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Sample records for tests main steam-line

  1. RELAP5 analysis of PKL, main steam line break test

    Energy Technology Data Exchange (ETDEWEB)

    Jonnet, J.R.; Stempniewicz, M.M., E-mail: stempniewicz@nrg.eu; With, A. de; Wakker, P.H.

    2013-12-15

    Highlights: • RELAP5/MOD 3.2 code validation is performed by analyzing a main steam line break test in the PKL large scale test facility. • The RELAP5 model reproduces well the important events of the PKL test. • RELAP5 transient results show noticeable sensitivity to small differences in the initial conditions. • Accurate prediction of the coolant temperature is essential for the assessment of potential core re-criticality. - Abstract: PKL is a large scale test facility of the primary system owned by AREVA NP GmbH. It is used for extensive experimental investigations to study the integral behavior of Pressurized Water Reactor (PWR) plants under accident conditions. Since 2001, the test program is a part of an international cooperation project (SETH, followed by PKL1 and PKL2) set up by the OECD. The aim of the present work was to perform a short validation study of the thermo-hydraulics code RELAP5. A model of the PKL test facility has been developed, tested and applied to one of the experiments performed at the PKL. The chosen experiment was the test G3.1. In that experiment, a main steam line break occurs, causing a rapid depressurization of the affected steam generator. This leads to an increase of the heat transfer from the primary to the secondary side and thereby to a fast cool-down transient on the primary side. The main objective of this analysis was the qualification of the RELAP5 code results against heat transfer from the primary to the secondary side in both affected and intact loops, and temperatures in the primary system. The calculation results have been compared to the experimental results. It was concluded that the most important events during the test are reproduced relatively well by the model. The calculated coolant temperature in the core is higher than in the experiment. The minimum temperature is about 5% higher than measured. The secondary pressures in SG-1, 3, and 4 is in very good agreement with the experimental value, but in the

  2. Code Assessment of SPACE 2.19 using LSTF 10% Main Steam-Line-Break Test

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Minhee; Kim, Seyun [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    The Safety and Performance Analysis Code for Nuclear Power Plants (SPACE) has been developed in recent years by the Korea Hydro and Nuclear Power Co. through collaborative works with other Korean nuclear industries and research institutes. As a result of the development, the 2.19 version of the code was released through the successive various verification and validation works. In this study, results produced by the SPACE 2.19 code were compared with the experimental data from JAERI's LSTF Test Run SBSL- 01 for a 10% main steam line break transient in a pressurized water reactor. The LSTF 10% main steam line break test were simulated using the SPACE 2.19 for code V and V work. The overall comparisons between the SPACE 2.19 code prediction and the LSTF Test Run SB-SL-01 experimental data are reasonably satisfactory. The comparisons were conducted in terms of the variations of mass flow rate, void fraction, pressure, collapsed liquid level, temperature, and system flow rate for the transient. In addition, the input model was modified for simulation accuracy of PZR pressure based on the calculated results. The correction of PORV setpoint affects to simulate the PORV open and close phenomena similarly with experiments. From the modification, the computed results show a reasonable agreement with experimental data in overall transient time.

  3. NPP Krsko Containment Response Following Main Steam Line Break

    International Nuclear Information System (INIS)

    Spalj, S.; Grgic, D.; Cavlina, N.

    2000-01-01

    This paper presents the calculation of thermohydraulic environmental parameters (pressure and temperature) inside containment of Krsko NPP after postulated Main Steam Line Break (MSLB) accident. This analysis was done as a part of the ambient parameters specification in the frame of the NPP Krsko Equipment Qualification (EQ) project. The RELAP5/mod2 computer code was used for the determination of MSLB mass and energy release and computer code GOTHIC was used to calculate pressure and temperature profiles inside NPP Krsko containment. The analysis was performed for spectrum of break sizes to account for possible steam superheating during accidents with smaller break sizes. (author)

  4. Characteristics of fluctuating pressure generated in BWR main steam lines

    International Nuclear Information System (INIS)

    Takahashi, Shiro; Okuyama, Keita; Tamura, Akinori

    2009-01-01

    The BWR-3 steam dryer in the Quad Cities Unit 2 Nuclear Power Plant was damaged by high cycle fatigue due to acoustic-induced vibration. The dryer failure was as attributed to flow-induced acoustic resonance at the stub pipes of safety relief valves (SRVs) in the main steam lines (MSLs). The acoustic resonance was considered to be generated by interaction between the sound field and an unstable shear layer across the closed side branches with SRV stub pipes. We have started a research program on BWR dryers to develop their loading evaluation methods. Moreover, it has been necessary to evaluate the dryer integrity of BWR-5 plants which are the main type of BWR in Japan. In the present study, we used 1/10-scale BWR tests and analyses to investigate the flow-induced acoustic resonance and acoustic characteristics in MSLs. The test apparatus consisted of a steam dryer, a steam dome and 4 MSLs with 20 SRV stub pipes. A finite element method (FEM) was applied for the calculation of three-dimensional wave equations in acoustic analysis. We demonstrated that remarkable fluctuating pressures occurred in high and low frequency regions. High frequency fluctuating pressures was generated by the flow-induced acoustic resonance in the SRV stub pipes. Low frequency fluctuating pressure was generated in an MSL with the dead leg. The frequency of the latter almost coincided with the natural frequency of the MSL with the dead leg. The amplitude of the fluctuating pressures in the multiple stub pipes became more intense because of interaction between them compared with that in the single stub pipe. Acoustic analysis results showed that the multiple stub pipes caused several natural frequencies in the vicinity of the natural frequency of the single stub pipe and several modes of the standing wave in the MSLs. (author)

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

    International Nuclear Information System (INIS)

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

    1976-08-01

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

  6. Modeling bubble condenser containment with computer code COCOSYS: post-test calculations of the main steam line break experiment at ELECTROGORSK BC V-213 test facility

    International Nuclear Information System (INIS)

    Lola, I.; Gromov, G.; Gumenyuk, D.; Pustovit, V.; Sholomitsky, S.; Wolff, H.; Arndt, S.; Blinkov, V.; Osokin, G.; Melikhov, O.; Melikhov, V.; Sokoline, A.

    2005-01-01

    Containment of the WWER-440 Model 213 nuclear power plant features a Bubble Condenser, a complex passive pressure suppression system, intended to limit pressure rise in the containment during accidents. Due to lack of experimental evidence of its successful operation in the original design documentation, the performance of this system under accidents with ruptures of large high-energy pipes of the primary and secondary sides remains a known safety concern for this containment type. Therefore, a number of research and analytical studies have been conducted by the countries operating WWER-440 reactors and their Western partners in the recent years to verify Bubble Condenser operation under accident conditions. Comprehensive experimental research studies at the Electrogorsk BC V-213 test facility, commissioned in 1999 in Electrogorsk Research and Engineering Centre (EREC), constitute essential part of these efforts. Nowadays this is the only operating large-scale facility enabling integral tests on investigation of the Bubble Condenser performance. Several large international research projects, conducted at this facility in 1999-2003, have covered a spectrum of pipe break accidents. These experiments have substantially improved understanding of the overall system performance and thermal hydraulic phenomena in the Bubble Condenser Containment, and provided valuable information for validating containment codes against experimental results. One of the recent experiments, denoted as SLB-G02, has simulated steam line break. The results of this experiment are of especial value for the engineers working in the area of computer code application for WWER-440 containment analyses, giving an opportunity to verify validity of the code predictions and identify possibilities for model improvement. This paper describes the results of the post-test calculations of the SLB-G02 experiment, conducted as a joint effort of GRS, Germany and Ukrainian technical support organizations for

  7. Quad Cities Unit 2 Main Steam Line Acoustic Source Identification and Load Reduction

    International Nuclear Information System (INIS)

    DeBoo, Guy; Ramsden, Kevin; Gesior, Roman

    2006-01-01

    The Quad Cities Units 1 and 2 have a history of steam line vibration issues. The implementation of an Extended Power Up-rate resulted in significant increases in steam line vibration as well as acoustic loading of the steam dryers, which led to equipment failures and fatigue cracking of the dryers. This paper discusses the results of extensive data collection on the Quad Cities Unit 2 replacement dryer and the Main Steam Lines. This data was taken with the intent of identifying acoustic sources in the steam system. Review of the data confirmed that vortex shedding coupled column resonance in the relief and safety valve stub pipes were the principal sources of large magnitude acoustic loads in the main steam system. Modifications were developed in sub-scale testing to alter the acoustic properties of the valve standpipes and add acoustic damping to the system. The modifications developed and installed consisted of acoustic side branches that were attached to the Electromatic Relief Valve (ERV) and Main Steam Safety Valve (MSSV) attachment pipes. Subsequent post-modification testing was performed in plant to confirm the effectiveness of the modifications. The modifications have been demonstrated to reduce vibration loads at full Extended Power Up-rate (EPU) conditions to levels below those at Original Licensed Thermal Power (OLTP). (authors)

  8. Development of LBB Piping Evaluation Diagram for APR 1000 Main Steam Line Piping

    International Nuclear Information System (INIS)

    Yang, J. S.; Jeong, I. L.; Park, C. Y.; Bai, S. Y.

    2010-01-01

    This paper presents the piping evaluation diagram (PED) to assess the applicability of Leak-Before- Break(LBB) for APR 1000 main steam line piping. LBB-PED of APR 1000 main steam line piping is independent of its piping geometry and has a function of the loads applied in piping system. Also, in order to evaluate LBB applicability during construction process with only the comparative evaluation of material properties between actually used and expected, the expected changes of material properties are considered in the LBB-PED. The LBB-PED, therefore, can be used for quick LBB evaluation of APR 1000 main steam line piping of both design and construction

  9. Steam line rupture experiments with the PPOOLEX test facility

    International Nuclear Information System (INIS)

    Laine, J.; Puustinen, M.

    2008-07-01

    The results of the steam line rupture experiment series in 2007 with the scaled down PPOOLEX test facility designed and constructed at Lappeenranta University of Technology are reported. The test facility is a closed stainless steel vessel divided into two compartments, dry well and wet well. Air was blown into the dry well compartment and from there through a DN200 blowdown pipe to the condensation pool. Altogether five experiments, each consisting of several blows (tests), were carried out. The main purpose of the experiment series was to study the initial phase of a postulated steam line break accident inside a BWR containment. Specifically, thermal stratification in the dry well compartment and ejection of water plug from the blowdown pipe were of interest. In addition, the effect of counterpressure on bubble dynamics was studied. A temperature difference of approximately 15 deg. C between the upper and lower part of the dry well was measured. In the wet well gas space, a temperature difference of more than 30 deg. C was registered. These were measured during the compression period of the tests. Towards the end of the tests the temperature differences tended to disappear. To get a more detailed picture of temperature distribution in the wet well, especially close to the water level, a dense net of measurements is required in future experiments. In longer experiments, heat conduction to structures and heat losses to surroundings should also be taken into account. Ejection of water plugs from the blowdown pipe did not cause notable loads to the structures due to the suppressing effect of the dry well compartment. The maximum measured pressure pulse at the pool bottom was only 10 kPa and the maximum strain amplitude at the pool bottom rounding was negligible both in axial and circumferential direction. As the counterpressure of the system increased, but the flow rate remained the same, the maximum size of the air bubbles at the blowdown pipe outlet got smaller and

  10. Steam line rupture experiments with the PPOOLEX test facility

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-15

    The results of the steam line rupture experiment series in 2007 with the scaled down PPOOLEX test facility designed and constructed at Lappeenranta University of Technology are reported. The test facility is a closed stainless steel vessel divided into two compartments, dry well and wet well. Air was blown into the dry well compartment and from there through a DN200 blowdown pipe to the condensation pool. Altogether five experiments, each consisting of several blows (tests), were carried out. The main purpose of the experiment series was to study the initial phase of a postulated steam line break accident inside a BWR containment. Specifically, thermal stratification in the dry well compartment and ejection of water plug from the blowdown pipe were of interest. In addition, the effect of counterpressure on bubble dynamics was studied. A temperature difference of approximately 15 deg. C between the upper and lower part of the dry well was measured. In the wet well gas space, a temperature difference of more than 30 deg. C was registered. These were measured during the compression period of the tests. Towards the end of the tests the temperature differences tended to disappear. To get a more detailed picture of temperature distribution in the wet well, especially close to the water level, a dense net of measurements is required in future experiments. In longer experiments, heat conduction to structures and heat losses to surroundings should also be taken into account. Ejection of water plugs from the blowdown pipe did not cause notable loads to the structures due to the suppressing effect of the dry well compartment. The maximum measured pressure pulse at the pool bottom was only 10 kPa and the maximum strain amplitude at the pool bottom rounding was negligible both in axial and circumferential direction. As the counterpressure of the system increased, but the flow rate remained the same, the maximum size of the air bubbles at the blowdown pipe outlet got smaller and

  11. Main steam line break accident simulation of APR1400 using the model of ATLAS facility

    Science.gov (United States)

    Ekariansyah, A. S.; Deswandri; Sunaryo, Geni R.

    2018-02-01

    A main steam line break simulation for APR1400 as an advanced design of PWR has been performed using the RELAP5 code. The simulation was conducted in a model of thermal-hydraulic test facility called as ATLAS, which represents a scaled down facility of the APR1400 design. The main steam line break event is described in a open-access safety report document, in which initial conditions and assumptionsfor the analysis were utilized in performing the simulation and analysis of the selected parameter. The objective of this work was to conduct a benchmark activities by comparing the simulation results of the CESEC-III code as a conservative approach code with the results of RELAP5 as a best-estimate code. Based on the simulation results, a general similarity in the behavior of selected parameters was observed between the two codes. However the degree of accuracy still needs further research an analysis by comparing with the other best-estimate code. Uncertainties arising from the ATLAS model should be minimized by taking into account much more specific data in developing the APR1400 model.

  12. Flow Instabilities and Main Steam Line Vibrations in a Pressurized Water Reactor

    International Nuclear Information System (INIS)

    Henriksson, Mats; Westin, Johan; Granhall, Tord; Andersson, Lars; Bjerke, Lars-Erik

    2002-01-01

    Severe vibrational problems occurred in the main steam system of a PWR nuclear power plant, about 18 months after a steam generator replacement had been carried out. The magnitude of the vibrations reached levels at which the operators had to reduce power in order to stay within the operating limits imposed by the nuclear inspectorate. To solve the problem the following analyses methods were employed: - Testing the influence on vibration level from different modes of plant operation; - Analyses of plant measurement data; - Calculations of: hydraulic behaviour of the system, structural dynamic behaviour of the system, flow at the steam generator outlet. Scale model testing of the steam generator outlet region. Hydraulic flow disturbances in the main steam system were measured using pressure and strain gauges, which made it possible to track individual pressure pulses propagating through the main steam system. Analyses showed that the pressure pulses causing the vibration originated from the vicinity of the steam generator outlet. By using computer codes for network fluid flow analyses the pressure pulses found in the measurement traces could be generated in calculations. Careful studies of the flow at the steam generator outlet region, using model testing in a 1:3 scale model as well as transient 3D CFD calculations, gave clear indications that flow separation occurred at the steam generator outlet nozzle and at the first bend. Finally, by substituting the outlet nozzle for a different design with a multi-port nozzle, the steam line vibration problem has been solved. (authors)

  13. AREVA main steam line break fully coupled methodology based on CATHARE-ARTEMIS - 15496

    International Nuclear Information System (INIS)

    Denis, L.; Jasserand, L.; Tomatis, D.; Segond, M.; Royere, C.; Sauvage, J.Y.

    2015-01-01

    The CATHARE code developed since 1979 by AREVA, CEA, EDF and IRSN is one of the major thermal-hydraulic system codes worldwide. In order to have at disposal realistic methodologies based on CATHARE for the whole transient and accident analysis in Chapter 15 of Safety Reports, a coupling with the code ARTEMIS was developed. ARTEMIS is the core code in AREVA's new reactor simulator system ARCADIA, using COBRA-FLX to model the thermal-hydraulics in the core. The Fully Coupled Methodology was adapted to the CATHARE-ARTEMIS coupling to perform Main Steam Line Break studies. This methodology, originally applied to the MANTA-SMART-FLICA coupling, is dedicated to Main Steam Line Break transients at zero power. The aim of this paper is to present the coupling between CATHARE and ARTEMIS and the application of the Fully Coupled Methodology in a different code environment. (authors)

  14. Assessment of vibration anomalies of main steam lines at Palo Verde-3

    International Nuclear Information System (INIS)

    Amr, A.; Landstrom, C.; Maxwell, H.; Miller, J.S.; Lynch, J.J.

    1996-01-01

    Historically, flow induced vibration in piping systems that transport liquid has presented problems for plant designers. When evaluating a vibration problem, it is always important to determine the forcing frequencies from different phenomena and the natural frequencies of the system as an integral part of establishing the root cause of the problem. Since in most cases of large vibration and noise levels, the natural frequency of the system and the frequency of the flow induced vibration are very close, determining the natural frequency of the system is important. Palo Verde Unit-3 exhibited a vibration problem where identification of the root cause was difficult. A Palo Verde team was created which consisted of engineers from different on-site departments and support from consultants. The process used to determine the root cause for the vibration/noise problem on Main Steam Supply System (MSSS) steam line 2 at Palo Verde Unit 3 is discussed in this paper. Since the root cause was not readily apparent, a finite element model was constructed to determine the natural frequency of the piping system. The finite element model consisted of a portion of the main steam lines, including a sample line which traverses the main steam line

  15. Simulation of a main steam line break with steam generator tube rupture using trace

    Energy Technology Data Exchange (ETDEWEB)

    Gallardo, S.; Querol, A.; Verdu, G. [Departamento de Ingenieria Quimica Y Nuclear, Universitat Politecnica de Valencia, Camino de Vera s/n, 46022, Valencia (Spain)

    2012-07-01

    A simulation of the OECD/NEA ROSA-2 Project Test 5 was made with the thermal-hydraulic code TRACE5. Test 5 performed in the Large Scale Test Facility (LSTF) reproduced a Main Steam Line Break (MSLB) with a Steam Generator Tube Rupture (SGTR) in a Pressurized Water Reactor (PWR). The result of these simultaneous breaks is a depressurization in the secondary and primary system in loop B because both systems are connected through the SGTR. Good approximation was obtained between TRACE5 results and experimental data. TRACE5 reproduces qualitatively the phenomena that occur in this transient: primary pressure falls after the break, stagnation of the pressure after the opening of the relief valve of the intact steam generator, the pressure falls after the two openings of the PORV and the recovery of the liquid level in the pressurizer after each closure of the PORV. Furthermore, a sensitivity analysis has been performed to know the effect of varying the High Pressure Injection (HPI) flow rate in both loops on the system pressures evolution. (authors)

  16. Validation cases of CATHARE 2 for VVER-1000 main steam line break analysis

    International Nuclear Information System (INIS)

    Kolev, Nikolay P.; Petrov, Nikolay; Donov, Jordan; Sabotinov, Luben; Nikonov, Sergey

    2008-01-01

    Recent coupled code benchmarks identified coolant mixing in the reactor vessel as an unresolved issue in the analysis of complex plan transients with reactivity insertion. Thus, Phase 2 of the OECD VVER-1000 Coolant Transient Benchmark (V1000CT-2) was defined. The benchmark includes calculation of vessel mixing tests and main steam line break (MSLB) analysis. The reference plant is Kozloduy-6 in Bulgaria. The general objective is the assessment of system codes for VVER safety analysis and specifically for their use in the analysis of reactivity transients. A specific objective is the testing of different scale mixing models (mixing matrix, multi-1D, coarse-3D and CFD), and analysis of MSLB transients with improved vessel thermal hydraulic models. The benchmark is sponsored by CEA-France and OECD and is jointly prepared by CEA and INRNE, in collaboration with the Kozloduy NPP, IRSN and PSU. This paper summarizes CATHARE2 code assessment calculations using multi-1D vessel thermal hydraulics with cross flow. Test cases are the OECD V1000CT-1 pump start-up benchmark and the V1000CT-2 benchmarks. Emphasis is put on vessel mixing aspects. Separate effects in the lower plenum as well as component and integral system tests are considered. The comparison shows that a six-sector vessel mixing model informed by plant data or validated CFD calculations in the initial state was able to correctly reproduce the channel average temperatures at the core inlet as well as the vessel outlet temperatures. Testing at system level including code-to-experiment and CATHARE-ATHLET comparison shows that the considered CATHARE VVER-1000 system model is capable of MSLB simulation. (author)

  17. R.B. pressure and temperature transient following main steam line break

    International Nuclear Information System (INIS)

    Das, M.; Bhawal, R.N.; Prakash, P.

    1989-01-01

    The R.B. containment plays an important role in mitigating the consequences of any accident core. The analysis of Main Steam Line Break (MSLB), though not of relevance from activity release considerations, is essentially from structural integrity point of view. In this paper the outline of the likely scenario is drawn and the approach for thermal hydraulic simulation of the system for carrying out transient blowdown analysis is discussed. The results of the containment pressure and temperature transient analysis are also presented. (author). 4 refs., 7 figs

  18. Analysis of the OECD main steam line break benchmark using ANC-K/MIDAC code

    International Nuclear Information System (INIS)

    Aoki, Shigeaki; Tahara, Yoshihisa; Suemura, Takayuki; Ogawa, Junto

    2004-01-01

    A three-dimensional (3D) neutronics and thermal-and-hydraulics (T/H) coupling code ANC-K/MIDAC has been developed. It is the combination of the 3D nodal kinetic code ANC-K and the 3D drift flux thermal hydraulic code MIDAC. In order to verify the adequacy of this code, we have performed several international benchmark problems. In this paper, we show the calculation results of ''OECD Main Steam Line Break Benchmark (MSLB benchmark)'', which gives the typical local power peaking problem. And we calculated the return-to-power scenario of the Phase II problem. The comparison of the results shows the very good agreement of important core parameters between the ANC-K/MIDAC and other participant codes. (author)

  19. Development of phenomena identification and ranking table for APR1400 main steam line break

    International Nuclear Information System (INIS)

    Song, J. H.; Chung, B. D.; Jeong, J. J.

    2003-01-01

    A Phenomena Identification and Ranking Table (PIRT) was developed for the Main Steam Line Break (MSLB) event of an APR-1400 (Advanced Power Reactor-1400). A team of experts from research institutes, industries, and regulatory bodies participated in the development. The selected event was a double-ended steam line break at full power with the reactor coolant pump running. The panel selected the fuel performance as the primary safety criterion for ranking. The plant design data, the results of APR-1400 safety analysis, and the results of additional best estimate analysis by MARS2.1 were utilized. Three phases of pre-trip, rapid cool-down, and safety injection phase were identified. Then, the ranking of a system, components, phenomenon/process based on the relative importance to the primary evaluation criterion were followed for each time phase. Finally, the knowledge-level for each important process in the component was ranked in terms of the existing knowledge. The highly ranked phenomena identified for APR-1400 MSLB are tube wall heat transfer at the steam generator shell, void distribution at the steam generator shell, liquid entrainment in the separators, mixture level in the separators, boron mixing in the upper down comer, boron transport and thermal mixing in the lower plenum, stored energy release in the upper head, and flow to and/from the upper head. The PIRT will be used as a guide in planning cost effective experimental programs and code development efforts, especially for the quantification of the process and/or phenomena, which have a high importance but low knowledge level

  20. Lungmen ABWR containment analyses during short-term main steam line break LOCA using GOTHIC

    International Nuclear Information System (INIS)

    Chen, Yen-Shu; Yuann, Yng-Ruey; Dai, Liang-Che

    2012-01-01

    Highlights: ► The Lungmen ABWR containment responses due to the main steam line break are analyzed. ► In the Lungmen FSAR, the peak drywell temperature is greater than the designed value. ► GOTHIC is used to calculate the containment responses in this study. ► With more realistic conditions, the drywell temperature can be reasonably suppressed. - Abstract: Lungmen Nuclear Power Plant in Taiwan is a GE-designed twin-unit Advanced Boiling Water Reactor (ABWR) plant with rated thermal power of 3926 MWt. Both units are currently under construction. In the Lungmen Final Safety Analysis Report (FSAR) section 6.2, the calculated peak drywell temperature during the short-term Main Steam Line Break (MSLB) event is 176.3 °C, which is greater than the designed temperature of 171.1 °C. It resulted in a controversial issue in the FSAR review process conducted by the Atomic Energy Council in Taiwan. The purpose of this study is to independently investigate the Lungmen ABWR containment pressure and temperature responses to the MSLB using the GOTHIC program. Blowdown conditions are either calculated by using a simplified reactor vessel volume in GOTHIC model, or provided by the RELAP5 transient analysis. The blowdown flow rate from the steam header side is calculated with a more reasonable pressure loss coefficient of the open main steam isolation valves, and the peak drywell temperature is then reduced. By using the RELAP5 blowdown data, the peak drywell temperature can be further reduced because of the initial liquid entrainment in the blowdown flow. The drywell space is either treated as a single volume, or separated into a upper drywell and a lower drywell to reflect the real configuration of the Lungmen containment. It is also found that a single drywell volume may not present the overheating of the upper drywell. With more realistic approaches and assumptions, the drywell temperature can be reasonably below the design value and the Lungmen containment integrity

  1. Simulation and analysis of a main steam line transient with isolation valves closure and subsequent pipe break

    Energy Technology Data Exchange (ETDEWEB)

    Stevanovic, Vladimir; Studovic, Milovan [Faculty of Mechanical Engineering, University of Belgrade, Belgrade (Yugoslavia); Bratic, Aleksandar [Thermal Power Plant Nikola Tesla (Yugoslavia)

    1993-11-01

    Simulation and analysis of a real main steam line break transient at the coal fired 300 MW Drmno Thermal Power Plant have been performed by the computer code TEA-01. The methods and procedures used could be applied to a nuclear power plant. 9 refs., 6 figs.

  2. Structural analysis of steam generator internals following feed water main steam line break: DLF approach

    International Nuclear Information System (INIS)

    Bhasin, Vivek; Kushwaha, H.S.; Mahajan, S.C.; Kakodkar, A.

    1993-01-01

    In order to evaluate the possible release of radioactivity in extreme events, some postulated accidents are analysed and studied during the design stage of Steam Generator (SG). Among the various accidents postulated, the most important are Feed Water Line Break (FWLB) and Main Steam Line Break (MSLB). This report concerns with dynamic structural analysis of SG internals following FWLB/MSLB. The pressure/drag-force time histories considered were corresponding to the conditions leading to the accident of maximum potential. The SG internals were analysed using two approaches of structural dynamics. In first approach simplified DLF method was adopted. This method yields an upper bound values of stresses and deflection. In the second approach time history analysis by Mode Superposition Technique was adopted. This approach gives more realistic results. The structure was qualified as per ASME B and PV Code SecIII NB. It was concluded that in all the components except perforated flow distribution plate, the stress values based on elastic analysis are within the limits specified by ASME Code. In case of perforated flow distribution plate during the MSLB transient the stress values based on elastic analysis are higher than the ASME Code limits. Therefore, its limit load analysis had to be done. Finally, the collapse pressure evaluated using limit load analysis was shown to be within the limits of ASME B and PV Code SecIII Nb. (author). 31 refs., 94 figs., 16 tabs

  3. Signal analysis of acoustic and flow-induced vibrations of BWR main steam line

    Energy Technology Data Exchange (ETDEWEB)

    Espinosa-Paredes, G., E-mail: gepe@xanum.uam.mx [División de Ciencias Básicas e Ingeniería, Universidad Autónoma Metropolitana-Iztapalapa, México, D.F. 09340 (Mexico); Prieto-Guerrero, A. [División de Ciencias Básicas e Ingeniería, Universidad Autónoma Metropolitana-Iztapalapa, México, D.F. 09340 (Mexico); Núñez-Carrera, A. [Comisión Nacional de Seguridad Nuclear y Salvaguardias, Doctor Barragán 779, Col. Narvarte, México, D.F. 03020 (Mexico); Vázquez-Rodríguez, A. [División de Ciencias Básicas e Ingeniería, Universidad Autónoma Metropolitana-Iztapalapa, México, D.F. 09340 (Mexico); Centeno-Pérez, J. [Instituto Politécnico Nacional, Escuela Superior de Física y Matemáticas Unidad Profesional “Adolfo López Mateos”, Av. IPN, s/n, México, D.F. 07738 (Mexico); Espinosa-Martínez, E.-G. [Departamento de Sistemas Energéticos, Universidad Nacional Autónoma de México, México, D.F. 04510 (Mexico); and others

    2016-05-15

    Highlights: • Acoustic and flow-induced vibrations of BWR are analyzed. • BWR performance after extended power uprate is considered. • Effect of acoustic side branches (ASB) is analyzed. • The ASB represents a reduction in the acoustic loads to the steam dryer. • Methodology developed for simultaneous analyzing the signals in the MSL. - Abstract: The aim of this work is the signal analysis of acoustic waves due to phenomenon known as singing in Safety Relief Valves (SRV) of the main steam lines (MSL) in a typical BWR5. The acoustic resonance in SRV standpipes and fluctuating pressure is propagated from SRV to the dryer through the MSL. The signals are analyzed with a novel method based on the Multivariate Empirical Mode Decomposition (M-EMD). The M-EMD algorithm has the potential to find common oscillatory modes (IMF) within multivariate data. Based on this fact, we implement the M-EMD technique to find the oscillatory mode in BWR considering the measurements obtained collected by the strain gauges located around the MSL. These IMF, analyzed simultaneously in time, allow obtaining an estimation of the effects of the multiple-SRV in the MSL. Two scenarios are analyzed: the first is the signal obtained before the installation of the acoustic dampers (ASB), and the second, the signal obtained after installation. The results show the effectiveness of the ASB to damp the strong resonances when the steam flow increases, which represents an important reduction in the acoustic loads to the steam dryer.

  4. Numerical simulations of pressure fluctuations at branch piping in BWR main steam line

    International Nuclear Information System (INIS)

    Morita, Ryo; Inada, Fumio; Yoshikawa, Kazuhiro; Takahashi, Shiro

    2009-01-01

    The power uprating of a nuclear power plant may increase/accelerate degradation phenomena such as flow-induced vibration and wall thinking. A steam dryer was damaged by a high cycle fatigue due to an acoustic-induced vibration at the branch piping of safety relief valves (SRVs) in main steam lines. In this study, we conducted the numerical simulations of steam/air flow around a simplified branch piping to clarify the basic characteristics of resonance. LES simulations were conducted in ordinary pressure/temperature air and steam under BWR plant conditions. In both cases, the excitation of the pressure fluctuations at the branch was observed under some inlet velocity conditions. These fluctuations and inlet conditions were normalized and the obtained results were compared. The normalized results showed that the range and maximum amplitude of pressure fluctuations were almost the same in low-pressure/temperature air and high-pressure/temperature steam. We found that ordinary pressure/temperature air experiments and simulations can possibly clarify the characteristics of the resonance in high-pressure/temperature steam. (author)

  5. Main Steam Line Break Mass/Energy and Pressure/Temperature Analysis for the Environmental Qualification

    International Nuclear Information System (INIS)

    Park, Yong-Chan; Song, Dong-Soo; Jun, Hwang-Yong

    2006-01-01

    The Main steam line break(MSLB) occurring inside a reactor containment structure may result in significant releases of high energy fluid to the containment, possibly result in high containment pressure and temperature. The MSLB accident, along with the Loss Of Coolant Accident, is a design basis accident for determining the peak containment pressure and temperature. The analysis for a MSLB for inside containment should be performed to justify the structural integrity and equipment qualification in accordance with revision 1 of Reg. Guide 1.89. Rev1(1984), which is also required as part of obtaining the extended operating license for WestingHouse(WH) 3-Loops Nuclear Power Plant(NPP). Now, the WH NPP has been performed power uprating. Therefore, all initial conditions, setpoints and uncertainties were considered with MSLB analysis for environment qualification(EQ). The transient was analyzed to determine the worst set of mass and energy releases that impact the EQ aspects of safety related equipment inside containment. The most limiting single failure in this event was determined by a sensitivity study. The MSLB event was analyzed for a full set of power conditions and break sizes

  6. Signal analysis of acoustic and flow-induced vibrations of BWR main steam line

    International Nuclear Information System (INIS)

    Espinosa-Paredes, G.; Prieto-Guerrero, A.; Núñez-Carrera, A.; Vázquez-Rodríguez, A.; Centeno-Pérez, J.; Espinosa-Martínez, E.-G.

    2016-01-01

    Highlights: • Acoustic and flow-induced vibrations of BWR are analyzed. • BWR performance after extended power uprate is considered. • Effect of acoustic side branches (ASB) is analyzed. • The ASB represents a reduction in the acoustic loads to the steam dryer. • Methodology developed for simultaneous analyzing the signals in the MSL. - Abstract: The aim of this work is the signal analysis of acoustic waves due to phenomenon known as singing in Safety Relief Valves (SRV) of the main steam lines (MSL) in a typical BWR5. The acoustic resonance in SRV standpipes and fluctuating pressure is propagated from SRV to the dryer through the MSL. The signals are analyzed with a novel method based on the Multivariate Empirical Mode Decomposition (M-EMD). The M-EMD algorithm has the potential to find common oscillatory modes (IMF) within multivariate data. Based on this fact, we implement the M-EMD technique to find the oscillatory mode in BWR considering the measurements obtained collected by the strain gauges located around the MSL. These IMF, analyzed simultaneously in time, allow obtaining an estimation of the effects of the multiple-SRV in the MSL. Two scenarios are analyzed: the first is the signal obtained before the installation of the acoustic dampers (ASB), and the second, the signal obtained after installation. The results show the effectiveness of the ASB to damp the strong resonances when the steam flow increases, which represents an important reduction in the acoustic loads to the steam dryer.

  7. Large scale multi-zone creep finite element modelling of a main steam line branch intersection

    International Nuclear Information System (INIS)

    Payten, Warwick

    2006-01-01

    A number of papers detail the non-linear creep finite element analysis of branch pieces. Predominately these models have incorporated only a single material zone representing the parent material. Multi-zone models incorporating weld material and heat affected zones have primarily been two-dimensional analyses, in part due to the large number of elements required to adequately represent all of the zones. This paper describes a non-linear creep analysis of a main steam line branch intersection using creep properties to represent the parent metal, weld metal, and heat affected zone (HAZ), the stress redistribution over 100,000 h is examined. The results show that the redistribution leads to a complex stress state, particularly at the heat affected zone. Although, there is damage on the external surface of the branch piece as expected, the results indicate that the damage would be more widespread through extensive sections of the heat affected zone. This would appear to indicate that the time between damage indications on the surface using techniques such as replication and full thickness damage may be more limited then previously expected

  8. Containment fan cooler heat transfer calculation during main steam line break for Maanshan PWR plant

    Energy Technology Data Exchange (ETDEWEB)

    Yuann, Yng-Ruey, E-mail: ryyuann@iner.gov.tw; Kao, Lain-Su, E-mail: lskao@iner.gov.tw

    2013-10-15

    Highlights: • Evaluate component cooling water (CCW) thermal response during MSLB for Maanshan. • Using GOTHIC to calculate CCW temperature and determine time required to boil CCW. • Both convective and condensation heat transfer from the air side are considered. • Boiling will not occur since T{sub B} is sufficiently longer than CCW pump restart time. -- Abstract: A thermal analysis has been performed for the Containment Fan Cooler Unit (FCU) during Main Steam Line Break (MSLB) accident, concurrent with loss of offsite power, for Maanshan PWR plant. The analysis is performed in order to address the waterhammer and two-phase flow issues discussed in USNRC's Generic Letter 96-06 (GL 96-06). Maanshan plant is a twin-unit Westinghouse 3-loop PWR currently operated at rated core thermal power of 2822 MWt for each unit. The design basis for containment temperature is Main Steam Line Break (MSLB) accident at power of 2830.5 MWt, which results in peak vapor temperature of 387.6 °F. The design is such that when MSLB occurs concurrent with loss of offsite power (MSLB/LOOP), both the coolant pump on the secondary side and the fan on the air side of the FCU loose power and coast down. The pump has little inertia and coasts down in 2–3 s, while the FCU fan coasts down over much longer period. Before the pump is restored through emergency diesel generator, there is potential for boiling the coolant in the cooling coils by the high-temperature air/steam mixture entering the FCU. The time to boiling depends on the operating pressure of the coolant before the pump is restored. The prediction of the time to boiling is important because it determines whether there is potential for waterhammer or two-phase flow to occur before the pump is restored. If boiling occurs then there exists steam region in the pipe, which may cause the so called condensation induced waterhammer or column closure waterhammer. In either case, a great amount of effort has to be spent to

  9. Sensitivity Studies for Main Steam Line Break Exercises 2 and 3 with RELAP5/PANBOX

    International Nuclear Information System (INIS)

    Boeer, Rainer; Knoll, Alfred

    2003-01-01

    This paper presents and discusses results obtained with the nuclear plant safety analysis code system RELAP5/PANBOX (R/P/C) for the return-to-power scenario of exercises 2 and 3 of the Organization for Economic Cooperation and Development/Nuclear Energy Agency Main Steam Line Break (MSLB) Benchmark. Both the external and internal coupling options of R/P/C have been considered for exercise 3; i.e., the COBRA module of PANBOX was used to calculate the core thermal hydraulics in the external coupling option, whereas the core thermal hydraulics of RELAP5 was used in the internal coupling option. For the representation of thermal-hydraulic channels, a fine channel geometry based on the 177 fuel assemblies was selected for the external coupling option, and a coarse channel geometry based on 19 coarse channels has been investigated for the internal coupling option. The comparison of the results shows very good agreement of important core parameters between the considered coupling variants. Both exercises 2 and 3 have been investigated with respect to local safety parameters like fuel centerline temperatures and minimum departure from nucleate boiling ratios using the on-line hot subchannel analysis capability of R/P/C in the external coupling option. The results show that both quantities are far from the safety-related limits.The benchmark demonstrates, that R/P/C - as part of the integrated CASCADE-3D core analysis system of Framatome ANP GmbH - has proven to be a powerful tool for detailed analyses of an MSLB accident

  10. Main Steam Line Break Analysis for the Fully Passive Safety System of SMART

    International Nuclear Information System (INIS)

    Kim, Seong Wook; Chun, Ji Han; Bae, Kyoo Hwan; Kim, Keung Koo

    2013-01-01

    The standard design approval of SMART (System-integrated Modular Advanced ReacTor) developed by KAERI and KEPCO consortium was issued on July 4, 2012. Although SMART has enhanced safety compared to the conventional reactor, there is a demand to meet the 'passive safety performance requirements' after the Fukushima accident. The passive safety performance requirements are the capabilities to maintain the plant at a safe shutdown condition for a minimum of 72 hours without AC power supply or operator action in case of design basis accident (DBA). To satisfy the requirements, KAERI is developing a safety enhanced SMART by adopting a passive safety injection system. The passive safety injection system developed for SMART is a gravity-driven injection system, which consists of four trains, each of which includes a pressure balance line, core makeup tank (CMT), safety injection tank (SIT) and injection line. The CMT plays an important role to inject borated water into the RCS to prevent or dissolve the return to power (re-criticality) condition during the event of increase in heat removal by the secondary system. The main steam line break accident (MSLB) is the most limiting accident for an increase in heat removal by the secondary system. In this study, the safety analysis results of MSLBs at hot full power condition and at hot zero power condition in view of re-criticality are given. The MSLB accident has been analyzed for the SMART adopting fully passive safety system in the aspect of re-criticality. The results show that the core remains subcritical condition throughout the transient due to the borated water injected by the CMT. As further works, many kinds of analyses and sensitivity studies should be performed for the design establishment and improvement of the fully passive system of SMART

  11. Effects of the Pressurized Water Reactor Main Steam Line Break Location on the Blowdown Loading

    International Nuclear Information System (INIS)

    Jo, Jong Chull; Kang, Soon Ho; Chan, Won Joon

    2016-01-01

    The thermal hydraulic analysis has been performed generally using a simple lumped model or one dimensional numerical model. However, those models have limitations in predicting the transient variations of the steam velocity, pressure and hydrodynamic load at a local point and the most conservative condition. Furthermore, it cannot be confirmed if the blowdown loads predicted by either of the models are conservative to evaluate every part of the SG internal structures. In this study, the transient hydraulic response of the SG secondary side to the MSLB case where the pipe break is assumed to occur at the SG outlet nozzle end, another weld point on the MSL, was numerically simulated using a CFD code. The present CFD calculation results was compared with those in ref. to investigate the effect of break location (friction loss) on the blowdown load in the SG secondary side. The result shows that the friction loss along the steam line span between the SG nozzle end and the MSIV would cause reduction in steam velocity disturbance or dynamic pressure. It implies that the consequence of the MSLB at the SG nozzle end would be much severer that those of other MSLB cases where the break locations are far from the SG. Therefore, to assure a conservative safety evaluation of the SG structural integrity, the blowdown loading on the SG internal structures including tubes during a MSLB accident in terms of the transient steam velocity, dynamic pressure and decompression wave fluctuations should be assessed for the MSLB case where the break is assumed to occur at the SG nozzle end.

  12. Effects of the Pressurized Water Reactor Main Steam Line Break Location on the Blowdown Loading

    Energy Technology Data Exchange (ETDEWEB)

    Jo, Jong Chull; Kang, Soon Ho; Chan, Won Joon [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The thermal hydraulic analysis has been performed generally using a simple lumped model or one dimensional numerical model. However, those models have limitations in predicting the transient variations of the steam velocity, pressure and hydrodynamic load at a local point and the most conservative condition. Furthermore, it cannot be confirmed if the blowdown loads predicted by either of the models are conservative to evaluate every part of the SG internal structures. In this study, the transient hydraulic response of the SG secondary side to the MSLB case where the pipe break is assumed to occur at the SG outlet nozzle end, another weld point on the MSL, was numerically simulated using a CFD code. The present CFD calculation results was compared with those in ref. to investigate the effect of break location (friction loss) on the blowdown load in the SG secondary side. The result shows that the friction loss along the steam line span between the SG nozzle end and the MSIV would cause reduction in steam velocity disturbance or dynamic pressure. It implies that the consequence of the MSLB at the SG nozzle end would be much severer that those of other MSLB cases where the break locations are far from the SG. Therefore, to assure a conservative safety evaluation of the SG structural integrity, the blowdown loading on the SG internal structures including tubes during a MSLB accident in terms of the transient steam velocity, dynamic pressure and decompression wave fluctuations should be assessed for the MSLB case where the break is assumed to occur at the SG nozzle end.

  13. Simulation of the OECD Main-Steam-Line-Break Benchmark Exercise 3 Using the Coupled RELAP5/PANTHER Codes

    International Nuclear Information System (INIS)

    Schneidesch, Christophe R.; Zhang Jinzhao

    2004-01-01

    The RELAP5 best-estimate thermal-hydraulic system code has been coupled with the PANTHER three-dimensional neutron kinetics code via the TALINK dynamic data exchange control and processing tool. The coupled RELAP5/PANTHER code package has been qualified and will be used at Tractebel Engineering (TE) for analyzing asymmetric pressurized water reactor (PWR) accidents with strong core-system interactions. The Organization for Economic Cooperation and Development/U.S. Nuclear Regulatory Commission PWR main-steam-line-break benchmark problem was analyzed as part of the qualification efforts to demonstrate the capability of the coupled code package of simulating such transients. This paper reports the main results of TE's contribution to the benchmark Exercise 3

  14. HEXTRAN-SMABRE calculation of the 6th AER Benchmark, main steam line break in a WWER-440 NPP

    International Nuclear Information System (INIS)

    Haemaelaeinen, A.; Kyrki-Rajamaeki, R.

    2003-01-01

    The sixth AER benchmark is the second AER benchmark for couplings of the thermal hydraulic codes and three dimensional neutron kinetic core models. It concerns a double end break of one main steam line in a WWER-440 plant. The core is at the end of its first cycle in full power conditions. In VTT HEXTRAN2.9 is used for the core kinetics and dynamics and SMABRE4.8 as a thermal hydraulic model for the primary and secondary loop. The plant model for SMABRE consists mainly of two input models, Loviisa model and a standard WWER-440/213 plant model. The primary loop includes six separate loops, the pressure vessel is divided into six parallel channels in SMABRE and the whole core calculation is performed in the core with HEXTRAN. The horizontal steam generators are modelled with heat transfer tubes in five levels and vertically with two parts, riser and downcomer. With this kind of detailed modelling of steam generators there occurs strong flashing after break opening. As a sequence of the main steam line break at nominal power level, the reactor trip is followed quite soon. The liquid temperature continues to decrease in one core inlet sector which may lead to recriticality and neuron power increase. The situation is very sensitive to small changes in the steam generator and break flow modelling and therefore several sensitivity calculations have been done. Also two stucked control rods have been assumed. Due to boric acid concentration in the high pressure safety injection subcriticality is finally guaranteed in the transient (Authors)

  15. Study of condensation heat transfer following a main steam line break inside containment

    Energy Technology Data Exchange (ETDEWEB)

    Cho, J.H.; Elia, F.A. Jr.; Lischer, D.J. [Stone & Webster Engineering Corporation, Boston, MA (United States)

    1995-09-01

    An alternative model for calculating condensation heat transfer following a main stream line break (MSLB) accident is proposed. The proposed model predictions and the current regulatory model predictions are compared to the results of the Carolinas Virginia Tube Reactor (CVTR) test. The very conservative results predicted by the current regulatory model result from: (1) low estimate of the condensation heat transfer coefficient by the Uchida correlation and (2) neglecting the convective contribution to the overall heat transfer. Neglecting the convection overestimates the mass of steam being condensed and does not permit the calculation of additional convective heat transfer resulting from superheated conditions. In this study, the Uchida correlation is used, but correction factors for the effects of convection an superheat are derived. The proposed model uses heat and mass transfer analogy methods to estimate to convective fraction of the total heat transfer and bases the steam removal rate on the condensation heat transfer portion only. The results predicted by the proposed model are shown to be conservative and more accurate than those predicted by the current regulatory model when compared with the results of the CVTR test. Results for typical pressurized water reactors indicate that the proposed model provides a basis for lowering the equipment qualification temperature envelope, particularly at later times following the accident.

  16. Analysis of containment parameters during the main steam line break with the failure of the feedwater control valves

    International Nuclear Information System (INIS)

    Fabjan, L.; Petelin, S.; Mavko, B.; Gortnar, O.; Tiselj, I.

    1992-01-01

    U.S. Nuclear Regulatory Commission (NRC) information notice 91-69: 'Errors in Main Steam Line Break Analyses for Determining Containment Parameters' shows the possibility of an accident which could lead to beyond design containment pressure and temperature. Such accident would be caused by the continuation of feedwater flow following a main stream line break (MSLB) inside the containment. Krsko power plant already experienced problems with main feedwater control valves. For that reason, analysis of MSLB has been performed taking into account continuous feedwater addition scenario and different containment safety systems capabilities availability. Steam and water released into the containment during MSLB was calculated using RELAP5/MOD2 computer code. The containment response to MSLB was calculated using CONTEMPT-LT/028 computer code. The results indicated that the continuous feedwater flow following a MSLB could lead to beyond design containment pressure. The peak pressure and temperature depend on isolation time for main- and auxiliary-feedwater supply. In the case of low boron concentration injection, the core recriticality is characteristic for this type of accidents. It was concluded that the presented analysis of MSLB with continuous feedwater addition scenario is the worst case for containment design

  17. Evaluation of acoustic resonance at branch section in main steam line. Part 1. Effects of steam wetness on acoustic resonance

    International Nuclear Information System (INIS)

    Uchiyama, Yuta; Morita, Ryo

    2011-01-01

    The power uprating of the nuclear power plant (NPP) is conducted in United States, EU countries and so on, and also is planned in Japan. However, the degradation phenomena such as flow-induced vibration and wall thinning may increase or expose in the power uprate condition. In U.S. NPP, the dryer had been damaged by high cycle fatigue due to acoustic-induced vibration under a 17% extended power uprating (EPU) condition. This is caused by acoustic resonance at the stub pipes of safety relief valves (SRVs) in the main steam lines (MSL). Increased velocity by uprating excites the pressure fluctuations and makes large amplitude resonance. To evaluate the acoustic resonance at the stub pipes of SRVs in actual BWR, it is necessary to clarify the acoustic characteristics in steam flow. Although there are several previous studies about acoustic resonance, most of them are not steam flow but air flow. Therefore in this study, to investigate the acoustic characteristics in steam flow, we conducted steam flow experiments in each dry and wet steam conditions, and also nearly saturated condition. We measured pressure fluctuation at the top of the single stub pipe and in main steam piping. As a result, acoustic resonance in dry steam flow could be evaluated as same as that in air flow. It is clarified that resonance amplitude of fluctuating pressure at the top of the stub pipe in wet steam was reduced to one-tenth compared with that in dry. (author)

  18. Evaluation of the Main Steam Line Break Accident for the APR+ Standard Design using MARS-KS

    Energy Technology Data Exchange (ETDEWEB)

    Park, M. H.; Kim, Y. S.; Hwang, Min Jeong; Sim, S. K. [Environment Energy Technology, Daejeon (Korea, Republic of); Bang, Young Seok [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2015-05-15

    As a part of licensing evaluation of the APR+ (Advanced Power Reactor +) standard design, Korea Institute of Nuclear Safety(KINS) performed safety evaluation of the APR+ Standard Safety Analysis Report(SSAR). The results of the safety evaluation of the APR+ Main Steam Line Break(MSLB) accident is presented for the most limiting post-trip return-to-power case with the single failure assumption of the Loss Of Offsite Power(LOOP). MARS-KS regulatory safety analysis code has been used to evaluate safety as well as the system behavior during MSLB accident. The MARS-KS analysis results are compared with the results of the MSLB safety analysis presented in the SSAR of the APR+. Independent safety evaluation has been performed using MARS-KS regulatory safety analysis code for the APR+ MSLB accident inside containment for the limiting case of the full power post-trip return-to-power. The results of MARS-KS analysis were compared with the results of the MSLB safety analysis presented in the APR+ SSAR. Due to higher cooldown of the MARS-KS analysis, the MARS-KS analysis results in a higher positive reactivity insertion into the core by the negative moderator and fuel temperature reactivity coefficients than the APR+ SSAR analysis. Both results show no return-to-power during the limiting case of the MSLB inside containment. However, APR+ SSAR moderator temperature reactivity insertion should be evaluated against the MARS-KS moderator density reactivity insertion for is conservatism. This study also clearly shows asymmetric thermal hydraulic behavior during the MSLB accident at intact and affected sides of the downcomer and the core. These asymmetric phenomena should be further investigated for the effects on the system design.

