A simple analytical scaling method for a scaled-down test facility simulating SB-LOCAs in a passive PWR

International Nuclear Information System (INIS)

Lee, Sang Il

1992-02-01

A Simple analytical scaling method is developed for a scaled-down test facility simulating SB-LOCAs in a passive PWR. The whole scenario of a SB-LOCA is divided into two phases on the basis of the pressure trend ; depressurization phase and pot-boiling phase. The pressure and the core mixture level are selected as the most critical parameters to be preserved between the prototype and the scaled-down model. In each phase the high important phenomena having the influence on the critical parameters are identified and the scaling parameters governing the high important phenomena are generated by the present method. To validate the model used, Marviken CFT and 336 rod bundle experiment are simulated. The models overpredict both the pressure and two phase mixture level, but it shows agreement at least qualitatively with experimental results. In order to validate whether the scaled-down model well represents the important phenomena, we simulate the nondimensional pressure response of a cold-leg 4-inch break transient for AP-600 and the scaled-down model. The results of the present method are in excellent agreement with those of AP-600. It can be concluded that the present method is suitable for scaling the test facility simulating SB-LOCAs in a passive PWR

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

Saltstone studies using the scaled continuous processing facility

Energy Technology Data Exchange (ETDEWEB)

Fowley, M. D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Cozzi, A. D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Hansen, E. K. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2015-08-01

The Savannah River National Laboratory (SRNL) has supported the Saltstone Facility since its conception with bench-scale laboratory experiments, mid-scale testing at vendor facilities, and consultations and testing at the Saltstone Facility. There have been minimal opportunities for the measurement of rheological properties of the grout slurry at the Saltstone Production Facility (SPF); thus, the Scaled Continuous Processing Facility (SCPF), constructed to provide processing data related to mixing, transfer, and other operations conducted in the SPF, is the most representative process data for determining the expected rheological properties in the SPF. These results can be used to verify the laboratory scale experiments that support the SPF using conventional mixing processes that appropriately represent the shear imparted to the slurry in the SPF.

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.

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...

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.

Scaling and design analyses of a scaled-down, high-temperature test facility for experimental investigation of the initial stages of a VHTR air-ingress accident

International Nuclear Information System (INIS)

Arcilesi, David J.; Ham, Tae Kyu; Kim, In Hun; Sun, Xiaodong; Christensen, Richard N.; Oh, Chang H.

2015-01-01

Highlights: • A 1/8th geometric-scale test facility that models the VHTR hot plenum is proposed. • Geometric scaling analysis is introduced for VHTR to analyze air-ingress accident. • Design calculations are performed to show that accident phenomenology is preserved. • Some analyses include time scale, hydraulic similarity and power scaling analysis. • Test facility has been constructed and shake-down tests are currently being carried out. - Abstract: A critical event in the safety analysis of the very high-temperature gas-cooled reactor (VHTR) is an air-ingress accident. This accident is initiated, in its worst case scenario, by a double-ended guillotine break of the coaxial cross vessel, which leads to a rapid reactor vessel depressurization. In a VHTR, the reactor vessel is located within a reactor cavity that is filled with air during normal operating conditions. Following the vessel depressurization, the dominant mode of ingress of an air–helium mixture into the reactor vessel will either be molecular diffusion or density-driven stratified flow. The mode of ingress is hypothesized to depend largely on the break conditions of the cross vessel. Since the time scales of these two ingress phenomena differ by orders of magnitude, it is imperative to understand under which conditions each of these mechanisms will dominate in the air ingress process. Computer models have been developed to analyze this type of accident scenario. There are, however, limited experimental data available to understand the phenomenology of the air-ingress accident and to validate these models. Therefore, there is a need to design and construct a scaled-down experimental test facility to simulate the air-ingress accident scenarios and to collect experimental data. The current paper focuses on the analyses performed for the design and operation of a 1/8th geometric scale (by height and diameter), high-temperature test facility. A geometric scaling analysis for the VHTR, a time

Real-time graphic display system for ROSA-V Large Scale Test Facility

International Nuclear Information System (INIS)

Kondo, Masaya; Anoda, Yoshinari; Osaki, Hideki; Kukita, Yutaka; Takigawa, Yoshio.

1993-11-01

A real-time graphic display system was developed for the ROSA-V Large Scale Test Facility (LSTF) experiments simulating accident management measures for prevention of severe core damage in pressurized water reactors (PWRs). The system works on an IBM workstation (Power Station RS/6000 model 560) and accommodates 512 channels out of about 2500 total measurements in the LSTF. It has three major functions: (a) displaying the coolant inventory distribution in the facility primary and secondary systems; (b) displaying the measured quantities at desired locations in the facility; and (c) displaying the time histories of measured quantities. The coolant inventory distribution is derived from differential pressure measurements along vertical sections and gamma-ray densitometer measurements for horizontal legs. The color display indicates liquid subcooling calculated from pressure and temperature at individual locations. (author)

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

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)

A scaling study of the natural circulation flow of the ex-vessel core catcher cooling system of a 1400MW PWR for designing a scale-down test facility

International Nuclear Information System (INIS)

Rhee, Bo. W.; Ha, K. S.; Park, R. J.; Song, J. H.

2012-01-01

A scaling study on the steady state natural circulation flow along the flow path of the ex-vessel core catcher cooling system of 1400MWe PWR is described. The scaling criteria for reproducing the same thermalhydraulic characteristics of the natural circulation flow as the prototype core catcher cooling system in the scale-down test facility is derived and the resulting natural circulation flow characteristics of the prototype and scale-down facility analyzed and compared. The purpose of this study is to apply the similarity law to the prototype EU-APR1400 core catcher cooling system and the model test facility of this prototype system and derive a relationship between the heating channel characteristics and the down-comer piping characteristics so as to determine the down-comer pipe size and the orifice size of the model test facility. As the geometry and the heating wall heat flux of the heating channel of the model test facility will be the same as those of the prototype core catcher cooling system except the width of the heating channel is reduced, the axial distribution of the coolant quality (or void fraction) is expected to resemble each other between the prototype and model facility. Thus using this fact, the down-comer piping design characteristics of the model facility can be determined from the relationship derived from the similarity law

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

Performance test results of mock-up model test facility with a full-scale reaction tube for HTTR hydrogen production system. Contract research

Energy Technology Data Exchange (ETDEWEB)

Inagaki, Yoshiyuki; Hayashi, Koji; Kato, Michio [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment] [and others

2003-03-01

Research on a hydrogen production system by steam reforming of methane, chemical reaction; CH{sub 4} + H{sub 2}O {yields} 3H{sub 2}O + CO, has been carried out to couple with the HTTR for establishment of high-temperature nuclear heat utilization technology and contribution to hydrogen energy society in future. The mock-up test facility with a full-scale reaction tube test facility, a model simulating one reaction tube of a steam reformer of the HTTR hydrogen production system in full scale, was fabricated to perform tests on controllability, hydrogen production performance etc. under the same pressure and temperature conditions as those of the HTTR hydrogen production system. The design and fabrication of the test facility started from 1997, and the all components were installed until September in 2001. In a performance test conducted from October in 2001 to February in 2002, performance of each component was examined and hydrogen of 120m{sup 3}{sub N}/h was successfully produced with high-temperature helium gas. This report describes the performance test results on components performance, hydrogen production characteristics etc., and main troubles and countermeasures. (author)

Performance test results of mock-up model test facility with a full-scale reaction tube for HTTR hydrogen production system. Contract research

International Nuclear Information System (INIS)

Inagaki, Yoshiyuki; Hayashi, Koji; Kato, Michio

2003-03-01

Research on a hydrogen production system by steam reforming of methane, chemical reaction; CH 4 + H 2 O → 3H 2 O + CO, has been carried out to couple with the HTTR for establishment of high-temperature nuclear heat utilization technology and contribution to hydrogen energy society in future. The mock-up test facility with a full-scale reaction tube test facility, a model simulating one reaction tube of a steam reformer of the HTTR hydrogen production system in full scale, was fabricated to perform tests on controllability, hydrogen production performance etc. under the same pressure and temperature conditions as those of the HTTR hydrogen production system. The design and fabrication of the test facility started from 1997, and the all components were installed until September in 2001. In a performance test conducted from October in 2001 to February in 2002, performance of each component was examined and hydrogen of 120m 3 N /h was successfully produced with high-temperature helium gas. This report describes the performance test results on components performance, hydrogen production characteristics etc., and main troubles and countermeasures. (author)

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

Test Plan for the Wake Steering Experiment at the Scaled Wind Farm Technology (SWiFT) Facility.

Energy Technology Data Exchange (ETDEWEB)

Naughton, Brian Thomas [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2017-02-01

This document is a test plan describing the objectives, configuration, procedures, reporting, roles, and responsibilities for conducting the joint Sandia National Laboratories and National Renewable Energy Laboratory Wake Steering Experiment at the Sandia Scaled Wind Farm Technology (SWiFT) facility near Lubbock, Texas in 2016 and 2017 . The purpose of this document is to ensure the test objectives and procedures are sufficiently detailed such that al l involved personnel are able to contribute to the technical success of the test. This document is not intended to address safety explicitly which is addressed in a separate document listed in the references titled Sandia SWiFT Facility Site Operations Manual . Both documents should be reviewed by all test personnel.

Full scaled tests of the KERENA trademark containment cooling condenser at the INKA test facility

International Nuclear Information System (INIS)

Leyer, Stephan; Maisberger, Fabian; Lineva, Natalia; Wagner, Thomas; Doll, Mathias; Herbst, Vasilli; Wich, Michael

2010-01-01

KERENA trademark is a medium-capacity boiling water reactor. It combines passive safety systems with active safety equipment of service-proven design. The passive systems utilize basic laws of physics, such as gravity and natural convection, enabling them to function without any power supply or actuation by instrumentation and control (I and C) equipment. They are designed to bring the plant to a safe and stable condition without the aid of active systems. Furthermore, the passive safety features partially replace the active systems, which reduces costs significantly and provides a safe, reliable and economically competitive plant design. At the new test facility at Karlstein called INKA (Integral Test Stand Karlstein), the key components of the KERENA trademark passive safety concept - the Emergency Condenser (EC), the Containment Cooling Condenser (CCC) and the passive core flooding system (PCFS) - are presently under full-scale testing,. Integral system tests will also be performed to show how the passive safety systems interact under various anticipated accident conditions and to demonstrate the ability of the passive systems to bring the plant to a safe and stable condition without the aid of active systems or actuation by I and C signals. The passive pressure pulse transmitter (PPPT) will be included in these integral tests. In this report the experimental setup and the first test results with the full scaled Containment Cooling Condenser will be described. (orig.)

Experimental results of the SMART ECC injection performance with reduced scale of test facility

Energy Technology Data Exchange (ETDEWEB)

Cho, Young Il; Cho, Seok; Ko, Yung Joo; Shin, Yong Cheol; Kwon, Tae Soon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2011-05-15

SMART pressurized water reactor type is different from the existing integral NSSS commercial pressurized water reactor system which is equipped with the main features. In addition, RCS piping is removed and the feature of the SBLOCA is a major design break accident. SWAT (SMART ECC Water Asymmetric Two-phase choking test facility) test facility is to simulate the 2 inch SBLOCA of the SMART using with reduced scale. The Test was performed to produce experimental data for the validation of the TASS/SMR-S thermal hydraulic analysis code, and to investigate the related thermal hydraulic phenomena in the down-comer region during the 2 inch SBLOCA of the safety inject line. The particular phenomena for the observation are ECC bypass and multi-dimensional flow characteristics to verify the effectiveness and performance of the safety injection system. In this paper, the corresponding steady state test conditions, including initial and boundary conditions along with major measuring parameters, and related experimental results were described

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

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.

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

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.)

Test design requirements: Canister-scale heater test

International Nuclear Information System (INIS)

Schauer, M.I.; Craig, P.A.; Stickney, R.G.

1986-03-01

This document establishes the Test Design Requirements for the design of a canister scale heater test to be performed in the Exploratory Shaft test facility. The purpose of the test is to obtain thermomechanical rock mass response data for use in validation of the numerical models. The canister scale heater test is a full scale simulation of a high-level nuclear waste container in a prototypic emplacement borehole. Electric heaters are used to simulate the heat loads expected in an actual waste container. This document presents an overview of the test including objectives and justification for the test. A description of the test as it is presently envisioned is included. Discussions on Quality Assurance and Safety are also included in the document. 12 refs., 1 fig

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.)

Counterpart experimental study of ISP-42 PANDA tests on PUMA facility

International Nuclear Information System (INIS)

Yang, Jun; Choi, Sung-Won; Lim, Jaehyok; Lee, Doo-Yong; Rassame, Somboon; Hibiki, Takashi; Ishii, Mamoru

2013-01-01

Highlights: ► Counterpart tests were performed on two large-scale BWR integral facilities. ► Similarity of post-LOCA system behaviors observed between two tests. ► Passive core and containment cooling systems work as design in both tests. -- Abstract: A counterpart test to the Passive Nachwärmeabfuhr und Druckabbau Test Anlage (Passive Decay Heat Removal and Depressurization Test Facility, PANDA) International Standard Problem (ISP)-42 test was conducted at the Purdue University Multi-Dimensional Integral Test Assembly (PUMA) facility. Aimed to support code validation on a range of light water reactor (LWR) containment issues, the ISP-42 test consists of six sequential phases (Phases A–F) with separately defined initial and boundary conditions, addressing different stages of anticipated accident scenario and system responses. The counterpart test was performed from Phases A to D, which are within the scope of the normal integral tests performed on the PUMA facility. A scaling methodology was developed by using the PANDA facility as prototype and PUMA facility as test model, and an engineering scaling has been applied to the PUMA facility. The counterpart test results indicated that functions of passive safety systems, such as passive containment cooling system (PCCS) start-up, gravity-driven cooling system (GDCS) discharge, PCCS normal operation and overload function were confirmed in both the PANDA and PUMA facilities with qualitative similarities

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

A testing facility for large scale models at 100 bar and 3000C to 10000C

International Nuclear Information System (INIS)

Zemann, H.

1978-07-01

A testing facility for large scale model tests is in construction under support of the Austrian Industry. It will contain a Prestressed Concrete Pressure Vessel (PCPV) with hot linear (300 0 C at 100 bar), an electrical heating system (1.2 MW, 1000 0 C), a gas supply system, and a cooling system for the testing space. The components themselves are models for advanced high temperature applications. The first main component which was tested successfully was the PCPV. Basic investigation of the building materials, improvements of concrete gauges, large scale model tests and measurements within the structural concrete and on the liner from the beginning of construction during the period of prestressing, the period of stabilization and the final pressurizing tests have been made. On the basis of these investigations a computer controlled safety surveillance system for long term high pressure, high temperature tests has been developed. (author)

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

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

Remote-handling demonstration tests for the Fusion Materials Irradiation Test (FMIT) Facility

International Nuclear Information System (INIS)

Shen, E.J.; Hussey, M.W.; Kelly, V.P.; Yount, J.A.

1982-01-01

The mission of the Fusion Materials Irradiation Test (FMIT) Facility is to create a fusion-like environment for fusion materials development. Crucial to the success of FMIT is the development and testing of remote handling systems required to handle materials specimens and maintenance of the facility. The use of full scale mock-ups for demonstration tests provides the means for proving these systems

Large scale FCI experiments in subassembly geometry. Test facility and model experiments

International Nuclear Information System (INIS)

Beutel, H.; Gast, K.

A program is outlined for the study of fuel/coolant interaction under SNR conditions. The program consists of a) under water explosion experiments with full size models of the SNR-core, in which the fuel/coolant system is simulated by a pyrotechnic mixture. b) large scale fuel/coolant interaction experiments with up to 5kg of molten UO 2 interacting with liquid sodium at 300 deg C to 600 deg C in a highly instrumented test facility simulating an SNR subassembly. The experimental results will be compared to theoretical models under development at Karlsruhe. Commencement of the experiments is expected for the beginning of 1975

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

Air pollution control system testing at the DOE offgas components test facility

International Nuclear Information System (INIS)

Burns, D.B.; Speed, D.; VanPelt, W.; Burns, H.H.

1997-01-01

In 1997, the Department of Energy (DOE) Savannah River Site (SRS) plans to begin operation of the Consolidated Incineration Facility (CIF) to treat solid and liquid RCRA hazardous and mixed wastes. The Savannah River Technology Center (SRTC) leads an extensive technical support program designed to obtain incinerator and air pollution control equipment performance data to support facility start-up and operation. A key component of this technical support program includes the Offgas Components Test Facility (OCTF), a pilot-scale offgas system test bed. The primary goal for this test facility is to demonstrate and evaluate the performance of the planned CIF Air Pollution Control System (APCS). To accomplish this task, the OCTF has been equipped with a 1/10 scale CIF offgas system equipment components and instrumentation. In addition, the OCTF design maximizes the flexibility of APCS operation and facility instrumentation and sampling capabilities permit accurate characterization of all process streams throughout the facility. This allows APCS equipment performance to be evaluated in an integrated system under a wide range of possible operating conditions. This paper summarizes the use of this DOE test facility to successfully demonstrate APCS operability and maintainability, evaluate and optimize equipment and instrument performance, and provide direct CIF start-up support. These types of facilities are needed to permit resolution of technical issues associated with design and operation of systems that treat and dispose combustible hazardous, mixed, and low-level radioactive waste throughout and DOE complex

Development of a test facility for analyzing transients in supercritical water-cooled reactors by fractional scaling analysis

Energy Technology Data Exchange (ETDEWEB)

Roberto, Thiago D., E-mail: thiagodbtr@gmail.com [Instituto de Engenharia Nuclear (IEN/CNEN—RJ), Rua Hélio de Almeida, 75 21941-972, Rio de Janeiro Caixa-Postal: 68550, RJ (Brazil); Silva, Mário A. B. da, E-mail: mabs500@gmail.com [Departamento de Energia Nuclear (CTG/UFPE), Av. Professor Luiz Freire, 1000, Recife 50740-540, PE (Brazil); Lapa, Celso M.F., E-mail: lapa@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN—RJ), Rua Hélio de Almeida, 75 21941-972, Rio de Janeiro Caixa-Postal: 68550, RJ (Brazil)

2016-01-15

The feasibility of performing experiments using water under supercritical conditions is limited by technical and financial difficulties. These difficulties can be overcome by using model fluids that are characterized by feasible supercritical conditions, that is, lower critical pressure and critical temperature. Experimental investigations are normally used to determine the conditions under which model fluids reliably represent supercritical fluids under steady-state conditions. A fluid-to-fluid scaling approach has been proposed to determine the model fluids that represent supercritical fluids in a transient state. Recently, a similar technique known as fractional scaling analysis was developed to establish the conditions under which experiments can be performed using models that represent transients in prototypes. This paper presents a fractional scaling analysis application to determine parameters for a test facility in which transient conditions in supercritical water-cooled reactors are simulated by using carbon dioxide as a model fluid, whose critical point conditions are more feasible than those of water. Similarity is obtained between water (prototype) and carbon dioxide (model) by depressurization in a simple vessel. The main parameters required for the construction of a future test facility are obtained using the proposed method.

Development of a test facility for analyzing transients in supercritical water-cooled reactors by fractional scaling analysis

International Nuclear Information System (INIS)

Roberto, Thiago D.; Silva, Mário A. B. da; Lapa, Celso M.F.

2016-01-01

The feasibility of performing experiments using water under supercritical conditions is limited by technical and financial difficulties. These difficulties can be overcome by using model fluids that are characterized by feasible supercritical conditions, that is, lower critical pressure and critical temperature. Experimental investigations are normally used to determine the conditions under which model fluids reliably represent supercritical fluids under steady-state conditions. A fluid-to-fluid scaling approach has been proposed to determine the model fluids that represent supercritical fluids in a transient state. Recently, a similar technique known as fractional scaling analysis was developed to establish the conditions under which experiments can be performed using models that represent transients in prototypes. This paper presents a fractional scaling analysis application to determine parameters for a test facility in which transient conditions in supercritical water-cooled reactors are simulated by using carbon dioxide as a model fluid, whose critical point conditions are more feasible than those of water. Similarity is obtained between water (prototype) and carbon dioxide (model) by depressurization in a simple vessel. The main parameters required for the construction of a future test facility are obtained using the proposed method.

Preparation for Scaling Studies of Ice-Crystal Icing at the NRC Research Altitude Test Facility

Science.gov (United States)

Struk, Peter M.; Bencic, Timothy J.; Tsao, Jen-Ching; Fuleki, Dan; Knezevici, Daniel C.

2013-01-01

This paper describes experiments conducted at the National Research Council (NRC) of Canadas Research Altitiude Test Facility between March 26 and April 11, 2012. The tests, conducted collaboratively between NASA and NRC, focus on three key aspects in preparation for later scaling work to be conducted with a NACA 0012 airfoil model in the NRC Cascade rig: (1) cloud characterization, (2) scaling model development, and (3) ice-shape profile measurements. Regarding cloud characterization, the experiments focus on particle spectra measurements using two shadowgraphy methods, cloud uniformity via particle scattering from a laser sheet, and characterization of the SEA Multi-Element probe. Overviews of each aspect as well as detailed information on the diagnostic method are presented. Select results from the measurements and interpretation are presented which will help guide future work.

Full-scale mark II CRT program facility description report

International Nuclear Information System (INIS)

Namatame, Ken; Kukita, Yutaka; Ito, Hideo; Yamamoto, Nobuo; Shiba, Masayoshi

1980-03-01

Started in fiscal year 1977, the Full-Scale Mark II CRT (Containment Response Test) Program is proceeding for the period of five years. The primary objective of the CRT Program is to provide a data base for evaluation of the pressure suppression pool hydrodynamic loads associated with a postulated loss-of-coolant accident in the BWR Mark II containment system. The test facility was designed and constructed from fiscal year 1977 to 1978, and completed in March 1979. It is 1/18 in volume and has a wetwell which is a full-scale replica of one 20 0 -sector of that of a reference Mark II. This report describes design concepts, dimensions and constructions of the test facility, as well as specifications, locations and installation schemes of the measuring equipments. Results of soil structure inspection, vacuum breaker test and shaker test of the containment shell are given in the appendices. (author)

Pilot-scale reverse osmosis testing for the F and H Area Effluent Treatment Facility

International Nuclear Information System (INIS)

Kessler, J.L.

1984-01-01

Pilot-scale reverse osmosis (RO) tests were completed with a 10 gpm unit to demonstrate the performance of RO in the F and H Area Effluent Treatment Facility (F/H ETF). RO will be used in the WMETF to remove soluble salts and soluble radioactivity. The advantage of using RO (over ion exchange) is that it is nondescriminanting and removes virtually all dissolved solids species, regardless of ionic charge. RO also generates less than half the waste volume produced by ion exchange. Test results using a 200-Area nonradioactive effluent simulant demonstrated salt rejections of 98% and water recoveries of 94% by using recycle on a single stage pilot unit. For a full-scale, multi-staged unit overall salt rejections will be 95% (DF = 20) while obtaining a 94% water recovery (94% discharge, 6% concentrated waste stream). Identical performance is expected on actual radioactive streams, based on shielded cells testing performed by Motyka and Stimson. Similarly, if the WMETF RO system is configured in the same manner as the SRL ECWPF, a DF of 20 and a water recvery of 94% should be obtained

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

Performance test results of mock-up test facility of HTTR hydrogen production system

International Nuclear Information System (INIS)

Ohashi, Hirofumi; Inaba, Yoshitomo; Nishihara, Tetsuo

2004-01-01

For the purpose to demonstrate effectiveness of high-temperature nuclear heat utilization, Japan Atomic Energy Research Institute has been developing a hydrogen production system and has planned to connect the hydrogen production system to High Temperature Engineering Test Reactor (HTTR). Prior to construction of a HTTR hydrogen production system, a mock-up test facility was constructed to investigate transient behavior of the hydrogen production system and to establish system controllability. The Mock-up test facility with a full-scale reaction tube is an approximately 1/30-scale model of the HTTR hydrogen production system and an electric heater is used as a heat source instead of a reactor. After its construction, a performance test of the test facility was carried out in the same pressure and temperature conditions as those of the HTTR hydrogen production system to investigate its performance such as hydrogen production ability, controllability and so on. It was confirmed that hydrogen was stably produced with a hot helium gas about 120m 3 /h, which satisfy the design value, and thermal disturbance of helium gas during the start-up could be mitigated within the design value by using a steam generator. The mock-up test of the HTTR hydrogen production system using this facility will continue until 2004. (author)

Transition from depressurization to long term cooling in AP600 scaled integral test facilities

International Nuclear Information System (INIS)

Bessette, D.E.; Marzo, M. di

1999-01-01

A novel light water reactor design called the AP600 has been proposed by the Westinghouse Electric Corporation. In the evaluation of this plant's behavior during a small break loss of coolant accident (LOCA), the crucial transition to low pressure, long-term cooling is marked by the injection of the gravitationally driven flow from the in-containment refueling water storage tank (IRWST). The onset of this injection is characterized by intermittency in the IRWST flow. This happens at a time when the reactor vessel reaches its minimum inventory. Therefore, it is important to understand and scale the behavior of the integral experimental test facilities during this portion of the transient. The explanation is that the periodic liquid drains and refills of the pressurizer are the reason for the intermittent behavior. The momentum balance for the surge line yields the nondimensional parameter controlling this process. Data from one of the three experimental facilities represent the phenomena well at the prototypical scale. The impact of the intermittent IRWST injection on the safe plant operation is assessed and its implications are successfully resolved. The oscillation is found to result from, in effect, excess water in the primary system and it is not of safety significance. (orig.)

WTP Pilot-Scale Evaporation Tests

International Nuclear Information System (INIS)

QURESHI, ZAFAR

2004-01-01

This report documents the design, assembly, and operation of a Pilot-Scale Evaporator built and operated by SRTC in support of Waste Treatment Plant (WTP) Project at the DOE's Hanford Site. The WTP employs three identical evaporators, two for the Waste Feed and one for the Treated LAW. The Pilot-Scale Evaporator was designed to test simulants for both of these waste streams. The Pilot-Scale Evaporator is 1/76th scale in terms of evaporation rates. The basic configuration of forced circulation vacuum evaporator was employed. A detailed scaling analysis was performed to preserve key operating parameters such as basic loop configuration, system vacuum, boiling temperature, recirculation rates, vertical distances between important hardware pieces, reboiler heat transfer characteristics, vapor flux, configuration of demisters and water spray rings. Three evaporation test campaigns were completed. The first evaporation run used water in order to shake down the system. The water runs were important in identifying a design flaw that inhibited mixing in the evaporator vessel, thus resulting in unstable boiling operation. As a result the loop configuration was modified and the remaining runs were completed successfully. Two simulant runs followed the water runs. Test 1: Simulated Ultrafiltration Recycles with HLW SBS, and Test 2: Treated AN102 with Envelop C LAW. Several liquid and offgas samples were drawn from the evaporator facility for regulatory and non-regulatory analyses. During Test 2, the feed and the concentrate were spiked with organics to determine organic partitioning. The decontamination factor (DF) for Test 1 was measured to be 110,000 (more than the expected value of 100,000). Dow Corning Q2-3183A antifoam agent was tested during both Tests 1 and 2. It was determined that 500 ppm of this antifoam agent was sufficient to control the foaminess to less than 5 per cent of the liquid height. The long-term testing (around 100 hours of operation) did not show any

Passive BWR integral LOCA testing at the Karlstein test facility INKA

Energy Technology Data Exchange (ETDEWEB)

Drescher, Robert [AREVA GmbH, Erlangen (Germany); Wagner, Thomas [AREVA GmbH, Karlstein am Main (Germany); Leyer, Stephan [TH University of Applied Sciences, Deggendorf (Germany)

2014-05-15

KERENA is an innovative AREVA GmbH boiling water reactor (BWR) with passive safety systems (Generation III+). In order to verify the functionality of the reactor design an experimental validation program was executed. Therefore the INKA (Integral Teststand Karlstein) test facility was designed and erected. It is a mockup of the BWR containment, with integrated pressure suppression system. While the scaling of the passive components and the levels match the original values, the volume scaling of the containment compartments is approximately 1:24. The storage capacity of the test facility pressure vessel corresponds to approximately 1/6 of the KERENA RPV and is supplied by a benson boiler with a thermal power of 22 MW. In March 2013 the first integral test - Main Steam Line Break (MSLB) - was executed. The test measured the combined response of the passive safety systems to the postulated initiating event. The main goal was to demonstrate the ability of the passive systems to ensure core coverage, decay heat removal and to maintain the containment within defined limits. The results of the test showed that the passive safety systems are capable to bring the plant to stable conditions meeting all required safety targets with sufficient margins. Therefore the test verified the function of those components and the interplay between them. The test proved that INKA is an unique test facility, capable to perform integral tests of passive safety concepts under plant-like conditions. (orig.)

