, 250 °C) at the IFMIF beam center are as follows: average target thickness = 26.08 ± 0.09 mm (2σ), mean wave amplitude = 0.26 ± 0.01 mm (2σ), and maximum wave amplitude = 1.46 ± 0.25 mm (2σ). Of the total wave components, 99.7% are within the design requirement. The analytically predicted target thickness is in excellent agreement with the experimental data, resulting in successful characterization of the Li target thickness.

  20. Measurement of lithium target surface velocity in the IFMIF/EVEDA lithium test loop

    Energy Technology Data Exchange (ETDEWEB)

    Kanemura, Takuji, E-mail: kanemura.takuji@jaea.go.jp [Japan Atomic Energy Agency, 4002 Narita, O-arai, Higashi-Ibaraki-gun, Ibaraki 311-1393 (Japan); Kondo, Hiroo; Furukawa, Tomohiro; Hirakawa, Yasushi [Japan Atomic Energy Agency, 4002 Narita, O-arai, Higashi-Ibaraki-gun, Ibaraki 311-1393 (Japan); Hoashi, Eiji [Osaka University, 2-1 Yamada-oka, Suita, Osaka 565-0871 (Japan); Yoshihashi, Sachiko; Horiike, Hiroshi [Fukui University of Technology, Gakuen 3-6-1, Fukui-shi, Fukui 910-8505 (Japan); Wakai, Eiichi [Japan Atomic Energy Agency, 4002 Narita, O-arai, Higashi-Ibaraki-gun, Ibaraki 311-1393 (Japan)

    2016-11-01

    Highlights: • The objective is to measure the free-surface velocity field of the IFMIF Li target. • The Li target has an important role to remove 10 MW heat input from a deuteron beam. • The free-surface of the Li target is under the most severe heat load condition. • Measured surface velocities are almost equal to cross-sectional average velocities. • It was confirmed that the IFMIF Li target has adequate heat removal performance. - Abstract: In the framework of the Engineering Validation and Engineering Design Activities (EVEDA) project of the International Fusion Materials Irradiation Facility (IFMIF), we measured surface velocity fields of a lithium (Li) target at the EVEDA Li test loop under specifically-designated IFMIF conditions (target speeds of 10, 15, and 20 m/s, vacuum pressure of 10{sup −3} Pa, and Li temperature of 250 °C). In the current design of the IFMIF, the free surface of the Li target is under a most severe heat load condition with respect to Li boiling. The objective of this study is to measure the actual free-surface velocity under these IFMIF conditions to evaluate the heat removal performance of the Li target. The measured results (using the surface-wave tracking method that our team developed) showed two-dimensional time-averaged velocity distributions around the IFMIF beam footprint being virtually uniform, and close to the cross-sectional average velocity. The uniformity of the velocity distributions was less than 1 m/s. The comparison between the measured and analyzed surface velocity at the beam center showed that the analysis accurately predicts the measurement results within a margin of 3%. Finally, it was confirmed that the Li target delivers adequate heat removal performance in the IFMIF as designed.

  1. Preliminary Design of the AEGIS Test Facility

    CERN Document Server

    Dassa, Luca; Cambiaghi, Danilo

    2010-01-01

    The AEGIS experiment is expected to be installed at the CERN Antiproton Decelerator in a very close future, since the main goal of the AEGIS experiment is the measurement of gravity impact on antihydrogen, which will be produced on the purpose. Antihydrogen production implies very challenging environmental conditions: at the heart of the AEGIS facility 50 mK temperature, 1e-12 mbar pressure and a 1 T magnetic field are required. Interfacing extreme cryogenics with ultra high vacuum will affect very strongly the design of the whole facility, requiring a very careful mechanical design. This paper presents an overview of the actual design of the AEGIS experimental facility, paying special care to mechanical aspects. Each subsystem of the facility – ranging from the positron source to the recombination region and the measurement region – will be shortly described. The ultra cold region, which is the most critical with respect to the antihydrogen formation, will be dealt in detail. The assembly procedures will...

  2. Team Update on North American Proton Facilities for Radiation Testing

    Science.gov (United States)

    Label, Kenneth A.; Turflinger, Thomas; Haas, Thurman; George, Jeffrey; Moss, Steven; Davis, Scott; Kostic, Andrew; Wie, Brian; Reed, Robert; Guertin, Steven; hide

    2016-01-01

    In the wake of the closure of the Indiana University Cyclotron Facility (IUCF), this presentation provides an overview of the options for North American proton facilities. This includes those in use by the aerospace community as well as new additions from the cancer therapy regime. In addition, proton single event testing background is provided for understanding the criteria needed for these facilities for electronics testing.

  3. Vanadium—lithium in-pile loop for comprehensive tests of vanadium alloys and multipurpose coatings

    Science.gov (United States)

    Lyublinski, I. E.; Evtikhin, V. A.; Ivanov, V. B.; Kazakov, V. A.; Korjavin, V. M.; Markovchev, V. K.; Melder, R. R.; Revyakin, Y. L.; Shpolyanskiy, V. N.

    1996-10-01

    The reliable information on design and material properties of self-cooled Li sbnd Li blanket and liquid metal divertor under neutron radiation conditions can be obtained using the concept of combined technological and material in-pile tests in a vanadium—lithium loop. The method of in-pile loop tests includes studies of vanadium—base alloys resistance, weld resistance under mechanical stress, multipurpose coating formation processes and coatings' resistance under the following conditions: high temperature (600-700°C), lithium velocities up to 10 m/s, lithium with controlled concentration of impurities and technological additions, a neutron load of 0.4-0.5 MW/m 2 and level of irradiation doses up to 5 dpa. The design of such an in-pile loop is considered. The experimental data on corrosion and compatibility with lithium, mechanical properties and welding technology of the vanadium alloys, methods of coatings formation and its radiation tests in lithium environment in the BOR-60 reactor (fast neutron fluence up to 10 26 m -2, irradiation temperature range of 500-523°C) are presented and analyzed as a basis for such loop development.

  4. Techniques and results of examination of fission product release from VVER fuel rods with artificial defects and a burnup of ∼60 MWd/kgU at the MIR loop facility

    International Nuclear Information System (INIS)

    Burukin, A.; Goryachev, A.; Ilyenko, S.; Izhutov, A.; Konyashov, V.; Shishin, V.; Shulimov, V.; Luzanova, L.; Miglo, V.

    2009-01-01

    Complex of equipment and several techniques for examination of radioactive fission product release from defective fuel rods were developed, prepared and tested at the PV-1 loop facility of the MIR reactor. During the first test, which was conducted at the PV-1 loop facility and aimed at testing of developed equipment and techniques, measurement of radioactive fission product release from an experimental re-fabricated fuel rod with a burnup of ∼60 MWd/kgU and an artificial defect was performed under design-basis steady-state operating conditions of the VVER-1000 reactor. PIE of all main parameters of the experimental defective fuel rod did not reveal any state peculiarities which could be caused by the artificial defect, i.e. fuel and cladding characteristics in the defect area did not differ from the initial ones (before testing) as well as their characteristics in areas distant from the defect; they are typical for fuel rods with a similar irradiation history in the VVER NPP. The gap in the experimental fuel rod was bridged due to close contact between fuel and cladding at increased fuel burnup; it can appreciable reduce release of radioactive fission products into the PV-1 primary coolant. This suggestion and quantitative characteristics of effect of gap bridging in a high-burnup fuel rod on radioactive fission product release should be investigated during the next tests performed at the PV-1 loop facility. Values of radioactive fission product release measured during the first test at the PV-1 loop facility in the MIR reactor will be used for development of an empirical engineering model in order to take into account high burnup effects and their impact on fission product release from fuel and defective fuel rods

  5. Conceptual design of a First Wall mock-up experiment in preparation for the qualification of breeding blanket technologies in the Helium Loop Karlsruhe (HELOKA) facility

    Energy Technology Data Exchange (ETDEWEB)

    Zeile, C., E-mail: christian.zeile@kit.edu [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Abou-Sena, A.; Boccaccini, L.V.; Ghidersa, B.E.; Kang, Q.; Kunze, A. [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Lamberti, L. [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Dipartimento Energia, Politecnico di Torino (Italy); Maione, I.A.; Rey, J.; Weth, A. von der [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2016-11-01

    Highlights: • Experiment in preparation for the qualification of Breeding Blanket technologies in HELOKA facility is proposed. • Experimental capabilities, instrumentation of the mock-up and experimental program are presented. • Design and manufacturing of the mock-up is described. • Design of modular attachment system to obtain different stress levels and distributions on the mock-up is discussed. - Abstract: An experimental program based on a First Wall mock-up is presented as preparation for the qualification of breeding blanket mock-ups at high heat flux in the Helium Loop Karlsruhe (HELOKA) facility. Two objectives of the experimental program have been defined: testing of the experimental setup and a first validation of FE models. The design and manufacturing of mock-up representing about 1/3 of the heated zone of an ITER Test Blanket Module (TBM) First Wall is discussed. A modular attachment system concept has been developed for the fixation of the mock-up in order to be able to generate different stress distributions and levels on the plate, which is confirmed by thermo-mechanical analyses. The HELOKA facility is able to provide a TBM relevant helium cooling system and to generate the required surface heat flux by an electron beam gun. An installed IR camera can be used to measure the temperature distribution on the surface.

  6. Cryogenic aspects of the mirror fusion test facility

    International Nuclear Information System (INIS)

    Sterbentz, W.H.; Nelson, R.L.

    1979-01-01

    This paper covers the design and construction of the MFTF cryogenic system and a description of the operating procedures throughout the many functional modes. The coils and the cryopanels for maintaining the high vacuum environment weigh 417,000 kg (920,000 lb) and must be cooled from room temperature to 4.5 k. The cryogenic system for MFTF consists of a closed-loop helium system with a 3000-W helium refrigerator that uses gas-bearing expansion turbines and oil-flooded screw compressors. In addition, liquid helium storage facilities have adequate capacity for standby operation, and a complete helium-purification plant is capable of processing 17 m 3 /min (600 scfm). An open-loop liquid nitrogen system (with provision for later addition of a nitrogen recondenser) provides the required refrigeration for the radiation shields that must be maintained at 85 K

  7. Vehicle Test Facilities at Aberdeen Test Center and Yuma Test Center

    Science.gov (United States)

    2012-02-27

    fiber optic and bi-directional microwave telemetry links that provide high speed data transfer and real-time test control, and secure, single...Patton Hilly Trails is a 4.2 km (2.6 mi) loop course situated on a Basalt Hill Range site which is located on a series of relic beach terraces. The...Mountain and consists of primarily stretches of very cobbled surface interspersed with rock outcrops and bedrock. The course terrain is basalt hills

  8. 200 Area treated effluent disposal facility operational test report

    International Nuclear Information System (INIS)

    Crane, A.F.

    1995-01-01

    This document reports the results of the 200 Area Treated Effluent Disposal Facility (200 Area TEDF) operational testing activities. These completed operational testing activities demonstrated the functional, operational and design requirements of the 200 Area TEDF have been met

  9. STG-ET: DLR electric propulsion test facility

    Directory of Open Access Journals (Sweden)

    Andreas Neumann

    2017-04-01

    Full Text Available DLR operates the High Vacuum Plume Test Facility Göttingen – Electric Thrusters (STG-ET. This electric propulsion test facility has now accumulated several years of EP-thruster testing experience. Special features tailored to electric space propulsion testing like a large vacuum chamber mounted on a low vibration foundation, a beam dump target with low sputtering, and a performant pumping system characterize this facility. The vacuum chamber is 12.2m long and has a diameter of 5m. With respect to accurate thruster testing, the design focus is on accurate thrust measurement, plume diagnostics, and plume interaction with spacecraft components. Electric propulsion thrusters have to run for thousands of hours, and with this the facility is prepared for long-term experiments. This paper gives an overview of the facility, and shows some details of the vacuum chamber, pumping system, diagnostics, and experiences with these components.

  10. Overview of US fast-neutron facilities and testing capabilities

    International Nuclear Information System (INIS)

    Evans, E.A.; Cox, C.M.; Jackson, R.J.

    1982-01-01

    Rather than attempt a cataloging of the various fast neutron facilities developed and used in this country over the last 30 years, this paper will focus on those facilities which have been used to develop, proof test, and explore safety issues of fuels, materials and components for the breeder and fusion program. This survey paper will attempt to relate the evolution of facility capabilities with the evolution of development program which use the facilities. The work horse facilities for the breeder program are EBR-II, FFTF and TREAT. For the fusion program, RTNS-II and FMIT were selected

  11. Hardware-in-the-loop (HIL) Test of Demand as Frequency Controlled Reserve (DFR)

    DEFF Research Database (Denmark)

    Wu, Qiuwei; Zimmermann, K.; Østergaard, Jacob

    2016-01-01

    This paper presents the hardware-in-the-loop (HIL) test of the demand as frequency controlled reserve (DFR). The HIL test refers to a test in which parts of a pure simulation have been replaced by actual physical components. It is used to understand the behavior of a new device or controller....... The DFR has been tested by offline simulations to illustrate the efficacy of this technology. The DFR control logics have been implemented in the SmartBox. The HIL was conducted by having the SmartBox connected to the real time simulations and the performance of the SmartBox was tested with difference...