  19. Effects of Secondary Circuit Modeling on Results of Pressurized Water Reactor Main Steam Line Break Benchmark Calculations with New Coupled Code TRAB-3D/SMABRE

    International Nuclear Information System (INIS)

    Daavittila, Antti; Haemaelaeinen, Anitta; Kyrki-Rajamaeki, Riitta

    2003-01-01

    All of the three exercises of the Organization for Economic Cooperation and Development/Nuclear Regulatory Commission pressurized water reactor main steam line break (PWR MSLB) benchmark were calculated at VTT, the Technical Research Centre of Finland. For the first exercise, the plant simulation with point-kinetic neutronics, the thermal-hydraulics code SMABRE was used. The second exercise was calculated with the three-dimensional reactor dynamics code TRAB-3D, and the third exercise with the combination TRAB-3D/SMABRE. VTT has over ten years' experience of coupling neutronic and thermal-hydraulic codes, but this benchmark was the first time these two codes, both developed at VTT, were coupled together. The coupled code system is fast and efficient; the total computation time of the 100-s transient in the third exercise was 16 min on a modern UNIX workstation. The results of all the exercises are similar to those of the other participants. In order to demonstrate the effect of secondary circuit modeling on the results, three different cases were calculated. In case 1 there is no phase separation in the steam lines and no flow reversal in the aspirator. In case 2 the flow reversal in the aspirator is allowed, but there is no phase separation in the steam lines. Finally, in case 3 the drift-flux model is used for the phase separation in the steam lines, but the aspirator flow reversal is not allowed. With these two modeling variations, it is possible to cover a remarkably broad range of results. The maximum power level reached after the reactor trip varies from 534 to 904 MW, the range of the time of the power maximum being close to 30 s. Compared to the total calculated transient time of 100 s, the effect of the secondary side modeling is extremely important

  20. Evaluation of a main steam line break with induced, multiple tube ruptures: A comparison of NUREG 1477 (Draft) and transient methodologies Palo Verde Nuclear Generating Station

    Energy Technology Data Exchange (ETDEWEB)

    Parrish, K.R.

    1995-09-01

    This paper presents the approach taken to analyze the radiological consequences of a postulated main steam line break event, with one or more tube ruptures, for the Palo Verde Nuclear Generating Station. The analysis was required to support the restart of PVNGS Unit 2 following the steam generator tube rupture event on March 14, 1993 and to justify continued operation of Units 1 and 3. During the post-event evaluation, the NRC expressed concern that Unit 2 could have been operating with degraded tubes and that similar conditions could exist in Units 1 and 3. The NRC therefore directed that a safety assessment be performed to evaluate a worst case scenario in which a non-isolable main steam line break occurs inducing one or more tube failures in the faulted steam generator. This assessment was to use the generic approach described in NUREG 1477, Voltage-Based Interim Plugging Criteria for Steam Generator Tubes - Task Group Report. An analysis based on the NUREG approach was performed but produced unacceptable results for off-site and control room thyroid doses. The NUREG methodology, however, does not account for plant thermal-hydraulic transient effects, system performance, or operator actions which could be credited to mitigate dose consequences. To deal with these issues, a more detailed analysis methodology was developed using a modified version of the Combustion Engineering Plant Analysis Code, which examines the dose consequences for a main steam line break transient with induced tube failures for a spectrum equivalent to 1 to 4 double ended guillotine U-tube breaks. By incorporating transient plant system responses and operator actions, the analysis demonstrates that the off-site and control room does consequences for a MSLBGTR can be reduced to acceptable limits. This analysis, in combination with other corrective and recovery actions, provided sufficient justification for continued operation of PVNGS Units 1 and 3, and for the subsequent restart of Unit 2.

  1. Evaluation of a main steam line break with induced, multiple tube ruptures: A comparison of NUREG 1477 (Draft) and transient methodologies Palo Verde Nuclear Generating Station

    International Nuclear Information System (INIS)

    Parrish, K.R.

    1995-01-01

    This paper presents the approach taken to analyze the radiological consequences of a postulated main steam line break event, with one or more tube ruptures, for the Palo Verde Nuclear Generating Station. The analysis was required to support the restart of PVNGS Unit 2 following the steam generator tube rupture event on March 14, 1993 and to justify continued operation of Units 1 and 3. During the post-event evaluation, the NRC expressed concern that Unit 2 could have been operating with degraded tubes and that similar conditions could exist in Units 1 and 3. The NRC therefore directed that a safety assessment be performed to evaluate a worst case scenario in which a non-isolable main steam line break occurs inducing one or more tube failures in the faulted steam generator. This assessment was to use the generic approach described in NUREG 1477, Voltage-Based Interim Plugging Criteria for Steam Generator Tubes - Task Group Report. An analysis based on the NUREG approach was performed but produced unacceptable results for off-site and control room thyroid doses. The NUREG methodology, however, does not account for plant thermal-hydraulic transient effects, system performance, or operator actions which could be credited to mitigate dose consequences. To deal with these issues, a more detailed analysis methodology was developed using a modified version of the Combustion Engineering Plant Analysis Code, which examines the dose consequences for a main steam line break transient with induced tube failures for a spectrum equivalent to 1 to 4 double ended guillotine U-tube breaks. By incorporating transient plant system responses and operator actions, the analysis demonstrates that the off-site and control room does consequences for a MSLBGTR can be reduced to acceptable limits. This analysis, in combination with other corrective and recovery actions, provided sufficient justification for continued operation of PVNGS Units 1 and 3, and for the subsequent restart of Unit 2

  2. Phenomena identification and ranking tables for Westinghouse AP600 small break loss-of-coolant accident, main steam line break, and steam generator tube rupture scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, G.E.; Fletcher, C.D.; Davis, C.B. [and others

    1997-06-01

    This report revision incorporates new experimental evidence regarding AP600 behavior during small break loss-of-coolant accidents. This report documents the results of Phenomena Identification and Ranking Table (PIRT) efforts for the Westinghouse AP600 reactor. The purpose of this PIRT is to identify important phenomena so that they may be addressed in both the experimental programs and the RELAP5/MOD3 systems analysis computer code. In Revision of this report, the responses of AP600 during small break loss-of-coolant accident, main steam line break, and steam generator tube rupture accident scenarios were evaluated by a committee of thermal-hydraulic experts. Committee membership included Idaho National Engineering and Environmental Laboratory staff and recognized thermal-hydraulic experts from outside of the laboratory. Each of the accident scenarios was subdivided into separate, sequential periods or phases. Within each phase, the plant behavior is controlled by, at most, a few thermal-hydraulic processes. The committee identified the phenomena influencing those processes, and ranked & influences as being of high, medium, low, or insignificant importance. The primary product of this effort is a series of tables, one for each phase of each accident scenario, describing the thermal-hydraulic phenomena judged by the committee to be important, and the relative ranking of that importance. The rationales for the phenomena selected and their rankings are provided. This document issue incorporates an update of the small break loss-of-coolant accident portion of the report. This revision is the result of the release of experimental evidence from AP600-related integral test facilities (ROSA/AP600, OSU, and SPES) and thermal-hydraulic expert review. The activities associated with this update were performed during the period from June 1995 through November 1996. 8 refs., 26 figs., 42 tabs.

  3. Phenomena identification and ranking tables for Westinghouse AP600 small break loss-of-coolant accident, main steam line break, and steam generator tube rupture scenarios

    International Nuclear Information System (INIS)

    Wilson, G.E.; Fletcher, C.D.; Davis, C.B.

    1997-06-01

    This report revision incorporates new experimental evidence regarding AP600 behavior during small break loss-of-coolant accidents. This report documents the results of Phenomena Identification and Ranking Table (PIRT) efforts for the Westinghouse AP600 reactor. The purpose of this PIRT is to identify important phenomena so that they may be addressed in both the experimental programs and the RELAP5/MOD3 systems analysis computer code. In Revision of this report, the responses of AP600 during small break loss-of-coolant accident, main steam line break, and steam generator tube rupture accident scenarios were evaluated by a committee of thermal-hydraulic experts. Committee membership included Idaho National Engineering and Environmental Laboratory staff and recognized thermal-hydraulic experts from outside of the laboratory. Each of the accident scenarios was subdivided into separate, sequential periods or phases. Within each phase, the plant behavior is controlled by, at most, a few thermal-hydraulic processes. The committee identified the phenomena influencing those processes, and ranked ampersand influences as being of high, medium, low, or insignificant importance. The primary product of this effort is a series of tables, one for each phase of each accident scenario, describing the thermal-hydraulic phenomena judged by the committee to be important, and the relative ranking of that importance. The rationales for the phenomena selected and their rankings are provided. This document issue incorporates an update of the small break loss-of-coolant accident portion of the report. This revision is the result of the release of experimental evidence from AP600-related integral test facilities (ROSA/AP600, OSU, and SPES) and thermal-hydraulic expert review. The activities associated with this update were performed during the period from June 1995 through November 1996. 8 refs., 26 figs., 42 tabs

  4. Comparison of the updated solutions of the 6th dynamic AER Benchmark - main steam line break in a NPP with WWER-440

    International Nuclear Information System (INIS)

    Kliem, S.

    2003-01-01

    The 6 th dynamic AER Benchmark is used for the systematic validation of coupled 3D neutron kinetic/thermal hydraulic system codes. It was defined at The 10 th AER-Symposium. In this benchmark, a hypothetical double ended break of one main steam line at full power in a WWER-440 plant is investigated. The main thermal hydraulic features are the consideration of incomplete coolant mixing in the lower and upper plenum of the reactor pressure vessel and an asymmetric operation of the feed water system. For the tuning of the different nuclear cross section data used by the participants, an isothermal re-criticality temperature was defined. The paper gives an overview on the behaviour of the main thermal hydraulic and neutron kinetic parameters in the provided solutions. The differences in the updated solution in comparison to the previous ones are described. Improvements in the modelling of the transient led to a better agreement of a part of the results while for another part the deviations rose up. The sensitivity of the core power behaviour on the secondary side modelling is discussed in detail (Authors)

  5. Best estimate plus uncertainty analysis of departure from nucleate boiling limiting case with CASL core simulator VERA-CS in response to PWR main steam line break event

    Energy Technology Data Exchange (ETDEWEB)

    Brown, C.S., E-mail: csbrown3@ncsu.edu [Department of Nuclear Engineering, North Carolina State University, 2500 Stinson Drive, Raleigh, NC 27695-7909 (United States); Zhang, H., E-mail: Hongbin.Zhang@inl.gov [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3870 (United States); Kucukboyaci, V., E-mail: kucukbvn@westinghouse.com [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States); Sung, Y., E-mail: sungy@westinghouse.com [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

    2016-12-01

    Highlights: • Best estimate plus uncertainty (BEPU) analyses of PWR core responses under main steam line break (MSLB) accident. • CASL’s coupled neutron transport/subchannel code VERA-CS. • Wilks’ nonparametric statistical method. • MDNBR 95/95 tolerance limit. - Abstract: VERA-CS (Virtual Environment for Reactor Applications, Core Simulator) is a coupled neutron transport and thermal-hydraulics subchannel code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS was applied to simulate core behavior of a typical Westinghouse-designed 4-loop pressurized water reactor (PWR) with 17 × 17 fuel assemblies in response to two main steam line break (MSLB) accident scenarios initiated at hot zero power (HZP) at the end of the first fuel cycle with the most reactive rod cluster control assembly stuck out of the core. The reactor core boundary conditions at the most DNB limiting time step were determined by a system analysis code. The core inlet flow and temperature distributions were obtained from computational fluid dynamics (CFD) simulations. The two MSLB scenarios consisted of the high and low flow situations, where reactor coolant pumps either continue to operate with offsite power or do not continue to operate since offsite power is unavailable. The best estimate plus uncertainty (BEPU) analysis method was applied using Wilks’ nonparametric statistical approach. In this demonstration of BEPU application, 59 full core simulations were performed for each accident scenario to provide the minimum departure from nucleate boiling ratio (MDNBR) at the 95/95 (95% probability with 95% confidence level) tolerance limit. A parametric goodness-of-fit approach was also applied to the results to obtain the MDNBR value at the 95/95 tolerance limit. Initial sensitivity analysis was performed with the 59 cases per accident scenario by use of Pearson correlation coefficients. The results show that this typical PWR core

  6. Best estimate plus uncertainty analysis of departure from nucleate boiling limiting case with CASL core simulator VERA-CS in response to PWR main steam line break event

    International Nuclear Information System (INIS)

    Brown, C.S.; Zhang, H.; Kucukboyaci, V.; Sung, Y.

    2016-01-01

    Highlights: • Best estimate plus uncertainty (BEPU) analyses of PWR core responses under main steam line break (MSLB) accident. • CASL’s coupled neutron transport/subchannel code VERA-CS. • Wilks’ nonparametric statistical method. • MDNBR 95/95 tolerance limit. - Abstract: VERA-CS (Virtual Environment for Reactor Applications, Core Simulator) is a coupled neutron transport and thermal-hydraulics subchannel code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS was applied to simulate core behavior of a typical Westinghouse-designed 4-loop pressurized water reactor (PWR) with 17 × 17 fuel assemblies in response to two main steam line break (MSLB) accident scenarios initiated at hot zero power (HZP) at the end of the first fuel cycle with the most reactive rod cluster control assembly stuck out of the core. The reactor core boundary conditions at the most DNB limiting time step were determined by a system analysis code. The core inlet flow and temperature distributions were obtained from computational fluid dynamics (CFD) simulations. The two MSLB scenarios consisted of the high and low flow situations, where reactor coolant pumps either continue to operate with offsite power or do not continue to operate since offsite power is unavailable. The best estimate plus uncertainty (BEPU) analysis method was applied using Wilks’ nonparametric statistical approach. In this demonstration of BEPU application, 59 full core simulations were performed for each accident scenario to provide the minimum departure from nucleate boiling ratio (MDNBR) at the 95/95 (95% probability with 95% confidence level) tolerance limit. A parametric goodness-of-fit approach was also applied to the results to obtain the MDNBR value at the 95/95 tolerance limit. Initial sensitivity analysis was performed with the 59 cases per accident scenario by use of Pearson correlation coefficients. The results show that this typical PWR core

  7. Analysis of the main steam line break accident with loss of offsite power using the fully coupled RELAP5/PANTHER/COBRA code package

    International Nuclear Information System (INIS)

    Ruben Van Parys; Sandrine Bosso; Christophe Schneidesch; Jinzhao Zhang

    2005-01-01

    Full text of publication follows: A coupled thermal hydraulics-neutronics code package (RELAP5/PANTHER/COBRA) has been qualified for accident analysis at Tractebel Engineering. In the TE coupled code package, the best estimate thermal-hydraulic system code, RELAP5/MOD2.5, is coupled with the full three-dimensional reactor core kinetics code, PANTHER, via a dynamic data exchange control and processing tool, TALINK. An interface between PANTHER code and the sub-channel thermal-hydraulic analysis code COBRA-IIIC is developed in order to perform online calculation of Departure from Nucleate Boiling Ratio (DNBR). The TE coupled code package has been applied to develop a MSLB accident analysis methodology using the TE deterministic bounding approach. The methodology has been applied for MSLB accident analysis in support of licensing of the power up-rate and steam generator replacement of the Doel 2 plant. The results of coupled thermal-hydraulic and neutronic analysis of SLB show that there exists an important margin in the traditional FSAR MSLB accident analysis. As a specific licensing requirement, the main steam line break accident with loss of offsite power has to be analyzed. In the standard methodology with the coupled RELAP5/PANTHER code, and some corrective methods has to be taken in order to overcome the limitations due to the close-channel T/H model in PANTHER at low flow conditions. The results show that the steam line break accident with loss of offsite power is far less limiting. In order to verify the effect of the cross-flow at low flow conditions, the fully dynamic coupling of RELAP5/PANTHER/COBRA code package is used for reanalysis of this case, in which the PANTHER close-channel T/H model is replaced by the COBRA sub-channel T/H model with crossflow option. It has been demonstrated that, although the consideration of cross-flow in this challenging situation may lead to higher core return to power and slightly lower DNBR than in the standard methodology

  8. Analysis of the main steam line break accident with loss of offsite power using the fully coupled RELAP5/PANTHER/COBRA code package

    Energy Technology Data Exchange (ETDEWEB)

    Ruben Van Parys; Sandrine Bosso; Christophe Schneidesch; Jinzhao Zhang [Nuclear Department, Suez-Tractebel Engineering, avenue Ariane 5, B-1200 Brussels (Belgium)

    2005-07-01

    Full text of publication follows: A coupled thermal hydraulics-neutronics code package (RELAP5/PANTHER/COBRA) has been qualified for accident analysis at Tractebel Engineering. In the TE coupled code package, the best estimate thermal-hydraulic system code, RELAP5/MOD2.5, is coupled with the full three-dimensional reactor core kinetics code, PANTHER, via a dynamic data exchange control and processing tool, TALINK. An interface between PANTHER code and the sub-channel thermal-hydraulic analysis code COBRA-IIIC is developed in order to perform online calculation of Departure from Nucleate Boiling Ratio (DNBR). The TE coupled code package has been applied to develop a MSLB accident analysis methodology using the TE deterministic bounding approach. The methodology has been applied for MSLB accident analysis in support of licensing of the power up-rate and steam generator replacement of the Doel 2 plant. The results of coupled thermal-hydraulic and neutronic analysis of SLB show that there exists an important margin in the traditional FSAR MSLB accident analysis. As a specific licensing requirement, the main steam line break accident with loss of offsite power has to be analyzed. In the standard methodology with the coupled RELAP5/PANTHER code, and some corrective methods has to be taken in order to overcome the limitations due to the close-channel T/H model in PANTHER at low flow conditions. The results show that the steam line break accident with loss of offsite power is far less limiting. In order to verify the effect of the cross-flow at low flow conditions, the fully dynamic coupling of RELAP5/PANTHER/COBRA code package is used for reanalysis of this case, in which the PANTHER close-channel T/H model is replaced by the COBRA sub-channel T/H model with crossflow option. It has been demonstrated that, although the consideration of cross-flow in this challenging situation may lead to higher core return to power and slightly lower DNBR than in the standard methodology

  9. VVER-1000 main steam line break analysis using the coupled code system DYN3D/ATHLET

    International Nuclear Information System (INIS)

    Kliem, Soeren; Hoehne, Thomas; Rohde, Ulrich; Weiss, Frank-Peter; Kozmenkov, Yaroslav

    2008-01-01

    Calculations using the coupled code system DYN3D/ATHLET were performed in the frame of the OECD/NEA MSLB benchmark for a VVER-1000 reactor. The coolant mixing inside the reactor pressure vessel was treated using a validated empirical mixing model implemented into the DYN3D/ATHLET code. Using very conservative boundary conditions (reduced scram worth, two stuck rods, running MCP throughout the whole transient) a return-to-power was predicted. For the assessment of the empirical mixing model a time dependent calculation using the computational fluid dynamics code CFX-10 was performed. For that analysis, a detailed model of the reactor pressure vessel consisting of the inlets nozzles, downcomer, lower plenum and a part of the core and having 4.67 million unstructured tetra cell elements was used. For the considered case with running main coolant pumps, this calculation shows a sector formation at the core inlet with a certain amount of mixing at the edges of the sector. A core calculation using these CFX results as boundary conditions predicted also a return-to-power with a maximum value being about 200 MW lower than in the coupled code calculation. This variation calculation confirms the applicability of the empirical mixing model. The comparison shows also, that in this way results with a reasonable degree of conservatism can be obtained. (authors)

  10. Digital implementation, simulation and tests in MATLAB of the models of Steam line, the turbines, the pressure regulator of a BWR type nucleo electric power plant

    International Nuclear Information System (INIS)

    Lopez R, A.

    2004-01-01

    In this phase of the project they were carried out exhaustive tests to the models of the steam lines, turbines and pressure regulator of a BWR type nucleo electric central for to verify that their tendencies and behaviors are it more real possible. For it, it was necessary to also analyze the transfer functions of the different components along the steam line until the power generator. Such models define alone the dominant poles of the system, what is not limitation to reproduce a wide range of anticipated transitoriness of a power station operation. In the same manner, it was integrated and proved the integrated model form with the models of feeding water of the SUN-RAH, simulating the nuclear reactor starting from predetermined entrances of the prospective values of the vessel. Also it was coupled with the graphic interface developed with the libraries DirectX implementing a specific monitoring panel for this system. (Author)

  11. The applicability of CFD to simulate and study the mixing process and the thermo-hydraulic consequences of a main steam line break in PWR model

    Directory of Open Access Journals (Sweden)

    Farkas Istvan

    2017-01-01

    Full Text Available This paper focuses on the validation and applicability of CFD to simulate and analyze the thermo-hydraulic consequences of a main steam line break. Extensive validation data come from experiments performed using the Rossendorf coolant mixing model facility. For the calculation, the range of 9 to 12 million hexahe¬dral cells was constructed to capture all details in the interrogation domain in the system. The analysis was performed by running a time-dependent calculation, Detailed analyses were made at different cross-sections in the system to evaluate not only the value of the maximum and minimum temperature, but also the loca¬tion and the time at which it occurs during the transient which is considered to be indicator for the quality of mixing in the system. CFD and experimental results were qualitatively compared; mixing in the cold legs with emergency core cooling systems was overestimated. This could be explained by the sensitivity to the bound¬ary conditions. In the downcomer, the experiments displayed higher mixing: by our assumption this related to the dense measurement grid (they were not modelled. The temperature distribution in the core inlet plane agreed with the measurement results. Minor deviations were seen in the quantitative comparisons: the maximum temperature difference was 2ºC.

  12. Development and application of an entrainment model for the PWR U-tube steam generators for main steam line break analysis

    International Nuclear Information System (INIS)

    Song, Dong-Soo; Park, Young-Chan

    2004-01-01

    The purpose of this paper is to present the analyses that were performed to develop and use an entrainment model for pressurized water reactor U-tube steam generators (SG) for main steam line break (MSLB) analyses. The entrainment model was developed using the RETRAN-3D computer program, and the model was benchmarked against experimental data of moisture carryover during a simulated MSLB accident. The application methodology was also developed to incorporate into the MSLB mass and energy release calculations for Kori Unit 1. This methodology utilizes LOFTRAN and RETRAN-3D codes in an iterative sequence of cases in which the LOFTRAN nuclear steam supply system model provides boundary conditions for the RETRAN-3D broken loop steam generator model, and the RETRAN-3D model provides the entrainment data that is input back into the LOFTRAN model. FORTRAN programs were developed to facilitate the sequencing of these iterative calculations. As a result of applying the entrainment model to Kori Unit 1, the temperature calculated inside Containment during MSLB accident using the CONTEMP-LT computer program decreased by about 25degC. Consequently this entrainment model provides a significant benefit by decreasing the temperature envelop for environment qualification as well as decreasing the peak Containment pressure. (author)

  13. Three-dimensional neutron kinetics-thermal-hydraulics VVER 1000 main steam line break analysis by RELAP5-3D code

    International Nuclear Information System (INIS)

    Frisani, A.; Parisi, C.; D'Auria, F.

    2007-01-01

    After the development and the assessment of Three-Dimensional (3D) Neutron Kinetics (NK) - 1D Thermal-Hydraulics (TH) coupled codes analyses methods, deterministic nuclear safety technology is nowadays producing noticeable efforts for the validation of 3D NK - 3D TH coupled codes analyses methods too. Thus, the purpose of this work was to address the capability of the RELAP5-3D 3D NK-3D TH code to reproduce VVER 1000 Nuclear Power Plant (NPP) core dynamic in simulating the mixing effects that could happen in the vessel downcomer and lower plenum during some scenarios. The work was developed in three steps. The first step dealt with the 3D TH modeling of the Kozloduy-6 VVER 1000 reactor pressure vessel. Then this model was validated following a Steam Generator Isolation transient. The second step has been the development of a 3D NK nodalization for the reactor core region. Then the 3D NK model was directly coupled with the previously developed 3D TH model. The third step was the calculation of a Main Steam Line Break (MSLB) transient. The 3D NK global nuclear parameters were then compared with the 0-D results showing a good agreement; nevertheless only the 3D NK- 3D TH model allowed the calculation of each single assembly power trend for this strong NK-TH asymmetric transient. (author)

  14. OECD/NEZ Main Steam Line Break Benchmark Problem Exercise I Simulation Using the SPACE Code with the Point Kinetics Model

    International Nuclear Information System (INIS)

    Kim, Yohan; Kim, Seyun; Ha, Sangjun

    2014-01-01

    The Safety and Performance Analysis Code for Nuclear Power Plants (SPACE) has been developed in recent years by the Korea Nuclear Hydro and Nuclear Power Co. (KHNP) through collaborative works with other Korean nuclear industries. The SPACE is a best-estimated two-phase three-field thermal-hydraulic analysis code to analyze the safety and performance of pressurized water reactors (PWRs). The SPACE code has sufficient features to replace outdated vendor supplied codes and to be used for the safety analysis of operating PWRs and the design of advanced reactors. As a result of the second phase of the development, the 2.14 version of the code was released through the successive various V and V works. The topical reports on the code and related safety analysis methodologies have been prepared for license works. In this study, the OECD/NEA Main Steam Line Break (MSLB) Benchmark Problem Exercise I was simulated as a V and V work. The results were compared with those of the participants in the benchmark project. The OECD/NEA MSLB Benchmark Problem Exercise I was simulated using the SPACE code. The results were compared with those of the participants in the benchmark project. Through the simulation, it was concluded that the SPACE code can effectively simulate PWR MSLB accidents

  15. Three Mile Island Unit 1 Main Steam Line Break Three-Dimensional Neutronics/Thermal-Hydraulics Analysis: Application of Different Coupled Codes

    International Nuclear Information System (INIS)

    D'Auria, Francesco; Moreno, Jose Luis Gago; Galassi, Giorgio Maria; Grgic, Davor; Spadoni, Antonino

    2003-01-01

    A comprehensive analysis of the double ended main steam line break (MSLB) accident assumed to occur in the Babcock and Wilcox Three Mile Island Unit 1 (TMI-1) has been carried out at the Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione of the University of Pisa, Italy, in cooperation with the University of Zagreb, Croatia. The overall activity has been completed within the framework of the participation in the Organization for Economic Cooperation and Development-Committee on the Safety of Nuclear Installations-Nuclear Science Committee pressurized water reactor MSLB benchmark.Thermal-hydraulic system codes (various versions of Relap5), three-dimensional (3-D) neutronics codes (Parcs, Quabbox, and Nestle), and one subchannel code (Cobra) have been adopted for the analysis. Results from the following codes (or code versions) are assumed as reference:1. Relap5/mod3.2.2, beta version, coupled with the 3-D neutron kinetics Parcs code parallel virtual machine (PVM) coupling2. Relap5/mod3.2.2, gamma version, coupled with the 3-D neutron kinetics Quabbox code (direct coupling)3. Relap5/3D code coupled with the 3-D neutron kinetics Nestle code.The influence of PVM and of direct coupling is also discussed.Boundary and initial conditions of the system, including those relevant to the fuel status, have been supplied by Pennsylvania State University in cooperation with GPU Nuclear Corporation (the utility, owner of TMI) and the U.S. Nuclear Regulatory Commission. The comparison among the results obtained by adopting the same thermal-hydraulic nodalization and the coupled code version is discussed in this paper.The capability of the control rods to recover the accident has been demonstrated in all the cases as well as the capability of all the codes to predict the time evolution of the assigned transient. However, one stuck control rod caused some 'recriticality' or 'return to power' whose magnitude is largely affected by boundary and initial conditions

  16. Coupled simulation of steam line break accident

    International Nuclear Information System (INIS)

    Royer, E.; Raimond, E.; Caruge, D.

    2000-01-01

    The steam line break is a PWR type reactor design accident, which concerns coupled physical phenomena. To control these problems simulation are needed to define and validate the operating procedures. The benchmark OECD PWR MSLB (Main Steam Line Break) has been proposed by the OECD to validate the feasibility and the contribution of the multi-dimensional tools in the simulation of the core transients. First the benchmark OECD PWR MSLB is presented. Then the analysis of the three exercises (system with pinpoint kinetic, three-dimensional core and whole system with three-dimensional core) are discussed. (A.L.B.)

  17. A coupled 3-D kinetics/system thermal-hydraulic analysis of main steam line break accident for Optimized Power Reactor 1000

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Yung Kwon; Choi, Chul Jin; Kim, Eun Kee; Lee, Sang Yong [Korea Power Engineering Company, Inc, 150 Deokjin-dong, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of)

    2006-07-01

    This paper presents the results of the coupled 3-D neutronics/thermal-hydraulic analysis of hypothetical main steam line break (MSLB) accident for Optimized Power Reactor 1000. One of the major concerns of this accident is a return-to-power occurrence accompanied with extremely large radial peaking near the stuck Control Element Assembly (CEA). The conventional point kinetics application does not properly account for this kind of asymmetric and local core behavior. Therefore, the current licensing method of point kinetics application introduces some uncertainties and conservatisms in the physics parameters generation, e.g., the static net scram rod worth, moderator cooldown reactivity, Doppler reactivity, and a 3-D peaking factor. The recently developed UNICORN-TM code system is applied for the 3-D coupled calculation, where neutronics code MASTER is coupled with the best-estimate system transient code RETRAN. The 3-D coupled results were assessed in comparison with those by point kinetics application using stand-alone RETRAN application. To quantify the 3-D reactivity benefits over point kinetics, both calculations assumed the accidents to be initiated from the same core state, e.g., end of cycle burnup, fuel and CEA configuration with the same initial moderator and Doppler temperature coefficient, and with initial system thermal-hydraulic condition. The core physics parameters required for point kinetics application were produced using MASTER with the method and procedure consistent with the current licensing application. The occurrence of return-to-power was simulated by intentionally reducing the net CEA worth in order to assess the spatial power distribution and local T-H effect on the dynamic reactivity feedback. The results have demonstrated that the 3-D analysis removes some of the conservatisms inherent in point kinetics analysis mainly caused by the inability to properly account for local reactivity feedback effects during return-to-power transient

  18. A coupled 3-D kinetics/system thermal-hydraulic analysis of main steam line break accident for Optimized Power Reactor 1000

    International Nuclear Information System (INIS)

    Jin, Yung Kwon; Choi, Chul Jin; Kim, Eun Kee; Lee, Sang Yong

    2006-01-01

    This paper presents the results of the coupled 3-D neutronics/thermal-hydraulic analysis of hypothetical main steam line break (MSLB) accident for Optimized Power Reactor 1000. One of the major concerns of this accident is a return-to-power occurrence accompanied with extremely large radial peaking near the stuck Control Element Assembly (CEA). The conventional point kinetics application does not properly account for this kind of asymmetric and local core behavior. Therefore, the current licensing method of point kinetics application introduces some uncertainties and conservatisms in the physics parameters generation, e.g., the static net scram rod worth, moderator cooldown reactivity, Doppler reactivity, and a 3-D peaking factor. The recently developed UNICORN-TM code system is applied for the 3-D coupled calculation, where neutronics code MASTER is coupled with the best-estimate system transient code RETRAN. The 3-D coupled results were assessed in comparison with those by point kinetics application using stand-alone RETRAN application. To quantify the 3-D reactivity benefits over point kinetics, both calculations assumed the accidents to be initiated from the same core state, e.g., end of cycle burnup, fuel and CEA configuration with the same initial moderator and Doppler temperature coefficient, and with initial system thermal-hydraulic condition. The core physics parameters required for point kinetics application were produced using MASTER with the method and procedure consistent with the current licensing application. The occurrence of return-to-power was simulated by intentionally reducing the net CEA worth in order to assess the spatial power distribution and local T-H effect on the dynamic reactivity feedback. The results have demonstrated that the 3-D analysis removes some of the conservatisms inherent in point kinetics analysis mainly caused by the inability to properly account for local reactivity feedback effects during return-to-power transient

  19. Process of characterization of vibration in Cofrentes NPP SRVs - scale model of main steam line; Proceso de caracterizacion de vibraciones en SRVs de C.N. Cofrentes-Modelo a escala linea de vapor principal

    Energy Technology Data Exchange (ETDEWEB)

    Galbally, D.; Hernando, J.; Garcia, G.; Barral, M.

    2014-07-01

    The Cofrentes Nuclear power plant has experienced different events anomalous related to its relief and system (SRVs) main steam safety valves. After various studies is determined that the existence of dynamics of pressure oscillations in the interior of the main steam lines is the cause of many of the events that occurred in the SRVs. To monitor these vibrations, Iberdrola performed the installation of a measuring system of vibration in SRVs and actuators during the recharge 18 (September - October 2011) with a total of 40 accelerometers distributed in 6 of the 16 existing valves. (Author)

  20. Numerical and computational aspects of the coupled three-dimensional core/ plant simulations: organization for economic cooperation and development/ U.S. nuclear regulatory commission pressurized water reactor main-steam-line-break benchmark-II. 3. Analysis of the OECD TMI-1 Main-Steam- Line-Break Benchmark Accident Using the Coupled RELAP5/PANTHER Codes

    International Nuclear Information System (INIS)

    Schneidesch, C.R.; Guisset, J.P.; Zhang, J.; Bryce, P.; Parkes, M.

    2001-01-01

    The RELAP5 best-estimate thermal-hydraulic system code has been coupled with the PANTHER three-dimensional (3-D) neutron kinetics code via the TALINK dynamic data exchange control and processing tool. The coupled RELAP5/PANTHER code package is being qualified and will be used at British Energy (BE) and Tractebel Energy Engineering (TEE), independently, to analyze pressurized water reactor (PWR) transients where strong core-system interactions occur. The Organization for Economic Cooperation and Development/Nuclear Energy Agency PWR Main-Steam-Line-Break (MSLB) Benchmark problem was performed to demonstrate the capability of the coupled code package to simulate such transients, and this paper reports the BE and TEE contributions. In the first exercise, a point-kinetics (PK) calculation is performed using the RELAP5 code. Two solutions have been derived for the PK case. The first corresponds to scenario, 1 where calculations are carried out using the original (BE) rod worth and where no significant return to power (RTP) occurs. The second corresponds to scenario 2 with arbitrarily reduced rod worth in order to obtain RTP (and was not part of the 'official' results). The results, as illustrated in Fig. 1, show that the thermalhydraulic system response and rod worth are essential in determining the core response. The second exercise consists of a 3-D neutron kinetics transient calculation driven by best-estimate time-dependent core inlet conditions on a 18 T and H zones basis derived from TRAC-PF1/MOD2 (PSU), again analyzing two scenarios of different rod worths. Two sets of PANTHER solutions were submitted for exercise 2. The first solution uses a spatial discretization of one node per assembly and 24 core axial layers for both flux and T and H mesh. The second is characterized by spatial refinement (2 x 2 nodes per assembly, 48 core layers for flux, and T and H calculation), time refinement (half-size time steps), and an increased radial discretization for solution

  1. Evaluation of acoustic resonance at branch section in main steam line. Part 2. Proposal of method for predicting resonance frequency in steam flow

    International Nuclear Information System (INIS)

    Uchiyama, Yuta; Morita, Ryo

    2012-01-01

    Flow-induced acoustic resonances of piping system containing closed side-branches are sometimes encountered in power plants. Acoustic standing waves with large amplitude pressure fluctuation in closed side-branches are excited by the unstable shear layer which separates the mean flow in the main piping from the stagnant fluid in the branch. In U.S. NPP, the steam dryer had been damaged by high cycle fatigue due to acoustic-induced vibration under a power uprating condition. Our previous research developed the method for evaluating the acoustic resonance at the branch sections in actual power plants by using CFD. In the method, sound speed in wet steam is evaluated by its theory on the assumption of homogeneous flow, although it may be different from practical sound speed in wet steam. So, it is necessary to consider and introduce the most suitable model of practical sound speed in wet steam. In addition, we tried to develop simplified prediction method of the amplitude and frequency of pressure fluctuation in wet steam flow. Our previous experimental research clarified that resonance amplitude of fluctuating pressure at the top of the branch in wet steam. However, the resonance frequency in steam condition could not be estimated by using theoretical equation as the end correction in steam condition and sound speed in wet steam is not clarified as same reason as CFD. Therefore, in this study, we tried to evaluate the end correction in each dry and wet steam and sound speed of wet steam from experimental results. As a result, method for predicting resonance frequency by using theoretical equation in each wet and dry steam condition was proposed. (author)

  2. Neutronic calculations for Angra-1 steam line break accident

    International Nuclear Information System (INIS)

    Ponzoni Filho, Pedro; Sato, Sadakatu

    2000-01-01

    The reduction of boron concentration in the Boron Injection Tank (BIT), to the room temperature solubility level, makes necessary a reanalysis of the steam line break accident of Angra 1 NPP. This paper describes the neutronic calculation related to this reanalysis. The main steps of the work were: review of reactivity parameters used in the accident simulation; search of xenon profiles that cause the most severe core power distribution; calculation of hot channel factors and other neutronic parameters necessary for DNBR determination. The final conclusion, related to the steam line break accident, states the BIT concentration may be reduced to 2000 ppm. (author)

  3. Simulation of A Main Steam Line Break Accident Using the Coupled 'System Thermal-Hydraulics, 3D reactor Kinetics, and Hot Channel' Analysis Capability of MARS 3.0

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Jae Jun; Chung, Bub Dong

    2005-09-15

    For realistic analysis of thermal-hydraulics (T-H) transients in light water reactors, KAERI has developed the best-estimate T-H system code, MARS. The code has been improved from the consolidated version of the RELAP5/MOD3 and COBRA-TF codes. Then, the MARS code was coupled with a three-dimensional (3-D) reactor kinetics code, MASTER. This coupled calculation feature, in conjunction with the existing hot channel analysis capabilities of the MARS and MASTER codes, allows for more realistic simulations of nuclear system transients. In this work, a main steam line break (MSLB) accident is simulated using the coupled 'system T-H, 3-D reactor kinetics, and hot channel analysis' feature of the MARS code. Two coupled calculations are performed for demonstration. First, a coupled calculation of the 'system T-H and 3-D reactor kinetics' with a refined core T-H nodalization is carried out to obtain global core power and local departure from nucleate boiling (DNB) ratio (DNBR) behaviors. Next, for a more accurate DNBR prediction, another coupled calculation with subchannel meshes for the hot channels is performed. The results of the coupled calculations are very reasonable and consistent so that these can be used to remove the excessive conservatism in the conventional safety analysis.

  4. Simulation of A Main Steam Line Break Accident Using the Coupled 'System Thermal-Hydraulics, 3D reactor Kinetics, and Hot Channel' Analysis Capability of MARS 3.0

    International Nuclear Information System (INIS)

    Jeong, Jae Jun; Chung, Bub Dong

    2005-09-01

    For realistic analysis of thermal-hydraulics (T-H) transients in light water reactors, KAERI has developed the best-estimate T-H system code, MARS. The code has been improved from the consolidated version of the RELAP5/MOD3 and COBRA-TF codes. Then, the MARS code was coupled with a three-dimensional (3-D) reactor kinetics code, MASTER. This coupled calculation feature, in conjunction with the existing hot channel analysis capabilities of the MARS and MASTER codes, allows for more realistic simulations of nuclear system transients. In this work, a main steam line break (MSLB) accident is simulated using the coupled 'system T-H, 3-D reactor kinetics, and hot channel analysis' feature of the MARS code. Two coupled calculations are performed for demonstration. First, a coupled calculation of the 'system T-H and 3-D reactor kinetics' with a refined core T-H nodalization is carried out to obtain global core power and local departure from nucleate boiling (DNB) ratio (DNBR) behaviors. Next, for a more accurate DNBR prediction, another coupled calculation with subchannel meshes for the hot channels is performed. The results of the coupled calculations are very reasonable and consistent so that these can be used to remove the excessive conservatism in the conventional safety analysis

  5. Hydrodynamic and acoustic analysis in 3-D of a section of main steam line to EPU conditions; Analisis hidrodinamico y acustico en 3D de una seccion de linea de vapor principal a condiciones de EPU

    Energy Technology Data Exchange (ETDEWEB)

    Centeno P, J.; Castillo J, V.; Espinosa P, G. [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Area de Ingenieria en Recursos Energeticos, Av. San Rafael Atlixco No. 186, Col. Vicentina, 09340 Mexico D. F. (Mexico); Nunez C, A.; Polo L, M. A., E-mail: baldepeor21@hotmail.com [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose Ma. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

    2013-10-15

    The objective of this word is to study the hydrodynamic and acoustic phenomenon in the main steam lines (MSLs). For this study was considered the specific case of a pipe section of the MSL, where is located the standpipe of the pressure and/or safety relief valve (SRV). In the SRV cavities originates a phenomenon known as whistling that generates a hydrodynamic disturbance of acoustic pressure waves with different tones depending of the reactor operation conditions. In the SRV cavities the propagation velocity of the wave can originate mechanical damage in the structure of the steam dryer and on free parts. The importance of studying this phenomenon resides in the safety of the integrity of the reactor pressure vessel. To dissipate the energy of the pressure wave, acoustic side branches (ASBs) are used on the standpipe of the SRVs. The ASBs are arrangements of compacted lattices similar to a porous medium, where the energy of the whistling phenomenon is dissipate and therefore the acoustic pressure load that impacts in particular to the steam dryers, and in general to the interns of the vessel, diminishes. For the analysis of the whistling phenomenon two three-dimensional (3-D) models were built, one hydrodynamic in stationary state and other acoustic for the harmonic times in transitory regimen, in which were applied techniques of Computational Fluid Dynamics. The study includes the reactor operation analysis under conditions of extended power up rate (EPU) with ASB and without ASB. The obtained results of the gauges simulated in the MSL without ASB and with ASB, show that tones with values of acoustic pressure are presented in frequency ranges between 160 and 200 Hz around 12 MPa and of 7 MPa, respectively. This attenuation of tones implies the decrease of the acoustic loads in the steam dryer and in the interns of the vessel that are designed to support pressures not more to 7.5 MPa approximately. With the above-mentioned is possible to protect the steam dryer

  6. Valve for closing a steam line

    International Nuclear Information System (INIS)

    Meyer, W.; Potrykus, G.

    1976-01-01

    Instead of several control elements, the quick-closing valve, especially in the main-steam line between steam generator and turbine of a power station has the valve cone itself as the only movable part, acting with its inner surface as a piston within a second cylinder space. The valve shaft is at the same time a piston rod with a stepped piston at the upper end. This piston is loaded in a cylinder at the upspace below the valve cover on one hand by a spring, on the other hand by its own medium. Two non-return valves, one of it in a bore of the valve cone, connect the first-mentioned cylinder space with the steam-loaded inlet resp. outlet side of the valve. For controlling the valve, a magnet valve is sufficient. By automatic control of the valve cone coupled with several pistons several control lines can be omitted. There are also no pressurized control lines outside the valve which could be damaged by exterior influences. (ERA) [de

  7. Steam line break analysis in CAREM-25 reactor

    International Nuclear Information System (INIS)

    Zanocco, Pablo; Gimenez, Marcelo O.; Vertullo, Alicia; Schlamp, Miguel A.; Garcia, Alicia E.

    2000-01-01

    The main objective of this report is to analyze the reactor response during a steam line break postulated accident with RELAP5, a plant code using a separated flow model. The steam line break caused a rapid blowdown of the secondary coolant increasing the heat removal in the steam generator. As a consequence and due to reactor features the core power is also increased. As maximum removed power in the secondary side is highly dependant on the total water volume evaporated during the accident a detailed model of feed water and outlet steam pipes is provided. Different cases are analyzed with and without feedwater system and considering the fail or success of the First Shutdown System. In all the sequences the DNBR and CPR remain above the minimum safety values established by design. Further calculations concerning depressurization ramps and steam generator feed water pumps response during depressurization are advised. (author)

  8. Steam Line Break Analysis in CAREM-25 Reactor

    International Nuclear Information System (INIS)

    Zanocco, Pablo; Gimenez, Marcelo; Vertullo, Alicia; Garcia, A; Schlamp, Miguel

    2000-01-01

    The main objective of this report is to analyze the reactor response during a steam line break postulated accident with RELAP5, a plant code using a separated flow model.The steam line break caused a rapid blowdown of the secondary coolant increasing the heat removal in the steam generator.As a consequence and due to reactor features the core power is also increased.As maximum removed power in the secondary side is highly dependant on the total water volume evaporated during the accident a detailed model of feed water and outlet steam pipes is provided.Different cases are analyzed with and without feedwater system and considering the fail or success of the First Shutdown System.In all the sequences the DNBR and CPR remain above the minimum safety values established by design.Further calculations concerning depressurization ramps and steam generator feed water pumps response during depressurization are advised

  9. TRAC-PF1/MOD 1 independent assessment: Semiscale MOD-2A feedwater-line break (S-SF-3) and steam-line break (S-SF-5) tests

    International Nuclear Information System (INIS)

    Dobranich, D.

    1985-11-01

    The TRAC-PF1/MOD1 independent assessment project at Sandia is part of an overall effort funded by the NRC to determine the ability of various systems codes to predict the detailed thermal/hydraulic response of LWRs during accident and off-normal conditions. As part of this effort, calculations for Semiscale Mod-2A test S-SF-3, a feedwater-line break test, and S-SF-5, a steam-line break test, were performed with TRAC-PF1/MOD1. Most aspects of both the S-SF-3 and S-SF-5 steady-state calculations were found to be in good agreement with data. However, the need for a better steam separator model was identified from the S-SF-3 calculation. Overall, the qualitative behavior of both transients was calculated reasonably well; however, there were some discrepancies in the prediction of the quantitative behavior. The results for the S-SF-3 transient calculation indicate that the primary-to-secondary heat transfer degradation began too early. This was possibly due to overprediction of entrainment in the steam generator boiler, leading to an incorrect calculation of the secondary-side fluid distribution during the steady state. However, there was insufficient data to verify this. Results for the S-SF-5 transient calculation indicate that the primary-side fluid temperature response to a steam-line break was in reasonable agreement with data but the pressure response did not coincide with the data. Uncertainties in the data are sufficient to prevent us from determining the exact cause of this discrepancy, but there is indirect evidence that the calculated rate of phase change in the pressurizer was incorrect. 16 refs., 73 figs., 13 tabs

  10. Diagnostics of metal state in steam lines

    International Nuclear Information System (INIS)

    Gofman, Yu.M.; Kazantseva, N.S.; Losev, L.Ya.; Nevolina, G.S.

    1986-01-01

    A series of micropore detection methods is suggested: light microscopy, electron microscopy, hydrostatic weighing; and comparative investigations of pore-formation processes in 12Kh1MF steel steam lines, which have operated for about 100 thousand hours at t=550 deg C and 47-55 MPa stresses are conducted using these methods. It is shown, that electron microscpy method can be applied at the early stages damaging, when embrionic micropores of 0.1 μm in size appear. Optical metallography allows one to detect pores of about 1 μm in size. Damage in density using the hydrostatic weighing method is estimated in the following way: at creep stages 1-2-0.1; at stage 3-0.4-0.6; at predestruction stage the degree of damage equals to 0.7-0.8

  11. Detection of leaks in steam lines by distributed fibre-optic temperature sensing (DTS)

    Energy Technology Data Exchange (ETDEWEB)

    Craik, N G [Maritime Nuclear, Fredericton, N.B. (Canada)

    1997-12-31

    This paper describes an instrumentation system concept which should be capable of early detection of a leak-before-break in main steam lines. Distributed fibre-optic Temperature Sensing (DTS) systems have been used in commercial application for a few years now, but in other industries and applications. DTS uses very long fibre optical cable both as a temperature sensor and as a means of bringing the information back from the sensor to the terminal equipment. The entire length of the fibre is sensitive to temperature and each resolvable section of fibre is equivalent to a point sensor. This commercially available DTS system could be adapted to indicate leaks in steam lines. The fibre-optic cable could either be run either just underneath the aluminium sheathing covering the installation over a steam line, or between the two layers of insulation. This would detect an increase in the temperature of the insulation due to a steam leak. 1 ref., 4 figs.

  12. Detection of leaks in steam lines by distributed fibre-optic temperature sensing (DTS)

    International Nuclear Information System (INIS)

    Craik, N.G.