Testing of components on the shaking table facilities of AEP and contribution to full scale dynamic testing of Kozloduy NPP. Final report

International Nuclear Information System (INIS)

Ambriashvili, Y.

1996-01-01

This final report summarizes the results of components testing on the shaking table facilities of 'Atomenergoproject' which are considered as a contribution to the full scale dynamic testing of the Kozloduy nuclear power plant Units 5 and 6. It was designed on 1.0 g according to the calculations that were based on accelerograms which included artificial and already known recordings of real earthquakes. Maximum acceleration of the designed spectrum and new spectrum which are recommended are now within the range of frequencies 2.5-20 Hz. Active reactor and the primary loop are seismic stable as well as the tested equipment tested by 'Atomenergoproject'

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.

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.

Pilot-scale ultrafiltration testing for the F and H area effluent treatment facility

International Nuclear Information System (INIS)

Kessler, J.L.

1984-01-01

An F and H Area Effluent Treatment Facility (F/H ETF) is being designed to treat low activity aqueous effluents which are produced from F and H Area daily operations. The treatment scheme for the F/H ETF will include pretreatment (pH adjustment and filtration) followed by Reverse Osmosis and/or Ion Exchange to remove dissolved species. Several alternative treatment processes are being considered for the F/H ETF. One of the alternatives in the pretreatment step is tubular Ultrafiltration (UF), using a dynamically formed zirconium oxide membrane supported on a porous stainless steel backing. Pilot-scale testing with a single membrane module (13 ft 2 area) and 200-Area effluent simulant has demonstrated that UF is a viable filtration option for the F/H ETF. UF testing at TNX has defined the operating conditions necessary for extended operation and also demonstrated excellent filtration performance (filtrate SDI 2 /day) flux and will provide excellent pretreatment for both reverse osmosis and ion exchange. 2 refs

Preliminary conceptual study of engineering-scale pyroprocess demonstration facility

International Nuclear Information System (INIS)

Moon, Seong-In; Chong, Won-Myung; You, Gil-Sung; Ku, Jeong-Hoe; Kim, Ho-Dong

2013-01-01

Highlights: ► The conceptual design of a pyroprocess demonstration facility was performed. ► The design requirements for the pyroprocess hot cell and equipment were determined. ► The maintenance concept for the pyroprocess hot cell was presented. -- Abstract: The development of an effective management technology of spent fuel is important to enhance environmental friendliness, cost viability and proliferation resistance. In Korea, pyroprocess technology has been considered as a fuel cycle option to solve the spent fuel accumulation problems. PRIDE (PyRoprocess Integrated inactive DEmonstration facility) has been developed from 2007 to 2012 in Korea as a cold test facility to support integrated pyroprocessing and an equipment demonstration, which is essential to verify the pyroprocess technology. As the next stage of PRIDE, the design requirements of an engineering-scale demonstration facility are being developed, and the preliminary conceptual design of the facility is being performed for the future. In this paper, the main design requirements for the engineering-scale pyroprocess demonstration facility were studied in the throughput of 10tHM a year. For the preliminary conceptual design of the facility, the design basis of the pyroprocess hot cell was suggested, and the main equipment, main process area, operation area, maintenance area, and so on were arranged in consideration of the effective operation of the hot cells. Also, the argon system was designed to provide and maintain a proper inert environment for the pyroprocess. The preliminary conceptual design data will be used to review the validity of the engineering-scale pyroprocess demonstration facility that enhances both safety and nonproliferation

ROSA-V large scale test facility (LSTF) system description for the third and fourth simulated fuel assemblies

Energy Technology Data Exchange (ETDEWEB)

Suzuki, Mitsuhiro; Nakamura, Hideo; Ohtsu, Iwao [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

2003-03-01

The Large Scale Test Facility (LSTF) is a full-height and 1/48 volumetrically scaled test facility of the Japan Atomic Energy Research Institute (JAERI) for system integral experiments simulating the thermal-hydraulic responses at full-pressure conditions of a 1100 MWe-class pressurized water reactor (PWR) during small break loss-of-coolant accidents (SBLOCAs) and other transients. The LSTF can also simulate well a next-generation type PWR such as the AP600 reactor. In the fifth phase of the Rig-of-Safety Assessment (ROSA-V) Program, eighty nine experiments have been conducted at the LSTF with the third simulated fuel assembly until June 2001, and five experiments have been conducted with the newly-installed fourth simulated fuel assembly until December 2002. In the ROSA-V program, various system integral experiments have been conducted to certify effectiveness of both accident management (AM) measures in beyond design basis accidents (BDBAs) and improved safety systems in the next-generation reactors. In addition, various separate-effect tests have been conducted to verify and develop computer codes and analytical models to predict non-homogeneous and multi-dimensional phenomena such as heat transfer across the steam generator U-tubes under the presence of non-condensable gases in both current and next-generation reactors. This report presents detailed information of the LSTF system with the third and fourth simulated fuel assemblies for the aid of experiment planning and analyses of experiment results. (author)

Large scale sodium-water reaction tests for Monju steam generators

International Nuclear Information System (INIS)

Sato, M.; Hiroi, H.; Hori, M.

1976-01-01

To demonstrate the safe design of the steam generator system of the prototype fast reactor Monju against the postulated large leak sodium-water reaction, a large scale test facility SWAT-3 was constructed. SWAT-3 is a 1/2.5 scale model of the Monju secondary loop on the basis of the iso-velocity modeling. Two tests have been conducted in SWAT-3 since its construction. The test items using SWAT-3 are discussed, and the description of the facility and the test results are presented

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...

Sampling based uncertainty analysis of 10% hot leg break LOCA in large scale test facility

International Nuclear Information System (INIS)

Sengupta, Samiran; Kraina, V.; Dubey, S. K.; Rao, R. S.; Gupta, S. K.

2010-01-01

Sampling based uncertainty analysis was carried out to quantify uncertainty in predictions of best estimate code RELAP5/MOD3.2 for a thermal hydraulic test (10% hot leg break LOCA) performed in the Large Scale Test Facility (LSTF) as a part of an IAEA coordinated research project. The nodalisation of the test facility was qualified for both steady state and transient level by systematically applying the procedures led by uncertainty methodology based on accuracy extrapolation (UMAE); uncertainty analysis was carried out using the Latin hypercube sampling (LHS) method to evaluate uncertainty for ten input parameters. Sixteen output parameters were selected for uncertainty evaluation and uncertainty band between 5 th and 95 th percentile of the output parameters were evaluated. It was observed that the uncertainty band for the primary pressure during two phase blowdown is larger than that of the remaining period. Similarly, a larger uncertainty band is observed relating to accumulator injection flow during reflood phase. Importance analysis was also carried out and standard rank regression coefficients were computed to quantify the effect of each individual input parameter on output parameters. It was observed that the break discharge coefficient is the most important uncertain parameter relating to the prediction of all the primary side parameters and that the steam generator (SG) relief pressure setting is the most important parameter in predicting the SG secondary pressure

Improvement works report on mock-up model test facility with a full-scale reaction tube for HTTR hydrogen production system (Contract research)

International Nuclear Information System (INIS)

Sakaki, Akihiro; Kato, Michio; Hayashi, Koji; Fujisaki, Katsuo; Aita, Hideki; Ohashi, Hirofumi; Takada, Shoji; Shimizu, Akira; Morisaki, Norihiro; Maeda, Yukimasa; Sato, Hiroyuki; Hanawa, Hiromi; Yonekawa, Hideo; Inagaki, Yoshiyuki

2005-04-01

In order to establish the system integration technology to connect a hydrogen production system to a high temperature gas cooled reactor; the mock-up test facility with a full-scale reaction tube for the steam reforming HTTR hydrogen production system was constructed in fiscal year 2001 and its functional test operation was performed in the year. Seven experimental test operations were performed from fiscal year 2001 to 2004. On a period of each test operation, there happened some troubles. For each trouble, the cause was investigated and the countermeasures and the improvement works were performed to succeed the experiments. The tests were successfully achieved according to plan. This report describes the improvement works on the test facility performed from fiscal year 2001 to 2004. (author)

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

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

Outcomes of Domestic Standard Problem-03 : Scaling Capability of Facility Data

Energy Technology Data Exchange (ETDEWEB)

Park, Yusun; Youn, Bumsu; Lee, Seung-won; Kim, Won-tae; Kang, Kyoung-ho; Choi, Ki-yong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

The Domestic Standard Problem (DSP) previous two DSPs provided good research opportunities to many nuclear organizations to understand the capability of the current system-scale safety analysis codes and to find a way for further code development area. Thus, the third DSP program was launched in the 2nd half of the year 2012. As the third DSP exercise (DSP-03), a double-ended guillotine break of the main steam line at an 8% power without loss of off-site power (LOOP) was decided a target scenario. Seventeen domestic organizations joined this DSP exercise. This DSP exercise was performed in an open calculation environment similar to the previous ones. In the present DSP-03, taking into accounts the different levels of code experiences and expertise, three sub-topics were suggested by operating agency. Among them, the investigation on scaling capability of facility data which was the topic of Group A will be discussed in this paper. Agreed participants should perform two calculations with the ATLAS model and the APR1400 model. By comparing major and detailed local parameters from both calculation models, scaling capability of the facility data was investigated. The 38.6 mm MSLB in ATLAS test facility was calculated using SPACE and MARS-KS code. To analyze the effect of scaling on the system behavior, MSLB in APR-1400 was also simulated in this study and following results were obtained. - The code predicted appropriately the overall MSLB experimental data obtained from ATLAS test facility. - The break flow calculated by code was lower than that of experimental data. - And the difference between calculated value and measured value was attributed to the measurement of mass from break flow. - The temperatures of core inlet and outlet of ATLAS test facility were predicted lower than those of experimental data.

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.

Production of U3O8 by uranyl formate precipitation and calcination in a full-scale pilot facility

International Nuclear Information System (INIS)

Kendrick, L.S.; Wilson, W.A.; Mosley, W.C.

1984-08-01

The uranyl formate process for the production of U 3 O 8 with a controlled particle size has been extensively studied on a laboratory scale. Based on this study, a pilot-scale facility (the Uranyl Formate Facility) was built to investigate the key steps of the process on a larger scale. These steps were the precipitation of a uranyl formate monohydrate salt and the calcination of this salt to U 3 O 8 . Tests of the facility and process were conducted at conditions recommended by the laboratory-scale studies for a full-scale production facility. These tests demonstrated that U 3 O 8 of the required particle size for the PM process can be produced on a plant scale by the calcination of uranyl formate crystals. The performance of the U 3 O 8 produced by the uranyl formate process in fuel tube fabrication was also investigated. Small-scale extrusion tests of U 3 O 8 -Al cores which used the U 3 O 8 produced in the Uranyl Formate Facility were conducted. These tests demonstrated that the U 3 O 8 quality was satisfactory for the PM process

Validation of a methodology to develop a test facility in reduced scale related to boron dispersion in a pressurizer of an iPWR

Energy Technology Data Exchange (ETDEWEB)

Nascimento, Samira R.V.; Lira, Carlos A.B.O.; Lapa, Celso M.F.; Lima, Fernando R.A.; Bezerra, Jair L.; Silva, Mário A.B., E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Centro de Tecnologia e Geociências. Departamento de Energia Nuclear; Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil); Centro Regional de Ciências Nucleares do Nordeste (CRCN/CNEN-PE), Recife, PE (Brazil)

2017-11-01

The conception and the project of a 1:200 reduced scale test facility have been developed in earlier researches. Such a facility aims to investigate boron homogenization process inside the pressurizer of an iPWR (integral PWR) by considering water mixing from this component with that coming from the reactor core. For this kind of reactor, the pressurizer is located at the top of the pressure vessel demanding the need of identifying the proper mechanisms in order to warrant an adequate homogenization for the water mixture. Once the installation of the experimental setup was concluded, its behavior has been analyzed by considering the concentration of a tracer diluted in the circulation water, whose measurements were obtained at the pressurizer outlet orifices. Two experiments representing boration(boron concentration increase)/deboration(boron concentration decrease) scenarios have been accomplished. Sample acquisition was carried out for every ten minutes during a total time equal to 180 minutes. Results showed that the combination of Fractional Scaling Analysis with local Froude number consisted of an appropriate methodology to provide the reduced scale test facility parameters, inasmuch the measured concentrations from the experiments reproduced the theoretical behavior with sufficient accuracy. (author)

Validation of a methodology to develop a test facility in reduced scale related to boron dispersion in a pressurizer of an iPWR

International Nuclear Information System (INIS)

Nascimento, Samira R.V.; Lira, Carlos A.B.O.; Lapa, Celso M.F.; Lima, Fernando R.A.; Bezerra, Jair L.; Silva, Mário A.B.

2017-01-01

The conception and the project of a 1:200 reduced scale test facility have been developed in earlier researches. Such a facility aims to investigate boron homogenization process inside the pressurizer of an iPWR (integral PWR) by considering water mixing from this component with that coming from the reactor core. For this kind of reactor, the pressurizer is located at the top of the pressure vessel demanding the need of identifying the proper mechanisms in order to warrant an adequate homogenization for the water mixture. Once the installation of the experimental setup was concluded, its behavior has been analyzed by considering the concentration of a tracer diluted in the circulation water, whose measurements were obtained at the pressurizer outlet orifices. Two experiments representing boration(boron concentration increase)/deboration(boron concentration decrease) scenarios have been accomplished. Sample acquisition was carried out for every ten minutes during a total time equal to 180 minutes. Results showed that the combination of Fractional Scaling Analysis with local Froude number consisted of an appropriate methodology to provide the reduced scale test facility parameters, inasmuch the measured concentrations from the experiments reproduced the theoretical behavior with sufficient accuracy. (author)

Test on large-scale seismic isolation elements

International Nuclear Information System (INIS)

Mazda, T.; Shiojiri, H.; Oka, Y.; Fujita, T.; Seki, M.

1989-01-01

Demonstration test of seismic isolation elements is considered as one of the most important items in the application of seismic isolation system to fast breeder reactor (FBR) plant. Facilities for testing seismic isolation elements have been built. This paper reports on tests for fullscale laminated rubber bearing and reduced scale models are conducted. From the result of the tests, the laminated rubber bearings turn out to satisfy the specification. Their basic characteristics are confirmed from the tests with fullscale and reduced scale models. The ultimate capacity of the bearings under the condition of ordinary temperature are evaluated

Introduction to flow visualization system in SPARC test facility

International Nuclear Information System (INIS)

Lee, Wooyoung; Song, Simon; Na, Young Su; Hong, Seong Wan

2016-01-01

The released hydrogen can be accumulated and mixed by steam and air depending on containment conditions under severe accident, which generates flammable mixture. Hydrogen explosion induced by ignition source cause severe damage to a structure or facility. Hydrogen risk regarding mixing, distribution, and combustion has been identified by several expert groups and studied actively since TMI accident. A large-scale thermal-hydraulic experimental facility is required to simulate the complex severe accident phenomena in the containment building. We have prepared the test facility, called the SPARC (Spray, Aerosol, Recombiner, Combustion), to resolve the international open issues regarding hydrogen risk. Gas mixing and stratification test using helium instead of hydrogen and estimation of a stratification surface erosion of helium owing to the vertical jet flow will be performed in SPARC. The measurement system is need to observe the gas flow in the large scale test facility such as SPARC. The PIV (particle image velocimetry) system have been installed to visualize gas flow. We are preparing the test facility, called the SPARC, for estimation the thermal-hydraulic process of hydrogen in a closed containment building and the PIV system for quantitative assessment of gas flow. In particular, we will perform gas mixing and erosion of stratification surface test using helium which is the replacement of hydrogen. It will be evaluated by measuring 2D velocity field using the PIV system. The PIV system mainly consists of camera, laser and tracer particle. Expected maximum size of FOV is 750 x 750 mm 2 limited by focal length of lens and high power laser corresponding to 425mJ/pulse at 532 wavelength is required due to large FOV

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.

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.)

Chemical Transfer (Single Small-Scale) Facility

Data.gov (United States)

Federal Laboratory Consortium — Description/History: Chemistry laboratoryThe Chemical Transfer Facility (CTF)  is the only U.S. single small-scale  facility, a single repository for the Army’s...

106-AN grout pilot-scale test HGTP-93-0501-02

International Nuclear Information System (INIS)

Bagaasen, L.M.

1993-05-01

The Grout Treatment Facility (GTF) at Hanford, Washington will process the low-level fraction of selected double-shell tank (DST) wastes into a cementitious waste form. This facility, which is operated by Westinghouse Hanford Company (WHC), mixes liquid waste with cementitious materials to produce a waste form that immobilizes hazardous constituents through chemical reactions and/or microencapsulation. Over 1,000,000 gal of Phosphate/Sulfate Waste were solidified in the first production campaign with this facility. The next tank scheduled for treatment is 106-AN. After conducting laboratory studies to select the grout formulation, part of the normal formulation verification process is to conduct tests using the 1/4-scale pilot facilities at the Pacific Northwest Laboratory (PNL). The major objectives of these pilot-scale tests were to determine if the proposed grout formulation could be processed in the pilot-scale equipment and to collect thermal information to help determine the best way to manage the grout hydration heat

Heater test planning for the near surface test facility at the Hanford reservation

International Nuclear Information System (INIS)

DuBois, A.; Binnall, E.; Chan, T.; McEvoy, M.; Nelson, P.; Remer, J.

1979-03-01

The underground test facility NSTF being constructed at Gable Mountain, is the site for a group of experiments designed to evaluate the thermo-mechanical suitability of a deep basalt stratum as a permanent repository for nuclear waste. Thermo-mechanical modeling was performed to help design the instrumentation arrays for the three proposed heater tests (two full scale tests and one time scale test) and predict the thermal environment of the heaters and instruments. The modeling does not reflect recent RHO revisions to the in situ heater experiment plan. Heaters, instrumentation, and data acquisition system designs and recommendations were adapted from those used in Sweden

Current Status and Performance Tests of Korea Heat Load Test Facility KoHLT-EB

Energy Technology Data Exchange (ETDEWEB)

Kim, Sukkwon; Jin, Hyunggon; Shin, Kyuin; Choi, Boguen; Lee, Eohwak; Yoon, Jaesung; Lee, Dongwon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Duckhoi; Cho, Seungyon [National Fusion Research Institute, Daejeon (Korea, Republic of)

2013-05-15

A commissioning test has been scheduled to establish the installation and preliminary performance experiments of the copper hypervapotron mockups. And a qualification test will be performed to evaluate the CuCrZr duct liner in the ITER neutral beam injection facility and the ITER first wall small-scale mockups of the semi-prototype, at up to 1.5 and 5 MW/m{sup 2} high heat flux. Also, this system will be used to test other PFCs for ITER and materials for tokamak reactors. Korean high heat flux test facility(KoHLT-EB; Korea Heat Load Test facility - Electron Beam) by using an electron beam system has been constructed in KAERI to perform the qualification test for ITER blanket FW semi-prototype mockups, hypervapotron cooling devices in fusion devices, and other ITER plasma facing components. The commissioning and performance tests with the supplier of e-gun system have been performed on November 2012. The high heat flux test for hypervapotron cooling device and calorimetry were performed to measure the surface heat flux, the temperature profile and cooling performance. Korean high heat flux test facility for the plasma facing components of nuclear fusion machines will be constructed to evaluate the performance of each component. This facility for the plasma facing materials will be equipped with an electron beam system with a 60 kV acceleration gun.

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)

Pilot scale, alpha disassembly and decontamination facility at the Savannah River Laboratory

International Nuclear Information System (INIS)

Cadieux, J.R.; Becker, G.W. Jr.; Richardson, G.W.; Coogler, A.L.

1982-01-01

An alpha-contained pilot facility is being built at the Savannah River Laboratory (SRL) for research into the disassembly and dcontamination of noncombustible, Transuranic (TRU) waste. The design and program objectives for the facility are presented along with the initial test results from laboratory scale decontamination experiments with Pu-238 and Cm-244

Phase III (full scale) agitated mixing test plan

International Nuclear Information System (INIS)

Ruff, D.T.

1994-01-01

Waste Receiving and Processing Facility Module 2A (WRAP 2A) is the proposed second module of the WRAP facility. This facility will provide the required treatment for contact Handled (CH) Low Level (LL) Mixed Waste (MW) to allow its permanent disposal. Solidification of a portion of this waste using a cement based grout has been selected in order to reduce the toxicity and mobility of the waste in the disposal site. Mixing of the waste with the cement paste and material handling constraints/requirements associated with the mixed material is, therefore, a key process in the overall treatment strategy. This test plan addresses Phase 3, Full Scale Testing. The objectives of these tests are to determine if there are scale-up issues associated with the mixing results obtained in Phase 1 and 2 mixing tests, verify the workability of mixtures resulting from previous formulation development efforts (Waste Immobilization Development [WID]), and provide a baseline for WRAP 2A mixing equipment design. To this end, the following objectives are of particular interest: determine geometric influence of mixing blade at full scale (i.e., size, type, and location: height/offset); determine if similar results in terms of mixing effectiveness and product quality are achievable at this scale; determine if vibration is as effective at this larger scale in fluidizing the mixture and aiding in cleaning the vessel; determine if baffles or sweeping blades are needed to aid in mixing at the larger size and for cleaning the vessel; and determine quality of the poured monolithic product and investigate exotherm and filling influences at this larger size

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

Experimental Evaluation for the Microvibration Performance of a Segmented PC Method Based High Technology Industrial Facility Using 1/2 Scale Test Models

Directory of Open Access Journals (Sweden)

Sijun Kim

2017-01-01

Full Text Available The precast concrete (PC method used in the construction process of high technology industrial facilities is limited when applied to those with greater span lengths, due to the transport length restriction (maximum length of 15~16 m in Korea set by traffic laws. In order to resolve this, this study introduces a structural system with a segmented PC system, and a 1/2 scale model with a width of 9000 mm (hereafter Segmented Model is manufactured to evaluate vibration performance. Since a real vibrational environment cannot be reproduced for vibration testing using a scale model, a comparative analysis of their relative performances is conducted in this study. For this purpose, a 1/2 scale model with a width of 7200 mm (hereafter Nonsegmented Model of a high technology industrial facility is additionally prepared using the conventional PC method. By applying the same experiment method for both scale models and comparing the results, the relative vibration performance of the Segmented Model is observed. Through impact testing, the natural frequencies of the two scale models are compared. Also, in order to analyze the estimated response induced by the equipment, the vibration responses due to the exciter are compared. The experimental results show that the Segmented Model exhibits similar or superior performances when compared to the Nonsegmented Model.

Primary functions of the first Japanese large-scale facility for exposing small animals to radon

International Nuclear Information System (INIS)

Ishimori, Yuu; Tanaka, Hiroshi; Mitsunobu, Fumihiro; Yamaoka, Kiyonori; Kataoka, Takahiro; Sakoda, Akihiro

2010-01-01

The Japan Atomic Energy Agency (JAEA) and Okayama University have carried out the experimental animal study and its related studies since 2007 in order to examine the physiological effects of radon in detail. Thus, a radon test facility for small animals was developed in order to increase the statistical certainty of our animal tests. This paper illustrates the performances of that facility, the first large-scale facility of its types in Japan. The facility has a potential of approximately 150 mouse-scale tests at the same time. The apparatus for exposing small animals to radon has six animal chamber groups each of which consists of five independent cages. Different radon concentrations in each animal chamber group are available. The major functions of the facility controlling radon and avoiding thoron were shown theoretically and experimentally. The relative standard deviation of radon concentration at the highest concentration group was about 5%, although the lower concentration groups seemed to be affected by variations in background radon. (author)

Operating experience with sodium valves in the TNO-sodium test facilities

International Nuclear Information System (INIS)

Gasselt, M.L.G. van

1974-01-01

The development of sodium components for the SNR-300 in Holland has reached the stage where full scale testing in sodium has almost been finished and construction is at its height. It is against this background that a review is given of the weaknesses in one area or the other of the commercially available types of sodium valves used in TNO's smaller test facilities at Apeldoorn and TNO's 50 MW sodium components test facility at Hengelo. (U.S.)

Integrated Human Test Facilities at NASA and the Role of Human Engineering

Science.gov (United States)

Tri, Terry O.

2002-01-01

Integrated human test facilities are a key component of NASA's Advanced Life Support Program (ALSP). Over the past several years, the ALSP has been developing such facilities to serve as a large-scale advanced life support and habitability test bed capable of supporting long-duration evaluations of integrated bioregenerative life support systems with human test crews. These facilities-targeted for evaluation of hypogravity compatible life support and habitability systems to be developed for use on planetary surfaces-are currently in the development stage at the Johnson Space Center. These major test facilities are comprised of a set of interconnected chambers with a sealed internal environment, which will be outfitted with systems capable of supporting test crews of four individuals for periods exceeding one year. The advanced technology systems to be tested will consist of both biological and physicochemical components and will perform all required crew life support and habitability functions. This presentation provides a description of the proposed test "missions" to be supported by these integrated human test facilities, the overall system architecture of the facilities, the current development status of the facilities, and the role that human design has played in the development of the facilities.

Comparison of SBLOCA Test Results with the FESTA Facility for the SMART Design

Energy Technology Data Exchange (ETDEWEB)

Ryu, Hyobong; Park, Hyun--Sik; Bae, Hwang; Ryu, Sung-Uk; Ko, Young-Joo; Yi, Sung-Jae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-10-15

The FESTA facility is a full height, 1/49-volume scaled test facility with four trains of a secondary system and PRHRS, and can be used to investigate the integral performance of the interconnected components and possible thermal-hydraulic phenomena occurring in the SMART (System-Integrated Modular Advanced Reactor) design, and to validate its safety for various design basis accidents and broad transient scenarios. The role of FESTA can be extended to examine and verify the normal, abnormal, and emergency operating procedures required during the construction phases of SMART. During the design of the FESTA facility, the height is preserved to the full scale, and its area and volume are scaled down to 1/49 compared with the prototype plant, SMART. The scaling ratios adopted in FESTA with respect to SMART are summarized in Table 1. The maximum core power is 2..0 MW, which is about 30% of the scaled full power. The design pressure and temperature of SMART-ITL can simulate the maximum operating conditions, that is, 18.0 MPa and 350 .deg. C. A preliminary analysis of small-break loss of coolant accident (SBLOCA) tests using the MARS/KS code for FESTA was previously conducted. In addition, major test results of SBLOCA scenarios with the VISTA-ITL facility for the SMART design were discussed. In this research, three SBLOCA experimental tests of a safety injection system (SIS) line break, shutdown cooling system (SCS) line break and pressurizer safety valve (PSV) line break for the SMART design were successfully performed and its major results have been compared and discussed. An integral effect test has been performed for the SBLOCA scenario for the SMART design with the FESTA facility.

A practical view of the insights from scaling thermal-hydraulic tests

Energy Technology Data Exchange (ETDEWEB)

Levin, A.E.; McPherson, G.D.

1995-09-01

The authors review the broad concept of scaling of thermal-hydraulic test facilities designed to acquire data for application to modeling the behavior of nuclear power plants, especially as applied to the design certification of passive advanced light water reactors. Distortions and uncertainties in the scaling process are described, and the possible impact of these effects on the test data are discussed. A practical approach to the use of data from the facilities is proposed, with emphasis on the insights to be gained from the test results rather than direct application of test results to behavior of a large plant.

Post-test analysis of the experiment 5.2C - total loss of feed water at the BETHSY test facility

Energy Technology Data Exchange (ETDEWEB)

Krepper, E; Schaefer, F

1998-10-01

The BETHSY-test facility is a 1:100 scaled thermohydraulic model of a 900 MW(el) pressurized water reactor (FRAMATOME). The test facility is mainly designed to investigate various accident scenarios and to provide an experimental data base for code validation and for the verification of accident management measures. (orig.)

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

Heat removal tests on dry storage facilities for nuclear spent fuels

International Nuclear Information System (INIS)

Wataru, M.; Saegusa, T.; Koga, T.; Sakamoto, K.; Hattori, Y.

1999-01-01

In Japan, spent fuel generated in NPP is controlled and stored in dry storage facility away-from reactor. Natural convection cooling system of the storage facility is considered advantageous from both safety and economic point of view. In order to realize this type of facility it is necessary to develop an evaluation method for natural convection characteristics and to make a rational design taking account safety and economic factors. Heat removal tests with the reduces scale models of storage facilities (cask, vault and silo) identified the the flow pattern in the test modules. The temperature and velocity distributions were obtained and the heat transfer characteristics were evaluated

Health and Safety Management for Small-scale Methane Fermentation Facilities

Science.gov (United States)

Yamaoka, Masaru; Yuyama, Yoshito; Nakamura, Masato; Oritate, Fumiko

In this study, we considered health and safety management for small-scale methane fermentation facilities that treat 2-5 ton of biomass daily based on several years operation experience with an approximate capacity of 5 t·d-1. We also took account of existing knowledge, related laws and regulations. There are no qualifications or licenses required for management and operation of small-scale methane fermentation facilities, even though rural sewerage facilities with a relative similar function are required to obtain a legitimate license. Therefore, there are wide variations in health and safety consciousness of the operators of small-scale methane fermentation facilities. The industrial safety and health laws are not applied to the operation of small-scale methane fermentation facilities. However, in order to safely operate a small-scale methane fermentation facility, the occupational safety and health management system that the law recommends should be applied. The aims of this paper are to clarify the risk factors in small-scale methane fermentation facilities and encourage planning, design and operation of facilities based on health and safety management.