  12. Stored energy analysis in the scaled-down test facilities

    International Nuclear Information System (INIS)

    Deng, Chengcheng; Chang, Huajian; Qin, Benke; Wu, Qiao

    2016-01-01

    Highlights: • Three methods are developed to evaluate stored energy in the scaled-down test facilities. • The mechanism behind stored energy distortion in the test facilities is revealed. • The application of stored energy analysis is demonstrated for the ACME facility of China. - Abstract: In the scaled-down test facilities that simulate the accident transient process of the prototype nuclear power plant, the stored energy release in the metal structures has an important influence on the accuracy and effectiveness of the experimental data. Three methods of stored energy analysis are developed, and the mechanism behind stored energy distortion in the test facilities is revealed. Moreover, the application of stored energy analysis is demonstrated for the ACME test facility newly built in China. The results show that the similarity requirements of three methods analyzing the stored energy release decrease gradually. The physical mechanism of stored energy release process can be characterized by the dimensionless numbers including Stanton number, Fourier number and Biot number. Under the premise of satisfying the overall similarity of natural circulation, the stored energy release process in the scale-down test facilities cannot maintain exact similarity. The results of the application of stored energy analysis illustrate that both the transient release process and integral total stored energy of the reactor pressure vessel wall of CAP1400 power plant can be well reproduced in the ACME test facility.

  13. Space nuclear thermal propulsion test facilities accommodation at INEL

    International Nuclear Information System (INIS)

    Hill, T.J.; Reed, W.C.; Welland, H.J.

    1993-01-01

    The U.S. Air Force (USAF) has proposed to develop the technology and demonstrate the feasibility of a particle bed reactor (PBR) propulsion system that could be used to power an advanced upper stage rocket engine. The U.S. Department of Energy (DOE) is cooperating with the USAF in that it would host the test facility if the USAF decides to proceed with the technology demonstration. Two DOE locations have been proposed for testing the PBR technology, a new test facility at the Nevada Test Site, or the modification and use of an existing facility at the Idaho National Engineering Laboratory. The preliminary evaluations performed at the INEL to support the PBR technology testing has been completed. Additional evaluations to scope the required changes or upgrade needed to make the proposed USAF PBR test facility meet the requirements for testing Space Exploration Initiative (SEI) nuclear thermal propulsion engines are underway

  14. Space nuclear thermal propulsion test facilities accommodation at INEL

    Science.gov (United States)

    Hill, Thomas J.; Reed, William C.; Welland, Henry J.

    1993-01-01

    The U.S. Air Force (USAF) has proposed to develop the technology and demonstrate the feasibility of a particle bed reactor (PBR) propulsion system that could be used to power an advanced upper stage rocket engine. The U.S. Department of Energy (DOE) is cooperating with the USAF in that it would host the test facility if the USAF decides to proceed with the technology demonstration. Two DOE locations have been proposed for testing the PBR technology, a new test facility at the Nevada Test Site, or the modification and use of an existing facility at the Idaho National Engineering Laboratory. The preliminary evaluations performed at the INEL to support the PBR technology testing has been completed. Additional evaluations to scope the required changes or upgrade needed to make the proposed USAF PBR test facility meet the requirements for testing Space Exploration Initiative (SEI) nuclear thermal propulsion engines are underway.

  15. Experiments on natural circulation of lead-bismuth in the TALL test facility

    International Nuclear Information System (INIS)

    Ma, W.M.; Karbojian, A.; Sehgal, B.R.

    2005-01-01

    Full text of publication follows: Lead-bismuth eutectic (LBE) is a potential candidate coolant for next generation liquid metal reactors due to its favorable properties such as being chemical inert and low melting point, in comparison with sodium and lead considered as coolants in FBRs. Having a high atomic number of LBE allows it be well suited as a spallation target for accelerator-driven systems (ADS) which have been proposed for the transmutation of nuclear waste. Due to its strong buoyancy, the LBE-cooled system should also have significant natural circulation, which is desirable for so-called Generation IV nuclear reactors, which like to employ passive safety and reliability. But so far, very little experimental data have been published on the natural circulation thermal-hydraulics of LBE-cooled systems. Motivated by the increasing interest in LBE-cooled fast reactors and ADS, a test facility called Thermal-hydraulic ADS Lead-bismuth Loop (TALL) was designed and constructed at KTH to investigate the thermalhydraulic characteristics of liquid LBE. The facility consists of a primary loop (LBE loop) and a secondary loop (oil loop). The LBE loop consists of sump tank, core tank, expansion tank, heat exchanger, EM pump, EM flowmeter, electric heaters and instrumentation. The heating of LBE in the core tank and its cooling in the heat exchanger allows natural convection flows as should occur in the prototypic vessel. Recently, our experimental study on natural circulation was performed on the TALL test facility. This paper will present the experimental results and analysis. The facility is of 6.8 m height which is comparable to the full height of the LBE heat exchange circuit in the ANSALDO ADS reactor vessel design, and has been scaled for prototypic (power/volume) ratio to represent the main components. Their LBE volume, flow velocity and heating rates correspond to one tube of the heat exchanger design chosen. During the experiments, the main adjustable

  16. FY11 Facility Assessment Study for Aeronautics Test Program

    Science.gov (United States)

    Loboda, John A.; Sydnor, George H.

    2013-01-01

    This paper presents the approach and results for the Aeronautics Test Program (ATP) FY11 Facility Assessment Project. ATP commissioned assessments in FY07 and FY11 to aid in the understanding of the current condition and reliability of its facilities and their ability to meet current and future (five year horizon) test requirements. The principle output of the assessment was a database of facility unique, prioritized investments projects with budgetary cost estimates. This database was also used to identify trends for the condition of facility systems.

  17. Fusion fuel purification during the Tritium Systems Test Assembly 3-week loop experiment

    International Nuclear Information System (INIS)

    Willms, R.S.

    1989-01-01

    During the time period from April 19, 1989--May 5, 1989, the Tritium Systems Test Assembly (TSTA) at Los Alamos National Laboratory (LANL) conducted its longest continuous integrated loop operation to date. This provided an opportunity to test some hitherto unproven capabilities of the TSTA Fuel Cleanup System (FCU). Previous FCU tests were reported. The purpose of the FCU is to remove impurities from a stream of hydrogen isotopes (Q 2 ) representative of torus exhaust gas. During this run impurities loadings ranging from 60 to 179 sccm of 90% N 2 and 10% CH 4 were fed to the FCU. Each of the two FCU main flow molecular sieve beds (MSB's) were filled to breakthrough three times. The MSB's were regenerated during loop operations. 2 refs., 6 figs., 2 tabs

  18. Project assembling and commissioning of a rewetting test facility

    International Nuclear Information System (INIS)

    Rezende, H.C.

    1985-08-01

    A test facility (ITR - Instalacao de Testes de Remolhamento) has been erected at the Thermal-hydraulics Laboratory of CDTN, dedicated to the investigation of the basic phenomena that can occur during the reflood phase of a Loss of Coolant Accident (LOCA) in a Pressurized Water Reactor (PWR), utilizing tubular and annular test sections. The present work consists in a presentation of the facility design and a report of its commissioning. The mechanical aspects of the facility, its power supply system and its instrumentation are described. The results of the instruments calibration and two operational tests are presented and a comparison is done with calculations perfomed usign a computer code. (Author) [pt

  19. Development of Aerosol Scrubbing Test Loop for Containment Filtered Venting System

    International Nuclear Information System (INIS)

    Lee, Doo Yong; Jung, Woo Young; Lee, Hyun Chul; Lee, Jong Chan; Kim, Gyu Tae

    2016-01-01

    The scrubber tank is filled with scrubbing water with the chemical additives. The droplet separator based on a cyclone is installed above the scrubbing water pool to remove the large droplets that may clog a metal fiber filter installed at the upper section of the scrubber tank. The outlet piping is connected from the scrubber tank to the molecular sieve to chemically remove the gaseous iodine. The aerosol as a particle is physically captured in the scrubbing water pool passing through the scrubbing nozzle as well as the metal fiber filter. The gaseous iodine such as molecular iodine as well as organic iodide is chemically removed in the scrubbing water pool and molecular sieve. The thermal-hydraulic as well as scrubbing performance for the CFVS should be verified with the experiments. The experiment can be divided into the filtration component based experiment and whole system based one. In this paper, the aerosol scrubbing test loop developed to test the thermal-hydraulic and aerosol scrubbing performance of the scrubbing nozzle with the scrubbing water pool is introduced. The aerosol scrubbing test loop has been developed as a part of the Korean CFVS project. In this loop, the filtration components such as the scrubbing nozzle submerged in the scrubbing water pool as well as the cyclone as droplet separator can be tested under the CFVS operating conditions. The aerosol scrubbing performance of the filtration components including pool scrubbing behavior can be tested with the aerosol generation and feeding system and aerosol measurement system.

  20. Analyses of the Anticipated Operational Occurrences for the HANARO Fuel Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    Park, S. K.; Sim, B. S.; Chi, D. Y.; Lee, C. Y.; Ahn, S. H

    2007-12-15

    The analyses of anticipated operational occurrences of the HANARO fuel test loop have been carried out by using the MARS/FTL{sub A} code, which is a modified version of the MARS code. A critical heat flux correlation on the three rods with triangular array was implemented in the MARS/FTL{sub A} code. The correlation was obtained from the critical heat fluxes measured at a test section, which is the same geometry of the in-pile test section of the HANARO fuel test loop. The anticipated operational occurrences of the HANARO fuel test loop are the inadvertent closure of the isolation valves, the over-power transient of the HANARO, the stuck open of the safety valves, and the loss of HANARO class IV power. A minimum DNBR (Departure from Nucleate Boiling Ratio) was predicted in the inadvertent closure of the isolation valves. It is indicated that the minimum DNBR of 1.85 is greater than the design limit DNBR of 1.39. The maximum coolant pressure calculated in the anticipated operational occurrences is also less than the 110 percents of the design pressure.

  1. Analyses of the Anticipated Operational Occurrences for the HANARO Fuel Test Loop

    International Nuclear Information System (INIS)

    Park, S. K.; Sim, B. S.; Chi, D. Y.; Lee, C. Y.; Ahn, S. H.

    2007-12-01

    The analyses of anticipated operational occurrences of the HANARO fuel test loop have been carried out by using the MARS/FTL A code, which is a modified version of the MARS code. A critical heat flux correlation on the three rods with triangular array was implemented in the MARS/FTL A code. The correlation was obtained from the critical heat fluxes measured at a test section, which is the same geometry of the in-pile test section of the HANARO fuel test loop. The anticipated operational occurrences of the HANARO fuel test loop are the inadvertent closure of the isolation valves, the over-power transient of the HANARO, the stuck open of the safety valves, and the loss of HANARO class IV power. A minimum DNBR (Departure from Nucleate Boiling Ratio) was predicted in the inadvertent closure of the isolation valves. It is indicated that the minimum DNBR of 1.85 is greater than the design limit DNBR of 1.39. The maximum coolant pressure calculated in the anticipated operational occurrences is also less than the 110 percents of the design pressure

  2. Development of Aerosol Scrubbing Test Loop for Containment Filtered Venting System

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Doo Yong; Jung, Woo Young; Lee, Hyun Chul; Lee, Jong Chan; Kim, Gyu Tae [FNC Technology, Yongin (Korea, Republic of)

    2016-05-15

    The scrubber tank is filled with scrubbing water with the chemical additives. The droplet separator based on a cyclone is installed above the scrubbing water pool to remove the large droplets that may clog a metal fiber filter installed at the upper section of the scrubber tank. The outlet piping is connected from the scrubber tank to the molecular sieve to chemically remove the gaseous iodine. The aerosol as a particle is physically captured in the scrubbing water pool passing through the scrubbing nozzle as well as the metal fiber filter. The gaseous iodine such as molecular iodine as well as organic iodide is chemically removed in the scrubbing water pool and molecular sieve. The thermal-hydraulic as well as scrubbing performance for the CFVS should be verified with the experiments. The experiment can be divided into the filtration component based experiment and whole system based one. In this paper, the aerosol scrubbing test loop developed to test the thermal-hydraulic and aerosol scrubbing performance of the scrubbing nozzle with the scrubbing water pool is introduced. The aerosol scrubbing test loop has been developed as a part of the Korean CFVS project. In this loop, the filtration components such as the scrubbing nozzle submerged in the scrubbing water pool as well as the cyclone as droplet separator can be tested under the CFVS operating conditions. The aerosol scrubbing performance of the filtration components including pool scrubbing behavior can be tested with the aerosol generation and feeding system and aerosol measurement system.