    1996-01-01

    This paper describes an instrumentation system concept which should be capable of early detection of a leak-before-break in main steam lines. Distributed fibre-optic Temperature Sensing (DTS) systems have been used in commercial application for a few years now, but in other industries and applications. DTS uses very long fibre optical cable both as a temperature sensor and as a means of bringing the information back from the sensor to the terminal equipment. The entire length of the fibre is sensitive to temperature and each resolvable section of fibre is equivalent to a point sensor. This commercially available DTS system could be adapted to indicate leaks in steam lines. The fibre-optic cable could either be run either just underneath the aluminium sheathing covering the installation over a steam line, or between the two layers of insulation. This would detect an increase in the temperature of the insulation due to a steam leak. 1 ref., 4 figs

  13. Digital implementation, simulation and tests in MATLAB of the models of Steam line, the turbines, the pressure regulator of a BWR type nucleo electric power plant; Implementacion digital, simulacion y pruebas en MATLAB de los modelos de la linea de vapor, las turbinas y el regulador de presion de una central Nucleoelectrica tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Lopez R, A [UNAM, Laboratorio de Analisis de Ingenieria de Reactores Nucleares, DEPFI, Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)

    2004-07-01

    In this phase of the project they were carried out exhaustive tests to the models of the steam lines, turbines and pressure regulator of a BWR type nucleo electric central for to verify that their tendencies and behaviors are it more real possible. For it, it was necessary to also analyze the transfer functions of the different components along the steam line until the power generator. Such models define alone the dominant poles of the system, what is not limitation to reproduce a wide range of anticipated transitoriness of a power station operation. In the same manner, it was integrated and proved the integrated model form with the models of feeding water of the SUN-RAH, simulating the nuclear reactor starting from predetermined entrances of the prospective values of the vessel. Also it was coupled with the graphic interface developed with the libraries DirectX implementing a specific monitoring panel for this system. (Author)

  14. Track 3: growth of nuclear technology and research numerical and computational aspects of the coupled three-dimensional core/plant simulations: organization for economic cooperation and development/U.S. nuclear regulatory commission pressurized water reactor main-steam-line-break benchmark-I. 5. Analyses of the OECD MSLB Benchmark with the Codes DYN3D and DYN3D/ATHLET

    International Nuclear Information System (INIS)

    Grundmann, U.; Kliem, S.

    2001-01-01

    The code DYN3D coupled with ATHLET was used for the analysis of the OECD Main-Steam-Line-Break (MSLB) Benchmark, which is based on real plant design and operational data of the TMI-1 pressurized water reactor (PWR). Like the codes RELAP or TRAC,ATHLET is a thermal-hydraulic system code with point or one-dimensional neutron kinetic models. ATHLET, developed by the Gesellschaft for Anlagen- und Reaktorsicherheit, is widely used in Germany for safety analyses of nuclear power plants. DYN3D consists of three-dimensional nodal kinetic models and a thermal-hydraulic part with parallel coolant channels of the reactor core. DYN3D was coupled with ATHLET for analyzing more complex transients with interactions between coolant flow conditions and core behavior. It can be applied to the whole spectrum of operational transients and accidents, from small and intermediate leaks to large breaks of coolant loops or steam lines at PWRs and boiling water reactors. The so-called external coupling is used for the benchmark, where the thermal hydraulics is split into two parts: DYN3D describes the thermal hydraulics of the core, while ATHLET models the coolant system. Three exercises of the benchmark were simulated: Exercise 1: point kinetics plant simulation (ATHLET) Exercise 2: coupled three-dimensional neutronics/core thermal-hydraulics evaluation of the core response for given core thermal-hydraulic boundary conditions (DYN3D) Exercise 3: best-estimate coupled core-plant transient analysis (DYN3D/ATHLET). Considering the best-estimate cases (scenarios 1 of exercises 2 and 3), the reactor does not reach criticality after the reactor trip. Defining more serious tests for the codes, the efficiency of the control rods was decreased (scenarios 2 of exercises 2 and 3) to obtain recriticality during the transient. Besides the standard simulation given by the specification, modifications are introduced for sensitivity studies. The results presented here show (a) the influence of a reduced

  15. Main Test Floor (MTF)

    Data.gov (United States)

    Federal Laboratory Consortium — Purpose: The MTF is employed to validate advanced structural concepts and verify new analytical methodologies. Test articles range in size from subcomponent to full...

  16. Engineering task plan for steam line ramp calculations

    International Nuclear Information System (INIS)

    DeSantis, G.N.; Freeman, R.D.

    1994-01-01

    The purpose of this document is to provide an approved work plan to perform calculations that verify the load limits of a proposed ramp over a steam line at the back side (East side) of SY Farm in support of work package 2W-94-00812/K. The objective of this supporting document is to provide Operations with a set of checked calculations that verify the ramp over the steam line at SY Farm will support a fully loaded concrete mixer truck without affecting the steam line. The calculations will be performed by an engineers from Facility Systems and independently checked and reviewed by another engineer. The calculations may then be added to the work package. If Operations decides to make any configuration changes to the steam line or surrounding area, Operations shall have these changes documented by an Engineering Change Notice (ECN). This ECN can be done by Facility Systems or any other engineering organization at the direction of Operations

  17. Loss of feed flow, steam generator tube rupture and steam line break thermohydraulic experiments

    Energy Technology Data Exchange (ETDEWEB)

    Mendler, O J; Takeuchi, K; Young, M Y

    1986-10-01

    The Westinghouse Model Boiler No. 2 (MB-2) steam generator test model at the Engineering Test Facility in Tampa, Florida, was reinstrumented and modified for performing a series of tests simulating steam generator accident transients. The transients simulated were: loss of feed flow, steam generator tube rupture, and steam line break events. This document presents a description of (1) the model boiler and the associated test facility, (2) the tests performed, and (3) the analyses of the test results.

  18. Loss of feed flow, steam generator tube rupture and steam line break thermohydraulic experiments

    International Nuclear Information System (INIS)

    Mendler, O.J.; Takeuchi, K.; Young, M.Y.

    1986-10-01

    The Westinghouse Model Boiler No. 2 (MB-2) steam generator test model at the Engineering Test Facility in Tampa, Florida, was reinstrumented and modified for performing a series of tests simulating steam generator accident transients. The transients simulated were: loss of feed flow, steam generator tube rupture, and steam line break events. This document presents a description of (1) the model boiler and the associated test facility, (2) the tests performed, and (3) the analyses of the test results

  19. Numerical and computational aspects of the coupled three-dimensional core/ plant simulations: organization for economic cooperation and development/ U.S. nuclear regulatory commission pressurized water reactor main-steam-line-break benchmark-II. 2. TRAB-3D/SMABRE Calculation of the OECD/ NRC PWR MSLB Benchmark

    International Nuclear Information System (INIS)

    Daavittila, A.; Haemaelaeinen, A.; Kyrki-Rajamaki, R.

    2001-01-01

    All three exercises of the OECD/NRC Pressurized Water Reactor (PWR) Main-Steam-Line-Break (MSLB) Benchmark were calculated at VTT Energy. The SMABRE thermal-hydraulics code was used for the first exercise, the plant simulation with point-kinetics neutronics. The second exercise was calculated with the TRAB-3D three-dimensional reactor dynamics code. The third exercise was calculated with the combination TRAB-3D/SMABRE. Both codes have been developed at VTT Energy. The results of all the exercises agree reasonably well with those of the other participants; thus, instead of reporting the results, this paper concentrates on describing the computational aspects of the calculation with the foregoing codes and on some observations of the sensitivity of the results. In the TRAB-3D neutron kinetics, the two-group diffusion equations are solved in homogenized fuel assembly geometry with an efficient two-level nodal method. The point of the two-level iteration scheme is that only one unknown variable per node, the average neutron flux, is calculated during the inner iteration. The nodal flux shapes and cross sections are recalculated only once in the outer iteration loop. The TRAB-3D core model includes also parallel one-dimensional channel hydraulics with detailed fuel models. Advanced implicit time discretization methods are used in all submodels. SMABRE is a fast-running five-equation model completed by a drift-flux model, with a time discretization based on a non-iterative semi-implicit algorithm. For the third exercise of the benchmark, the TMI-1 models of TRAB-3D and SMABRE were coupled. This was the first time these codes were coupled together. However, similar coupling of the HEXTRAN and SMABRE codes has been shown to be stable and efficient, when used in safety analyses of Finnish and foreign VVER-type reactors. The coupling used between the two codes is called a parallel coupling. SMABRE solves the thermal hydraulics both in the cooling circuit and in the core

  20. Main Propulsion Test Article (MPTA)

    Science.gov (United States)

    Snoddy, Cynthia

    2010-01-01

    Scope: The Main Propulsion Test Article integrated the main propulsion subsystem with the clustered Space Shuttle Main Engines, the External Tank and associated GSE. The test program consisted of cryogenic tanking tests and short- and long duration static firings including gimbaling and throttling. The test program was conducted on the S1-C test stand (Position B-2) at the National Space Technology Laboratories (NSTL)/Stennis Space Center. 3 tanking tests and 20 hot fire tests conducted between December 21 1 1977 and December 17, 1980 Configuration: The main propulsion test article consisted of the three space shuttle main engines, flightweight external tank, flightweight aft fuselage, interface section and a boilerplate mid/fwd fuselage truss structure.

  1. Rupture of steam lines between blocks D and G

    International Nuclear Information System (INIS)

    1999-01-01

    Analysis of steam lines rupture between blocks D and G of Ignalina NPP was performed. Model for evaluation of thermo hydrodynamic parameters was developed. Structural analysis of the shaft building was done as well. State of the art codes such as RELAP5, ALGOR, NEPTUNE were used in these calculations

  2. Condensation of the steam in the horizontal steam line during cold water flooding

    International Nuclear Information System (INIS)

    Strubelj, L.; Tiselj, I.

    2006-01-01

    Direct contact condensation and condensation induced water-hammer in a horizontal pipe was experimentally investigated at PMK-2 test facility of the Hungarian Atomic Energy Research Institute KFKI. The experiment is preformed in the horizontal section of the steam line of the PMK-2 integral test facility. As liquid water floods the horizontal part of the pipeline, the counter current horizontally stratified flow is being observed. During the flooding of the steam line, the vapour-liquid interface area increases and therefore the vapour condensation rate and the vapour velocity also increase. Similar phenomena can occur in the cold/hot leg of the primary loop of PWR nuclear power plant during loss of coolant accident, when emergency core cooling system is activated. Water level at one cross-section and four local void fraction and temperature at the top of steam line was measured and compared with simulation. Condensed steam increases the water temperature that is why the local temperature measurements are the most important information, from which condensation rate can be estimated, since mass of condensed steam was not measured. Free surface simulation of the experiment with thermal phase change model is presented. Surface renewal concept with small eddies is used for calculation of heat transfer coefficient. With surface renewal theory we did not get results similar to experiment, that is why heat transfer coefficient was increased by factor 20. In simulation with heat transfer coefficient calculated with surface renewal concept bubble entrapment is due to reflection of the wave from the end of the pipe. When heat transfer coefficient is increased, condensation rate and steam velocity are also increased, bubble entrapment is due to Kelvin-Helmholtz instability of the free surface, and the results become similar to the measurements. (author)

  3. Numerical and computational aspects of the coupled three-dimensional core/ plant simulations: organization for economic cooperation and development/ U.S. nuclear regulatory commission pressurized water reactor main-steam-line-break benchmark-II. 5. TMI-1 Benchmark Performed by Different Coupled Three-Dimensional Neutronics Thermal- Hydraulic Codes

    International Nuclear Information System (INIS)

    D'Auria, F.; Galassi, G.M.; Spadoni, A.; Gago, J.L.; Grgic, D.

    2001-01-01

    A comprehensive analysis of a double-ended main-steam-line-break (MSLB) accident assumed to have occurred in the Babcock and Wilcox Three Mile Island (TMI) Unit 1 nuclear power plant (NPP) has been carried out at the Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione of the University of Pisa, Italy. The research has been carried out in cooperation with the University of Zagreb, Croatia, and with partial financial support from the European Union through a grant to one of the authors. The overall activity has been completed within the framework of the participation in the Organization for Economic Cooperation and Development Committee on the Safety of Nuclear Installations-Nuclear Science Committee PWR MSLB Benchmark. Different code versions have been adopted in the analysis. Results from the following codes (or code versions) are described in this paper: 1. RELAP5/mod 3.2.2, gamma version, coupled with the three-dimensional (3-D) neutron kinetics PARCS code; 2. RELAP5/mod 3.2.2, gamma version, coupled with the 3-D neutron kinetics QUABBOX code; 3. RELAP5/3D code coupled with the 3-D neutron kinetics NESTLE code. Boundary and initial conditions of the system, including those relevant to the fuel status, have been supplied by The Pennsylvania State University in cooperation with GPU Nuclear (the utility, owner of TMI) and the U.S. Nuclear Regulatory Commission (NRC). The main challenge for the calculation was the prediction of the return to power (RTP) following the inlet of cold water into the core and one 'stuck-withdrawn' control rod. Non-realistic assumptions were proposed to augment the core power peak following scram. Zero-dimensional neutronics codes were capable of detecting the RTP after scram. However, the application of 3-D neutronics codes to the same scenario allowed the calculation of a similar value for overall core power peak but showed power increase occurrence in about one-tenth of the core volume. The results achieved in phase 1 of

  4. Preliminary Analysis of a Steam Line Break Accident with the MARS-KS code for the SMART Design with Passive Safety Systems

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Doohyuk; Ko, Yungjoo; Suh, Jaeseung [Hannam Univ., Daejeon (Korea, Republic of); Bae, Hwang; Ryu, Sunguk; Yi, Sungjae; Park, Hyunsik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    SMART has been developed by KAERI, and SMART-Standard Design Approval (SDA) was recently granted in 2012. A SMART design with Passive Safety System (PSS) features (called SMART-PSS) is being developed and added to the standard design of SMART by KAERI to improve its safety system. Active safety systems such as safety injection pumps will be replaced by a passive safety system, which is actuated only by the gravity force caused by the height difference. All tanks for the passive safety systems are higher than the injection nozzle, which is located around the reactor coolant pumps (RCPs). In this study, a preliminary analysis of the main steam line break accident (MSLB) was performed using the MARS-KS code to understand the general behavior of the SMART-PSS design and to prepare its validation test with the SMART-ITL (FESTA) facility. An anticipated accident for the main steam line break (MSLB) was performed using the MARS-KS code to understand the thermal-hydraulic behaviors of the SMART-PSS design. The preliminary analysis provides good insight into the passive safety system design features of the SMART-PSS and the thermal-hydraulic characteristics of the SMART design. The analysis results of the MSLB showed that the core water collapsed level inside the core support barrel was maintained high over the active core top level during the transient period. Therefore, the SMART-PSS design has satisfied the requirements to maintain the plant at a safe shutdown condition during 72 hours without AC power or operator action after an anticipated accident.

  5. Analyses in support of installation of steam-dump-to-atmosphere valves at steam lines of the Dukovany NPP

    International Nuclear Information System (INIS)

    Kral, P.

    1998-01-01

    Four conservative analyses were carried out with a view to examining the cooldown capacity of the super-emergency feedwater pump (SEFWP) → steam generator (SG) → steam dump to atmosphere/main steam line (SDA/MSL) chain. This emergency cooldown capacity was investigated for a postulated accident associated with a main steam header break + main feedwater header break + closing of all main steam lines, and for an artificial accident with SCRAM + isolation of all MSLs + loss of feedwater. The RELAP5/MOD3.1 code and a detailed 3-loop input model of the Dukovany plant were employed. Conservative assumptions with respect to the initial reactor power, decay heat evolution, and other input parameters were applied. The results gave evidence that the capacity of both the 2SEFWP → 2SG → 2SDA/SG and 1SEFWP → 1SG → 1SDA/SG chains is sufficient for the decay heat to be removed from the reactor; however, a considerably long time allowing for a sufficient drop of the decay heat is necessary for a deep cooldown of the primary circuit. For the event encompassing main steam header break + main feedwater header break with isolation of all MSLs and with cooling by 2SEFWPs, a time-consuming calculation gave evidence of the feasibility of passing to the water-water regime and primary system cooldown to below 93 deg C in the hot legs

  6. Long-term rupture strength as a criterion of operational durability of steam line metal

    International Nuclear Information System (INIS)

    Gofman, Yu.M.

    2000-01-01

    The method for substantiation of the steam line service life prolongation, depending on the achieved level of the metal vulnerability to damage, is proposed. The methodology for evaluating the metal state is developed on the basis of the durability bond with the level of the vulnerability to damage through micropores and the ferrite dislocation structure state. The main changes in the metal at the 1-3 stages of its creep are presented. The micropores are absent at the 1 stage. the micropores of about 0.1 μm in diameter are identified at the beginning of the 2 stage. The ferrite grains on the transition from the 2 to the 3 creep stage are mainly fragmentary. There takes place further micropores growth on the grain boundaries up to 1 - 3 μm. Significant number of recrystallized volumes in the ferrite is observed at the 3 creep stage. The number of micropores of 1 - 3μm in size sharply increases, and, as a rule, chains of micropores are observed. The pores of 5 μm in size are formed at the pre-destruction stage, the fusion whereof leads to microcracks formation [ru

  7. Evaluation on operation of liquid relief valves for steam line break accidents by RELAP5/CANDU+ code

    International Nuclear Information System (INIS)

    Yang, C. Y.; Bang, Y. S.; Kim, H. J.

    2001-01-01

    A development of RELAP5/CANDU+ code for regulatory audits of accident analysis of CANDU nuclear power plants is on progress. This paper is undertaken in a procedure of a verification and validation for RELAP5/CANDU+ code by analyzing main steam line break accidents of WS 2/3/4. Following the ECC injection in sequence of the steam line breaks, the mismatch in heat transfer between the primary and the secondary systems makes pressure of the primary system instantly peaked to the open setpoint of liquid relief valves. The event sequence follows the result of WS 2/3/4 FSAR, but there is a difference in pressure transient after ECC injection. Sensitivity analysis for main factors dependent on the peak pressure such as control logics of liquid relief valves. ECC flow path and feedwater flow is performed. Because the pressure increase is continued for a long time and its peaking is high, open and close of the liquid relief valves are repeated several times, which is obviously different from those of WS 2/3/4 FSAR. As a result, it is evaluated that conservative modeling for the above variables is required in the analysis

  8. Operational reliability of high pressure steam lines of pearlitic steels after 150-200 thousand h service

    International Nuclear Information System (INIS)

    Veksler, E.Ya.; Chajkovskij, V.M.; Osasyuk, V.V.

    1980-01-01

    Usage of both calculational and physical methods is recommended to estimate a service operating life of long-term working steam line materials. Application of these methods is demonstrated when studying steam line bends made of 12MKh and 12Kh1MF pearlitic steels. Good coincidence of results for the determination of residual durability of steam lines is obtained using these two methods [ru

  9. Technical evaluation: 300 Area steam line valve accident

    International Nuclear Information System (INIS)

    1993-08-01

    On June 7, 1993, a journeyman power operator (JPO) was severely burned and later died as a result of the failure of a 6-in. valve that occurred when he attempted to open main steam supply (MSS) valve MSS-25 in the U-3 valve pit. The pit is located northwest of Building 331 in the 300 Area of the Hanford Site. Figure 1-1 shows a layout of the 300 Area steam piping system including the U-3 steam valve pit. Figure 1-2 shows a cutaway view of the approximately 10- by 13- by 16-ft-high valve pit with its various steam valves and connecting piping. Valve MSS-25, an 8-in. valve, is located at the bottom of the pit. The failed 6-in. valve was located at the top of the pit where it branched from the upper portion of the 8-in. line at the 8- by 8- by 6-in. tee and was then ''blanked off'' with a blind flange. The purpose of this technical evaluation was to determine the cause of the accident that led to the failure of the 6-in. valve. The probable cause for the 6-in. valve failure was determined by visual, nondestructive, and destructive examination of the failed valve and by metallurgical analysis of the fractured region of the valve. The cause of the accident was ultimately identified by correlating the observed failure mode to the most probable physical phenomenon. Thermal-hydraulic analyses, component stress analyses, and tests were performed to verify that the probable physical phenomenon could be reasonably expected to produce the failure in the valve that was observed

  10. Response of the steam generator VVER 1000 to a steam line break

    International Nuclear Information System (INIS)

    Novotny, J.; Novotny, J. Jr.

    2003-01-01

    Dynamic effects of a steam line break in the weld of the steam pipe and the steam collector on the steam generator system are analyzed. Modelling of a steam line break may concern two cases. The steam line without a restraint and the steam line protected by a whip restraint with viscous elements applied at the postulated break cross-section. The second case is considered. Programme SYSTUS offers a special element the stiffness and viscous damping coefficients of which may be defined as dependent on the relative displacement and velocity of its nodes respectively. A circumferential crack is simulated by a sudden decrease of longitudinal and lateral stiffness coefficients of these special SYSTUS elements to zero. The computation has shown that one can simulate the pipe to behave like completely broken during a time interval of 0,0001 s or less. These elements are used to model the whip restraint with viscous elements and viscous dampers of the GERB type as well. In the case of a whip restraint model the stiffness coefficient-displacement relation and damping coefficient - velocity relation are chosen to fit the given characteristics of the restraint. The special SYSTUS elements are used to constitute Maxwell elements modelling the elasto-plastic and viscous properties of the GERB dampers applied to the steam generator. It has been ascertained that a steam line break at the postulated weld crack between the steam pipe and the steam generator collector cannot endanger the integrity of the system even in a case of the absence of a whip restraint effect. (author)

  11. Three-Dimensional Modeling of a Steam-Line Break in a Boiling Water Reactor

    International Nuclear Information System (INIS)

    Tinoco, Hernan

    2002-01-01

    Because of weld problems, the core grids of Units 1 and 2 at the Forsmark nuclear power plant have been replaced by grids of a new design, consisting of a single machined piece without welds. The qualifying structural analysis has been carried out considering dynamic loads, which implies that even loss-of-coolant accidents have to be included. Therefore, a detailed time description of the loads acting on the different internal parts of the reactor is needed. To achieve sufficient space and time resolution, a computational fluid dynamics (CFD) analysis was considered to be a viable alternative.A CFD analysis of a steam-line break in the boiling water reactor of Unit 2 is the subject of this work. The study is based on the assumption that the timescale of the transient analysis is smaller than the relaxation time of the water-steam system. Therefore, a simulation of only the upper, steam part of the reactor with no two-phase effects (flashing) is feasible.The results obtained display a rather complex behavior of the decompression process, forcing the analysis of the pressure field to be accomplished through animation. In contrast, the computed instantaneous forces over different internal parts oscillate regularly and are approximately twice the forces estimated in the past by simpler methods, with frequencies of 30 to 40 Hz; top amplitudes of ∼1.64 MN; and relatively low damping, ∼25% after 0.5 s.According to the present results, this type of modeling is physically meaningful for simulation timescales smaller than the water-steam relaxation time, i.e., ∼0.5 s at reactor conditions. At larger times, a two-phase model is necessary to describe the decompression process since two-phase effects are dominant. The results have not yet been validated with experiments, but validation computations will be run in the future for comparison with results of the Marviken tests

  12. Space Shuttle Main Engine Public Test Firing

    Science.gov (United States)

    2000-01-01

    A new NASA Space Shuttle Main Engine (SSME) roars to the approval of more than 2,000 people who came to John C. Stennis Space Center in Hancock County, Miss., on July 25 for a flight-certification test of the SSME Block II configuration. The engine, a new and significantly upgraded shuttle engine, was delivered to NASA's Kennedy Space Center in Florida for use on future shuttle missions. Spectators were able to experience the 'shake, rattle and roar' of the engine, which ran for 520 seconds - the length of time it takes a shuttle to reach orbit.

  13. Modal analysis of main steam line piping under high energy line break condition

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Tae-Jin; Kim, Seung Hyun; Je, Sang-Yun; Chang, Yoon-Suk [Kyung Hee University, Yongin (Korea, Republic of)

    2016-10-15

    If HELB (High Energy Line Break) occurs in NPPs (Nuclear Power Plants), not only environmental effect like release of radioactive material but also secondary structural defects should be considered. Jet impingement phenomenon caused by sudden pipe rupture may lead to severe damage on neighboring safe-related components and other structure. Lots of studies have been conducted to assess dynamic behaviors of the SG and MSL piping while pipe whip restraints and jet impingement shields are taken into account during design stage. Arroyo et al. performed modal analyses of a simple square component to examine the jet impingement phenomenon. Also, structural characteristics were predicted to assure structural integrity against the HELB. In this study, we examined dynamic characteristics of SG and MSL piping in a typical 1000MWe NPP. Simulation was performed by using two commercial computational softwares. In particular, modal analyses were conducted to determine mode shapes and natural frequencies of the structure and maximum displacements. The data obtain from each software were compared and observation was discussed in relation to the jet impingement phenomenon. In this research, modal analyses on the SG and MSL piping were carried out to get natural frequencies, vibration mode shapes and maximum displacements. Thereby, the following key finding was observed. (1) Maximum displacement was calculated at the top of SG outlet nozzle with y-directional bending at the third mode. (2) The differences between two models were respectively 7% in natural frequencies and less than 1% in maximum displacements.

  14. Main coolant pump testing at Ontario Hydro

    International Nuclear Information System (INIS)

    Hartlen, R.

    1991-01-01

    This article describes Ontario Hydro Research Division's experience with a computerized data acquisition and analysis system for monitoring mechanical vibration in reactor coolant pumps. The topics covered include bench-marking of the computer system and the coolant pumps, signatures of normal and malfunctioning pumps, analysis of data collected by the monitoring system, simulation of faults, and concerns that have been expressed about data interpretation, sensor types and locations, alarm/shutdown limits and confirmation of nondestructive examination testing. This presentation consists of overheads only

  15. The effect of steam separataor efficiency on transient following a steam line break

    International Nuclear Information System (INIS)

    Choi, J.H.; Ohn, M.Y.; Lee, N.H.; Hwang, S.T.; Lee, S.K.

    1996-01-01

    Detailed thermalhydraulic simulations for CANDU 6 steam line break inside containment are performed to predict the response of the primary and secondary circuits. The analysis is performed using the thermalhydraulic computer code, CATHENA, with a coupled primary and secondary circuit model. A two-loop representation of the primary and secondary circuits is modelled. The secondary circuit model includes the feedwater line from the deaerator storage tank, multi-node steam generators and the steam line up to the turbine. Two cases were carried out using different assumptions for the efficiency of the steam separators. Case 1 assumes the efficiency of the steam separators becomes zero when the water level in the steam drum increases to the elevation of primary cyclones, or the outlet flow from the steam generator becomes higher than 150 % of normal flow. Case 2 assumes the efficiency becomes zero only when the water level in the steam drum reaches the elevation of primary cyclones. The simulation results show that system responses are sensitive to the assumption for the efficiency of the steam separators and case 1 gives higher discharge energy. Fuel cooling is assured, since primary circuit is cooled down sufficiently by the steam generators for both cases. (author)

  16. Level-Swell Prediction With RETRAN-3D And Its Application To A BWR Steam-Line-Break Analysis

    International Nuclear Information System (INIS)

    Aounallah, Y.; Hofer, K.

    2003-01-01

    Level-swell experiments have often been simulated using system codes, such as TRAC and RELAP, but only cursory assessments have been performed with the operational-transient code RETRAN-3D, the main system code used within the STARS project. The present study, initiated in the framework of a BWR Steam-Line-Break (SLB) accident scenario, addresses this lacuna by performing RETRAN simulations of the General Electric Level-Swell experiments, and by investigating their implications on power plant accident analyses. Parameters to which the predicted level swell is sensitive have been identified, and recommendations on code options are made. The SLB analysis objective was to determine the amount of steam and liquid discharged through the break under specified boundary conditions, and to gauge the results against reference values. The impact of the nodalization of the upper part of the reactor pressure vessel was investigated and found to play an important role, whereas the level swell induced from flashing was found not to be the predominant factor for these simulations. (author)

  17. Analysis methodology for the post-trip return to power steam line break event

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chul Shin; Kim, Chul Woo; You, Hyung Keun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-06-01

    An analysis for Steam Line Break (SLB) events which result in a Return-to-Power (RTP) condition after reactor trip was performed for a postulated Yonggwang Nuclear Power Plant Unit 3 cycle 8. Analysis methodology for post-trip RTP SLB is quite different from that of non-RTP SLB and is more difficult. Therefore, it is necessary to develop a methodology to analyze the response of the NSSS parameters to the post-trip RTP SLB events and the fuel performance after the total reactivity exceeds the criticality. In this analysis, the cases with and without offsite power were simulated crediting 3-D reactivity feedback effect due to a local heatup in the vicinity of stuck CEA and compared with the cases without 3-D reactivity feedback with respect to post-trip fuel performance. Departure-to Nucleate Boiling Ratio (DNBR) and Linear Heat Generation Rate (LHGR). 36 tabs., 32 figs., 11 refs. (Author) .new.

  18. SPES-2, the full-height, full-pressure, test facility simulating the AP600 plant: Main results from the experimental campaign

    International Nuclear Information System (INIS)

    Medich, C.; Rigamonti, M.; Martinelli, R.; Tarantini, M.; Conway, L.

    1995-01-01

    The SPES-2 is a full height, full pressure experimental test facility reproducing the Westinghouse AP600 reactor with a scaling factor of 1/395. The experimental plant, designed and operated by SIET in Piacenza, consists of a full simulation of the AP600 primary core cooling system including all the passive and active safety systems. In 1992, Westinghouse, in cooperation with ENEL, ENEA, SIET and ANSALDO developed an experimental program to test the integrated behavior of the AP600 passive safety systems. The SPES-2 test matrix, concluded in November 1994, has examined the AP600 passive safety system response for a range of small break LOCAs at different locations on the primary system and on the passive system lines; single steam generator tube ruptures with both passive and active non-safety systems, and a main steam line break transient to demonstrate the capability of passive safety systems for rapid cooldown. Each of the tests has provided detailed experimental results for verification of the capability of the analysis methods to predict the integrated passive safety system behavior

  19. Evaluation of the PTS potential in a WWER-1000 following a steam line break

    International Nuclear Information System (INIS)

    Beghini, M.; D'Auria, F.; Galassi, G.M.; Vitale, E.

    1997-01-01

    A qualified nodalization for WWER-1000 is available at DCMN (Dipartimento di Costruzioni Meccaniche e Nucleari) of University of Pisa that is suitable for running with the thermohydraulic system code Relap5/mod3.2. The nodalization consists of about 1400 hydraulic nodes and more than 5000 mesh points for conduction heat transfer. The four loops of the NPP are separately modelled. Detailed information about the plant hardware has been gotten from contacts with Eastern Organizations in Bulgaria, Russia and Ukraine. The qualification of the nodalization has been achieved at a steady state level utilizing a procedure available at DCMN and at a transient level on the basis of operational (planned) transients performed in the Bulgarian Kozloduy-5 NPP and of the unplanned transient occurred at the Ukrainian Zaporosche NPP (April 1995). Data measured in steam generators have also been utilized. The nodalization has been widely applied to the analysis of accident scenarios in WWER-1000, including Large Break LOCA, Small Break LOCA, ATWS, Loss of Feedwater and Station Blackout. The present activity aims at evaluating the potential for PTS (Pressurized Thermal Shock) following a steam line break accident. The thermalhydraulic results were employed as input for a parametric Fracture Mechanics analysis based on conservative hypothesis of the shape and localization of a pre-existing defect. Stress analysis evidenced the effect of partial cooling of the vessel and gave some general indications of the risk for unstable crack propagation under the simulated PTS conditions. (author). 30 refs, 17 figs, 4 tabs

  20. Application of the coupled Relap5/Panther codes for PWR steam. Line break accident analysis

    International Nuclear Information System (INIS)

    Guisset, J.-P.; Bosso, S.; Charlier, A.; Delhaye, X.; Ergo, O.; Ouliddren, K.; Schneidesch, C.; Zhang, J.

    2001-01-01

    A dynamic coupling between the existing 1-dimensional thermal-hydraulics system code RELAP5 and the 3-dimensional neutronics code PANTHER is applied via the transient analysis code linkage program TALINK. An interface between PANTHER and the subchannel thermal-hydraulic analysis code COBRA 3C allows direct evaluation of the Departure from Nucleate Boiling Ratio in parallel with the coupled PANTHER/RELAP5 simulation. The coupled codes are applied to develop a Final Safety Analysis Report (FSAR) accident analysis methodology for the major Steam Line Break (SLB) accident at hot zero power in a typical three-loop pressurised water reactor. In this methodology, the uncertainties related to the plant, core thermal-hydraulic and neutronic parameters are combined in a deterministic bounding approach based on sensitivity studies. The results of coupled thermal-hydraulic and neutronic analysis of SLB are presented and discussed. It is shown that there exists an important margin in the traditional FSAR accident analysis for SLB, which can be attributed by the conservatism's introduced by de-coupling the plant sub-systems. (author)

  1. Application of the coupled Relap5/Panther codes for PWR steam. Line break accident analysis

    Energy Technology Data Exchange (ETDEWEB)

    Guisset, J.-P.; Bosso, S.; Charlier, A.; Delhaye, X.; Ergo, O.; Ouliddren, K.; Schneidesch, C.; Zhang, J. [Tractebel Energy Engineering, Brussels (Belgium)

    2001-07-01

    A dynamic coupling between the existing 1-dimensional thermal-hydraulics system code RELAP5 and the 3-dimensional neutronics code PANTHER is applied via the transient analysis code linkage program TALINK. An interface between PANTHER and the subchannel thermal-hydraulic analysis code COBRA 3C allows direct evaluation of the Departure from Nucleate Boiling Ratio in parallel with the coupled PANTHER/RELAP5 simulation. The coupled codes are applied to develop a Final Safety Analysis Report (FSAR) accident analysis methodology for the major Steam Line Break (SLB) accident at hot zero power in a typical three-loop pressurised water reactor. In this methodology, the uncertainties related to the plant, core thermal-hydraulic and neutronic parameters are combined in a deterministic bounding approach based on sensitivity studies. The results of coupled thermal-hydraulic and neutronic analysis of SLB are presented and discussed. It is shown that there exists an important margin in the traditional FSAR accident analysis for SLB, which can be attributed by the conservatism's introduced by de-coupling the plant sub-systems. (author)

  2. Analysis of steam line break of SMART using RETRAN-3D/INT

    International Nuclear Information System (INIS)

    Kim, Tae-Wan; Kim, Jong-Won; Park, Goon-Cherl

    2003-01-01

    RETRAN-3D has been modified to be suitable to safety analysis for integral type marine reactor with modular helical-coiled steam generator cassettes. The modified RETRAN-3D, RETRAN-3D/INT, has helical coil heat conductor model and heat transfer coefficient models for tube and shell sides of helical-coiled steam generator. In addition, moving models are added to simulate the effect of ship motions such as inclination, heaving, rolling and so on. RETRAN-3D/INT has been verified with natural circulation experiment conducted in Seoul National University and the analysis results for the first Japanese nuclear ship, MUTSU. In this study, the safety analysis for SMART, which has been developed by Korea Atomic Energy Research Institute, is performed to examine the applicability of RETRAN-3D/INT to the safety analysis of SMART. The steam line break is selected as reference case. The break type is assumed to the guillotine break. The loss of offsite power is considered as a coincident event and the failure of single train of passive residual heat removal system is assumed as single failure. From the results, it is found that RETRAN-3D/INT can appropriately simulate the transient of SMART and the improvement of non-condensable gas model is required. (author)

  3. A standing pressure wave hypothesis of oscillating forces generated during a steam line break

    International Nuclear Information System (INIS)

    Tinoco, H.

    2001-01-01

    A rapid glance at the figure depicting the net forces acting on the reactor vessel and internals, as obtained through a CFD simulation of a BWR steam line break, reveals an amazing oscillating regularity of these forces which is in glaring contrast to the chaotic behaviour of the steam pressure field in the steam annulus. Assuming that the decompression process excites and maintains standing pressure waves in the annular cylindrical region constituted by the steam annulus, it is possible to reconstruct the net forces acting on the reactor vessel and internals through the contribution of almost only the first dispersive mode. If a Neumann boundary condition is assumed at the section connecting the steam annulus to the steam dome, the frequency predicted is approximately % 5.9 higher than that of the CFD simulations. However, this connecting section allows wave transmission, and a more appropriate boundary condition should be one of the Robin type. Therefore, this section is modelled as an absorbing wall, and the corresponding normal impedance is calculated using the CFD simulations. Week non-linear effects can also be observed in the calculated forces through the presence of the first subharmonic. By the methodology described above, an estimate of the forces acting on the reactor vessel and internals of unit 3 of Forsmark Nuclear Power Plant has been obtained. (author)

  4. Evaluation of the PTS potential in a WWER-1000 following a steam line break

    Energy Technology Data Exchange (ETDEWEB)

    Beghini, M; D` Auria, F; Galassi, G M; Vitale, E [Universita degli Studi di Pisa, Dipt. di Costruzioni Meccaniche e Nucleari, Pisa (Italy)

    1997-09-01

    A qualified nodalization for WWER-1000 is available at DCMN (Dipartimento di Costruzioni Meccaniche e Nucleari) of University of Pisa that is suitable for running with the thermohydraulic system code Relap5/mod3.2. The nodalization consists of about 1400 hydraulic nodes and more than 5000 mesh points for conduction heat transfer. The four loops of the NPP are separately modelled. Detailed information about the plant hardware has been gotten from contacts with Eastern Organizations in Bulgaria, Russia and Ukraine. The qualification of the nodalization has been achieved at a steady state level utilizing a procedure available at DCMN and at a transient level on the basis of operational (planned) transients performed in the Bulgarian Kozloduy-5 NPP and of the unplanned transient occurred at the Ukrainian Zaporosche NPP (April 1995). Data measured in steam generators have also been utilized. The nodalization has been widely applied to the analysis of accident scenarios in WWER-1000, including Large Break LOCA, Small Break LOCA, ATWS, Loss of Feedwater and Station Blackout. The present activity aims at evaluating the potential for PTS (Pressurized Thermal Shock) following a steam line break accident. The thermalhydraulic results were employed as input for a parametric Fracture Mechanics analysis based on conservative hypothesis of the shape and localization of a pre-existing defect. Stress analysis evidenced the effect of partial cooling of the vessel and gave some general indications of the risk for unstable crack propagation under the simulated PTS conditions. (author). 30 refs, 17 figs, 4 tabs.

  5. Durability of bends in high-temperature steam lines under the conditions of long-term operation

    Science.gov (United States)

    Katanakha, N. A.; Semenov, A. S.; Getsov, L. B.

    2015-04-01

    The article presents the results of stress-strain state computations and durability of bent and steeply curved branches of high-temperature steam lines carried out on the basis of the finite element method using the modified Soderberg formula for describing unsteady creep processes with taking the accumulation of damage into account. The computations were carried out for bends made of steel grades that are most widely used for manufacturing steam lines (12Kh1MF, 15Kh1M1F, and 10Kh9MFB) and operating at different levels of inner pressure and temperature. The solutions obtained using the developed creep model are compared with those obtained using the models widely used in practice.

  6. Qualification test of a main coolant pump for SMART pilot

    International Nuclear Information System (INIS)

    Park, Sang Jin; Yoon, Eui Soo; Oh, Hyong Woo

    2006-01-01

    SMART Pilot is a multipurpose small capacity integral type reactor. Main Coolant Pump (MCP) of SMART Pilot is a canned-motor-type axial pump to circulate the primary coolant between nuclear fuel and steam generator in the primary system. The reactor is designed to operate under condition of 310 .deg. C and 14.7 MPa. Thus MCP has to be tested under same operating condition as reactor design condition to verify its performance and safety. In present work, a test apparatus to simulate real operating situations of the reactor has been designed and constructed to test MCP. And then functional tests, performance tests, and endurance tests have been carried out upon a prototype MCP. Canned motor characteristics, homologous head/torque curves, coast-down curves, NPSH curves and life-time performance variations were obtained from the qualification test as well as hydraulic performance characteristics of MCP

  7. Experiment Operating Specification for the Semiscale MOD-2C feedwater and steam line break experiment series. Appendix S-FS-6 and 7

    International Nuclear Information System (INIS)

    Boucher, T.J.; Owca, W.A.

    1985-05-01

    This document is the Semiscale MOD-2C feedwater and steam line break experiment series Experiment Operating Specification Appendix for tests S-FS-6 and S-FS-7. Test S-FS-6 is the third test in the series and simulates a 100% break in a steam generator bottom feedwater line downstream of the check valve accompanied by compounding factors (such as check valve failure, loss-of-offsite power at SIS and SIS delayed until low steam generator pressure signal). The test is terminated after plant stabilization and recovery procedures including unaffected loop steam and feed, pressurizer heater operation, pressurizer auxiliary spray operation, and normal charging/letdown operation. Test S-FS-7 is the fourth test in the series and simulates a 14.3% break in a steam generator bottom feedwater line downstream of the check valve, accompanied by compounding factors. The test is terminated after plant stabilization procedures including unaffected loop steam and feed, pressurizer heater operation, and normal charging/letdown operation. The test was followed by an affected loop secondary refill after isolating the break. The Appendix contains information on the major fluid systems, initial experiment conditions, experiment boundary conditions, and sequence of experiment events. Also included is a discussion of the scaling criteria and philosophy used to develop the experiment initial and boundary conditions and system configuration

  8. 49 CFR 229.31 - Main reservoir tests.

    Science.gov (United States)

    2010-10-01

    ... appropriately safe environment. (d) Each aluminum main reservoir before being placed in use and at intervals... working pressure fixed by the chief mechanical officer. The test date, place, and pressure shall be... be subjected to a hydrostatic pressure of at least 25 percent more than the maximum working pressure...

  9. Coupled RELAP5/PANTHER/COBRA steam line break accident analysis in support of licensing DOEL 2 power uprate and steam generator replacement

    International Nuclear Information System (INIS)

    Zhang, J.; Bosso, S.; Henno, X.; Ouliddren, K.; Schneidesch, C.R.; Hove, W. van

    2004-01-01

    The nuclear reactor accident analyses using best estimate codes provide a better understanding and more accurate modeling of the key physical phenomena, which allows a more realistic evaluation of the conservatism and margins in the final safety analysis report (FSAR) accident analysis. The use of the best estimate codes and methods is necessary to meet the increasing technical, licensing and regulatory requirements for major plant changes (e.g. steam generator replacement), power uprate, core design optimization (cycle extension), as well as Periodic Safety Review. Since 1992, Tractebel Engineering (TE) has developed and applied a deterministic bounding approach to FASR accident analysis using the best estimate system thermal hydraulic code RELAP5/MOD2.5 and the subchannel thermal hydraulic code COBRA-3C. This approach has been accepted by the Belgian Safety Authorities, and turned out to be cost effective for most of the non-LOCA transient analyses. Since this approach adapts a decoupled modeling of the core responses, the analysis results often involved too large un-quantified conservatisms, due to either simplistic approximations for asymmetric accidents with strong 3D core neutronics - plant thermal hydraulics interactions, or additional penalties introduced from 'incoherent' initial/boundary conditions for separate plant and core analyses. Therefore, an external dynamic coupling between the RELAP5/MOD2.5 code and the 3-D neutronic code PANTHER was implemented since 1997 via the transient analysis code linkage program TALINK. Furthermore, a static linkage between the PANTHER code and the COBRA-3C code was developed for on-line calculation of (Departure from Nucleate Boiling Ratio (DNBR). TE intends to use the coupled code package for licensing non-symmetric FSAR accident analysis. The TE coupled code package has been applied to develop a main steam line break (MSLB) accident analysis methodology [using the TE deterministic bounding approach. The methodology

  10. Fabrication and testing of main sodium pumps of Superphenix 1

    International Nuclear Information System (INIS)

    Noel, H.; Pasqualini, G.

    1985-01-01

    The complexity of the loads involved and the extremely fine analysis required necessitates extensive design calculations for the Superphenix 1 primary and secondary pumps and associated expansion tanks, aiming toward detailed design validation, after slight adjustments, mainly to the secondary pumps and expansion tanks. The component parts to be built were far larger than those for the previous pumps (Rapsodie, Phenix), with very low manufacturing tolerances, which led to precision machining and welding operations, together with numerous dimensional inspections and materials characterization tests to achieve the required quality standards

  11. LECOTELO - conceptual design, testings and realisation of the main vessel

    International Nuclear Information System (INIS)

    Ioan, M.; Hororoi, M.

    2013-01-01

    Lead Corrosion Testing Loop (LECOTELO) facility was conceived to assure all conditions requested by corrosion/erosion tests in pure hot lead for different materials. The main vessel will receive at least 36 different material samples; each of them must be swept on both sides by a lead flow at a very well known speed. Taking into account that the inner system of this vessel is rather complex, it is very important to know the behavior of the vessel at different speeds of the lead flow around the samples. After many simulations of different configurations of the inner components, it was obtained the best inner geometry of the flow which provides the minimum pressure loss between inlet and outlet vessel. Consequently, the design of vessel components was changed in accordance with these new results of simulations and in this moment they are in the manufacturing process. (authors)

  12. Beam-induced quench test of LHC main quadrupole

    CERN Document Server

    Priebe, A; Dehning, B; Effinger, E; Emery, J; Holzer, E B; Kurfuerst, C; Nebot Del Busto, E; Nordt, A; Sapinski, M; Steckert, J; Verweij, A; Zamantzas, C

    2011-01-01

    Unexpected beam loss might lead to a transition of the accelerator superconducting magnet to a normal conducting state. The LHC beam loss monitoring (BLM) system is designed to abort the beam before the energy deposited in the magnet coils reach a quench-provoking level. In order to verify the threshold settings generated by simulation, a series of beam-induced quench tests at various beam energies has been performed. The beam losses are generated by means of an orbital bump peaked in one of main quadrupole magnets (MQ). The analysis includes not only BLM data but also the quench protection system (QPS) and cryogenics data. The measurements are compared to Geant4 simulations of energy deposition inside the coils and corresponding BLM signal outside the cryostat.

  13. Influence of transient flow in the formation of condensate and in the calculation of steam line

    International Nuclear Information System (INIS)

    Bazzo, E.

    1989-01-01

    The piping design is analyzed in unsteady-state conditions, with the main goal of minimizing operational costs and initial investments of a plant. All heat losses are calculated by applying the control volume method. The results confirm the applicability of the method and show that the influence of the transient regime on the condensation rate and economical insulation thickness must be considered. (author)

  14. Sealing performance test for main flange of pressure vessel of T2 test section in HENDEL

    International Nuclear Information System (INIS)

    Ioka, Ikuo; Inagaki, Yoshiyuki; Matsumoto, Kiminori; Kondou, Yasuo; Suzuki, Kunihiko; Miyamoto, Yoshiaki; Asami, Masanobu.

    1990-12-01

    A pressure vessel of T 2 test section in helium engineering demonstration loop (HENDEL) was fabricated to the same scale of the reactor pressure vessel made of 2(1/4)Cr-1Mo steel in high temperature engineering test reactor (HTTR). Also, the sealing structure of a main flange of pressure vessel in T 2 test section was composed of the double metal O-rings and Ω-seal which would be used in the sealing structure of HTTR. The sealing performance test for the main flange of the pressure vessel in T 2 test section was carried out to confirm the integrity of sealing structure of a main flange in HTTR. T 2 test section has been operated about 7700 hours in previous 18 cycles. The leakage of helium gas from inner metal O-ring was measured by the static pressurized process under the operating condition of HTTR (helium gas: 400degC, 40kg/cm 2 G, 4gk/s). The calculated leakage of helium gas was less than 9.6x10 -7 atm·cm 3 /sec. From the result, it is expected that the sealing structure of main flange in HTTR would maintain the leak tightness in the life. (author)

  15. Test Protocols for Advanced Inverter Interoperability Functions – Main Document

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, Jay Dean [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gonzalez, Sigifredo [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Ralph, Mark E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Ellis, Abraham [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Broderick, Robert Joseph [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2013-11-01

    Distributed energy resources (DER) such as photovoltaic (PV) systems, when deployed in a large scale, are capable of influencing significantly the operation of power systems. Looking to the future, stakeholders are working on standards to make it possible to manage the potentially complex interactions between DER and the power system. In 2009, the Electric Power Research Institute (EPRI), Sandia National Laboratories (SNL) with the U.S. Department of Energy (DOE), and the Solar Electric Power Association (SEPA) initiated a large industry collaborative to identify and standardize definitions for a set of DER grid support functions. While the initial effort concentrated on grid-tied PV inverters and energy storage systems, the concepts have applicability to all DER. A partial product of this on-going effort is a reference definitions document (IEC TR 61850-90-7, Object models for power converters in distributed energy resources (DER) systems) that has become a basis for expansion of related International Electrotechnical Commission (IEC) standards, and is supported by US National Institute of Standards and Technology (NIST) Smart Grid Interoperability Panel (SGIP). Some industry-led organizations advancing communications protocols have also embraced this work. As standards continue to evolve, it is necessary to develop test protocols to independently verify that the inverters are properly executing the advanced functions. Interoperability is assured by establishing common definitions for the functions and a method to test compliance with operational requirements. This document describes test protocols developed by SNL to evaluate the electrical performance and operational capabilities of PV inverters and energy storage, as described in IEC TR 61850-90-7. While many of these functions are not currently required by existing grid codes or may not be widely available commercially, the industry is rapidly moving in that direction. Interoperability issues are already

  16. Modelling and simulation of the steam line, the high and low pressure turbines and the pressure regulator for the SUN-RAH nucleo electric university simulator

    International Nuclear Information System (INIS)

    Lopez R, A.

    2003-01-01

    In the following article the development of a simulator that allows to represent the dynamics of the following systems: steam line, nozzle, vapor separator, reheater, high pressure turbine, low pressure turbine, power generator and the pressure regulator of a nucleo electric power station. We start from the supposition that this plant will be modeled from a nuclear reactor type BWR (Boiling Water Reactor), using models of reduced order that represent the more important dynamic variables of the physical processes that happen along the steam line until the one generator. To be able to carry out the simulation in real time the Mat lab mathematical modeling software is used, as well as the specific simulation tool Simulink. It is necessary to point out that the platform on which the one is executed the simulator is the Windows operating system, to allow the intuitive use that only this operating system offers. The above-mentioned obeys to that the objective of the simulator it is to help the user to understand some of the dynamic phenomena that are present in the systems of a nuclear plant, and to provide a tool of analysis and measurement of variables to predict the desirable behavior of the same ones. The model of a pressure controller for the steam lines, the high pressure turbine and the low pressure turbine is also presented that it will be the one in charge of regulating the demand of the system according to the characteristics and critic restrictions of safety and control, assigned according to those wanted parameters of performance of this system inside the nucleo electric plant. This simulator is totally well defined and it is part of the University student nucleo electric simulator with Boiling Water Reactor (SUN-RAH), an integral project and of greater capacity. (Author)

  17. RELAP/MOD1.5 analysis of steam line break transients for a 3-loop and a 4-loop Westinghouse nuclear steam supply system

    International Nuclear Information System (INIS)

    Peeler, G.B.; McDonald, T.A.; Kennedy, M.F.