S.E.T., CSNI Separate Effects Test Facility Validation Matrix

International Nuclear Information System (INIS)

1997-01-01

1 - Description of test facility: The SET matrix of experiments is suitable for the developmental assessment of thermal-hydraulics transient system computer codes by selecting individual tests from selected facilities, relevant to each phenomena. Test facilities differ from one another in geometrical dimensions, geometrical configuration and operating capabilities or conditions. Correlation between SET facility and phenomena were calculated on the basis of suitability for model validation (which means that a facility is designed in such a way as to stimulate the phenomena assumed to occur in a plant and is sufficiently instrumented); limited suitability for model variation (which means that a facility is designed in such a way as to stimulate the phenomena assumed to occur in a plant but has problems associated with imperfect scaling, different test fluids or insufficient instrumentation); and unsuitability for model validation. 2 - Description of test: Whereas integral experiments are usually designed to follow the behaviour of a reactor system in various off-normal or accident transients, separate effects tests focus on the behaviour of a single component, or on the characteristics of one thermal-hydraulic phenomenon. The construction of a separate effects test matrix is an attempt to collect together the best sets of openly available test data for code validation, assessment and improvement, from the wide range of experiments that have been carried out world-wide in the field of thermal hydraulics. In all, 2094 tests are included in the SET matrix

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...

Startup of the remote laboratory-scale waste-treatment facility

International Nuclear Information System (INIS)

Knox, C.A.; Siemens, D.H.; Berger, D.N.

1981-01-01

The Remote Laboratory-Scale Waste-Treatment Facility was designed as a system to solidify small volumes of radioactive liquid wastes. The objectives in operating this facility are to evaluate solidification processes, determine the effluents generated, test methods for decontaminating the effluents, and provide radioactive solidified waste products for evaluation. The facility consists of a feed-preparation module, a waste-solidification module and an effluent-treatment module. The system was designed for remote installation and operation. Several special features for remotely handling radioactive materials were incorporated into the design. The equipment was initially assembled outside of a radiochemical cell to size and fabricate the connecting jumpers between the modules and to complete some preliminary design-verification tests. The equipment was then disassembled and installed in the radiochemical cell. When installation was completed the entire system was checked out with water and then with a nonradioactive simulated waste solution. The purpose of these operations was to start up the facility, find and solve operational problems, verify operating procedures and train personnel. The major problems experienced during these nonradioactive runs were plugging of the spray calciner nozzle and feed tank pumping failures. When these problems were solved, radioactive operations were started. This report describes the installation of this facility, its special remote design feature and the startup operations

Development and Implementation of a Scaled Saltstone Facility at Savannah River National Laboratory - 13346

International Nuclear Information System (INIS)

Reigel, Marissa M.; Fowley, Mark D.; Hansen, Erich K.; Hera, Kevin R.; Marzolf, Athneal D.; Cozzi, Alex D.

2013-01-01

The Savannah River National Laboratory (SRNL) has supported the Saltstone Production Facility (SPF) since its conception. However, bench scaled tests have not always provided process or performance data related to the mixing, transfer, and other operations utilized in the SPF. A need was identified to better understand the SPF processes and to have the capabilities at SRNL to simulate the SPF unit operations to support an active low-level radioactive waste (LLW) processing facility. At the SPF, the dry premix is weighed, mixed and transferred to the Readco '10-inch' continuous mixer where it is mixed with the LLW salt solution from the Salt Feed Tank (SFT) to produce fresh Saltstone slurry. The slurry is discharged from the mixer into a hopper. The hopper feeds the grout pump that transfers the slurry through at least 457.2 meters of piping and discharges it into the Saltstone Disposal Units (SDU) for permanent disposal. In conjunction with testing individual SPF processes over several years, SRNL has designed and fabricated a scaled Saltstone Facility. Scaling of the system is primarily based on the volume capacity of the mixer and maintaining the same shear rate and total shear at the wall of the transfer line. At present, SRNL is utilizing the modular capabilities of the scaled Saltstone Facility to investigate the erosion issues related to the augers and paddles inside the SPF mixer. Full implementation of the scaled Saltstone Facility is still ongoing, but it is proving to be a valuable resource for testing alternate Saltstone formulations, cleaning sequences, the effect of pumping Saltstone to farther SDU's, optimization of the SPF mixer, and other operational variables before they are implemented in the SPF. (authors)

Results of the Preliminary Test in the 1/4-Scale RCCS of the PMR200 VHTR

Energy Technology Data Exchange (ETDEWEB)

Kim, Jong-Hwan; Bae, Yoon-Yeong; Hong, Sung-Deok; Kim, Chan-Soo; Cho, Bong-Hyun; Kim, Min-Hwan [Nuclear Hydrogen Reactor Technology Development Dep., Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-10-15

The Reactor Cavity Cooling System (RCCS) is a key ex-vessel passive safety system that will ensure the safety of the PMR200, and its performance needs to be verified. For the difficulty of the full-scale test, a 1/4-scale RCCS facility, NACEF (Natural Cooling Experimental Facility), has been constructed at KAERI, and a shakedown test has been performed. A brief design and the preliminary test results of this facility are described. A 1/4-scale RCCS mockup of PMR200, NACEF, was constructed and tested preliminarily. The functioning of the facility worked quite well. Moreover, the preliminary test results show a fairly good agreement with past work except for the conductive heat transfer region in the riser bottom. After a remedy such as the installation of more precise flow meters and a more improved insulation, the test facility is likely to work well.

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...

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.

Liquefied Gaseous Fuels Spill Test Facility: Overview of STF capabilities

International Nuclear Information System (INIS)

Gray, H.E.

1993-01-01

The Liquefied Gaseous Fuels Spill Test Facility (STF) constructed at the Department of Energy's Nevada Test Site is a basic research tool for studying the dynamics of accidental releases of various hazardous liquids. This Facility is designed to (1) discharge, at a controlled rate, a measured volume of hazardous test liquid on a prepared surface of a dry lake bed (Frenchman Lake); (2) monitor and record process operating data, close-in and downwind meteorological data, and downwind gaseous concentration levels; and (3) provide a means to control and monitor these functions from a remote location. The STF will accommodate large and small-scale testing of hazardous test fluid release rates up to 28,000 gallons per minute. Spill volumes up to 52,800 gallons are achievable. Generic categories of fluids that can be tested are cryogenics, isothermals, aerosol-forming materials, and chemically reactive. The phenomena that can be studied include source definition, dispersion, and pool fire/vapor burning. Other capabilities available at the STF include large-scale wind tunnel testing, a small test cell for exposing personnel protective clothing, and an area for developing mitigation techniques

Design, fabrication, and mockup testing in the remote maintenance development facility

International Nuclear Information System (INIS)

Carter, J.A.; Jacobs, R.T.; Bingham, G.E.

1978-01-01

The Remote Maintenance Development Facility at the Idaho National Engineering Laboratory was installed and used extensively for full-scale development, mockup, and testing of remote maintenance requirements for the New Waste Calcining Facility (NWCF). By performing remote handling tests, the NWCF handling concepts, techniques, and remote capabilities were proven workable prior to construction. A description of the RMDF and its purpose, functions, and handling capabilities as they were used in support of the NWCF is presented

Performance of the first Japanese large-scale facility for radon inhalation experiments with small animals

International Nuclear Information System (INIS)

Ishimori, Y.; Mitsunobu, F.; Yamaoka, K.; Tanaka, H.; Kataoka, T.; Sakoda, A.

2011-01-01

A radon test facility for small animals was developed in order to increase the statistical validity of differences of the biological response in various radon environments. This paper illustrates the performances of that facility, the first large-scale facility of its kind in Japan. The facility has a capability to conduct approximately 150 mouse-scale tests at the same time. The apparatus for exposing small animals to radon has six animal chamber groups with five independent cages each. Different radon concentrations in each animal chamber group are available. Because the first target of this study is to examine the in vivo behaviour of radon and its effects, the major functions to control radon and to eliminate thoron were examined experimentally. Additionally, radon progeny concentrations and their particle size distributions in the cages were also examined experimentally to be considered in future projects. (authors)

Control and data acquisition of the ITER full-scale ion source for the neutral beam test facility

International Nuclear Information System (INIS)

Luchetta, Adriano; Manduchi, Gabriele; Taliercio, Cesare; Paolucci, Francesco; Sartori, Filippo; Svensson, Lennart; Labate, Carmelo Vincenzo; Breda, Mauro; Capobianco, Roberto; Molon, Federico; Moressa, Modesto; Simionato, Paola; Zampiva, Enrico; Barbato, Paolo; Polato, Sandro

2015-01-01

Highlights: • This paper describes the requirements and architecture of the control and data acquisition system of the ITER full-ion source experiment in the neutral beam test facility. • The system architecture integrates various popular software frameworks. • Slow control is based on the EPICS (Experimental Physics and Industrial Control System) framework. • Fast control is based on the MARTe (Multi-threaded Application Real-Time executor) framework. • Data acquisition is based on the MDSplus framework. - Abstract: The neutral beam test facility, which is under construction in Padova, Italy, is developing the ITER full-scale ion source for the ITER heating neutral beam injectors, referred to as the SPIDER experiment, and the full-size prototype injector, referred to as MITICA. The SPIDER control and data acquisition system (CODAS) has been developed and its construction will start in 2014. Slow control and data acquisition will be based on the ITER CODAC core system software suite that has been designed to facilitate the integration of ITER plant systems with CODAC. Fast control and data acquisition will use solutions specific to the test facility, as the corresponding concepts are not ready-to-use in the ITER design. The ITER hardware catalog for fast control has been taken into consideration. The software development will be based on the integration of MDSplus and MARTe, two framework software packages that are well known in the fusion community, targeting data organization and fast real-time control, respectively. The paper revises the system requirements and the system design and shows the results already achieved in terms of system integration. In addition, the paper will report the experience in the usage of different cooperating software frameworks and in the integration of industrial procured plant systems.

Control and data acquisition of the ITER full-scale ion source for the neutral beam test facility

Energy Technology Data Exchange (ETDEWEB)

Luchetta, Adriano, E-mail: adriano.luchetta@igi.cnr.it [Consorzio RFX (CNR, ENEA, INFN, Università di Padova, Acciaierie Venete SpA), Padova (Italy); Manduchi, Gabriele; Taliercio, Cesare [Consorzio RFX (CNR, ENEA, INFN, Università di Padova, Acciaierie Venete SpA), Padova (Italy); Paolucci, Francesco; Sartori, Filippo [Fusion for Energy, Barcelona (Spain); Svensson, Lennart [ITER Organization, Route de Vinon-sur-Verdon, CS 90046 St. Paul Lez Durance (France); Labate, Carmelo Vincenzo [Association ENEA-CREATE, Department of Engineering, University of Naples “Parthenope” (Italy); Breda, Mauro; Capobianco, Roberto; Molon, Federico; Moressa, Modesto; Simionato, Paola; Zampiva, Enrico; Barbato, Paolo; Polato, Sandro [Consorzio RFX (CNR, ENEA, INFN, Università di Padova, Acciaierie Venete SpA), Padova (Italy)

2015-10-15

Highlights: • This paper describes the requirements and architecture of the control and data acquisition system of the ITER full-ion source experiment in the neutral beam test facility. • The system architecture integrates various popular software frameworks. • Slow control is based on the EPICS (Experimental Physics and Industrial Control System) framework. • Fast control is based on the MARTe (Multi-threaded Application Real-Time executor) framework. • Data acquisition is based on the MDSplus framework. - Abstract: The neutral beam test facility, which is under construction in Padova, Italy, is developing the ITER full-scale ion source for the ITER heating neutral beam injectors, referred to as the SPIDER experiment, and the full-size prototype injector, referred to as MITICA. The SPIDER control and data acquisition system (CODAS) has been developed and its construction will start in 2014. Slow control and data acquisition will be based on the ITER CODAC core system software suite that has been designed to facilitate the integration of ITER plant systems with CODAC. Fast control and data acquisition will use solutions specific to the test facility, as the corresponding concepts are not ready-to-use in the ITER design. The ITER hardware catalog for fast control has been taken into consideration. The software development will be based on the integration of MDSplus and MARTe, two framework software packages that are well known in the fusion community, targeting data organization and fast real-time control, respectively. The paper revises the system requirements and the system design and shows the results already achieved in terms of system integration. In addition, the paper will report the experience in the usage of different cooperating software frameworks and in the integration of industrial procured plant systems.

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

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...

Experiments on hydraulically-compensated Compressed Air Energy Storage (CAES) system using large-diameter vertical pipe two-phase flow test facility: test facility and test procedure

International Nuclear Information System (INIS)

Ohtsu, Iwao; Murata, Hideo; Kukita, Yutaka; Kumamaru, Hiroshige.

1996-07-01

JAERI, the University of Tokyo, the Central Research Institute of Electric Power Industry and Shimizu Corporation jointing performed and experimental study on two-phase flow in the hydraulically-compensated Compressed Air Energy Storage (CAES) system with a large-diameter vertical pipe two-phase flow test facility from 1993 to 1995. A hydraulically-compensated CAES system is a proposed, conceptual energy storage system where energy is stored in the form of compressed air in an underground cavern which is sealed by a deep (several hundred meters) water shaft. The shaft water head maintains a constant pressure in the cavern, of several mega Pascals, even during loading or unloading of the cavern with air. The dissolved air in the water, however, may create voids in the shaft when the water rises through the shaft during the loading, being forced by the air flow into the cavern. The voids may reduce the effective head of the shaft, and thus the seal may fail, if significant bubbling should occur in the shaft. This bubbling phenomenon (termed 'Champaign effect') and potential failure of the water seal ('blowout') are simulated in a scaled-height, scaled-diameter facility. Carbon dioxide is used to simulate high solubility of air in the full-height, full-pressure system. This report describes the expected and potential two-phase flow phenomena in a hydraulically-compensated CAES system, the test facility and the test procedure, a method to estimate quantities which are not directly measured by using measured quantities and hydrodynamic basic equations, and desirable additional instrumentation. (author)

Modelling and operation strategies of DLR's large scale thermocline test facility (TESIS)

Science.gov (United States)

Odenthal, Christian; Breidenbach, Nils; Bauer, Thomas

2017-06-01

In this work an overview of the TESIS:store thermocline test facility and its current construction status will be given. Based on this, the TESIS:store facility using sensible solid filler material is modelled with a fully transient model, implemented in MATLAB®. Results in terms of the impact of filler site and operation strategies will be presented. While low porosity and small particle diameters for the filler material are beneficial, operation strategy is one key element with potential for optimization. It is shown that plant operators have to ponder between utilization and exergetic efficiency. Different durations of the charging and discharging period enable further potential for optimizations.

Design, fabrication, and mockup testing in the Remote Maintenance Development Facility

International Nuclear Information System (INIS)

Carter, J.A.; Jacobs, R.T.; Bingham, G.E.

1978-01-01

The Remote Maintenance Development Facility (RMDF) at the Idaho National Engineering Laboratory (INEL) was installed and used extensively for full-scale development, mockup and testing of remote maintenance requirements for the New Waste Calcining Facility (NWCF). By performing remote handling tests, the NWCF handling concepts, techniques and remote capabilities were proven workable prior to construction. Presented in this paper is a description of the RMDF and its purpose, functions, and handling capabilities as they were used in support of the NWCF

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

In November of 2009, Clemson University was awarded a competitive grant from the U.S. Department of Energy to design, build and operate a facility for full-scale, highly accelerated mechanical testing of next-generation wind turbine drivetrain technologies. The primary goal of the project was to design, construct, commission, and operate a state-of-the-art sustainable facility that permits full-scale highly accelerated testing of advanced drivetrain systems for large wind turbines. The secondary goal was to meet the objectives of the American Recovery and Reinvestment Act of 2009, especially in job creation, and provide a positive impact on economically distressed areas in the United States, and preservation and economic recovery in an expeditious manner. The project was executed according to a managed cooperative agreement with the Department of Energy and was an extraordinary success. The resultant new facility is located in North Charleston, SC, providing easy transportation access by rail, road or ship and operates on an open access model such that it is available to the U.S. Wind Industry for research, analysis, and evaluation activities. The 72 m by 97 m facility features two mechanical dynamometer test bays for evaluating the torque and blade dynamic forces experienced by the rotors of wind turbine drivetrains. The dynamometers are rated at 7.5 MW and 15 MW of low speed shaft power and are configured as independent test areas capable of simultaneous operation. All six degrees of freedom, three linear and three rotational, for blade and rotor dynamics are replicated through the combination of a drive motor, speed reduction gearbox and a controllable hydraulic load application unit (LAU). This new LAU setup readily supports accelerated lifetime mechanical testing and load analysis for the entire drivetrain system of the nacelle and easily simulates a wide variety of realistic operating scenarios in a controlled laboratory environment. The development of these

Cryogenics for a vertical test stand facility for testing superconducting radio frequency cavities at RRCAT

International Nuclear Information System (INIS)

Gupta, Prabhat Kumar; Kumar, Manoj; Kush, P.K.

2015-01-01

Vertical Test Stand (VTS) Facility is located in a newly constructed building of Cryo-Engineering and Cryo-Module Development Division (CCDD). This test facility is one of the important facilities to develop SCRF technologies for superconducting accelerators like Indian Spallation Neutron Source. VTS has to be used for regular testing of the Superconducting Radio Frequency (SRF) Niobium cavities at nominal frequency of 1.3 GHz/ 650 MHz at 4 K / 2 K liquid helium (LHe) bath temperatures. Testing of these cavities at 2 K evaluates cavity processing methods, procedures and would also serve as a pre-qualification test for cavity to test it in horizontal cryostat, called horizontal test stand, with other cavity components such as tuner and helium vessel. Cryogenic technologies play a major role in these cavity testing facilities. Achieving and maintaining a stable temperature of 2 K in these test stands on regular and reliable basis is a challenging task and require broad range of cryogenic expertise, large scale system level understanding and many in-house technological and process developments. Furthermore this test stand will handle large amount of liquid helium. Therefore, an appropriately designed infrastructure is required to handle such large amount of helium gas generated during the operation of VTS .This paper describes the different cryogenic design aspects, initial cryogenic operation results and different cryogenic safety aspects. (author)

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.

Novel electro-magnetic test facility for the calibration of a propulsor fluctuating force module

International Nuclear Information System (INIS)

Schofield, N.; Lonsdale, A.; Hodges, A.Y.

2004-01-01

The testing of scale model propulsors is an essential part of any marine propulsion design process. The fluctuating force module (FFM) is a self-contained, instrumented propulsor drive system designed to be an integral part of a scaled propulsor test facility. This paper describes a novel electro-magnetic test facility which provides a static axial thrust of 0-1 kN and triaxial dynamic forces of 0.3-3 Nrms, at frequencies of 80-800 Hz, to an equivalent propulsor mass rotating at speeds of 0-900 rpm, in order to calibrate the FFM force measurement systems

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...

Analysis of Elektrogorsk 108 test facility experimental data

International Nuclear Information System (INIS)

Urbonas, R.

2001-01-01

In the paper an evaluation of experimental data obtained at Russian Elektrogorsk 108 (E-108) test facility is presented. E-108 facility is a scaled model of Russian RBMK design reactor. An attempt to validate state-of-the-art thermal hydraulic codes on the basis of E-108 test facility was made. Originally these codes were developed and validated for BWRs and PWRs. Since state-of-art thermal hydraulic codes are widely used for simulation of RBMK reactors further codes' implementation and validation is required. The facility was modelled by employing RELAP5 (INEEL, USA) thermal hydraulic system analysis best estimate code. The results show dependence from number of nodes used in the heated channels, frictional and form losses employed. The obtained oscillatory behaviour is resulted by density wave and critical heat flux. It is shown that codes are able to predict thermal hydraulic instability and sudden heat structure temperature excursion, when critical heat flux is approached, well. In addition, an uncertainty analysis of one of the experiments was performed by employing GRS developed System for Uncertainty and Sensitivity Analysis (SUSA). It was one of the first attempts to use this statistic-based methodology in Lithuania.(author)

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...

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...

Design and Calibration of a Full Scale Active Magnetic Bearing Based Test Facility for Investigating Rotordynamic Properties of Turbomachinery Seals in Multiphase Flow

DEFF Research Database (Denmark)

Voigt, Andreas Jauernik; Mandrup-Poulsen, Christian; Nielsen, Kenny Krogh

2016-01-01

of the subsea installed rotating machinery. It is well known that careful design of turbomachinery seals, such as interstage and balance piston seals, is pivotal for the performance of pumps and compressors. Consequently, the ability to predict the complex interaction between fluid dynamics and rotordynamics...... University of Denmark and Lloyd's Register Consulting are currently establishing a purpose built state of the art multiphase seal test facility, which is divided into three modules. Module I consists of a full scale Active Magnetic Bearing (AMB) based rotor dynamic test bench. The internally designed custom...... AMBs are equipped with an embedded Hall sensor system enabling high precision non-contact seal force quantification. Module II is a fully automatised calibration facility for the Hall sensor based force quantification system. Module III consists of the test seal housing assembly. This paper provides...

GENIUS-TF: a test facility for the GENIUS project

OpenAIRE

Baudis, L.; Dietz, A.; Heusser, G.; Majorovits, B.; Strecker, H.; Klapdor--Kleingrothaus, H. V.

2000-01-01

GENIUS is a proposal for a large scale detector of rare events. As a first step of the experiment, a small test version, the GENIUS test facility, will be build up at the Laboratorio Nazionale del Gran Sasso (LNGS). With about 40 kg of natural Ge detectors operated in liquid nitrogen, GENIUS-TF could exclude (or directly confirm) the DAMA annual modulation signature within about two years of measurement.

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

Large-Scale Spray Releases: Additional Aerosol Test Results

Energy Technology Data Exchange (ETDEWEB)

Daniel, Richard C.; Gauglitz, Phillip A.; Burns, Carolyn A.; Fountain, Matthew S.; Shimskey, Rick W.; Billing, Justin M.; Bontha, Jagannadha R.; Kurath, Dean E.; Jenks, Jeromy WJ; MacFarlan, Paul J.; Mahoney, Lenna A.

2013-08-01

One of the events postulated in the hazard analysis for the Waste Treatment and Immobilization Plant (WTP) and other U.S. Department of Energy (DOE) nuclear facilities is a breach in process piping that produces aerosols with droplet sizes in the respirable range. The current approach for predicting the size and concentration of aerosols produced in a spray leak event involves extrapolating from correlations reported in the literature. These correlations are based on results obtained from small engineered spray nozzles using pure liquids that behave as a Newtonian fluid. The narrow ranges of physical properties on which the correlations are based do not cover the wide range of slurries and viscous materials that will be processed in the WTP and in processing facilities across the DOE complex. To expand the data set upon which the WTP accident and safety analyses were based, an aerosol spray leak testing program was conducted by Pacific Northwest National Laboratory (PNNL). PNNL’s test program addressed two key technical areas to improve the WTP methodology (Larson and Allen 2010). The first technical area was to quantify the role of slurry particles in small breaches where slurry particles may plug the hole and prevent high-pressure sprays. The results from an effort to address this first technical area can be found in Mahoney et al. (2012a). The second technical area was to determine aerosol droplet size distribution and total droplet volume from prototypic breaches and fluids, including sprays from larger breaches and sprays of slurries for which literature data are mostly absent. To address the second technical area, the testing program collected aerosol generation data at two scales, commonly referred to as small-scale and large-scale testing. The small-scale testing and resultant data are described in Mahoney et al. (2012b), and the large-scale testing and resultant data are presented in Schonewill et al. (2012). In tests at both scales, simulants were used

Accelerating sustainability in large-scale facilities

CERN Multimedia

Marina Giampietro

2011-01-01

Scientific research centres and large-scale facilities are intrinsically energy intensive, but how can big science improve its energy management and eventually contribute to the environmental cause with new cleantech? CERN’s commitment to providing tangible answers to these questions was sealed in the first workshop on energy management for large scale scientific infrastructures held in Lund, Sweden, on the 13-14 October.   Participants at the energy management for large scale scientific infrastructures workshop. The workshop, co-organised with the European Spallation Source (ESS) and  the European Association of National Research Facilities (ERF), tackled a recognised need for addressing energy issues in relation with science and technology policies. It brought together more than 150 representatives of Research Infrastrutures (RIs) and energy experts from Europe and North America. “Without compromising our scientific projects, we can ...

Analysis of a shield design for a DT neutron generator test facility.

Science.gov (United States)

Chichester, D L; Pierce, G D

2007-10-01

Independent numerical simulations have been performed using the MCNP5 and SCALE5 radiation transport codes to evaluate the effectiveness of a concrete facility designed to shield personnel from neutron radiation emitted from DT neutron generators. The analysis considered radiation source terms of 14.1 MeV monoenergetic neutrons located at three discrete locations within the two test vaults in the facility, calculating neutron and photon dose rates at 44 locations around the facility using both codes. In addition, dose rate contours were established throughout the facility using the MCNP5 mesh tally feature. Neutron dose rates calculated outside of the facility are predicted to be below 0.01 mrem/h at all locations when all neutron generator source terms are operating within the facility. Similarly, the neutron dose rate in one empty test vault when the adjacent test vault is being utilized is also less then 0.01 mrem/h. For most calculation locations outside the facility the photon dose rates were less then the neutron dose rates by a factor of 10 or more.

Healy Clean Coal Project: Healy coal firing at TRW Cleveland Test Facility

Energy Technology Data Exchange (ETDEWEB)

Koyama, T.; Petrill, E.; Sheppard, D.

1991-08-01

A test burn of two Alaskan coals was conducted at TRW's Cleveland test facility in support of the Healy Clean Coal Project, as part of Clean Coal Technology III Program in which a new power plant will be constructed using a TRW Coal Combustion System. This system features ash slagging technology combined with NO{sub x} and SO{sub x} control. The tests, funded by the Alaska Industrial Development and Export Authority (AIDEA) and TRW, were conducted to verify that the candidate Healy station coals could be successfully fired in the TRW coal combustor, to provide data required for scale-up to the utility project size requirements, and to produce sufficient flash-calcined material (FCM) for spray dryer tests to be conducted by Joy/NIRO. The tests demonstrated that both coals are viable candidates for the project, provided the data required for scale-up, and produced the FCM material. This report describes the modifications to the test facility which were required for the test burn, the tests run, and the results of the tests.

Scaling and instrumentation of the LOFT facility

International Nuclear Information System (INIS)

Modro, S.M.; Goodrich, L.D.; McPherson, G.D.

1985-01-01

This paper describes the LOFT experimental facility and instrumentation of the facility during small break loss-of-coolant experiments. Basic scaling considerations applied in the facility design are presented. Because LOFT was not designed with emphasis on small break LOCA some atypicalities with regard to small break transients are discussed. Review of important small break LOCA phenomena observed during the experiments and their measurability is provided

Multiloop Integral System Test (MIST): MIST Facility Functional Specification

International Nuclear Information System (INIS)

Habib, T.F.; Koksal, C.G.; Moskal, T.E.; Rush, G.C.; Gloudemans, J.R.

1991-04-01

The Multiloop Integral System Test (MIST) is part of a multiphase program started in 1983 to address small-break loss-of-coolant accidents (SBLOCAs) specific to Babcock and Wilcox designed plants. MIST is sponsored by the US Nuclear Regulatory Commission, the Babcock ampersand Wilcox Owners Group, the Electric Power Research Institute, and Babcock and Wilcox. The unique features of the Babcock and Wilcox design, specifically the hot leg U-bends and steam generators, prevented the use of existing integral system data or existing integral facilities to address the thermal-hydraulic SBLOCA questions. MIST was specifically designed and constructed for this program, and an existing facility -- the Once Through Integral System (OTIS) -- was also used. Data from MIST and OTIS are used to benchmark the adequacy of system codes, such as RELAP5 and TRAC, for predicting abnormal plant transients. The MIST Functional Specification documents as-built design features, dimensions, instrumentation, and test approach. It also presents the scaling basis for the facility and serves to define the scope of work for the facility design and construction. 13 refs., 112 figs., 38 tabs

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...