  3. IR1 flow tube and In-Pile Test Section Pressure drop test for the 3-Pin Fuel Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    Lee, H. H.; Park, K. N.; Chi, D. Y.; Sim, B. S.; Park, S. K.; Lee, J. M.; Lee, C. Y.; Kim, H. N

    2006-02-15

    The in-pile Section (IPS) of 3-pin Fuel Test Loop(FTL) shall be installed in the vertical hole call IR1 of HANARO reactor core. In order to verify the pressure drop and flow rate both the inside region of IPS at the annular region between IPS and IR1 flow tube, a pressure drop was measured by varing the flow rate on both regions. The measured pressure drop in the annular region is 209kpa at 14.9kg/s which meets the limiting condition of operation of 200kpa. The measured pressure drop in side the IPS becomes 260.25kpa which is lower than the designed value of 306.65kpa. As the pressure drop is lower than the design value, it is quite conservative from the safety and operating point of view.

  4. PANDA: A Multipurpose Integral Test Facility for LWR Safety Investigations

    International Nuclear Information System (INIS)

    Paladino, D.; Dreier, J.

    2012-01-01

    The PANDA facility is a large scale, multicompartmental thermal hydraulic facility suited for investigations related to the safety of current and advanced LWRs. The facility is multipurpose, and the applications cover integral containment response tests, component tests, primary system tests, and separate effect tests. Experimental investigations carried on in the PANDA facility have been embedded in international projects, most of which under the auspices of the EU and OECD and with the support of a large number of organizations (regulatory bodies, technical dupport organizations, national laboratories, electric utilities, industries) worldwide. The paper provides an overview of the research programs performed in the PANDA facility in relation to BWR containment systems and those planned for PWR containment systems.

  5. Fast flux test facility noise data management

    International Nuclear Information System (INIS)

    Thie, J.A.

    1988-01-01

    An extensive collection of spectra from an automated data collection system at the Fast Flux Facility has features from neutron data extracted and managed by database software. Inquiry techniques, including screening, applied to database results show the influences of control rods on wideband noise and, more generally, abilities to detect diverse types of off-normal noise. Uncovering a temporary 0.1-Hz resonance shift gave additional diagnostic information on a 13-Hz mechanical motion characterized by the interference of two resonances. The latter phenomenon is discussed generically for possible application to other reactor types. (author)

  6. A review of experiments and results from the transient reactor test (TREAT) facility

    International Nuclear Information System (INIS)

    Deitrich, L. W.

    1998-01-01

    The TREAT Facility was designed and built in the late 1950s at Argonne National Laboratory to provide a transient reactor for safety experiments on samples of reactor fuels. It first operated in 1959. Throughout its history, experiments conducted in TREAT have been important in establishing the behavior of a wide variety of reactor fuel elements under conditions predicted to occur in reactor accidents ranging from mild off normal transients to hypothetical core disruptive accidents. For much of its history, TREAT was used primarily to test liquid-metal reactor fuel elements, initially for the Experimental Breeder Reactor-II (EBR-II), then for the Fast Flux Test Facility (FFTF), the Clinch River Breeder Reactor Plant (CRBRP), the British Prototype Fast Reactor (PFR), and finally, for the Integral Fast Reactor (IFR). Both oxide and metal elements were tested in dry capsules and in flowing sodium loops. The data obtained were instrumental in establishing the behavior of the fuel under off-normal and accident conditions, a necessary part of the safety analysis of the various reactors. In addition, TREAT was used to test light-water reactor (LWR) elements in a steam environment to obtain fission-product release data under meltdown conditions. Studies are now under way on applications of TREAT to testing of the behavior of high-burnup LWR elements under reactivity-initiated accident (RIA) conditions using a high-pressure water loop

  7. Elaboration of a guideline for counterpart testing of integral loop systems

    International Nuclear Information System (INIS)

    Karwat, H.

    1988-01-01

    A previous study on the problems of scaling and extrapolation of experimental results indicated the necessity to overcome some uncertainty in the achievable code accuracy when applied to full size reactor systems. The empirical approach in simulating interfacial relationships within systems behaviour codes in combination with the liberty of nodalisation requires a careful assessment of the overall validity of the code verification concept. Several new test facilities are presently under construction to study fluid-dynamic effects of interest for the behaviour of PWRs under small break LOCA conditions in a larger dimension and with somewhat modified scaling considerations. The test results expected from these new facilities may be useful to assess the overall validity of the analytical simulation concept of the applied computer codes. Test facilities of interest within the European Community are the - LOBI Test Facility (vol. scaling 1/700) at JRC-Ispra - SPES Test Facility (vol. scaling 1/425) in Italy - BETHSY Facility (vol scaling 1/100) in France and outside Europe the - ROSA IV Facility (vol. scaling 1/50) in Japan. Main objective of the study was the elaboration of a guideline for the particular requirements of meaningful counterpart testing with the final aim to qualify the overall validity of code simulation concepts and the expected predictive accuracy of code applications for full size prototype PWRs

  8. Altitude simulation facility for testing large space motors

    Science.gov (United States)

    Katz, U.; Lustig, J.; Cohen, Y.; Malkin, I.

    1993-02-01

    This work describes the design of an altitude simulation facility for testing the AKM motor installed in the 'Ofeq' satellite launcher. The facility, which is controlled by a computer, consists of a diffuser and a single-stage ejector fed with preheated air. The calculations of performance and dimensions of the gas extraction system were conducted according to a one-dimensional analysis. Tests were carried out on a small-scale model of the facility in order to examine the design concept, then the full-scale facility was constructed and operated. There was good agreement among the results obtained from the small-scale facility, from the full-scale facility, and from calculations.

  9. Test facility for PLT TF coils

    International Nuclear Information System (INIS)

    Hearney, J.; File, J.; Dreskin, S.

    1975-01-01

    Past experience with the model C stellerator and other toroidal field devices indicates that mechanical and electrical tests of a toroidal field coil prior to maximum field operation of the device is prudent and desirable. This paper describes a test program for the PLT-TF coils. The test stand consists of one test coil, two background coils and a steel supporting structure. The three coil configuration produces a 67.5 kG field at the inner conductor (38 kG at the bore center) and simulates a 1/R field distribution in the bore of the test coil. The resolution of the field force system and resultant stresses within the test structure are discussed. A test procedure is described which maximizes the information obtained from a 100,000 pulse program

  10. PLC based control system for RAM assembly test facility

    International Nuclear Information System (INIS)

    Kulkarni, S.S.; Kumar, Vinaya; Chandra, Umesh

    1994-01-01

    The flexibility, expandability, ease of programming and diagnostic features makes the programmable logic controller (PLC) suitable for a variety of control applications in engineering system test facilities. A PLC based control system for RAM assembly test facility (RATF) and for testing the related hydraulic components is being developed and installed at BARC. This paper describes the approach taken for meeting the control requirements and illustrates the PLC software that has been developed. (author). 1 fig

  11. Propeller Test Facilities Â

    Data.gov (United States)

    Federal Laboratory Consortium — Description: Three electrically driven whirl test stands are used to determine propeller (or other rotating device) performance at various rotational speeds. These...

  12. Air water loop - an experimental facility to study thermal hydraulics of AHWR steam drum

    International Nuclear Information System (INIS)

    Bagul, R.K.; Pilkhwal, D.S.; Jain, V.; Vijayan, P.K.

    2014-05-01

    In the proposed Indian Advanced Heavy Water Reactor (AHWR) the coolant recirculation in the primary system is achieved by two-phase natural circulation. The two-phase steam-water mixture from the reactor core is separated in steam drum by gravity. Gravity separation of phases may lead to undesirable phenomena - carryover and carryunder. Carryover is the entrainment of liquid droplets in the vapor phase.Carryover needs to be minimized to avoid erosion corrosion of turbine blades. Carryunder is the entrainment of vapor bubbles with liquid flowing back to reactor core. Significant carryunder may in turn lead to reduced flow resulting in reduced CHF margin and stability in the coolant channel. An Air-Water Loop (AWL) has been designed to carry out the experiments relevant to AHWR steam drum. The design features and scaling philosophy is described in this report. (author)

  13. BWR Full Integral Simulation Test (FIST) program: facility description report

    International Nuclear Information System (INIS)

    Stephens, A.G.

    1984-09-01

    A new boiling water reactor safety test facility (FIST, Full Integral Simulation Test) is described. It will be used to investigate small breaks and operational transients and to tie results from such tests to earlier large-break test results determined in the TLTA. The new facility's full height and prototypical components constitute a major scaling improvement over earlier test facilities. A heated feedwater system, permitting steady-state operation, and a large increase in the number of measurements are other significant improvements. The program background is outlined and program objectives defined. The design basis is presented together with a detailed, complete description of the facility and measurements to be made. An extensive component scaling analysis and prediction of performance are presented

  14. Project W-049H disposal facility test report

    International Nuclear Information System (INIS)

    Buckles, D.I.

    1995-01-01

    The purpose of this Acceptance Test Report (ATR) for the Project W-049H, Treated Effluent Disposal Facility, is to verify that the equipment installed in the Disposal Facility has been installed in accordance with the design documents and function as required by the project criteria

  15. Cryogenic systems for the Mirror Fusion Test Facility

    International Nuclear Information System (INIS)

    Slack, D.S.; Chronis, W.C.; Nelson, R.L.

    1986-01-01

    This paper will include an in-depth discussion of the design, fabrication, and operation of the Mirror Fusion Test Facility (MFTF) cryogenic system located at Lawrence Livermore National Laboratory (LLNL). Each subsystem will be discussed to present a basic composite of the entire facility

  16. 702AZ aging waste ventilation facility year 2000 test procedure

    International Nuclear Information System (INIS)

    Winkelman, W.D.

    1998-01-01

    This test procedure was developed to determine if the 702AZ Tank Ventilation Facility system is Year 2000 Compliant. The procedure provides detailed instructions for performing the operations necessary and documenting the results. This verification procedure will document that the 702AZ Facility Systems are year 2000 compliant and will correctly meet the criteria established in this procedure

  17. Field Lysimeter Test Facility for protective barriers: Experimental plan

    International Nuclear Information System (INIS)

    Kirkham, R.R.; Gee, G.W.; Downs, J.L.

    1987-12-01

    This document was first written in October 1986 and has been used to guide the design of the Field Lysimeter Test Facility (FLTF) and to promote discussions between research and engineering staff regarding the selection of barrier treatments for inclusion in the FLTF. The construction of the lysimeter facility was completed June 28, 1987. This document describes the facility, the treatments placed in each lysimeter, types of measurements made in each lysimeter, and a brief discussion of project activities related to quality assurance, safety, and funding requirements. The treatment description and figures have been updated to reflect the lysimeter facility as constructed. 12 refs., 6 figs., 5 tabs

  18. An experimental test facility to support development of the fluoride-salt-cooled high-temperature reactor

    International Nuclear Information System (INIS)

    Yoder, Graydon L.; Aaron, Adam; Cunningham, Burns; Fugate, David; Holcomb, David; Kisner, Roger; Peretz, Fred; Robb, Kevin; Wilgen, John; Wilson, Dane

    2014-01-01

    Highlights: • • A forced convection test loop using FLiNaK salt was constructed to support development of the FHR. • The loop is built of alloy 600, and operating conditions are prototypic of expected FHR operation. • The initial test article is designed to study pebble bed heat transfer cooled by FLiNaK salt. • The test facility includes silicon carbide test components as salt boundaries. • Salt testing with silicon carbide and alloy 600 confirmed acceptable loop component lifetime. - Abstract: The need for high-temperature (greater than 600 °C) energy transport systems is significantly increasing as the world strives to improve energy efficiency and develop alternatives to petroleum-based fuels. Liquid fluoride salts are one of the few energy transport fluids that have the capability of operating at high temperatures in combination with low system pressures. The fluoride-salt-cooled high-temperature reactor design uses fluoride salt to remove core heat and interface with a power conversion system. Although a significant amount of experimentation has been performed with these salts, specific aspects of this reactor concept will require experimental confirmation during the development process. The experimental facility described here has been constructed to support the development of the fluoride-salt-cooled high-temperature reactor concept. The facility is capable of operating at up to 700 °C and incorporates a centrifugal pump to circulate FLiNaK salt through a removable test section. A unique inductive heating technique is used to apply heat to the test section, allowing heat transfer testing to be performed. An air-cooled heat exchanger removes added heat. Supporting loop infrastructure includes a pressure control system, a trace heating system, and a complement of instrumentation to measure salt flow, temperatures, and pressures around the loop. The initial experiment is aimed at measuring fluoride-salt heat transfer inside a heated pebble bed

  19. Fast Flux Test Facility (FFTF) standby plan

    Energy Technology Data Exchange (ETDEWEB)

    Hulvey, R.K.