    1984-01-01

    RELAP/MOD1.5 (Cycle 31 and 34) calculations were made to assess the assumptions used by Westinghouse (W) to analyze mainsteam line break transients. Models of a W 3-loop and 4-loop nuclear steam supply system were used. Sensitivity studies were performed to determine the effect of the availability of offsite power, break size and initial core power. Comparison with W results indicated that if the assumptions used by W are replicated within the RELAP5 framework, then the W methodology for prediction of the Nuclear Steam Supply System (NSSS) response is conservative for steam line break transients

  18. First-ever evening public engine test of a Space Shuttle Main Engine

    Science.gov (United States)

    2001-01-01

    Thousands of people watch the first-ever evening public engine test of a Space Shuttle Main Engine at NASA's John C. Stennis Space Center. The spectacular test marked Stennis Space Center's 20th anniversary celebration of the first Space Shuttle mission.

  19. Performance Evaluation and Quality Assurance Management during the Series Power Tests of LHC Main Lattice Magnets

    CERN Document Server

    Siemko, A

    2008-01-01

    Within the LHC magnet program a series production of superconducting dipoles and quadrupoles has recently been completed in industry and all magnets were cold tested at CERN. The main features of these magnets are: two-in-one structure, 56 mm aperture, two layer coils wound from 15.1 mm wide Nb-Ti cables, and all-polyimide insulation. This paper reviews the process of the power test quality assurance and performance evaluation, which was applied during the LHC magnet series tests. The main test results of magnets tested in both supercritical and superfluid helium, including the quench training, the conductor performance, the magnet protection efficiency and the electrical integrity are presented and discussed in terms of the design parameters and the requirements of the LHC project.

  20. Development and evaluation of the NSSS model with four steam lines for the LVNP using the SCDAPSIM code

    International Nuclear Information System (INIS)

    Salazar C, J.H.; Nunez C, A.; Camargo C, R.

    2005-01-01

    The present work shows the pattern of the NSSS considering the four main vapor lines as well as their evaluation. The pattern was developed by the National Commission of Nuclear Security and Safeguards (CNSNS) and it has as main objective to account with a model of the Laguna Verde Nuclear power plant (CNLV) for the simulation and analysis of transitory events where are involved some of main vapor lines, or some relief valves and safety (SRV's). The model was evaluated with data of the CNLV. In 1996 the Federal Commission of Electricity (CFE) request to the CNSNS permission to operate the Unit 2 until the first recharge, having the main vapor line 'B' isolated and operating with a level of power corresponding to a flow of total vapor of 85% of the nominal one (of 1931 MWt). The obtained values were compared with the obtained registrations of the CNLV in order to evaluate the model. Those results show relative errors inferior to 3% among the CNLV reported value and the one calculated by the SCDAPSIM code. (Author)

  1. Main maintenance operations for Test Blanket Systems in ITER TBM port cells

    Energy Technology Data Exchange (ETDEWEB)

    Pascal, R., E-mail: romain.pascal@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Cortes, P.; Friconneau, J.-P.; Giancarli, L.M.; Gotewal, K.K.; Iseli, M.; Kim, B.Y.; Levesy, B.; Martins, J.-P.; Merola, M. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Nevière, J.-C. [Comex-Nucleaire, 13115 Saint Paul Lez Durance (France); Patisson, L. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Siarras, A. [Sogetti, Parc de la Duranne, 13857 Aix-en-Provence (France); Tesini, A. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2013-10-15

    Highlights: • The Test Blanket System components layout in Port Cell room is described. • The maintenance of the two Test Blanket Systems in ITER port cell is addressed. • The overall replacement/maintenance strategy is defined. • The main maintenance tasks of the systems are discussed. • The maintenance strategy and required tools are presented. -- Abstract: Each Test Blanket System in ITER is formed by an in-vessel component, the Test Blanket Module, and several associated ancillary systems (coolant and Tritium systems, instrumentation and control systems). The paper describes the overall replacement/maintenance strategy and the main maintenance tasks that have to be considered in the design of the systems. It shows that there are no critical issues.

  2. Overview of the main challenges for the engineering design of the test facilities system of IFMIF

    International Nuclear Information System (INIS)

    Molla, J.; Nakamura, K.

    2009-01-01

    High intense radiation fields were demanded to IFMIF to address the lack of information on effects in materials due to radiation fields with fusion reactor features. Such intense radiation fields will also produce a number of unwanted effects in exposed materials and components. The main difficulties to achieve a reliable engineering design of the Test Facilities System during the Engineering Validation and the Engineering Design phase of IFMIF now under development are reviewed in this paper. The most challenging activities will be the design of the high flux test module, the creep fatigue test module, the test cell and the remote handling system. The intense radiation fields in the irradiation area and the high availability required for IFMIF (70%) are the main reasons for these difficulties.

  3. Application Research on Testing Efficiency of Main Drainage Pump in Coal Mine Using Thermodynamic Theories

    OpenAIRE

    Shang, Deyong

    2017-01-01

    The efficiency of a drainage pump should be tested at regular intervals to master the status of the drainage pump in real time and thus achieve the goal of saving energy. The ultrasonic flowmeter method is traditionally used to measure the flow of the pump. But there are some defects in this kind of method of underground coal mine. This paper first introduces the principle of testing the main drainage pump efficiency in coal mine using thermodynamic theories, then analyzes the energy transfor...

  4. History and Benefits of Engine Level Testing Throughout the Space Shuttle Main Engine Program

    Science.gov (United States)

    VanHooser, Katherine; Kan, Kenneth; Maddux, Lewis; Runkle, Everett

    2010-01-01

    Rocket engine testing is important throughout a program s life and is essential to the overall success of the program. Space Shuttle Main Engine (SSME) testing can be divided into three phases: development, certification, and operational. Development tests are conducted on the basic design and are used to develop safe start and shutdown transients and to demonstrate mainstage operation. This phase helps form the foundation of the program, demands navigation of a very steep learning curve, and yields results that shape the final engine design. Certification testing involves multiple engine samples and more aggressive test profiles that explore the boundaries of the engine to vehicle interface requirements. The hardware being tested may have evolved slightly from that in the development phase. Operational testing is conducted with mature hardware and includes acceptance testing of flight assets, resolving anomalies that occur in flight, continuing to expand the performance envelope, and implementing design upgrades. This paper will examine these phases of testing and their importance to the SSME program. Examples of tests conducted in each phase will also be presented.

  5. Application Research on Testing Efficiency of Main Drainage Pump in Coal Mine Using Thermodynamic Theories

    Directory of Open Access Journals (Sweden)

    Deyong Shang

    2017-01-01

    Full Text Available The efficiency of a drainage pump should be tested at regular intervals to master the status of the drainage pump in real time and thus achieve the goal of saving energy. The ultrasonic flowmeter method is traditionally used to measure the flow of the pump. But there are some defects in this kind of method of underground coal mine. This paper first introduces the principle of testing the main drainage pump efficiency in coal mine using thermodynamic theories, then analyzes the energy transformation during the process of draining water, and finally derives the calculation formulae of the pump efficiency, which meet the on-site precision of engineering. On the basis of analyzing the theories, the protective sleeve and the base of the temperature sensor are designed to measure the water temperature at inlet and outlet of the pump. The efficiencies of pumps with two specifications are measured, respectively, by using the thermodynamic method and ultrasonic flowmeter method. By contrast, the results show that thermodynamic method can satisfy the precision of the testing requirements accuracy for high-flow and high-lift drainage pump under normal temperatures. Moreover, some measures are summed up to improve the accuracy of testing the pump efficiency, which are of guiding significance for on-site testing of the main drainage pump efficiency in coal mine.

  6. Technology, safety, and costs of decommissioning reference nuclear research and test reactors. Main report

    International Nuclear Information System (INIS)

    Konzek, G.J.; Ludwick, J.D.; Kennedy, W.E. Jr.; Smith, R.I.

    1982-03-01

    Safety and Cost Information is developed for the conceptual decommissioning of two representative licensed nuclear research and test reactors. Three decommissioning alternatives are studied to obtain comparisons between costs (in 1981 dollars), occupational radiation doses, potential radiation dose to the public, and other safety impacts. The alternatives considered are: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). The study results are presented in two volumes. Volume 1 (Main Report) contains the results in summary form

  7. Charge measurement and mitigation for the main test masses of the GEO 600 gravitational wave observatory

    International Nuclear Information System (INIS)

    Hewitson, M; Danzmann, K; Grote, H; Hild, S; Hough, J; Lueck, H; Rowan, S; Smith, J R; Strain, K A; Willke, B

    2007-01-01

    Spurious charging of the test masses in gravitational wave interferometers is a well-known problem. Typically, concern arises due to the possibility of increased thermal noise due to a lowering of the quality factor of modes of the test-mass suspension, or due to the potential for increased displacement noise arising from charge migration on the surface of the test masses. Recent experience gained at the GEO 600 gravitational wave detector has highlighted an additional problem. GEO 600 uses electrostatic actuators to control the longitudinal position of the main test masses. The presence of charge on the test masses is shown to strongly affect the performance of the electrostatic actuators. This paper reports on a measurement scheme whereby the charge state of the GEO 600 test masses can be measured using the electrostatic actuators. The resulting measurements are expressed in terms of an effective bias voltage on the electrostatic actuators. We also describe attempts to remove the charge from the test masses and we show that the use of UV illumination was the most successful. Using UV illumination we were able to discharge and re-charge the test masses

  8. Coolant Mixing in a Pressurized Water Reactor: Deboration Transients, Steam-Line Breaks, and Emergency Core Cooling Injection

    International Nuclear Information System (INIS)

    Prasser, Horst-Michael; Grunwald, Gerhard; Hoehne, Thomas; Kliem, Soeren; Rohde, Ulrich; Weiss, Frank-Peter

    2003-01-01

    The reactor transient caused by a perturbation of boron concentration or coolant temperature at the inlet of a pressurized water reactor (PWR) depends on the mixing inside the reactor pressure vessel (RPV). Initial steep gradients are partially lessened by turbulent mixing with coolant from the unaffected loops and with the water inventory of the RPV. Nevertheless the assumption of an ideal mixing in the downcomer and the lower plenum of the reactor leads to unrealistically small reactivity inserts. The uncertainties between ideal mixing and total absence of mixing are too large to be acceptable for safety analyses. In reality, a partial mixing takes place. For realistic predictions it is necessary to study the mixing within the three-dimensional flow field in the complicated geometry of a PWR. For this purpose a 1:5 scaled model [the Rossendorf Coolant Mixing Model (ROCOM) facility] of the German PWR KONVOI was built. Compared to other experiments, the emphasis was put on extensive measuring instrumentation and a maximum of flexibility of the facility to cover as much as possible different test scenarios. The use of special electrode-mesh sensors together with a salt tracer technique provided distributions of the disturbance within downcomer and core entrance with a high resolution in space and time. Especially, the instrumentation of the downcomer gained valuable information about the mixing phenomena in detail. The obtained data were used to support code development and validation. Scenarios investigated are the following: (a) steady-state flow in multiple coolant loops with a temperature or boron concentration perturbation in one of the running loops, (b) transient flow situations with flow rates changing with time in one or more loops, such as pump startup scenarios with deborated slugs in one of the loops or onset of natural circulation after boiling-condenser-mode operation, and (c) gravity-driven flow caused by large density gradients, e.g., mixing of cold

  9. Neutron radiography and other NDE tests of main rotor helicopter blades

    International Nuclear Information System (INIS)

    Beer, F.C. de; Coetzer, M.; Fendeis, D.; Silva, A. da Costa E

    2004-01-01

    A few nondestructive examination (NDE) techniques are extensively being used worldwide to investigate aircraft structures for all types of defects. The detection of corrosion and delaminations, which are believed to be the major initiators of defects leading to aircraft structural failures, are addressed by various NDE techniques. In a combined investigation by means of visual inspection, X-ray radiography and shearography on helicopter main rotor blades, neutron radiography (NRad) at SAFARI-1 research reactor operated by Necsa, was performed to introduce this form of NDE testing to the South African aviation industry to be evaluated for applicability. The results of the shearography, visual inspection and NRad techniques are compared in this paper. The main features and advantages of neutron radiography, within the framework of these investigations, will be highlighted

  10. Purpose-in-Life Test: Comparison of the Main Models in Patients with Mental Disorders.

    Science.gov (United States)

    García-Alandete, Joaquín; Marco, José H; Pérez, Sandra

    2017-06-27

    The aim of this study was to compare the main proposed models for the Purpose-In-Life Test, a scale for assessing meaning in life, in 229 Spanish patients with mental disorders (195 females and 34 males, aged 13-68, M = 34.43, SD = 12.19). Confirmatory factor-analytic procedures showed that the original model of the Purpose-In-Life Test, a 20-item unidimensional scale, obtained a better fit than the other analyzed models, SBχ2(df) = 326.27(170), SBχ2/df = 1.92, TLI = .93, CFI = .94, IFI = .94, RMSEA = .063 (90% CI [.053, .074]), CAIC = -767.46, as well as a high internal consistency, (α = .90). The main conclusion is that the original version of the Purpose-In-Life shows a robust construct validity in a clinical population. However, authors recommend an in-depth psychometric analysis of the Purpose-In-Life Test among clinical population. Likewise, the importance of assessing meaning in life in order to enhance psychotherapeutic treatment is noted.

  11. HTTR hydrogen production system. Structure and main specifications of mock-up test facility (Contract research)

    International Nuclear Information System (INIS)

    Kato, Michio; Aita, Hideki; Inagaki, Yoshiyuki; Hayashi, Koji; Ohashi, Hirofumi; Sato, Hiroyuki; Iwatsuki, Jin; Takada, Shoji; Inaba, Yoshitomo

    2007-03-01

    The mock-up test facility was fabricated to investigate performance of the steam generator for mitigation of the temperature fluctuation of helium gas and transient behavior of the hydrogen production system for HTTR and to obtain experimental data for verification of a dynamic analysis code. The test facility has an approximate hydrogen production capacity of 120Nm 3 /h and the steam reforming process of methane; CH 4 +H 2 O=3H 2 +CO, was used for hydrogen production of the test facility. An electric heater was used as a heat source instead of the reactor in order to heat helium gas up to 880degC (4MPa) at the chemical reactor inlet which is the same temperature as the HTTR hydrogen production system. Fabrication of the test facility was completed in February in 2002, and seven cycle operations were carried out from March in 2002 to December in 2004. This report describes the structure and main specifications of the test facility. (author)

  12. Reactor protection system software test-case selection based on input-profile considering concurrent events and uncertainties

    International Nuclear Information System (INIS)

    Khalaquzzaman, M.; Lee, Seung Jun; Cho, Jaehyun; Jung, Wondea

    2016-01-01

    Recently, the input-profile-based testing for safety critical software has been proposed for determining the number of test cases and quantifying the failure probability of the software. Input-profile of a reactor protection system (RPS) software is the input which causes activation of the system for emergency shutdown of a reactor. This paper presents a method to determine the input-profile of a RPS software which considers concurrent events/transients. A deviation of a process parameter value begins through an event and increases owing to the concurrent multi-events depending on the correlation of process parameters and severity of incidents. A case of reactor trip caused by feedwater loss and main steam line break is simulated and analyzed to determine the RPS software input-profile and estimate the number of test cases. The different sizes of the main steam line breaks (e.g., small, medium, large break) with total loss of feedwater supply are considered in constructing the input-profile. The uncertainties of the simulation related to the input-profile-based software testing are also included. Our study is expected to provide an option to determine test cases and quantification of RPS software failure probability. (author)

  13. MK-III function tests in JOYO. Primary main cooling pump

    International Nuclear Information System (INIS)

    Isozaki, Kazunori; Saito, Takakazu; Sumino, Kouzo; Karube, Kouji; Terano, Toshihiro; Sakaba, Hideo; Nakai, Satoru

    2004-06-01

    MK-III function test (SKS-1) that was carried out from October 17, 2001 through October 23, 2001 using MK-III transition core configuration and MK-III function tests (SKS-2) was carried out from January 27, 2003 through February 13, 2003 using MK-III core configuration. The major function tests results of primary cooling system were shown as follows; (1) The stability of the primary main pump flow control system was confirmed on both CAS (cascade) mode and Man (manual) mode. Also no divergence of flow and revolution of the pump were observed at step flow change disturbance. (2) The main motor was shifted to run-back flow control operation in about 54 seconds after scram. The flow rate and pump revolution at run-back operation of A and B cooling system were 167 m 3 /h and 117 rpm, 185m 3 /h and 118 rpm respectively. The pump revolution was within the design target revolution 122 rpm ± 8 rpm and the flow was over the 10% of the rated flow. (3) The pony motor was engaged in operation in about 39 seconds after the primary main pump trip. The flow rate and pump revolution at the pony motor operation of A and B cooling system were 180 m 3 /h and 124 rpm, 190 m 3 /h and 123 rpm respectively. These values were satisfied the design low limit of 93 rpm and 10% of the rated flow. (4) Free flow coast down time constant was longer than 10 seconds that was design shortest time at both the primary pump trip and run-back operation. (5) Pump over flow column sodium levels of both A and B cooling system at rated operating condition were NL-1550 mm and, NL-1468 mm respectively and were lower than NL-1581 mm of the design value. This result shows the new IHX pressure loss estimation was conservative. (6) It was confirmed that the primary main pump could operate with out scram for up to 0.6 seconds of external power supply loss. (author)

  14. Zircaloy PWR fuel cladding deformation tests under mainly convective cooling conditions

    International Nuclear Information System (INIS)

    Hindle, E.D.; Mann, C.A.

    1980-01-01

    In a loss-of-coolant accident the temperature of the cladding of the fuel rods may rise to levels (650-810 0 C) where the ductility of Zircaloy is high (approximately 80%). The net outward pressure which will obtain if the coolant pressure falls to a small fraction of its normal working value produces stresses in the cladding which can result in large strain through secondary creep. An earlier study of the deformation of specimens of PWR Zircaloy cladding tubing 450 mm long under internal pressure had shown that strains of over 50% could be produced over considerable lengths (greater than twenty tube diameters). Extended deformation of this sort might be unacceptable if it occurred in a fuel element. The previous tests had been carried out under conditions of uniform radiative heat loss, and the work reported here extends the study to conditions of mainly convective heat loss believed to be more representative of a fuel element following a loss of coolant. Zircaloy-4 cladding specimens 450 mm long were filled with alumina pellets and tested at temperatures between 630 and 845 0 C in flowing steam at atmospheric pressure. Internal test pressures were in the range 2.9-11.0 MPa (400-1600 1b/in 2 ). Maximum strains were observed of the same magnitude as those seen in the previous tests, but the shape of the deformation differed; in these tests the deformation progressively increased in the direction of the steam flow. These results are compared with those from multi-rod tests elsewhere, and it is suggested that heat transfer has a dominant effect in determining deformation. The implications for the behaviour of fuel elements in a loss-of-coolant accident are outlined. (author)

  15. Quick Look Report for Semiscale MOD-2C Test S-FS-2

    International Nuclear Information System (INIS)

    Boucher, T.J.; Chen, T.H.

    1985-01-01

    Results of a preliminary analysis of the first test performed in the Semiscale MOD-2C Steam Generator Feedwater and Steam Line Break (FS) experiment series are presented. Test S-FS-2 simulated a pressurized water reactor transient initiated by a double-ended offset shear of a steam generator main steam line upstream of the flow restrictor. Initial conditions represented normal ''hot-standby'' operation. The transient included an initial 600-s period in which only automatic plant protection systems responded to the initiating event. This period was followed by a series of operator actions necessary to stabilize the plant at conditions required to allow a natural circulation cooldown. The test results provided a measured evaluation of the effectiveness of the automatic responses in minimizing primary system overcooling and operator actions in stabilizing the plant. Test data also provided a basis for comparison with other tests in the series of the effects of break size on primary overcooling and primary-to-secondary heat transfer. 57 figs., 3 tabs

  16. Use of an expert system data analysis manager for space shuttle main engine test evaluation

    Science.gov (United States)

    Abernethy, Ken

    1988-01-01

    The ability to articulate, collect, and automate the application of the expertise needed for the analysis of space shuttle main engine (SSME) test data would be of great benefit to NASA liquid rocket engine experts. This paper describes a project whose goal is to build a rule-based expert system which incorporates such expertise. Experiential expertise, collected directly from the experts currently involved in SSME data analysis, is used to build a rule base to identify engine anomalies similar to those analyzed previously. Additionally, an alternate method of expertise capture is being explored. This method would generate rules inductively based on calculations made using a theoretical model of the SSME's operation. The latter rules would be capable of diagnosing anomalies which may not have appeared before, but whose effects can be predicted by the theoretical model.

  17. Results of temperature test 6 in the Asse salt mine. Volume 1 - Main report

    International Nuclear Information System (INIS)

    Feddersen, H.; Flach, D.; Flentge, I.

    1986-01-01

    In the year 1985 a heater test with a mean heat load of 50 kW was carried out in the Asse salt mine for 78 days. Its main aims were to investigate possible fracturing of the rock; investigations on the transport of brine and gases; comparison of the measured mechanical stresses and temperatures, as compared to those determined by numerical methods. The evaluation of the measurement results was impeded by premature failure of some of the heaters, which proved to be a handicap to the symmetry of the experiment. It was possible, nevertheless, to find a good agreement between the measured and the numerically calculated temperatures. The mechanical stress measurements showed, as compared to the 2D-FE-calculations, that the measured stresses lay within the expected range. Fracturing was detected by means of seismic observations, especially after termination of the heating. Brine transport was ascertained using geoelectric four point -and self-potential measurements. The staining test showed no sharp fracturing of the rock salt, but a loosened-up zone at the grain boundaries impregnated with staining oil

  18. Pretest analysis document for Semiscale Test S-FS-1

    International Nuclear Information System (INIS)

    Chen, T.H.

    1985-02-01

    This report documents the pretest analysis calculation completed with the RELAP5/MOD2/CY21 code for Semiscale Test S-FS-1. The test will simulate the double-ended offset shear of the main steam line at the exit of the broken loop steam generator (downstream of the flow restrictor) and the subsequent plant recovery. The recovery portion of the test consists of a plant stabilization phase and a plant cooldown phase. The recovery procedures involve normal charging/letdown operation, pressurizer heater operation, secondary steam and feed of the unaffected steam generator, and pressurizer auxiliary spray. The test will be terminated after the unaffected steam generator and pressurizer pressures and liquid levels are stable, and the average priamry fluid temperature is stable at about 480 K (405 0 F) for at least 10 minutes

  19. Examining Innovative Divertor and Main Chamber Options for a National Divertor Test Tokamak

    Science.gov (United States)

    Labombard, B.; Umansky, M.; Brunner, D.; Kuang, A. Q.; Marmar, E.; Wallace, G.; Whyte, D.; Wukitch, S.

    2016-10-01

    The US fusion community has identified a compelling need for a National Divertor Test Tokamak. The 2015 Community Planning Workshop on PMI called for a national working group to develop options. Important elements of a NDTT, adopted from the ADX concept, include the ability to explore long-leg divertor `solutions for power exhaust and particle control' (Priority Research Direction B) and to employ inside-launch RF actuators combined with double-null topologies as `plasma solution for main chamber wall components, including tools for controllable sustained operation' (PRD-C). Here we examine new information on these ideas. The projected performance of super-X and X-point target long-leg divertors is looking very promising; a stable fully-detached divertor condition handling an order-of-magnitude increase in power handling over conventional divertors may be possible. New experiments on Alcator C-Mod are addressing issues of high-field side versus low-field side heat flux sharing in double-null topologies and the screening of impurities that might originate from RF actuators placed in the high-field side - both with favorable results. Supported by USDoE Awards DE-FC02-99ER54512 and DE-AC52-07NA27344.

  20. Test Results of the Third LHC Main Quadrupole Magnet Prototype at CEA/Saclay

    CERN Document Server

    Derégel, J; Gourdin, C; Hervieu, M; Ogitsu, T; Peyrot, M; Rifflet, J M; Schild, T; Simon, F; Tortschanoff, Theodor; Tsuchiya, K

    2002-01-01

    The construction of the third second-generation main quadrupole magnet prototype for LHC has been completed at CEA/Saclay in November 2000. The magnet was tested at 1.9 K. Similarly to the two first ones, this prototype has exceeded the operating current in one training step and exhibited excellent training memory after a thermal cycle. This paper describes the quench performance and quench start localization determined by means of voltage-taps and a quench antenna system developed in collaboration with KEK. As this magnet was equipped with capacitive gauges, the stresses during cool-down and powering have been recorded and are in agreement with FE computations. The newly designed quench heaters have improved efficiency and reproducibility compared to those of the first generation. Magnetic measurements have been performed at various stages. The cold measurements show minor differences with those at room temperature and are similar to those of the two first magnets of this design. These results prove that the...

  1. Assessment of SPACE Code Using the LSTF 10% MSLB Test

    International Nuclear Information System (INIS)

    Kim, Yo Han; Yang, Chang Keun; Ha, Sang Jun

    2012-01-01

    The Korea Nuclear Hydro and Nuclear Power Co. (KHNP) has developed a multipurpose nuclear safety analysis code called the Safety and Performance Analysis Code for Nuclear Power Plants (SPACE). The SPACE is a best-estimated two-phase three-field thermal-hydraulic analysis code used to analyze the safety and performance of pressurized water reactors (PWRs). As in the second phase of the project, the beta version of the code has been developed through the validation and verification (V and V) using integral loop test data or plant operating data and the complement of code to solve the SPACE code user problem and resolution reports. In this study, the Large Scale Test Facility (LSTF) 10% main steam line break (MSLB) test, SB-SL-01, was simulated as a V and V work. The results were compared with the experimental data and those of the RELAP5/MOD3.1 code simulation

  2. Comparative analysis of methods for workload assessment of the main control room operators of NPP

    International Nuclear Information System (INIS)

    Georgiev, V.; Petkov, G.

    2008-01-01

    The paper presents benchmarking workload results obtained by a method for operator workload assessment – NASA Task Load Index, and a method for human error probability assessment - Performance Evaluation of Teamwork. Based on the archives of FSS-1000 training in the accident “Main Steam Line Tube Rupture at the WWER-1000 Containment” the capacities of the two methods for direct and indirect workload assessment are evaluated

  3. HIV Testing and Awareness of Partner's HIV Status Among Chinese Men Who Have Sex with Men in Main Partnerships.

    Science.gov (United States)

    Wei, Chongyi; Yan, Hongjing; Raymond, H Fisher; Shi, Ling-En; Li, Jianjun; Yang, Haitao; McFarland, Willi

    2016-04-01

    Many men who have sex with men (MSM) do not use condoms with their main partners, especially if both parties are of the same HIV status. However, significant proportions of MSM have never tested or recently tested and are unaware of their main partners' HIV status. A cross-sectional survey was conducted among 524 MSM in Jiangsu, China in 2013-2014. Time-location sampling and online convenience sampling were used to recruit participants. We compared awareness of HIV status and recent HIV testing between participants who had main partners versus those who did not, and identified factors associated with recent HIV testing among men in main partnerships. Participants in main partnerships were significantly more likely to report recent HIV testing and being HIV-negative instead of HIV-unknown compared to participants in casual partnerships only. Overall, 74.5 % of participants were aware of their main partners' HIV status. Among participants in main partnerships, those who had 2-5 male anal sex partners in the past 6 months and those who reported that their partners were HIV-negative had 2.36 (95 % CI 1.12, 4.97) and 4.20 (95 % CI 2.03, 8.70) fold greater odds of being tested in the past year compared to those who had main partners only and those whose partners were HIV-positive/unknown, respectively. Chinese MSM in main partnerships might be practicing serosorting and may be at lower risk for HIV infection due to increased awareness of main partners' HIV status and higher uptake of recent testing.

  4. International Space Station Lithium-Ion Main Battery Thermal Runaway Propagation Test

    Science.gov (United States)

    Dalton, Penni J.; North, Tim

    2017-01-01

    In 2010, the ISS Program began the development of Lithium-Ion (Li-Ion) batteries to replace the aging Ni-H2 batteries on the primary Electric Power System (EPS). After the Boeing 787 Li-Ion battery fires, the NASA Engineering and Safety Center (NESC) Power Technical Discipline Team was tasked by ISS to investigate the possibility of Thermal Runaway Propagation (TRP) in all Li-Ion batteries used on the ISS. As part of that investigation, NESC funded a TRP test of an ISS EPS non-flight Li-Ion battery. The test was performed at NASA White Sands Test Facility in October 2016. This paper will discuss the work leading up to the test, the design of the test article, and the test results.

  5. Neutron radiography and other NDE tests of main rotor helicopter blades

    CSIR Research Space (South Africa)

    De Beer, FC

    2004-10-01

    Full Text Available leading to aircraft structural failures, are addressed by various NDE techniques. In a combined investigation by means of visual inspection, X-ray radiography and shearography on helicopter main rotor blades, neutron radiography (NRad) at SAFARI-1 research...

  6. Science Library of Test Items. Volume Twenty-Two. A Collection of Multiple Choice Test Items Relating Mainly to Skills.

    Science.gov (United States)

    New South Wales Dept. of Education, Sydney (Australia).

    As one in a series of test item collections developed by the Assessment and Evaluation Unit of the Directorate of Studies, items are made available to teachers for the construction of unit tests or term examinations or as a basis for class discussion. Each collection was reviewed for content validity and reliability. The test items meet syllabus…

  7. Science Library of Test Items. Volume Eighteen. A Collection of Multiple Choice Test Items Relating Mainly to Chemistry.

    Science.gov (United States)

    New South Wales Dept. of Education, Sydney (Australia).

    As one in a series of test item collections developed by the Assessment and Evaluation Unit of the Directorate of Studies, items are made available to teachers for the construction of unit tests or term examinations or as a basis for class discussion. Each collection was reviewed for content validity and reliability. The test items meet syllabus…

  8. Science Library of Test Items. Volume Twenty. A Collection of Multiple Choice Test Items Relating Mainly to Physics, 1.

    Science.gov (United States)

    New South Wales Dept. of Education, Sydney (Australia).

    As one in a series of test item collections developed by the Assessment and Evaluation Unit of the Directorate of Studies, items are made available to teachers for the construction of unit tests or term examinations or as a basis for class discussion. Each collection was reviewed for content validity and reliability. The test items meet syllabus…

  9. Science Library of Test Items. Volume Seventeen. A Collection of Multiple Choice Test Items Relating Mainly to Biology.

    Science.gov (United States)

    New South Wales Dept. of Education, Sydney (Australia).

    As one in a series of test item collections developed by the Assessment and Evaluation Unit of the Directorate of Studies, items are made available to teachers for the construction of unit tests or term examinations or as a basis for class discussion. Each collection was reviewed for content validity and reliability. The test items meet syllabus…

  10. Science Library of Test Items. Volume Nineteen. A Collection of Multiple Choice Test Items Relating Mainly to Geology.

    Science.gov (United States)

    New South Wales Dept. of Education, Sydney (Australia).

    As one in a series of test item collections developed by the Assessment and Evaluation Unit of the Directorate of Studies, items are made available to teachers for the construction of unit tests or term examinations or as a basis for class discussion. Each collection was reviewed for content validity and reliability. The test items meet syllabus…

  11. Leak before break evaluation for main steam piping system made of SA106 Gr.C

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Kyoung Mo; Jee, Kye Kwang; Pyo, Chang Ryul; Ra, In Sik [Korea Power Engineering Company, Seoul (Korea, Republic of)

    1997-04-01

    The basis of the leak before break (LBB) concept is to demonstrate that piping will leak significantly before a double ended guillotine break (DEGB) occurs. This is demonstrated by quantifying and evaluating the leak process and prescribing safe shutdown of the plant on the basis of the monitored leak rate. The application of LBB for power plant design has reduced plant cost while improving plant integrity. Several evaluations employing LBB analysis on system piping based on DEGB design have been completed. However, the application of LBB on main steam (MS) piping, which is LBB applicable piping, has not been performed due to several uncertainties associated with occurrence of steam hammer and dynamic strain aging (DSA). The objective of this paper is to demonstrate the applicability of the LBB design concept to main steam lines manufactured with SA106 Gr.C carbon steel. Based on the material properties, including fracture toughness and tensile properties obtained from the comprehensive material tests for base and weld metals, a parametric study was performed as described in this paper. The PICEP code was used to determine leak size crack (LSC) and the FLET code was used to perform the stability assessment of MS piping. The effects of material properties obtained from tests were evaluated to determine the LBB applicability for the MS piping. It can be shown from this parametric study that the MS piping has a high possibility of design using LBB analysis.

  12. Integration of test modules in the main blanket and vacuum vessel design

    International Nuclear Information System (INIS)

    Nakahira, Masataka; Kurasawa, Toshimasa; Sato, Satoshi; Furuya, Kazuyuki; Togami, Ikuhide; Hashimoto, Toshiyuki; Takatsu, Hideyuki; Kuroda, Toshimasa.

    1995-07-01

    Typical test modules for water-cooled and helium-cooled ceramic breeder blankets have been designed, and their major design parameters are summarized. Among various candidates studied in Japan at present, BOT (Breeder Out of Tube) type of blanket is exemplified here. The integration scheme of the test module into ITER basic machine is also shown. Even with other type of blanket, the integration scheme won't be affected. The composition and space requirement of cooling and tritium recovery systems for the test module have also been studied. (author)

  13. Test study on safety features of station blackout accident for nuclear main pump

    International Nuclear Information System (INIS)

    Liu Xiajie; Wang Dezhong; Zhang Jige; Liu Junsheng; Yang Zhe

    2009-01-01

    The theoretical and experimental studies of reactor coolant pump accidents encountered nation-wide and world-wide were described. To investigate the transient hydrodynamic performance of reactor coolant pump (RCP) during the period of rotational inertia in the station blackout accident, some theoretical and experimental studies were carried out, and the analysis of the test results was presented. The experiment parameters, conditions and test methods were introduced. The flow-rate, rotate speed and vibrations were analyzed emphatically. The quadruplicate polynomial curve equation was used to simulate the flow-rate,rotate speed along with time. The test results indicate that the flow-rate and rotator speed decrease rapidly at the very beginning of cut power and the test results accord with the regulation of safety standard. The vibrant displacement of bearing seat is intensified at the moment of lose power, but after a certain period rotor shaft libration changes. The test and analysis results help to understand the hydrodynamic performance of nuclear primary pump under lost of power accident, and provide the basic reference for safety evaluation. (authors)

  14. The benefits of conducting factory performance tests for main mine fans

    Energy Technology Data Exchange (ETDEWEB)

    Ray, R.E.Jr. [PB Americas Inc., New York, NY (United States); Gamble, G.A. [Clarage Twin City Fan Co., Akron, OH (United States)

    2010-07-01

    Axial flow fans used in underground mining are also commonly used in subway tunnel ventilation fans to provide an evacuation path during a tunnel fire emergency. The axial flow fans provide sufficient air velocity to the fire site to prevent backlayering of smoke against the incoming airflow. Since the tunnels are used by the public, advance testing of fans and motors is conducted to confirm that the equipment will perform as specified during a fire. This paper discussed some of the advantages derived from conducting fan factory tests for tunnel projects that would also apply to mining applications. It also described other benefits from testing that are unique to mining. External factors that may cause the fan performance to vary considerably from the predicted performance measured at the factory were also discussed. These included air density changes and system effects produced by poorly designed shaft configurations and fan inlet ductwork. 11 refs., 6 figs.

  15. A consistency test of white dwarf and main sequence ages: NGC 6791

    Directory of Open Access Journals (Sweden)

    Córsico A.H.

    2013-03-01

    Full Text Available NGC 6791 is an open cluster that it is so close to us that can be imaged down to very faint luminosities. The main sequence turn-off age (∼8 Gyr and the age derived from the cut-off of the white dwarf luminosity function (∼6 Gyr were found to be significantly different. Here we demonstrate that the origin of this age discrepancy lies in an incorrect evaluation of the white dwarf cooling ages, and we show that when the relevant physical separation processes are included in the calculation of white dwarf sequences both ages are coincident.

  16. Testing the structure of earthquake networks from multivariate time series of successive main shocks in Greece

    Science.gov (United States)

    Chorozoglou, D.; Kugiumtzis, D.; Papadimitriou, E.

    2018-06-01

    The seismic hazard assessment in the area of Greece is attempted by studying the earthquake network structure, such as small-world and random. In this network, a node represents a seismic zone in the study area and a connection between two nodes is given by the correlation of the seismic activity of two zones. To investigate the network structure, and particularly the small-world property, the earthquake correlation network is compared with randomized ones. Simulations on multivariate time series of different length and number of variables show that for the construction of randomized networks the method randomizing the time series performs better than methods randomizing directly the original network connections. Based on the appropriate randomization method, the network approach is applied to time series of earthquakes that occurred between main shocks in the territory of Greece spanning the period 1999-2015. The characterization of networks on sliding time windows revealed that small-world structure emerges in the last time interval, shortly before the main shock.

  17. The main directions of prospective cohort study of population living around the Semipalatinsk nuclear test site

    OpenAIRE

    ZHUNUSSOVA T.; GROSCHE B.; APSALIKOV K.; BELIKHINA T.; PIVINA L.; MULDAGALIEV T.

    2014-01-01

    In the paper we have presented the possibilities of prospective cohort study of health status in the radiation exposed population living around the Semipalatinsk nuclear test site. It was substantiated the necessity of international cooperation of scientists from Kazakhstan, Europe, Japan and the United States for long-term study of radiation effects for the people and the environment.

  18. Public views evening engine test of a Space Shuttle Main Engine

    Science.gov (United States)

    2001-01-01

    Over the past year, more than 20,000 people came to Stennis Space Center to witness the 'shake, rattle and roar' of one of the world's most sophisticated engines. Stennis Space Center in south Mississippi is NASA's lead center for rocket propulsion testing. StenniSphere, the visitor center for Stennis Space Center, hosted more than 250,000 visitors in its first year of operation. Of those visitors, 26.4 percent were from Louisiana.

  19. CATHENA analysis of CANDU 6 steam generators for steam main break at a remote location

    Energy Technology Data Exchange (ETDEWEB)

    Liu, T. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2009-07-01

    CATHENA (Canadian Algorithm for THErmalhydraulic Network Analysis) is a nonequilibrium, two-phase, two fluid network analysis code that has been in use for over two decades by various groups in Canada and around the world. It is the primary Thermalhydraulics network analysis tool used by Atomic Energy Canada Ltd. (AECL) in the design, safety and licensing analysis of power and research reactors as well as test facilities. In the thermalhydraulic model, the liquid and vapor phases may have different pressures, velocities, and temperatures. The objective of the present paper is to present the detailed modeling of a CANDU 6 Steam Generator (SG) using the transient, thermalhydraulics network code CATHENA. The model represents the secondary side, primary side and the main steam system including the main steam line up to the assumed break location. The present model is designed such that the transient pressure drops across Tube Support Plates (TSP) could be extracted. The resistances of degraded/fouled TSPs were modeled by using the increased/reduced flow area of the TSPs. CATHENA then calculates the flow resistance in two-phase flow based on the area contraction/expansion at the TSPs. Three sets of simulations were performed; one with the degraded steam generator data provided by the utility users, and the other two with waterlanced (cleaned with high pressure water jet) TSPs. One run assumed the flow area increased by 25 percent, the other run assumed the flow area increased by 50 percent. on the hot side of the SG. No significant changes in the break discharge rates were observed between the runs. However, the steam generator downcomer flow for the waterlanced case did not reverse during the blowdown as was calculated for the degraded case. As expected, the pressure drop across the TSPs was decreased in the waterlanced cases comparing with degraded cases. The CATHENA simulation provides estimates of the velocity, density, and quality in the tube bundle as well as

  20. Main boiler feed pump for fast breeder test reactor. Failure analysis and remedial measures

    International Nuclear Information System (INIS)

    Iyer, M.A.K.; Chande, S.K.; Raghuvir, A.D.; Baskar, S.; Kale, R.D.

    1994-01-01

    A small capacity ten stage 670 kw feed water pump is used for supplying feed water at a temperature of 190 deg C to a once through steam generator in the Fast Breeder Test Reactor at Kalpakkam. During preparatory heating up stage to commission the steam generator the pump suffered a severe loss of suction which resulted in failure of hydrostatic journal bearings and extensive damage to pump internals. This paper discusses the detailed mechanism of loss of suction, details of damage to the pump and various modifications carried out to prevent recurrence of the problem. (author). 4 refs., 3 figs., 2 tabs

  1. Analyzing and comparing the dynamic response of test reactor main workshop

    International Nuclear Information System (INIS)

    Wang Jiachun; Fu Jiyang; Cai Laizhong

    2001-01-01

    Analyzing soil-structure interaction is an important section in anti-seismic design and analysis of nuclear engineering. The factors that influence on the response of nuclear structures include the properties of earthquake, soil and structures. So the soil-structure interaction in the non-rock foundation is different from that in the surface free field. And the interaction must be considered under the anti-seismic design standard of test reactors. The FLUSH program and SASSI2000 are applied to dynamic analysis. Moreover, comparing the obtained data and diagrams draws some conclusions

  2. Radiation detectors for use in major public events: classification, requirements, main features, tests and lessons learned

    International Nuclear Information System (INIS)

    Souza, Elder Magalhães de

    2017-01-01

    Since September 11, 2001, we have entered a new terrorism era. The possibility of the use of lost/stolen radioactive materials increases the probability of a radiological threat. The real goal intended with the use of a Radiological Dispersal Device (RRD or dirty bomb) or a Radiation Exposure Device (RDE) could be psychological in nature. Panic in the venues and surrounding area would cause more deaths than the RDD itself, therefore these attempts could cause chaos, injury, fear and terror, the main target of terrorists. The response of the national authorities with the support and aid of the IAEA served as an increase of the capability of detection and identification of nuclear and radiological materials. But this response could not be limited only to the MPE, because if the country has radioactive or nuclear facilities they also should be considered in terms of theft, sabotage, illegal transfer, unauthorized access, and any other malicious acts. In 2007, Rio de Janeiro, received the first Brazilian Major Public Event in this new era. This was the first Brazilian operation which largely utilized detectors (personal radiations detectors -PRD- radiological identification detectors, -RID or RIID- and spectral radiations scanners, -backpacks-, HPGe detectors, car-borne and air-borne systems) to protect the venues, the athletes, the population and the environment. (author)

  3. Radiation detectors for use in major public events: classification, requirements, main features, tests and lessons learned

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Elder Magalhães de, E-mail: elder@ird.gov.br [Instituto de Radioproteção e Dosimetria (DIRAD/IRD/CNEN-RJ), Rio de Janeiro, RJ (Brazil). Div. de Radiometria

    2017-07-01

    Since September 11, 2001, we have entered a new terrorism era. The possibility of the use of lost/stolen radioactive materials increases the probability of a radiological threat. The real goal intended with the use of a Radiological Dispersal Device (RRD or dirty bomb) or a Radiation Exposure Device (RDE) could be psychological in nature. Panic in the venues and surrounding area would cause more deaths than the RDD itself, therefore these attempts could cause chaos, injury, fear and terror, the main target of terrorists. The response of the national authorities with the support and aid of the IAEA served as an increase of the capability of detection and identification of nuclear and radiological materials. But this response could not be limited only to the MPE, because if the country has radioactive or nuclear facilities they also should be considered in terms of theft, sabotage, illegal transfer, unauthorized access, and any other malicious acts. In 2007, Rio de Janeiro, received the first Brazilian Major Public Event in this new era. This was the first Brazilian operation which largely utilized detectors (personal radiations detectors -PRD- radiological identification detectors, -RID or RIID- and spectral radiations scanners, -backpacks-, HPGe detectors, car-borne and air-borne systems) to protect the venues, the athletes, the population and the environment. (author)

  4. Upgrade of the cryogenic infrastructure of SM18, CERN main test facility for superconducting magnets and RF cavities

    Science.gov (United States)

    Perin, A.; Dhalla, F.; Gayet, P.; Serio, L.

    2017-12-01

    SM18 is CERN main facility for testing superconducting accelerator magnets and superconducting RF cavities. Its cryogenic infrastructure will have to be significantly upgraded in the coming years, starting in 2019, to meet the testing requirements for the LHC High Luminosity project and for the R&D program for superconducting magnets and RF equipment until 2023 and beyond. This article presents the assessment of the cryogenic needs based on the foreseen test program and on past testing experience. The current configuration of the cryogenic infrastructure is presented and several possible upgrade scenarios are discussed. The chosen upgrade configuration is then described and the characteristics of the main newly required cryogenic equipment, in particular a new 35 g/s helium liquefier, are presented. The upgrade implementation strategy and plan to meet the required schedule are then described.

  5. Empirical tests of pre-main-sequence stellar evolution models with eclipsing binaries

    Science.gov (United States)

    Stassun, Keivan G.; Feiden, Gregory A.; Torres, Guillermo

    2014-06-01

    We examine the performance of standard pre-main-sequence (PMS) stellar evolution models against the accurately measured properties of a benchmark sample of 26 PMS stars in 13 eclipsing binary (EB) systems having masses 0.04-4.0 M⊙ and nominal ages ≈1-20 Myr. We provide a definitive compilation of all fundamental properties for the EBs, with a careful and consistent reassessment of observational uncertainties. We also provide a definitive compilation of the various PMS model sets, including physical ingredients and limits of applicability. No set of model isochrones is able to successfully reproduce all of the measured properties of all of the EBs. In the H-R diagram, the masses inferred for the individual stars by the models are accurate to better than 10% at ≳1 M⊙, but below 1 M⊙ they are discrepant by 50-100%. Adjusting the observed radii and temperatures using empirical relations for the effects of magnetic activity helps to resolve the discrepancies in a few cases, but fails as a general solution. We find evidence that the failure of the models to match the data is linked to the triples in the EB sample; at least half of the EBs possess tertiary companions. Excluding the triples, the models reproduce the stellar masses to better than ∼10% in the H-R diagram, down to 0.5 M⊙, below which the current sample is fully contaminated by tertiaries. We consider several mechanisms by which a tertiary might cause changes in the EB properties and thus corrupt the agreement with stellar model predictions. We show that the energies of the tertiary orbits are comparable to that needed to potentially explain the scatter in the EB properties through injection of heat, perhaps involving tidal interaction. It seems from the evidence at hand that this mechanism, however it operates in detail, has more influence on the surface properties of the stars than on their internal structure, as the lithium abundances are broadly in good agreement with model predictions. The

  6. A Single Test Combining Blood Markers and Elastography is More Accurate Than Other Fibrosis Tests in the Main Causes of Chronic Liver Diseases.

    Science.gov (United States)

    Ducancelle, Alexandra; Leroy, Vincent; Vergniol, Julien; Sturm, Nathalie; Le Bail, Brigitte; Zarski, Jean Pierre; Nguyen Khac, Eric; Salmon, Dominique; de Ledinghen, Victor; Calès, Paul

    2017-08-01

    International guidelines suggest combining a blood test and liver stiffness measurement (LSM) to stage liver fibrosis in chronic hepatitis C (CHC) and non-alcoholic fatty liver disease (NAFLD). Therefore, we compared the accuracies of these tests between the main etiologies of chronic liver diseases. Overall, 1968 patients were included in 5 etiologies: CHC: 698, chronic hepatitis B: 152, human immunodeficiency virus/CHC: 628, NAFLD: 225, and alcoholic liver disease (ALD): 265. Sixteen tests [13 blood tests, LSM (Fibroscan), 2 combined: FibroMeters] were evaluated. References were Metavir staging and CHC etiology. Accuracy was evaluated mainly with the Obuchowski index (OI) and accessorily with area under the receiver operating characteristics (F≥2, F≥3, cirrhosis). OIs in CHC were: FibroMeters: 0.812, FibroMeters: 0.785 to 0.797, Fibrotest: 0.762, CirrhoMeters: 0.756 to 0.771, LSM: 0.754, Hepascore: 0.752, FibroMeter: 0.750, aspartate aminotransferase platelet ratio index: 0.742, Fib-4: 0.741. In other etiologies, most tests had nonsignificant changes in OIs. In NAFLD, CHC-specific tests were more accurate than NAFLD-specific tests. The combined FibroMeters had significantly higher accuracy than their 2 constitutive tests (FibroMeters and LSM) in at least 1 diagnostic target in all etiologies, except in ALD where LSM had the highest OI, and in 3 diagnostic targets (OIs and 2 area under the receiver operating characteristics) in CHC and NAFLD. Some tests developed in CHC outperformed other tests in their specific etiologies. Tests combining blood markers and LSM outperformed single tests, validating recent guidelines and extending them to main etiologies. Noninvasive fibrosis evaluation can thus be simplified in the main etiologies by using a unique test: either LSM alone, especially in ALD, or preferably combined to blood markers.