The first DC performance test and analysis of CC conductor short sample at ASIPP conductor test facility

International Nuclear Information System (INIS)

Shi Yi; Wu Yu; Liu Huajun; Long Feng; Qian Li; Ren Zhibin; Li Shaolei; Liu Bo; Chen Jinglin

2012-01-01

Highlights: ► In this study the first DC performance experiments of ITER correction coil conductor short sample have been carried out in ASIPP test facility. ► A CC conductor short sample was fabricated and tested to confirm the capability of this test facility for qualification tests of CC conductors. ► There is no obvious impact of cycling on DC performance measurement. ► Those measured results of current sharing temperature are in agreement with the expected results from strand scaling - Abstract: The first DC performance experiments of ITER correction coil (CC) conductor short sample have been carried out in the conductor test facility of Institute of Plasma Physics, CAS (ASIPP) in January this year. Those experiments aim to investigate the DC performance of ITER CC conductor. The tested conductor short sample is bended as a half circle with the diameter of 270 mm to meet the background magnetic field shape. The half circle part of sample is longer than the final twist pitch. The current sharing temperature (T cs ) in the 3.86 T external magnetic field (B ex ), ≤12 kA could be measured including the critical current (I c ) run. There is no obvious impact of 1000 cycles on DC performance. Those measured T cs results are in agreement with the expected results from strand scaling.

Design for the National RF Test Facility at ORNL

International Nuclear Information System (INIS)

Gardner, W.L.; Hoffman, D.J.; Becraft, W.R.

1983-01-01

Conceptual and preliminary engineering design for the National RF Test Facility at Oak Ridge National Laboratory (ORNL) has been completed. The facility will comprise a single mirror configuration embodying two superconducting development coils from the ELMO Bumpy Torus Proof-of-Principle (EBT-P) program on either side of a cavity designed for full-scale antenna testing. The coils are capable of generating a 1.2-T field at the axial midpoint between the coils separated by 1.0 m. The vacuum vessel will be a stainless steel, water-cooled structure having an 85-cm-radius central cavity. The facility will have the use of a number of continuous wave (cw), radio-frequency (rf) sources at levels including 600 kW at 80 MHz and 100 kW at 28 GHz. Several plasma sources will provide a wide range of plasma environments, including densities as high as approx. 5 x 10 13 cm -3 and temperatures on the order of approx. 10 eV. Furthermore, a wide range of diagnostics will be available to the experimenter for accurate appraisal of rf testing

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.

Sodium-water reaction test facility (SWAT-3)

International Nuclear Information System (INIS)

Shimazu, Hisashi; Ukechi, Kazutoshi; Sasakura, Kazutake; Kusunoki, Junichi

1976-01-01

In the development of the liquid metal cooled fast breeder reactor (LMFBR), the steam generator (SG) is considered one of the most important components. The Power Reactor and Nuclear Fuel Development Corporation (PNC) is now promoting the research and development of the SG system used with the prototype fast breeder reactor ''Monju''. In this research, the phenomena of the sodium-water reaction in the SG are the key which must be investigated for the solution of problems. The test facility (SWAT-3) simulating Monju's SG on the scale of 1/2.5 was designed, fabricated and installed by IHI at Oarai Engineering Center of PNC, its pre-operation being accomplished in February 1975. The purpose of SWAT-3 is summarized as follows: (1) To perform an overall test on the safety of Monju's SG and intermediate heat transport system under the design condition against sodium-water reaction accidents. (2) To investigate the damage of the SG structure caused by the sodium-water reaction, and the possibility of repair and recovery operations. The first test was accomplished successfully on June 9, 1975. As a result of the test, the fundamental function of this test facility was proven to be satisfactory as expected. (auth.)

Small-Scale Spray Releases: Additional Aerosol Test Results

Energy Technology Data Exchange (ETDEWEB)

Schonewill, Philip P.; Gauglitz, Phillip A.; Kimura, Marcia L.; Brown, G. N.; Mahoney, Lenna A.; Tran, Diana N.; Burns, Carolyn A.; Kurath, Dean E.

2013-08-01

One of the events postulated in the hazard analysis at the Waste Treatment and Immobilization Plant (WTP) and other U.S. Department of Energy (DOE) nuclear facilities is a breach in process piping that produces aerosols with droplet sizes in the respirable range. The current approach for predicting the size and concentration of aerosols produced in a spray leak involves extrapolating from correlations reported in the literature. These correlations are based on results obtained from small engineered spray nozzles using pure liquids with Newtonian fluid behavior. The narrow ranges of physical properties on which the correlations are based do not cover the wide range of slurries and viscous materials that will be processed in the WTP and across processing facilities in the DOE complex. To expand the data set upon which the WTP accident and safety analyses were based, an aerosol spray leak testing program was conducted by Pacific Northwest National Laboratory (PNNL). PNNL’s test program addressed two key technical areas to improve the WTP methodology (Larson and Allen 2010). The first technical area was to quantify the role of slurry particles in small breaches where slurry particles may plug the hole and prevent high-pressure sprays. The results from an effort to address this first technical area can be found in Mahoney et al. (2012a). The second technical area was to determine aerosol droplet size distribution and total droplet volume from prototypic breaches and fluids, including sprays from larger breaches and sprays of slurries for which literature data are largely absent. To address the second technical area, the testing program collected aerosol generation data at two scales, commonly referred to as small-scale and large-scale. The small-scale testing and resultant data are described in Mahoney et al. (2012b) and the large-scale testing and resultant data are presented in Schonewill et al. (2012). In tests at both scales, simulants were used to mimic the

Test on large-scale seismic isolation elements, 2

International Nuclear Information System (INIS)

Mazda, T.; Moteki, M.; Ishida, K.; Shiojiri, H.; Fujita, T.

1991-01-01

Seismic isolation test program of Central Research Inst. of Electric Power Industry (CRIEPI) to apply seismic isolation to Fast Breeder Reactor (FBR) plant was started in 1987. In this test program, demonstration test of seismic isolation elements was considered as one of the most important research items. Facilities for testing seismic isolation elements were built in Abiko Research Laboratory of CRIEPI. Various tests of large-scale seismic isolation elements were conducted up to this day. Many important test data to develop design technical guidelines was obtained. (author)

Performance test results of helium gas circulator of mock-up test facility with full-scale reaction tube for HTTR hydrogen production system. Contract research

Energy Technology Data Exchange (ETDEWEB)

Shimizu, Akira; Kato, Michio; Hayashi, Koji [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment] [and others

2003-03-01

Hydrogen production system by steam reforming of methane will be connected to the High Temperature Engineering Test Reactor (HTTR) of the Japan Atomic Energy Research Institute (JAERI) against development of nuclear heat utilization system. To obtain design and safety database of the HTTR hydrogen production system, mock-up test facility with full-scale reaction was constructed in FY 2001 and hydrogen of 120m{sup 3}N{sub /}h was successfully produced in overall performance test. This report describes performance test results of a helium gas circulator in this facility. The circulator performance curves regarding to pressure-rise, input power and adiabatic thermal efficiency at standard revolution number were made based on the measured flow-rate, temperature and pressure data in overall performance test. The circulator performance prediction code was made based on these performance curves. The code can calculate revolution number, electric power and temperature-rise of the circulator using flow-rate, inlet temperature, inlet pressure and pressure-rise data. The verification of the code was carried out with the test data in FY 2002. Total pressure loss of the helium gas circulation loop was also evaluated. The circulator should be operated in conditions such as pressure from 2.7MPa to 4.0MPa and flow-rate from 250g/s to 400g/s and at maximum pressure-rise of 250 kPa in test operation. It was confirmed in above verification and evaluations that the circulator had performance to satisfy above conditions within operation limitation of the circulator such as maximum input-power of 150 kW and maximum revolution number of 12,000 rpm. (author)

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

Confirmatory tests of full-scale condensers for SBWR

International Nuclear Information System (INIS)

Masoni, P.; Botti, S.; Fitzsimmons, G.W.

1993-01-01

A full-scale isolation condenser and a full-scale passive containment cooling condenser for the Simplified Boiling Water Reactor (SBWR) will be tested to confirm the thermal-hydraulic and structural design characteristics of these components. The condensers provide vital roles in removing heat from the reactor vessel and the containment during certain design basis events. This paper describes the condensers and the test facilities which are under construction and summarizes the test objectives, the planned instrumentation, and the conditions to be tested. The results of some pre-test performance predictions, calculated with the TRACG code are presented. The results of the testing program are expected to demonstrate that the condenser designs will provide the required heat removal capacity and will survive the design basis temperature/pressure cycles without structural damage

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...

GERDA test facility for pressurized water reactors with straight tube steam generators

International Nuclear Information System (INIS)

Ahrens, G.; Haury, G.; Lahner, K.; Schatz, A.

1983-01-01

A number of large-scale experimental facilities have been constructed and operate in order to experiment on the thermodynamic and thermohydraulic behaviour of nuclear facilities in case of LOCA. Most of them were designed for ''large leak'' accidents, but as ''small leak'' accidents became the focus of interest, such experiments were also carried out. Experiments carried out with this arrangement for PWR-type reactors with straight-tube steam generators are only partially evaluable. BBR and B and W therefore cooperated in the construction of the test facility GERDA, designed for testing reactors of BBR design. It supplied relevant experimental results for the nuclear power plant at Muelheim-Kaerlich. (orig.) [de

The Transrapid test facility between system development and system application

Energy Technology Data Exchange (ETDEWEB)

Baur, L [MVP GmbH, Muenchen (Germany)

1996-12-31

In the development of a new rail technology, such as the magnetic levitation, there is - in contrast to the further development of the railway technology - not the possibility to use existing routes for the technical verification of the system technology until the application. Instead of this there are two possibilities: Cost-effective preliminary development on test beds and small test facilities up to a development stage which justifies the (relatively serious) risk of realising a service route early and to conclude the system trials and verification there; cost-intensive construction of a large-scale test facility which permits an application-related verification of all important system functions and thus creates the technical pre-requisites for a low-risk system application; The presentation deals with the technical requirements of the system at the test facility the challenges and chances linked to its realisation and adjustment to the rapidly progressing state-of-the-art and which this way opens up for a minimisation of the technical application risk. (orig./HW)

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...

Large-scale testing of in-vessel debris cooling through external flooding of the reactor pressure vessel in the CYBL facility

International Nuclear Information System (INIS)

Chu, T.Y.; Bentz, J.H.; Bergeron, K.D.; Slezak, S.E.; Simpson, R.B.

1994-01-01

The possibility of achieving in-vessel core retention by flooding the reactor cavity, or the ''flooded cavity'', is an accident management concept currently under consideration for advanced light water reactors (ALWR), as well as for existing light water reactors (LWR). The CYBL (CYlindrical BoiLing) facility is a facility specifically designed to perform large-scale confirmatory testing of the flooded cavity concept. CYBL has a tank-within-a-tank design; the inner 3.7 m diameter tank simulates the reactor vessel, and the outer tank simulates the reactor cavity. The energy deposition on the bottom head is simulated with an array of radiant heaters. The array can deliver a tailored heat flux distribution corresponding to that resulting from core melt convection. The present paper provides a detailed description of the capabilities of the facility, as well as results of recent experiments with heat flux in the range of interest to those required for in-vessel retention in typical ALWRs. The paper concludes with a discussion of other experiments for the flooded cavity applications

Large-Scale testing of in-vessel debris cooling through external flooding of the reactor pressure vessel in the CYBL facility

International Nuclear Information System (INIS)

Chu, T.Y.; Bentz, J.H.; Bergeron, K.D.; Slezak, S.E.; Simpson, R.B.

1994-01-01

The possibility of achieving in-vessel core retention by flooding the reactor cavity, or the open-quotes flooded cavityclose quotes, is an accident management concept currently under consideration for advanced light water reactors (ALWR), as well as for existing light water reactors (LWR). The CYBL (CYlindrical BoiLing) facility is a facility specifically designed to perform large-scale confirmatory testing of the flooded cavity concept. CYBL has a tank-within-a-tank design; the inner 3.7 m diameter tank simulates the reactor vessel, and the outer tank simulates the reactor cavity. The energy deposition on the bottom head is simulated with an array of radiant heaters. The array can deliver a tailored heat flux distribution corresponding to that resulting from core melt convection. The present paper provides a detailed description of the capabilities of the facility, as well as results of recent experiments with heat flux in the range of interest to those required for in-vessel retention in typical ALWRs. The paper concludes with a discussion of other experiments for the flooded cavity applications

Ripple scalings in geothermal facilities, a key to understand the scaling process

Science.gov (United States)

Köhl, Bernhard; Grundy, James; Baumann, Thomas

2017-04-01

Scalings are a widespread problem among geothermal plants which exploit the Malm Aquifer in the Bavarian Molasse Zone. They effect the technical and economic efficiency of geothermal plants. The majority of the scalings observed at geothermal facilities exploring the Malm aquifer in the Bavarian Molasse Basin are carbonates. They are formed due to a disruption of the lime-carbonic-acid equilibrium during production caused by degassing of CO2. These scalings are found in the production pipes, at the pumps and at filters and can nicely be described using existing hydrogeochemical models. This study proposes a second mechanism for the formation of scalings in ground-level facilities. We investigated scalings which accumulated at the inlet to the heat exchanger. Interestingly, the scalings were recovered after the ground level facilities had been cleaned. The scalings showed distinct ripple structures, which is likely a result of solid particle deposition. From the ripple features the the flow conditions during their formation were calculated based on empirical equations (Soulsby, 2012). The calculations suggest that the deposits were formed during maintenance works. Thin section images of the sediments indicate a two-step process: deposition of sediment grains, followed by stabilization with a calcite layer. The latter likely occured during maintenance. To prevent this type of scalings blocking the heat exchangers, the maintenance procedure has to be revised. References: Soulsby, R. L.; Whitehouse, R. J. S.; Marten, K. V.: Prediction of time-evolving sand ripples in shelf seas. Continental Shelf Research 2012, 38, 47-62

LOCA assessment experiments in a full-elevation, CANDU-typical test facility

International Nuclear Information System (INIS)

Ingham, P.J.; McGee, G.R.; Krishnan, V.S.

1990-01-01

The RD-14 thermal-hydraulics test facility, located at the Whiteshell Nuclear Research Establishment, is a full-elevation model representative of a CANDU primary heat transport system. The facility is scaled to accommodate a single, full-scale (5.0 MW, 21 kg/s), electrically heated channel per pass. The steam generators, pumps, headers, feeders and heated channels are arranged in a typical CANDU figure-of-eight geometry. The loop has an emergency coolant injection system (ECI) that may be operated in several modes, including typical features of the various ECI systems found in CANDU reactors. A series of experiments has been performed in RD-14 to investigate the thermal-hydraulic behaviour during the blowdown and injection phases of a loss-of-coolant accident (LOCA). The tests were designed to cover a full range of break sizes from feeder-sized breaks to guillotine breaks in either an inlet or an outlet header. Breaks resulting in channel flow stagnation were also investigated. This paper reviews the results of some of the LOCA tests carried out in RD-14, and discusses some of the behaviour observed. Plans for future experiments in a multiple-channel RD-14 facility, modified to contain multiple flow channels, are outlined. (orig.)

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.

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

PEROXIDE DESTRUCTION TESTING FOR THE 200 AREA EFFLUENT TREATMENT FACILITY

International Nuclear Information System (INIS)

Halgren, D.L.

2010-01-01

The hydrogen peroxide decomposer columns at the 200 Area Effluent Treatment Facility (ETF) have been taken out of service due to ongoing problems with particulate fines and poor destruction performance from the granular activated carbon (GAC) used in the columns. An alternative search was initiated and led to bench scale testing and then pilot scale testing. Based on the bench scale testing three manganese dioxide based catalysts were evaluated in the peroxide destruction pilot column installed at the 300 Area Treated Effluent Disposal Facility. The ten inch diameter, nine foot tall, clear polyvinyl chloride (PVC) column allowed for the same six foot catalyst bed depth as is in the existing ETF system. The flow rate to the column was controlled to evaluate the performance at the same superficial velocity (gpm/ft 2 ) as the full scale design flow and normal process flow. Each catalyst was evaluated on peroxide destruction performance and particulate fines capacity and carryover. Peroxide destruction was measured by hydrogen peroxide concentration analysis of samples taken before and after the column. The presence of fines in the column headspace and the discharge from carryover was generally assessed by visual observation. All three catalysts met the peroxide destruction criteria by achieving hydrogen peroxide discharge concentrations of less than 0.5 mg/L at the design flow with inlet peroxide concentrations greater than 100 mg/L. The Sud-Chemie T-2525 catalyst was markedly better in the minimization of fines and particle carryover. It is anticipated the T-2525 can be installed as a direct replacement for the GAC in the peroxide decomposer columns. Based on the results of the peroxide method development work the recommendation is to purchase the T-2525 catalyst and initially load one of the ETF decomposer columns for full scale testing.

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

Fermilab R and D test facility for SSC magnets

International Nuclear Information System (INIS)

Strait, J.; Bleadon, M.; Hanft, R.; Lamm, M.; McGuire, K.; Mantsch, P.; Mazur, P.O.; Orris, D.; Pachnik, J.

1989-01-01

The test facility used for R and D testing of full scale development dipole magnets for the SSC is described. The Fermilab Magnet Test Facility, originally built for production testing of Tevatron magnets, has been substantially modified to allow testing also of SSC magnets. Two of the original six test stands have been rebuilt to accommodate testing of SSC magnets at pressures between 1.3 Atm and 4 Atm and at temperatures between 1.8 K and 4.8 K and the power system has been modified to allow operation to at least 8 kA. Recent magnets have been heavily instrumented with voltage taps to allow detailed study of quench location and propagation and with strain gage based stress, force and motion transducers. A data acquisition system has been built with a capacity to read from each SSC test stand up to 220 electrical quench signals, 32 dynamic pressure, temperature and mechanical transducer signals during quench and up to 200 high precision, low time resolution, pressure, temperature and mechanical transducer signals. The quench detection and protection systems is also described. 23 refs., 4 figs. 2 tabs

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...

Design of Test Facility to Evaluate Boric Acid Precipitation Following a LOCA

Energy Technology Data Exchange (ETDEWEB)

Suh, Jeong-Kwan; Song, Yong-Jae [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

The U.S.NRC has identified a concern that debris associated with generic safety issue (GSI) - 191 may affect the potential precipitation of boric acid due to one or more of the following phenomena: - Reducing mass transport (i.e. mixing) between the core and the lower plenum (should debris accumulate at the core inlet) - Reduced lower plenum volume (should debris settle in the lower plenum), and, - Increased potential for boric acid precipitation (BAP) in the core (should debris accumulate in suspension in the core) To address these BAP issues, KHNP is planning to conduct validation tests by constructing a BAP test facility. This paper describes the design of test facility to evaluate BAP following a LOCA. The design of BAP test facility has been developed by KHNP. To design the test facility, test requirements and success criteria were established, and scaling analysis of power-to-volume method, Ishii-Kataoka method, and hierarchical two-tiered method were investigated. The test section is composed of two fuel assemblies with half of full of prototypic FA height. All the fuel rods are heated by the electric power supplier. The BAP tests in the presence of debris, buffering agents, and boron will be performed following the test matrix.

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

International Nuclear Information System (INIS)

Sugawara, S.

1995-01-01

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

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

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...

Role of non-destructive examinations in leak testing of glove boxes for industrial scale plutonium handling at nuclear fuel fabrication facility along with case study

International Nuclear Information System (INIS)

Aher, Sachin

2015-01-01

Non Destructive Examinations has the prominent role at Nuclear Fuel Fabrication Facilities. Specifically NDE has contributed at utmost stratum in Leak Testing of Glove Boxes and qualifying them as a Class-I confinement for safe Plutonium handling at industrial scale. Advanced Fuel Fabrication Facility, BARC, Tarapur is engaged in fabrication of Plutonium based MOX (PuO 2 , DDUO 2 ) fuel with different enrichments for first core of PFBR reactor. Alpha- Leak Tight Glove Boxes along with HEPA Filters and dynamic ventilation form the promising engineering system for safe and reliable handling of plutonium bearing materials considering the radiotoxicity and risk associated with handling of plutonium. Leak Testing of Glove Boxes which involves the leak detection, leak rectification and leak quantifications is major challenging task. To accomplish this challenge, various Non Destructive Testing methods have assisted in promising way to achieve the stringent leak rate criterion for commissioning of Glove Box facilities for plutonium handling. This paper highlights the Role of various NDE techniques like Soap Solution Test, Argon Sniffer Test, Pressure Drop/Rise Test etc. in Glove Box Leak Testing along with procedure and methodology for effective rectification of leakage points. A Flow Chart consisting of Glove Box leak testing procedure starting from preliminary stage up to qualification stage along with a case study and observations are discussed in this paper. (author)

The FENIX [Fusion ENgineering International EXperimental] test facility

International Nuclear Information System (INIS)

Slack, D.S.; Patrick, R.E.; Chaplin, M.R.; Miller, J.R.; Shen, S.S.; Summers, L.T.; Kerns, J.A.

1989-01-01

The Fusion ENgineering International EXperimental Magnet Facility (FENIX), under construction at Lawrence Livermore National Laboratory (LLNL), is a significant step forward in meeting the testing requirements necessary for the development of superconductor for large-scale, superconducting magnets. A 14-T, transverse field over a test volume of 150 x 60 x 150 mm in length will be capable of testing conductors the size of the International Thermonuclear Experimental Reactor (ITER). Proposed conductors for ITER measure ∼35 mm on one side and will operate at currents of up to 40 kA at fields of ∼14 T. The testing of conductors and associated components, such as joints, will require large-bore, high-field magnet facilities. FENIX is being constructed using the existing A 2o and A 2i magnets from the idle MFTF. The east and west A 2 pairs will be mounted together to form a split-pair solenoid. The pairs of magnets will be installed in a 4.0-m cryostat vessel located in the HFTF building at LLNL. Each magnet is enclosed in its own cryostat, the existing 4.0-m vessel serving only as a vacuum chamber. 4 refs., 8 figs

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

Annual report on experimental operations and maintenances of mock-up model test facility with a full-scale reaction tube for HTTR hydrogen production system in 2003 fiscal year (Contract research)

International Nuclear Information System (INIS)

Hayashi, Koji; Morisaki, Norihiro; Ohashi, Hirofumi; Kato, Michio; Aita, Hideki; Takeda, Tetsuaki; Nishihara, Tetsuo; Inaba, Yoshitomo; Takada, Shoji; Inagaki, Yoshiyuki

2006-03-01

This is a report on the experimental operations and maintenances of the mock-up test facility with a full-scale reaction tube for the HTTR hydrogen production system in 2003 fiscal year. The fourth and fifth experimental test operations were performed, from May to July and from October to December in 2003, for the following tests; (a) start-up and shutdown operation test, (b) process change test, (c) continuous hydrogen-production test and (d) chemical reaction shutdown test. From the results, a long time-range stability of the hydrogen production system was confirmed, a behavior of the helium-gas cooling system, consists of steam generator and radiator; during chemical reaction shutdown, was understood, and so on. Periodic inspections on boiler equipment and high-pressure gas production facilities were performed from end of July 2003. This report is summarized on outlines and results of the tests, outlines and results of the periodic inspections, and operation records of the mock-up test facility. (author)

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...

Fire exposed facades: Numerical modelling of the LEPIR2 testing facility

Directory of Open Access Journals (Sweden)

Dréan Virginie

2016-01-01

Full Text Available LEPIR2 testing facility is aimed to evaluate the fire behaviour of construction solutions implemented on facade according with the experimental evaluation required by the French Technical Specification 249 (IT249 of the safety regulation. It aims to limit the risks of fire spreading by facades to upper levels. This facility involves a wood crib fire in the lower compartment of a full scale two levels high structure. Flames are coming outside from the compartment through windows openings and develop in front of the facade. Computational fluids dynamics simulations are carried out with the FDS code (Fire Dynamics Simulator for two full-scale experiments performed by Efectis France laboratory. The first objective of this study is to evaluate the ability of numerical model to reproduce quantitative results in terms of gas temperatures and heat flux on the tested facade for further evaluation of fire performances of an insulation solution. When experimental results are compared with numerical calculations, good agreement is found out for every quantities and each test. The proposed models for wood cribs and geometry give correct thermal loads and flames shape near the tested facade.

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

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...

Current status of the Demonstration Test of Underground Cavern-Type Disposal Facilities

International Nuclear Information System (INIS)

Akiyama, Yoshihiro; Terada, Kenji; Oda, Nobuaki; Yada, Tsutomu; Nakajima, Takahiro

2011-01-01

In Japan, the underground cavern-type disposal facilities for low-level waste (LLW) with relatively high radioactivity, mainly generated from power reactor decommissioning, and for certain transuranic (TRU) waste, mainly from spent fuel reprocessing, are designed to be constructed in a cavern 50-100 m underground and to employ an engineered barrier system (EBS) made of bentonite and cement materials. To advance a disposal feasibility study, the Japanese government commissioned the Demonstration Test of Underground Cavern-Type Disposal Facilities in fiscal year (FY) 2005. Construction of a full-scale mock-up test facility in an actual subsurface environment started in FY 2007. The main test objective is to establish the construction methodology and procedures that ensure the required quality of the EBS on-site. A portion of the facility was constructed by 2010, and the test has demonstrated both the practicability of the construction and the achievement of quality standards: low permeability of less than 5x10 -13 m/s and low-diffusion of less than 1x10 -12 m 2 /s at the completion of construction. This paper covers the test results from the construction of certain parts using bentonite and cement materials. (author)

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...

Information on the Advanced Plant Experiment (APEX) Test Facility

International Nuclear Information System (INIS)

Smith, Curtis Lee

2015-01-01

The purpose of this report provides information related to the design of the Oregon State University Advanced Plant Experiment (APEX) test facility. Information provided in this report have been pulled from the following information sources: Reference 1: R. Nourgaliev and et.al, 'Summary Report on NGSAC (Next-Generation Safety Analysis Code) Development and Testing,' Idaho National Laboratory, 2011. Note that this is report has not been released as an external report. Reference 2: O. Stevens, Characterization of the Advanced Plant Experiment (APEX) Passive Residual Heat Removal System Heat Exchanger, Master Thesis, June 1996. Reference 3: J. Reyes, Jr., Q. Wu, and J. King, Jr., Scaling Assessment for the Design of the OSU APEX-1000 Test Facility, OSU-APEX-03001 (Rev. 0), May 2003. Reference 4: J. Reyes et al, Final Report of the NRC AP600 Research Conducted at Oregon State University, NUREG/CR-6641, July 1999. Reference 5: K. Welter et al, APEX-1000 Confirmatory Testing to Support AP1000 Design Certification (non-proprietary), NUREG-1826, August 2005.

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...

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)

Structural evaluation of spent nuclear fuel storage facilities under aircraft crash impact (2). Horizontal impact test onto reduced scale metal cask due to aircraft engine missile

International Nuclear Information System (INIS)

Namba, Kosuke; Shirai, Koji; Saegusa, Toshiari

2009-01-01

In this study, to confirm the sealing performance of a metal cask subjected to impact force due to possible commercial aircraft crash against a spent fuel storage facility, the horizontal impact test was carried out. In the test, an aircraft engine missile with a speed of 57.3 m/s attacked the reduced scale metal cask containing helium gas, which stands vertically. Then the leak rate and sliding displacement of the lid were measured. The leak rate increased rapidly and reached to 4.0 x 10 -6 Pa·m 3 /sec. After that, the leak rate decreased slowly and converged to 1.0x10 -6 Pa·m 3 /sec after 20 hours from the impact test. The leak rate of a full scale cask was evaluated using that of reduced scale cask obtained by the test. Then the leak rate of the full scale cask was 3.5x10 -5 Pa·m 3 /sec. This result showed that the sealing performance of the full scale metal cask would not be affected immediately by the horizontal impact of the aircraft engine with a speed of 57.3 m/s. (author)

Investigation of small break loss-of-coolant phenomena in a small scale nonnuclear test facility

International Nuclear Information System (INIS)

Cozzuol, J.M.; Fauble, T.J.; Harvego, E.A.

1980-01-01

A small-scale nonnuclear integral test facility designed to simulate a pressurized water reactor (PWR) system was used to evaluate the effects of a small break loss-of-coolant accident (LOCA) on the system thermal-hydraulic response. The experiment approximated a 2.5% (11-cm diameter) communicative break in the cold leg of a PWR, and included initial conditions which were similar to conditions in a PWR operating at full power. The 2.5% break size ensured that the nominal break flow rate was greater than the high pressure injection system (HPIS) flow rate, thus providing the potential for a continuous system depressurization. The sequence of events was similar to that used in evaluation model analysis of small break loss-of-coolant accidents, and included simulated reactor scram and loss of offsite power. Comparisions of experimental data with computer code calculations are used to demonstrate the capabilities and limitations of integral system calculations used to predict phenomena which can be important in the assessment of a small break LOCA in a PWR

The Testing Behind The Test Facility: The Acoustic Design of the NASA Glenn Research Center's World-Class Reverberant Acoustic Test Facility

Science.gov (United States)

Hozman, Aron D.; Hughes, William O.; McNelis, Mark E.; McNelis, Anne M.