    1997-03-06

    The FFTF Standby Plan, Revision 0, provides changes to the major elements and project baselines to maintain the FFTF plant in a standby condition and to continue washing sodium from irradiated reactor fuel. The Plan is consistent with the Memorandum of Decision approved by the Secretary of Energy on January 17, 1997, which directed that FFTF be maintained in a standby condition to permit the Department to make a decision on whether the facility should play a future role in the Department of Energy`s dual track tritium production strategy. This decision would be made in parallel with the intended December 1998 decision on the selection of the primary, long- term source of tritium. This also allows the Department to review the economic and technical feasibility of using the FFTF to produce isotopes for the medical community. Formal direction has been received from DOE-RL and Fluor 2020 Daniel Hanford to implement the FFTF standby decision. The objective of the Plan is maintain the condition of the FFTF systems, equipment and personnel to preserve the option for plant restart within three and one-half years of a decision to restart, while continuing deactivation work which is consistent with the standby mode.

  20. Fast Flux Test Facility (FFTF) standby plan

    International Nuclear Information System (INIS)

    Hulvey, R.K.

    1997-01-01

    The FFTF Standby Plan, Revision 0, provides changes to the major elements and project baselines to maintain the FFTF plant in a standby condition and to continue washing sodium from irradiated reactor fuel. The Plan is consistent with the Memorandum of Decision approved by the Secretary of Energy on January 17, 1997, which directed that FFTF be maintained in a standby condition to permit the Department to make a decision on whether the facility should play a future role in the Department of Energy's dual track tritium production strategy. This decision would be made in parallel with the intended December 1998 decision on the selection of the primary, long- term source of tritium. This also allows the Department to review the economic and technical feasibility of using the FFTF to produce isotopes for the medical community. Formal direction has been received from DOE-RL and Fluor 2020 Daniel Hanford to implement the FFTF standby decision. The objective of the Plan is maintain the condition of the FFTF systems, equipment and personnel to preserve the option for plant restart within three and one-half years of a decision to restart, while continuing deactivation work which is consistent with the standby mode

  1. Final design of ITER port plug test facility

    Energy Technology Data Exchange (ETDEWEB)

    Cerisier, Thierry, E-mail: thierry.cerisier@yahoo.fr [ITER Organization, Route de Vinon-sur-Verdon, CS 90046, St Paul-lez-Durance Cedex, 13067 (France); Levesy, Bruno [ITER Organization, Route de Vinon-sur-Verdon, CS 90046, St Paul-lez-Durance Cedex, 13067 (France); Romannikov, Alexander [Institution “Project Center ITER”, Kurchatov sq. 1, Building 3, Moscow 123182 (Russian Federation); Rumyantsev, Yuri [JSC “Cryogenmash”, Moscow reg., Balashikha 143907 (Russian Federation); Cordier, Jean-Jacques; Dammann, Alexis [ITER Organization, Route de Vinon-sur-Verdon, CS 90046, St Paul-lez-Durance Cedex, 13067 (France); Minakov, Victor; Rosales, Natalya; Mitrofanova, Elena [JSC “Cryogenmash”, Moscow reg., Balashikha 143907 (Russian Federation); Portone, Sergey; Mironova, Ekaterina [Institution “Project Center ITER”, Kurchatov sq. 1, Building 3, Moscow 123182 (Russian Federation)

    2016-11-01

    Highlights: • We introduce the port plug test facility (purpose and status of the design). • We present the PPTF sub-systems. • We present the environmental and functional tests. • We present the occupational and nuclear safety functions. • We conclude on the achievements and next steps. - Abstract: To achieve the overall ITER machine availability target, the availability of diagnostics and heating port plugs shall be as high as 99.5%. To fulfill this requirement, it is mandatory to test the port plugs at operating temperature before installation on the machine and after refurbishment. The ITER port plug test facility (PPTF) is composed of several test stands that can be used to test the port plugs whereas at the end of manufacturing (in a non-nuclear environment), or after refurbishment in the ITER hot cell facility. The PPTF provides the possibility to perform environmental (leak tightness, vacuum and thermo-hydraulic performances) and functional tests (radio frequency acceptance tests, behavior of the plugs’ steering mechanism and calibration of diagnostics) on upper and equatorial port plugs. The final design of the port plug test facility is described. The configuration of the standalone test stands and the integration in the hot cell facility are presented.

  2. Lead Coolant Test Facility Technical and Functional Requirements, Conceptual Design, Cost and Construction Schedule

    International Nuclear Information System (INIS)

    Soli T. Khericha

    2006-01-01

    This report presents preliminary technical and functional requirements (T and FR), thermal hydraulic design and cost estimate for a lead coolant test facility. The purpose of this small scale facility is to simulate lead coolant fast reactor (LFR) coolant flow in an open lattice geometry core using seven electrical rods and liquid lead or lead-bismuth eutectic. Based on review of current world lead or lead-bismuth test facilities and research need listed in the Generation IV Roadmap, five broad areas of requirements of basis are identified: Develop and Demonstrate Prototype Lead/Lead-Bismuth Liquid Metal Flow Loop Develop and Demonstrate Feasibility of Submerged Heat Exchanger Develop and Demonstrate Open-lattice Flow in Electrically Heated Core Develop and Demonstrate Chemistry Control Demonstrate Safe Operation and Provision for Future Testing. These five broad areas are divided into twenty-one (21) specific requirements ranging from coolant temperature to design lifetime. An overview of project engineering requirements, design requirements, QA and environmental requirements are also presented. The purpose of this T and FRs is to focus the lead fast reactor community domestically on the requirements for the next unique state of the art test facility. The facility thermal hydraulic design is based on the maximum simulated core power using seven electrical heater rods of 420 kW; average linear heat generation rate of 300 W/cm. The core inlet temperature for liquid lead or Pb/Bi eutectic is 420 C. The design includes approximately seventy-five data measurements such as pressure, temperature, and flow rates. The preliminary estimated cost of construction of the facility is $3.7M. It is also estimated that the facility will require two years to be constructed and ready for operation

  3. Fast Flux Test Facility project plan. Revision 2

    International Nuclear Information System (INIS)

    Hulvey, R.K.

    1995-11-01

    The Fast Flux Test Facility (FFTF) Transition Project Plan, Revision 2, provides changes to the major elements and project baseline for the deactivation activities necessary to transition the FFTF to a radiologically and industrially safe shutdown condition

  4. Fast Flux Test Facility project plan. Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    Hulvey, R.K.

    1995-11-01

    The Fast Flux Test Facility (FFTF) Transition Project Plan, Revision 2, provides changes to the major elements and project baseline for the deactivation activities necessary to transition the FFTF to a radiologically and industrially safe shutdown condition.

  5. Technical Evaluation of Oak Ridge Filter Test Facility

    CERN Document Server

    Kriskovich, J R

    2002-01-01

    Two evaluations of the Oak Ridge Department of Energy (DOE) Filter Test Facility (FTF) were performed on December 11 and 12, 2001, and consisted of a quality assurance and a technical evaluation. This report documents results of the technical evaluation.

  6. Super Conducting and Conventional Magnets Test & Mapping Facilities

    Data.gov (United States)

    Federal Laboratory Consortium — Vertical Magnet Test Facility: Accommodate a device up to 3.85 m long, 0.61 m diameter, and 14,400 lbs. Configured for 5 psig sub-cooled liquid helium bath cooling...

  7. Scaling analysis for the OSU AP600 test facility (APEX)

    International Nuclear Information System (INIS)

    Reyes, J.N.

    1998-01-01

    In this paper, the authors summarize the key aspects of a state-of-the-art scaling analysis (Reyes et al. (1995)) performed to establish the facility design and test conditions for the advanced plant experiment (APEX) at Oregon State University (OSU). This scaling analysis represents the first, and most comprehensive, application of the hierarchical two-tiered scaling (H2TS) methodology (Zuber (1991)) in the design of an integral system test facility. The APEX test facility, designed and constructed on the basis of this scaling analysis, is the most accurate geometric representation of a Westinghouse AP600 nuclear steam supply system. The OSU APEX test facility has served to develop an essential component of the integral system database used to assess the AP600 thermal hydraulic safety analysis computer codes. (orig.)

  8. Micro-Combined Heat and Power Device Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — NIST has developed a test facility for micro-combined heat and power (micro-CHP) devices to measure their performance over a range of different operating strategies...

  9. The high-temperature helium test facility (HHV)

    International Nuclear Information System (INIS)

    Noack, G.; Weiskopf, H.

    1977-03-01

    The report describes the high-temperature helium test facility (HHV). Construction of this plant was started in 1972 by Messrs. BBC, Mannheim, on behalf of the Kernforschungsanlage Juelich. By the end of 1976, the construction work is in its last stage, so that the plant may start operation early in 1977. First of all, the cycle system and the arrangement of components are dealt with, followed by a discussion of individual components. Here, emphasis is laid on components typical for HHT systems, while conventional components are mentioned without further structural detail. The projected test programme for the HHV facility in phase IB of the HHT project is shortly dealt with. After this, the potential of this test facility with regard to the possible use of test components and to fluid- and thermodynamic boundary conditions is pointed out. With the unique potential the facility offers here, aspects of shortened service life at higher cycle temperatures do not remain disregarded. (orig./UA) [de

  10. Qualification tests and facilities for the ITER superconductors

    International Nuclear Information System (INIS)

    Bruzzone, P.; Wesche, R.; Stepanov, B.; Cau, F.; Bagnasco, M.; Calvi, M.; Herzog, R.; Vogel, M.

    2009-01-01

    All the ITER superconductors are tested as short length samples in the SULTAN test facility at CRPP. Twenty-four TF conductor samples with small layout variations were tested since February 2007 with the aim of verifying the design and qualification of the manufacturers. The sample assembly and the measurement techniques at CRPP are discussed. Starting in 2010, another test facility for ITER conductors, named EDIPO, will be operating at CRPP to share with SULTAN the load of the samples for the acceptance tests during the construction of ITER.

  11. An assessment of testing requirement impacts on nuclear thermal propulsion ground test facility design

    International Nuclear Information System (INIS)

    Shipers, L.R.; Ottinger, C.A.; Sanchez, L.C.

    1993-01-01

    Programs to develop solid core nuclear thermal propulsion (NTP) systems have been under way at the Department of Defense (DoD), the National Aeronautics and Space Administration (NASA), and the Department of Energy (DOE). These programs have recognized the need for a new ground test facility to support development of NTP systems. However, the different military and civilian applications have led to different ground test facility requirements. The Department of Energy (DOE) in its role as landlord and operator of the proposed research reactor test facilities has initiated an effort to explore opportunities for a common ground test facility to meet both DoD and NASA needs. The baseline design and operating limits of the proposed DoD NTP ground test facility are described. The NASA ground test facility requirements are reviewed and their potential impact on the DoD facility baseline is discussed

  12. Evaluating Past and Future USCG Use of Ohmsett Test Facility

    Science.gov (United States)

    2016-10-01

    of Pages 22 22. Price Evaluating Past and Future USCG Use of Ohmsett Test Facility iv UNCLAS//Public | | CG-926 RDC | M. Fitzpatrick, et al...Opportunity Skimming System WEC Wave energy converter Evaluating Past and Future USCG Use of Ohmsett Test Facility x UNCLAS//Public | | CG-926 RDC | M...Date Summary of Effort OCT-NOV 1993 Vessel of Opportunity Skimming System (VOSS) (5 Weeks) APR-JUN 1996 Spilled Oil Recovery System (SORS) (8 Weeks

  13. String 2, test facility for the LHC

    CERN Multimedia

    Patrice Loïez

    2002-01-01

    String 2 is the long chain seen to the right, representing one complete cell of bending dipoles, focusing quadrupoles and corrector magnets. On 17 June 2002 the test string reached the nominal running current of 11 860 A and magnetic field of 8.335 T for the LHC.

  14. Calibration and use of filter test facility orifice plates

    Science.gov (United States)

    Fain, D. E.; Selby, T. W.

    1984-07-01

    There are three official DOE filter test facilities. These test facilities are used by the DOE, and others, to test nuclear grade HEPA filters to provide Quality Assurance that the filters meet the required specifications. The filters are tested for both filter efficiency and pressure drop. In the test equipment, standard orifice plates are used to set the specified flow rates for the tests. There has existed a need to calibrate the orifice plates from the three facilities with a common calibration source to assure that the facilities have comparable tests. A project has been undertaken to calibrate these orifice plates. In addition to reporting the results of the calibrations of the orifice plates, the means for using the calibration results will be discussed. A comparison of the orifice discharge coefficients for the orifice plates used at the seven facilities will be given. The pros and cons for the use of mass flow or volume flow rates for testing will be discussed. It is recommended that volume flow rates be used as a more practical and comparable means of testing filters. The rationale for this recommendation will be discussed.