  7. Influence of lubricant on the pitting capacity of gears. Comparison and discussion of main pitting test methods

    Energy Technology Data Exchange (ETDEWEB)

    Hoehn, Bernd-Robert; Oster, Peter; Tobie, Thomas; Hergesell, Maria [Forschungsstelle fuer Zahnraeder und Getriebebau, Technische Univ. Muenchen, Garching (Germany)

    2009-07-01

    The lubricant in a gearbox should provide a separating lubricating film between the meshing teeth. Furthermore its main task is the evacuation of the developing heat. Hence the lubricant is also an important influence parameter on the pitting life time of gearsets. Due to the constantly rising demands for higher power transmission of gearsets, also higher demands are made against the performance of the lubricants. In order to quantify the capability and to choose the appropriate lubricant for each application, reliable testing methods are necessary. Therefor today different testing methods are available. Two of the main pitting testing methods are the Standard Pittingtest according to FVA 2/IV [2] and its advancement, the Practice Relevant Pittingtest according to FVA 371 [4]. For the Practice Relevant Pittingtest test gears with profile modification and lengthwise crowning are used, which are superfinished. For the introduction of the Practice Relevant Pittingtest into industrial application a round robin test was performed in co-operation with FVA and DGMK. This paper is discussing the today usual testing methods regarding the pitting load capacity of lubricants and their characteristics in comparison. Furthermore the results of the round robin test and further example results, including pictures of typical pittings, are presented. In addition it is demonstrated, how to deal with the results. On the basis of test experience, knowledge from research projects and other examples typical questions concerning the different testing methods are discussed. Thereby essential requirements for reliable and reproducible results are pointed out and possible limits for the application of the test methods are represented. (orig.)

  8. Application to a commercial plant of PKL results in a main steam line break is shown; Aplicacion a una planta comercial de los resultados de PKL en una rotura de la Linea de Vapor Principal

    Energy Technology Data Exchange (ETDEWEB)

    Villanueva, J. F.; Carlos, S.; Sanchez, F.; Martorell, S.

    2012-07-01

    The applicability of the results of the experimental facility and its implications from the point of view of security of the same. have been studied Likewise within the experiment possible corrective measures for implementation to keep the plant in stable situation against the accidental happening.

  9. First-principles simulation and comparison with beam tests for transverse instabilities and damper performance in the Fermilab Main Injector

    International Nuclear Information System (INIS)

    Nicklaus, Dennis; Foster, G.William; Kashikhin, Vladimir

    2005-01-01

    An end-to-end performance calculation and comparison with beam tests was performed for the bunch-by-bunch digital transverse damper in the Fermilab Main Injector. Time dependent magnetic wakefields responsible for ''Resistive Wall'' transverse instabilities in the Main Injector were calculated with OPERA-2D using the actual beam pipe and dipole magnet lamination geometry. The leading order dipole component was parameterized and used as input to a bunch-by-bunch simulation which included the filling pattern and injection errors experienced in high-intensity operation of the Main Injector. The instability growth times, and the spreading of the disturbance due to newly misinjected batches was compared between simulations and beam data collected by the damper system. Further simulation models the effects of the damper system on the beam

  10. Reliability and main findings of the FEES-Tensilon Test in patients with myasthenia gravis and dysphagia.

    Science.gov (United States)

    Im, Sun; Suntrup-Krueger, Sonja; Colbow, Sigrid; Sauer, Sonja; Claus, Inga; Meuth, Sven G; Dziewas, Rainer; Warnecke, Tobias

    2018-05-26

    Diagnosis of pharyngeal dysphagia caused by myasthenia gravis (MG) based on clinical examination alone is often challenging. Flexible endoscopic evaluation of swallowing (FEES) combined with Tensilon (edrophonium) application, referred to as the FEES-Tensilon Test, was developed to improve diagnostic accuracy and to detect the main symptoms of pharyngeal dysphagia in MG. Here we investigated inter- and intra-rater reliability of the FEES-Tensilon Test and analyzed the main endoscopic findings. Four experienced raters reviewed a total of 20 FEES-Tensilon-Test videos in randomized order. Residue severity was graded at 4 different pharyngeal spaces before and after Tensilon administration. All interpretations were performed twice per rater, 4 weeks apart (a total of 160 scorings). Intra-rater test-retest reliability and inter-rater reliability levels were calculated. The most frequent FEES findings in MG patients before Tensilon application were prominent residues of semi solids spread all over the hypopharynx in varying locations. The reliability level in the interpretation of the FEES-Tensilon test was excellent regardless of the raters' profession or years of experience with FEES. All 4 raters showed high inter- and intra- reliability levels in interpreting the FEES-Tensilon Test based on residue clearance (kappa=0.922, 0.981). Degree of residue normalization in the vallecular space after Tensilon application showed the highest inter- and intra-rater reliability level (kappa=0.863, 0.957) followed by the epiglottis (kappa=0.813, 0.946) and pyriform sinuses (kappa=0.836, 0.929). Interpretation of the FEES-Tensilon Test based on residue severity and degree of Tensilon clearance, especially in the vallecular space, is consistent and reliable. This article is protected by copyright. All rights reserved. This article is protected by copyright. All rights reserved.

  11. 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.)

  12. Influences on domestic well water testing behavior in a Central Maine area with frequent groundwater arsenic occurrence.

    Science.gov (United States)

    Flanagan, Sara V; Marvinney, Robert G; Zheng, Yan

    2015-02-01

    In 2001 the Environmental Protection Agency (EPA) adopted a new standard for arsenic (As) in drinking water of 10 μg/L, replacing the old standard of 50 μg/L. However, for the 12% of the U.S. population relying on unregulated domestic well water, including half of the population of Maine, it is solely the well owner's responsibility to test and treat the water. A mailed household survey was implemented in January 2013 in 13 towns of Central Maine with the goal of understanding the population's testing and treatment practices and the key behavior influencing factors in an area with high well-water dependency and frequent natural groundwater As. The response rate was 58.3%; 525 of 900 likely-delivered surveys to randomly selected addresses were completed. Although 78% of the households reported that their well has been tested, half of it was more than 5 years ago. Among the 58.7% who believe they have tested for As, most do not remember the results. Better educated, higher income homeowners who more recently purchased their homes are most likely to have included As when last testing. While households agree that water and As-related health risks can be severe, they feel low personal vulnerability and there are low testing norms overall. Significant predictors of including As when last testing include: having knowledge that years of exposure increases As-related health risks (risk knowledge), knowing who to contact to test well water (action knowledge), believing that regular testing does not take too much time (instrumental attitude), and having neighbors who regularly test their water (descriptive norm). Homeowners in As-affected communities have the tendency to underestimate their As risks compared to their neighbors. The reasons for this optimistic bias require further study, but low testing behaviors in this area may be due to the influence of a combination of norm, ability, and attitude factors and barriers. Copyright © 2014 Elsevier B.V. All rights reserved.

  13. Hydraulic model of the steam-lines network of the Cerro Prieto, B.C., geothermal field; Modelo hidraulico de la red de vaporductos del campo geotermico de Cerro Prieto, B.C.

    Energy Technology Data Exchange (ETDEWEB)

    Salaices, E; Garcia, A; Martinez J I; Ovando, R; Cecenas, M; Hernandez A F [Instituto de Investigaciones Electricas, Cuernavaca, Morelos (Mexico)]. E-mail: salaices@iie.org.mx; Canchola, I; Mora, O; Miranda, C; Herandez, M; Lopez, S; Murillo, I [Comision Federal de Electricidad, B.C (Mexico)

    2007-01-15

    The steam-line network of the Cerro Prieto geothermal field is composed of 184 wells, and 162 of the wells are integrated and connected by pipes. Thirteen power units, with an installed electrical capacity of 720 MW, are fed by that network. The network length is 120 km, including pipes of several diameters with branches and interconnections. The extension and complexity of the steam-line system make it difficult to analyze the transport and supply of steam to the power plants. For that it was necessary to have a tool capable of analyzing the system and the performance of the network as a whole, as well as the direction and flow volumes in each part of the system. In this paper, a hydraulic model of the Cerro Prieto steam-line network is presented. The model can determine the performance of the whole network by quantifying the pressure drops, flows and heat losses of the components. The model analyses the consequences of changes in operating conditions, steam production, maintenance activities and design (such as the integration of new wells). The model was developed using PIPEPHASE 9.0, a numeric simulator of multi-phase flow in steady state with heat transfer. It is used to model systems and pipe networks for steam- and condensate-transport. [Spanish] La red de vaporductos del campo geotermico de Cerro Prieto esta compuesta por un conjunto de 184 pozos, de los cuales 162 son pozos integrados, interconectados entre si a traves de una red de tuberias. Por medio de esta red se alimentan 13 unidades generadoras de electricidad con una capacidad total instalada de 720 MWe. La red tiene una longitud aproximada de 120 kilometros y esta compuesta por tuberias de diferentes diametros, ramales, interconexiones, etc. La complejidad y extension del sistema de vaporductos hace muy dificil el analisis del transporte y suministro de vapor a las plantas generadoras. Lo anterior creo la necesidad de contar con una herramienta que ayudara en el analisis del sistema con el fin de

  14. Testing Scaling Relations for Solar-like Oscillations from the Main Sequence to Red Giants Using Kepler Data

    DEFF Research Database (Denmark)

    Huber, D.; Bedding, T.R.; Stello, D.

    2011-01-01

    ), and oscillation amplitudes. We show that the difference of the Δν-νmax relation for unevolved and evolved stars can be explained by different distributions in effective temperature and stellar mass, in agreement with what is expected from scaling relations. For oscillation amplitudes, we show that neither (L/M) s......We have analyzed solar-like oscillations in ~1700 stars observed by the Kepler Mission, spanning from the main sequence to the red clump. Using evolutionary models, we test asteroseismic scaling relations for the frequency of maximum power (νmax), the large frequency separation (Δν...... scaling nor the revised scaling relation by Kjeldsen & Bedding is accurate for red-giant stars, and demonstrate that a revised scaling relation with a separate luminosity-mass dependence can be used to calculate amplitudes from the main sequence to red giants to a precision of ~25%. The residuals show...

  15. Ergonomic Investigation On The Layout And Design Of The Main Control Room Of (MCR)Reactor Thermalhydraulic Testing Loop

    International Nuclear Information System (INIS)

    DARLlS; WIDAGDO, SUHARYO

    2000-01-01

    Ergonomics investigation on the layout and design of the reactor thermalhydraulic testing loop main control room has been done. This reason is needed to be done as the primary step for evaluating of operator workload. The operator work load be influence on the operator performance, and finally would influencing the installation operation safety. Generally, the factors that is influencing on operator performance are the layout and design of MCR and its supporting physical environments factors for instance lighting, noising and climatic condition respectively. From investigation had been done, cod be identified that ergonomics point of view not implemented yet on the main control console design, especially on the alarm panel, and also found a little bit brightness problem. Otherwise the temperature and noise room are still in the tolerance boundary

  16. BWR 2 % main recirculation line break LOCA tests RUNs 915 and 920 without HPCS in ROSA-III program

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Nakamura, Hideo; Anoda, Yoshinari; Kumamaru, Hiroshige; Yonomoto, Taisuke; Koizumi, Yasuo; Tasaka, Kanji

    1987-03-01

    This report presents the experimental results of BWR LOCA integral tests, RUNs 915 and 920, which are performed in the ROSA-III program simulating 2 % main recirculation line break LOCA tests with and without pressure control system operation. The ROSA-III test facility simulates a BWR system with volume scale of 1/424 and has four half-length electrically heated fuel bundles, two active recirculation loops, four types of ECCS's, and steam and feedwater systems. The report presents (1) the experimental results of 2 % small break LOCA phanomena in the ROSA-III system and (2) the effects of the pressure control system on the LOCA phenomena. The pressure control system contributed to (A) prevent bulk flashing in the early blowdown phase, (B) early closure of MSIV by L2 level trip, (C) early actuation of ADS by L1 level trip. However, the core thermal responses of the two tests were similar because of the similar mass inventory in PV after the ADS actuation in both tests. (author)

  17. Characterization of Pump-Induced Acoustics in Space Launch System Main Propulsion System Liquid Hydrogen Feedline Using Airflow Test Data

    Science.gov (United States)

    Eberhart, C. J.; Snellgrove, L. M.; Zoladz, T. F.

    2015-01-01

    High intensity acoustic edgetones located upstream of the RS-25 Low Pressure Fuel Turbo Pump (LPFTP) were previously observed during Space Launch System (STS) airflow testing of a model Main Propulsion System (MPS) liquid hydrogen (LH2) feedline mated to a modified LPFTP. MPS hardware has been adapted to mitigate the problematic edgetones as part of the Space Launch System (SLS) program. A follow-on airflow test campaign has subjected the adapted hardware to tests mimicking STS-era airflow conditions, and this manuscript describes acoustic environment identification and characterization born from the latest test results. Fluid dynamics responsible for driving discrete excitations were well reproduced using legacy hardware. The modified design was found insensitive to high intensity edgetone-like discretes over the bandwidth of interest to SLS MPS unsteady environments. Rather, the natural acoustics of the test article were observed to respond in a narrowband-random/mixed discrete manner to broadband noise thought generated by the flow field. The intensity of these responses were several orders of magnitude reduced from those driven by edgetones.

  18. TESTING SCALING RELATIONS FOR SOLAR-LIKE OSCILLATIONS FROM THE MAIN SEQUENCE TO RED GIANTS USING KEPLER DATA

    Energy Technology Data Exchange (ETDEWEB)

    Huber, D.; Bedding, T. R.; Stello, D. [Sydney Institute for Astronomy (SIfA), School of Physics, University of Sydney, NSW 2006 (Australia); Hekker, S. [Astronomical Institute ' Anton Pannekoek' , University of Amsterdam, Science Park 904, 1098 XH Amsterdam (Netherlands); Mathur, S. [High Altitude Observatory, NCAR, P.O. Box 3000, Boulder, CO 80307 (United States); Mosser, B. [LESIA, CNRS, Universite Pierre et Marie Curie, Universite Denis, Diderot, Observatoire de Paris, 92195 Meudon cedex (France); Verner, G. A.; Elsworth, Y. P.; Hale, S. J.; Chaplin, W. J. [School of Physics and Astronomy, University of Birmingham, Birmingham B15 2TT (United Kingdom); Bonanno, A. [INAF Osservatorio Astrofisico di Catania (Italy); Buzasi, D. L. [Eureka Scientific, 2452 Delmer Street Suite 100, Oakland, CA 94602-3017 (United States); Campante, T. L. [Centro de Astrofisica da Universidade do Porto, Rua das Estrelas, 4150-762 Porto (Portugal); Kallinger, T. [Department of Physics and Astronomy, University of British Columbia, Vancouver (Canada); Silva Aguirre, V. [Max-Planck-Institut fuer Astrophysik, Karl-Schwarzschild-Str. 1, 85748 Garching (Germany); De Ridder, J. [Instituut voor Sterrenkunde, K.U.Leuven (Belgium); Garcia, R. A. [Laboratoire AIM, CEA/DSM-CNRS, Universite Paris 7 Diderot, IRFU/SAp, Centre de Saclay, 91191, Gif-sur-Yvette (France); Appourchaux, T. [Institut d' Astrophysique Spatiale, UMR 8617, Universite Paris Sud, 91405 Orsay Cedex (France); Frandsen, S. [Danish AsteroSeismology Centre (DASC), Department of Physics and Astronomy, Aarhus University, DK-8000 Aarhus C (Denmark); Houdek, G., E-mail: dhuber@physics.usyd.edu.au [Institute of Astronomy, University of Vienna, 1180 Vienna (Austria); and others

    2011-12-20

    We have analyzed solar-like oscillations in {approx}1700 stars observed by the Kepler Mission, spanning from the main sequence to the red clump. Using evolutionary models, we test asteroseismic scaling relations for the frequency of maximum power ({nu}{sub max}), the large frequency separation ({Delta}{nu}), and oscillation amplitudes. We show that the difference of the {Delta}{nu}-{nu}{sub max} relation for unevolved and evolved stars can be explained by different distributions in effective temperature and stellar mass, in agreement with what is expected from scaling relations. For oscillation amplitudes, we show that neither (L/M){sup s} scaling nor the revised scaling relation by Kjeldsen and Bedding is accurate for red-giant stars, and demonstrate that a revised scaling relation with a separate luminosity-mass dependence can be used to calculate amplitudes from the main sequence to red giants to a precision of {approx}25%. The residuals show an offset particularly for unevolved stars, suggesting that an additional physical dependency is necessary to fully reproduce the observed amplitudes. We investigate correlations between amplitudes and stellar activity, and find evidence that the effect of amplitude suppression is most pronounced for subgiant stars. Finally, we test the location of the cool edge of the instability strip in the Hertzsprung-Russell diagram using solar-like oscillations and find the detections in the hottest stars compatible with a domain of hybrid stochastically excited and opacity driven pulsation.

  19. Closure Report for Corrective Action Unit 425: Area 9 Main Lake Construction Debris Disposal Area, Tonopah Test Range, Nevada

    Energy Technology Data Exchange (ETDEWEB)

    K. B. Campbell

    2003-03-01

    Corrective Action Unit (CAU) 425 is located on the Tonopah Test Range, approximately 386 kilometers (240 miles) northwest of Las Vegas, Nevada. CAU 425 is listed in the Federal Facility Agreement and Consent Order (FFACO, 1996) and is comprised of one Corrective Action Site (CAS). CAS 09-08-001-TA09 consisted of a large pile of concrete rubble from the original Hard Target and construction debris associated with the Tornado Rocket Sled Tests. CAU 425 was closed in accordance with the FFACO and the Nevada Division of Environmental Protection-approved Streamlined Approach for Environmental Restoration Plan for CAU 425: Area 9 Main Lake Construction Debris Disposal Area, Tonopah Test Range, Nevada (U.S. Department of Energy, Nevada Operations Office, 2002). CAU 425 was closed by implementing the following corrective actions: The approved corrective action for this unit was clean closure. Closure activities included: (1) Removal of all the debris from the site. (2) Weighing each load of debris leaving the job site. (3) Transporting the debris to the U.S. Air Force Construction Landfill for disposal. (4) Placing the radioactive material in a U.S. Department of Transportation approved container for proper transport and disposal. (5) Transporting the radioactive material to the Nevada Test Site for disposal. (6) Regrading the job site to its approximate original contours/elevation.

  20. Test results for cables used in nuclear power plants by a new environmental testing method

    Energy Technology Data Exchange (ETDEWEB)

    Handa, Katsue; Fujimura, Shun-ichi; Hayashi, Toshiyasu; Takano, Keiji; Oya, Shingo

    1982-12-01

    In the nuclear power plants using PWRs or BWRs in Japan, environmental tests are provided, in which simulated LOCA conditions are considered so as to conform with Japanese conditions, and many cables which passed these tests are presently employed. Lately, the new environmental testing, in which a credible accident called MSLB (main steam line breakage) is taken into account, is investigated in PWR nuclear power plants, besides LOCA. This paper reports on the results of evaluating some PWR cables for this new environmental testing conditions. The several cables tested were selected out of PH cables (fire-retardant, ethylene propylene rubber insulated, chlorosulfonated polyethylene sheathed cables) as the cables for safety protecting circuits and to be used in containment vessels where the cables are to be exposed to severe environmental test conditions of 2 x 10/sup 8/ Rad ..gamma..-irradiation and simulated LOCA. All these cables have been accepted after the vertical tray burning test provided in the IEEE Standard 383. The new testing was carried out by sequentially applying thermal deterioration, ..gamma..-irradiation, and the exposure to steam (twice 300 s exposures to 190 deg C superheated steam). After completing each step, tensile strength, elongation, insulation resistance and breakdown voltage were measured, respectively. Every cable tested showed satisfactory breakdown voltage after the exposure to steam, thus it was decided to be acceptable. In future, it is required to investigate the influence of the rate of temperature rise on the cable to be tested in MSLB simulation.

  1. Test results for cables used in nuclear power plants by a new environmental testing method

    International Nuclear Information System (INIS)

    Handa, Katsue; Fujimura, Shun-ichi; Hayashi, Toshiyasu; Takano, Keiji; Oya, Shingo

    1982-01-01

    In the nuclear power plants using PWRs or BWRs in Japan, environmental tests are provided, in which simulated LOCA conditions are considered so as to conform with Japanese conditions, and many cables which passed these tests are presently employed. Lately, the new environmental testing, in which a credible accident called MSLB (main steam line breakage) is taken into account, is investigated in PWR nuclear power plants, besides LOCA. This paper reports on the results of evaluating some PWR cables for this new environmental testing conditions. The several cables tested were selected out of PH cables (fire-retardant, ethylene propylene rubber insulated, chlorosulfonated polyethylene sheathed cables) as the cables for safety protecting circuits and to be used in containment vessels where the cables are to be exposed to severe environmental test conditions of 2 x 10 8 Rad γ-irradiation and simulated LOCA. All these cables have been accepted after the vertical tray burning test provided in the IEEE Standard 383. The new testing was carried out by sequentially applying thermal deterioration, γ-irradiation, and the exposure to steam (twice 300 s exposures to 190 deg C superheated steam). After completing each step, tensile strength, elongation, insulation resistance and breakdown voltage were measured, respectively. Every cable tested showed satisfactory breakdown voltage after the exposure to steam, thus it was decided to be acceptable. In future, it is required to investigate the influence of the rate of temperature rise on the cable to be tested in MSLB simulation. (Wakatsuki, Y.)

  2. Development of 20 kW input power coupler for 1.3 GHz ERL main linac. Component test at 30 kW IOT test stand

    International Nuclear Information System (INIS)

    Sakai, Hiroshi; Umemori, Kensei; Sakanaka, Shogo; Takahashi, Takeshi; Furuya, Takaaki; Shinoe, Kenji; Ishii, Atsushi; Nakamura, Norio; Sawamura, Masaru

    2009-01-01

    We started to develop an input coupler for a 1.3 GHz ERL superconducting cavity. Required input power is about 20 kW for the cavity acceleration field of 20 MV/m and the beam current of 100 mA in energy recovery operation. The input coupler is designed based on the STF-BL input coupler and some modifications are applied to the design for the CW 20 kW power operation. We fabricated input coupler components such as ceramic windows and bellows and carried out the high-power test of the components by using a 30 kW IOT power source and a test stand constructed for the highpower test. In this report, we mainly describe the results of the high-power test of ceramic window and bellows. (author)

  3. Results of the 1986 NASA/FAA/DFVLR main rotor test entry in the German-Dutch wind tunnel (DNW)

    Science.gov (United States)

    Brooks, Thomas F.; Martin, Ruth M.

    1987-10-01

    An acoustics test of a 40%-scale MBB BO-105 helicopter main rotor was conducted in the Deutsch-Niederlandischer Windkanal (DNW). The research, directed by NASA Langley Research Center, concentrated on the generation and radiation of broadband noise and impulsive blade-vortex interaction (BVI) noise over ranges of pertinent rotor operational envelopes. Both the broadband and BVI experimental phases are reviewed, along with highlights of major technical results. For the broadband portion, significant advancement is the demonstration of the accuracy of prediction methods being developed for broadband self noise, due to boundary layer turbulence. Another key result is the discovery of rotor blade-wake interaction (BWI) as an important contributor to mid frequency noise. Also the DNW data are used to determine for full scale helicopters the relative importance of the different discrete and broadband noise sources. For the BVI test portion, a comprehensive data base documents the BVI impulsive noise character and directionality as functions of rotor flight conditions. The directional mapping of BVI noise emitted from the advancing side as well as the retreating side of the rotor constitutes a major advancement in the understanding of this dominant discrete mechanism.

  4. Evaluation of reflooding effects on an overheated boiling water reactor core in a small steam-line break accident using MAAP, MELCOR, and SCDAP/RELAP5 computer codes

    International Nuclear Information System (INIS)

    Lindholm, I.; Pekkarinen, E.; Sjoevall, H.

    1995-01-01

    Selected core reflooding situations were investigated in the case of a Finnish boiling water reactor with three severe accident analysis computer codes (MAAP, MELCOR, and SCDAP/RELAP5). The unmitigated base case accident scenario was a 10% steam-line break without water makeup to the reactor pressure vessel initially. The pumping of water to the core was started with the auxiliary feed water system when the maximum fuel cladding temperature reached 1,500 K. The auxiliary feedwater system pumps water (temperature 303 K) through the core spray spargers (core spray) on the top of the core and through feedwater nozzles to the downcomer (downcomer injection). The scope of the study was restricted to cases where the overheated core was still geometrically intact at the start of the reflooding. The following different core reflooding situations were investigated: (1) auxiliary feedwater injection to core spray (45 kg/s); (2) auxiliary feedwater injection to downcomer (45 kg/s); (3) auxiliary feedwater injection to downcomer (45 kg/s) and to core spray (45 kg/s); (4) no reflooding of the core. All the three codes predicted debris formation after the water addition, and in all MAAP and MELCOR reflooding results the core was quenched. The major difference between the code predictions was in the amount of H 2 produced, though the trends in H 2 production were similar. Additional steam production during the quenching process accelerated the oxidation in the unquenched parts of the core. This result is in accordance with several experimental observations

  5. Design, fabrication and test of a liquid hydrogen titanium honeycomb cryogenic test tank for use as a reusable launch vehicle main propellant tank

    Science.gov (United States)

    Stickler, Patrick B.; Keller, Peter C.

    1998-01-01

    Reusable Launch Vehicles (RLV's) utilizing LOX\\LH2 as the propellant require lightweight durable structural systems to meet mass fraction goals and to reduce overall systems operating costs. Titanium honeycomb sandwich with flexible blanket TPS on the windward surface is potentially the lightest-weight and most operable option. Light weight is achieved in part because the honeycomb sandwich tank provides insulation to its liquid hydrogen contents, with no need for separate cryogenic insulation, and in part because the high use temperature of titanium honeycomb reduces the required surface area of re-entry thermal protection systems. System operability is increased because TPS needs to be applied only to surfaces where temperatures exceed approximately 650 K. In order to demonstrate the viability of a titanium sandwich constructed propellant tank, a technology demonstration program was conducted including the design, fabrication and testing of a propellant tank-TPS system. The tank was tested in controlled as well as ambient environments representing ground hold conditions for a RLV main propellant tank. Data collected during each test run was used to validate predictions for air liquefaction, outside wall temperature, boil-off rates, frost buildup and its insulation effects, and the effects of placing a thermal protection system blanket on the external surface. Test results indicated that titanium honeycomb, when used as a RLV propellant tank material, has great promise as a light-weight structural system.

  6. Signal analysis of steam line acoustics

    International Nuclear Information System (INIS)

    Martin, C. Samuel

    2003-01-01

    The vibration of nuclear steam piping is usually associated with pressure fluctuations emanating from flow disturbances such as steam generator nozzles, bends, or other pipe fittings. Flow separation at pipe tees and within steam chest manifolds or headers generate pressure fluctuations that propagate both upstream to steam generators as well as downstream to the steam turbine. Steady-state acoustic oscillations at various frequencies occur within the piping, possibly exciting structural vibrations. This paper focuses on the assessment of the origin of the disturbances using signal analyses of two dynamic pressure recordings from pressure transducers located along straight runs in the steam piping. The technique involves performing the cross spectrum to two dynamic pressure signals in piping between (1) the steam generator and steam chest header, and (2) between the header and steam turbine outlet. If, at a specified frequency, no causality occurs between the two signals then the cross spectra magnitude will be negligible. Of interest here is the value of the phase between the two signals for frequencies for which the magnitude of the cross spectrum is not negligible. It is shown in the paper that the direction of the dominant waves at all frequencies can be related to the phase angle from the cross spectrum. It has to be realized that pressure waves emanating from one source such as a steam generator will propagate along uniform steam pipes with little transformation or attenuation, but will be reflected at fittings and at inlets and outlets. Hence, the eventual steady-state time record at a given location in the piping is a result of not only the disturbance, but also reflections of earlier pulsations. Cross-spectral analyses has been employed to determine the direction of the dominant acoustic waves in the piping for various frequencies for which there are signals. To prove the technique, synthetic spectra are generated comprised of harmonic waves moving both upstream and downstream between two locations. The method shows that the direction of the larger magnitude wave -- whether propagating upstream or downstream -- is directly related to the slope of the unwrapped phase angle versus frequency correlation. Indeed, the slope of this line can be related to the acoustic velocity of the wave. The method is then applied to dynamic pressure recordings obtained in a nuclear steam system. Plots of cross-spectra phase versus frequency taken in straight runs of steam piping yield correlations that are nearly linear, and, moreover, the slope of the line is closely related to the acoustic velocity at the corresponding steam pressure. (author)

  7. Reliability Verification of DBE Environment Simulation Test Facility by using Statistics Method

    International Nuclear Information System (INIS)

    Jang, Kyung Nam; Kim, Jong Soeg; Jeong, Sun Chul; Kyung Heum

    2011-01-01

    In the nuclear power plant, all the safety-related equipment including cables under the harsh environment should perform the equipment qualification (EQ) according to the IEEE std 323. There are three types of qualification methods including type testing, operating experience and analysis. In order to environmentally qualify the safety-related equipment using type testing method, not analysis or operation experience method, the representative sample of equipment, including interfaces, should be subjected to a series of tests. Among these tests, Design Basis Events (DBE) environment simulating test is the most important test. DBE simulation test is performed in DBE simulation test chamber according to the postulated DBE conditions including specified high-energy line break (HELB), loss of coolant accident (LOCA), main steam line break (MSLB) and etc, after thermal and radiation aging. Because most DBE conditions have 100% humidity condition, in order to trace temperature and pressure of DBE condition, high temperature steam should be used. During DBE simulation test, if high temperature steam under high pressure inject to the DBE test chamber, the temperature and pressure in test chamber rapidly increase over the target temperature. Therefore, the temperature and pressure in test chamber continue fluctuating during the DBE simulation test to meet target temperature and pressure. We should ensure fairness and accuracy of test result by confirming the performance of DBE environment simulation test facility. In this paper, in order to verify reliability of DBE environment simulation test facility, statistics method is used

  8. Low priority main reason not to participate in a colorectal cancer screening program with a faecal occult blood test

    NARCIS (Netherlands)

    van Rijn, A. F.; van Rossum, L. G. M.; Deutekom, M.; Laheij, R. J. F.; Fockens, P.; Bossuyt, P. M. M.; Dekker, E.; Jansen, J. B. M. J.

    2008-01-01

    Compared with screening programs for breast and cervical cancer, reported participation rates for colorectal cancer (CRC) screening are low. The effectiveness of a screening program is strongly influenced by the participation rate. The aim of this study was to investigate the main reasons not to

  9. Low priority main reason not to participate in a colorectal cancer screening program with a faecal occult blood test.

    NARCIS (Netherlands)

    Rijn, A.F. van; Rossum, L.G.M. van; Deutekom, M.; Laheij, R.J.F.; Fockens, P.; Bossuyt, P.M.; Dekker, E. den; Jansen, J.B.M.J.

    2008-01-01

    BACKGROUND: Compared with screening programs for breast and cervical cancer, reported participation rates for colorectal cancer (CRC) screening are low. The effectiveness of a screening program is strongly influenced by the participation rate. The aim of this study was to investigate the main

  10. Field Demonstration of Innovative Leak Detection/Location in Conjunction with Pipe Wall Thickness Testing for Water Mains

    Science.gov (United States)

    The U.S. Environmental Protection Agency (EPA) sponsored a large-scale field demonstration of innovative leak detection/location and condition assessment technologies on a 76-year old, 2,000-ft long, cement-lined, 24-in. cast iron water main in Louisville, KY from July through Se...

  11. Testing the main prediction of the Interpersonal Theory of Suicide in a representative sample of the German general population.

    Science.gov (United States)

    Glaesmer, Heide; Hallensleben, Nina; Forkmann, Thomas; Spangenberg, Lena; Kapusta, Nestor; Teismann, Tobias

    2017-03-15

    To evaluate the main prediction of the Interpersonal Theory of Suicide (IPTS): 3-way-interaction of perceived burdensomeness (PB), thwarted belongingness (TB), and acquired capability (AC) for the prediction of suicidal behavior in a representative population sample. A total of 2513 participants completed measures of suicidal behavior, TB, PB, acquired capability (AC-FAD), and symptoms of depression and anxiety. The two-way-interaction of TB and PB, and the three-way interaction of TB, PB and AC-FAD predict suicidality. Given the cross-sectional nature of the data, conclusions on causality should be handled carefully. The main prediction of the IPTS has been proven in a general population sample. Copyright © 2017 Elsevier B.V. All rights reserved.

  12. Fission products and nuclear fuel behaviour under severe accident conditions part 1: Main lessons learnt from the first VERDON test

    Science.gov (United States)

    Pontillon, Y.; Geiger, E.; Le Gall, C.; Bernard, S.; Gallais-During, A.; Malgouyres, P. P.; Hanus, E.; Ducros, G.

    2017-11-01

    This paper describes the first VERDON test performed at the end of September 2011 with special emphasis on the behaviour of fission products (FP) and actinides during the accidental sequence itself. Two other papers discuss in detail the post-test examination results (SEM, EPMA and SIMS) of the VERDON-1 sample. The first VERDON test was devoted to studying UO2 fuel behaviour and fission product releases under reducing conditions at very high temperature (∼2883 K), which was able to confirm the very good performance of the VERDON loop. The fuel sample did not lose its integrity during this test. According to the FP behaviour measured by the online gamma station (fuel sight), the general classification of the FP in relation to their released fraction is very accurate, and the burn-up effect on the release rate is clearly highlighted.

  13. Modelling and simulation of the steam line, the high and low pressure turbines and the pressure regulator for the SUN-RAH nucleo electric university simulator; Modelado y simulacion de la linea de vapor, las turbinas de alta y de baja presion y el regulador de presion para el simulador universitario de nucleo electricas SUN RAH

    Energy Technology Data Exchange (ETDEWEB)

    Lopez R, A. [DEPFI, Campus Morelos, en IMTA Jiutepec, Morelos, UNAM (Mexico)]. e-mail: andyskamx@yahoo.com.mx

    2003-07-01

    In the following article the development of a simulator that allows to represent the dynamics of the following systems: steam line, nozzle, vapor separator, reheater, high pressure turbine, low pressure turbine, power generator and the pressure regulator of a nucleo electric power station. We start from the supposition that this plant will be modeled from a nuclear reactor type BWR (Boiling Water Reactor), using models of reduced order that represent the more important dynamic variables of the physical processes that happen along the steam line until the one generator. To be able to carry out the simulation in real time the Mat lab mathematical modeling software is used, as well as the specific simulation tool Simulink. It is necessary to point out that the platform on which the one is executed the simulator is the Windows operating system, to allow the intuitive use that only this operating system offers. The above-mentioned obeys to that the objective of the simulator it is to help the user to understand some of the dynamic phenomena that are present in the systems of a nuclear plant, and to provide a tool of analysis and measurement of variables to predict the desirable behavior of the same ones. The model of a pressure controller for the steam lines, the high pressure turbine and the low pressure turbine is also presented that it will be the one in charge of regulating the demand of the system according to the characteristics and critic restrictions of safety and control, assigned according to those wanted parameters of performance of this system inside the nucleo electric plant. This simulator is totally well defined and it is part of the University student nucleo electric simulator with Boiling Water Reactor (SUN-RAH), an integral project and of greater capacity. (Author)

  14. Cold flow testing of the Space Shuttle Main Engine alternate turbopump development high pressure fuel turbine model

    Science.gov (United States)

    Gaddis, Stephen W.; Hudson, Susan T.; Johnson, P. D.

    1992-01-01

    NASA's Marshall Space Flight Center has established a cold airflow turbine test program to experimentally determine the performance of liquid rocket engine turbopump drive turbines. Testing of the SSME alternate turbopump development (ATD) fuel turbine was conducted for back-to-back comparisons with the baseline SSME fuel turbine results obtained in the first quarter of 1991. Turbine performance, Reynolds number effects, and turbine diagnostics, such as stage reactions and exit swirl angles, were investigated at the turbine design point and at off-design conditions. The test data showed that the ATD fuel turbine test article was approximately 1.4 percent higher in efficiency and flowed 5.3 percent more than the baseline fuel turbine test article. This paper describes the method and results used to validate the ATD fuel turbine aerodynamic design. The results are being used to determine the ATD high pressure fuel turbopump (HPFTP) turbine performance over its operating range, anchor the SSME ATD steady-state performance model, and validate various prediction and design analyses.

  15. A proactive alarm reduction method and its human factors validation test for a main control room for SMART

    International Nuclear Information System (INIS)

    Jang, Gwi-sook; Suh, Sang-moon; Kim, Sa-kil; Suh, Yong-suk; Park, Je-yun

    2013-01-01

    Highlights: ► A proactive alarm reduction method improves effectiveness on the alarm reduction. ► The method suppresses alarms based on the ECA rules and facts for the alarm reduction under an alarm flood situation. ► The alarm reduction logics are supplemented to a high hit ratio of the reduction logics during on-line operations. ► The method is validated by human factors validation test based on regulatory requirements. -- Abstract: Conventional alarm systems tend to overwhelm operators during a transient because of a large number of nearly simultaneous annunciator activations with varying degrees of relevance to operator tasks. Thus alarm processing techniques have developed to support operators in coping with the volume of alarms, to identify which alarms are significant, and to reduce the need for operators to infer the plant conditions. This paper proposes a proactive alarm reduction method for SMART (System-integrated Modular Advanced ReacTor) whereby based on the contents of the past operating effects alarm reduction is carried out during the next transient. We designed and implemented the proactive alarm reduction system and constructed the environment for the human factors validation test. Also, eight subjects actually working in a nuclear power plant (NPP) tested the practical effectiveness of the proposed proactive alarm reduction method according to the procedure of human factors validation test under a dynamic simulation of a partial scope for an NPP.

  16. Science Library of Test Items. Volume Twenty-One. A Collection of Multiple Choice Test Items Relating Mainly to Physics, 2.

    Science.gov (United States)

    New South Wales Dept. of Education, Sydney (Australia).

    As one in a series of test item collections developed by the Assessment and Evaluation Unit of the Directorate of Studies, items are made available to teachers for the construction of unit tests or term examinations or as a basis for class discussion. Each collection was reviewed for content validity and reliability. The test items meet syllabus…

  17. Pre-Test Analysis of Major Scenarios for ATLAS

    Energy Technology Data Exchange (ETDEWEB)

    Euh, Dong-Jin; Choi, Ki-Yong; Park, Hyun-Sik; Kwon, Tae-Soon

    2007-02-15

    A thermal-hydraulic integral effect test facility, ATLAS was constructed at the Korea Atomic Energy Research Institute (KAERI). The ATLAS is a 1/2 reduced height and 1/288 volume scaled test facility based on the design features of the APR1400. The simulation capability of the ATLAS for major design basis accidents (DBAs), including a large-break loss-of-coolant (LBLOCA), DVI line break and main steam line break (MSLB) accidents, is evaluated by the best-estimate system code, MARS, with the same control logics, transient scenarios and nodalization scheme. The validity of the applied scaling law and the thermal-hydraulic similarity between the ATLAS and the APR1400 for the major design basis accidents are assessed. It is confirmed that the ATLAS has a capability of maintaining an overall similarity with the reference plant APR1400 for the major design basis accidents considered in the present study. However, depending on the accident scenarios, there are some inconsistencies in certain thermal hydraulic parameters. It is found that the inconsistencies are mainly due to the reduced power effect and the increased stored energy in the structure. The present similarity analysis was successful in obtaining a greater insight into the unique design features of the ATLAS and would be used for developing the optimized experimental procedures and control logics.

  18. Pre-Test Analysis of Major Scenarios for ATLAS

    International Nuclear Information System (INIS)

    Euh, Dong-Jin; Choi, Ki-Yong; Park, Hyun-Sik; Kwon, Tae-Soon

    2007-02-01

    A thermal-hydraulic integral effect test facility, ATLAS was constructed at the Korea Atomic Energy Research Institute (KAERI). The ATLAS is a 1/2 reduced height and 1/288 volume scaled test facility based on the design features of the APR1400. The simulation capability of the ATLAS for major design basis accidents (DBAs), including a large-break loss-of-coolant (LBLOCA), DVI line break and main steam line break (MSLB) accidents, is evaluated by the best-estimate system code, MARS, with the same control logics, transient scenarios and nodalization scheme. The validity of the applied scaling law and the thermal-hydraulic similarity between the ATLAS and the APR1400 for the major design basis accidents are assessed. It is confirmed that the ATLAS has a capability of maintaining an overall similarity with the reference plant APR1400 for the major design basis accidents considered in the present study. However, depending on the accident scenarios, there are some inconsistencies in certain thermal hydraulic parameters. It is found that the inconsistencies are mainly due to the reduced power effect and the increased stored energy in the structure. The present similarity analysis was successful in obtaining a greater insight into the unique design features of the ATLAS and would be used for developing the optimized experimental procedures and control logics

  19. BWR 1 % main recirculation line break LOCA tests, RUNs 917 and 918, without HPCS at ROSA-III program

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Okazaki, Motoaki; Anoda, Yoshinari; Kumamaru, Hiroshige; Nakamura, Hideo; Yonomoto, Taisuke; Koizumi, Yasuo; Tasaka, Kanji

    1988-07-01

    In a case of small break loss-of-coolant accident (LOCA) at a boiling water reactor (BWR) system, it is important to lower the system pressure to cool down the reactor system by using either the high pressure core spray (HPCS) or the automatic depressurization system (ADS). The report presents characteristic test results of RUNs 918 and 917, which were performed at the rig-of-safety assessment (ROSA)-III program simulating a 1 % break BWR LOCA with an assumption of HPCS failure, and clarifies effects of the ADS delay time on a small break LOCA. The ROSA-III test facility simulates principal components of a BWR/6 system with volumetric scaling factor of 1/424. It is experimentally concluded that the ADS delay time shorter than 4 minutes results in a similar PCT as that in a standard case, in which the PCT is observed after actuation of the low pressure core spray (LPCS). And the ADS delay time longer than 4 minutes results in higher PCT than in the standard case. In the latter, the PCT depends on the ADS time, a 220 K higher PCT, for example, in a case of 10 minutes ADS delay compared with the standard case. (author) 52 refs. 299 figs

  20. Main Memory

    NARCIS (Netherlands)

    P.A. Boncz (Peter); L. Liu (Lei); M. Tamer Özsu

    2008-01-01

    htmlabstractPrimary storage, presently known as main memory, is the largest memory directly accessible to the CPU in the prevalent Von Neumann model and stores both data and instructions (program code). The CPU continuously reads instructions stored there and executes them. It is also called Random

  1. The definition of achievement and the construction of tests for its measurement: A review of the main trends

    Directory of Open Access Journals (Sweden)

    Salvador Algarabel

    2001-01-01

    Full Text Available En esta revisión se analizan diferentes definiciones de rendimiento y se exploran posibilidades en la construcción de tests para su medida. Una primera caracterización del rendimiento se consigue a través del análisis de la representación del constructo. Desde esta perspectiva, la aproximación conductual, se centra más en el resultado final, mientras el enfoque cognitivo se centra más en el proceso. En segundo lugar, esta revisión analiza los datos sobre amplitud nomotética: relación entre rendimiento y aptitudes, status socioeconómico y cambios en el tiempo. La sección final ofrece una visión de las posibilidades y dificultades implicadas en el intento de sustituir los métodos tradicionalmente utilizados en la evaluación del rendimiento. Dada su dificultad y coste en términos del tiempo necesario para desarrollarlos, puntuarlos y otras variables, se concluye atribuyendo un peso mayor a las aplicaciones informáticas en evaluación, para que la evaluación conductual pueda tener mayor difusión.

  2. Main Memory

    OpenAIRE

    Boncz, Peter; Liu, Lei; Özsu, M.

    2008-01-01

    htmlabstractPrimary storage, presently known as main memory, is the largest memory directly accessible to the CPU in the prevalent Von Neumann model and stores both data and instructions (program code). The CPU continuously reads instructions stored there and executes them. It is also called Random Access Memory (RAM), to indicate that load/store instructions can access data at any location at the same cost, is usually implemented using DRAM chips, which are connected to the CPU and other per...

  3. Space Shuttle Main Engine Low Pressure Oxidizer Turbo-Pump Inducer Dynamic Environment Characterization through Water Model and Hot-Fire Testing

    Science.gov (United States)

    Arellano, Patrick; Patton, Marc; Schwartz, Alan; Stanton, David

    2006-01-01

    The Low Pressure Oxidizer Turbopump (LPOTP) inducer on the Block II configuration Space Shuttle Main Engine (SSME) experienced blade leading edge ripples during hot firing. This undesirable condition led to a minor redesign of the inducer blades. This resulted in the need to evaluate the performance and the dynamic environment of the redesign, relative to the current configuration, as part of the design acceptance process. Sub-scale water model tests of the two inducer configurations were performed, with emphasis on the dynamic environment due to cavitation induced vibrations. Water model tests were performed over a wide range of inlet flow coefficient and pressure conditions, representative of the scaled operating envelope of the Block II SSME, both in flight and in ground hot-fire tests, including all power levels. The water test hardware, facility set-up, type and placement of instrumentation, the scope of the test program, specific test objectives, data evaluation process and water test results that characterize and compare the two SSME LPOTP inducers are discussed. In addition, dynamic characteristics of the two water models were compared to hot fire data from specially instrumented ground tests. In general, good agreement between the water model and hot fire data was found, which confirms the value of water model testing for dynamic characterization of rocket engine turbomachinery.

  4. Computer programs for locating and fitting full energie peak in γ-ray spectra. Test and rules for an estimation of the main results

    International Nuclear Information System (INIS)

    1980-12-01

    After the different interlaboratory tests on gamma spectrum analysis organised by the 'Laboratoire de Metrologie des Rayonnements Ionisants' and by the International Atomic Energy Agency, it looked useful to manage a same type of intercomparison with the different supplies of Data acquisition and Analysis systems including mini-ordinator or microprocessor. Four spectrum have been chosen between those of the interlaboratory tests. The test dealt with the investigation of total absorption peaks of different levels in a complex spectrum and the calculation of their main parameters. Four supplies participed in the intercomparison with their own logicial. The result allow to suggest a few tests in order to try a new logicial, or to compare results with standards [fr

  5. Extrapolating the Trends of Test Drop Data with Opening Shock Factor Calculations: the Case of the Orion Main and Drogue Parachutes Inflating to 1st Reefed Stage

    Science.gov (United States)

    Potvin, Jean; Ray, Eric

    2017-01-01

    We describe a new calculation of the opening shock factor C (sub k) characterizing the inflation performance of NASA's Orion spacecraft main and drogue parachutes opening under a reefing constraint (1st stage reefing), as currently tested in the Capsule Parachute Assembly System (CPAS) program. This calculation is based on an application of the Momentum-Impulse Theorem at low mass ratio (R (sub m) is less than 10 (sup -1)) and on an earlier analysis of the opening performance of drogues decelerating point masses and inflating along horizontal trajectories. Herein we extend the reach of the Theorem to include the effects of payload drag and gravitational impulse during near-vertical motion - both important pre-requisites for CPAS parachute analysis. The result is a family of C (sub k) versus R (sub m) curves which can be used for extrapolating beyond the drop-tested envelope. The paper proves this claim in the case of the CPAS Mains and Drogues opening while trailing either a Parachute Compartment Drop Test Vehicle or a Parachute Test Vehicle (an Orion capsule boiler plate). It is seen that in all cases the values of the opening shock factor can be extrapolated over a range in mass ratio that is at least twice that of the test drop data.

  6. A data base and analysis program for shuttle main engine dynamic pressure measurements. Appendix C: Data base plots for SSME tests 902-214 through 902-314

    Science.gov (United States)

    Coffin, T.

    1986-01-01

    A dynamic pressure data base and data base management system developed to characterize the Space Shuttle Main Engine (SSME) dynamic pressure environment is reported. The data base represents dynamic pressure measurements obtained during single engine hot firing tests of the SSME. Software is provided to permit statistical evaluation of selected measurements under specified operating conditions. An interpolation scheme is included to estimate spectral trends with SSME power level. Flow Dynamic Environments in High Performance Rocket Engines are described.

  7. A data base and analysis program for shuttle main engine dynamic pressure measurements. Appendix F: Data base plots for SSME tests 750-120 through 750-200

    Science.gov (United States)

    Coffin, T.

    1986-01-01

    A dynamic pressure data base and data base management system developed to characterize the Space Shuttle Main Engine (SSME) dynamic pressure environment is presented. The data base represents dynamic pressure measurements obtained during single engine hot firing tests of the SSME. Software is provided to permit statistical evaluation of selected measurements under specified operating conditions. An interpolation scheme is also included to estimate spectral trends with SSME power level.

  8. Main findings

    International Nuclear Information System (INIS)

    2014-01-01

    Licensing regimes vary from country to country. When the license regime involves several regulators and several licenses, this may lead to complex situations. Identifying a leading organisation in charge of overall coordination including preparation of the licensing decision is a useful practice. Also, if a stepwise licensing process is implemented, it is important to fix in legislation decisions and/or time points and to identify the relevant actors. There is considerable experience in civil and mining engineering that can be applied when constructing a deep geological disposal facility. Specific challenges are, however, the minimization of disturbances to the host rock and the understanding of its long-term behavior. Construction activities may affect the geo-hydraulic and geochemical properties of the various system components which are important safety features of the repository system. Clearly defined technical specifications and an effective quality management plan are important in ensuring successful repository implementation which is consistent with safety requirements. Monitoring plan should also be defined in advance. The regulatory organization should prepare itself to the licensing review before construction by allocating sufficient resources. It should increase its competence, e.g., by interacting early with the implementer and through its own R and D. This will allow the regulator to define appropriate technical conditions associated to the construction license and to elaborate a relevant inspection plan of the construction work. After construction, obtaining the operational license is the most important and crucial step. Main challenges include (a) establishing sufficient confidence so that the methods for closing the individual disposal units comply with the safety objectives and (b) addressing the issue of ageing of materials during a 50-100 years operational period. This latter challenge is amplified when reversibility/retrievability is required

  9. Post-test analysis of ROSA-III experiment RUNs 705 and 706

    International Nuclear Information System (INIS)

    Koizumi, Yasuo; Soda, Kunihisa; Kikuchi, Osamu; Tasaka, Kanji; Shiba, Masayoshi

    1980-07-01

    The purpose of ROSA-III experiment with a scaled BWR Test facility is to examine primary coolant thermal-hydraulic behavior and performance of ECCS during a postulated loss-of-coolant accident of BWR. The results provide the information for verification and improvement of reactor safety analysis codes. RUNs 705 and 706 assumed a 200% double-ended break at the recirculation pump suction. RUN 705 was an isothermal blowdown test without initial power and initial core flow. In RUN 706 for an average core power and no ECCS, the main steam line and feed water line were isolated immediately on the break. Post-test analysis of RUNs 705 and 706 was made with computer code RELAP4J. The agreement in system pressure between calculation and experiment was satisfactory. However, the calculated heater rod surface temperature were significantly higher than the experimental ones. The calculated axial temperature profile was different in tendency from the experimental one. The calculated mixture level behavior in the core was different from the liquid void distribution observed in experiment. The rapid rise of fuel rod surface temperature was caused by the reduction of heat transfer coefficient attributed to the increase of quality. The need was indicated for improvement of analytical model of void distribution in the core, and also to performe a characteristic test of recirculation line under reverse flow and to examine the core inlet flow rate experimentally and analytically. (author)

  10. Performance Data from a Wind-Tunnel Test of Two Main-rotor Blade Designs for a Utility-Class Helicopter

    Science.gov (United States)

    Singleton, Jeffrey D.; Yeager, William T., Jr.; Wilbur, Matthew L.