2011-01-01

The National Aeronautics and Space Administration (NASA) Glenn Research Center (GRC) is leading the design and build of the new world-class vibroacoustic test capabilities at the NASA GRC's Plum Brook Station in Sandusky, Ohio, USA. Benham Companies, LLC is currently constructing modal, base-shake sine and reverberant acoustic test facilities to support the future testing needs of NASA's space exploration program. The large Reverberant Acoustic Test Facility (RATF) will be approximately 101,000 cu ft in volume and capable of achieving an empty chamber acoustic overall sound pressure level (OASPL) of 163 dB. This combination of size and acoustic power is unprecedented amongst the world's known active reverberant acoustic test facilities. The key to achieving the expected acoustic test spectra for a range of many NASA space flight environments in the RATF is the knowledge gained from a series of ground acoustic tests. Data was obtained from several NASA-sponsored test programs, including testing performed at the National Research Council of Canada's acoustic test facility in Ottawa, Ontario, Canada, and at the Redstone Technical Test Center acoustic test facility in Huntsville, Alabama, USA. The majority of these tests were performed to characterize the acoustic performance of the modulators (noise generators) and representative horns that would be required to meet the desired spectra, as well as to evaluate possible supplemental gas jet noise sources. The knowledge obtained in each of these test programs enabled the design of the RATF sound generation system to confidently advance to its final acoustic design and subsequent on-going construction.

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

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)

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)

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)

Healy Clean Coal Project: Healy coal firing at TRW Cleveland Test Facility. Final report

Energy Technology Data Exchange (ETDEWEB)

Koyama, T.; Petrill, E.; Sheppard, D.

1991-08-01

A test burn of two Alaskan coals was conducted at TRW`s Cleveland test facility in support of the Healy Clean Coal Project, as part of Clean Coal Technology III Program in which a new power plant will be constructed using a TRW Coal Combustion System. This system features ash slagging technology combined with NO{sub x} and SO{sub x} control. The tests, funded by the Alaska Industrial Development and Export Authority (AIDEA) and TRW, were conducted to verify that the candidate Healy station coals could be successfully fired in the TRW coal combustor, to provide data required for scale-up to the utility project size requirements, and to produce sufficient flash-calcined material (FCM) for spray dryer tests to be conducted by Joy/NIRO. The tests demonstrated that both coals are viable candidates for the project, provided the data required for scale-up, and produced the FCM material. This report describes the modifications to the test facility which were required for the test burn, the tests run, and the results of the tests.

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

NASA Data Acquisition System Software Development for Rocket Propulsion Test Facilities

Science.gov (United States)

Herbert, Phillip W., Sr.; Elliot, Alex C.; Graves, Andrew R.

2015-01-01

Current NASA propulsion test facilities include Stennis Space Center in Mississippi, Marshall Space Flight Center in Alabama, Plum Brook Station in Ohio, and White Sands Test Facility in New Mexico. Within and across these centers, a diverse set of data acquisition systems exist with different hardware and software platforms. The NASA Data Acquisition System (NDAS) is a software suite designed to operate and control many critical aspects of rocket engine testing. The software suite combines real-time data visualization, data recording to a variety formats, short-term and long-term acquisition system calibration capabilities, test stand configuration control, and a variety of data post-processing capabilities. Additionally, data stream conversion functions exist to translate test facility data streams to and from downstream systems, including engine customer systems. The primary design goals for NDAS are flexibility, extensibility, and modularity. Providing a common user interface for a variety of hardware platforms helps drive consistency and error reduction during testing. In addition, with an understanding that test facilities have different requirements and setups, the software is designed to be modular. One engine program may require real-time displays and data recording; others may require more complex data stream conversion, measurement filtering, or test stand configuration management. The NDAS suite allows test facilities to choose which components to use based on their specific needs. The NDAS code is primarily written in LabVIEW, a graphical, data-flow driven language. Although LabVIEW is a general-purpose programming language; large-scale software development in the language is relatively rare compared to more commonly used languages. The NDAS software suite also makes extensive use of a new, advanced development framework called the Actor Framework. The Actor Framework provides a level of code reuse and extensibility that has previously been difficult

Experience in the Application of INES scale to events in the Spanish Radioactive facilities

International Nuclear Information System (INIS)

Ramirez, M. L.; Alvarez, C.

2002-01-01

In February 2001, the International Atomic Energy Agency (IAEA) and the Nuclear energy Agency of the OECD (NEA) published a new edition of the INES User's Manual for the classification of nuclear events. One of the new developments introduced with respect to the scope of the former Manual was the inclusion within the INES of any event associated with radioactive material and/or radiation. This would include events occurred in radioactive facilities so the INES would apply not only to events in nuclear facilities. During the publication process some doubts rose about the applicability of INES to other non nuclear types of events. The IAEA was open to the future development of more practical guidance for the application of the scale. Since the beginning of 2001 the Consejo de Seguridad Nuclear (CSN) has been using INES to test the applicability of the system to classify events in radioactive facilities. A total of 31 events occurred at Spanish radioactive facilities has been classified applying INES scale and a report was sent to IAEA to publish our experience. The objective of this presentation is to introduce the experience obtained by the application of the International Nuclear Events Scale (INES) to classify events in radioactive facilities in Spain and to present several issues raised during its application that may need further development in a practical guidance. (Author)

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

Material testing facilities and programs for plasma-facing component testing

Science.gov (United States)

Linsmeier, Ch.; Unterberg, B.; Coenen, J. W.; Doerner, R. P.; Greuner, H.; Kreter, A.; Linke, J.; Maier, H.

2017-09-01

Component development for operation in a large-scale fusion device requires thorough testing and qualification for the intended operational conditions. In particular environments are necessary which are comparable to the real operation conditions, allowing at the same time for in situ/in vacuo diagnostics and flexible operation, even beyond design limits during the testing. Various electron and neutral particle devices provide the capabilities for high heat load tests, suited for material samples and components from lab-scale dimensions up to full-size parts, containing toxic materials like beryllium, and being activated by neutron irradiation. To simulate the conditions specific to a fusion plasma both at the first wall and in the divertor of fusion devices, linear plasma devices allow for a test of erosion and hydrogen isotope recycling behavior under well-defined and controlled conditions. Finally, the complex conditions in a fusion device (including the effects caused by magnetic fields) are exploited for component and material tests by exposing test mock-ups or material samples to a fusion plasma by manipulator systems. They allow for easy exchange of test pieces in a tokamak or stellarator device, without opening the vessel. Such a chain of test devices and qualification procedures is required for the development of plasma-facing components which then can be successfully operated in future fusion power devices. The various available as well as newly planned devices and test stands, together with their specific capabilities, are presented in this manuscript. Results from experimental programs on test facilities illustrate their significance for the qualification of plasma-facing materials and components. An extended set of references provides access to the current status of material and component testing capabilities in the international fusion programs.

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

PWR station blackout transient simulation in the INER integral system test facility

International Nuclear Information System (INIS)

Liu, T.J.; Lee, C.H.; Hong, W.T.; Chang, Y.H.

2004-01-01

Station blackout transient (or TMLB' scenario) in a pressurized water reactor (PWR) was simulated using the INER Integral System Test Facility (IIST) which is a 1/400 volumetrically-scaled reduce-height and reduce-pressure (RHRP) simulator of a Westinghouse three-loop PWR. Long-term thermal-hydraulic responses including the secondary boil-off and the subsequent primary saturation, pressurization and core uncovery were simulated based on the assumptions of no offsite and onsite power, feedwater and operator actions. The results indicate that two-phase discharge is the major depletion mode since it covers 81.3% of the total amount of the coolant inventory loss. The primary coolant inventory has experienced significant re-distribution during a station blackout transient. The decided parameter to avoid the core overheating is not the total amount of the coolant inventory remained in the primary core cooling system but only the part of coolant left in the pressure vessel. The sequence of significant events during transient for the IIST were also compared with those of the ROSA-IV large-scale test facility (LSTF), which is a 1/48 volumetrically-scaled full-height and full-pressure (FHFP) simulator of a PWR. The comparison indicates that the sequence and timing of these events during TMLB' transient studied in the RHRP IIST facility are generally consistent with those of the FHFP LSTF. (author)

Commercial-scale biotherapeutics manufacturing facility for plant-made pharmaceuticals.

Science.gov (United States)

Holtz, Barry R; Berquist, Brian R; Bennett, Lindsay D; Kommineni, Vally J M; Munigunti, Ranjith K; White, Earl L; Wilkerson, Don C; Wong, Kah-Yat I; Ly, Lan H; Marcel, Sylvain

2015-10-01

Rapid, large-scale manufacture of medical countermeasures can be uniquely met by the plant-made-pharmaceutical platform technology. As a participant in the Defense Advanced Research Projects Agency (DARPA) Blue Angel project, the Caliber Biotherapeutics facility was designed, constructed, commissioned and released a therapeutic target (H1N1 influenza subunit vaccine) in manufacturing facilities, with the capacity to process over 3500 kg of plant biomass per week in an automated multilevel growing environment using proprietary LED lighting. The facility can commission additional plant grow rooms that are already built to double this capacity. In addition to the commercial-scale manufacturing facility, a pilot production facility was designed based on the large-scale manufacturing specifications as a way to integrate product development and technology transfer. The primary research, development and manufacturing system employs vacuum-infiltrated Nicotiana benthamiana plants grown in a fully contained, hydroponic system for transient expression of recombinant proteins. This expression platform has been linked to a downstream process system, analytical characterization, and assessment of biological activity. This integrated approach has demonstrated rapid, high-quality production of therapeutic monoclonal antibody targets, including a panel of rituximab biosimilar/biobetter molecules and antiviral antibodies against influenza and dengue fever. © 2015 Society for Experimental Biology, Association of Applied Biologists and John Wiley & Sons Ltd.

Characterization results for 106-AN grout produced in a pilot-scale test

International Nuclear Information System (INIS)

Lokken, R.O.; Bagaasen, L.M.; Martin, P.F.C.; Palmer, S.E.; Anderson, C.M.

1993-06-01

The Grout Treatment Facility (GTF) at Hanford. Washington, will process the low-level fraction of selected double-shell tank (DST) wastes into a cementitious waste form. This facility, which is operated by Westinghouse Hanford Company (WHC), mixes liquid waste with cementitious materials to produce a waste form that immobilizes hazardous constituents through chemical reactions and/or microencapsulation. Over one million gallons of phosphate/sulfate waste were solidified in the first production campaign with this facility. The next tank waste scheduled for treatment is 106-AN (the waste from Tank 241-AN-106). After laboratory studies were conducted to select the grout formulation, tests using the 1/4-scale pilot facilities at the Pacific Northwest Laboratory (PNL) were conducted as part of the formulation verification process. The major objectives of these pilot-scale tests were to determine if the proposed grout formulation could be processed in the pilotscale equipment. to collect thermal information to help determine the best way to manage the grout hydration heat, and to characterize the solidified grout

In Situ Decommissioning Sensor Network, Meso-Scale Test Bed - Phase 3 Fluid Injection Test Summary Report

International Nuclear Information System (INIS)

Serrato, M. G.

2013-01-01

The DOE Office of Environmental management (DOE EM) faces the challenge of decommissioning thousands of excess nuclear facilities, many of which are highly contaminated. A number of these excess facilities are massive and robust concrete structures that are suitable for isolating the contained contamination for hundreds of years, and a permanent decommissioning end state option for these facilities is in situ decommissioning (ISD). The ISD option is feasible for a limited, but meaningfull number of DOE contaminated facilities for which there is substantial incremental environmental, safety, and cost benefits versus alternate actions to demolish and excavate the entire facility and transport the rubble to a radioactive waste landfill. A general description of an ISD project encompasses an entombed facility; in some cases limited to the blow-grade portion of a facility. However, monitoring of the ISD structures is needed to demonstrate that the building retains its structural integrity and the contaminants remain entombed within the grout stabilization matrix. The DOE EM Office of Deactivation and Decommissioning and Facility Engineering (EM-13) Program Goal is to develop a monitoring system to demonstrate long-term performance of closed nuclear facilities using the ISD approach. The Savannah River National Laboratory (SRNL) has designed and implemented the In Situ Decommissioning Sensor Network, Meso-Scale Test Bed (ISDSN-MSTB) to address the feasibility of deploying a long-term monitoring system into an ISD closed nuclear facility. The ISDSN-MSTB goal is to demonstrate the feasibility of installing and operating a remote sensor network to assess cementitious material durability, moisture-fluid flow through the cementitious material, and resulting transport potential for contaminate mobility in a decommissioned closed nuclear facility. The original ISDSN-MSTB installation and remote sensor network operation was demonstrated in FY 2011-12 at the ISDSN-MSTB test cube

In Situ Decommissioning Sensor Network, Meso-Scale Test Bed - Phase 3 Fluid Injection Test Summary Report

Energy Technology Data Exchange (ETDEWEB)

Serrato, M. G.

2013-09-27

The DOE Office of Environmental management (DOE EM) faces the challenge of decommissioning thousands of excess nuclear facilities, many of which are highly contaminated. A number of these excess facilities are massive and robust concrete structures that are suitable for isolating the contained contamination for hundreds of years, and a permanent decommissioning end state option for these facilities is in situ decommissioning (ISD). The ISD option is feasible for a limited, but meaningfull number of DOE contaminated facilities for which there is substantial incremental environmental, safety, and cost benefits versus alternate actions to demolish and excavate the entire facility and transport the rubble to a radioactive waste landfill. A general description of an ISD project encompasses an entombed facility; in some cases limited to the blow-grade portion of a facility. However, monitoring of the ISD structures is needed to demonstrate that the building retains its structural integrity and the contaminants remain entombed within the grout stabilization matrix. The DOE EM Office of Deactivation and Decommissioning and Facility Engineering (EM-13) Program Goal is to develop a monitoring system to demonstrate long-term performance of closed nuclear facilities using the ISD approach. The Savannah River National Laboratory (SRNL) has designed and implemented the In Situ Decommissioning Sensor Network, Meso-Scale Test Bed (ISDSN-MSTB) to address the feasibility of deploying a long-term monitoring system into an ISD closed nuclear facility. The ISDSN-MSTB goal is to demonstrate the feasibility of installing and operating a remote sensor network to assess cementitious material durability, moisture-fluid flow through the cementitious material, and resulting transport potential for contaminate mobility in a decommissioned closed nuclear facility. The original ISDSN-MSTB installation and remote sensor network operation was demonstrated in FY 2011-12 at the ISDSN-MSTB test cube

Investigations on efficiency of the emergency cooling by means of large-scale tests

International Nuclear Information System (INIS)

Hicken, E.F.

1982-01-01

The RSK guidelines contain the maximum permissible loads (max. cladding tube temperature 1200 0 C, max. Zr/H 2 O-reaction of 1% Zr). Their observance implies that only a small number of fuel rods fail. The safety research has to produce the evidence that the limiting loads are not exceeded. The analytical investigations on the emergency cooling behaviour could so far only be verified in scaled-down test facilities. After about 100 tests in four different large-scale test facilities the experimental investigations on the blow-down phase for large cracks are finished in the main. With the refill- and flood process the systems behaviour in scaled down test stands, the multidimensional conditions in the reactor pressure vessel can, however, only be simulated on the original scale. More experiments are planned as part of the 2D/3D-project (CCTF , SCTF, UPTF) and as part of the PKL-tests, so that more than 200 tests in seven plants will be available then. As to the small cracks the physical phenomena are known. The current investigations are used to increase the reliability of statement. After their being finished approximately 300 tests in seven plants will be available. (orig./HP) [de

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...

Annual report on experimental operations and maintenances of mock-up model test facility with a full-scale reaction tube for HTTR hydrogen production system in 2002 fiscal year (Contract research)

International Nuclear Information System (INIS)

Hayashi, Koji; Ohashi, Hirofumi; Inaba, Yoshitomo; Kato, Michio; Aita, Hideki; Morisaki, Norihiro; Takeda, Tetsuaki; Nishihara, Tetsuo; Takada, Shoji; Inagaki, Yoshiyuki

2006-03-01

This report describes 2002 fiscal-year experimental test operations of the mock-up test facility with a full-scale reaction tube for the HTTR hydrogen production system. The improvement works were performed in May 2002. The second experimental test operation was performed from June 2002 and the performances of the improved parts were confirmed. Periodic inspections on boiler equipment and high-pressure gas production facilities were performed from end of July 2002. The third experimental test operation was performed, from October 2002, for (a) start-up and shutdown test, (b) process change test, (c) chemical reaction shutdown test and (d) characteristics test on steam reformer. It was confirmed that the changes of helium gas temperature, caused at steam reformer, could be mitigated into the target range by the steam generator. Maintenance works of high-pressure gas production facilities were also performed in February 2003. This report is summarized with the outline and the results of the test, maintenance works and inspections, and operation records in mentioned above. (author)

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...

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...

Integral test facilities for validation of the performance of passive safety systems and natural circulation

International Nuclear Information System (INIS)

Choi, J. H.

2010-10-01

Passive safety systems are becoming an important component in advanced reactor designs. This has led to an international interest in examining natural circulation phenomena as this may play an important role in the operation of these passive safety systems. Understanding reactor system behaviour is a challenging process due to the complex interactions between components and associated phenomena. Properly scaled integral test facilities can be used to explore these complex interactions. In addition, system analysis computer codes can be used as predictive tools in understanding the complex reactor system behaviour. However, before the application of system analysis computer codes for reactor design, it is capability in making predictions needs to be validated against the experimental data from a properly scaled integral test facility. The IAEA has organized a coordinated research project (CRP) on natural circulation phenomena, modeling and reliability of passive systems that utilize natural circulation. This paper is a part of research results from this CRP and describes representative international integral test facilities that can be used for data collection for reactor types in which natural circulation may play an important role. Example experiments were described along with the analyses of these example cases in order to examine the ability of system codes to model the phenomena that are occurring in the test facilities. (Author)

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

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.

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.

Thermal-hydraulic tests with out-of-pile test facility for BOCA development

International Nuclear Information System (INIS)

Kitagishi, Shigeru; Aoyama, Masashi; Tobita, Masahiro; Inaba, Yoshitomo; Yamaura, Takayuki

2012-01-01

The fuel transient test facility was prepared for power ramping tests of light-water-reactor (LWR) fuels in the Japan Materials Testing Reactor (JMTR) under a contract project with the Nuclear Industrial Safety Agent (NISA) of the Ministry of Economy, Trade and Industry (METI). It is necessary to develop high accuracy analysis procedure for power ramping tests after restart of the JMTR. The out-of-pile test facility to simulate thermal-hydraulic conditions of the fuel transient test facility was therefore developed. Applicability of the analysis code ACE-3D was examined for thermal-hydraulic analysis of power ramping tests for 10x10 BWR fuels by the fuel transient test facility. As the results, the calculated temperature was 304°C in comparison with measured value of 304.9-317.4°C in the condition of 600 W/cm. There is a bright prospect of high accuracy power ramping tests by the fuel transient test facility in JMTR. (author)

Mark I 1/5-scale boiling water reactor pressure suppression experiment facility report

International Nuclear Information System (INIS)

Altes, R.G.; Pitts, J.H.; Ingraham, R.F.; Collins, E.K.; McCauley, E.W.

1977-01-01

An accurate Mark I 1 / 5 -scale, boiling water reactor (BWR), pressure suppression facility was designed and constructed at Lawrence Livermore Laboratory (LLL) in 11 months. Twenty-seven air tests using the facility are described. Cost was minimized by utilizing equipment borrowed from other LLL programs. The total value of borrowed equipment exceeded the program's budget of $2,020,000. Substantial flexibility in the facility was used to permit independent variation in the drywell pressure-time history, initial pressure in the drywell and toroidal wetwells, initial toroidal wetwell water level and downcomer length, vent line flow resistance, and vent line flow asymmetry. The two- and three-dimensional sectors of the toroidal wetwell provided significant data

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

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

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)

Goethite Bench-scale and Large-scale Preparation Tests

Energy Technology Data Exchange (ETDEWEB)

Josephson, Gary B.; Westsik, Joseph H.

2011-10-23

The Hanford Waste Treatment and Immobilization Plant (WTP) is the keystone for cleanup of high-level radioactive waste from our nation's nuclear defense program. The WTP will process high-level waste from the Hanford tanks and produce immobilized high-level waste glass for disposal at a national repository, low activity waste (LAW) glass, and liquid effluent from the vitrification off-gas scrubbers. The liquid effluent will be stabilized into a secondary waste form (e.g. grout-like material) and disposed on the Hanford site in the Integrated Disposal Facility (IDF) along with the low-activity waste glass. The major long-term environmental impact at Hanford results from technetium that volatilizes from the WTP melters and finally resides in the secondary waste. Laboratory studies have indicated that pertechnetate ({sup 99}TcO{sub 4}{sup -}) can be reduced and captured into a solid solution of {alpha}-FeOOH, goethite (Um 2010). Goethite is a stable mineral and can significantly retard the release of technetium to the environment from the IDF. The laboratory studies were conducted using reaction times of many days, which is typical of environmental subsurface reactions that were the genesis of this new process. This study was the first step in considering adaptation of the slow laboratory steps to a larger-scale and faster process that could be conducted either within the WTP or within the effluent treatment facility (ETF). Two levels of scale-up tests were conducted (25x and 400x). The largest scale-up produced slurries of Fe-rich precipitates that contained rhenium as a nonradioactive surrogate for {sup 99}Tc. The slurries were used in melter tests at Vitreous State Laboratory (VSL) to determine whether captured rhenium was less volatile in the vitrification process than rhenium in an unmodified feed. A critical step in the technetium immobilization process is to chemically reduce Tc(VII) in the pertechnetate (TcO{sub 4}{sup -}) to Tc(Iv)by reaction with the

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

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.

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.

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

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.

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

LS-DYNA Analysis of a Full-Scale Helicopter Crash Test

Science.gov (United States)

Annett, Martin S.

2010-01-01

A full-scale crash test of an MD-500 helicopter was conducted in December 2009 at NASA Langley's Landing and Impact Research facility (LandIR). The MD-500 helicopter was fitted with a composite honeycomb Deployable Energy Absorber (DEA) and tested under vertical and horizontal impact velocities of 26 ft/sec and 40 ft/sec, respectively. The objectives of the test were to evaluate the performance of the DEA concept under realistic crash conditions and to generate test data for validation of a system integrated LS-DYNA finite element model. In preparation for the full-scale crash test, a series of sub-scale and MD-500 mass simulator tests was conducted to evaluate the impact performances of various components, including a new crush tube and the DEA blocks. Parameters defined within the system integrated finite element model were determined from these tests. The objective of this paper is to summarize the finite element models developed and analyses performed, beginning with pre-test and continuing through post test validation.

Design Optimization for a Truncated Catenary Mooring System for Scale Model Test

Directory of Open Access Journals (Sweden)

Climent Molins

2015-11-01

Full Text Available One of the main aspects when testing floating offshore platforms is the scaled mooring system, particularly with the increased depths where such platforms are intended. The paper proposes the use of truncated mooring systems to emulate the real mooring system by solving an optimization problem. This approach could be an interesting option when the existing testing facilities do not have enough available space. As part of the development of a new spar platform made of concrete for Floating Offshore Wind Turbines (FOWTs, called Windcrete, a station keeping system with catenary shaped lines was selected. The test facility available for the planned experiments had an important width constraint. Then, an algorithm to optimize the design of the scaled truncated mooring system using different weights of lines was developed. The optimization process adjusts the quasi-static behavior of the scaled mooring system as much as possible to the real mooring system within its expected maximum displacement range, where the catenary line provides the restoring forces by its suspended line length.

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.

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

Full Scale Drinking Water System Decontamination at the Water Security Test Bed

Data.gov (United States)

U.S. Environmental Protection Agency — The EPA’s Water Security Test Bed (WSTB) facility is a full-scale representation of a drinking water distribution system. In collaboration with the Idaho National...

Utilization of the Pilot Scale Demonstration Facility for Vitrification of Low and Intermediate Level Radioactive Wastes

International Nuclear Information System (INIS)

Oh, Won Zin; Choi, W. K.; Jung, C. H.; Won, H. J.; Song, P. S.; Min, B. Y.; Park, H. S.; Jung, K. K.; Yun, K. S.

2005-10-01

A series of maintenance and repair work for normalization of the pilot scale vitrification demonstration facility was completed successfully to develop the waste treatment in high temperature and melting technology. It was investigated that the treatment of combustible and non-combustible wastes produced at the KAERI site is technically feasible in the pilot scale vitrification demonstration facility which is designed to be able to treat various kinds of radioactive wastes such as combustible and non-combustible wastes including soil and concrete. The vitrification test facility can be used as the R and D and the technology demonstration facility for melt decontamination of the metallic wastes which have a fixed specification. The modification of the RI storage room in the pilot scale vitrification demonstration facility and the licensing according to the facility modification were completed for the R and D on melt decontamination of dismantled metallic wastes which is carrying out as one of the national long-term R and D projects on nuclear energy. The lab-scale melt decontamination apparatus was installed in modified RI storage room and the characteristics of melt decontamination will be examined using various metallic wastes. It is expected that the economical feasibility on the volume reduction and recycle of metallic wastes will be escalated in the present situation when the unit cost for waste disposal has the tendency to grow up gradually. Therefore, the pilot scale vitrification demonstration facility can be used for the technology development for the volume reduction and recycle of the metallic wastes generated from on-going projects on the decommissioning of research reactors and the environmental restoration of uranium conversion plant, and for the treatment of radioactive solid wastes produced at the KAERI site

Utilization of the Pilot Scale Demonstration Facility for Vitrification of Low and Intermediate Level Radioactive Wastes

Energy Technology Data Exchange (ETDEWEB)

Oh, Won Zin; Choi, W. K.; Jung, C. H.; Won, H. J.; Song, P. S.; Min, B. Y.; Park, H. S.; Jung, K. K.; Yun, K. S

2005-10-15

A series of maintenance and repair work for normalization of the pilot scale vitrification demonstration facility was completed successfully to develop the waste treatment in high temperature and melting technology. It was investigated that the treatment of combustible and non-combustible wastes produced at the KAERI site is technically feasible in the pilot scale vitrification demonstration facility which is designed to be able to treat various kinds of radioactive wastes such as combustible and non-combustible wastes including soil and concrete. The vitrification test facility can be used as the R and D and the technology demonstration facility for melt decontamination of the metallic wastes which have a fixed specification. The modification of the RI storage room in the pilot scale vitrification demonstration facility and the licensing according to the facility modification were completed for the R and D on melt decontamination of dismantled metallic wastes which is carrying out as one of the national long-term R and D projects on nuclear energy. The lab-scale melt decontamination apparatus was installed in modified RI storage room and the characteristics of melt decontamination will be examined using various metallic wastes. It is expected that the economical feasibility on the volume reduction and recycle of metallic wastes will be escalated in the present situation when the unit cost for waste disposal has the tendency to grow up gradually. Therefore, the pilot scale vitrification demonstration facility can be used for the technology development for the volume reduction and recycle of the metallic wastes generated from on-going projects on the decommissioning of research reactors and the environmental restoration of uranium conversion plant, and for the treatment of radioactive solid wastes produced at the KAERI site.

The X-625 Groundwater Treatment Facility: A field-scale test of trichloroethylene dechlorination using iron filings for the X-120/X-749 groundwater plume

International Nuclear Information System (INIS)

Liang, L.; West, O.R.; Korte, N.E.

1997-09-01

The dehalogenation of chlorinated solvents by zero-valence iron has recently become the subject of intensive research and development as a potentially cost-effective, passive treatment for contaminated groundwater through reactive barriers. Because of its successful application in the laboratory and other field sites, the X-625 Groundwater Treatment Facility (GTF) was constructed to evaluate reactive barrier technology for remediating trichloroethylene (TCE)-contaminated groundwater at the Portsmouth Gaseous Diffusion Plant (PORTS). The X-625 GTF was built to fulfill the following technical objectives: (1) to test reactive barrier materials (e.g., iron filings) under realistic groundwater conditions for long term applications, (2) to obtain rates at which TCE degrades and to determine by-products for the reactive barrier materials tested, and (3) to clean up the TCE-contaminated water in the X-120 plume. The X-625 is providing important field-scale and long-term for the evaluation and design of reactive barriers at PORTS. The X-625 GTS is a unique facility not only because it is where site remediation is being performed, but it is also where research scientists and process engineers can test other promising reactive barrier materials. In addition, the data collected from X-625 GTF can be used to evaluate the technical and economic feasibility of replacing the activated carbon units in the pump-and-treat facilities at PORTS

Current state of the construction of SPARC test facility for observing hydrogen′s behavior

Energy Technology Data Exchange (ETDEWEB)

Na, Young Su; Hong, Seong-Ho; Park, Ki Han; Hong, Seong-Wan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

Hydrogen combustion can make a dynamic load, which can cause severe damage to a structure or facility. Many studies on hydrogen behavior, such as distribution, combustion and mitigation, have been conducted since the TMI accident, and they were recently summarized in. A large-scaled experimental facility is required for simulating the complex severe accident phenomena in a closed containment building. We are preparing the test facility, called the SPARC (Spray, Aerosol, Recombiner, Combustion), to resolve the international open issues regarding hydrogen risk as well as the validation of the Korean PAR (Passive Auto-catalytic Recombiner). This paper summarized the previous study submitted to the NUTHOS-11, which introduced the SPARC test facility. KAERI (Korea Atomic Energy Research Institute) is preparing a test facility, called the SPARC for an assessment of the containment integrity under a severe accident. In the SPARC test facility, the hydrogen behavior such as mixing with steam and air, distribution, and combustion will be observed under various thermal-hydraulic conditions. We will carry out the performance tests of the safety systems such as the spray, cooling fan, PAR, and igniter. The SPARC test facility consists of a pressure vessel with a 9.5 m height and 3.4 m diameter, and an operating system to control and measure the thermal hydraulic conditions. In a commissioning test, we verified the controllable thermal conditions. It took about 8,400 seconds to increase up to 5 bar. The increment rate of the atmosphere temperature is about 34° C/h from room temperature to 100° C.