  15. Pre-test analysis for identification of natural circulation instabilities in TALL-3D facility

    Energy Technology Data Exchange (ETDEWEB)

    Kööp, Kaspar, E-mail: kaspar@safety.sci.kth.se; Jeltsov, Marti, E-mail: marti@safety.sci.kth.se; Grishchenko, Dmitry, E-mail: dmitry@safety.sci.kth.se; Kudinov, Pavel, E-mail: pavel@safety.sci.kth.se

    2017-04-01

    Highlights: • Global optimum search method was used to identify a region of instability. • Parametric study was used for detailed investigation of system behavior modes. • The results include identification of sustained mass flow rate oscillations. • Recommendations are made for selection of optimal experimental conditions. - Abstract: TALL-3D facility is a lead-bismuth eutectic (LBE) thermal-hydraulic loop designed to provide experimental data on thermal-hydraulics phenomena for validation of stand-alone and coupled System Thermal Hydraulics (STH) and Computational Fluid Dynamics (CFD) codes. Pre-test analysis is crucial for proper choice of experimental conditions at which the experimental data would be most useful for code validation and benchmarking. The goal of this work is to identify these conditions at which the experiment is challenging for the STH codes yet minimizes the 3D-effects from the test section on the loop dynamics. The analysis is focused on the identification of limit cycle flow oscillations in the TALL-3D facility main heater leg using a global optimum search tool GA-NPO to find a general region in the parameter space where oscillatory behavior is expected. As a second step a grid study is conducted outlining the boundaries between different stability modes. Phenomena, simulation results and methodology for selection of the test parameters are discussed in detail and recommendations for experiments are provided.

  16. Preliminary reactor physics calculations for Exxon LWR fuel testing in the power burst facility

    International Nuclear Information System (INIS)

    Olson, W.O.; Nigg, D.W.

    1981-05-01

    The PFB reactor is being considered as an irradiation facility to test LWR fuel rods for Exxon Nuclear Company. Requested test conditions are 18 kW/ft axial peak steady state power in 2.5% initial enrichment, 20,000 MWd/Tu exposed rods. Multigroup transport theory calculations (S/sub n/ and Monte Carlo) showed that this was unattainable in the standard PBF test loop. Thus, a flux multiplier was developed in the form of a Zr-2-clad 0.15-inch thick cylindrical shell of 35% enriched, 88% T.D. UO 2 replacing the flow divider, surrounding the rod within the in-pile tube in PFB. With this flux multiplier installed and assuming an average water density of 0.86 g/cm 3 within the test loop, a Figure of Merit (FOM) for a single-rod test assembly of 0.86 kW/ft-MW +- 5% (at 95% confidence level) was calculated. This FOM is the axial peak linear test rod power per megawatt of reactor power. A reactor power of about 21 megawatts will therefore be required to supply the requested linear test rod axial peak heating rate of 18 kW/ft

  17. The use of an automated flight test management system in the development of a rapid-prototyping flight research facility

    Science.gov (United States)

    Duke, Eugene L.; Hewett, Marle D.; Brumbaugh, Randal W.; Tartt, David M.; Antoniewicz, Robert F.; Agarwal, Arvind K.

    1988-01-01

    An automated flight test management system (ATMS) and its use to develop a rapid-prototyping flight research facility for artificial intelligence (AI) based flight systems concepts are described. The ATMS provides a flight test engineer with a set of tools that assist in flight planning and simulation. This system will be capable of controlling an aircraft during the flight test by performing closed-loop guidance functions, range management, and maneuver-quality monitoring. The rapid-prototyping flight research facility is being developed at the Dryden Flight Research Facility of the NASA Ames Research Center (Ames-Dryden) to provide early flight assessment of emerging AI technology. The facility is being developed as one element of the aircraft automation program which focuses on the qualification and validation of embedded real-time AI-based systems.

  18. UPTF test 21D counterpart test in the MIDAS test facility

    International Nuclear Information System (INIS)

    Yoon, B. C.; Ah, D. J.; Joo, I. C.; Kwon, T. S.; Park, W. M.; Song, C. H.

    2002-01-01

    This paper describes the experimental results of UPTF Test 21D counterpart tests in the downcomer during the late reflood phase of LBLOCA. The experiments have been performed in the MIDAS test facility using superheated steam and water. The test condition was determined,based on the test results of UPTF Test 21D, by applying the 'modified linear scaling method of 1/4.077 length scale. The tests of ECC direct bypass and void height are performed separately to estimate each phenomena quantitatively. The tests were carried out by varying the injection steam flow rate of intact cold legs widely to investigate the effect of steam flow rate on the direct bypass fraction and void height. In the tests, separate effect tests have been performed in cases of DVI-1,DVI- 2 and DVI-1 and 2 injections to see the direct bypass fraction according to the DVI nozzle combination. From the tests, we found that the fraction of direct ECC bypass and the void height observed in the MIDAS test facility reasonably well agree with those of UPTF test 21- D. It confirms that the applied 'modified linear scaling law' reproduces major thermal hydraulics phenomena in the downcomer during the LBLOCA reflood phase

  19. Aerospace Structures Test Facility Environmental Test Chambers (ETC)

    Data.gov (United States)

    Federal Laboratory Consortium — Purpose: The ETCs test the structural integrity of aerospace structures in representative operating temperatures and aerodynamic load distributions. The test article...

  20. The Application of Hardware in the Loop Testing for Distributed Engine Control

    Science.gov (United States)

    Thomas, George L.; Culley, Dennis E.; Brand, Alex

    2016-01-01

    The essence of a distributed control system is the modular partitioning of control function across a hardware implementation. This type of control architecture requires embedding electronics in a multitude of control element nodes for the execution of those functions, and their integration as a unified system. As the field of distributed aeropropulsion control moves toward reality, questions about building and validating these systems remain. This paper focuses on the development of hardware-in-the-loop (HIL) test techniques for distributed aero engine control, and the application of HIL testing as it pertains to potential advanced engine control applications that may now be possible due to the intelligent capability embedded in the nodes.

  1. Analysis of the SBLOCAs in HANARO pool for the 3-pin fuel test loop

    International Nuclear Information System (INIS)

    Park, S. K.; Chi, D. Y.; Sim, B. S.; Park, K. N.; Ahn, S. H.; Lee, J. M.; Lee, C. Y.; Kim, Y. J.

    2004-09-01

    Fuel Test Loop(FTL) has been developed to meet the increasing demand on fuel irradiation and burn up test required the development of new fuels in Korea. It is designed to provide the test conditions of high pressure and temperature like the commercial PWR and CANDU power plants. And also the FTL have the cooling capability to sufficiently remove the thermal power of the in-pile test section for normal operation, Anticipated Operational Occurrences(AOOs), and Design Basis Accidents(DBAs). This report deals with the Small Break Loss Of Coolant Accidents (SBLOCAs) in HANARO pool for the 3-pin fuel test loop. The MARS code has been used for the prediction of the emergency core cooling capability of the FTL and the peak cladding temperature of the test fuels for the SBLOCAs. The location of the pipe break is assumed at the hill taps connecting the cold and hot legs in HANARO pool to the inlet and outlet nozzles of the In-Pile test Section (IPS). The break size is also assumed less than 20% of the cross section area of the pipe. The test fuels are heated up when the cold leg break occur. However, they are not heated up when the hot leg break occur. The maximum Peak Cladding Temperatures (PCT) are predicted to be about 906.9 .deg. C for the cold leg break accident in PWR fuel test mode and 971.9 .deg. C in CANDU fuel test mode respectively. The critical break size is about the 6% of the cross section area of the pipe for PWR fuel test mode and the 8% for CANDU fuel test mode. The PCTs meet the design criterion of commercial PWR fuel that the maximum PCT is lower than 1204 .deg. C

  2. Biaxial wheel/hub test facility. Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, G.; Grubisic, V. [eds.

    2000-07-01

    The 4{sup th} meeting aims to exchange the experience and knowledge of engineers during several presentations and discussions about new developments required for a reliable, time and cost reducing validation of the wheel/hub assembly. Tremendous development of the wheel performance, described by the ratio of the rated load (kg) versus the wheel weight (kg) had taken place during the last 5000 years. Starting from the ratio of 3 for wooden 2-piece-disc-wheels in Mesopotamia it needed nearly 1000 years to increase the ratio to approx 5 at light-weight spoke wheels for fighting carriages, found in the grave of king Tutenchamon in Egypt. Modern light alloy wheels of commercial vehicles reach values up to 160 kg/kg. Additionally the comlex design of the modern systems for cars and commercial vehicles comprising wheel, brake, hub, bearing, spindle and hub carrier, including different materials and their treatment, fasteners, press-fits, require an appropriate testing procedure. The variable loading conditions, caused by operational wheel forces, brake and torque moments including heating, may result in changing tolerances and press-fits during operation and consequently in different damage mechanisms. This can be simulated in the Biaxial Wheel Test Machine, whereby corresponding load programs are necessary. An overview about all biaxial test machines in usage at the end of 1999 is shown in the introduction. The total number is 17 for cars, 7 for commercial vehicles and 1 for trains. The six presentations of this meeting were consequently concentrated on: (a) recommendations for a standardization of load programs of the German Wheel Committee, (b) the simulation of brake and torque events and (c) the possibility for a numerical stress analyses and fatigue life assessment. (orig./AKF)

  3. The construction of solid waste form test facility

    International Nuclear Information System (INIS)

    Park, Hun Hwee; Kim, Joon Hyung; Lee, Byung Jik; Koo, Jun Mo; Kim, Jeong Guk; Jung, In Ha

    1990-03-01

    The solid waste form test facility (SWFTF) to test and/or evaluate the characteristics of waste forms, such as homogeniety, mechanical properties, thermal properties, waste resistance and leachability, have been constructed, and some equipments for testing actual waste forms has been purchased; radiocative monitoring system, glove box for the manipulator repair room, and uninteruppted power supply system, et al. Classifications of radioactive wastes, basic requirements and criteria to be considered during waste management were also reviewed. Some of the described items above have been standardized for the purpose of indigenigation. Therefore, safety assurance of waste forms, as well as increase in the range of participating of domestic companies in construction of further nuclear facilities could be obtained as results through constructing this facility. In the furture this facility is going to be utilized not only for the inspection of waste forms but also for the periodic decontamination for extending the life time of some expensive radiological equipments using remote handling techniques. (author)

  4. Test facilities for evaluating nuclear thermal propulsion systems

    International Nuclear Information System (INIS)

    Beck, D.F.; Allen, G.C.; Shipers, L.R.; Dobranich, D.; Ottinger, C.A.; Harmon, C.D.; Fan, W.C.; Todosow, M.

    1992-01-01

    Interagency panels evaluating nuclear thermal propulsion (NTP) development options have consistently recognized the need for constructing a major new ground test facility to support fuel element and engine testing. This paper summarizes the requirements, configuration, and baseline performance of some of the major subsystems designed to support a proposed ground test complex for evaluating nuclear thermal propulsion fuel elements and engines being developed for the Space Nuclear Thermal Propulsion (SNTP) program. Some preliminary results of evaluating this facility for use in testing other NTP concepts are also summarized

  5. Stored energy analysis in scale-down test facility

    International Nuclear Information System (INIS)

    Deng Chengcheng; Qin Benke; Fang Fangfang; Chang Huajian; Ye Zishen

    2013-01-01

    In the integral test facilities that simulate the accident transient process of the prototype nuclear power plant, the stored energy in the metal components has a direct influence on the simulation range and the test results of the facilities. Based on the heat transfer theory, three methods analyzing the stored energy were developed, and a thorough study on the stored energy problem in the scale-down test facilities was further carried out. The lumped parameter method and power integration method were applied to analyze the transient process of energy releasing and to evaluate the average total energy stored in the reactor pressure vessel of the ACME (advanced core-cooling mechanism experiment) facility, which is now being built in China. The results show that the similarity requirements for such three methods to analyze the stored energy in the test facilities are reduced gradually. Under the condition of satisfying the integral similarity of natural circulation, the stored energy releasing process in the scale-down test facilities can't maintain exact similarity. The stored energy in the reactor pressure vessel wall of ACME, which is released quickly during the early stage of rapid depressurization of system, will not make a major impact on the long-term behavior of system. And the scaling distortion of integral average total energy of the stored heat is acceptable. (authors)

  6. Large scale reflood test with cylindrical core test facility (CCTF). Core I. FY 1979 tests

    International Nuclear Information System (INIS)

    Murao, Yoshio; Akimoto, Hajime; Okubo, Tsutomu; Sudoh, Takashi; Hirano, Kenmei

    1982-03-01

    This report presents the results of analysis of the data obtained in the CCTF Core I test series (19 tests) in FY. 1979 as an interim report. The Analysis of the test results showed that: (1) The present safety evaluation model on the reflood phenomena during LOCA conservatively represents the phenomena observed in the tests except for the downcomer thermohydrodynamic behavior. (2) The downcomer liquid level rose slowly and it took long time for the water to reach a terminal level or the spill-over level. It was presume that such a results was due to an overly conservative selection of the ECC flow rate. This presumption will be checked against a future test result for an increased flow rate. The loop-seal-water filling test was unsuccessful due to a premature power shutdown by the core protection circuit. The test will be conducted again. The tests to be performed in the future are summerized. Tests for investigation of the refill phenomena were also proposed. (author)