    1990-01-01

    An investigation was conducted in the NASA Langley Transonic Dynamics Tunnel to evaluate an advanced main rotor designed for use on a utility class helicopter, specifically the U.S. Army UH-60A Blackhawk. This rotor design incorporated advanced twist, airfoil cross sections, and geometric planform. For evaluation purposes, the current UH-60A main rotor was also tested and is referred to as the baseline blade set. A total of four blade sets were tested. One set of both the baseline and the advanced rotors were dynamically scaled to represent a full scale helicopter rotor blade design. The remaining advanced and baseline blade sets were not dynamically scaled so as to isolate the effects of structural elasticity. The investigation was conducted in hover and at rotor advance ratios ranging from 0.15 to 0.4 at a range of nominal test medium densities from 0.00238 to 0.009 slugs/cu ft. This range of densities, coupled with varying rotor lift and propulsive force, allowed for the simulation of several vehicle gross weight and density altitude combinations. Performance data are presented for all blade sets without analysis; however, cross referencing of data with flight condition may be useful to the analyst for validating aeroelastic theories and design methodologies as well as for evaluating advanced design parameters.

  11. Shaking table test of a base isolated model in main control room of nuclear power plant using LRB (lead rubber bearing)

    International Nuclear Information System (INIS)

    Ham, K. W.; Lee, K. J.; Suh, Y. P.

    2005-01-01

    LRB(Lead Rubber Bearing) is a widely used isolation system which is installed between equipment and foundation to reduce seismic vibration from ground. LRB is consist of bearings which are resistant to lateral motion and torsion and has a high vertical stiffness. For that reason, several studies are conducted to apply LRB to the nuclear power plant. In this study, we designed two types of main control floor systems (type I, type II) and a number of shaking table tests with and without isolation system were conducted to evaluate floor isolation effectiveness of LRB

  12. Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 3E. Kozloduy NPP units 5/6: Analysis/testing. Working material

    International Nuclear Information System (INIS)

    1996-01-01

    The Co-ordinated research programme on the benchmark study for the seismic analysis and testing of WWER-type nuclear power plants was initiated subsequent to the request from representatives of Member States. The conclusions adopted at the Technical Committee Meeting on Seismic Issues related to existing nuclear power plants held in Tokyo in 1991 called for the harmonization of methods and criteria used in Member States in issues related to seismic safety. The Consulltants' Meeting which followed resulted in producing a working document for CRP. It was decided that a benchmark study is the most effective way to achieve the principal objective. Two types of WWER reactors (WWER-440/213 and WWER-1000) were selected as prototypes for the benchmark exercise to be tested on a full scale using explosions and/or vibration generators. The two prototypes are Kozloduy Units 5/6 for WWER-1000 and Paks for WWER-440/213 nuclear power plants. This volume of Working material contains reports on data related to floor response spectra of Kozloduy NPP; calculational-experimental examination and ensuring of equipment and pipelines seismic resistance at starting and operating WWER-type NPPs; analysis of design floor response spectra and testing of the electrical systems; experimental investigations and seismic analysis Kozloduy NPP; testing of components on the shaking table facilities and contribution to full scale dynamic testing of Kozloduy NPP; seismic evaluation of the main steam line, piping systems, containment pre-stressing and steel ventilation chimney of Kozloduy NPP

  13. Streamlined Approach for Environmental Restoration Plan for Corrective Action Unit 425: Area 9 Main Lake Construction Debris Disposal Area, Tonopah Test Range, Nevada; TOPICAL

    International Nuclear Information System (INIS)

    K. B. Campbell

    2002-01-01

    This Streamlined Approach for Environmental Restoration (SAFER) Plan addresses the action necessary for the closure of Corrective Action Unit (CAU) 425, Area 9 Main Lake Construction Debris Disposal Area. This CAU is currently listed in Appendix III of the Federal Facility Agreement and Consent Order (FFACO, 1996). This site will be cleaned up under the SAFER process since the volume of waste exceeds the 23 cubic meters (m(sup 3)) (30 cubic yards[yd(sup 3)]) limit established for housekeeping sites. CAU 425 is located on the Tonopah Test Range (TTR) and consists of one Corrective Action Site (CAS) 09-08-001-TA09, Construction Debris Disposal Area (Figure 1). CAS 09-08-001-TA09 is an area that was used to collect debris from various projects in and around Area 9. The site is located approximately 81 meters (m) (265 feet[ft]) north of Edwards Freeway northeast of Main Lake on the TTR. The site is composed of concrete slabs with metal infrastructure, metal rebar, wooden telephone poles, and concrete rubble from the Hard Target and early Tornado Rocket sled tests. Other items such as wood scraps, plastic pipes, soil, and miscellaneous nonhazardous items have also been identified in the debris pile. It is estimated that this site contains approximately 2280 m(sup 3) (3000 yd(sup 3)) of construction-related debris

  14. Assessment of the SPACE Code Using the ATLAS SLB-GB-01 Test

    International Nuclear Information System (INIS)

    Kim, Yo Han; Yang, Chang Keun; Kim, Seyun

    2013-01-01

    The Korea Nuclear Hydro and Nuclear Power Co. (KHNP) has developed a safety analysis code, called the Safety and Performance Analysis Code for Nuclear Power Plants (SPACE) by collaborative works with other Korean nuclear industries. The SPACE is a general-purpose best-estimated two-phase three-field thermal-hydraulic analysis code to analyze the safety and performance of pressurized water reactors (PWRs). The SPACE code has sufficient functions and capabilities to replace outdated vendor supplied codes and to be used for the safety analysis of operating PWRs and the design of advanced reactors. As a result of the second phase of the SPACE code development project, the 2.14 version of the code was released through the successive various V and V works using integral loop test data or plant operating data. In this study, the ATLAS main steam-line break (MSLB) test, SLB-GB-01, was simulated as a V and V work. The results were compared with the measured data. The ATALS MSLB test, SLB-GB-01, was simulated using the SPACE code. The results were compared with experimental data. Through the simulation, it was concluded that the SPACE code can effectively simulate MSLB accidents

  15. Post-test sensitivity analysis of OECD/CSNI ISP42 panda experiment by Relap5 code

    International Nuclear Information System (INIS)

    Zanocco, P.; D'Auria, F.; Galassi, G.M.

    2001-01-01

    The present document deals with Relap5/Mod3.2 analysis of the International Standard Problem (ISP-42) exercise performed in PANDA facility on April 21-22, 1998. PANDA is installed at PSI (Paul Scherrer Institute). PANDA is a large-scale thermal-hydraulic test facility suitable for the simulation of passive containment for Advanced Light Water Reactors (ALWR). The work focuses phase A of the ISP-42 experiment, including the break in the main steam line, and the Passive Containment Cooling System Start-Up. The objective is to investigate the start-up phenomenology of passive cooling system when steam is injected into cold vessel filled with air and to observe the resulting system behavior. A detailed nodalization was set-up at the University of Pisa, in order to model 3-D flow paths with a 1-D code. The comparison between pre-test predictions and experimental data is discussed. Overall time behavior is reasonably well predicted, showing a rather good and robust overall code behavior in the simulation of the global test scenario. The results of a preliminary post-test analysis are discussed, including the comparison with the experimental data. (authors)

  16. Supplemental Environmental Baseline Survey for Proposed Land Use Permit Modification for Expansion of the Dynamic Explosive Test Site (DETS) 9940 Main Complex Parking Lot

    Energy Technology Data Exchange (ETDEWEB)

    Peek, Dennis W. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-10-01

    The “subject property” is comprised of a parcel of land within the Kirtland Military Reservation, Bernalillo County, New Mexico, as shown on the map in Appendix B of this document. The land requirement for the parking lot addition to the 9940 Main Complex is approximately 2.7 acres. The scope of this Supplemental Environmental Baseline Survey (SEBS) is for the parking lot addition land transfer only. For details on the original 9940 Main Complex see Environmental Baseline Survey, Land Use Permit Request for the 9940 Complex PERM/0-KI-00-0001, August 21, 2003, and for details on the 9940 Complex Expansion see Environmental Baseline Survey, Proposed Land Use Permit Expansion for 9940 DETS Complex, June 24, 2009. The 2.7-acre parcel of land for the new parking lot, which is the subject of this EBS (also referred to as the “subject property”), is adjacent to the southwest boundary of the original 12.3- acre 9940 Main Complex. No testing is known to have taken place on the subject property site. The only activity known to have taken place was the burial of overhead utility lines in 2014. Adjacent to the subject property, the 9940 Main Complex was originally a 12.3-acre site used by the Department of Energy (DOE) under a land use permit from the United States Air Force (USAF). Historical use of the site, dating from 1964, included arming, fusing, and firing of explosives and testing of explosives systems components. In the late 1970s and early 1980s experiments at the 9940 Main Complex shifted toward reactor safety issues. From 1983 to 1988, fuel coolant interaction (FCI) experiments were conducted, as were experiments with conventional high explosives (HE). Today, the land is used for training of the Nuclear Emergency Response community and for research on energetic materials. In 2009, the original complex was expanded to include four additional 20-acre areas: 9940 Training South, 9940 Training East, T-Range 6, and Training West Landing Zone. The proposed use of

  17. Commissioning of an Integral Effect Test Loop for SMART

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-05-15

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

  18. Development and evaluation of the NSSS model with four steam lines for the LVNP using the SCDAPSIM code; Desarrollo y evaluacion del modelo del NSSS con cuatro lineas de vapor para la CNLV utilizando el codigo SCDAPSIM

    Energy Technology Data Exchange (ETDEWEB)

    Salazar C, J.H.; Nunez C, A.; Camargo C, R. [CNSNS, Dr. Barragan No. 779, Col. Narvarte, 03020 Mexico D.F. (Mexico)

    2005-07-01

    The present work shows the pattern of the NSSS considering the four main vapor lines as well as their evaluation. The pattern was developed by the National Commission of Nuclear Security and Safeguards (CNSNS) and it has as main objective to account with a model of the Laguna Verde Nuclear power plant (CNLV) for the simulation and analysis of transitory events where are involved some of main vapor lines, or some relief valves and safety (SRV's). The model was evaluated with data of the CNLV. In 1996 the Federal Commission of Electricity (CFE) request to the CNSNS permission to operate the Unit 2 until the first recharge, having the main vapor line 'B' isolated and operating with a level of power corresponding to a flow of total vapor of 85% of the nominal one (of 1931 MWt). The obtained values were compared with the obtained registrations of the CNLV in order to evaluate the model. Those results show relative errors inferior to 3% among the CNLV reported value and the one calculated by the SCDAPSIM code. (Author)

  19. Fluorescent antibody test, quantitative polymerase chain reaction pattern and clinical aspects of rabies virus strains isolated from main reservoirs in Brazil

    Directory of Open Access Journals (Sweden)

    Camila Appolinário

    2015-09-01

    Full Text Available Rabies virus (RABV isolated from different mammals seems to have unique characteristics that influence the outcome of infection. RABV circulates in nature and is maintained by reservoirs that are responsible for the persistence of the disease for almost 4000 years. Considering the different pattern of pathogenicity of RABV strains in naturally and experimentally infected animals, the aim of this study was to analyze the characteristics of RABV variants isolated from the main Brazilian reservoirs, being related to a dog (variant 2, Desmodus rotundus (variant 3, crab eating fox, marmoset, and Myotis spp. Viral replication in brain tissue of experimentally infected mouse was evaluated by two laboratory techniques and the results were compared to clinical evolution from five RABV variants. The presence of the RABV was investigated in brain samples by fluorescent antibody test (FAT and real time polymerase chain reaction (qRT-PCR for quantification of rabies virus nucleoprotein gene (N gene. Virus replication is not correlated with clinical signs and evolution. The pattern of FAT is associated with RABV replication levels. Virus isolates from crab eating fox and marmoset had a longer evolution period and higher survival rate suggesting that the evolution period may contribute to the outcome. RABV virus variants had independent characteristics that determine the clinical evolution and survival of the infected mice.

  20. Main designations and attributions

    International Nuclear Information System (INIS)

    2010-01-01

    The chapter presents the main designations and attributions of the LNMRI - Brazilian National Laboratory of Metrology of Ionizing Radiation, the Cooperative Center in Radiation Protection and Medical Preparations for Accidents with Radiation; the Treaty for fully banning of nuclear tests and the Regional Center for Training of IAEA

  1. Full scale BWR containment LOCA response test at the INKA test facility

    International Nuclear Information System (INIS)

    Wagner, Thomas; Leyer, Stephan

    2015-01-01

    KERENA is an innovative boiling water reactor concept with passive safety systems (Generation III+) of AREVA. The reactor is an evolutionary design of operating BWRs (Generation II). In order to verify the functionality and performance of the KERENA safety concept required for the transient and accident management, the test facility “Integral Teststand Karlstein” (INKA) was built at Karlstein (Germany). It is a mock-up of the KERENA boiling water reactor containment, with integrated pressure suppression system. The complete chain of passive safety components is available. The passive components and the levels are represented in full scale. The volume scaling of the containment compartments is approximately 1:24. The reactor pressure vessel (RPV) is simulated via the steam accumulator of the Karlstein Large Valve Test Facility. This vessel provides an energy storage capacity of approximately 1/6 of the KERENA RPV and is supplied by a Benson boiler with a thermal power of 22 MW. With respect to the available power supply, the containment- and system-sizing of the facility is by far the largest one of its kind worldwide. From 2009 to 2012, several single component tests were conducted (Emergency Condenser, Containment Cooling Condenser, Core Flooding System etc.). On March 21st, 2013, the worldwide first large-scale only passively managed integral accident test of a boiling water reactor was simulated at INKA. The integral test measured the combined response of the KERENA passive safety systems to the postulated initiating event was the “Main Steam Line Break” (MSLB) inside the Containment with decay heat simulation. The results of the performed integral test (MSLB) showed that the passive safety systems alone 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 as response to an anticipated accident scenario

  2. Background reduction of the KATRIN spectrometers. Transmission function of the pre-spectrometer and systematic tests of the main-spectrometer wire electrode

    Energy Technology Data Exchange (ETDEWEB)

    Prall, Matthias

    2011-07-04

    The KArlsruhe TRItium Neutrino experiment, KATRIN will determine the mass of the anti {nu}{sub e} with a sensitivity of 0.2 eV (90% C.L.) via a measurement of the {beta}-spectrum of tritium decaying in a windowless gaseous molecular tritium source near its endpoint of 18.57 keV. This approach relies exclusively on the relativistic kinematics of the decay products rendering the experiment model independent and reducing the systematic uncertainty. An ultra-low background of a few mHz and an energy resolution of 0.93 eV are among the requirements to reach the sensitivity. These demands are fulfilled with the main spectrometer (MS). While the {beta}-decay electrons are guided by a magnetic field through the experiment, the MS acts as a high-pass filter for the {beta}-decay electrons. Only those above an energy barrier, the retarding potential, are transmitted to the detector. The last about 30 eV of the T{sub 2} {beta}-spectrum will be scanned in this way. The MS is equipped with a 650 m{sup 2}, two-layered, UHV compatible and quasi-massless wire electrode suppressing secondary electron background originating at the main-spectrometer walls and caused by residual radioactivity and cosmic muons. Its energy resolution of 0.93 eV is only achieved, if a large part of the 248 wire electrode modules, which determine the electric field inside the MS, has a mechanical precision of 0.2 mm. Not a single of the about 28.000 wires of the electrode must break during the lifetime of KATRIN. A 2-dimensional laser sensor for contact-less position (precision about 0.01 mm) and tension (precision about 0.04 N) measurements was developed and applied, to firstly, verify the mechanical precision of the electrode modules and secondly, to examine their reliability. A 3-dimensional coordinate measurement table was automated to perform these measurements in a clean room. This table was also used to verify the precision of components using a camera system and image recognition methods (0.05 mm

  3. Studies to demonstrate the adequacy of testing results of the qualification tests for the actuator of main steam safety relive valves (MSSRV) in an advanced boiling water reactor (ABWR)

    International Nuclear Information System (INIS)

    Gou, P.F.; Patel, R.; Curran, G.; Henrie, D.; Solorzano, E.

    2005-01-01

    This paper presents several studies performed to demonstrate that the testing results from the qualification tests for the actuator of the Main Steam Safety Relief Valves (MSSRV; also called SRV in this paper) in GE's Advanced Boiling Water Reactor (ABWR) are in compliance with the qualification guidelines stipulated in the applicable IEEE standards. The safety-related function of the MSSRV is to relieve pressure in order to protect the reactor pressure vessel from over-pressurization condition during normal operation and design basis events. In order to perform this function, the SRV must actuate at a given set pressure while maintaining the pressure and structural integrity of the SRV. The valves are provided with an electro-pneumatic actuator assembly that opens the valve upon receipt of an automatic or manually initiated electric signal to allow depressurization of the reactor pressure vessel (RPV). To assure the SRV can perform its intended safety related functions properly, qualification tests are needed in addition to analysis, to demonstrate that the SRV can withstand the specified environmental, dynamic and seismic design basis conditions without impairing its safety related function throughout their installed life under the design conditions including postulated design basis events such as OBE loads and Faulted (SSE) events. The guidelines used for the test methods, procedures and acceptance criteria for the qualification tests are established in IEEE std 344-1987 and IEEE std 382-1985. In the qualification tests, the specimen consists of the actuator, control valve assembly, limit switches, and limit switch support structure. During the functional, dynamic and seismic tests, the test specimen was mounted on a SRV. Qualification of safety related equipment to meet the guidelines of the IEEE standards is typically a two-step process: 1) environmental aging and 2) design basis events qualification. The purpose of the first step is to put the equipment in an

  4. RELAP5 capabilities in thermal-hydraulic prediction of SBWR containment behaviour: PANDA steady state and transient tests evaluation

    International Nuclear Information System (INIS)

    Faluomi, V.; Aksan, S.N.

    2000-01-01

    This paper summarizes the results of the qualification activity of RELAP5/Mod3.2 code performed using PANDA steady state and integral test experimental data. The steady state tests evaluate the PCC performances in removing decay heat power in presence and in absence of non-condensable gases, while the considered integral test (M3) simulates the transient following a break in the main steam line of the SBWR, using, as nominal initial conditions, those calculated for the SBWR under SSAR assumptions at one hour into the LOCA. The results obtained simulating both types of tests show a rather good and robust overall code behavior both in the simulation of steady state test and in the representation of the integral test considered: most of the main experimental results (WW/DW pressures, PCC heat exchange) were well represented by the code. The different studies performed indicated that: Different models of PCC pool lead a different trend of system pressure, and sometimes to an opening of vacuum breaker valves, that does not occur in the transient; The code underestimate the heat exchanged between PCC pool and tubes: n the considered test the system pressure is slightly overestimated (maximum 2% more than the experimental value). This fact is also proved by the differences in the temperature of the condensing mixture in the PCC, quite large in all the performed studies; The treatment of the non condensable gases, as implemented in the code, lead some errors in the calculation of the heat transfer coefficient in the PCC components and generally slow down the overall calculation. In general terms, the RELAP5/Mod3.2 was found to be suitable to represent the SBWR containment behavior under the conditions specified in the experimental side. (author)

  5. Post test analysis of the LOBI BT01 experiment

    International Nuclear Information System (INIS)

    Hozer, Z.; Takacs, A.

    1994-12-01

    The LOBI experimental facility and the BT01 experiment is described. The experiment represents a small break transient in the secondary side (steam line) followed by special conditions for the establishment of pressurized thermal shock and accident management procedures. The computational analysis has been performed by the CATHARE thermal hydraulic system code. The results of calculations are in satisfactory agreement with the experimental values. A comparison has been made with a secondary side break test performed on the PMK-2 facility. (author). 14 refs., 26 figs., 6 tabs

  6. 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

  7. A data base and analysis program for shuttle main engine dynamic pressure measurements. Appendix B: Data base plots for SSME tests 901-290 through 901-414

    Science.gov (United States)

    Coffin, T.

    1986-01-01

    A dynamic pressure data base and data base management system developed to characterize the Space Shuttle Main Engine (SSME) dynamic pressure environment is described. The data base represents dynamic pressure measurements obtained during single engine hot firing tesets of the SSME. Software is provided to permit statistical evaluation of selected measurements under specified operating conditions. An interpolation scheme is also included to estimate spectral trends with SSME power level. Flow dynamic environments in high performance rocket engines are discussed.

  8. Members of the team responsable for the strength test stand on the plug for the main CMS shaft on which 2500 tonnes of steel blocks have been placed

    CERN Multimedia

    Maximilien Brice

    2006-01-01

    The plug over the CMS shaft, which will be required to bear the weight of the various detector sub-assemblies when they are lowered into the experiment hall, has just passed a strength test. The plug, a huge 2.2-metre-thick rectangular block of reinforced concrete measuring 15 x 20 metres and weighing 2000 tonnes, underwent its first strength test on 15 May.

  9. SPES-2, AP600 intergral system test S01007 2 inch CL to core make-up tank pressure balance line break

    Energy Technology Data Exchange (ETDEWEB)

    Bacchiani, M.; Medich, C.; Rigamonti, M. [SIET S.p.A. Piacenza (Italy)] [and others

    1995-09-01

    The SPES-2 is a full height, full pressure experimental test facility reproducing the Westinghouse AP600 reactor with a scaling factor of 1/395. The experimental plant, designed and operated by SIET in Piacenza, consists of a full simulation of the AP600 primary core cooling system including all the passive and active safety systems. In 1992, Westinghouse, in cooperation with ENEL (Ente Nazionale per l` Energia Elettrica), ENEA (Enter per le numove Technlogie, l` Energia e l` Ambient), Siet (Societa Informazioni Esperienze Termoidraulich) and ANSALDO developed an experimental program to test the integrated behaviour of the AP600 passive safety systems. The SPES-2 test matrix, concluded in November 1994, has examined the AP600 passive safety system response for a range of small break LOCAs at different locations on the primary system and on the passive system lines; single steam generator tube ruptures with passive and active safety systems and a main steam line break transient to demonstrate the boration capability of passive safety systems for rapid cooldown. Each of the tests has provided detailed experimental results for verification of the capability of the analysis methods to predict the integrated passive safety system behaviour. Cold and hot shakedown tests have been performed on the facility to check the characteristics of the plant before starting the experimental campaign. The paper first presents a description of the SPES-2 test facility then the main results of S01007 test {open_quotes}2{close_quotes} Cold Leg (CL) to Core Make-up Tank (CMT) pressure balance line break{close_quotes} are reported and compared with predictions performed using RELAP5/mod3/80 obtained by ANSALDO through agreement with U.S.N.R.C. (U.S. Nuclear Regulatory Commission). The SPES-2 nodalization and all the calculations here presented were performed by ANSALDO and sponsored by ENEL as a part of pre-test predictions for SPES-2.

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

    International Nuclear Information System (INIS)

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

    1978-06-01

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

  11. Field and laboratory tests of etched track detectors for 222Rn: summer-vs-winter variations and tightness effects in Maine houses

    International Nuclear Information System (INIS)

    Hess, C.T.; Fleischer, R.L.; Turner, L.G.

    1985-01-01

    Effects of tightness of homes of bedrock character on indoor 222 Rn concentrations were sought in 70 homes in the state of Maine by means of four 6- to 8-month-long surveys over a 1.5-yr period. Laboratory experiments were also performed that document the reliability of the track etching system used for the measurements. In this survey the Rn in tight homes was on the average 3.5 times that in drafty ones, and areas with granitic bedrock led to homes having 2.3 times the Rn as for homes on chlorite-biotite-rich bedrock. Winter-to-summer ratios ranged from 0.5-7, and averaged 1.5, implying that surveys of individual homes require a full year of monitoring

  12. POLCA-T simulation of OECD/NRC BWR turbine trip benchmark exercise 3 best estimate scenario TT2 test and four extreme scenarios

    International Nuclear Information System (INIS)

    Panayotov, D.

    2004-01-01

    Westinghouse transient code POLCA-T brings together the system thermal-hydraulics plant models and the 3D neutron kinetics core model. Code validation plan includes the calculations of Peach Bottom end of cycle 2 turbine trip transients and low-flow stability tests. The paper describes the objectives, method, and results of analyses performed in the final phase of OECD/NRC Peach Bottom 2 Boiling Water Reactor Turbine Trip Benchmark. Brief overview of the code features, the method of simulation, the developed 3D core model and system input deck for Peach Bottom 2 are given. The paper presents the results of benchmark exercise 3 best estimate scenario: coupled 3D core neutron kinetics with system thermal-hydraulics analyses. Performed sensitivity studies cover the SCRAM initiation, carry-under, and decay power. Obtained results including total power, steam dome, core exit, lower and upper plenum, main steam line and turbine inlet pressures showed good agreement with measured plant data Thus the POLCA-T code capabilities for correct simulation of turbine trip transients were proved The performed calculations and obtained results for extreme cases demonstrate the POLCA-T code wide range capabilities to simulate transients when scram, steam bypass, and safety and relief valves are not activated. The code is able to handle such transients even when the reactor power and pressure reach values higher than 600 % of rated power, and 10.8 MPa. (authors)

  13. Survey of ketolactia, determining the main predisposing management factors and consequences in Hungarian dairy herds by using a cow-side milk test

    Science.gov (United States)

    Zechner, Gerhard; Csorba, Csaba; Könyves, László

    2018-01-01

    The aims of the survey were to determine the prevalence of ketosis in dairy herds by measuring the concentration of beta-hydroxybutyrate (BHBA) in milk by Keto-Test (Sanwa Kagaku Kenkyusho, Nagoya, Japan); risk factors and the relationship with postpartum diseases were investigated. 1667 early lactating (days in milk 0–75) cows were tested in 52 dairy herds in 2013 and 2014 years. In total, 29.3 per cent of samples were positive (BHBAMILK ≥100 µmol/l), including 3.7 per cent high positives (BHBAMILK ≥500 µmol/l). The prevalence was similar in herds with less than or more than 9000 kg milk yield (0.34 and 0.38, respectively, P=0.4); however, it was higher in the herds with more than 1000 cows than in smaller herds (ketosis (P<0.001). The results confirm the high prevalence of ketolactia in Hungarian dairy herds and its links to herd-related and cow-related risk factors and diseases occurring commonly in fresh cows. PMID:29868171

  14. Main Memory DBMS

    NARCIS (Netherlands)

    P.A. Boncz (Peter); L. Liu (Lei); M. Tamer Özsu

    2008-01-01

    htmlabstractA main memory database system is a DBMS that primarily relies on main memory for computer data storage. In contrast, normal database management systems employ hard disk based persisntent storage.

  15. Main Maralinga test sites now clean

    International Nuclear Information System (INIS)

    Anon

    2000-01-01

    Landscaping and revegetating work will soon he complete, bringing the Maralinga Rehabilitation Project to a conclusion on time and within its 108 million budget. The project involved removing surface soil from the resulting plutonium-contaminated areas and the treatment of a number of contaminated debris pits.The Australian Radiation Protection and Nuclear Safety Agency (ARPANSA) monitored the clean-up operation to ensure that the clearance criteria has been met. ARPANSA has indicated that the burial trenches have been constructed consistent with the national Code of Practice for the Near-Surface Disposal of Radioactive Waste

  16. Determination of Functional Capabilities, the Level of Physical Performance and the State of Main Physiological Body Systems in the First Hours after the Accomplishment of Long-term Space Flights ("Field Test")

    Science.gov (United States)

    Kozlovskaya, Inesa; Tomilovskaya, Elena; Rukavishnikov, Ilya; Kitov, Vladimir; Reschke, Millard; Kofman, Igor

    2014-01-01

    Long-term stay in weightlessness is accompanied by alterations in the activity of main physiological body systems including sensory-motor, skeletal-muscular disturbances and cardiovascular deconditioning. However, up to now, there are no data on the state and level of functional performance of cosmonauts/astronauts directly after flight, nor are there data to help define the dynamic recovery of functional characteristics and work efficiency which are greatly needed to provide the safety and planning of their activity once they reach space objects. The Russian and American scientists are currently engaged in a joint experiment known as the "Field Test" with the goal of studying the functional performance and the state of main physiological body systems directly after landing and their temporal recovery dynamics. The functional performance is identified during the test by temporal characteristics of the movements of spatial translation, the stability of the vertical stance for 3.5 min, and the kinematic characteristics of walking - non-complicated and complicated. The following characteristics are identified as physiological characteristics of the test: a) orthostatic tolerance during stand test, b) back muscle tone; c) vertical stability - by characteristics of the correction responses to unexpected perturbations of the vertical stance, and d) support reactions during the performance of the full battery of tests. To date, a pilot version of the "Field Test" has been conducted with participation from four Russian cosmonauts. The results of studies have shown that in 1 - 5 hours after landing the functional abilities of the cosmonauts are considerably reduced. All the test movements at this time are considerably slower than preflight and the more complicated the task is, the greater significant reduction in orthostatic tolerance: during the first test that occurs 1 - 5 hours after landing. two of four cosmonauts declined to continue the task after the orthostatic test

  17. Maine's Employability Skills Program

    Science.gov (United States)

    McMahon, John M.; Wolffe, Karen E.; Wolfe, Judy; Brooker, Carrie

    2013-01-01

    This Practice Report describes the development and implementation of the "Maine Employability Skills Program," a model employment program developed by the Maine Division for the Blind and Visually Impaired (DBVI). The program was designed to support the efforts of the chronically unemployed or underemployed. These consumers were either…

  18. Turbine main engines

    CERN Document Server

    Main, John B; Herbert, C W; Bennett, A J S

    1965-01-01

    Turbine Main Engines deals with the principle of operation of turbine main engines. Topics covered include practical considerations that affect turbine design and efficiency; steam turbine rotors, blades, nozzles, and diaphragms; lubricating oil systems; and gas turbines for use with nuclear reactors. Gas turbines for naval boost propulsion, merchant ship propulsion, and naval main propulsion are also considered. This book is divided into three parts and begins with an overview of the basic mode of operation of the steam turbine engine and how it converts the pressure energy of the ingoing ste

  19. Maine highway safety plan

    Science.gov (United States)

    2010-01-01

    Each September 1, the MeBHS must provide NHTSA a comprehensive plan to reduce : traffic crashes and resulting deaths, injuries and property damage. The Highway Safety : Plan (HSP) serves as Maines application for available federal funds for these ...

  20. Maine Field Station

    Data.gov (United States)

    Federal Laboratory Consortium — In 2000 NOAA's National Marine Fisheries Service established the Maine Field Station in Orono, ME to have more direct involvement in the conservation of the living...

  1. Maine's forests 2008

    Science.gov (United States)

    George L. McCaskill; William H. McWilliams; Charles J. Barnett; Brett J. Butler; Mark A. Hatfield; Cassandra M. Kurtz; Randall S. Morin; W. Keith Moser; Charles H. Perry; Christopher W. Woodall

    2011-01-01

    The second annual inventory of Maine's forests was completed in 2008 after more than 3,160 forested plots were measured. Forest land occupies almost 17.7 million acres, which represents 82 percent of the total land area of Maine. The dominant forest-type groups are maple/beech/yellow birch, spruce/fir, white/red/jack pine, and aspen/white birch. Statewide volume...

  2. Maine Forests 2013

    Science.gov (United States)

    George L. McCaskill; Thomas Albright; Charles J. Barnett; Brett J. Butler; Susan J. Crocker; Cassandra M. Kurtz; William H. McWilliams; Patrick D. Miles; Randall S. Morin; Mark D. Nelson; Richard H. Widmann; Christopher W. Woodall

    2016-01-01

    The third 5-year annualized inventory of Maine's forests was completed in 2013 after more than 3170 forested plots were measured. Maine contains more than 17.6 million acres of forest land, an area that has been quite stable since 1960, covering more than 82 percent of the total land area. The number of live trees greater than 1 inch in diameter are approaching 24...

  3. Blind pre-analysis of the main building complex WWER-440/213 Paks for comparison of analytical and experimental results obtained by explosive testing (task 7a of workplan 95/96)

    International Nuclear Information System (INIS)

    1999-01-01

    Within the research programme on Benchmark studies of seismic analysis of WWER type reactors the blind pre-analysis must be prepared for the main building complex of Paks NPP, based on given excitations derived from explosion tests. The aim of the investigation was to validate different idealization concepts (mathematical models for the idealization of the structures and the soil) as well as investigation procedures (time domain and frequency domain analysis) and finally the software tools by comparing dynamic properties (eigenfrequencies, eigenmodes, modal values) and structural response results (time histories and response spectra). This report contains results of the blind pre-analysis performed by using three dimensional idealization of the main building complex (reactor building, turbine house, galleries) by means of time and frequency domian calculation procedures

  4. Blind pre-analysis of the main building complex WWER-1000 Kozloduy. Comparison of analytical and experimental results obtained by explosive testing (task 8a of workplan 96/97)

    International Nuclear Information System (INIS)

    Krutzik, N.J.

    1999-01-01

    In accordance with the 96/97 workplan of the Research Programme on 'Benchmark Studies for Seismic Analysis and Testing of WWER-Type Nuclear Power Plants', blind pre analyses were prepared for the main building complex of the WWER-1000 based on given excitations derived from explosive tests. The investigations were performed by several institutions based on various mathematical models and procedures for consideration of soil-structure interaction effects, but on the same explosive test input data recently obtained. The methods of calculation and software tools used will also be different. The aim of this investigation is to validate different idealization concepts (mathematical models for the idealization of the structures and the soil) as well as investigation procedures (time domain and frequency domain analysis) and finally the software tools by comparing structural response results (time histories and response spectra). This report contains the results of the blind pre analysis performed by Siemens using an equivalent beam model of the main building of the WWER 1000. The calculations were performed by means of a frequency domain calculation

  5. Pre- and post-test analyses of a KKL turbine trip test at 109% power using RETRAN-3D

    Energy Technology Data Exchange (ETDEWEB)

    Coddington, P

    2001-03-01

    As part of the PSI/HSK STARS project, pre-test calculations have been performed for a KKL turbine trip test at 109% power using the RETRAN-3D code. In this paper, we first present the results of these calculations, together with a description of the test and a comparison of the results with the measured plant data, and then discuss in more detail the differences between the pre-test results and the plant measurements, including the differences in the initial and boundary conditions, and how these differences influenced the calculated results. Finally, we comment on a series of post-test and sensitivity analyses, which were performed to resolve some of the discrepancies. The results of the pre-test (blind) calculations show good overall agreement with the experimental data, particularly for the maximum in the steam-line mass flow rate following the opening of the turbine bypass valves. This is of critical importance, since the steam-line flow has the least margin to the reactor scram limit. The agreement is especially good since the control rod banks used for the selected rod insertion were changed from those given in the test specification. Following a review of the comparison of the pre-test calculations with the measured data, several deficiencies in the RETRAN-3D model for KKL were identified and corrected as part of the post-test analysis. This allowed for both an improvement in the calculated results, and a deeper understanding of the behaviour of the turbine trip transient. (author)

  6. Pre- and post-test analyses of a KKL turbine trip test at 109% power using RETRAN-3D

    International Nuclear Information System (INIS)

    Coddington, P.

    2001-01-01

    As part of the PSI/HSK STARS project, pre-test calculations have been performed for a KKL turbine trip test at 109% power using the RETRAN-3D code. In this paper, we first present the results of these calculations, together with a description of the test and a comparison of the results with the measured plant data, and then discuss in more detail the differences between the pre-test results and the plant measurements, including the differences in the initial and boundary conditions, and how these differences influenced the calculated results. Finally, we comment on a series of post-test and sensitivity analyses, which were performed to resolve some of the discrepancies. The results of the pre-test (blind) calculations show good overall agreement with the experimental data, particularly for the maximum in the steam-line mass flow rate following the opening of the turbine bypass valves. This is of critical importance, since the steam-line flow has the least margin to the reactor scram limit. The agreement is especially good since the control rod banks used for the selected rod insertion were changed from those given in the test specification. Following a review of the comparison of the pre-test calculations with the measured data, several deficiencies in the RETRAN-3D model for KKL were identified and corrected as part of the post-test analysis. This allowed for both an improvement in the calculated results, and a deeper understanding of the behaviour of the turbine trip transient. (author)

  7. Main sequence mass loss

    International Nuclear Information System (INIS)

    Brunish, W.M.; Guzik, J.A.; Willson, L.A.; Bowen, G.

    1987-01-01

    It has been hypothesized that variable stars may experience mass loss, driven, at least in part, by oscillations. The class of stars we are discussing here are the δ Scuti variables. These are variable stars with masses between about 1.2 and 2.25 M/sub θ/, lying on or very near the main sequence. According to this theory, high rotation rates enhance the rate of mass loss, so main sequence stars born in this mass range would have a range of mass loss rates, depending on their initial rotation velocity and the amplitude of the oscillations. The stars would evolve rapidly down the main sequence until (at about 1.25 M/sub θ/) a surface convection zone began to form. The presence of this convective region would slow the rotation, perhaps allowing magnetic braking to occur, and thus sharply reduce the mass loss rate. 7 refs

  8. FERMILAB: Main Injector

    International Nuclear Information System (INIS)

    Anon.

    1993-01-01

    The Fermilab Main Injector (FMI) project is the centerpiece of the Laboratory's Fermilab III programme for the 1990s. Designed to support a luminosity of at least 5x10 31 cm -2 s -1 in the Tevatron collider, it will also provide new capabilities for rare neutral kaon decay and neutrino oscillation studies. The Fermilab Main Injector 8-150 GeV synchrotron is designed to replace the existing Main Ring which seriously limits beam intensities for the Tevatron and the antiproton production target. The project has passed several significant milestones and is now proceeding rapidly towards construction. The project received a $11.65M appropriation in 1992 and has been given $15M for the current fiscal year. Through the Energy Systems Acquisition Advisory Board (ESAAB) process, the US Department of Energy (DoE) has authorized funds for construction of the underground enclosure and service building where the Main Injector will touch the Tevatron, and to the preparation of bids for remaining project construction

  9. FERMILAB: Main Injector

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    1993-06-15

    The Fermilab Main Injector (FMI) project is the centerpiece of the Laboratory's Fermilab III programme for the 1990s. Designed to support a luminosity of at least 5x10{sup 31} cm{sup -2} s{sup -1} in the Tevatron collider, it will also provide new capabilities for rare neutral kaon decay and neutrino oscillation studies. The Fermilab Main Injector 8-150 GeV synchrotron is designed to replace the existing Main Ring which seriously limits beam intensities for the Tevatron and the antiproton production target. The project has passed several significant milestones and is now proceeding rapidly towards construction. The project received a $11.65M appropriation in 1992 and has been given $15M for the current fiscal year. Through the Energy Systems Acquisition Advisory Board (ESAAB) process, the US Department of Energy (DoE) has authorized funds for construction of the underground enclosure and service building where the Main Injector will touch the Tevatron, and to the preparation of bids for remaining project construction.

  10. Main facts 1995

    International Nuclear Information System (INIS)

    1996-01-01

    This report presents the main facts of the studies carried out by the Direction des Etudes et Recherches (DER) of Electricite de France: new applications of electricity, classical and nuclear thermal power plants, electrical equipment, environment protection, monitoring and plants operations

  11. Maine Bouguer Gravity Grid

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — A 2 kilometer Bouguer anomaly grid for the state of Maine. Number of columns is 197 and number of rows is 292. The order of the data is from the lower left to the...

  12. Main facts 1993

    International Nuclear Information System (INIS)

    1994-01-01

    This report presents the main facts of the studies carried out by the Direction des Etudes et Recherches (DER) of Electricite de France: new applications of electricity, classical and nuclear thermal power plants, electrical equipment, environment protection, network analysis, information and informatic equipment

  13. 17,20β-P and cortisol are the main in vitro metabolites of 17-hydroxy-progesterone produced by spermiating testes of Micropogonias furnieri (Desmarest, 1823 (Perciformes: Sciaenidae

    Directory of Open Access Journals (Sweden)

    Denise Vizziano Cantonnet

    Full Text Available The aim was to investigate the major C21 steroids produced by spermiating white croaker Micropogonias furnieri (Sciaenidae in order to establish the potential mediator of gamete maturation in males of this species. The testes steroid production at the spawning season was identified incubating the 3H-17-hydroxy-4-pregnene-3,20-dione precursor through thin layer chromatography, high pressure liquid chromatography, enzymatic oxydation, acetylation and immunochemistry analyses. 17,20β-Dihydroxy-4-pregnen-3-one (17,20β-P and 11β,17,21-Trihydroxy-4-pregnene-3,20-dione (cortisol were the main metabolites produced. Contrary to what we expected, 17,20β,21-Trihydroxy-4-pregnen-3-one was not detected. Circulating levels of 17,20β-P were undetectable in immature testes and in those at the first spermatogenesis stages, while a clear increase was observed during the whole spermatogenesis and spermiation phases (from undetectable to 1047 pg mL-1. In vitro studies together with plasma detection suggest that 17,20β-P is a good steroid candidate involved in M. furnieri testes maturation. The role of cortisol during late phases of testes development needs further studies.

  14. Renovating the Main Building

    CERN Multimedia

    CERN Bulletin

    2010-01-01

    CERN's "Main Building" is exactly that. The Organization's central hub, with hundreds of staff and visitors passing through its doors every day, will soon be getting a well-earned facelift. Refurbishment work will proceed in phases, starting with the Salle des Pas Perdus, the concourse between the Council Chamber and the Main Auditorium. By the end of August, informal seating areas will be installed, electronic display panels will provide practical information and improved sound insulation will enhance conditions in the auditoria and surrounding meeting rooms.   In light green the area that will undergo the facelift. Work will start in July. The ground floor is home to the entrance to Restaurant No. 1, the bank, the post office, the travel agent, the Users Office, the Staff Association, the notice boards etc. Step up to the first floor to access CERN's largest lecture theatre, the Council Chamber and its "Pas Perdus" lobby. Everyone who works at or visits CERN i...

  15. Fermilab Main Injector plan

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    1990-07-15

    The Fermilab Main Injector is the centrepiece of the 'Fermilab III' scheme to significantly upgrade the Laboratory's existing accelerator complex. The new accelerator is designed to provide increased particle beam levels to boost the collision rate in the Tevatron proton-antiproton collider (luminosity in excess of 5 x 10{sup 31} per sq cm per s) and, if approved, would provide increased flexibility in all areas of high energy physics research.

  16. Fermilab Main Injector plan

    International Nuclear Information System (INIS)

    Anon.

    1990-01-01

    The Fermilab Main Injector is the centrepiece of the 'Fermilab III' scheme to significantly upgrade the Laboratory's existing accelerator complex. The new accelerator is designed to provide increased particle beam levels to boost the collision rate in the Tevatron proton-antiproton collider (luminosity in excess of 5 x 10 31 per sq cm per s) and, if approved, would provide increased flexibility in all areas of high energy physics research

  17. Maine coast winds

    Energy Technology Data Exchange (ETDEWEB)

    Avery, Richard

    2000-01-28

    The Maine Coast Winds Project was proposed for four possible turbine locations. Significant progress has been made at the prime location, with a lease-power purchase contract for ten years for the installation of turbine equipment having been obtained. Most of the site planning and permitting have been completed. It is expect that the turbine will be installed in early May. The other three locations are less suitable for the project, and new locations are being considered.

  18. Project evaluation: main characteristics

    OpenAIRE

    Moutinho, Nuno

    2010-01-01

    — The evaluation process of real investment projects must consider not only the traditional financial approach, but also non-financial aspects. Non financial analysis can provide additional relevant information about projects. We investigate financial and non-financial areas most relevant in project appraisal. We present main critical success factors and areas of analysis that lead to the perception of project success. Finally, companies are segmented to verify its financial and non-financial...

  19. Marketing Maine Tablestock Potatoes

    OpenAIRE

    Berney, Gerald; Grajewski, Gregory; Hinman, Don; Prater, Marvin E.; Taylor, April

    2010-01-01

    The Marketing Services Division of USDA’s Agricultural Marketing Service (AMS) was asked by USDA’s Agricultural Research Service (ARS) National Program Leader and ARS’s New England Soil and Water Research Laboratory personnel to help with existing efforts to assist Maine fresh potato farmers in their search for alternative marketing strategies, and reverse the recent decline in the profitability of their operations. ARS researchers previously had conducted an exhaustive study defining possibl...

  20. Summary of main points

    International Nuclear Information System (INIS)

    2006-01-01

    In conjunction with its 6. annual meeting, the WPDD in close co-operation with the FSC held a Topical session on 'Stakeholder Involvement in Decommissioning' on November 14, 2005. The session was attended by 36 participants totally representing 14 NEA member countries and 2 international organisations. Two keynote addresses were given at the Topical Session. The first one treated of what is needed for robust decisions and how to bring all stakeholders into the debate. In the second keynote address a summary was made on what have been said on stakeholder involvement in decommissioning during earlier meetings of the WPDD. The main part of the session was then devoted to views from different stakeholders regarding their role and their involvement. This part contained viewpoints from local communities (Kaevlinge in Sweden and Port Hope in Canada), authorities (Scottish Executive and CSNC) and operators (EDF from France and EWN from Germany). Case studies from the decommissioning of Dounrey in the UK and from Trojan and Main Yankee in the USA were presented in the end part of the Topical session followed by a summary and lessons learnt report by the Rapporteur. A detailed programme of the Topical session can be seen in Appendix 1

  1. Preliminary observations of gate valve flow interruption tests, Phase 2

    International Nuclear Information System (INIS)

    Steele, R. Jr.; DeWall, K.G.

    1990-01-01

    This paper presents preliminary observations from the US Nuclear Regulatory Commission/Idaho National Engineering Laboratory Flexible Wedge Gate Valve Qualification and Flow Interruption Test Program, Phase 2. The program investigated the ability of selected boiling water reactor (BWR) process line valves to perform their containment isolation function at high energy pipe break conditions and other more normal flow conditions. The fluid and valve operating responses were measured to provide information concerning valve and operator performance at various valve loadings so that the information could be used to assess typical nuclear industry motor operator sizing equations. Six valves were tested, three 6-in. isolation valves representative of those used in reactor water cleanup systems in BWRs and three 10-in. isolation valves representative of those used in BWR high pressure coolant injection (HPCI) steam lines. The concern with these normally open isolation valves is whether they will close in the event of a downstream pipe break outside of containment. The results of this testing will provide part of the technical insights for NRC efforts regarding Generic Issue 87 (GI-87), Failure of the HPCI Steam Line Without Isolation, which includes concerns about the uncertainties in gate valve motor operator sizing and torque switch settings for these BWR containment isolation valves. As of this writing, the Phase 2 test program has just been completed. Preliminary observations made in the field confirmed most of the results from the Phase 1 test program. All six valves closing in high energy water, high energy steam, and high pressure cold water require more force to close than would be calculated using the typical variables in the standard industry motor operator sizing equations

  2. TRIGA reactor main systems

    International Nuclear Information System (INIS)

    Boeck, H.; Villa, M.

    2007-01-01

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

  3. JWST-MIRI spectrometer main optics design and main results

    Science.gov (United States)

    Navarro, Ramón; Schoenmaker, Ton; Kroes, Gabby; Oudenhuysen, Ad; Jager, Rieks; Venema, Lars

    2017-11-01

    MIRI ('Mid InfraRed Instrument') is the combined imager and integral field spectrometer for the 5-29 micron wavelength range under development for the James Webb Space Telescope JWST. The flight acceptance tests of the Spectrometer Main Optics flight models (SMO), part of the MIRI spectrometer, are completed in the summer of 2008 and the system is delivered to the MIRI-JWST consortium. The two SMO arms contain 14 mirrors and form the MIRI optical system together with 12 selectable gratings on grating wheels. The entire system operates at a temperature of 7 Kelvin and is designed on the basis of a 'no adjustments' philosophy. This means that the optical alignment precision depends strongly on the design, tolerance analysis and detailed knowledge of the manufacturing process. Because in principle no corrections are needed after assembly, continuous tracking of the alignment performance during the design and manufacturing phases is important. The flight hardware is inspected with respect to performance parameters like alignment and image quality. The stability of these parameters is investigated after exposure to various vibration levels and successive cryogenic cool downs. This paper describes the philosophy behind the acceptance tests, the chosen test strategy and reports the results of these tests. In addition the paper covers the design of the optical test setup, focusing on the simulation of the optical interfaces of the SMO. Also the relation to the SMO qualification and verification program is addressed.

  4. 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

  5. Test

    DEFF Research Database (Denmark)

    Bendixen, Carsten

    2014-01-01

    Bidrag med en kortfattet, introducerende, perspektiverende og begrebsafklarende fremstilling af begrebet test i det pædagogiske univers.......Bidrag med en kortfattet, introducerende, perspektiverende og begrebsafklarende fremstilling af begrebet test i det pædagogiske univers....