10-MWe pilot-plant-receiver panel test requirements document solar thermal test facility

Energy Technology Data Exchange (ETDEWEB)

1978-08-25

Testing plans for a full-scale test receiver panel and supporting hardware which essentially duplicate both physically and functionally, the design planned for the Barstow Solar Pilot Plant are presented. Testing is to include operation during normal start and shutdown, intermittent cloud conditions, and emergencies to determine the panel's transient and steady state operating characteristics and performance under conditions equal to or exceeding those expected in the pilot plant. The effects of variations of input and output conditions on receiver operation are also to be investigated. Test hardware are described, including the pilot plant receiver, the test receiver assembly, receiver panel, flow control, electrical control and instrumentation, and structural assembly. Requirements for the Solar Thermal Test Facility for the tests are given. The safety of the system is briefly discussed, and procedures are described for assembly, installation, checkout, normal and abnormal operations, maintenance, removal and disposition. Also briefly discussed are quality assurance, contract responsibilities, and test documentation. (LEW)

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...

Buffer Construction Methodology in Demonstration Test For Cavern Type Disposal Facility

International Nuclear Information System (INIS)

Yoshihiro, Akiyama; Takahiro, Nakajima; Katsuhide, Matsumura; Kenji, Terada; Takao, Tsuboya; Kazuhiro, Onuma; Tadafumi, Fujiwara

2009-01-01

A number of studies concerning a cavern type disposal facility have been carried out for disposal of low level radioactive waste mainly generated by power plant decommissioning in Japan. The disposal facility is composed of an engineered barrier system with concrete pit and bentonite buffer, and planed to be constructed in sub-surface 50 - 100 meters depth. Though the previous studies have mainly used laboratory and mock-up tests, we conducted a demonstration test in a full-size cavern. The main objectives of the test were to study the construction methodology and to confirm the quality of the engineered barrier system. The demonstration test was planned as the construction of full scale mock-up. It was focused on a buffer construction test to evaluate the construction methodology and quality control in this paper. Bentonite material was compacted to 1.6 Mg/m 3 in-site by large vibrating roller in this test. Through the construction of the buffer part, a 1.6 Mg/m 3 of the density was accomplished, and the data of workability and quality is collected. (authors)

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

Remote handling of the blanket segments: testing of 1/3 scale mock-ups at the Robertino facility

International Nuclear Information System (INIS)

Maisonnier, D.; Amelotti, F.; Chiasera, A.; Gaggini, P.; Damiani, C.; Degli Esposti, L.; Gatti, G.; Castillo, E.; Caravati, D.; Farfalletti-Casali, F.; Gritzmann, P.; Ruiz, E.

1995-01-01

The remote replacement of blanket segments inside the vacuum vessel of a fusion reactor is probably the most complex task from the maintenance standpoint. Its success will rely on the definition of appropriate handling concepts and equipment, but also on a ''maintenance friendly'' reactor layout and blanket design. The key difficulty is the lack of rigidity of the segments which results in considerable deformations since they cannot be gripped above their centre of gravity. These deformations may be up to five times greater than the assembly clearance and one order of magnitude larger than the required positioning accuracy. Experimental activities have been undertaken to select appropriate handling devices and procedures, to assess the design of the components handled, and to review specific technical issues such as kinematics and dynamics performance, trajectory planning and control and sensors requirement for the handling devices. Work was performed in the Robertino facility where two handling concepts have been tested at a 1/3 scale. (orig.)

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.

Sandia Wake Imaging System Field Test Report: 2015 Deployment at the Scaled Wind Farm Technology (SWiFT) Facility.

Energy Technology Data Exchange (ETDEWEB)

Naughton, Brian Thomas [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Herges, Thomas [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2015-10-01

This report presents the objectives, configuration, procedures, reporting , roles , and responsibilities and subsequent results for the field demonstration of the Sandia Wake Imaging System (SWIS) at the Sandia Scaled Wind Farm Technology (SWiFT) facility near Lubbock, Texas in June and July 2015.

Assembly and installation of the large coil test facility test stand

International Nuclear Information System (INIS)

Queen, C.C. Jr.

1983-01-01

The Large Coil Test Facility (LCTF) was built to test six tokamak-type superconducting coils, with three to be designed and built by US industrial teams and three provided by Japan, Switzerland, and Euratom under an international agreement. The facility is designed to test these coils in an environment which simulates that of a tokamak. The heart of this facility is the test stand, which is made up of four major assemblies: the Gravity Base Assembly, the Bucking Post Assembly, the Torque Ring Assembly, and the Pulse Coil Assembly. This paper provides a detailed review of the assembly and installation of the test stand components and the handling and installation of the first coil into the test stand

Research on the Construction Management and Sustainable Development of Large-Scale Scientific Facilities in China

Science.gov (United States)

Guiquan, Xi; Lin, Cong; Xuehui, Jin

2018-05-01

As an important platform for scientific and technological development, large -scale scientific facilities are the cornerstone of technological innovation and a guarantee for economic and social development. Researching management of large-scale scientific facilities can play a key role in scientific research, sociology and key national strategy. This paper reviews the characteristics of large-scale scientific facilities, and summarizes development status of China's large-scale scientific facilities. At last, the construction, management, operation and evaluation of large-scale scientific facilities is analyzed from the perspective of sustainable development.

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

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

Large scale high strain-rate tests of concrete

Directory of Open Access Journals (Sweden)

Kiefer R.

2012-08-01

Full Text Available This work presents the stages of development of some innovative equipment, based on Hopkinson bar techniques, for performing large scale dynamic tests of concrete specimens. The activity is centered at the recently upgraded HOPLAB facility, which is basically a split Hopkinson bar with a total length of approximately 200 m and with bar diameters of 72 mm. Through pre-tensioning and suddenly releasing a steel cable, force pulses of up to 2 MN, 250 μs rise time and 40 ms duration can be generated and applied to the specimen tested. The dynamic compression loading has first been treated and several modifications in the basic configuration have been introduced. Twin incident and transmitter bars have been installed with strong steel plates at their ends where large specimens can be accommodated. A series of calibration and qualification tests has been conducted and the first real tests on concrete cylindrical specimens of 20cm diameter and up to 40cm length have commenced. Preliminary results from the analysis of the recorded signals indicate proper Hopkinson bar testing conditions and reliable functioning of the facility.

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...

GENIUS-TF: a test facility for the GENIUS project

International Nuclear Information System (INIS)

Klapdor-Kleingrothaus, H.V.; Baudis, L.; Dietz, A.; Heusser, G.; Krivosheina, I.; Majorovits, B.; Strecker, H.

2002-01-01

GENIUS is a proposal for a large scale detector of rare events. As a first step of the experiment, a small test version, the Genius Test-Facility will be built at the Laboratori Nazionali del Gran Sasso. With about 40 kg of natural Ge detectors operated in liquid nitrogen, GENIUS-TF could exclude (or directly confirm) the DAMA annual modulation seasonal modulation signature within about 2 yr of measurement using both, signal and signature of the claimed WIMP Dark Matter. The construction of the experiment has already been started, and four 2.5 kg germanium detectors with an extreme low threshold of 500 eV have been produced

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.

Full Scale Test of a SSP 34m boxgirder 2

DEFF Research Database (Denmark)

Jensen, Find Mølholt; Branner, Kim; Nielsen, Per Hørlyk

was part of a proof of concept investigation for a patent. The tests were performed at the Blaest test facility in August 2007. The tests are an important part of a research project established in cooperation between Risø National Laboratory for sustainable energy – Technical university of Denmark, SSP......This report presents the setup and result from three static full-scale tests of the reinforced glass fiber/epoxy box girder used in a 34m wind turbine blade. One test was without reinforcement one with cap reinforcement and the final test was with rib reinforcement. The cap reinforcement test......-Technology A/S and Blaest (Blade test centre A/S) and it has been performed as a part of Find Mølholt Jensen’s PhD thesis. This report is the second data report containing the complete test data for the three full-scale tests. This report deals only with the test methods and the obtained results...

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.

Lead coolant test facility systems design, thermal hydraulic analysis and cost estimate

Energy Technology Data Exchange (ETDEWEB)

Khericha, Soli, E-mail: slk2@inel.gov [Battelle Energy Alliance, LLC, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Harvego, Edwin; Svoboda, John; Evans, Robert [Battelle Energy Alliance, LLC, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Dalling, Ryan [ExxonMobil Gas and Power Marketing, Houston, TX 77069 (United States)

2012-01-15

The Idaho National Laboratory prepared a 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 coolant. Based on review of current world lead or lead-bismuth test facilities and research needs listed in the Generation IV Roadmap, five broad areas of requirements were identified as listed below: Bullet Develop and demonstrate feasibility of submerged heat exchanger. Bullet Develop and demonstrate open-lattice flow in electrically heated core. Bullet Develop and demonstrate chemistry control. Bullet Demonstrate safe operation. Bullet Provision for future testing. This paper discusses the preliminary design of systems, thermal hydraulic analysis, and simplified cost estimated. 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 4200 Degree-Sign 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 (in 2006 $). It is also estimated that the facility will require two years to be constructed and ready for operation.

Full-Scale Mark II CRT program data report No. 11 (TEST 1204)

International Nuclear Information System (INIS)

Kukita, Yutaka; Takeshita, Isao; Yamamoto, Nobuo; Namatame, Ken; Shiba, Masayoshi

1981-03-01

Recorded data for TEST 1204 conducted on the Full-Scale Mark II CRT (Containment Response Test) Facility are presented. The TEST 1204 is the fourth test run of a series of steam discharge pool swell tests. The test conditions are similar to those of the TEST 1203 except for lower initial pool temperature. The test was successful and the maximum level of pool surface was fairly lower than in the TEST 1203 due to the lower pool temperature. (author)

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.

Use of Ground Penetrating Radar at the FAA's National Airport Pavement Test Facility

Science.gov (United States)

Injun, Song

2015-04-01

The Federal Aviation Administration (FAA) in the United States has used a ground-coupled Ground Penetrating Radar (GPR) at the National Airport Pavement Test Facility (NAPTF) since 2005. One of the primary objectives of the testing at the facility is to provide full-scale pavement response and failure information for use in airplane landing gear design and configuration studies. During the traffic testing at the facility, a GSSI GPR system was used to develop new procedures for monitoring Hot Mix Asphalt (HMA) pavement density changes that is directly related to pavement failure. After reviewing current setups for data acquisition software and procedures for identifying different pavement layers, dielectric constant and pavement thickness were selected as dominant parameters controlling HMA properties provided by GPR. A new methodology showing HMA density changes in terms of dielectric constant variations, called dielectric sweep test, was developed and applied in full-scale pavement test. The dielectric constant changes were successfully monitored with increasing airplane traffic numbers. The changes were compared to pavement performance data (permanent deformation). The measured dielectric constants based on the known HMA thicknesses were also compared with computed dielectric constants using an equation from ASTM D4748-98 Standard Test Method for Determining the Thickness of Bound Pavement Layers Using Short-Pulse Radar. Six inches diameter cylindrical cores were taken after construction and traffic testing for the HMA layer bulk specific gravity. The measured bulk specific gravity was also compared to monitor HMA density changes caused by aircraft traffic conditions. Additionally this presentation will review the applications of the FAA's ground-coupled GPR on embedded rebar identification in concrete pavement, sewer pipes in soil, and gage identifications in 3D plots.

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...

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

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.

Development of turbopump cavitation performance test facility and the test of inducer performance

International Nuclear Information System (INIS)

Sohn, Dong Kee; Kim, Chun Tak; Yoon, Min Soo; Cha, Bong Jun; Kim, Jin Han; Yang, Soo Seok

2001-01-01

A performance test facility for turbopump inducer cavitation was developed and the inducer cavitation performance tests were performed. Major components of the performance test facility are driving unit, test section, piping, water tank, and data acquisition and control system. The maximum of testing capability of this facility are as follows: flow rate - 30kg/s; pressure - 13 bar, rotational speed - 10,000rpm. This cavitation test facility is characterized by the booster pump installed at the outlet of the pump that extends the flow rate range, and by the pressure control system that makes the line pressure down to vapor pressure. The vacuum pump is used for removing the dissolved air in the water as well as the line pressure. Performance tests were carried out and preliminary data of test model inducer were obtained. The cavitation performance test and cavitation bubble flow visualization were also made. This facility is originally designed for turbopump inducer performance test and cavitation test. However it can be applied to the pump impeller performance test in the future with little modification

Analysis of panthers full-scale heat transfer tests with RELAP5

International Nuclear Information System (INIS)

Parlatan, Y.; Boyer, B.D.; Jo, J.; Rohatgi, S.

1996-01-01

The RELAP5 code is being assessed on the full-scale Passive Containment Cooling System (PCCS) in the Performance ANalysis and Testing of HEat Removal Systems (PANTHERS) facility at Societa Informazioni Termoidrauliche (SIET) in Italy. PANTHERS is a test facility with fall-size prototype beat exchangers for the PCCS in support of the General Electric's (GE) Simplified Boiling Water Reactor (SBWR) program. PANTHERS tests with a low noncondensable gas concentration and with a high noncondensable gas concentration were analyzed with RELAP5. The results showed that beat transfer rate decreases significantly along the PCCS tubes. In the test case with a higher inlet noncondensable gas fraction, the PCCS removed 35% less heat than in the test case with the lower noncondensable gas fraction. The dominant resistance to the overall heat transfer is the condensation beat transfer resistance inside the tubes. This resistance increased by about 5-fold between the inlet and exit of the tube due to the build up of noncondensable gases along the tube. The RELAP5 calculations also predicted that 4% to 5% of the heat removed to the PCCS pool occurs in the inlet steam piping and PCCS upper and lower headers. These piping needs to be modeled for other tests systems. The full-scale PANTHERS predictions are also compared against 1/400 scale GIRAFFE tests. GIRAFFE has 33% larger heat surface area, but its efficiency is only 15% and 23% higher than PANTHERS for the two cases analyzed This was explained by the high heat transfer resistance inside the tubes near the exit

Comparison of steam-generator liquid holdup and core uncovery in two facilities of differing scale

International Nuclear Information System (INIS)

Motley, F.; Schultz, R.

1987-01-01

This paper reports on Run SB-CL-05, a test similar to Semiscale Run S-UT-8. The test results show that the core was uncovered briefly during the accident and that the rods overheated at certain core locations. Liquid holdup on the upflow side of the steam-generator tubes was observed. After the loop seal cleared, the core refilled and the rods cooled. These behaviors were similar to those observed in the Semiscale run. The Large-Scale Test Facility (LSTF) Run SB-CL-06 is a counterpart test to Semiscale Run S-LH-01. The comparison of the results of both tests shows similar phenomena. The similarity of phenomena in these two facilities build confidence that these results can be expected to occur in a PWR. Similar holdup has now been observed in the 6 tubes of Semiscale and in the 141 tubes of LSTF. It is now more believable that holdup may occur in a full-scale steam generator with 3000 or more tubes. These results confirm the scaling of these phenomena from Semiscale (1/1705) to LSTF (1/48). The TRAC results for SB-CL-05 are in reasonable agreement with the test data. TRAC predicted the core uncovery and resulting rod heatup. The liquid holdup on the upflow side of the steam-generator tubes was also correctly predicted. The clearing of the loop seal allowed core recovery and cooled the overheated rods just as it had in the data. The TRAC analysis results of Run SB-CL-05 are similar to those from Semiscale Run S-UT-8. The ability of the TRAC code to calculate the phenomena equally well in the two experiments of different scales confirms the scalability of the many models in the code that are important in calculating this small break

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

Initial operation of a solar heating and cooling system in a full-scale solar building test facility

Science.gov (United States)

Knoll, R. H.; Miao, D.; Hamlet, I. L.; Jensen, R. N.

1976-01-01

The Solar Building Test Facility (SBTF) located at Hampton, Virginia became operational in early summer of 1976. This facility is a joint effort by NASA-Lewis and NASA-Langley to advance the technology for heating and cooling of office buildings with solar energy. Its purposes are to (1) test system components which include high-performing collectors, (2) test performance of complete solar heating and cooling system, (3) investigate component interactions and (4) investigate durability, maintenance and reliability of components. The SBTF consists of a 50,000 square foot office building modified to accept solar heated water for operation of an absorption air conditioner and for the baseboard heating system. A 12,666 square foot solar collector field with a 30,000 gallon storage tank provides the solar heated water. A description of the system and the collectors selected is given here, along with the objectives, test approach, expected system performance and some preliminary results.

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

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

Annual report on experimental operations and maintenance of mock-up model test facility with a full-scale reaction tube for HTTR hydrogen production system in 2004 fiscal year (Contract research)

International Nuclear Information System (INIS)

Hayashi, Koji; Ohashi, Hirofumi; Morisaki, Norihiro; Kato, Michio; Aita, Hideki; Takeda, Tetsuaki; Nishihara, Tetsuo; Inaba, Yoshitomo; Takada, Shoji; Inagaki, Yoshiyuki

2006-03-01

This is annual report on the experimental test operations and maintenances of the mock-up test facility with a full-scale reaction tube for the HTTR hydrogen production system in 2004 fiscal year. The improvement work of catalyst dust filter in combustion system was performed in May 2004, and the performance was confirmed. The sixth experimental test operation was performed from June to July 2004. Periodic inspections on boiler equipment and high-pressure gas production facilities were performed from end of July to September 2004. The seventh experimental test operation was performed from October to December 2004 for chemical reaction shutdown test. From the results, a behavior of the helium-gas cooling system, consists of steam generator and radiator, during chemical reaction shutdown was confirmed. This report is summarized with the outline and the results of the test, maintenance works and inspections, and operation records in mentioned above. (author)

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

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.

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

Experiments to investigate direct containment heating phenomena with scaled models of the Zion Nuclear Power Plant in the Surtsey Test Facility

International Nuclear Information System (INIS)

Allen, M.D.; Pilch, M.M.; Blanchat, T.K.; Griffith, R.O.; Nichols, R.T.

1994-05-01

The Surtsey Facility at Sandia National Laboratories (SNL) is used to perform scaled experiments that simulate hypothetical high-pressure melt ejection (HPME) accidents in a nuclear power plant (NPP). These experiments are designed to investigate the effect of specific phenomena associated with direct containment heating (DCH) on the containment load, such as the effect of physical scale, prototypic subcompartment structures, water in the cavity, and hydrogen generation and combustion. In the Integral Effects Test (IET) series, 1:10 linear scale models of the Zion NPP structures were constructed in the Surtsey vessel. The RPV was modeled with a steel pressure vessel that had a hemispherical bottom head, which had a 4-cm hole in the bottom head that simulated the final ablated hole that would be formed by ejection of an instrument guide tube in a severe NPP accident. Iron/alumina/chromium thermite was used to simulate molten corium that would accumulate on the bottom head of an actual RPV. The chemically reactive melt simulant was ejected by high-pressure steam from the RPV model into the scaled reactor cavity. Debris was then entrained through the instrument tunnel into the subcompartment structures and the upper dome of the simulated reactor containment building. The results of the IET experiments are given in this report

Experiments to investigate direct containment heating phenomena with scaled models of the Zion Nuclear Power Plant in the Surtsey Test Facility

Energy Technology Data Exchange (ETDEWEB)

Allen, M.D.; Pilch, M.M.; Blanchat, T.K.; Griffith, R.O. [Sandia National Labs., Albuquerque, NM (United States); Nichols, R.T. [Ktech Corp., Albuquerque, NM (United States)

1994-05-01

The Surtsey Facility at Sandia National Laboratories (SNL) is used to perform scaled experiments that simulate hypothetical high-pressure melt ejection (HPME) accidents in a nuclear power plant (NPP). These experiments are designed to investigate the effect of specific phenomena associated with direct containment heating (DCH) on the containment load, such as the effect of physical scale, prototypic subcompartment structures, water in the cavity, and hydrogen generation and combustion. In the Integral Effects Test (IET) series, 1:10 linear scale models of the Zion NPP structures were constructed in the Surtsey vessel. The RPV was modeled with a steel pressure vessel that had a hemispherical bottom head, which had a 4-cm hole in the bottom head that simulated the final ablated hole that would be formed by ejection of an instrument guide tube in a severe NPP accident. Iron/alumina/chromium thermite was used to simulate molten corium that would accumulate on the bottom head of an actual RPV. The chemically reactive melt simulant was ejected by high-pressure steam from the RPV model into the scaled reactor cavity. Debris was then entrained through the instrument tunnel into the subcompartment structures and the upper dome of the simulated reactor containment building. The results of the IET experiments are given in this report.

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

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

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.

The Inter Facility Testing of a Standard Oscillating Water Column (OWC) Type Wave Energy Converter (WEC)

DEFF Research Database (Denmark)

Andersen, Morten Thøtt; Thomsen, Jonas Bjerg

This report describes the behavior and preliminary performance of a simplified standard oscillating water column (OWC) wave energy converter (WEC). The same tests will be conducted at different scales at 6 different test facilities and the results obtained will be used for comparison. This project...

Conceptual development of a test facility for spent fuel management

Energy Technology Data Exchange (ETDEWEB)

Park, S.W.; Lee, H.H.; Lee, J.Y.; Lee, J.S.; Ro, S.G. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1997-12-31

Spent fuel management is an important issue for nuclear power program, requiring careful planning and implementation. With the wait-and-see policy on spent fuel management in Korea, research efforts are directed at KAERI to develop advanced technologies for safer and more efficient management of the accumulating spent fuels. In support of these research perspectives, a test facility of pilot scale is being developed with provisions for integral demonstration of a multitude of technical functions required for spent fuel management. The facility, baptized SMART (Spent fuel MAnagement technology Research and Test facility), is to be capable of handling full size assembly of spent PWR fuel (as well as CANDU fuel) with a maximum capacity of 10 MTU/y (about 24 assemblies of PWR type). Major functions of the facility are consolidation of spent PWR fuel assembly into a half-volume package and optionally transformation of the fuel rod into a fuel of CANDU type (called DUPIC). Objectives of these functions are to demonstrate volume reduction of spent fuel (for either longer-term dry storage or direct disposal ) in the former case and direct refabrication of the spent PWR fuel into CANDU-type DUPIC fuel for reuse in CANDU reactors in the latter case, respectively. In addition to these major functions, there are other associated technologies to be demonstrated : such as waste treatment, remote maintenance, safeguards, etc. As the facility is to demonstrate not only the functional processes but also the safety and efficiency of the test operations, engineering criteria equivalent to industrial standards are incorporated in the design concept. The hot cell structure enclosing the radioactive materials is configured in such way to maximize costs within the given functional and operational requirements. (author). 3 tabs., 4 figs.

Conceptual development of a test facility for spent fuel management

International Nuclear Information System (INIS)

Park, S.W.; Lee, H.H.; Lee, J.Y.; Lee, J.S.; Ro, S.G.

1997-01-01

Spent fuel management is an important issue for nuclear power program, requiring careful planning and implementation. With the wait-and-see policy on spent fuel management in Korea, research efforts are directed at KAERI to develop advanced technologies for safer and more efficient management of the accumulating spent fuels. In support of these research perspectives, a test facility of pilot scale is being developed with provisions for integral demonstration of a multitude of technical functions required for spent fuel management. The facility, baptized SMART (Spent fuel MAnagement technology Research and Test facility), is to be capable of handling full size assembly of spent PWR fuel (as well as CANDU fuel) with a maximum capacity of 10 MTU/y (about 24 assemblies of PWR type). Major functions of the facility are consolidation of spent PWR fuel assembly into a half-volume package and optionally transformation of the fuel rod into a fuel of CANDU type (called DUPIC). Objectives of these functions are to demonstrate volume reduction of spent fuel (for either longer-term dry storage or direct disposal ) in the former case and direct refabrication of the spent PWR fuel into CANDU-type DUPIC fuel for reuse in CANDU reactors in the latter case, respectively. In addition to these major functions, there are other associated technologies to be demonstrated : such as waste treatment, remote maintenance, safeguards, etc. As the facility is to demonstrate not only the functional processes but also the safety and efficiency of the test operations, engineering criteria equivalent to industrial standards are incorporated in the design concept. The hot cell structure enclosing the radioactive materials is configured in such way to maximize costs within the given functional and operational requirements. (author). 3 tabs., 4 figs

A Counterpart Test of the VISTA-ITL SB-SIS-07 using the FESTA (SMART-ITL) Facility for the SMART Design

Energy Technology Data Exchange (ETDEWEB)

Park, Hyunsik; Bae, Hwang; Ryu, Seonguk; Yi, Sungjae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Ko, Yungjoo [System Engineering and Technology Co., Ltd., Daejeon (Korea, Republic of)

2014-05-15

The SMART design is characterized by the introduction of simplified and improved safety systems such as PRHRS, and its integral arrangement of the reactor vessel assembly. The integrated reactor vessel assembly consists of a reactor core, eight steam generators, four reactor coolant pumps, and a steam pressurizer, and it has four trains of a secondary system and passive residual heat removal system. The SMART design was fully assessed through various thermal-hydraulic validation tests during the licensing review process. Among them, a small-scale integral effect test facility of VISTA-ITL was used to investigate various thermal-hydraulic phenomena during design basis accident scenarios such as SBLOCA and CLOF, and the acquired data were used to validate the related thermal-hydraulic models of the safety analysis codes such as MARS-KS and TASS/SMR-S. The VISTA-ITL facility is a reduced height, 1/1310-volume scaled test facility with a single train of a secondary system and PRHRS. Recently, a large-scale integral effect test facility of FESTA was constructed in KAERI and a set of integral effect tests for design basis accident scenarios is also being performed. The role of FESTA could be extended to examine and verify the normal, abnormal, and emergency operating procedures required during the construction phases of SMART. It could be called as a counterpart test when similar experiments are performed in differently scaled facilities. It is clear that transient scenarios measured in the experimental rigs cannot be directly extrapolated to the plant conditions. Nevertheless one of the objectives of counter-part tests is to evaluate the influence of the geometric dimensions of the loops upon the evolution of a given accident. In this paper the counterpart test results with both the VISTA-ITL and FESTA facilities will be compared on a SBLOCA scenario of the SIS line break for the SMART design. As counter-part tests for an SBLOCA for the SMART design, two integral effect

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.)

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

Capabilities of the Large-Scale Sediment Transport Facility

Science.gov (United States)

2016-04-01

pump flow meters, sediment trap weigh tanks , and beach profiling lidar. A detailed discussion of the original LSTF features and capabilities can be...ERDC/CHL CHETN-I-88 April 2016 Approved for public release; distribution is unlimited. Capabilities of the Large-Scale Sediment Transport...describes the Large-Scale Sediment Transport Facility (LSTF) and recent upgrades to the measurement systems. The purpose of these upgrades was to increase

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

An Integrated Assessment of Location-Dependent Scaling for Microalgae Biofuel Production Facilities

Energy Technology Data Exchange (ETDEWEB)

Coleman, Andre M.; Abodeely, Jared; Skaggs, Richard; Moeglein, William AM; Newby, Deborah T.; Venteris, Erik R.; Wigmosta, Mark S.

2014-06-19

Successful development of a large-scale microalgae-based biofuels industry requires comprehensive analysis and understanding of the feedstock supply chain—from facility siting/design through processing/upgrading of the feedstock to a fuel product. The evolution from pilot-scale production facilities to energy-scale operations presents many multi-disciplinary challenges, including a sustainable supply of water and nutrients, operational and infrastructure logistics, and economic competitiveness with petroleum-based fuels. These challenges are addressed in part by applying the Integrated Assessment Framework (IAF)—an integrated multi-scale modeling, analysis, and data management suite—to address key issues in developing and operating an open-pond facility by analyzing how variability and uncertainty in space and time affect algal feedstock production rates, and determining the site-specific “optimum” facility scale to minimize capital and operational expenses. This approach explicitly and systematically assesses the interdependence of biofuel production potential, associated resource requirements, and production system design trade-offs. The IAF was applied to a set of sites previously identified as having the potential to cumulatively produce 5 billion-gallons/year in the southeastern U.S. and results indicate costs can be reduced by selecting the most effective processing technology pathway and scaling downstream processing capabilities to fit site-specific growing conditions, available resources, and algal strains.