  7. Flow analysis of HANARO flow simulated test facility

    International Nuclear Information System (INIS)

    Park, Yong-Chul; Cho, Yeong-Garp; Wu, Jong-Sub; Jun, Byung-Jin

    2002-01-01

    The HANARO, a multi-purpose research reactor of 30 MWth open-tank-in-pool type, has been under normal operation since its initial critical in February, 1995. Many experiments should be safely performed to activate the utilization of the NANARO. A flow simulated test facility is being developed for the endurance test of reactivity control units for extended life times and the verification of structural integrity of those experimental facilities prior to loading in the HANARO. This test facility is composed of three major parts; a half-core structure assembly, flow circulation system and support system. The half-core structure assembly is composed of plenum, grid plate, core channel with flow tubes, chimney and dummy pool. The flow channels are to be filled with flow orifices to simulate core channels. This test facility must simulate similar flow characteristics to the HANARO. This paper, therefore, describes an analytical analysis to study the flow behavior of the test facility. The computational flow analysis has been performed for the verification of flow structure and similarity of this test facility assuming that flow rates and pressure differences of the core channel are constant. The shapes of flow orifices were determined by the trial and error method based on the design requirements of core channel. The computer analysis program with standard k - ε turbulence model was applied to three-dimensional analysis. The results of flow simulation showed a similar flow characteristic with that of the HANARO and satisfied the design requirements of this test facility. The shape of flow orifices used in this numerical simulation can be adapted for manufacturing requirements. The flow rate and the pressure difference through core channel proved by this simulation can be used as the design requirements of the flow system. The analysis results will be verified with the results of the flow test after construction of the flow system. (author)

  8. Analysis of th SBLOCAs in the room 1 for the 3-pin fuel test loop

    International Nuclear Information System (INIS)

    Park, S. K.; Chi, D. Y.; Sim, B. S.; Park, K. N.; Ahn, S. H.; Lee, J. M.; Lee, C. Y.; Kim, Y. J.

    2004-10-01

    Fuel Test Loop(FTL) has been developed to meet the increasing demand on fuel irradiation and burn up test required the development of new fuels in Korea. It is designed to provide the test conditions of high pressure and temperature like the commercial PWR and CANDU power plants. And also the FTL have the cooling capability to sufficiently remove the thermal power of the in-pile test section for normal operation, Anticipated Operational Occurrences(AOOs), and Design Basis Accidents(DBAs). This report deals with the Small Break Loss of Coolant Accidents (SBLOCAs) in the Room 1 for the 3-pin fuel test loop. The MARS code has been used for the prediction of the emergency core cooling capability of the FTL and the peak cladding temperature of the test fuels for the SBLOCAs. The location of the pipe break is assumed at the downstream of the main cooling water pump and the upstream of the main cooler in the room 1. The break size is also assumed less than 20% of the cross section area of the pipe. The test fuels are heated up when the cold leg break occur. However, they are not heated up when the hot leg break occur. The maximum Peak Cladding Temperature (PCT) is predicted to be about 931.4 .deg. C for the cold leg break accident in PWR fuel test mode and 931.6 .deg. C in CANDU fuel test mode respectively. The critical break size is about the 8% of the cross section area of the pipe for PWR fuel test mode and the 10% for CANDU fuel test mode. The PCTs meet the design criterion of commercial PWR fuel that the maximum PCT is lower than 1204 .deg. C

  9. Analysis of the LBLOCAs in the HANARO pool for the 3-pin fuel test loop

    International Nuclear Information System (INIS)

    Park, S. K.; Chi, D. Y.; Sim, B. S.; Park, K. N.; Ahn, S. H.; Lee, J. M.; Lee, C. Y.; Kim, Y. J.

    2004-12-01

    The Fuel Test Loop(FTL) has been developed to meet the increasing demand on fuel irradiation and burn up test required the development of new fuels in Korea. It is designed to provide the test conditions of high pressure and temperature like the commercial PWR and CANDU power plants. And also the FTL have the cooling capability to sufficiently remove the thermal power of the in-pile test section for normal operation, Anticipated Operational Occurrences(AOOs), and Design Basis Accidents(DBAs). This report deals with the Large Break Loss of Coolant Accidents (LBLOCAs) in HANARO pool for the 3-pin fuel test loop. The MARS code has been used for the prediction of the emergency core cooling capability of the FTL and the peak cladding temperature of the test fuels for the LBLOCAs. The location of the pipe break is assumed at the hill taps connecting the cold and hot legs in HANARO pool to the inlet and outlet nozzles of the In-Pile test Section (IPS). Double ended guillotine break is assumed for the large break loss of coolant accidents. The discharge coefficients of 0.1, 0.33, 0.67, 1.0 are investigated for the LBLOCAs. The test fuels for PWR and CANDU test modes are not heated up for the LBLOCAs caused by the double ended guillotine break in the HANARO pool. The reason is that the sufficient emergency cooling water to cool down the test fuels is supplied continuously to the in-pile test section. Therefore the PCTs for the LBLOCAs in the HANARO pool meet the design criterion of commercial PWR fuel that maximum PCT is lower than 1204 .deg. C

  10. Australian national networked tele-test facility for integrated systems

    Science.gov (United States)

    Eshraghian, Kamran; Lachowicz, Stefan W.; Eshraghian, Sholeh

    2001-11-01

    The Australian Commonwealth government recently announced a grant of 4.75 million as part of a 13.5 million program to establish a world class networked IC tele-test facility in Australia. The facility will be based on a state-of-the-art semiconductor tester located at Edith Cowan University in Perth that will operate as a virtual centre spanning Australia. Satellite nodes will be located at the University of Western Australia, Griffith University, Macquarie University, Victoria University and the University of Adelaide. The facility will provide vital equipment to take Australia to the frontier of critically important and expanding fields in microelectronics research and development. The tele-test network will provide state of the art environment for the electronics and microelectronics research and the industry community around Australia to test and prototype Very Large Scale Integrated (VLSI) circuits and other System On a Chip (SOC) devices, prior to moving to the manufacturing stage. Such testing is absolutely essential to ensure that the device performs to specification. This paper presents the current context in which the testing facility is being established, the methodologies behind the integration of design and test strategies and the target shape of the tele-testing Facility.

  11. High temperature high vacuum creep testing facilities

    International Nuclear Information System (INIS)

    Matta, M.K.

    1985-01-01

    Creep is the term used to describe time-dependent plastic flow of metals under conditions of constant load or stress at constant high temperature. Creep has an important considerations for materials operating under stresses at high temperatures for long time such as cladding materials, pressure vessels, steam turbines, boilers,...etc. These two creep machines measures the creep of materials and alloys at high temperature under high vacuum at constant stress. By the two chart recorders attached to the system one could register time and temperature versus strain during the test . This report consists of three chapters, chapter I is the introduction, chapter II is the technical description of the creep machines while chapter III discuss some experimental data on the creep behaviour. Of helium implanted stainless steel. 13 fig., 3 tab

  12. The TALL-3D facility design and commissioning tests for validation of coupled STH and CFD codes

    Energy Technology Data Exchange (ETDEWEB)

    Grishchenko, Dmitry, E-mail: dmitry@safety.sci.kth.se; Jeltsov, Marti, E-mail: marti@safety.sci.kth.se; Kööp, Kaspar, E-mail: kaspar@safety.sci.kth.se; Karbojian, Aram, E-mail: karbojan@kth.se; Villanueva, Walter, E-mail: walter@safety.sci.kth.se; Kudinov, Pavel, E-mail: pavel@safety.sci.kth.se

    2015-08-15

    Highlights: • Design of a heavy liquid thermal-hydraulic loop for CFD/STH code validation. • Description of the loop instrumentation and assessment of measurement error. • Experimental data from forced to natural circulation transient. - Abstract: Application of coupled CFD (Computational Fluid Dynamics) and STH (System Thermal Hydraulics) codes is a prerequisite for computationally affordable and sufficiently accurate prediction of thermal-hydraulics of complex systems. Coupled STH and CFD codes require validation for understanding and quantification of the sources of uncertainties in the code prediction. TALL-3D is a liquid Lead Bismuth Eutectic (LBE) loop developed according to the requirements for the experimental data for validation of coupled STH and CFD codes. The goals of the facility design are to provide (i) mutual feedback between natural circulation in the loop and complex 3D mixing and stratification phenomena in the pool-type test section, (ii) a possibility to validate standalone STH and CFD codes for each subsection of the facility, and (iii) sufficient number of experimental data to separate the process of input model calibration and code validation. Description of the facility design and its main components, approach to estimation of experimental uncertainty and calibration of model input parameters that are not directly measured in the experiment are discussed in the paper. First experimental data from the forced to natural circulation transient is also provided in the paper.

  13. Test system design for Hardware-in-Loop evaluation of PEM fuel cells and auxiliaries

    Energy Technology Data Exchange (ETDEWEB)

    Randolf, Guenter; Moore, Robert M. [Hawaii Natural Energy Institute, University of Hawaii, Honolulu, HI (United States)

    2006-07-14

    In order to evaluate the dynamic behavior of proton exchange membrane (PEM) fuel cells and their auxiliaries, the dynamic capability of the test system must exceed the dynamics of the fastest component within the fuel cell or auxiliary component under test. This criterion is even more critical when a simulated component of the fuel cell system (e.g., the fuel cell stack) is replaced by hardware and Hardware-in-Loop (HiL) methodology is employed. This paper describes the design of a very fast dynamic test system for fuel cell transient research and HiL evaluation. The integration of the real time target (which runs the simulation), the test stand PC (that controls the operation of the test stand), and the programmable logic controller (PLC), for safety and low-level control tasks, into one single integrated unit is successfully completed. (author)

  14. Analysis of the October 5, 1979 lithium spill and fire in the Lithium Processing Test Loop

    International Nuclear Information System (INIS)

    Maroni, V.A.; Beatty, R.A.; Brown, H.L.; Coleman, L.F.; Foose, R.M.; McPheeters, C.C.; Slawecki, M.; Smith, D.L.; Van Deventer, E.H.; Weston, J.R.

    1981-12-01

    On October 5, 1979, the Lithium Processing Test Loop (LPTL) developed a lithium leak in the electromagnetic (EM) pump channel, which damaged the pump, its surrounding support structure, and the underlying floor pan. A thorough analysis of the causes and consequences of the pump failure was conducted by personnel from CEN and several other ANL divisions. Metallurgical analyses of the elliptical pump channel and adjacent piping revealed that there was a significant buildup of iron-rich crystallites and other solid material in the region of the current-carrying bus bars (region of high magnetic field), which may have resulted in a flow restriction that contributed to the deterioration of the channel walls. The location of the failure was in a region of high residual stress (due to cold work produced during channel fabrication); this failure is typical of other cold work/stress-related failures encountered in components operated in forced-circulation lithium loops. Another important result was the isolation of crystals of a compound characterized as Li/sub x/CrN/sub y/. Compounds of this type are believed to be responsible for much of the Fe, Cr, and Ni mass transfer encountered in lithium loops constructed of stainless steel. The importance of nitrogen in the mass-transfer mechanism has long been suspected, but the existence of stable ternary Li-M-N compounds (M = Fe, Cr, Ni) had not previously been verified

  15. Determination of Importance Evaluation for Exploratory Studies Facility (ESF) Subsurface Testing Activities

    International Nuclear Information System (INIS)

    C.J. Byrne

    2001-01-01

    This Determination of Importance Evaluation (DIE) applies to the Subsurface Exploratory Studies Facility (ESF), encompassing the Topopah Spring (TS) Loop from Station 0+00 meters (m) at the North Portal to breakthrough at the South Portal (approximately 78+77 m), and ancillary test and operation support areas including the Enhanced Characterization of the Repository Block (ECRB) Cross Drift. This evaluation applies specifically to site characterization testing activities ongoing and planned in the Subsurface ESF. ESF site characterization activities are being performed to obtain the information necessary to determine whether the Yucca Mountain Site is suitable as a geologic repository for spent nuclear fuel and high-level radioactive waste. A more detailed description of these testing activities is provided in Section 6 of this DIE. Generally, the construction and operation of excavations associated with these testing activities are evaluated in the DIE for the Subsurface ESF (CRWMS M and O 1999a) and the DIE for the ESF ECRB Cross Drift (CRWMS M and O 2000a). The scope of this DIE also entails the proposed Unsaturated Zone (UZ) Transport Test at Busted Butte. Although, not a part of the TS Loop or ECRB Cross Drift, the associated testing activities are Subsurface testing activities. Busted Butte is located to the south south-east of the TS Loop and is outside the Conceptual Controlled Area Boundary (CCAB). These activities provide access to the Calico Hills (CH) geologic structure. In the case of Busted Butte, construction and operation of excavations are evaluated herein (since this activity was not previously evaluated in CRWMS M and O 1999a). The objectives of this DIE are to determine whether Subsurface ESF testing, and associated activities, could potentially impact site characterization testing and/or the waste isolation capabilities of the site. Controls needed to limit any potential impacts are identified in Section 13. The validity and veracity of the