  6. Space Shuttle main engine product improvement

    Science.gov (United States)

    Lucci, A. D.; Klatt, F. P.

    1985-01-01

    The current design of the Space Shuttle Main Engine has passed 11 certification cycles, amassed approximately a quarter million seconds of engine test time in 1200 tests and successfully launched the Space Shuttle 17 times of 51 engine launches through May 1985. Building on this extensive background, two development programs are underway at Rocketdyne to improve the flow of hot gas through the powerhead and evaluate the changes to increase the performance margins in the engine. These two programs, called Phase II+ and Technology Test Bed Precursor program are described. Phase II+ develops a two-tube hot-gas manifold that improves the component environment. The Precursor program will evaluate a larger throat main combustion chamber, conduct combustion stability testing of a baffleless main injector, fabricate an experimental weld-free heat exchanger tube, fabricate and test a high pressure oxidizer turbopump with an improved inlet, and develop and test methods for reducing temperature transients at start and shutdown.

  7. Achados principais de exames laboratoriais no diagnóstico de apendicite aguda: uma avaliação prospectiva Main findings in laboratory tests diagnosis of acute appendicitis: a prospective evaluation

    Directory of Open Access Journals (Sweden)

    Rafael Nunes Goulart

    2012-06-01

    Full Text Available RACIONAL: Apendicite aguda é a doença abdominal cirúrgica mais comum nas unidades de emergência. Embora o diagnóstico seja clínico, a realização de exames complementares pode ser útil na dúvida diagnóstica. OBJETIVO: Avaliar as principais alterações de exames laboratoriais em pacientes com apendicite aguda, assim como sua relação com a fase evolutiva da doença. MÉTODOS: Avaliação prospectiva de pacientes com diagnóstico de apendicite aguda submetidos ao tratamento cirúrgico. RESULTADOS: Cento e setenta e nove pacientes participaram deste estudo, a maioria do sexo masculino. A idade média foi de 26 anos. Em relação à contagem de leucócitos, 46,9% apresentavam valores BACKGROUND: Acute appendicitis is the most common surgical abdominal disease in the emergency room. Although the diagnosis is clinical the complementary tests may be useful in doubt. AIM: To evaluate the main laboratory tests in patients with acute appendicitis, as well as its relationship with the evolutionary stage of the disease. METHODS: Prospective evaluation of patients with acute appendicitis who underwent surgical treatment. RESULTS: A total of 179 patients participated in this study, most were male. The mean age was 26 years. For leukocyte count 46.9% had values ​​<15.000mm3. The mean percentage of polymorphonuclear cells was 81,7%, 1,2% of sticks, 1% eosinophils, lymphocytes 12,8% and 2,9% monocytes. C-reactive protein was required for 54 patients. It was <10 mg/dl in 19, between 10 and 50 mg/dl in 24 and greater than or equal to 50 mg/dl in 11. Regarding the evolutionary phase 64% patients had early stage (stages 1 and 2, 16,2% stage 3 and 35 stage 4. A total of 57% of patients with white blood cell count greater than or equal to 20.000/mm3 had appendicular perforation (p<0,05. The percentage of polymorphonuclear leukocytes from patients with early stages was lower than the later stages (79,8% and 85,1%, respectively, with p<0,05. Patients

  8. Main challenges of residential areas

    Directory of Open Access Journals (Sweden)

    Oana Luca

    2017-06-01

    Full Text Available The present article is a position paper aiming to initiate a professional debate related to the aspects related to the urban dysfunctions leading to the wear of the residential areas. The paper proposes a definition of the wear process, identify the main causes leading to its occurrence and propose a number of solutions to neutralise the dysfunctions. The three wearing phases of residential areas components are emphasized, exploring their lifecycle. In order to perform the study of urban wear, the status of the residential areas components can be established and monitored, and also the variables of the function that can mathematically model the specific wear process may be considered. The paper is considered a first step for the model adjustment, to be tested and validated in the following steps. Based on the mathematical method and model, there can be created, in a potential future research, the possibility of determining the precarity degree for residential areas/neighbourhoods and cities, by minimising the subjective component of the analyses preceding the decision for renovation or regeneration.

  9. Main heat transfer components for SNR-300

    International Nuclear Information System (INIS)

    De Haas Van Dorsser, A.H.

    1976-01-01

    Early in the joint German-Belgium-Dutch fast breeder programme it was decided that all main components should be tested, if possible at full scale, before fabrication of the actual SNR-300 components. Descriptions are given of the results of testing, and subsequent modifications, of the pumps, intermediate heat exchangers, and steam generators. A full scale model of the primary pump, free surface vertical shaft centrifugal type, was constructed and tested in the 5000 cubic metres per hour pump test facility erected at Bensberg. A 70 MW model of an intermediate heat exchanger, straight tube type with floating head, was tested in the 50 MW steam generator test station at Hengelo. Also tested in the Hengelo facility was an almost full scale straight tube 50 MW steam generator and subsequently a 50 MW helical tube evaporator. The latter tests were of more than 3000 h operation and resulted in minor changes in design and manufacturing operation. (U.K.)

  10. Major Achievements and Prospect of the ATLAS Integral Effect Tests

    International Nuclear Information System (INIS)

    Choi, K.; Kim, Y.; Song, C.; Baek, W.

    2012-01-01

    A large-scale thermal-hydraulic integral effect test facility, ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation), has been operated by KAERI. The reference plant of ATLAS is the APR1400 (Advanced Power Reactor, 1400 MWe). Since 2007, an extensive series of experimental works were successfully carried out, including large break loss of coolant accident tests, small break loss of coolant accident tests at various break locations, steam generator tube rupture tests, feed line break tests, and steam line break tests. These tests contributed toward an understanding of the unique thermal-hydraulic behavior, resolving the safety-related concerns and providing validation data for evaluation of the safety analysis codes and methodology for the advanced pressurized water reactor, APR1400. Major discoveries and lessons found in the past integral effect tests are summarized in this paper. As the demand for integral effect tests is on the rise due to the active national nuclear R and D program in Korea, the future prospects of the application of the ATLAS facility are also discussed.

  11. Additively Manufactured Main Fuel Valve Housing

    Science.gov (United States)

    Eddleman, David; Richard, Jim

    2015-01-01

    Selective Laser Melting (SLM) was utilized to fabricate a liquid hydrogen valve housing typical of those found in rocket engines and main propulsion systems. The SLM process allowed for a valve geometry that would be difficult, if not impossible to fabricate by traditional means. Several valve bodies were built by different SLM suppliers and assembled with valve internals. The assemblies were then tested with liquid nitrogen and operated as desired. One unit was also burst tested and sectioned for materials analysis. The design, test results, and planned testing are presented herein.

  12. Seismic testing

    International Nuclear Information System (INIS)

    Sollogoub, Pierre

    2001-01-01

    This lecture deals with: qualification methods for seismic testing; objectives of seismic testing; seismic testing standards including examples; main content of standard; testing means; and some important elements of seismic testing

  13. Diagnosing in building main pipelines

    Energy Technology Data Exchange (ETDEWEB)

    Telegin, L.G.; Gorelov, A.S.; Kurepin, B.N.; Orekhov, V.I.; Vasil' yev, G.G.; Yakovlev, Ye. I.

    1984-01-01

    General principles are examined for technical diagnosis in building main pipelines. A technique is presented for diagnosis during construction, as well as diagnosis of the technical state of the pipeline-construction machines and mechanisms. The survey materials could be used to set up construction of main pipelines.

  14. Maine Agricultural Foods. Project SEED.

    Science.gov (United States)

    Beaulieu, Peter; Ossenfort, Pat

    This paper describes an activity-based program that teaches students in grades 4-12 about the importance of Maine agriculture in their lives. Specifically, the goal is to increase student awareness of how the foods they eat are planted, harvested, and processed. The emphasis is on crops grown in Maine such as potatoes, broccoli, peas, blueberries,…

  15. Left main percutaneous coronary intervention.

    Science.gov (United States)

    Teirstein, Paul S; Price, Matthew J

    2012-10-23

    The introduction of drug-eluting stents and advances in catheter techniques have led to increasing acceptance of percutaneous coronary intervention (PCI) as a viable alternative to coronary artery bypass graft (CABG) for unprotected left main disease. Current guidelines state that it is reasonable to consider unprotected left main PCI in patients with low to intermediate anatomic complexity who are at increased surgical risk. Data from randomized trials involving patients who are candidates for either treatment strategy provide novel insight into the relative safety and efficacy of PCI for this lesion subset. Herein, we review the current data comparing PCI with CABG for left main disease, summarize recent guideline recommendations, and provide an update on technical considerations that may optimize clinical outcomes in left main PCI. Copyright © 2012 American College of Cardiology Foundation. Published by Elsevier Inc. All rights reserved.

  16. Main: FEB3 [TP Atlas

    Lifescience Database Archive (English)

    Full Text Available nt to sterilization and rinsing - One of the main components of biofilms is polysaccharides - Some pit-formi...ng bacteria such as Sphingomonas species A1 possess superchannels that directly incorporate and decompose polysaccharides - Detai...e entrance of the superchannel have been elucidated - We have obtained the crystals of ABC importer complexe...of water pipes and dental plaque are examples of biofilms. One of the main components of biofilms is polysac

  17. Decontamination of main coolant pumps

    International Nuclear Information System (INIS)

    Roofthooft, R.

    1988-01-01

    Last year a number of main coolant pumps in Belgian nuclear power plants were decontaminated. A new method has been developed to reduce the time taken for decontamination and the volume of waste to be treated. The method comprises two phases: Oxidation with permanganate in nitric acid and dissolution in oxalic acid. The decontamination of main coolant pumps can now be achieved in less than one day. The decontamination factors attained range between 15 and 150. (orig.) [de

  18. New Main Ring control system

    International Nuclear Information System (INIS)

    Seino, K.; Anderson, L.; Ducar, R.; Franck, A.; Gomilar, J.; Hendricks, B.; Smedinghoff, J.

    1990-03-01

    The Fermilab Main Ring control system has been operational for over sixteen years. Aging and obsolescence of the equipment make the maintenance difficult. Since the advent of the Tevatron, considerable upgrades have been made to the controls of all the Fermilab accelerators except the Main Ring. Modernization of the equipment and standardization of the hardware and software have thus become inevitable. The Tevatron CAMAC serial system has been chosen as a basic foundation in order to make the Main Ring control system compatible with the rest of the accelerator complex. New hardware pieces including intelligent CAMAC modules have been designed to satisfy unique requirements. Fiber optic cable and repeaters have been installed in order to accommodate new channel requirements onto the already saturated communication medium system. 8 refs., 2 figs

  19. Main: FBB2 [TP Atlas

    Lifescience Database Archive (English)

    Full Text Available ion of the c-ring - A subunit packing model of E. coli c-ring has been proposed - The main chain secondary s...tructure of thermophile c-ring has been obtained ATP synthase is a general term for an enzyme that can synth

  20. Main: FEA5 [TP Atlas

    Lifescience Database Archive (English)

    Full Text Available al or an anti-cancer drug, is the main cause of hospital-acquired infection - Dru...e will elucidate the entire structure of the transport machinery in action to understand its functions in detail. FEA5.csml ...

  1. Residential Energy Efficiency Potential: Maine

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, Eric J [National Renewable Energy Laboratory (NREL), Golden, CO (United States)

    2017-11-02

    Energy used by Maine single-family homes that can be saved through cost-effective improvements. Prepared by Eric Wilson and Noel Merket, NREL, and Erin Boyd, U.S. Department of Energy Office of Energy Policy and Systems Analysis.

  2. CENTRE OF THE MAIN INTERESTS

    Directory of Open Access Journals (Sweden)

    DIANA DELEANU

    2013-05-01

    Full Text Available The centre of the main interests of the debtor is a legal tool meant to settle conflicts that can arise between jurisdictions in cross-border insolvencies, based on the principles of mutual recognition and co-operation.

  3. ONKALO - Main drawings in 2007

    International Nuclear Information System (INIS)

    2008-05-01

    The first overall site characterisation programme for a Finnish repository of spent nuclear fuel was introduced in 1982. This programme already suggested that the site confirmation for a detailed repository design and safety assessment should include characterisation performed in an underground rock characterisation facility (URCF). This idea was confirmed during the detailed site characterisation. International views have also emphasised the importance of underground characterisation before the final decision to construct the repository is taken. The underground rock characterisation facility (ONKALO) is excavated at Olkiluoto in the municipality of Eurajoki. ONKALO should be constructed to allow characterisation work for site confirmation without jeopardising long-term safety of the repository site. It should also be possible to link ONKALO later to the repository as to a part of it. The construction of ONKALO was started in 2004 and will be completed in 2014. The characterisation work has started in ONKALO and will focus on the disposal depth. In the main drawings stage, ONKALO was described at the level of detail needed for a construction permit in 2003. This meant description of the location, final structures and final systems. This summary report describes the development of design to updated main drawings in 2007 at the same level of detail (no temporary arrangements are described). The main changes are the added exhaust air shaft and advancing the controlled area's inlet air shaft to the ONKALO phase. Also the layout and the depth of the characterisation levels have been updated according to the current bedrock information. Some buildings on the surface will house sets of equipment directly connected with underground facility and this equipment is described in this report. No buildings or other equipment are described in this report, because they are not directly connected with the underground facility. The main element of ONKALO is a system of

  4. A main sequence for quasars

    Science.gov (United States)

    Marziani, Paola; Dultzin, Deborah; Sulentic, Jack W.; Del Olmo, Ascensión; Negrete, C. A.; Martínez-Aldama, Mary L.; D'Onofrio, Mauro; Bon, Edi; Bon, Natasa; Stirpe, Giovanna M.

    2018-03-01

    The last 25 years saw a major step forward in the analysis of optical and UV spectroscopic data of large quasar samples. Multivariate statistical approaches have led to the definition of systematic trends in observational properties that are the basis of physical and dynamical modeling of quasar structure. We discuss the empirical correlates of the so-called “main sequence” associated with the quasar Eigenvector 1, its governing physical parameters and several implications on our view of the quasar structure, as well as some luminosity effects associated with the virialized component of the line emitting regions. We also briefly discuss quasars in a segment of the main sequence that includes the strongest FeII emitters. These sources show a small dispersion around a well-defined Eddington ratio value, a property which makes them potential Eddington standard candles.

  5. A Main Sequence for Quasars

    Directory of Open Access Journals (Sweden)

    Paola Marziani

    2018-03-01

    Full Text Available The last 25 years saw a major step forward in the analysis of optical and UV spectroscopic data of large quasar samples. Multivariate statistical approaches have led to the definition of systematic trends in observational properties that are the basis of physical and dynamical modeling of quasar structure. We discuss the empirical correlates of the so-called “main sequence” associated with the quasar Eigenvector 1, its governing physical parameters and several implications on our view of the quasar structure, as well as some luminosity effects associated with the virialized component of the line emitting regions. We also briefly discuss quasars in a segment of the main sequence that includes the strongest FeII emitters. These sources show a small dispersion around a well-defined Eddington ratio value, a property which makes them potential Eddington standard candles.

  6. At ISR Main Control Room

    CERN Multimedia

    1983-01-01

    After 13 years the exploitation of the Intersecting Storage Rings as a beam-beam collider went to an end. In this last year the demands were very exacting, both in terms of operating time and diversified running conditions (Annual Report 1983 p. 123). Before dismantelement the photographer made a last tour, see photos 8310889X --> 8310667X. This photo shows the Main Control Room.

  7. Physical data generation methodology for return-to-power steam line break analysis

    Energy Technology Data Exchange (ETDEWEB)

    Zee, Sung Kyun; Lee, Chung Chan; Lee, Chang Kue [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-02-01

    Current methodology to generate physics data for steamline break accident analysis of CE-type nuclear plant such as Yonggwang Unit 3 is valid only if the core reactivity does not reach the criticality after shutdown. Therefore, the methodology requires tremendous amount of net scram worth, specially at the end of the cycle when moderator temperature coefficient is most negative. Therefore, we need a new methodology to obtain reasonably conservation physics data, when the reactor returns to power condition. Current methodology used ROCS which include only closed channel model. But it is well known that the closed channel model estimates the core reactivity too much negative if core flow rate is low. Therefore, a conservative methodology is presented which utilizes open channel 3D HERMITE model. Current methodology uses ROCS which include only closed channel model. But it is well known that the closed channel model estimates the core reactivity too much negative if core flow rate is low. Therefore, a conservative methodology is presented which utilizes open channel 3D HERMITE model. Return-to-power reactivity credit is produced to assist the reactivity table generated by closed channel model. Other data includes hot channel axial power shape, peaking factor and maximum quality for DNBR analysis. It also includes pin census for radiological consequence analysis. 48 figs., 22 tabs., 18 refs. (Author) .new.

  8. Main building fire drill safely concluded

    CERN Document Server

    CERN Bulletin

    2015-01-01

    Last week, a simulated fire in the stairwell of the Main Building put CERN’s emergency response procedures to the test.   Firefighters descend the stairwell in the Main Building as the simulated fire rises.   At 2 p.m. on 22 September, alarms sounded around CERN’s Main Building as an evacuation exercise got underway. A simulated fire in the  stairwell, complete with very realistic smoke, led to the evacuation of one of the busiest places at CERN. The Main Building complex includes the Carlson Wagonlit travel agency, the post office, UBS, Uniqa, the Users Office, the Staff Association and the Novae restaurant as well as the Main Auditorium, the Council Chamber and the Charpak meeting room. It was impressive to see how quickly the smoke propagated in the staircase as well as into the corridors, and equally impressive to see how smoothly, quickly and efficiently the evacuation proceeded. The...

  9. Main ring transition crossing simulations

    International Nuclear Information System (INIS)

    Kourbanis, I.; Ng, King-Yuen.

    1990-10-01

    We used ESME to simulate transition crossing in the Main Ring (MR). For the simulations, we followed the MR 29 cycle used currently for bar p production with a flat top of 120 GeV. In Sect. II, some inputs are discussed. In Sect. III, we present simulations with space charge turned off so that the effect of nonlinearity can be studied independently. When space charge is turned on in Sect. IV, we are faced with the problem of statistical errors due to binning, an analysis of which is given in the Appendices. Finally in Sects. V and VI, the results of simulations with space charge are presented and compared with the experimental measurements. 7 refs., 6 figs

  10. Improvement of main control room

    International Nuclear Information System (INIS)

    Chae, Sung Ki; Ham, Chang Sik; Kwon, Ki Chun

    1991-07-01

    Information display system, advanced alarm system and fiber optical communication system were developed to improve the main control room in nuclear power plant. Establishing the new hierachical information structure of plant operation data, plant overview status board(POSB) and digital indicator(DI) were designed and manufactured. The prototype advanced alarm system which employed the new alarm logics and algorithm compared with the conventional alarm system were developed and its effectiveness was proved. Optical communication system which has multi-drop feature and capability of upgrading to large-scale system by using BITBUS communication protocol which is proven technology, were developed. Reliability of that system was enhanced by using distributed control. (Author)

  11. The role of Main Institutions

    DEFF Research Database (Denmark)

    Persson, H Thomas R; Chabanet, Didier; Rakar, Fredrik

    2017-01-01

    ), in many countries the need emerged to understand the best methods to promote their establishment and continued success. In order to understand these issues, to contribute to the academic debate on SEs and to give useful policy advice on a truly enabling ecosystem, in November 2013 a consortium of 11...... Entrepreneurship”; to identify the “New Generation” of Social Entrepreneurs; to build an “Evolutionary Theory of Social Entrepreneurship”; to provide effective policy advices to stakeholders. In order to pursue and achieve these research objectives, the consortium implemented a complex research design...... in the social economy; - In the fifth chapter the authors address the role of the main institutions in developing (or hindering) social enterprises; - In the sixth chapter, stakeholder network maps are used to identify four ‘ecosystem types’ across the 10 partner countries; - The seventh chapter gives...

  12. Live insertion method used for main renewal

    International Nuclear Information System (INIS)

    Solkowitz, M.

    1992-01-01

    Baltimore Gas and Electric's pilot project using the live insertion method to replace a cast iron main provided excellent results. Its use on Eastern Avenue, a major state highway, was cost effective, provided gas service to customers during the work, required relatively short construction time and resulted in only minor traffic disruptions. Gas service transfers to the new main were done at customer convenience and resulted in outages of only a few hours per customer. This paper reports that the project involved inserting a 6-in. plastic line inside an existing 10-in. cast iron main. Miller Pipeline Corp., Indianapolis, supplier of the Insertec left-angle R right-angle live insertion method was contracted for the job. Miller technicians assisted BG and E forces by providing a load analysis of the main, a pushing machine and related supplies, foaming equipment and pipe cutting tools. Company forces were responsible for all preparatory work, including opening all excavations, installing bypasses, and fusing and testing the plastic pipe. Service transfers and renewals were also completed by company employees

  13. Challenges of the ILC Main Linac

    International Nuclear Information System (INIS)

    Ross, Marc

    2007-01-01

    With the completion of the ILC Reference Design Report (RDR), we begin the next phase of the project - development of the Engineering Design. Our strategy and priorities come from the identification, contained in the RDR, of scientific and engineering challenges of the ILC. First among these is the cost of the main linac which, including the associated earthworks and cooling/power systems, amounts to 60% of the ILC total cost. Next is the challenge to reach the highest practical gradient since this R and D has the largest cost leverage of any of the ongoing programs. Finally, we have to understand the beam dynamics and beam tuning processes in the main linac, as we will not have the opportunity to do full (or even large) scale tests of these before the linac is constructed.

  14. The LHC Main Quadrupoles during Series Fabrication

    CERN Document Server

    Tortschanoff, Theodor; Durante, M; Hagen, P; Klein, U; Krischel, D; Payn, A; Rossi, L; Schellong, B; Schmidt, P; Simon, F; Schirm, K-M; Todesco, E

    2006-01-01

    By the end of August 2005 about 320 of the 400 main LHC quadrupole magnets have been fabricated and about 220 of them assembled into their cold masses, together with corrector magnets. About 130 of them have been cold tested in their cryostats and most of the quadrupoles exceeded their nominal excitation, i.e. 12,000 A, after no more than two training quenches. During this series fabrication, the quality of the magnets and cold masses was thoroughly monitored by means of warm magnetic field measurements, of strict geometrical checking, and of various electrical verifications. A number of modifications were introduced in order to improve the magnet fabrication, mainly correction of the coil geometry for achieving the specified field quality and measures for avoiding coil insulation problems. Further changes concern the electrical connectivity and insulation of instrumentation, and of the corrector magnets inside the cold masses. The contact resistances for the bus-bar connections to the quench protection diode...

  15. The double main sequence of Omega Centauri

    Science.gov (United States)

    Bedin, L. R.; Piotto, G.; Anderson, J.; King, I. R.; Cassisi, S.; Momany, Y.

    Recent, high precision photometry of Omega Centauri, the biggest Galactic globular cluster, has been obtained with Hubble Space Telescope (HST). The color magnitude diagram reveals an unexpected bifurcation of colors in the main sequence (MS). The newly found double MS, the multiple turnoffs and subgiant branches, and other sequences discovered in the past along the red giant branch of this cluster add up to a fascinating but frustrating puzzle. Among the possible explanations for the blue main sequence an anomalous overabundance of helium is suggested. The hypothesis will be tested with a set of FLAMES@VLT data we have recently obtained (ESO DDT program), and with forthcoming ACS@HST images. Based on observations with the NASA/ESA Hubble Space Telescope, obtained at the Space Telescope Science Institute, which is operated by AURA, Inc., under NASA contract NAS 5-26555.

  16. Main injector synchronous timing system

    International Nuclear Information System (INIS)

    Blokland, W.; Steimel, J.

    1998-01-01

    The Synchronous Timing System is designed to provide sub-nanosecond timing to instrumentation during the acceleration of particles in the Main Injector. Increased energy of the beam particles leads to a small but significant increase in speed, reducing the time it takes to complete a full turn of the ring by 61 nanoseconds (or more than 3 rf buckets). In contrast, the reference signal, used to trigger instrumentation and transmitted over a cable, has a constant group delay. This difference leads to a phase slip during the ramp and prevents instrumentation such as dampers from properly operating without additional measures. The Synchronous Timing System corrects for this phase slip as well as signal propagation time changes due to temperature variations. A module at the LLRF system uses a 1.2 Gbit/s G-Link chip to transmit the rf clock and digital data (e.g. the current frequency) over a single mode fiber around the ring. Fiber optic couplers at service buildings split off part of this signal for a local module which reconstructs a synchronous beam reference signal. This paper describes the background, design and expected performance of the Synchronous Timing System. copyright 1998 American Institute of Physics

  17. Main injector synchronous timing system

    International Nuclear Information System (INIS)

    Blokland, Willem; Steimel, James

    1998-01-01

    The Synchronous Timing System is designed to provide sub-nanosecond timing to instrumentation during the acceleration of particles in the Main Injector. Increased energy of the beam particles leads to a small but significant increase in speed, reducing the time it takes to complete a full turn of the ring by 61 nanoseconds (or more than 3 rf buckets). In contrast, the reference signal, used to trigger instrumentation and transmitted over a cable, has a constant group delay. This difference leads to a phase slip during the ramp and prevents instrumentation such as dampers from properly operating without additional measures. The Synchronous Timing System corrects for this phase slip as well as signal propagation time changes due to temperature variations. A module at the LLRF system uses a 1.2 Gbit/s G-Link chip to transmit the rf clock and digital data (e.g. the current frequency) over a single mode fiber around the ring. Fiber optic couplers at service buildings split off part of this signal for a local module which reconstructs a synchronous beam reference signal. This paper describes the background, design and expected performance of the Synchronous Timing System

  18. Main technical topics in 1999

    International Nuclear Information System (INIS)

    2000-01-01

    This Safety Authority annual report strives to present current organizational provisions and future trends in nuclear safety supervision in France and to describe the most outstanding occurrences during the past year. A first part presents nine documents concerning the main topics of 1999: aging of nuclear installations, the Offsite Emergency Plans (PPI), the impact of nuclear activities on man and the environment, the criticality hazards, EDF in 1999, the EPR project, the Andra in 1999, the transport incidents, the nuclear safety in eastern Europe. The second part presents the missions and actions of the Nuclear Installations Safety in the domains of the liabilities, the organization of the nuclear safety control, the regulations of the INB, the public information, the international relations, the crisis management, the radioactive materials transportation, the radioactive wastes. The equipment, the radiation protection and the exploitation of the pressurized water reactors are also treated just as the experimental reactors, the fuel cycle installations and the the nuclear installations dismantling. (A.L.B.)

  19. Research in auditing: main themes

    Directory of Open Access Journals (Sweden)

    Marcelo Porte

    Full Text Available ABSTRACT The passage of the Sarbanes-Oxley Act (SOX was a turning point in auditing and in auditors practice for the academic world. Research concerning the characterization of academic production related to auditing is in its third decade. Its analysis is accomplished by means of definition of keywords, abstracts or title, and information on thematic association within the academic production itself in auditing is undisclosed. In order to revise this gap in auditing literature, this study identified the main themes in auditing and their association in post-SOX era by analyzing the content of objectives and hypothesis of 1,650 publications in Web of Science (2002-2014. The findings in this study extended those from the study by Lesage and Wechtler (2012 from 16 auditing thematic typologies to 22. The results demonstrate that the themes audit report & financial statement users, corporate governance, audit market, external audit, socio-economic data of the company, international regulation, and fraud risk & audit risk were the most addressed in the publications about auditing. Corporate governance has a broader association with the other themes in the area. Future researches may use these themes and relate them to the methodologies applied to audit studies.

  20. Description of the Main Features of the Series Production of the LHC Main Dipole Magnets

    CERN Document Server

    Savary, F; Chevret, P; de Rijk, G; Fessia, P; Liénard, P; Miles, J; Modena, M; Rossi, L; Tommasini, D; Vlogaert, J; Bresson, D; Grunblatt, G; Decoene, JF; Bressani, F; Drago, G; Gagliardi, P; Eysselein, F; Gärtner, W; Lublow, P

    2008-01-01

    The series production of the LHC main dipole magnets was completed in November 2006. This paper presents the organization implemented at CERN and the milestones fixed to fullfil the technical requirements and to respect the master schedule of the machine installation. The CERN organization for the production follow-up, the quality assurance and the magnet testing, as well as the organization of the three main contractors will be described. A description of the design work and procurement of most of the specific heavy tooling and key components will be given with emphasis on the advantages and drawbacks.

  1. Testing Testing Testing.

    Science.gov (United States)

    Deville, Craig; O'Neill, Thomas; Wright, Benjamin D.; Woodcock, Richard W.; Munoz-Sandoval, Ana; Gershon, Richard C.; Bergstrom, Betty

    1998-01-01

    Articles in this special section consider (1) flow in test taking (Craig Deville); (2) testwiseness (Thomas O'Neill); (3) test length (Benjamin Wright); (4) cross-language test equating (Richard W. Woodcock and Ana Munoz-Sandoval); (5) computer-assisted testing and testwiseness (Richard Gershon and Betty Bergstrom); and (6) Web-enhanced testing…

  2. Evaluation of main control room habitability in Japanese LWR (1). Outline of evaluation method and conditions

    International Nuclear Information System (INIS)

    Fujita, Yuko; Yoneda, Jiro; Okabayashi, Kazuki; Fukuda, Ryo

    2009-01-01

    It has been recognized that main control room habitability (CRH) is very important safety item. Its evaluation method and conditions was recently updated as a common method both for Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) in Japan. This paper describes the common procedure for CRH evaluation. Postulated accidents for CRH evaluation are Loss of Coolant Accident (LOCA) for BWR and PWR, Main Steam Line Break Accident(MSLBA) for BWR, and Steam Generator Tube Rupture(SGTR) for PWR. Evaluation period is thirty days after each accident occurs. Acceptable criterion for each operator is 100mSv as total effective dose equivalent for thirty days, considering the net staying period of an operator in a control room and frequency of alternation of working staff. Total effective dose equivalent is calculated not only for an operator in control room, but also for an operator outside a building for alternation, and then total effective dose is compared with the criterion. The routes for dose evaluation are classified into five patterns, covering three patterns of location of radioactive source (i.e. source inside a building, source released into air, and source leaked into a control room) and covering two patterns of location of an operator (i.e. in control room and outside a building for alternation). Total effective dose by a route for source which is released into air and leaked into a control room is typically the largest contributor, in case of PWR LOCA evaluation, as an example showing relatively severe result. Evaluated dose for this dominant route is significantly affected by two calculations. One is a source term for calculation of radioactive gas release and the other is dispersion calculation in the air for concentration of radioactive gas in the vicinity of a control room. For LOCA 100% Kr,Xe and 50% iodine of total core radionuclide inventory are assumed to be released immediately into containment vessel. This is the most conservative assumption

  3. Tinkering at the main-ring lattice

    Energy Technology Data Exchange (ETDEWEB)

    Ohnuma, S.

    1982-08-23

    To improve production of usable antiprotons using the proton beam from the main ring and the lossless injection of cooled antiprotons into the main ring, modifications of the main ring lattice are recommended.

  4. Seismic risks at Elsie Lake Main Dam

    International Nuclear Information System (INIS)

    McCammon, N.R.; Momenzadeh, M.; Hawson, H.H.; Nielsen, N.M.

    1991-01-01

    The Elsie Lake dams are located on Vancouver Island in an area of high seismic risk. A safety review in 1986 indicated potential deficiencies in the earthfill main dam with respect to modern earthquake design standards. A detailed field investigation program comprising drilling and penetration tests was carried out and the results used in an assessment of seismic stability. A 0.8 m thick less dense layer in the granular shell of the dam, possibly caused by wet construction conditions, would likely liquefy in a major earthquake but sufficient residual strength would likely remain to prevent catastrophic failure. The dam shell might undergo some distortion, and an assessment was initiated to determine the requirements for reservoir drawdown following an extreme earthquake to ensure the timely lowering of the reservoir for inspection and repair. It was suggested that an adequate evacuation capability would be 25% and 50% drawdown in not more than 30 and 50 days, respectively. 9 refs., 11 figs., 1 tab

  5. EPR compared to international requirements (Mainly EUR)

    International Nuclear Information System (INIS)

    Broecker, B.

    1996-01-01

    A number of European Utilities have entered an agreement to write common requirements dedicated to future light water nuclear power plants to be built in Europe. The activities are known under the sign EUR (European Utilities Requirements). EPR, the future European Pressurized water Reactor, is the first installation of this type which will be operational from the year 2000 onwards, must fulfill the European requirements. EPR will serve as a test whether these requirements are realistic and well balanced. At the basic design stage of EPR, this paper concentrates on four main topics: the requirements which are new compared with existing reactors and which put a major challenge to the designer; the requirements today still open and the way they can be met by the EPR or not; the points for which already today the EPR special requirements exceed the EUR; the examples where the design of the EPR has given feedback which has led to a change of the EUR. EPR and EUR are different approaches to the reactor of the future. EUR is a set of requirements which leaves a flexibility to the designer while EPR is a real project which defines the technical solutions. EPR will fulfill the EUR and will at the same time serve as a test whether these requirements are realistic. EPR will also fulfill international requirements with minor changes. (J.S.). 7 figs

  6. Quick look report for semiscale MOD-2C Test S-FS-11

    International Nuclear Information System (INIS)

    Plessinger, M.P.

    1985-11-01

    Results of a preliminary analysis of the fifth test performed in the Semiscale MOD-2C Steam Generator Feedwater and Steam Line Break (FS) experiment series are presented. Test S-FS-11 simulated a pressurized water reactor transient initiated by a 50% break in a steam generator bottom feedwater line downstream of the check valve. With the exception of primary pressure, the initial conditions represented the initial conditions used for the C-E System 80 Final Safety Analysis Report (FSAR) Appendix 15B calculations. The transient included an initial 600 s period in which only automatic plant protection systems responded to the initiating event. This period was followed by a series of operator actions necessary to stabilize the plant followed by break isolation and affected loop steam generator refill with auxiliary feedwater. The test results provided a measured evaluation of the effectiveness of the automatic responses in minimizing primary system overpressurization and operator actions in stabilizing the plant. Test data also provided a basis for comparison with other tests in the series of the effects of break size on primary overpressurization and primary-to-secondary heat transfer. 64 figs

  7. Automating Test Activities: Test Cases Creation, Test Execution, and Test Reporting with Multiple Test Automation Tools

    OpenAIRE

    Loke Mun Sei

    2015-01-01

    Software testing has become a mandatory process in assuring the software product quality. Hence, test management is needed in order to manage the test activities conducted in the software test life cycle. This paper discusses on the challenges faced in the software test life cycle, and how the test processes and test activities, mainly on test cases creation, test execution, and test reporting is being managed and automated using several test automation tools, i.e. Jira, ...

  8. High Precision Current Control for the LHC Main Power Converters

    CERN Document Server

    Thiesen, H; Hudson, G; King, Q; Montabonnet, V; Nisbet, D; Page, S

    2010-01-01

    Since restarting at the end of 2009, the LHC has reached a new energy record in March 2010 with the two 3.5 TeV beams. To achieve the performance required for the good functioning of the accelerator, the currents in the main circuits (Main Bends and Main Quadrupoles) must be controlled with a higher precision than ever previously requested for a particle accelerator at CERN: a few parts per million (ppm) of nominal current. This paper describes the different challenges that were overcome to achieve the required precision for the current control of the main circuits. Precision tests performed during the hardware commissioning of the LHC illustrate this paper.

  9. [Surgical angioplasty of the left main coronary artery].

    Science.gov (United States)

    Vranes, Mile; Velinović, Milos; Kocica, Mladen; Mikić, Aleksandar; Velimirović, Dusan; Djukić, Petar

    2010-01-01

    The conventional treatment for isolated stenosis of the left main coronary artery is bypass surgery (myocardial revascularization). However, the process of atherosclerosis is not arrested by myocardial revascularization and it will lead to the occlusion of the left main coronary artery. Revascularization will establish retrograde perfusion for 50-70% of the myocardium of the left ventricle. Direct surgical angioplasty of the left main coronary artery enables normal physiological perfusion of the whole myocardium and better myocardial function. The aim of our study is to point out a new surgical approach of treating left main coronary artery stenosis. Between October 2002 and October 2003, direct surgical angioplasty of the main left coronary artery was performed on three patients with isolated stenosis of the left main coronary artery using the anterior approach and the pericardium as a patch. The procedure was performed under total endotracheal anaesthesia and standard cardiopulmonary circulation, moderate hypothermia, anterograde St. Tomas cardioplegia and local cooling. Patients were followed clinically, echocardiographically and by load-tests. All three patients were without complications. In postoperative follow-up (54-68 months) neither angina pectoris nor electrocardiographically registered ischaemic changes were found. Load-tests performed every six months on all three patients were negative. Surgical angioplasty of isolated stenosis of the left main coronary artery is a preferred method for treating this type of coronary disease. Contraindications for this type of treatment are stenosis of the left main coronary artery with bifurcation and advanced calcification of the left main coronary artery.

  10. 75 FR 27863 - Savings Bank of Maine, MHC and Savings Bank of Maine, Gardiner, Maine; Approval of Conversion...

    Science.gov (United States)

    2010-05-18

    ... DEPARTMENT OF THE TREASURY Office of Thrift Supervision [AC-38: OTS Nos. 06947 and H 4709] Savings Bank of Maine, MHC and Savings Bank of Maine, Gardiner, Maine; Approval of Conversion Application Notice is hereby given that on May 7, 2010, the Office of Thrift Supervision approved the application of...

  11. Load Asymmetry Observed During Orion Main Parachute Inflation

    Science.gov (United States)

    Morris, Aaron L.; Taylor, Thomas; Olson, Leah

    2011-01-01

    The Crew Exploration Vehicle Parachute Assembly System (CPAS) has flight tested the first two generations of the Orion parachute program. Three of the second generation tests instrumented the dispersion bridles of the Main parachute with a Tension Measuring System. The goal of this load measurement was to better understand load asymmetry during the inflation process of a cluster of Main parachutes. The CPAS Main parachutes exhibit inflations that are much less symmetric than current parachute literature and design guides would indicate. This paper will examine loads data gathered on three cluster tests, quantify the degree of asymmetry observed, and contrast the results with published design guides. Additionally, the measured loads data will be correlated with videos of the parachute inflation to make inferences about the shape of the parachute and the relative load asymmetry. The goal of this inquiry and test program is to open a dialogue regarding asymmetrical parachute inflation load factors.

  12. 30 CFR 57.6160 - Main facilities.

    Science.gov (United States)

    2010-07-01

    ... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Main facilities. 57.6160 Section 57.6160...-Underground Only § 57.6160 Main facilities. (a) Main facilities used to store explosive material underground... facilities will not prevent escape from the mine, or cause detonation of the contents of another storage...

  13. Prototype fast breeder reactor main options

    International Nuclear Information System (INIS)

    Bhoje, S.B.; Chellapandi, P.

    1996-01-01

    Fast reactor programme gets importance in the Indian energy market because of continuous growing demand of electricity and resources limited to only coal and FBR. India started its fast reactor programme with the construction of 40 MWt Fast Breeder Test Reactor (FBTR). The reactor attained its first criticality in October 1985. The reactor power will be raised to 40 MWt in near future. As a logical follow-up of FBTR, it was decided to build a prototype fast breeder reactor, PFBR. Considering significant effects of capital cost and construction period on economy, systematic efforts are made to reduce the same. The number of primary and secondary sodium loops and components have been reduced. Sodium coolant, pool type concept, oxide fuel, 20% CW D9, SS 316 LN and modified 9Cr-1Mo steel (T91) materials have been selected for PFBR. Based on the operating experience, the integrity of the high temperature components including fuel and cost optimization aspects, the plant temperatures are recommended. Steam temperature of 763 K at 16.6 MPa and a single TG of 500 MWe gross output have been decided. PFBR will be located at Kalpakkam site on the coast of Bay of Bengal. The plant life is designed for 30 y and 75% load factor. In this paper the justifications for the main options chosen are given in brief. (author). 2 figs, 2 tabs

  14. Isolators Including Main Spring Linear Guide Systems

    Science.gov (United States)

    Goold, Ryan (Inventor); Buchele, Paul (Inventor); Hindle, Timothy (Inventor); Ruebsamen, Dale Thomas (Inventor)

    2017-01-01

    Embodiments of isolators, such as three parameter isolators, including a main spring linear guide system are provided. In one embodiment, the isolator includes first and second opposing end portions, a main spring mechanically coupled between the first and second end portions, and a linear guide system extending from the first end portion, across the main spring, and toward the second end portion. The linear guide system expands and contracts in conjunction with deflection of the main spring along the working axis, while restricting displacement and rotation of the main spring along first and second axes orthogonal to the working axis.

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

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  16. Looking for Synergy with Momentum in Main Asset Classes

    OpenAIRE

    Lukas Macijauskas; Dimitrios I. Maditinos

    2014-01-01

    As during turbulent market conditions correlations between main asset-classes falter, classical asset management concepts seem unreliable. This problem stimulates search for non-discretionary asset allocation methods. The aim of the paper is to test weather the concept of Momentum phenomena could be used as a stand alone investment strategy using all main asset classes. The study is based on exploring historical prices of various asset classes; statistical data analysis method is used. Result...

  17. Development of main steam safety valve set pressure evaluating system

    International Nuclear Information System (INIS)

    Oketani, Koichiro; Manabe, Yoshihisa.

    1991-01-01

    A main steam safety valve set pressure test is conducted for all valves during every refueling outage in Japan's PWRs. Almost all operations of the test are manually conducted by a skilled worker. In order to obtain further reliability and reduce the test time, an automatic test system using a personnel computer has been developed in accordance with system concept. Quality assurance was investigated to fix system specifications. The prototype of the system was manufactured to confirm the system reliability. The results revealed that this system had high accuracy measurement and no adverse influence on the safety valve. This system was concluded to be applicable for actual use. (author)

  18. Main approaches to the study of loss

    Directory of Open Access Journals (Sweden)

    Burina E. A.

    2016-07-01

    Full Text Available this article presents the main approaches to the concepts of grief and loss study. The article describes the contribution of E. Lindemann, Z. Freud, J. Bowlby, F.E. Vasiluk, and E. Kubler-Ross. The research also contains the main forms of grief and some stadial models within the scope of problematics.

  19. Nitrogen chronology of massive main sequence stars

    NARCIS (Netherlands)

    Köhler, K.; Borzyszkowski, M.; Brott, I.; Langer, N.; de Koter, A.

    2012-01-01

    Context. Rotational mixing in massive main sequence stars is predicted to monotonically increase their surface nitrogen abundance with time. Aims. We use this effect to design a method for constraining the age and the inclination angle of massive main sequence stars, given their observed luminosity,

  20. Engineering Design of KSTAR tokamak main structure

    International Nuclear Information System (INIS)

    Im, K.H.; Cho, S.; Her, N.I.

    2001-01-01

    The main components of the KSTAR (Korea Superconducting Tokamak Advanced Research) tokamak including vacuum vessel, plasma facing components, cryostat, thermal shield and magnet supporting structure are in the final stage of engineering design. Hundai Heavy Industries (HHI) has been involved in the engineering design of these components. The current configuration and the final engineering design results for the KSTAR main structure are presented. (author)

  1. Riigikogu maine ja armastus / Aune Past

    Index Scriptorium Estoniae

    Past, Aune, 1954-

    2007-01-01

    Autor arutleb Riigikogu maine üle, toetudes TÜ ajakirjanduse ja kommunikatsiooni osakonna 2007. aastal tehtud uuringu tulemustele. Riigikogu liikmete maine paraneks, kui neil endil oleks selge, mis eesmärgil Toompeale mindi ja sellest siis ka rahvale kõneldaks

  2. Investigation of coolant mixture in pressurized water reactors at the Rossendorf mixing test facility ROCOM

    International Nuclear Information System (INIS)

    Grunwald, G.; Hoehne, T.; Prasser, H.M.; Richter, K.; Weiss, F.P.

    1999-01-01

    During the so-called boron dilution or cold water transients at pressurized water reactors too weakly borated water or too cold water, respectively, might enter the reactor core. This results in the insertion of positive reactivity and possibly leads to a power excursion. If the source of unborated or subcooled water is not located in all coolant loops but in selected ones only, the amount of reactivity insertion depends on the coolant mixing in the downcomer and lower plenum of the reactor pressure vessel (RPV). Such asymmetric disturbances of the coolant temperature or boron concentration might e.g. be the result of a failure of the chemical and volume control system (CVCS) or of a main steam line break (MSLB) that does only affect selected steam generators (SG). For the analysis of boron dilution or MSLB accidents coupled neutron kinetics/thermo-hydraulic system codes have been used. To take into account coolant mixing phenomena in these codes in a realistic manner, analytical mixing models might be included. These models must be simple and fast running on the one hand, but must well describe the real mixing conditions on the other hand. (orig.)

  3. Design and main characteristics of HTGR fuel elements

    International Nuclear Information System (INIS)

    Chernikov, A.S.; Kolesov, V.S.; Permyakov, L.N.; Koshelev, Yu.V.; Mikhajlichenko, L.I.

    1983-01-01

    Two types of spherical fuel elements and coated particles were investigated under the operating conditions of the high temperature reactors in the Soviet Union (VGR-50 and VG-400). This paper gives the main characteristics of spherical fuel elements (thermal conductivity, static and dynamic strength, wear resistance, release of gaseous fission products, etc.) as determined in test facilities. (author)

  4. Hillshades for the main 8 Hawaiian Islands

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — These hillshade datasets were derived from USGS 7.5' DEM Quads for the main 8 Hawaiian Islands. Individual DEM quads were first converted to a common datum, and...

  5. Nuclear power station main control room habitability

    International Nuclear Information System (INIS)

    Paschal, W.B.; Knous, W.S.

    1989-01-01

    The main control room at a nuclear power station must remain habitable during a variety of plant conditions and postulated events. The control room habitability requirement and the function of the heating, ventilating, air-conditioning, and air treatment system are to control environmental factors, such as temperature, pressure, humidity, radiation, and toxic gas. Habitability requirements provide for the safety of personnel and enable operation of equipment required to function in the main control room. Habitability as an issue has been gaining prominence with the Advisor Committee of Reactor Safeguards and the Nuclear Regulatory Commission since the incident at Three Mile Island. Their concern is the ability of the presently installed habitability systems to control the main control room environment after an accident. This paper discusses main control room HVAC systems; the concern, requirements, and results of NRC surveys and notices; and an approach to control room habitability reviews

  6. Seasonal Composite Chlorophyll Concentrations - Gulf of Maine

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — This personal geodatabase contains raster images of chlorophyll concentrations in the Gulf of Maine. These raster images are seasonal composites, and were calculated...

  7. Collins' bypass for the main ring

    International Nuclear Information System (INIS)

    Ohnuma, S.

    1982-01-01

    Design of the bypass for the main ring at Fermilab is discussed. Specific design features discussed include space, path length, geometric closure, matching of betatron functions, and external dispersion. Bypass parameters are given

  8. Monthly Composite Chlorophyll Concentrations - Gulf of Maine

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — This personal geodatabase contains raster images of chlorophyll concentrations in the Gulf of Maine. These raster images are monthly composites, and were calculated...

  9. 2015 City of Portland, Maine, Lidar

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — 2015 City of Portland Maine Lidar Data Acquisition and Processing Woolpert Order No. 75564 Contractor: Woolpert, Inc. This task is for a high resolution data set of...

  10. 2016 USGS Lidar DEM: Maine QL2

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Product: These are Digital Elevation Model (DEM) data for Franklin, Oxford, Piscataquis, and Somerset Counties, Maine as part of the required deliverables for the...

  11. Radioecological monitoring of south Caucasus - main results

    International Nuclear Information System (INIS)

    Tsitskishvili, M.S; Chazaradze, R.E.; Katamadze, N.M.; Intskirveli, L.N.; Chxartishvili, A.G.; Gugushvili, B.S.; Tsitskishvili, N.B.; Saneblidze, O.I.

    2002-01-01

    Basing in surrounding ambience at present radioactive on its origin possible to split into two main groups: artificial and natural radioactive. How is obvious from the most names, natural based in the nature nearly with first days of its shaping and are its by the component. Artificial - not existed or not saved in the nature - having radioactive characteristics isotopes 'appeared' as a result artificial doing atoms. Getting into surrounding ambience as a result person activity artificial (systematically or episodic detectable there) possible conditionally split into three subgroups. Artificial radioactive isotopes, got into surrounding natural ambience as a result anthropogenic activity, in principal (ecological) are distinguished from the natural radioactive isotopes by fetters and particularities to migration on ecological chains, but, consequently, and nature 'influence'. Sufficiently remind that if in biosphere practically no ecological niches, in which goes an accumulation natural, capable to give significant dozing effect; for the artificial (isotopes of iodine, isotopes a strontium, caesium) exactly ability be accumulated in separate 'niches' ecological chain or in separate organs or weaving an organism (thyroid gland for the iodine) do artificial radioisotopes hygienic extremely dangerous. Location of Caucasus in the area of approximate location of firing ranges of test, (after the series 1961-1962 conducted by USSR in the North hemisphere this were test China) and damages on Chernobyl, in the area of most intensity stratosphere - troposphere exchange, manifests themselves: 1. Early approach spring-year maximum; 2. More clear maximum in the seasonal move; 3. The Greater fallout levels in contrast with other regions of country; 4. The Greater 'sensitivity' to 'fresh' products. Structure of global fallout on the under investigation region is stipulated: 1. Decreasing the fallout levels from the north on the south. 2. Vertical fallout levels (growth with the

  12. Molecular clusters of the main group elements

    CERN Document Server

    Driess, Matthias

    2008-01-01

    ""To summarize, Molecular Clusters of the Main Group Elements is certainly not a popular science book, nor is it a textbook; it is a very good, up-to-date collection of articles for the specialist. Als Fazit bleibt: Molecular Clusters of the Main Group Elements ist sicher kein populissenschaftliches Werk, auch kein Lehrbuch, aber eine gelungene, hoch aktuelle Zusammenstellung fen interessierten Fachmann."" -Michael Ruck, TU Dresden, Angewandte Chemie, 2004 - 116/36 + International Edition 2004 - 43/36

  13. Revisiting Reuse in Main Memory Database Systems

    OpenAIRE

    Dursun, Kayhan; Binnig, Carsten; Cetintemel, Ugur; Kraska, Tim

    2016-01-01

    Reusing intermediates in databases to speed-up analytical query processing has been studied in the past. Existing solutions typically require intermediate results of individual operators to be materialized into temporary tables to be considered for reuse in subsequent queries. However, these approaches are fundamentally ill-suited for use in modern main memory databases. The reason is that modern main memory DBMSs are typically limited by the bandwidth of the memory bus, thus query execution ...