Local control unit for ITER-India gyrotron test facility (IIGTF)

Energy Technology Data Exchange (ETDEWEB)

Rathod, Vipal, E-mail: vipal.rathod@iter-india.org; Shah, Ronak; Mandge, Deepak; Parmar, Rajvi; Rao, S.L.

2016-11-15

Highlights: • A dedicated full scale ITER prototype Local Control Unit for ITER-India Gyrotron test facility. • National Instruments® make PXIe system for real time control & data acquisition and Siemens® PLC for sequence control function. • Hardwired FPGA based fast protection interlock system. • High speed analog fiber optical transmission link using V/F and F/V technique. • Software framework based on LabVIEW™ platform and ITER CODAC Core System. - Abstract: Electron Cyclotron system on ITER, is one of the important RF ancillary systems based on high power Gyrotron RF sources, that is used for plasma heating and current drive applications. To operate a Gyrotron source, various auxiliary systems and services such as Super Conducting Magnet set, High Voltage Power Supplies, Auxiliary Power Supplies, Waveguide components, Cooling water system and a Local Control Unit (LCU) are required. The LCU plays a very crucial role for the safe and reliable operation of Gyrotron system. A dedicated full scale ITER prototype LCU is being developed for testing and commissioning of an ITER like Test Gyrotron at ITER-India Gyrotron Test facility (IIGTF). The main functions of LCU include Sequence Control, Local Interlock Protection and Real Time Data Acquisition. PLC based slow controller is used for implementing the Sequence Control & Slow Interlock functions. Critical Protection Interlocks are required to have a response time of <10 μs and are implemented using custom built hardware and PXIe based fast controller. Also PXIe system is used for implementing Real Time Data Acquisition function that is required to have slow and fast acquisition with online visualization and off line analysis facility. A Signal Conditioning Unit (SCU) is used to interface and faithfully transmit the field signals to the remote control systems. Necessary controller hardware is procured and several pre-prototype developments have been taken up to establish the critical subsystems such as

Local control unit for ITER-India gyrotron test facility (IIGTF)

International Nuclear Information System (INIS)

Rathod, Vipal; Shah, Ronak; Mandge, Deepak; Parmar, Rajvi; Rao, S.L.

2016-01-01

Highlights: • A dedicated full scale ITER prototype Local Control Unit for ITER-India Gyrotron test facility. • National Instruments® make PXIe system for real time control & data acquisition and Siemens® PLC for sequence control function. • Hardwired FPGA based fast protection interlock system. • High speed analog fiber optical transmission link using V/F and F/V technique. • Software framework based on LabVIEW™ platform and ITER CODAC Core System. - Abstract: Electron Cyclotron system on ITER, is one of the important RF ancillary systems based on high power Gyrotron RF sources, that is used for plasma heating and current drive applications. To operate a Gyrotron source, various auxiliary systems and services such as Super Conducting Magnet set, High Voltage Power Supplies, Auxiliary Power Supplies, Waveguide components, Cooling water system and a Local Control Unit (LCU) are required. The LCU plays a very crucial role for the safe and reliable operation of Gyrotron system. A dedicated full scale ITER prototype LCU is being developed for testing and commissioning of an ITER like Test Gyrotron at ITER-India Gyrotron Test facility (IIGTF). The main functions of LCU include Sequence Control, Local Interlock Protection and Real Time Data Acquisition. PLC based slow controller is used for implementing the Sequence Control & Slow Interlock functions. Critical Protection Interlocks are required to have a response time of <10 μs and are implemented using custom built hardware and PXIe based fast controller. Also PXIe system is used for implementing Real Time Data Acquisition function that is required to have slow and fast acquisition with online visualization and off line analysis facility. A Signal Conditioning Unit (SCU) is used to interface and faithfully transmit the field signals to the remote control systems. Necessary controller hardware is procured and several pre-prototype developments have been taken up to establish the critical subsystems such as

Pore pressure measurement plan of near field rock used on three dimensional groundwater flow analysis in demonstration test of cavern type disposal facility

International Nuclear Information System (INIS)

Onuma, Kazuhiro; Terada, Kenji; Matsumura, Katsuhide; Koyama, Toshihiro; Yajima, Kazuaki

2008-01-01

Demonstration test of underground cavern type disposal facilities is planed though carrying out construction of full scale engineering barrier system which simulated in the underground space in full scale and under actual environment. This test consists of three part, these are construction test, performance test and measurement test. Behavior of near field rock mass is measured about hydrological behavior under and after construction to evaluate effect at test facility. To make plan of pore pressure measurement, three dimensional groundwater flow analysis has been carried out. Based on comparison of analysis before and after test, detail plan has been studied. (author)

Qualification test for ITER HCCR-TBS mockups with high heat flux test facility

Energy Technology Data Exchange (ETDEWEB)

Kim, Suk-Kwon, E-mail: skkim93@kaeri.re.kr [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Park, Seong Dae; Jin, Hyung Gon; Lee, Eo Hwak; Yoon, Jae-Sung; Lee, Dong Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Cho, Seungyon [National Fusion Research Institute, Daejeon (Korea, Republic of)

2016-11-01

Highlights: • The test mockups for ITER HCCR (Helium Cooled Ceramic Reflector) TBS (Test Blanket System) in Korea were designed and fabricated. • A thermo-hydraulic analysis was performed using a high heat flux test facility by using electron beam. • The plan for qualification tests was developed to evaluate the thermo-hydraulic efficiency in accordance with the requirements of the ITER Organization. - Abstract: The test mockups for ITER HCCR (Helium Cooled Ceramic Reflector) TBS (Test Blanket System) in Korea were designed and fabricated, and an integrity and thermo-hydraulic performance test should be completed under the same or similar operation conditions of ITER. The test plan for a thermo-hydraulic analysis was developed by using a high heat flux test facility, called the Korean heat load test facility by using electron beam (KoHLT-EB). This facility is utilized for a qualification test of the plasma facing component (PFC) for the ITER first wall and DEMO divertor, and for the thermo-hydraulic experiments. In this work, KoHLT-EB will be used for the plan of the performance qualification test of the ITER HCCR-TBS mockups. This qualification tests should be performed to evaluate the thermo-hydraulic efficiency in accordance with the requirements of the ITER Organization (IO), which describe the specifications and qualifications of the heat flux test facility and test procedure for ITER PFC.

A study of return to saturation oscillations in the OSU APEX thermal hydraulic testing facility

Science.gov (United States)

Franz, Scott Cameron

The purpose of this paper is to describe the flow oscillations which occur in the AP600 long term cooling test facility at Oregon State University. The AP600 system is an advanced pressurized water reactor design utilizing passive emergency cooling systems. A few hours after the initiation of a cold leg break, the passive cooling systems inject gravity fed cold water at a rate allowing steam production in the reactor vessel. Steam production in the core causes the pressure in the upper head to increase leading to flow oscillations in all the connecting reactor systems. This paper will show that the oscillations have a definite region of onset and termination for specific conditions in the APEX testing facility. Tests performed at high powers, high elevation breaks, and small break sizes do not exhibit oscillations. The APOS (Advanced Plant Oscillation Simulator) computer code has been developed using a quasi-steady state analysis for flows and a transient analysis for the core node energy balance. The pressure in the reactor head is calculated using a modified perfect gas analysis. For tank liquid inventories, a simple conservation of mass analysis is used to estimate the tank elevations. Simulation logic gleaned from APEX data and photographic evidence have been incorporated into the code to predict termination of the oscillations. Areas which would make the work more complete include a better understanding of two-phase fluid behavior for a top offtake on a pipe, more instrumentation in the core region of the APEX testing facility, and a clearer understanding of fluid conditions in the reactor barrel. Scaling of the oscillations onset and pressure amplitude are relatively straightforward, but termination and period are difficult to scale to the full AP600 plant. Differences in the core power profile and other geometrical differences between the testing facility and the actual plant make the scaling of this phenomenon to the actual plant conditions very difficult.

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

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

Fermilab R and D test facility for SSC [Superconducting Super Collider] magnets

International Nuclear Information System (INIS)

Strait, J.; Bleadon, M.; Hanft, R.; Lamm, M.; McGuire, K.; Mantsch, P.; Mazur, P.O.; Orris, D.; Pachnik, J.

1989-02-01

The test facility used for R and D testing of full scale development dipole magnets for the SSC is described. The Fermilab Magnet Test Facility, originally built for production testing of Tevatron magnets, has been substantially modified to allow testing also of SSC magnets. Two of the original six test stands have been rebuilt to accommodate testing of SSC magnets at pressures between 1.3 Atm and 4 Atm and at temperatures between 1.8 K and 4.8 K and the power system has been modified to allow operation to at least 8 kA. Recent magnets have been heavily instrumented with voltage taps to allow detailed study of quench location and propagation and with strain gage based stress, force and motion transducers. A data acquisition system has been built with a capacity to read from each SSC test stand up to 220 electrical quench signals, 32 dynamic pressure, temperature and mechanical transducer signals during quench and up to 200 high precision, low time resolution, pressure, temperature and mechanical transducer signals. The quench detection and protection systems is also described. 23 refs., 4 figs., 2 tabs

Local flow distribution analysis inside the reactor pools of KALIMER-600 and PDRC performance test facility

International Nuclear Information System (INIS)

Jeong, Ji Hwan; Hwang, Seong Won; Choi, Kyeong Sik

2010-05-01

In the study, 3-dimensional thermal hydraulic analysis was carried out focusing on the thermal hydraulic behavior inside the reactor pools for both KALIMER-600 and one-fifth scale-down test facility. STAR-CD, one of the commercial CFD codes, was used to analyze 3-dimensional incompressible steady-state thermal hydraulic behavior in both designs of KALIMER-600 and the scale-down test facility. In the KALIMER-600 CFD analysis, the pressure drops in the core and IHX gave a good agreement within 1% error range. It was found that the porous media model was appropriate to analyze the pressure distribution inside reactor core and IHX. Also, a validation analysis showed the pressure drop through the porous media under the condition of 80% flow rate and thermal power was calculated 64% less than in 100% condition giving a physically reasonable analytic result. Since the temperatures in the hot-side pool and cold-side pool were estimated to be very close to 540 and 390 .deg. C specified on the design values respectively, the CFD models of heat source and sink was confirmed. Through the study, the methodology of 3-dimensional CFD analysis about KALIMER-600 has been established and proven. Performed with the methodology, the analysis data such as flow velocity, temperature and pressure distribution were compared by normalizing those data for the actual sized modeling and scale-down modeling. As a result, the characteristics of thermal hydraulic behavior were almost identical for the actual sized modeling and scale-down modeling and the similarity scaling law used in the design of the sodium test facility by KAERI was found to be correct

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.

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

Test facilities for radioactive materials transport packages (Chicago Bridge and Iron, USA)

International Nuclear Information System (INIS)

Gallagher, T.A.

1991-01-01

Chicago Bridge and Iron, Research and Development Center located in Plainfield, Illinois offers the total capabilities required to perform design verification testing of hazardous waste shipping containers. The tests, defined in the United States Code of Federal Regulations, Title 10, Part 71 (10CFR71), include vertical drop tests, puncture tests, crush tests, immersion tests, thermal tests, and container leak rate tests. Container structural design analysis, container manufacturing analysis, materials development testing plus dimensional analysis of individual components is also available. The test facilities meet or exceed the requirements given in the International Atomic Energy Agency (IAEA) Safety Guide, Safety Series No. 37, 1987. Additional capabilities for the design and fabrication of scale models and components for the test programme are also presented. (author)

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

DOE Coal Gasification Multi-Test Facility: fossil fuel processing technical/professional services

Energy Technology Data Exchange (ETDEWEB)

Hefferan, J.K.; Lee, G.Y.; Boesch, L.P.; James, R.B.; Rode, R.R.; Walters, A.B.

1979-07-13

A conceptual design, including process descriptions, heat and material balances, process flow diagrams, utility requirements, schedule, capital and operating cost estimate, and alternative design considerations, is presented for the DOE Coal Gasification Multi-Test Facility (GMTF). The GMTF, an engineering scale facility, is to provide a complete plant into which different types of gasifiers and conversion/synthesis equipment can be readily integrated for testing in an operational environment at relatively low cost. The design allows for operation of several gasifiers simultaneously at a total coal throughput of 2500 tons/day; individual gasifiers operate at up to 1200 tons/day and 600 psig using air or oxygen. Ten different test gasifiers can be in place at the facility, but only three can be operated at one time. The GMTF can produce a spectrum of saleable products, including low Btu, synthesis and pipeline gases, hydrogen (for fuel cells or hydrogasification), methanol, gasoline, diesel and fuel oils, organic chemicals, and electrical power (potentially). In 1979 dollars, the base facility requires a $288 million capital investment for common-use units, $193 million for four gasification units and four synthesis units, and $305 million for six years of operation. Critical reviews of detailed vendor designs are appended for a methanol synthesis unit, three entrained flow gasifiers, a fluidized bed gasifier, and a hydrogasifier/slag-bath gasifier.

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)

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

Scaling Studies for High Temperature Test Facility and Modular High Temperature Gas-Cooled Reactor

Energy Technology Data Exchange (ETDEWEB)

Richard R. Schult; Paul D. Bayless; Richard W. Johnson; James R. Wolf; Brian Woods

2012-02-01

The Oregon State University (OSU) High Temperature Test Facility (HTTF) is an integral experimental facility that will be constructed on the OSU campus in Corvallis, Oregon. The HTTF project was initiated, by the U.S. Nuclear Regulatory Commission (NRC), on September 5, 2008 as Task 4 of the 5-year High Temperature Gas Reactor Cooperative Agreement via NRC Contract 04-08-138. Until August, 2010, when a DOE contract was initiated to fund additional capabilities for the HTTF project, all of the funding support for the HTTF was provided by the NRC via their cooperative agreement. The U.S. Department of Energy (DOE) began their involvement with the HTTF project in late 2009 via the Next Generation Nuclear Plant (NGNP) project. Because the NRC's interests in HTTF experiments were only centered on the depressurized conduction cooldown (DCC) scenario, NGNP involvement focused on expanding the experimental envelope of the HTTF to include steady-state operations and also the pressurized conduction cooldown (PCC).

E-SCAPE: A scale facility for liquid-metal, pool-type reactor thermal hydraulic investigations

Energy Technology Data Exchange (ETDEWEB)

Van Tichelen, Katrien, E-mail: kvtichel@sckcen.be [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Mirelli, Fabio, E-mail: fmirelli@sckcen.be [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Greco, Matteo, E-mail: mgreco@sckcen.be [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Viviani, Giorgia, E-mail: giorgiaviviani@gmail.com [University of Pisa, Lungarno Pacinotti 43, 56126 Pisa (Italy)

2015-08-15

Highlights: • The E-SCAPE facility is a thermal hydraulic scale model of the MYRRHA fast reactor. • The focus is on mixing and stratification in liquid-metal pool-type reactors. • Forced convection, natural convection and the transition are investigated. • Extensive instrumentation allows validation of computational models. • System thermal hydraulic and CFD models have been used for facility design. - Abstract: MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is a flexible fast-spectrum research reactor under design at SCK·CEN. MYRRHA is a pool-type reactor with lead bismuth eutectic (LBE) as primary coolant. The proper understanding of the thermal hydraulic phenomena occurring in the reactor pool is an important issue in the design and licensing of the MYRRHA system and liquid-metal cooled reactors by extension. Model experiments are necessary for understanding the physics, for validating experimental tools and to qualify the design for the licensing. The E-SCAPE (European SCAled Pool Experiment) facility at SCK·CEN is a thermal hydraulic 1/6-scale model of the MYRRHA reactor, with an electrical core simulator, cooled by LBE. It provides experimental feedback to the designers on the forced and natural circulation flow patterns. Moreover, it enables to validate the computational methods for their use with LBE. The paper will elaborate on the design of the E-SCAPE facility and its main parameters. Also the experimental matrix and the pre-test analysis using computational fluid dynamics (CFD) and system thermal hydraulics codes will be described.

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

Utilisation of real-scale renewable energy test facility for validation of generic wind turbine and wind power plant controller models

DEFF Research Database (Denmark)

Zeni, Lorenzo; Gevorgian, Vahan; Wallen, Robb

2016-01-01

This article presents an example of application of a modern test facility conceived for experiments regarding the integration of renewable energy in the power system. The capabilities of the test facility are used to validate dynamic simulation models of wind power plants and their controllers....... The models are based on standard and generic blocks. The successful validation of events related to the control of active power (control phenomena in...

Letter report: Pre-conceptual design study for a pilot-scale Non-Radioactive Low-Level Waste Vitrification Facility

International Nuclear Information System (INIS)

Thompson, R.A.; Morrissey, M.F.

1996-03-01

This report presents a pre-conceptual design study for a Non-Radioactive Low-Level Waste, Pilot-Scale Vitrification System. This pilot plant would support the development of a full-scale LLW Vitrification Facility and would ensure that the full-scale facility can meet its programmatic objectives. Use of the pilot facility will allow verification of process flowsheets, provide data for ensuring product quality, assist in scaling to full scale, and support full-scale start-up. The facility will vitrify simulated non-radioactive LLW in a manner functionally prototypic to the full-scale facility. This pre-conceptual design study does not fully define the LLW Pilot-Scale Vitrification System; rather, it estimates the funding required to build such a facility. This study includes identifying all equipment necessary. to prepare feed, deliver it into the melter, convert the feed to glass, prepare emissions for atmospheric release, and discharge and handle the glass. The conceived pilot facility includes support services and a structure to contain process equipment

Annual report on experimental operation of mock-up model test facility with a full-scale reaction tube for HTTR hydrogen production system in 2001 fiscal year (Contract research)

International Nuclear Information System (INIS)

Hayashi, Koji; Inagaki, Yoshiyuki; Kato, Michio; Fujisaki, Katsuo; Aita, Hideki; Takeda, Tetsuaki; Nishihara, Tetsuo; Inaba, Yoshitomo; Ohashi, Hirofumi; Katanishi, Shoji; Takada, Shoji; Shimizu, Akira; Morisaki, Norihiro; Sakaki, Akihiro; Maeda, Yukimasa; Sato, Hiroyuki

2005-06-01

This is an annual report on the experimental operation of the mock-up test facility with a full-scale reaction tube for the HTTR hydrogen production system in 2001 fiscal year. The first experimental operation was performed during two weeks from March 1, 2002 to March 13, 2002 to test on the thermal hydraulic performance of the steam reformer and also to train the operators. The thermal hydraulic performance test of the steam reformer was performed to evaluate the heat transfer characteristics between helium gas and process gas in the steam reformer. This report is summarized with an overview of the test, the results and its operation records. (author)

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...

TOP 01-1-011B Vehicle Test Facilities at Aberdeen Test Center and Yuma Test Center

Science.gov (United States)

2017-12-12

Test Center 400 Colleran Road Aberdeen Proving Ground, MD 21005-5059 U.S. Army Yuma Proving Ground Yuma Test Center 301 C. Street Yuma, AZ...22 2.6 Munson Test Area (MTA) ..................................................... 24 2.7 Land Vehicle Maintenance Facility...127 3.6 Maintenance Facilities ........................................................... 143

Scientific design of the test facility for the KNGR DVI line small break LOCA

Energy Technology Data Exchange (ETDEWEB)

Yun, Byong Jo; Park, Choon Kyung; Jun, Hyung Gil; Cho, Seok; Kwon, Tae Soon; Song, Chul Hwa; Kim, Jung Taek

1999-03-01

Scientific design of the experimental facility (OASIS) for the KNGR (Korea Next Generation Reactor) DVI line SB-LOCA simulation is carried out. Main purpose of the OASIS is to produce thermal-hydraulic data base for determining the best location of the DVI (Direct Vessel Injection) injection nozzle of the KNGR as well as verifying its design performance in view of the ECCS (Emergency Core Cooling System) effectiveness. The experimental facility is designed based on the Ishii's three-level scaling law. The facility has 1/4 height and 1/341 area scaling ratio. It corresponds to the volume scale of 1/1364. The power scaling is 1/682 and the system pressure is prototypic. The OASIS consists of a core, a downcomer, two steam generators, two pump simulators, a break simulator, a collection tank, primary piping as well as a circulation pump for initial test condition. Each component is designed based on the Ishill's global scaling and boundary flow scaling of mass, energy and momentum. In addition, local phenomena scaling is carried out for the design of major components to preserve key local phenomena in each component. Most of the key phenomena are well preserved in the OASIS. However, the local scaling analysis shows that distortions of the void fraction and mixture level can not be avoided in the core. It comes from the basic features of the Ishill's scaling law in case of the reduced-height simulation. However, it is expected that these distortions will be analyzed properly by a best estimate system analysis code. (Author). 22 refs., 20 tabs., 25 figs.

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.

Design of the Caltrans Seismic Response Modification Device (SRMD) test facility

International Nuclear Information System (INIS)

Benzoni, G.; Seible, F.

1998-01-01

In the Seismic retrofit design of California's Toll Bridges, seismic isolation is used in several bridges to limit the seismic force input into the superstructure and to avoid costly superstructure retrofit measures which would require partial lane closures and traffic interruptions. Isolation bearings and dampers of the size required for these large span bridges have not been built or tested to date. This paper describes the design and construction of a full scale testing facility which will allow the real-time 6-DOF dynamic characterization of the seismic response modification devices designed for California's Toll Bridges. (author)

Full-scale vibration tests of Atucha II N.P.P. Part I: objectives, instrumentation and test description

International Nuclear Information System (INIS)

Konno, T.; Tsugawa, T.; Sala, G.; Friebe, T.M.; Prato, C.A.; Godoy, A.R.

1995-01-01

The main purpose of the tests was to provide experimental data on the dynamic characteristics of the main reactor building and adjacent structures of a full-scale nuclear power plant built on deep Quaternary soil deposits. Test results were intended to provide a benchmark case for control and calibration of state-of-the-art numerical techniques used for engineering design of new plants and assessment of existing facilities. Interpretation of test results and calibration of numerical analyses are described in other associated papers. (author). 5 figs

Development of a Test Facility to Simulate the Reactor Flow Distribution of APR+

International Nuclear Information System (INIS)

Euh, D. J.; Cho, S.; Youn, Y. J.; Kim, J. T.; Kang, H. S.; Kwon, T. S.

2011-01-01

Recently a design of new reactor, APR+, is being developed, as an advanced type of APR1400. In order to analyze the thermal margin and hydraulic characteristics of APR+, quantification tests for flow and pressure distribution with a conservation of flow geometry are necessary. Hetsroni (1967) proposed four principal parameters for a hydraulic model representing a nuclear reactor prototype: geometry, relative roughness, Reynolds number, and Euler number. He concluded that the Euler number should be similar in the prototype and model under the preservation of the aspect ratio on the flow path. The effect of the Reynolds number at its higher values on the Euler number is rather small, since the dependency of the form and frictional loss coefficients on the Reynolds number is seen to be small. ABB-CE has carried out several reactor flow model test programs, mostly for its prototype reactors. A series of tests were conducted using a 3/16 scale reactor model. (see Lee et al., 2001). Lee et al (1991) performed experimental studies using a 1/5.03 scale reactor flow model of Yonggwang nuclear units 3 and 4. They showed that the measured data met the acceptance criteria and were suitable for their intended use in terms of performance and safety analyses. The design of current test facility was based on the conservation of Euler number which is a ratio of pressure drop to dynamic pressure with a sufficiently turbulent region having a high Reynolds number. By referring to the previous study, the APR+ design is linearly reduced to 1/5 ratio with a 1/2 of the velocity scale, which yields a 1/39.7 of Reynolds number scaling ratio. In the present study, the design feature of the facilities, named 'ACOP', in order to investigate flow and pressure distribution are described

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

Science.gov (United States)

Cummings, Chase B.

2010-09-01

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

The progress and results of a demonstration test of a cavern-type disposal facility

International Nuclear Information System (INIS)

Akiyama, Yoshihiro; Terada, Kenji; Oda, Nobuaki; Yada, Tsutomu; Nakajima, Takahiro

2011-01-01

The cavern-type disposal facilities for low-level waste (LLW) with relatively high radioactivity levels mainly generated from power reactor decommissioning and for part of transuranic (TRU) waste mainly from spent fuel reprocessing are designed to be constructed in a cavern 50 to 100 meters below ground, and to employ an engineered barrier system (EBS) of a combination of bentonite and cement materials in Japan. In order to advance the feasibility study for these disposal, a government-commissioned research project named Demonstration Test of Cavern-Type Disposal Facility started in fiscal 2005, and since fiscal 2007 a full-scale mock-up test facility has been constructed under actual subsurface environment. The main objective of the test is to establish construction methodology and procedures which ensure the required quality of the EBS on-site. By fiscal 2009 some parts of the facility have been constructed, and the test has demonstrated both practicability of the construction and achievement of the quality. They are respectively taken as low-permeability of less than 5x10 13 m/s and low-diffusivity of less than 1x10 -12 m 2 /s at the time of completion of construction. This paper covers the project outline and the test results obtained by the construction of some parts of a bentonite and cement materials. (author)

High Power RF Test Facility at the SNS

CERN Document Server

Kang, Yoon W; Campisi, Isidoro E; Champion, Mark; Crofford, Mark; Davis, Kirk; Drury, Michael A; Fuja, Ray E; Gurd, Pamela; Kasemir, Kay-Uwe; McCarthy, Michael P; Powers, Tom; Shajedul Hasan, S M; Stirbet, Mircea; Stout, Daniel; Tang, Johnny Y; Vassioutchenko, Alexandre V; Wezensky, Mark

2005-01-01

RF Test Facility has been completed in the SNS project at ORNL to support test and conditioning operation of RF subsystems and components. The system consists of two transmitters for two klystrons powered by a common high voltage pulsed converter modulator that can provide power to two independent RF systems. The waveguides are configured with WR2100 and WR1150 sizes for presently used frequencies: 402.5 MHz and 805 MHz. Both 402.5 MHz and 805 MHz systems have circulator protected klystrons that can be powered by the modulator capable of delivering 11 MW peak and 1 MW average power. The facility has been equipped with computer control for various RF processing and complete dual frequency operation. More than forty 805 MHz fundamental power couplers for the SNS superconducting linac (SCL) cavitites have been RF conditioned in this facility. The facility provides more than 1000 ft2 floor area for various test setups. The facility also has a shielded cave area that can support high power tests of normal conducti...

Full-Scale Mark II CRT program data report no. 10 (TEST 1203)

International Nuclear Information System (INIS)

Kukita, Yutaka; Takeshita, Isao; Yamamoto, Nobuo; Namatame, Ken; Shiba, Masayoshi

1981-03-01

Recorded data for TEST 1203 conducted on the Full-Scale Mark II CRT (Containment Response Test) Facility are presented. The test 1203 is the third test run of a series of steam discharge pool swell test. It is one of the tests where break diameter was varied parametrically, i.e., TEST 1201 (200 mm), TEST 1202 (240 mm) and TEST 1203 (220 mm). The test was successfully conducted. A drywell initial pressurization rate of 188 kPa/s was obtained, which is approximately equal to what is postulated for a hypothetical DBA (Design Basis Accident) LOCA in the BWR Mark II containment. (author)

Full-Scale Mark II CRT program data report no. 12 (TEST 1205)

International Nuclear Information System (INIS)

Takeshita, Isao; Kukita, Yutaka; Yamamoto, Nubuo; Namatame, Ken; Shiba, Masayoshi

1981-03-01

Recorded data for TEST 1205 conducted on the Full-Scale Mark II CRT (Containment Response Test) Facility are presented. The TEST 1205 is the fifth test run of a series of steam discharge pool swell tests. The test conditions are similar to those of the TEST 1203 except that the vacuum breaker was locked close. The test was successful and the comparison of the TEST 1205 and the TEST 1203 results shows that the vacuum breaker may have some effects to reduce the wetwell airspace pressurization, and thus to reduce the upward force to diaphram floor during a pool swell. (author)

Numerical and experimental simulation of accident processes using KMS large-scale test facility under the program of training university students for nuclear power industry

International Nuclear Information System (INIS)

Aniskevich, Yu.N.

2005-01-01

The KMS large-scale test facility is being constructed at NITI site and designed to model accident processes in VVER reactor plants and provide experimental data for safety analysis of both existing and future NPPs. The KMS phase I is at the completion stage. This is a containment model of 2000 m3 volume intended for experimentally simulating heat and mass transfers of steam-gas mixtures and aerosols inside containment. The KMS phase II will incorporate a reactor model (1:27 scale) and be used for analysing a number of events including primary and secondary LOCA. The KMS program for background training of university students in the nuclear field will include preparation and conduction of experiments, analysis of experiment data. The KMS program for background training of university students in nuclear will include: participation in the development and application of experiment procedures, preparation and carrying out experiments; carrying out pretest and post-test calculations with different computer codes; on-the-job training as operators of experiment scenarios; training of specialists in measurement and information acquisition technologies. (author)

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.