  16. Determination of Importance Evaluation for Exploratory Studies Facility (ESF) Subsurface Testing Activities

    Energy Technology Data Exchange (ETDEWEB)

    C.J. Byrne

    2001-02-20

    This Determination of Importance Evaluation (DIE) applies to the Subsurface Exploratory Studies Facility (ESF), encompassing the Topopah Spring (TS) Loop from Station 0+00 meters (m) at the North Portal to breakthrough at the South Portal (approximately 78+77 m), and ancillary test and operation support areas including the Enhanced Characterization of the Repository Block (ECRB) Cross Drift. This evaluation applies specifically to site characterization testing activities ongoing and planned in the Subsurface ESF. ESF site characterization activities are being performed to obtain the information necessary to determine whether the Yucca Mountain Site is suitable as a geologic repository for spent nuclear fuel and high-level radioactive waste. A more detailed description of these testing activities is provided in Section 6 of this DIE. Generally, the construction and operation of excavations associated with these testing activities are evaluated in the DIE for the Subsurface ESF (CRWMS M&O 1999a) and the DIE for the ESF ECRB Cross Drift (CRWMS M&O 2000a). The scope of this DIE also entails the proposed Unsaturated Zone (UZ) Transport Test at Busted Butte. Although, not a part of the TS Loop or ECRB Cross Drift, the associated testing activities are Subsurface testing activities. Busted Butte is located to the south south-east of the TS Loop and is outside the Conceptual Controlled Area Boundary (CCAB). These activities provide access to the Calico Hills (CH) geologic structure. In the case of Busted Butte, construction and operation of excavations are evaluated herein (since this activity was not previously evaluated in CRWMS M&O 1999a). The objectives of this DIE are to determine whether Subsurface ESF testing, and associated activities, could potentially impact site characterization testing and/or the waste isolation capabilities of the site. Controls needed to limit any potential impacts are identified in Section 13. The validity and veracity of the individual

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-08-01

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

  18. Determination of Importance Evaluation for Exploratory Studies Facility (ESF) Subsurface Testing Activities

    International Nuclear Information System (INIS)

    Goodin, S.

    2002-01-01

    This Determination of Importance Evaluation (DIE) applies to the Subsurface Exploratory Studies Facility (ESF), encompassing the Topopah Spring (TS) Loop from Station 0+00 meters (m) at the North Portal to breakthrough at the South Portal (approximately 78+77 m), and ancillary test and operation support areas including the Enhanced Characterization of the Repository Block (ECRB) Cross Drift. This evaluation applies specifically to site characterization testing activities ongoing and planned in the Subsurface ESF. ESF site characterization activities are being performed to obtain the information necessary to determine whether the Yucca Mountain Site is suitable as a geologic repository for spent nuclear fuel and high-level radioactive waste. A more detailed description of these testing activities is provided in Section 6 of this DIE. Generally, the construction and operation of excavations associated with these testing activities are evaluated in the DIE for the Subsurface ESF (CRWMS M andO 1999a) and the DIE for the ESF ECRB Cross Drift (CRWMS M andO 2000a). The scope of this DIE also entails the proposed Unsaturated Zone (UZ) Transport Test at Busted Butte. Although, not a part of the TS Loop or ECRB Cross Drift, the associated testing activities are Subsurface testing activities. Busted Butte is located to the south south-east of the TS Loop and is outside the Conceptual Controlled Area Boundary (CCAB). These activities provide access to the Calico Hills (CH) geologic structure. In the case of Busted Butte, construction and operation of excavations are evaluated herein (since this activity was not previously evaluated in CRWMS M andO 1999a). The objectives of this DIE are to determine whether Subsurface ESF testing, and associated activities, could potentially impact site characterization testing and/or the waste isolation capabilities of the site. Controls needed to limit any potential impacts are identified in Section 13. The validity and veracity of the

  19. National RF Test Facility as a multipurpose development tool

    International Nuclear Information System (INIS)

    McManamy, T.J.; Becraft, W.R.; Berry, L.A.

    1983-01-01

    Additions and modifications to the National RF Test Facility design have been made that (1) focus its use for technology development for future large systems in the ion cyclotron range of frequencies (ICRF), (2) expand its applicability to technology development in the electron cyclotron range of frequencies (ECRF) at 60 GHz, (3) provide a facility for ELMO Bumpy Torus (EBT) 60-GHz ring physics studies, and (4) permit engineering studies of steady-state plasma systems, including superconducting magnet performance, vacuum vessel heat flux removal, and microwave protection. The facility will continue to function as a test bed for generic technology developments for ICRF and the lower hybrid range of frequencies (LHRF). The upgraded facility is also suitable for mirror halo physics experiments

  20. Research and test facilities required in nuclear science and technology

    International Nuclear Information System (INIS)

    2009-01-01

    Experimental facilities are essential research tools both for the development of nuclear science and technology and for testing systems and materials which are currently being used or will be used in the future. As a result of economic pressures and the closure of older facilities, there are concerns that the ability to undertake the research necessary to maintain and to develop nuclear science and technology may be in jeopardy. An NEA expert group with representation from ten member countries, the International Atomic Energy Agency and the European Commission has reviewed the status of those research and test facilities of interest to the NEA Nuclear Science Committee. They include facilities relating to nuclear data measurement, reactor development, neutron scattering, neutron radiography, accelerator-driven systems, transmutation, nuclear fuel, materials, safety, radiochemistry, partitioning and nuclear process heat for hydrogen production. This report contains the expert group's detailed assessment of the current status of these nuclear research facilities and makes recommendations on how future developments in the field can be secured through the provision of high-quality, modern facilities. It also describes the online database which has been established by the expert group which includes more than 700 facilities. (authors)

  1. Fusion Materials Irradiation Test Facility: a facility for fusion-materials qualification

    International Nuclear Information System (INIS)

    Trego, A.L.; Hagan, J.W.; Opperman, E.K.; Burke, R.J.

    1983-01-01

    The Fusion Materials Irradiation Test Facility will provide a unique testing environment for irradiation of structural and special purpose materials in support of fusion power systems. The neutron source will be produced by a deuteron-lithium stripping reaction to generate high energy neutrons to ensure damage similar to that of a deuterium-tritium neutron spectrum. The facility design is now ready for the start of construction and much of the supporting lithium system research has been completed. Major testing of key low energy end components of the accelerator is about to commence. The facility, its testing role, and the status and major aspects of its design and supporting system development are described

  2. RAMI strategies in the IFMIF Test Facilities design

    Energy Technology Data Exchange (ETDEWEB)

    Abal, Javier, E-mail: javier.abal@upc.edu [Fusion Energy Engineering Laboratory (FEEL), Technical University of Catalonia (UPC) Barcelona-Tech, Barcelona (Spain); Dies, Javier [Fusion Energy Engineering Laboratory (FEEL), Technical University of Catalonia (UPC) Barcelona-Tech, Barcelona (Spain); Arroyo, José Manuel [Laboratorio Nacional de Fusión por Confinamiento Magnético – CIEMAT, 28040 Madrid (Spain); Bargalló, Enric [Fusion Energy Engineering Laboratory (FEEL), Technical University of Catalonia (UPC) Barcelona-Tech, Barcelona (Spain); Casal, Natalia; García, Ángela [Laboratorio Nacional de Fusión por Confinamiento Magnético – CIEMAT, 28040 Madrid (Spain); Martínez, Gonzalo; Tapia, Carlos; De Blas, Alfredo [Fusion Energy Engineering Laboratory (FEEL), Technical University of Catalonia (UPC) Barcelona-Tech, Barcelona (Spain); Mollá, Joaquín; Ibarra, Ángel [Laboratorio Nacional de Fusión por Confinamiento Magnético – CIEMAT, 28040 Madrid (Spain)

    2013-10-15

    Highlights: • We have implemented fault tolerant design strategies so that the strong availability requirements are met. • The evolution to the present design of the signal and cooling lines inside the TTC has also been compared. • The RAMI analyses have demonstrated a strong capability in being a complementary tool in the design of IFMIF Test Facilities. -- Abstract: In this paper, a RAMI analysis of the different stages in Test Facilities (TF) design is described. The comparison between the availability results has been a milestone not only to evaluate the major unavailability contributors in the updates but also to implement fault tolerant design strategies when possible. These strategies encompass a wide range of design activities: from the definition of degraded modes of operation in the Test Facilities to specific modifications in the test modules in order to guarantee their fail safe operation.

  3. RAMI strategies in the IFMIF Test Facilities design

    International Nuclear Information System (INIS)

    Abal, Javier; Dies, Javier; Arroyo, José Manuel; Bargalló, Enric; Casal, Natalia; García, Ángela; Martínez, Gonzalo; Tapia, Carlos; De Blas, Alfredo; Mollá, Joaquín; Ibarra, Ángel

    2013-01-01

    Highlights: • We have implemented fault tolerant design strategies so that the strong availability requirements are met. • The evolution to the present design of the signal and cooling lines inside the TTC has also been compared. • The RAMI analyses have demonstrated a strong capability in being a complementary tool in the design of IFMIF Test Facilities. -- Abstract: In this paper, a RAMI analysis of the different stages in Test Facilities (TF) design is described. The comparison between the availability results has been a milestone not only to evaluate the major unavailability contributors in the updates but also to implement fault tolerant design strategies when possible. These strategies encompass a wide range of design activities: from the definition of degraded modes of operation in the Test Facilities to specific modifications in the test modules in order to guarantee their fail safe operation

  4. Fast Flux Test Facility replacement of a primary sodium pump

    International Nuclear Information System (INIS)

    Krieg, S.A.; Thomson, J.D.

    1985-01-01

    The Fast Flux Test Facility is a 400 MW Thermal Sodium Cooled Fast Reactor operated by Westinghouse Hanford Company for the US Department of Energy. During startup testing in 1979, the sodium level in one of the primary sodium pumps was inadvertently raised above the normal height. This resulted in distortion of the pump shaft. Pump replacement was carried out using special maintenance equipment. Nuclear radiation and contamination were not significant problems since replacement operations were carried out shortly after startup of the Fast Flux Test Facility

  5. A HWIL test facility of infrared imaging laser radar using direct signal injection

    Science.gov (United States)

    Wang, Qian; Lu, Wei; Wang, Chunhui; Wang, Qi

    2005-01-01

    Laser radar has been widely used these years and the hardware-in-the-loop (HWIL) testing of laser radar become important because of its low cost and high fidelity compare with On-the-Fly testing and whole digital simulation separately. Scene generation and projection two key technologies of hardware-in-the-loop testing of laser radar and is a complicated problem because the 3D images result from time delay. The scene generation process begins with the definition of the target geometry and reflectivity and range. The real-time 3D scene generation computer is a PC based hardware and the 3D target models were modeled using 3dsMAX. The scene generation software was written in C and OpenGL and is executed to extract the Z-buffer from the bit planes to main memory as range image. These pixels contain each target position x, y, z and its respective intensity and range value. Expensive optical injection technologies of scene projection such as LDP array, VCSEL array, DMD and associated scene generation is ongoing. But the optical scene projection is complicated and always unaffordable. In this paper a cheaper test facility was described that uses direct electronic injection to provide rang images for laser radar testing. The electronic delay and pulse shaping circuits inject the scenes directly into the seeker's signal processing unit.

  6. Electrical energy and cost for the Mirror Fusion Test Facility

    International Nuclear Information System (INIS)

    Pence, G.A.

    1983-01-01

    An operational scenario has been developed for the Mirror Fusion Test Facility (MFTF-B) based on the System Requirements, our experience with existing systems, and discussions with the project engineers and designers who are responsible for the systems. This scenario was used to predict the amount of electrical energy needed for running the facility. A generic type listing is included for the equipment considered in each system

  7. Electrical energy and cost for the Mirror Fusion Test Facility

    International Nuclear Information System (INIS)

    Pence, G.A.

    1983-02-01

    An operational scenario for the Mirror Fusion Test Facility has been developed based on System Requirements, experience with existing systems, and discussions with project engineers and designers who are responsible for the systems. This scenario was used to project the electrical energy required for the facility. Each system is listed showing the equipment that has been considered, the amount of power requested, and in most cases, the power that it is now connected

  8. Cryogenic systems for the Mirror Fusion Test Facility

    International Nuclear Information System (INIS)

    Slack, D.S.; Nelson, R.L.; Chronis, W.C.

    1985-08-01

    This paper includes an in-depth discussion of the design, fabrication, and operation of the Mirror Fusion Test Facility (MFTF) cryogenic system located at Lawrence Livermore National Laboratory (LLNL). Each subsystem discussed to present a basic composite of the entire facility. The following subsystems are included: 500kW nitrogen reliquefier, subcoolers, and distribution system; 15kW helium refrigerator/liquefier and distribution system; helium recovery and storage system; rough vacuum and high vacuum systems

  9. Operation of the Brookhaven national laboratory accelerator test facility

    International Nuclear Information System (INIS)

    Batchelor, K.; Ben-Zvi, I.; Botke, I.; Chou, T.S.; Fernow, R.; Fischer, J.; Fisher, A.; Gallardo, J.; Ingold, G.; Malone, R.; Palmer, R.; Parsa, Z.; Pogorelsky, I.; Rogers, J.; Sheehan, J.; Srinivasan-Rao, T.; Tsang, T.; Ulc, S.; Van Steenbergen, A.; Wang, X.J.; Woodle, M.; Yu, L.H.

    1992-01-01

    Early operation of the 50 MeV high brightness electron linac of the Accelerator Test Facility is described along with experimental data. This facility is designed to study new linear acceleration techniques and new radiation sources based on linacs in combination with free electron lasers. The accelerator utilizes a photo-excited, metal cathode, radio frequency electron gun followed by two travelling wave accelerating sections and an Experimental Hall for the study program. (Author) 5 refs., 4 figs., tab

  10. Operation of the Brookhaven National Laboratory Accelerator Test Facility

    International Nuclear Information System (INIS)

    Batchelor, K.; Ben-Zvi, I.; Botke, I.; Chou, T.S.; Fernow, R.; Fischer, J.; Fisher, A.; Gallardo, J.; Ingold, G.; Malone, R.; Palmer, R.; Parsa, Z.; Pogorelsky, I.; Rogers, J.; Sheehan, J.; Srinivasan-Rao, T.; Tsang, T.; Ulc, S.; van Steenbergen, A.; Wang, X.J.; Woodle, M.; Yu, L.H.

    1992-01-01

    Early operation of the 50 MeV high brightness electron linac of the Accelerator Test Facility is described along with experimental data. This facility is designed to study new linear acceleration techniques and new radiation sources based on linacs in combination with free electron lasers. The accelerator utilizes a photo-excited, metal cathode, radio frequency electron gun followed by two travelling wave accelerating sections and an Experimental Hall for the study program

  11. The forced flow high field test facility SULTAN

    International Nuclear Information System (INIS)

    Horvath, I.; Vecsey, G.; Weymuth, P.

    1984-01-01

    The construction of the 8 Tesla, 1 m bore Test Facility SULTAN - I, a common action of ENEA (I-Frascati), ECN (NL-Petten) and SIN (CH-Villigen), is completed. Results on assembly, cooldown and the first operation of the whole system are presented. The SULTAN facility provides a wide range of capability of parameter variations (field, current, cooling) for the investigation of steady state performance and stability of technical superconductors unders nominal and limiting conditions

  12. Acceptance test procedure: RMW Land Disposal Facility Project W-025

    International Nuclear Information System (INIS)

    Roscha, V.

    1994-01-01

    This ATP establishes field testing procedures to demonstrate that the electrical/instrumentation system functions as intended by design for the Radioactive Mixed Waste Land Disposal Facility. Procedures are outlined for the field testing of the following: electrical heat trace system; transducers and meter/controllers; pumps; leachate storage tank; and building power and lighting

  13. 200 area effluent treatment facility opertaional test report

    International Nuclear Information System (INIS)

    Crane, A.F.

    1995-01-01

    This document reports the results of the 200 Area Effluent Treatment Facility (200 Area ETF) operational testing activities. These Operational testing activities demonstrated that the functional, operational and design requirements of the 200 Area ETF have been met and identified open items which require retesting

  14. Qualification program for JHR fuel elements: Irradiation of the first JHR test assembly in the BR2-Evita loop

    International Nuclear Information System (INIS)

    Anselmet, M.-C.; Lemoine, P.; Koonen, E.; Benoit, P.; Gouat, P.; Claes, W.; Geens, F.; Miras, G.; Brisson, S.

    2010-01-01

    An experimental program has been designed by CEA to qualify the behaviour of the JHR fuel under conditions representative of the reactor operating ones. This program uses the SCK.CEN facilities, irradiating JHR lead test elements in the BR2 reactor, inside its central channel which has been particularly arranged for this objective (Evita loop). As a first step in the program, a two cycle irradiation (4 weeks by cycle) started mid-July 2009 and ended mid-November (EVITA-1). After a cooling phase, this first JHR lead test element will be submitted to post-irradiation examination. The second JHR test element began its irradiation in the first quarter of 2010; its unloading is planned before the end of 2010, after 5 cycles in the BR2 reactor. The results of these two experiments are expected as input information for the Safety Authority Report. This paper presents the qualification program with the objectives assigned to each phase (irradiation, examination). A first interpretation of the irradiation data for the first element is presented, so as the information available on the progress of the following phases of the programme. (author)

  15. Test facility for the evaluation of microwave transmission components

    International Nuclear Information System (INIS)

    Fong, C.G.; Poole, B.R.

    1985-01-01

    A Low Power Test Facility (LPTF) was developed to evaluate the performance of Electron Cyclotron Resonance Heating (ECRH) microwave transmission components for the Mirror Fusion Test Facility (MFTF-B). The facility generates 26 to 60 GHz in modes of TE 01 , TE 02 , or TE 03 launched at power levels of 1/2 milliwatt. The propagation of the rf as it radiates from either transmitting or secondary reflecting microwave transmission components is recorded by a discriminating crystal detector mechanically manipulated at constant radius in spherical coordinates. The facility is used to test, calibrate, and verify the design of overmoded, circular waveguide components, quasi-optical reflecting elements before high power use. The test facility consists of microwave sources and metering components, such as VSWR, power and frequency meters, a rectangular TE 10 to circular TE 01 mode transducer, mode filter, circular TE 01 to 2.5 in. diameter overmoded waveguide with mode converters for combination of TE 01 to TE 03 modes. This assembly then connects to a circular waveguide launcher or the waveguide component under test

  16. Development of a fault test experimental facility model using Matlab

    Energy Technology Data Exchange (ETDEWEB)

    Pereira, Iraci Martinez; Moraes, Davi Almeida, E-mail: martinez@ipen.br, E-mail: dmoraes@dk8.com.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    The Fault Test Experimental Facility was developed to simulate a PWR nuclear power plant and is instrumented with temperature, level and pressure sensors. The Fault Test Experimental Facility can be operated to generate normal and fault data, and these failures can be added initially small, and their magnitude being increasing gradually. This work presents the Fault Test Experimental Facility model developed using the Matlab GUIDE (Graphical User Interface Development Environment) toolbox that consists of a set of functions designed to create interfaces in an easy and fast way. The system model is based on the mass and energy inventory balance equations. Physical as well as operational aspects are taken into consideration. The interface layout looks like a process flowchart and the user can set the input variables. Besides the normal operation conditions, there is the possibility to choose a faulty variable from a list. The program also allows the user to set the noise level for the input variables. Using the model, data were generated for different operational conditions, both under normal and fault conditions with different noise levels added to the input variables. Data generated by the model will be compared with Fault Test Experimental Facility data. The Fault Test Experimental Facility theoretical model results will be used for the development of a Monitoring and Fault Detection System. (author)

  17. Development of a fault test experimental facility model using Matlab

    International Nuclear Information System (INIS)

    Pereira, Iraci Martinez; Moraes, Davi Almeida

    2015-01-01

    The Fault Test Experimental Facility was developed to simulate a PWR nuclear power plant and is instrumented with temperature, level and pressure sensors. The Fault Test Experimental Facility can be operated to generate normal and fault data, and these failures can be added initially small, and their magnitude being increasing gradually. This work presents the Fault Test Experimental Facility model developed using the Matlab GUIDE (Graphical User Interface Development Environment) toolbox that consists of a set of functions designed to create interfaces in an easy and fast way. The system model is based on the mass and energy inventory balance equations. Physical as well as operational aspects are taken into consideration. The interface layout looks like a process flowchart and the user can set the input variables. Besides the normal operation conditions, there is the possibility to choose a faulty variable from a list. The program also allows the user to set the noise level for the input variables. Using the model, data were generated for different operational conditions, both under normal and fault conditions with different noise levels added to the input variables. Data generated by the model will be compared with Fault Test Experimental Facility data. The Fault Test Experimental Facility theoretical model results will be used for the development of a Monitoring and Fault Detection System. (author)

  18. The construction of solid waste form test and inspection facility

    International Nuclear Information System (INIS)

    Park, Hun Hwee; Lee, Kang Moo; Jung, In Ha; Kim, Sung Hwan; Yoo, Jeong Woo; Lee, Jong Youl; Bae, Sang Min

    1988-01-01

    The solid waste form test and inspection facility is a facility to test and inspect the characteristics of waste forms, such as homogenity, mechanical structure, thermal behaviour, water resistance and leachability. Such kinds of characteristics in waste forms are required to meet a certain conditions for long-term storage or for final disposal of wastes. The facility will be used to evaluate safety for the disposal of wastes by test and inspection. At this moment, the efforts to search the most effective management of the radioactive wastes generated from power plants and radioisotope user are being executed by the people related to this field. Therefore, the facility becomes more significant tool because of its guidance of sucessfully converting wastes into forms to give a credit to the safety of waste disposal for managing the radioactive wastes. In addition the overall technical standards for inspecting of waste forms such as the standardized equipment and processes in the facility will be estabilished in the begining of 1990's when the project of waste management will be on the stream. Some of the items of the project have been standardized for the purpose of localization. In future, this facility will be utilized not only for the inspection of waste forms but also for the periodic decontamination apparatus by remote operation techniques. (Author)

  19. Prediction of the Long Term Cooling Performance for the 3-Pin Fuel Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    Park, S. K.; Chi, D. Y.; Sim, B. S.; Park, K. N.; Ahn, S. H.; Lee, J. M.; Lee, C. Y.; Kim, H. R

    2005-12-15

    In the long term cooling phase that the emergency cooling water injection ends, the performance of the residual heat removal for the 3-pin fuel test loop has been predicted by a simplified heat transfer model. In the long term cooling phase the residual heat is 1323W for PWR fuel test mode and 1449W for CANDU fuel test mode. The each residual heat is assumed as 2% of the fission power of the test fuel used in the anticipated operational occurrence and design basis accident analyses. The each fission power used for the analyses is 105% of the rated fission power in the normal operation. In the long term cooling phase the residual heat is removed to the HANARO pool through the double pressure vessels of the in-pile test section. Saturate pooling boiling is assumed on the test fuel and condensation heat transfer is expected on the inner wall of the fuel carrier and the flow divider. Natural convection heat transfer on a heated vertical wall is also assumed on the outer wall of the outer pressure vessel. The conduction heat transfer is only considered in the gap between the double pressure vessels charged with neon gas and in the downcomer filled with coolant. The heat transfer rate between the coolant temperature of 152 .deg. C in the in-pile test section and the water temperature of 45 .deg. C in the HANARO pool is predicted as about 1666W. The 152 .deg. C is the saturate temperature of the coolant pressure predicted from the MARS code. The cooling capacity of 1666W is greater than the residual heats of 1323W and 1449W. Consequently the long term cooling performance of the 3-pin fuel test loop is sufficient for the anticipated operational occurrences and design basis accidents.

  20. Analysis of the LBLOCAs in the room 1 for the 3-pin fuel test loop

    International Nuclear Information System (INIS)

    Park, S. K.; Chi, D. Y.; Sim, B. S.; Park, K. N.; Ahn, S. H.; Lee, J. M.; Lee, C. Y.; Kim, Y. J.

    2004-12-01

    Fuel Test Loop(FTL) has been developed to meet the increasing demand on fuel irradiation and burn up test required the development of new fuels in Korea. It is designed to provide the test conditions of high pressure and temperature like the commercial PWR and CANDU power plants. And also the FTL have the cooling capability to sufficiently remove the thermal power of the in-pile test section for normal operation, Anticipated Operational Occurrences(AOOs), and Design Basis Accidents(DBAs). This report deals with the Large Break Loss of Coolant Accidents (LBLOCAs) in the Room 1 for the 3-pin fuel test loop. The MARS code has been used for the prediction of the emergency core cooling capability of the FTL and the peak cladding temperature of the test fuels for the LBLOCAs. The location of the pipe break is assumed at the downstream of the main cooling water pump and the upstream of the main cooler in the room 1. Double ended guillotine break is assumed for the large break loss of coolant accidents. The discharge coefficients of 0.1, 0.33, 0.67, 1.0 are investigated for the LBLOCAs. The maximum Peak Cladding Temperature (PCT) is predicted to be about 734.7 .deg. C for the PWR fuel test mode and 850.4 .deg. C for the CANDU fuel test mode respectively. The maximum peak cladding temperatures meet the design criterion of commercial PWR fuel that the maximum PCT is lower than 1204 .deg. C