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...

SSC string test facility for superconducting magnets: Testing capabilities and program for collider magnets

International Nuclear Information System (INIS)

Kraushaar, P.; Burgett, W.; Dombeck, T.; McInturff, A.; Robinson, W.; Saladin, V.

1993-05-01

The Accelerator Systems String Test (ASST) R ampersand D Testing Facility has been established at the SSC Laboratory to test Collider and High Energy Booster (HEB) superconducting magnet strings. The facility is operational and has had two testing periods utilizing a half cell of collider prototypical magnets with the associated spool pieces and support systems. This paper presents a description of the testing capabilities of the facility with respect to components and supporting subsystems (cryogenic, power, quench protection, controls and instrumentation), the planned testing program for the collider magnets

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)

Vibration phenomena in large scale pressure suppression tests

International Nuclear Information System (INIS)

Aust, E.; Boettcher, G.; Kolb, M.; Sattler, P.; Vollbrandt, J.

1982-01-01

Structure und fluid vibration phenomena (acceleration, strain; pressure, level) were observed during blow-down experiments simulating a LOCA in the GKSS full scale multivent pressure suppression test facility. The paper describes first the source related excitations during the two regimes of condensation oscillation and of chugging, and deals then with the response vibrations of the facility's wetwell. Modal analyses of the wetwell were run using excitation by hammer and by shaker in order to separate phenomena that are particular to the GKSS facility from more general ones, i.e. phenomena specific to the fluid related parameters of blowdown and to the geometry of the vent pipes only. The lowest periodicities at about 12 and 16 Hz stem from the vent acoustics. A frequency of about 36 to 38 Hz prominent during chugging seems to result from the lowest local models of two of the wetwell's walls when coupled by the wetwell pool. Further peaks found during blowdown in the spectra of signals at higher frequencies correspond to global vibration modes of the wetwell. (orig.)

Utilization of the capsule out-pile test facilities(2000-2003)

Energy Technology Data Exchange (ETDEWEB)

Cho, M. S.; Oh, J. M.; Cho, Y. G. and others

2003-06-01

Two out-pile test facilities were installed and being utilized for the non-irradiation tests outside the HANARO. The names of the facilities are the irradiation equipment design verification test facilities and the one-channel flow test device. In these facilities, the performance test of all capsules manufactured before loading in the HANARO and the design verification test for newly developed capsules were performed. The tests in these facilities include loading/unloading, pressure drop, endurance and vibration test etc. of capsules. In the period 2000{approx}2003, the performance tests for 8 material capsules of 99M-01K{approx}02M-05U were carried out, and the design verification tests of creep and fuel capsules developed newly were performed. For development of the creep capsule, pressure drop measurement, operation test of heater, T/C, LVDT and stress loading test were performed. In the design stage of the fuel capsule, the endurance and vibration test besides the above mentioned tests were carried out for verification of the safe operation during irradiation test in the HANARO. And in-chimeny bracket and the capsule supporting system were fixed and the flow tubes and the handling tools were manufactured for use at the facilities.

ITER baffle module small-scale mock-ups: first wall thermo-mechanical testing results

International Nuclear Information System (INIS)

Severi, Y.; Giancarli, L.; Poitevin, Y.; Salavy, J.F.; Le Marois, G.; Roedig, M.; Vieider, G.

1998-01-01

The EU-home team is in charge of the R and D related to the ITER baffle first wall. Five small-scale mock-ups, using Be, CFC and W tiles and different armour/heat-sink material joints under development, have been fabricated and thermomechanically tested in FE200 (Le Creusot) and JUDITH (Juelich) electron beam facilities. The small-scale mock-ups have been submitted to thermo-mechanical fatigue tests (up to failure using accelerating techniques). The objective was to determine the performances of the armour material joints under high heat flux cycles. (orig.)

(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.

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

SBLOCA AND LOFW EXPERIMENTS IN A SCALED-DOWN IET FACILITY OF REX-10 REACTOR

Directory of Open Access Journals (Sweden)

YEON-GUN LEE

2013-06-01

Full Text Available This paper presents an experimental investigation of the small-break loss-of-coolant accident (SBLOCA and the loss-of-feedwater accident (LOFW in a scaled integral test facility of REX-10. REX-10 is a small integral-type PWR in which the coolant flow is driven by natural circulation, and the RCS is pressurized by the steam-gas pressurizer. The postulated accidents of REX-10 include the system depressurization initiated by the break of a nitrogen injection line connected to the steam-gas pressurizer and the complete loss of normal feedwater flow by the malfunction of control systems. The integral effect tests on SBLOCA and LOFW are conducted at the REX-10 Test Facility (RTF, a full-height full-pressure facility with reduced power by 1/50. The SBLOCA experiment is initiated by opening a flow passage out of the pressurizer vessel, and the LOFW experiment begins with the termination of the feedwater supply into the helical-coil steam generator. The experimental results reveal that the RTF can assure sufficient cooldown capability with the simulated PRHRS flow during these DBAs. In particular, the RTF exhibits faster pressurization during the LOFW test when employing the steam-gas pressurizer than the steam pressurizer. This experimental study can provide unique data to validate the thermal-hydraulic analysis code for REX-10.

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.

Oxy-Combustion Burner and Integrated Pollutant Removal Research and Development Test Facility

Energy Technology Data Exchange (ETDEWEB)

Mark Schoenfield; Manny Menendez; Thomas Ochs; Rigel Woodside; Danylo Oryshchyn

2012-09-30

A high flame temperature oxy-combustion test facility consisting of a 5 MWe equivalent test boiler facility and 20 KWe equivalent IPR® was constructed at the Hammond, Indiana manufacturing site. The test facility was operated natural gas and coal fuels and parametric studies were performed to determine the optimal performance conditions and generated the necessary technical data required to demonstrate the technologies are viable for technical and economic scale-up. Flame temperatures between 4930-6120F were achieved with high flame temperature oxy-natural gas combustion depending on whether additional recirculated flue gases are added to balance the heat transfer. For high flame temperature oxy-coal combustion, flame temperatures in excess of 4500F were achieved and demonstrated to be consistent with computational fluid dynamic modeling of the burner system. The project demonstrated feasibility and effectiveness of the Jupiter Oxygen high flame temperature oxy-combustion process with Integrated Pollutant Removal process for CCS and CCUS. With these technologies total parasitic power requirements for both oxygen production and carbon capture currently are in the range of 20% of the gross power output. The Jupiter Oxygen high flame temperature oxy-combustion process has been demonstrated at a Technology Readiness Level of 6 and is ready for commencement of a demonstration project.

Transfer of test samples and wastes between post-irradiation test facilities (FMF, AGF, MMF)

International Nuclear Information System (INIS)

Ishida, Yasukazu; Suzuki, Kazuhisa; Ebihara, Hikoe; Matsushima, Yasuyoshi; Kashiwabara, Hidechiyo

1975-02-01

Wide review is given on the problems associated with the transfer of test samples and wastes between post-irradiation test facilities, FMF (Fuel Monitoring Facility), AGF (Alpha Gamma Facility), and MMF (Material Monitoring Facility) at the Oarai Engineering Center, PNC. The test facilities are connected with the JOYO plant, an experimental fast reactor being constructed at Oarai. As introductory remarks, some special features of transferring irradiated materials are described. In the second part, problems on the management of nuclear materials and radio isotopes are described item by item. In the third part, the specific materials that are envisaged to be transported between JOYO and the test facilities are listed together with their geometrical shapes, dimensions, etc. In the fourth part, various routes and methods of transportation are explained with many block charts and figures. Brief explanation with lists and drawings is also given to transportation casks and vessels. Finally, some future problems are discussed, such as the prevention of diffusive contamination, ease of decontamination, and the identification of test samples. (Aoki, K.)

Investigations into NOx emissions and burnout for coals with high ash content in a bench scale test facility

Energy Technology Data Exchange (ETDEWEB)

Greul, U.; Kluger, F.; Peter, G.; Spliethoff, H.; Hein, K.R.G. [University of Stuttgart, Stuttgart (Germany). Inst. fuer Verfahrenstechnik und Dampfkesselwesen

2000-07-01

At the Stuttgart University's Institute of Process Engineering and Power Plant Technology (IVD) investigations of in-furnace DeNOx technologies with regard to their NOx reduction efficiency are carried out using an electrically heated bench-scale test facility to evaluate the effect of different process parameters independently. The DeNOx technologies of air and fuel staging have been demonstrated to be effective control techniques to reduce NOx from stationary sources. For a wide range of brown and hard coals from Europe, South Africa and Australia test runs with air-staged combustion have been carried out. The ash content of the hard coals used was in the range between 8 and 28%. The investigated parameters were temperature (1000-1300{degree}C), stoichiometry (1.25-0.55), and residence time (1-6 s) in the fuel rich primary zone. With increasing temperatures and residence times in fuel-rich conditions in air-staged combustion NOx emissions below 300 mg/m{sup 3} can be achieved even with hard coals. For a few brown coals NOx values lower than 100 mg/m{sup 3} are possible. Dependent on the coal rank individual parameters are more important than others. For low and medium volatile hard coals the increasing of the residence time is more effective than higher temperature or lower air ratios in the primary zone. However, with high volatile hard coal or brown coal as primary fuel the influence of temperature and stoichiometry in the primary zone plays a key role for NOx reduction effectiveness. The burnout led to restrictions in large scale applications for air-staged combustion especially with hard coals as primary fuel. Investigations at different primary air ratios and temperatures show the effect of these parameters on the burnout values along the course of combustion. 7 refs., 14 figs., 2 tabs.

Pilot-scale treatability test plan for the 100-HR-3 operable unit

International Nuclear Information System (INIS)

1994-08-01

This document presents the treatability test plan for pilot-scale pump-and-treat testing at the 100-HR-3 Operable Unit. The test will be conducted in fulfillment of interim Milestone M-15-06E to begin pilot-scale pump-and-treat operations by August 1994. The scope of the test was determined based on the results of lab/bench-scale tests (WHC 1993a) conducted in fulfillment of Milestone M-15-06B. These milestones were established per agreement between the U.S. Department of Energy (DOE), the Washington State Department of Ecology and the U.S. Environmental Protection Agency (EPA), and documented on Hanford Federal of Ecology Facility Agreement and Consent Order Change Control Form M-15-93-02. This test plan discusses a pilot-scale pump-and-treat test for the chromium plume associated with the D Reactor portion of the 100-HR-3 Operable Unit. Data will be collected during the pilot test to assess the effectiveness, operating parameters, and resource needs of the ion exchange (IX) pump-and-treat system. The test will provide information to assess the ability to remove contaminants by extracting groundwater from wells and treating extracted groundwater using IX. Bench-scale tests were conducted previously in which chromium VI was identified as the primary contaminant of concern in the 100-D reactor plume. The DOWEX 21K trademark resin was recommended for pilot-scale testing of an IX pump-and-treat system. The bench-scale test demonstrated that the system could remove chromium VI from groundwater to concentrations less than 50 ppb. The test also identified process parameters to monitor during pilot-scale testing. Water will be re-injected into the plume using wells outside the zone of influence and upgradient of the extraction well

The Field Lysimeter Test Facility (FLTF) at the Hanford Site: Installation and initial tests

International Nuclear Information System (INIS)

Gee, G.W.; Kirkham, R.R.; Downs, J.L.; Campbell, M.D.

1989-02-01

The objectives of this program are to test barrier design concepts and to demonstrate a barrier design that meets established performance criteria for use in isolating wastes disposed of near-surface at the Hanford Site. Specifically, the program is designed to assess how well the barriers perform in controlling biointrusion, water infiltration, and erosion, as well as evaluating interactions between environmental variables and design factors of the barriers. To assess barrier performance and design with respect to infiltration control, field lysimeters and small- and large-scale field plots are planned to test the performance of specific barrier designs under actual and modified (enhanced precipitation) climatic conditions. The Field Lysimeter Test Facility (FLTF) is located in the 600 Area of the Hanford Site just east of the 200 West Area and adjacent to the Hanford Meteorological Station. The FLTF data will be used to assess the effectiveness of selected protective barrier configurations in controlling water infiltration. The facility consists of 14 drainage lysimeters (2 m dia x 3 m deep) and four precision weighing lysimeters (1.5 m x 1.5 m x 1.7 m deep). The lysimeters are buried at grade and aligned in a parallel configuration, with nine lysimeters on each side of an underground instrument chamber. The lysimeters were filled with materials to simulate a multilayer protective barrier system. Data gathered from the FLTF will be used to compare key barrier components and to calibrate and test models for predicting long-term barrier performance

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...

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

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.

HIV testing in nonhealthcare facilities among adolescent MSM.

Science.gov (United States)

Marano, Mariette R; Stein, Renee; Williams, Weston O; Wang, Guoshen; Xu, Songli; Uhl, Gary; Cheng, Qi; Rasberry, Catherine N

2017-07-01

To describe the extent to which Centers for Disease Control and Prevention (CDC)-funded HIV testing in nonhealthcare facilities reaches adolescent MSM, identifies new HIV infections, and links those newly diagnosed to medical care. We describe HIV testing, newly diagnosed positivity, and linkage to medical care for adolescent MSM who received a CDC-funded HIV test in a nonhealthcare facility in 2015. We assess outcomes by race/ethnicity, HIV-related risk behaviors, and US geographical region. Of the 703 890 CDC-funded HIV testing events conducted in nonhealthcare facilities in 2015, 6848 (0.9%) were provided to adolescent MSM aged 13-19 years. Among those tested, 1.8% were newly diagnosed with HIV, compared with 0.7% among total tests provided in nonhealthcare facilities regardless of age and sex. The odds of testing positive among black adolescent MSM were nearly four times that of white adolescent MSM in multivariable analysis (odds ratio = 3.97, P adolescent MSM newly diagnosed with HIV, 67% were linked to HIV medical care. Linkage was lower among black (59%) and Hispanic/Latino adolescent MSM (71%) compared with white adolescent MSM (88%). CDC-funded nonhealthcare facilities can reach and provide HIV tests to adolescent MSM and identify new HIV infections; however, given the low rate of HIV testing overall and high engagement in HIV-related risk behaviors, there are opportunities to increase access to HIV testing and linkage to care for HIV-positive adolescent MSM. Efforts are needed to identify and address the barriers that prevent black and Hispanic/Latino adolescent MSM from being linked to HIV medical care in a timely manner.

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.

Applicability of laboratory data to large scale tests under dynamic loading conditions

International Nuclear Information System (INIS)

Kussmaul, K.; Klenk, A.

1993-01-01

The analysis of dynamic loading and subsequent fracture must be based on reliable data for loading and deformation history. This paper describes an investigation to examine the applicability of parameters which are determined by means of small-scale laboratory tests to large-scale tests. The following steps were carried out: (1) Determination of crack initiation by means of strain gauges applied in the crack tip field of compact tension specimens. (2) Determination of dynamic crack resistance curves of CT-specimens using a modified key-curve technique. The key curves are determined by dynamic finite element analyses. (3) Determination of strain-rate-dependent stress-strain relationships for the finite element simulation of small-scale and large-scale tests. (4) Analysis of the loading history for small-scale tests with the aid of experimental data and finite element calculations. (5) Testing of dynamically loaded tensile specimens taken as strips from ferritic steel pipes with a thickness of 13 mm resp. 18 mm. The strips contained slits and surface cracks. (6) Fracture mechanics analyses of the above mentioned tests and of wide plate tests. The wide plates (960x608x40 mm 3 ) had been tested in a propellant-driven 12 MN dynamic testing facility. For calculating the fracture mechanics parameters of both tests, a dynamic finite element simulation considering the dynamic material behaviour was employed. The finite element analyses showed a good agreement with the simulated tests. This prerequisite allowed to gain critical J-integral values. Generally the results of the large-scale tests were conservative. 19 refs., 20 figs., 4 tabs

Development and psychometric testing of the Supportive Supervisory Scale.

Science.gov (United States)

McGilton, Katherine S

2010-06-01

To describe the development and psychometric testing of the Supportive Supervisory Scale (SSS). The development of the items of the scale was based on Winnicott's relationship theory and on focus groups with 26 healthcare aides (HCAs) and 30 supervisors from six long-term care (LTC) facilities in Ontario, Canada. Content validity of the 15-item instrument was established by a panel of experts. Based on a secondary analysis of data collected from 222 HCAs in 10 LTC facilities in Ontario, Canada, the SSS was subjected to principal components analysis with oblique rotation. A two-factor solution was accepted, which is consistent with the theoretical conceptualization of the instrument. Factor I was labeled Respects Uniqueness and Factor II was labeled Being Reliable. Internal consistency of Factor I was .95, and that of Factor II was .91. Discriminant validity was also established. The focus groups revealed that "being available to staff" while "recognizing the HCA as an individual, and taking a moment to get to know them" was essential to feeling supported by their supervisor. The SSS is a reliable and valid measure of supervisory support of supervisors working in LTC facilities. At the core of supportive supervision is the supervisor's ability to develop and maintain positive relationships with each HCA. It is through respecting the uniqueness of each HCA and being reliable that supervisor-HCA relationships can flourish. Supportive leadership in LTC settings is a major contributor to HCAs' job satisfaction and retention and to quality of patient care. Therefore, a tool developed and tested to measure supervisors' supportive capacities in LTC is primal to evaluate the effectiveness of supervisors in these environments.

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

Current state of the construction of an integrated test facility for hydrogen risk

Energy Technology Data Exchange (ETDEWEB)

Na, Young Su; Hong, Seong-Ho; Hong, Seong-Wan [KAERI, Daejeon (Korea, Republic of)

2015-05-15

Experimental research on hydrogen as a combustible gas is important for an assessment of the integrity of a containment building under a severe accident. The Korea Atomic Energy Research Institute (KAERI) is preparing a large-scaled test facility, called SPARC (SPray-Aerosol-Recombiner-Combustion), to estimate the hydrogen behavior such as the distribution, combustion and mitigation. This paper introduces the experimental research activity on hydrogen risk, which was presented at International Congress on Advances in Nuclear Power Plants (ICAPP) this year. The KAERI is preparing a test facility, called SPARC (SPray-Aerosol-Recombiner-Combustion test facility), for an assessment of the hydrogen risk. In the SPARC, hydrogen behavior such as mixing with steam and air, distribution, and combustion in the containment atmosphere will be observed. The SPARC consists of a pressure vessel with a 9.5 m height and 3.4 m in diameter and the operating system to control the thermal hydraulic conditions up to 1.5 MPa at 453 K in a vessel. The temperature, pressure, and gas concentration at various locations will be measured to estimate the atmospheric behavior in a vessel. To install the SPARC, an experimental building, called LIFE (Laboratory for Innovative mitigation of threats from Fission products and Explosion), was constructed at the KAERI site. LIFE has an area of 480 m''2 and height of 18.6 m, and it was designed by considering the experimental safety and specification of a large-sized test facility.

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...

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

Technical bases for establishing a salt test facility

International Nuclear Information System (INIS)

1985-05-01

The need for a testing facility in which radioactive materials may be used in an underground salt environment is explored. No such facility is currently available in salt deposits in the United States. A salt test facility (STF) would demonstrate the feasibility of safely storing radioactive waste in salt and would provide data needed to support the design, construction, licensing, and operation of a radioactive waste repository in salt. Nineteen issues that could affect long-term isolation of waste materials in a salt repository are identified from the most pertinent recent literature. The issues are assigned an overall priority and a priority relative to the activities of the STF. Individual tests recommended for performance in the STF to resolve the 19 issues are described and organized under three groups: waste package performance, repository design and operation, and site characterization and evaluation. The requirements for a salt test facility are given in the form of functional criteria, and the approach that will be used in the design, execution, interpretation, and reporting of tests is discussed

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)

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)

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

Scale economies in a series of generic interim SNF storage facilities - 15104

International Nuclear Information System (INIS)

Rothwell, G.

2015-01-01

This paper describes a micro-economic, cost-engineering model of a centralized (Generic Interim Storage Facility - GISF) facility to monitor LWR irradiated fuel with particular attention to scale economies (e.g., to compare the likely costs at a power plant site or at regional, national and international facilities). This paper is based on the cost estimates of the Private Fuel Services Facility (PFSF) on the Skull Valley Band of Goshute Indians' Reservation in Utah, licensed by the US NRC in 2006 to centralize storage of 40.000 metric tons of heavy metal (MTHM) for 20 to 40 years. Assuming movement of the 40.000 MTHM every 40 years to a new facility, the levelized costs are 144 dollars/kg without high security and physical protection, and 208 dollars/kg with high security through 2111 (assuming disposal within a century), or about 0.50 dollars/MWh to 0.75 dollars/MWh depending on the burnup and thermal efficiency of the nuclear power plant. This cost estimate is generalized to explore scale economies for facilities with and without high security and physical protection. There are declining levelized costs with increasing size to 120.000 MTHM without high security, and to 500.000 MTHM with high security, i.e., the higher the level of security, the stronger the economies of scale. (author)

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

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.

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.)

Overview of the IFMIF test facility design in IFMIF/EVEDA phase

International Nuclear Information System (INIS)

Tian, Kuo; Abou-Sena, Ali; Arbeiter, Frederik; García, Ángela; Gouat, Philippe; Heidinger, Roland; Heinzel, Volker; Ibarra, Ángel; Leysen, Willem; Mas, Avelino; Mittwollen, Martin; Möslang, Anton; Theile, Jürgen; Yamamoto, Michiyoshi; Yokomine, Takehiko

2015-01-01

Highlights: • This paper summarizes the current design status of IFMIF EVEDA test facility. • The principle functions of the test facility and key components are described. • The brief specifications of the systems and key components are addressed. - Abstract: The test facility (TF) is one of the three major facilities of the International Fusion Material Irradiation Facility (IFMIF). Engineering designs of TF main systems and key components have been initiated and developed in the IFMIF EVEDA (Engineering Validation and Engineering Design Activities) phase since 2007. The related work covers the designs of a test cell which is the meeting point of the TF and accelerator facility and lithium facility, a series of test modules for experiments under different irradiation conditions, an access cell to accommodate remote handling systems, four test module handling cells for test module processing and assembling, and test facility ancillary systems for engineering support on energy, media, and control infrastructure. This paper summarizes the principle functions, brief specifications, and the current design status of the above mentioned IFMIF TF systems and key components.

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.

Waste Receiving and Processing (WRAP) Facility Weight Scale Analysis Fairbanks Weight Scale Evaluation Results

International Nuclear Information System (INIS)

JOHNSON, M.D.

1999-01-01

Fairbanks Weight Scales are used at the Waste Receiving and Processing (WRAP) facility to determine the weight of waste drums as they are received, processed, and shipped. Due to recent problems, discovered during calibration, the WRAP Engineering Department has completed this document which outlines both the investigation of the infeed conveyor scale failure in September of 1999 and recommendations for calibration procedure modifications designed to correct deficiencies in the current procedures

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

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

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.

Small-Scale Spray Releases: Initial Aerosol Test Results

Energy Technology Data Exchange (ETDEWEB)

Mahoney, Lenna A.; Gauglitz, Phillip A.; Kimura, Marcia L.; Brown, Garrett N.; Kurath, Dean E.; Buchmiller, William C.; Smith, Dennese M.; Blanchard, Jeremy; Song, Chen; Daniel, Richard C.; Wells, Beric E.; Tran, Diana N.; Burns, Carolyn A.

2013-05-29

One of the events postulated in the hazard analysis at the Waste Treatment and Immobilization Plant (WTP) and other U.S. Department of Energy (DOE) nuclear facilities is a breach in process piping that produces aerosols with droplet sizes in the respirable range. The current approach for predicting the size and concentration of aerosols produced in a spray leak involves extrapolating from correlations reported in the literature. These correlations are based on results obtained from small engineered spray nozzles using pure liquids with Newtonian fluid behavior. The narrow ranges of physical properties on which the correlations are based do not cover the wide range of slurries and viscous materials that will be processed in the WTP and across processing facilities in the DOE complex. Two key technical areas were identified where testing results were needed to improve the technical basis by reducing the uncertainty due to extrapolating existing literature results. The first technical need was to quantify the role of slurry particles in small breaches where the slurry particles may plug and result in substantially reduced, or even negligible, respirable fraction formed by high-pressure sprays. The second technical need was to determine the aerosol droplet size distribution and volume from prototypic breaches and fluids, specifically including sprays from larger breaches with slurries where data from the literature are scarce. To address these technical areas, small- and large-scale test stands were constructed and operated with simulants to determine aerosol release fractions and net generation rates from a range of breach sizes and geometries. The properties of the simulants represented the range of properties expected in the WTP process streams and included water, sodium salt solutions, slurries containing boehmite or gibbsite, and a hazardous chemical simulant. The effect of antifoam agents was assessed with most of the simulants. Orifices included round holes and

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...

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

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.)

Verification of a TRACE EPRTM model on the basis of a scaling calculation of an SBLOCA ROSA test

International Nuclear Information System (INIS)

Freixa, J.; Manera, A.

2011-01-01

Research highlights: → Verification of a TRACE input deck for the EPR TM generation III PWR. → Scaling simulation of an SBLOCA experiment of the integral test facility ROSA/LSTF. → The EPR TM model was compared with the TRACE results of the ROSA/LSTF model. - Abstract: In cooperation with the Finnish Radiation and Nuclear Safety Authority (STUK), a project has been launched at the Paul Scherrer Institute (PSI) aimed at performing safety evaluations of the Olkiluoto-3 nuclear power plant (NPP), the first EPR TM , a generation III pressurizer water reactor (PWR); with particular emphasis on small-and large-break loss-of-coolant-accidents (SB/LB-LOCAs) and main steam-line breaks. As a first step of this work, the best estimate system code TRACE has been used to develop a model of Olkiluoto-3. In order to test the nodalization, a scaling calculation from the rig of safety assessment (ROSA) test facility has been performed. The ROSA large scale test facility (LSTF) was built to simulate Westinghouse design pressurized water reactors (PWR) with a four-loop configuration. Even though there are differences between the EPR TM and the Westinghouse designs, the number of similarities is large enough to carry out scaling calculations on SBLOCA and LOCA cases from the ROSA facility; as a matter of fact, the main differences are located in the secondary side. Test 6-1 of the ROSA 1 programme, an SBLOCA with the break situated in the upper head of the reactor pressure vessel (RPV), was of special interest since a very good agreement with the experiment was obtained with a TRACE input deck. In order to perform such scaling calculation, the set-points of the secondary relief and safety valves in the EPR TM nodalization had to be changed to those used in the ROSA facility, the break size and the core power had to be scaled by a factor of 60 (according to the core power and core volume) and the pumps coast down had to be adapted to the ones of the test. The calculation showed

Small-Scale Spray Releases: Initial Aerosol Test Results

Energy Technology Data Exchange (ETDEWEB)

Mahoney, Lenna A.; Gauglitz, Phillip A.; Kimura, Marcia L.; Brown, Garrett N.; Kurath, Dean E.; Buchmiller, William C.; Smith, Dennese M.; Blanchard, Jeremy; Song, Chen; Daniel, Richard C.; Wells, Beric E.; Tran, Diana N.; Burns, Carolyn A.

2012-11-01

One of the events postulated in the hazard analysis at the Waste Treatment and Immobilization Plant (WTP) and other U.S. Department of Energy (DOE) nuclear facilities is a breach in process piping that produces aerosols with droplet sizes in the respirable range. The current approach for predicting the size and concentration of aerosols produced in a spray leak involves extrapolating from correlations reported in the literature. These correlations are based on results obtained from small engineered spray nozzles using pure liquids with Newtonian fluid behavior. The narrow ranges of physical properties on which the correlations are based do not cover the wide range of slurries and viscous materials that will be processed in the WTP and across processing facilities in the DOE complex. Two key technical areas were identified where testing results were needed to improve the technical basis by reducing the uncertainty due to extrapolating existing literature results. The first technical need was to quantify the role of slurry particles in small breaches where the slurry particles may plug and result in substantially reduced, or even negligible, respirable fraction formed by high-pressure sprays. The second technical need was to determine the aerosol droplet size distribution and volume from prototypic breaches and fluids, specifically including sprays from larger breaches with slurries where data from the literature are scarce. To address these technical areas, small- and large-scale test stands were constructed and operated with simulants to determine aerosol release fractions and generation rates from a range of breach sizes and geometries. The properties of the simulants represented the range of properties expected in the WTP process streams and included water, sodium salt solutions, slurries containing boehmite or gibbsite, and a hazardous chemical simulant. The effect of anti-foam agents was assessed with most of the simulants. Orifices included round holes and

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.