  14. Finnsjoen study site. Scope of activities and main results

    International Nuclear Information System (INIS)

    Ahlbom, K.; Andersson, J.E.; Andersson, Peter; Ittner, T.; Tiren, S.; Ljunggren, C.

    1992-12-01

    The Finnsjoen study site was selected in 1977 to provide input to the KBS-1 and KBS-2 performance assessments. The site was later used as a test site for testing new instruments and new site characterization methods, as well as a research site for studying mainly groundwater flow and groundwater transport. All together, the Finnsjoen studies have involved 11 cored boreholes, down to max 700 m depth, and extensive borehole geophysical, geochemical and geohydraulic measurements, as well as rock stress measurements and tracer tests. This report presents the scope of the Finnsjoen studies together with main results. Conceptual uncertainties in assumptions and models are discussed with emphasis on the models used for the performance assessment SKB91. Of special interest for the Finnsjoen study site is the strong influence caused by a subhorizontal fracture zone on groundwater flow, transport and chemistry

  15. Guide to the Main Ring DO overpass

    International Nuclear Information System (INIS)

    Turkot, F.

    1985-01-01

    The DO overpass is a modification of the beam orbit in Main Ring in order to better accommodate a Tevatron collider detector at DO. The orbit is moved up approx. 51 inches over most of the long straight section at DO, thus making the Main Ring the world's first non-planar proton synchrotron. A similar overpass, but with four times the displacement, is planned for the CDF detector at the BO straight section. The nominal separation between the beam orbit in the Main Ring and the orbit in the Tevatron is 25.5 inches. Early in the design study of a detector that would utilize the Tevatron is a anti pp collider, it was apparent that a larger separation at the detector was highly desirable. In 1981, Tom Collins proposed a specific lattice geometry in the Main Ring for achieving larger separation, called ''the screw beam''. His proposal has served as the basis for the design of both the BO and DO overpasses. The main purpose of this report is to describe in some detail the implementation of the DO overpass. Topics to be covered include: (a) geometry of the overpass orbit, (b) the new hardware in the tunnel, (c) the power supply system, (d) the control facility, (e) accelerator beam dynamics ramifications, and (f) commissioning experience. A secondary purpose is to provide a fairly complete ''bibliography'' to the sources of information on the overpass. 17 refs., 17 figs

  16. Post-test analysis of LOBI BT-01 using RELAP5/MOD2 and RELAP5/MOD3

    International Nuclear Information System (INIS)

    Holmes, B.J.

    1991-08-01

    LOBI is a high pressure, electrically heated integral system test facility simulating a KWU 1300 MW PWR scaled 1:712 by volume, although full scale has been maintained in the vertical direction. This report describes the results of an analysis of test BT-01, which simulates a 10% steam line break. The bulk of the analysis was performed using the Project Version of RELAP5/MOD2, with additional calculations using RELAP5/MOD3 for comparison. The codes provided generally good agreement with data. In particular, the break flows were well modelled, although the mass flow data proved to be unreliable, and this conclusion had to be derived from interpreting other signals. RELAP over-predicted primary/secondary heat transfer in the broken loop, however, leading to a more rapid cool-down of the primary circuit. Furthermore, the primary side pressure response was critically dependent upon the pressuriser behaviour, and the correct timing of the uncovery of the surge line. Inter-phase drag was not well predicted in the broken loop steam generator intermals, although some improvement was seen in the RELAP5/MOD3 predictions. MOD3 gave a reduction in primary/secondary heat transfer during the test pre-conditioning phase, resulting in a lower secondary side pressure at the start of the transient compared with MOD2. (author)

  17. Testing improved steam separators in third energy block of Chernobyl AES

    Energy Technology Data Exchange (ETDEWEB)

    Novosel' skiy, O Yu; Karasev, V B; Sakovich, E V; Lyutov, M A; An' kov, V I

    1984-12-01

    Improved steam separating drums are described. These have a four-row arrangement of the pipe unions of the steam-water supply lines, with increased diameter. Two collectors 400 mm in diameter from adjacent separators drums go to a single steam line 600 mm in diameter, which goes to the turbine. Tests of the system were conducted at a pressure of 7 MPa and at thermal power of 65, 83, 93, and 100% of rated value to determine the dependence of the moisture content at the output from the separator drum on the mass level at constant thermal power of the block. The separator drums, 2600 mm in diameter, were found to have a reserve for maximum permissible moisture content, and the moisture content of the steam in the central pipe unions did not exceed 0.02% with a level in the drums 200 mm above the rated level. Thus maintenance of the level above the submerged perforated plate by 100 mm above the planned level permits an increase of the water reserve in the multiple forced circulation circuit by 28 m/sup 3/ and does not hinder an increase of the steam productivity of the block.

  18. Main Coast Winds - Final Scientific Report

    Energy Technology Data Exchange (ETDEWEB)

    Jason Huckaby; Harley Lee

    2006-03-15

    The Maine Coast Wind Project was developed to investigate the cost-effectiveness of small, distributed wind systems on coastal sites in Maine. The restructuring of Maine's electric grid to support net metering allowed for the installation of small wind installations across the state (up to 100kW). The study performed adds insight to the difficulties of developing cost-effective distributed systems in coastal environments. The technical hurdles encountered with the chosen wind turbine, combined with the lower than expected wind speeds, did not provide a cost-effective return to make a distributed wind program economically feasible. While the turbine was accepted within the community, the low availability has been a negative.

  19. Transition crossing in the main injector

    International Nuclear Information System (INIS)

    Wei, J.

    1990-01-01

    This report summarizes the study of various longitudinal problems pertaining to the transition-energy crossing in the proposed Fermi Lab Main Injector. The theory indicates that the beam loss and bunch-area growth are mainly caused by the chromatic non-linear effect, which is enhanced by the space-charge force near transition. Computer simulation using the program TIBETAN shows that a ''γ T jump'' of about 1.5 unit within 1 ms is adequate to achieve a ''clean'' crossing in the currently proposed h=588 scenario. 19 refs., 4 figs

  20. The Fermilab Main Injector Technical Design Handbook

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    1994-08-01

    This report contains a description of the design, cost estimate, and construction schedule of the Fermilab Main Injector (FMI) Project. The technical, cost, and schedule baselines for the FMI Project have already been established and may be found in the Fermilab Main Injector Title I Design Report, issued in August 1992. This report updates and expands upon the design and schedule for construction of all subsystem components and associated civil construction described in the Title I Design Report. The facilities described have been designed in conformance with DOE 6430.1A, "United States Department of Energy General Design Criteria."

  1. Current Russian patriotism: matter, features, main directions

    Directory of Open Access Journals (Sweden)

    Lutovinov Vladimir Ilich

    2013-11-01

    Full Text Available The article considers understanding and the main point of patriotism as one of high cultural values. The main approaches that reveal different sides of this phenomenon, its role and importance in a history of Russia in the 21st century are inferred from the analysis of viewpoints of Russian thinkers and contemporary researchers. The patriotism formation problems in Russian society and their condition are defined, the need of patriotic level rise as one of the conditions for great Russia rebirth is substantiated.

  2. Bunch coalescing in the Fermilab Main Ring

    International Nuclear Information System (INIS)

    Wildman, D.; Martin, P.; Meisner, K.; Miller, H.W.

    1987-01-01

    A new RF system has been installed in the Fermilab Main Ring to coalesce up to 13 individual bunches of protons or antiprotons into a single high-intensity bunch. The coalescing process consists of adiabatically reducing the h=1113 Main Ring RF voltage from 1 MV to less than 1 kV, capturing the debunched beam in a linearized h=53 and h=106 bucket, rotating for a quarter of a synchrotron oscillation period, and then recapturing the beam in a single h=1113 bucket. The new system is described and the results of recent coalescing experiments are compared with computer-generated particle tracking simulations

  3. Failure analysis of a helicopter's main rotor bearing

    International Nuclear Information System (INIS)

    Shahzad, M.; Qureshi, A.H.; Waqas, H.; Hussain, N.; Ali, N.

    2011-01-01

    Presented results report some of the findings of a detailed failure analysis carried out on a main rotor hub assembly, which had symptoms of burning and mechanical damage. The analysis suggests environmental degradation of the grease which causes pitting on bearing-balls. The consequent inefficient lubrication raises the temperature which leads to the smearing of cage material (brass) on the bearing-balls and ultimately causes the failure. The analysis has been supported by the microstructural studies, thermal analysis and micro-hardness testing performed on the affected main rotor bearing parts. (author)

  4. Experimental investigation of iodine removal and containment depressurization in containment spray system test facility of 700 MWe Indian pressurized heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jain, Manish [Department of Mechanical Engineering, I.I.T., Bombay, Powai, Mumbai (India); Kandar, T.K.; Vhora, S.F.; Mohan, Nalini [Directorate of Technology Development, Nuclear Power Corporation of India Limited, Mumbai (India); Iyer, K.N. [Department of Mechanical Engineering, I.I.T., Bombay, Powai, Mumbai (India); Prabhu, S.V., E-mail: svprabhu@iitb.ac.in [Department of Mechanical Engineering, I.I.T., Bombay, Powai, Mumbai (India)

    2017-05-15

    Highlights: • Depressurization rate in a scaled down vessel filled with air and steam is studied. • Iodine removal rate in a scaled down vessel filled with steam/air is investigated. • Effect of SMD and vessel pressure on depressurization rate is studied. • Depressurization rate decreases with the increase in the droplet size (590 μm – 1 mm) • Decrease in pressure and iodine concentration with time follow exponential trend. - Abstract: As an additional safety measure in the new 700 MWe Indian pressurized heavy water reactors, the first of a kind system called containment Spray System is introduced. The system is designed to cater/mitigate the conditions after design basis accidents i.e., loss of coolant accident and main steam line break. As a contribution to the safety analysis of condition following loss-of-coolant accidents, experiments are carried out to establish the performance of the system. The loss of coolant is simulated by injecting saturated steam and iodine vapors into the containment vessel in which air is enclosed at atmospheric and room temperature, and then the steam-air mixture is cooled by sprays of water. The effect of water spray on the containment vessel pressure and the iodine scrubbing in a scaled down facility is investigated for the containment spray system of Indian pressurized heavy water reactors. The experiments are carried out in the scaled down vessel of the diameter of 2.0 m and height of 3.5 m respectively. Experiments are conducted with water at room temperature as the spray medium. Two different initial vessel pressure i.e. 0.7 bar and 1.0 bar are chosen for the studies as they are nearing the loss of coolant accident & main steam line break pressures in Indian pressurized heavy water reactors. These pressures are chosen based on the containment resultant pressures after a design basis accident. The transient temperature and pressure distribution of the steam in the vessel are measured during the depressurization

  5. Mains power synchronous conducted noise measurement in the 2 to 150 kHz band

    NARCIS (Netherlands)

    Keyer, Cornelis H.A.; Buesink, Frederik Johannes Karel; Leferink, Frank Bernardus Johannes

    2016-01-01

    When testing large stationary equipment for EMC compliance, on-site testing is often the only option. In case of a fixed mains connection, use of an artificial mains network is problematic. Within In the framework of an European Research project on EMI testing of large installations, research has

  6. Main principles of development stationary training facilities

    International Nuclear Information System (INIS)

    Tsiptsyura, R.D.

    1986-01-01

    The designation of stationary training facilities is shown and the main requirements for them are formulated. When considering the above-mentioned requirements, special attention was paid to obligatory correspondence between training experience and practical skill of an operator. It is shown, that the switchboard block is the major unit of the training facility, which should develop skills and habits of an operator

  7. Maine Project against Bullying. Final Report.

    Science.gov (United States)

    Saufler, Chuck; Gagne, Cyndi

    Noting that bullying among primary school-age children has become recognized as an antecedent to more violent behavior in later grades, the 3-year Maine Project Against Bullying examined currently available research on bullying and evaluated books, curricula, media materials, and programs to identify resources and strategies which can be applied…

  8. Modern Portfolio Theory: Some Main Results

    OpenAIRE

    Müller, Heinz H.

    2017-01-01

    This article summarizes some main results in modern portfolio theory. First, the Markowitz approach is presented. Then the capital asset pricing model is derived and its empirical testability is discussed. Afterwards Neumann-Morgenstern utility theory is applied to the portfolio problem. Finally, it is shown how optimal risk allocation in an economy may lead to portfolio insurance

  9. Enhancing training in the main control room

    International Nuclear Information System (INIS)

    McGuigan, K.; O'Leary, K.; Canavan, K.

    2004-01-01

    In 2003 Pickering B Nuclear of Ontario Power Generation installed a Desktop Simulator (DTS) in the Main Control Room (MCR) for training purposes. This paper will outline why this training enhancement was undertaken and the approach taken to secure its use in an active MCR environment while minimizing distractions to plant operations. (author)

  10. Main successes, achievements. Paths of development

    Directory of Open Access Journals (Sweden)

    A. A. Kubanova

    2015-01-01

    Full Text Available The article provides the overview of incidence of sexually transmitted infections and skin disorders over time in Russian Federation in 2004-2014 with its main positive tendencies; results of reorganisation of bed capacity of dermatovenerologic medical organizations; dermatovenerologic bed rates.

  11. Main facts 1995; Faits marquants 1995

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-11-01

    This report presents the main facts of the studies carried out by the Direction des Etudes et Recherches (DER) of Electricite de France: new applications of electricity, classical and nuclear thermal power plants, electrical equipment, environment protection, monitoring and plants operations.

  12. Space transportation main engine reliability and safety

    Science.gov (United States)

    Monk, Jan C.

    1991-01-01

    Viewgraphs are used to illustrate the reliability engineering and aerospace safety of the Space Transportation Main Engine (STME). A technology developed is called Total Quality Management (TQM). The goal is to develop a robust design. Reducing process variability produces a product with improved reliability and safety. Some engine system design characteristics are identified which improves reliability.

  13. Water Hammer in Pumped Sewer Mains

    DEFF Research Database (Denmark)

    Larsen, Torben

    This publication is intended for engineers seeking an introduction to the problem of water hammer in pumped pressure mains. This is a subject of increasing interest because of the development of larger and more integrated sewer systems. Consideration of water hammer is essential for structural...

  14. D.E.R. 91 main facts

    International Nuclear Information System (INIS)

    Schmidt, N.

    1991-01-01

    This report presents the main facts of the studies carried out by the Direction des Etudes et Recherches (DER) of Electricite de France: New applications of electricity, classical and nuclear thermal power plants, electrical equipment, environment protection, network analysis, information and informatic equipment

  15. VINKA, ten years on. Main scientific results

    International Nuclear Information System (INIS)

    1979-01-01

    The VINKA facility in the TRITON swimming-pool reactor at Fontenay-aux-Roses allows the irradiation of solids at low temperatures in order to study crystalline defects. After ten years of operation the main scientific results obtained in the fields of creep and growth (chapter I), point defects (chapter II), amorphisation (chapter III) and dechanneling of particles (chapter IV) are summarised [fr

  16. Results on Fermilab main injector dipole measurements

    International Nuclear Information System (INIS)

    Brown, B.C.; Baiod, R.; DiMarco, J.; Glass, H.D.; Harding, D.J.; Martin, P.S.; Mishra, S.; Mokhtarani, A.; Orris, D.F.; russell, O.A.; Tompkins, J.C.; Walbridge, D.G.C.

    1995-06-01

    Measurements of the Productions run of Fermilab Main Injector Dipole magnets is underway. Redundant strength measurements provide a set of data which one can fit to mechanical and magnetic properties of the assembly. Plots of the field contribution from the steel supplement the usual plots of transfer function (B/I) vs. I in providing insight into the measured results

  17. D.E.R. 92 - Main facts

    International Nuclear Information System (INIS)

    1992-01-01

    This report presents the main facts of the studies carried out by the Direction des Etudes et Recherches (DER) of Electricite de France: new applications of electricity, classical and nuclear thermal power plants, electrical equipment, environment protection, network analysis, information and informatic equipment

  18. Methodology for energy diagnosis in distribution steam lines; Metodologia para diagnostico de energia en lineas de distribucion de vapor

    Energy Technology Data Exchange (ETDEWEB)

    Almanza, M; Ambriz, J J; Romero P, H [Universidad Autonoma Metropolitana Iztapalapa, Mexico, D. F. (Mexico)

    1993-12-31

    This paper shows a methodology that results of great advantage in the development of the energy analysis of an industrial facility that utilizes steam as a mean of energy transport and where the steam operated equipment is physically located in a remote place, away from the generation site. Mention is made here of the equipment characteristics that can be used for such purpose, the most important parameters to identify in a rapid and efficient way the faults presented in the steam distribution system in the industrial plants and the formats that contribute to keep the corresponding records for efficiently maintain in operation the steam pipeline net in conjunction with the involved accessories. [Espanol] En el presenta trabajo se muestra una metodologia que resulta de gran utilidad en el desarrollo del analisis energetico de una planta industrial, en la cual se emplee el vapor como medio de transporte de la energia y en donde el equipo que utiliza el vapor se encuentre fisicamente en un lugar apartado de la zona de generacion. Aqui se mencionan las caracteristicas del equipo que se puede utilizar para dicho diagnostico, los parametros de mayor importancia para identificar en forma rapida y eficiente las fallas que se presentan en el sistema de distribucion de vapor en las plantas industriales y los formatos que contribuyen a llevar los registros correspondientes para mantener operando eficientemente la red de vapor, en conjunto con los accesorios que en ella se involucran.

  19. Methodology for energy diagnosis in distribution steam lines; Metodologia para diagnostico de energia en lineas de distribucion de vapor

    Energy Technology Data Exchange (ETDEWEB)

    Almanza, M.; Ambriz, J. J.; Romero P, H. [Universidad Autonoma Metropolitana Iztapalapa, Mexico, D. F. (Mexico)

    1992-12-31

    This paper shows a methodology that results of great advantage in the development of the energy analysis of an industrial facility that utilizes steam as a mean of energy transport and where the steam operated equipment is physically located in a remote place, away from the generation site. Mention is made here of the equipment characteristics that can be used for such purpose, the most important parameters to identify in a rapid and efficient way the faults presented in the steam distribution system in the industrial plants and the formats that contribute to keep the corresponding records for efficiently maintain in operation the steam pipeline net in conjunction with the involved accessories. [Espanol] En el presenta trabajo se muestra una metodologia que resulta de gran utilidad en el desarrollo del analisis energetico de una planta industrial, en la cual se emplee el vapor como medio de transporte de la energia y en donde el equipo que utiliza el vapor se encuentre fisicamente en un lugar apartado de la zona de generacion. Aqui se mencionan las caracteristicas del equipo que se puede utilizar para dicho diagnostico, los parametros de mayor importancia para identificar en forma rapida y eficiente las fallas que se presentan en el sistema de distribucion de vapor en las plantas industriales y los formatos que contribuyen a llevar los registros correspondientes para mantener operando eficientemente la red de vapor, en conjunto con los accesorios que en ella se involucran.

  20. Condition monitoring of main coolant pumps, Dhruva

    International Nuclear Information System (INIS)

    Prasad, V.; Satheesh, C.; Acharya, V.N.; Tikku, A.C.; Mishra, S.K.

    2002-01-01

    Full text: Dhruva is a 100 MW research reactor with natural uranium fuel, heavy water as moderator and primary coolant. Three Centrifugal pumps circulate the primary coolant across the core and the heat exchangers. Each pump is coupled to a flywheel (FW) assembly in order to meet operational safety requirements. All the 3 main coolant pump (MCP) sets are required to operate during operation of the reactor. The pump-sets are in operation since the year 1984 and have logged more than 1,00,000 hrs. Frequent breakdowns of its FW bearings were experienced during initial years of operation. Condition monitoring of these pumps, largely on vibration based parameters, was initiated on regular basis. Break-downs of main coolant pumps reduced considerably due to the fair accurate predictions of incipient break-downs and timely maintenance efforts. An effort is made in this paper to share the experience

  1. Alarm system for ABWR main control panels

    Energy Technology Data Exchange (ETDEWEB)

    Kobayashi, Yuji; Saito, Koji [Toshiba Corp., Yokohoma (Japan)

    1997-09-01

    TOSHIBA has developed integrated digital control and instrumentation system for ABWR, which is the third-generation man machine interface system for main control room that we call A-PODIA (Advanced PODIA). A-Podia has been introduced the first actual ABWR plant in Japan. in A-PODIA, TOSHIBA has realized improvement of alarm system that all operator crews in the control room can recognize plant anomalies easily. The alarm system can recognize essential alarms for plant safety easily and understand annunciators with each integrated annunciators and their prioritized color easily by classifying alarms into plant-level essential annunciators, system-level integrated annunciators and equipment level individual annunciators with hierarchical structure. This paper describes conventional alarm system and the design philosophy, alarm system design and operation of ``Alarm System for ABWR Main Control Panels``. (author). 5 refs, 8 figs, 1 tab.

  2. Organization structure. Main activities of the Division

    International Nuclear Information System (INIS)

    2008-01-01

    In this chapter the organization structure as well as main activities of the Division for radiation safety, NPP decommissioning and radioactive waste management are presented. This Division of the VUJE, a.s. consists of the following sections and departments: Section for economic and technical services; Section for radiation protection of employees; Department for management of emergency situations and risk assessment; Department for implementation of nuclear power facilities decommissioning and RAW management; Department for personnel and environmental dosimetry; Department for preparation of NPP decommissioning; Department for RAW treatment technologies; Department for chemical regimes and physico-chemical analyses; Department for management of nuclear power facilities decommissioning and RAW management. Main activities of this Division are presented.

  3. Chromaticity compensation scheme for the Main Injector

    International Nuclear Information System (INIS)

    Bogacz, S.A.

    1993-05-01

    The current Main Injector lattice is studied in the context of full chromaticity compensation in the presence of the eddy current, saturation and the end-pack sextupole fields generated by the dipole magnets. Two families of correcting sextupole magnets are placed to compensate these fields and to adjust the chromaticity (in both planes) to some desired value. Variation of the dipole induced sextupole fields with the B-field (changing along a ramp) are modeled according to recent experimental measurements of the Main Injector dipole magnet Analysis of the required sextupole strengths is carried out along two realistic momentum ramps. The results of our calculation give quantitative insight into the requisite performance of the sextupole magnets

  4. Installation Strategy for the LHC Main Dipoles

    CERN Multimedia

    Fartoukh, Stephane David

    2004-01-01

    All positions in the LHC machine are not equivalent in terms of beam requirements on the geometry and the field quality of the main dipoles. In the presence of slightly or strongly out-of tolerance magnets, a well-defined installation strategy will therefore contribute to preserve or even optimize the performance of the machine. Based on the present status of the production, we have anticipated a list of potential issues (geometry, transfer function, field direction and random b3) which, combined by order of priority, have been taken into account to define a simple but efficient installation algorithm for the LHC main dipoles. Its output is a prescription for installing the available dipoles in sequence while reducing to an absolute minimum the number of holes required by geometry or FQ issues.

  5. Alarm system for ABWR main control panels

    International Nuclear Information System (INIS)

    Kobayashi, Yuji; Saito, Koji

    1997-01-01

    TOSHIBA has developed integrated digital control and instrumentation system for ABWR, which is the third-generation man machine interface system for main control room that we call A-PODIA (Advanced PODIA). A-Podia has been introduced the first actual ABWR plant in Japan. in A-PODIA, TOSHIBA has realized improvement of alarm system that all operator crews in the control room can recognize plant anomalies easily. The alarm system can recognize essential alarms for plant safety easily and understand annunciators with each integrated annunciators and their prioritized color easily by classifying alarms into plant-level essential annunciators, system-level integrated annunciators and equipment level individual annunciators with hierarchical structure. This paper describes conventional alarm system and the design philosophy, alarm system design and operation of ''Alarm System for ABWR Main Control Panels''. (author). 5 refs, 8 figs, 1 tab

  6. INNOVATION DIFFUSION THEORY MAIN DEVELOPMENT STAGES

    Directory of Open Access Journals (Sweden)

    S. V. Lisafiev

    2011-01-01

    Full Text Available Abstract: Main innovation diffusion development theory stages are: Rogers model of moving new products to the market including characteristics of its segments; mathematic substantiation of this model by Bass; Moor model taking into account gaps between adjacent market segments; Goldenberg model making it possible to predict sales drops at new product life cycle initial stages. It is reasonable to use this theory while moving innovative products to the market.

  7. Space shuttle main engine vibration data base

    Science.gov (United States)

    Lewallen, Pat

    1986-01-01

    The Space Shuttle Main Engine Vibration Data Base is described. Included is a detailed description of the data base components, the data acquisition process, the more sophisticated software routines, and the future data acquisition methods. Several figures and plots are provided to illustrate the various output formats accessible to the user. The numerous vibration data recall and analysis capabilities available through automated data base techniques are revealed.

  8. The Main Properties of q-Functions

    OpenAIRE

    Nina O. Virchenko; Olena V. Ovcharenko

    2017-01-01

    Background.  The new generalization of the function of complex variable (q-function) is considered, its main properties are investigated. Such distributions have a special place among the special functions due to their widespread use in many areas of applied mathematics. Objective. The aim of the paper is to study the new generalization of the function of complex variable for application in applied sciences. Methods. To obtain scientific results the general methods of the mathematical a...

  9. Radiation shielding of the main injector

    International Nuclear Information System (INIS)

    Bhat, C.M.; Martin, P.S.

    1995-05-01

    The radiation shielding in the Fermilab Main Injector (FMI) complex has been carried out by adopting a number of prescribed stringent guidelines established by a previous safety analysis. Determination of the required amount of radiation shielding at various locations of the FMI has been done using Monte Carlo computations. A three dimensional ray tracing code as well as a code based upon empirical observations have been employed in certain cases

  10. Post-main-sequence planetary system evolution

    Science.gov (United States)

    Veras, Dimitri

    2016-01-01

    The fates of planetary systems provide unassailable insights into their formation and represent rich cross-disciplinary dynamical laboratories. Mounting observations of post-main-sequence planetary systems necessitate a complementary level of theoretical scrutiny. Here, I review the diverse dynamical processes which affect planets, asteroids, comets and pebbles as their parent stars evolve into giant branch, white dwarf and neutron stars. This reference provides a foundation for the interpretation and modelling of currently known systems and upcoming discoveries. PMID:26998326

  11. The BTeV main spectrometer

    International Nuclear Information System (INIS)

    Sheldon, P.D.

    2001-01-01

    BTeV is a second generation B-factory experiment that will use a double-arm, forward spectrometer in the C0 experimental hall at the Fermilab Tevatron. I will describe the motivation and design of the 'main spectrometer', consisting of a ring-imaging Cherenkov system for charged particle identification, an electromagnetic calorimeter of lead-tungstate crystals, a proportional tube muon system with magnetized filtering steel, and a straw-tube and silicon strip charged particle tracking system

  12. Main characteristics and development of NDT equipment

    International Nuclear Information System (INIS)

    Dubresson, J.

    1991-01-01

    Recent developments of non destructive testing with ionizing radiations are reviewed. Real time or differed time data processing is the key for new techniques and a renewal of classical techniques. Progress of radiation sources are also examined [fr

  13. Main-sequence photometry in NGC 2808

    International Nuclear Information System (INIS)

    Buonanno, R.; Corsi, C.E.; Fusi Pecci, F.; Harris, W.E.

    1984-01-01

    We have obtained a color-magnitude diagram for the southern globular cluster NGC 2808, to V/sub lim/approx. =21 (about 2 mag below the main-sequence turnoff). The internal photographic errors are sigma/sub V/approx. =0.02, sigma/sub B/-Vapprox. =0.03, small enough to permit a precise definition of the turnoff region and an estimate of the ''cosmic scatter'' along the main sequence. Fitting of the CMD to VandenBerg's [Astrophys. J. Suppl. 51, 29 (1983)] isochrones shows that an excellent match to the observations is achieved for model parameters of Yapprox. =0.2, Zapprox. =0.003 ([Fe/H]approx. =-0.8), and an age of (16 +- 2) billion years. All these characteristics are within the expected range from other observational constraints; no new clues from the main-sequence data alone have arisen to help explain the presence of the anomalous blue horizontal-branch stars

  14. Isolated systems with wind power. Main report

    Energy Technology Data Exchange (ETDEWEB)

    Lundsager, P.; Bindner, H.; Clausen, N.E.; Frandsen, S.; Hansen, L.H.; Hansen, J.C.

    2001-06-01

    The overall objective of this research project is to study the development of methods and guidelines rather than 'universal solutions' for the use of wind energy in isolated communities. The main specific objective of the project is to develop and present a more unified and generally applicable approach for assessing the technical and economical feasibility of isolated power supply systems with wind energy. As a part of the project the following tasks were carried out: Review of literature, field measurements in Egypt, development of an inventory of small isolated systems, overview of end-user demands, analysis of findings and development of proposed guidelines. The project is reported in one main report and four topical reports, all of them issued as Risoe reports. This is the Main Report Risoe-R-1256, summing up the activities and findings of the project and outlining an Implementation Strategy for Isolated Systems with Wind Power, applicable for international organisations such as donor agencies and development banks. (au)

  15. Progress on a prototype main ring rf cavity

    International Nuclear Information System (INIS)

    Swain, G.; Kandarian, R.; Thiessen, H.A.; Poirier, R.; Smythe, W.R.

    1989-01-01

    A prototype rf cavity and rf drive system for a hadron facility main ring has been designed and will be tested in the Proton Storage Ring (PSR) at Los Alamos as a part of a collaborative effort between LANL and TRIUMF. The cavity uses an orthogonally biased ferrite tuner. The design provides for accelerating gap voltages up to 200 kV for the 49.3 to 50.8 MHz range. Progress on the cavity construction and testing is described. 13 refs., 5 figs

  16. Component biomass equations for black spruce in Maine

    Science.gov (United States)

    M. M. Czapowskyj; D. J. Robison; R. D. Briggs; E. H. White; E. H. White

    1985-01-01

    Component biomass prediction equations are presented for young black spruce (Picea mariana B.S.P. (Mill,:)) in northern Maine. A weighted least squares model was used to construct the eq~iationsfo r small trees from 1 to 15 cm d.b.h., and an ordinary least squares model for trees less than 2 m in height. A linearized allometric model was also tested but was not used....

  17. MECAR (Main Ring Excitation Controller and Regulator): A real time learning regulator for the Fermilab Main Ring or the Main Injector synchrotron

    International Nuclear Information System (INIS)

    Flora, R.; Martin, K.; Moibenko, A.; Pfeffer, H.; Wolff, D.; Prieto, P.; Hays, S.

    1995-04-01

    The real time computer for controlling and regulating the FNAL Main Ring power supplies has been upgraded with a new learning control system. The learning time of the system has been reduced by an order of magnitude, mostly through the implementation of a 95 tap FIR filter in the learning algorithm. The magnet system consists of three buses, which must track each other during a ramp from 100 to 1700 amps at a 2.4 second repetition rate. This paper will present the system configuration and the tools used during development and testing

  18. Rates of the main thermonuclear reactions

    International Nuclear Information System (INIS)

    Abramovich, S.N.; Guzhovskii, B.Ya.; Dunaeva, S.A.; Fomushkin, E.F.

    1992-01-01

    The data on the cross sections of main thermonuclear reactions have been estimated with an account of the latest experimental results in a form of S-factor spline presentation. Based on this estimation, the reates of these reactions in 0.0001-1 MeV temperature range in the supposition of Maxwell distribution of relative velocities have been computed. The Maxwell-Boltzmann averaged -factors were calculated according to the table values of the reaction rates. Then the -factors were approximated with the 3 order spline-function. The necessity of the account of electron shielding and intramolecular movement at low temperatures is discussed (orig.)

  19. Musculoskeletal disorders in main battle tank personnel

    DEFF Research Database (Denmark)

    Nissen, Lars Ravnborg; Guldager, Bernadette; Gyntelberg, Finn

    2009-01-01

    PURPOSE: To compare the prevalence of musculoskeletal disorders of personnel in the main battle tank (MBT) units in the Danish army with those of personnel in other types of army units, and to investigate associations between job function in the tank, military rank, and musculoskeletal problems......, and ankle. RESULTS AND CONCLUSIONS: There were only 4 women in the MBT group; as a consequence, female personnel were excluded from the study. The participation rate was 58.0% (n = 184) in the MBT group and 56.3% (n = 333) in the reference group. The pattern of musculoskeletal disorders among personnel...

  20. Main problems of modern radiation hygiene

    International Nuclear Information System (INIS)

    Il'in, L.A.; Buldakov, L.A.; Knizhnikov, V.A.

    1982-01-01

    The results of investigations carried out in 1980-81 in the field of radiation hygiene as well as plans for 1981-85 are considered. Three main groups of problems which the radiation hygiene is facing at the present time are discussed. The determination of levels and study of regularities of ionizing radiation dose formation in the population and personnel working with ionizing radiation sources in one of the promissing directions of the investigations. Delayed irradiation aftereffects andcontaminant action ofirradiation and chemical substances are no less important. The third important problem lies in the development of protective measures and arrangements on improving state sanitary inspection in the field of radiation hygiene

  1. Some peat deposits in Penobscot County, Maine

    Science.gov (United States)

    Cameron, Cornelia Clermont; Anderson, Walter A.

    1979-01-01

    Twenty of the peat deposits in Penobscot County, Maine contain an estimated 29,282,000 short tons air-dried peat. The peat is chiefly sphagnum moss and reed-sedge of high quality according to ASTM standards for agricultural and horticultural use. Analyses show that this same volume has high fuel value, low sulfur and high hydrogen contents compared with lignite and sub-bituminous coal, which may indicate that it also has potential for fuel use. On the basis of the metallic trace element content, one area within the region containing the 20 deposits has been delineated for further bedrock studies.

  2. Main-chain supramolecular block copolymers.

    Science.gov (United States)

    Yang, Si Kyung; Ambade, Ashootosh V; Weck, Marcus

    2011-01-01

    Block copolymers are key building blocks for a variety of applications ranging from electronic devices to drug delivery. The material properties of block copolymers can be tuned and potentially improved by introducing noncovalent interactions in place of covalent linkages between polymeric blocks resulting in the formation of supramolecular block copolymers. Such materials combine the microphase separation behavior inherent to block copolymers with the responsiveness of supramolecular materials thereby affording dynamic and reversible materials. This tutorial review covers recent advances in main-chain supramolecular block copolymers and describes the design principles, synthetic approaches, advantages, and potential applications.

  3. Water Hammer in Pumped Sewer Mains

    DEFF Research Database (Denmark)

    Larsen, Torben

    of transients in pumped pipeline systems. This present publication can be understood as the second and revised edition of the pamphlet ”Transients in pumped sewer mains” (2006) which was published as a technical report by The EVA committee under The Danish Water Pollution Committee (The Danish Society......This publication is intended for students and engineers seeking an introduction to the problem of water transients in pumped sewer and water mains. This is a subject of increasing interest because of the development of larger and more integrated systems. Consideration of transients is essential...

  4. Main challenges in demulsifier research and application

    Science.gov (United States)

    Zhang, Fusheng; Liu, Guoliang; Ma, Junhan; Ouyang, Jian; Yi, Xiaoling; Su, Huimin

    2017-01-01

    Main challenges in demulsifier research, such as demulsification of ASP flooding produced liquid, demulsification of heavy oil produced liquid, low temperature demulsification and fast demulsification, are summarized. Some importance technology routes to solve the challenges are proposed according to demulsification mechanisms and emulsion characteristics. The proposed routes include increasing aromaticity, molecular weight and branch degree of demulsifiers, and introducing double-function groups to demulsifiers for W/O and O/W emulsions, or groups with alkyl matching with alkyl carbon number of the crude oil into demulsifier molecule. The demulsification mechanisms of the above-mentioned research routes are described in detail.

  5. Isolated systems with wind power. Main report

    DEFF Research Database (Denmark)

    Lundsager, P.; Bindner, Henrik W.; Clausen, Niels-Erik

    2001-01-01

    The overall objective of this research project is to study the development of methods and guidelines rather than "universal solutions" for the use of wind energy in isolated communities. The main specific objective of the project is to develop and present amore unified and generally applicable...... approach for assessing the technical and economical feasibility of isolated power supply systems with wind energy. As a part of the project the following tasks were carried out: Review of literature, fieldmeasurements in Egypt, development of an inventory of small isolated systems, overview of end...... for Isolated Systems with Wind Power, applicable for international organisations such as donoragencies and development banks....

  6. Forensic entomology and main challenges in Brazil.

    Science.gov (United States)

    Gomes, Leonardo; Von Zuben, Cláudio J

    2006-01-01

    Apart from an early case report from China (13th century), the first observations on insects and other arthropods as forensic indicators were documented in Germany and France during mass exhumations in the 1880s by Reinhard, who is considered a co-founder of the discipline. After the French publication of Mégnin's popular book on the applied aspects of forensic entomology, the concept quickly spread to Canada and United States. At that time, researchers recognized that the lack of systematic observations of insects of forensic importance jeopardized their use as indicators of postmortem interval. General advances in insect taxonomy and ecology helped to fill this gap over the following decades. After World Wars, few forensic entomology cases were reported in the scientific literature. From 1960s to the 1980s, Leclercq and Nuorteva were primarily responsible for maintaining the method in Central Europe, reporting isolated cases. Since then, basic research in the USA, Russia and Canada opened the way to the routine use of Entomology in forensic investigations. Identifications of insects associated with human cadavers are relatively few in the literature of the Neotropical region and have received little attention in Brazil. This article brings an overview of historic developments in this field, the recent studies and the main problems and challenges in South America and mainly in Brazil.

  7. Main Issues in Big Data Security

    Directory of Open Access Journals (Sweden)

    Julio Moreno

    2016-09-01

    Full Text Available Data is currently one of the most important assets for companies in every field. The continuous growth in the importance and volume of data has created a new problem: it cannot be handled by traditional analysis techniques. This problem was, therefore, solved through the creation of a new paradigm: Big Data. However, Big Data originated new issues related not only to the volume or the variety of the data, but also to data security and privacy. In order to obtain a full perspective of the problem, we decided to carry out an investigation with the objective of highlighting the main issues regarding Big Data security, and also the solutions proposed by the scientific community to solve them. In this paper, we explain the results obtained after applying a systematic mapping study to security in the Big Data ecosystem. It is almost impossible to carry out detailed research into the entire topic of security, and the outcome of this research is, therefore, a big picture of the main problems related to security in a Big Data system, along with the principal solutions to them proposed by the research community.

  8. A random walk down Main Street

    Directory of Open Access Journals (Sweden)

    David Matthew Levinson

    2016-08-01

    Full Text Available US suburbs have often been characterized by their relatively low walk accessibility compared to more urban environments, and US urban environments have been char- acterized by low walk accessibility compared to cities in other countries. Lower overall density in the suburbs implies that activities, if spread out, would have a greater distance between them. But why should activities be spread out instead of developed contiguously? This brief research note builds a positive model for the emergence of contiguous development along “Main Street” to illustrate the trade-offs that result in the built environment we observe. It then suggests some policy interventions to place a “thumb on the scale” to choose which parcels will develop in which sequence to achieve socially preferred outcomes.

  9. Mortise terrorism on the main pipelines

    Science.gov (United States)

    Komarov, V. A.; Nigrey, N. N.; Bronnikov, D. A.; Nigrey, A. A.

    2018-01-01

    The research aim of the work is to analyze the effectiveness of the methods of physical protection of main pipelines proposed in the article from the "mortise terrorism" A mathematical model has been developed that made it possible to predict the dynamics of "mortise terrorism" in the short term. An analysis of the effectiveness of physical protection methods proposed in the article to prevent unauthorized impacts on the objects under investigation is given. A variant of a video analytics system has been developed that allows detecting violators with recognition of the types of work they perform at a distance of 150 meters in conditions of complex natural backgrounds and precipitation. Probability of detection is 0.959.

  10. Siberian snakes for the Fermilab Main Injector

    International Nuclear Information System (INIS)

    Anferov, V.A.; Baiod, R.; Courant, E.D.

    1993-01-01

    Appropriate Siberian snakes were designed to maintain the proton beam polarization during acceleration in the Fermilab Main Injector from 8 to 150 GeV. Various snake designs were investigated to find one fitting into the 14 m straight section spaces with the required spin rotation axis and the minimum orbit excursion. The authors studied both cold and warm discrete magnet snakes as well as warm snakes with helical magnets. For the warm discrete magnet snake, obtaining small orbit excursions required a nearly longitudinal snake axis, while axes near ±45 degrees are needed when using two snakes in a ring. The authors found acceptable snakes either by using superconducting magnets or by using warm magnets with a helical dipole field

  11. CBA main magnet power supply ripple reduction

    International Nuclear Information System (INIS)

    Bagley, G.; Edwards, R.J.

    1983-01-01

    The preliminary results of a development program to minimize beam perturbation resulting from ripple current generated by the CBA Main Magnet Power Supply are presented. The assessment of the magnitude and causes of the ripple generated led to a modification of the SCR Gate Driver and the addition of a bandpass amplifier correction loop which gave significant improvement. A description of the changes made and the results obtained are included. A second design approach was developed in which the timing of the SCR gate pulses is directly determined by a VCO. The results reported with this VCO Loop indicate superior performance particularly at frequencies below 60 Hz. A shunt transistor regulator design is proposed to minimize higher SCR switching frequency harmonics

  12. BANKING ETHICS: MAIN CONCEPTIONS AND PROBLEMS

    Directory of Open Access Journals (Sweden)

    VALENTINA FETINIUC

    2014-10-01

    Full Text Available Banking ethics is a specialized set of ethical standards and rules that should be followed in the activities of financial institutions and employees of the banking sector. But despite the simplicity of the definition, in the modern world, this concept becomes complex and ambiguous. The importance of studying this subject is defined by the fact that the ethical behavior of the bank and bank employees promotes banking. At present there are several conceptions of banking ethics: general ethics, regulated ethics and ethical bank. The most common practice is to regulate internal and external relations of banks and bank workers with ethical codes. At the same time, studies show the existence of problems in the banking standards of ethics, which negatively affects the financial institution. This article is intended to reflect main tendencies and problems of banking ethics at international level and experience of Republic of Moldova in this field.

  13. Natural syntax : English interrogative main clauses

    Directory of Open Access Journals (Sweden)

    Janez Oresnik

    2007-12-01

    Full Text Available Natural Syntax is a developing deductive theory, a branch of Naturalness Theory. The naturalnessjudgements are couched in naturalness scales, whichfollow from the basic parameters (or «axioms» listed at the beginning of the paper. The predictions of the theory are calculated in deductions, whose chief components are apair of naturalness scales and the rules governing the alignment of corresponding naturalness values. Parallel and chiastic alignments are distinguished, in complementary distribution. Chiastic alignment is mandatory in deductions limited to unnatural environments. The paper deals with English interrogative main clauses. Within these, only the interrogatives containing wh-words exclusively insitu constitute an extremely unnatural environment and require chiastic alignment. Otherwiseparallel alignment is used. Earlier publications on Natural Syntax: Kavcic 2005a,b, Oresnik 1999, 2000a,b, 200la-f   2002, 2003a-c, 2002/03, 2004. This list cites only works written in English.

  14. PS main supply: motor-generator set.

    CERN Multimedia

    Maximilien Brice

    2002-01-01

    In picture 04 the motor is on the right in the background and the main view is of the generator. The peak power in each PS cycle drawn from the generator, up to 96 MW, is taken from the rotational kinetic energy of the rotor (a heavy-weight of 80 tons), which makes the rotational speed drop by only a few percent. The motor replenishes the average power of 2 to 4 MW. Photo 05: The motor-generator set is serviced every year and, in particular, bearings and slip-rings are carefully checked. To the left is the motor with its slip-rings visible. It has been detached from the axle and moved to the side, so that the rotor can be removed from the huge generator, looming at the right.

  15. VAT EVASION INLEBANON: CASES AND MAIN CAUSES

    Directory of Open Access Journals (Sweden)

    Rana Ismail

    2012-07-01

    Full Text Available The Value Added Tax (VAT is a very important source oftreasuryrevenuesinLebanon. It was initially introduced inLebanon in order to reduce the budgetdeficit and help contain the debt. However, VAT evasion growth inLebanon isleading to significant VAT revenue losses because of its size and frequency.In this paper, we will highlight the contribution of VAT to the treasury revenue. Inaddition, our research has led to figure out the most significant VAT evasion casesand the way tax payers evade paying their required VAT or try to have an illegalrefund. From these tax evasion cases, we will pinpoint the main causes of suchevasions and propose solutions to limit as much as possible VAT evasion inLebanon.

  16. Main clinical epidemiological features of lung cancer

    International Nuclear Information System (INIS)

    Costa Montane, Daniel Marino; Prado Lage, Yulien; Lozano Salazar; Jorge Luis

    2011-01-01

    A descriptive and cross-sectional study of 95 patients with lung cancer, discharged from Neumology Service at 'Dr Juan Bruno Zayas Alfonso' General Hospital in Santiago de Cuba, was carried out from January, 2008 to December, 2008 in order to identify the main clinical epidemiological features of the aforementioned disease. A malignancy predominance among men aged between 56 and 65 years old, belonging to urban areas and being heavy smoker (out of 30 cigarettes per day over 30 years ), was found. Those affected without a confirmed histological type and IV clinical stage epidermoid carcinoma were predominant. Most of them had the opportunity to be treated. Increasing and intensifying health promotion and disease prevention campaigns were recommended so as to achieve the population to avoid or quit the smoking habit. (author)

  17. BWR reactor water cleanup system flexible wedge gate isolation valve qualification and high energy flow interruption test

    International Nuclear Information System (INIS)

    DeWall, K.G.; Steele, R. Jr.

    1989-10-01

    This report presents the results of research performed to develop technical insights for the NRC effort regarding Generic Issue 87, ''Failure of HPCI Steam Line Without Isolation.'' Volume III of this report contains the data and findings from the original research performed to assess the qualification of the valves and reported in EGG-SSRE-7387, ''Qualification of Valve Assemblies in High Energy BWR Systems Penetrating Containment.'' We present the original work here to complete the documentation trail. The recommendations contained in Volume III of this report resulted in the test program described in Volume I and II. The research began with a survey to characterize the population of normally open containment isolation valves in those process lines that connect to the primary system and penetrate containment. The qualification methodology used by the various manufacturers identified in the survey is reviewed and deficiencies in that methodology are identified. Recommendations for expanding the qualification of valve assemblies for high energy pipe break conditions are presented. 11 refs., 1 fig., 2 tabs

  18. Main Findings: Lessons to be Learnt

    International Nuclear Information System (INIS)

    2010-01-01

    This section summarizes the main lessons to be learnt from the workshop: 1 - Workshop Methodology: This method of work has proven to be successful. Participants appreciated the high level of interaction with the other colleagues, especially in view of the variety of expertise that was represented at the workshop. The method also affords the participants the opportunity to learn about the status of waste management in the host country, and to come into contact with the main actors. Conversely, the method also affords the host country programme added visibility at the international level. 2 - National Regulations and International Guidance and Bases for Criteria and Regulatory Judgement: There is reasonable consensus amongst national regulations on fundamental regulatory objectives, but much less agreement on the most appropriate criteria. Consensus is nationally and internationally hampered by the lack of common definition of concepts and terms. International guidance is interpreted in different ways in each country. International guidance is rather difficult to interpret, understand and apply. It is important that stakeholders understand the bases for regulatory judgements. 3 - Optimisation: The fundamental goals of optimisation need to be clarified. Optimisation of long-term vs. short-term safety remains problematic. The process of performing optimisation is more important than the numerical or scientific result. A transparent, stepwise and iterative process of decision making is essential for optimisation. The basic, broad rules for decision making and involvement of stakeholders need to be defined in advance. 4 - Technical Indicators for Safe Performance: The relative importance of different safety indicators varies with timescale. There is still much to be done before reaching consensus on the relative importance of different time frames. More discussion is needed on time cut-offs for regulatory compliance. More discussion on the meaning and applicability of

  19. Maine Environmental Vulnerability Index (EVI) Atlas, Maine - 2007, maps in portable document format (NODC Accession 0036827)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — This four volume set of Environmental Vulnerability Index Maps depicts environmental resources along the coast of Maine most at risk from oil spilled into the marine...

  20. Ostial left main coronary stenosis in a frequent flyer.

    LENUS (Irish Health Repository)

    O'Sullivan, John F

    2009-05-15

    A 52 year old gentleman presented with chest pain, after a long distance flight from India; he had made long haul flights every 2 weeks over the last 5 years as part of his job. His ECG revealed T wave inversion in leads V1-3. Cardiac biomarkers including troponin were negative; we proceeded to exercise stress testing (EST). This revealed 2 mm ST depression at 2 min of the standard Bruce protocol, associated with chest pain. He was taken immediately to the coronary catheterization laboratory; engagement of the left main caused pressure damping with 6 French, then 5 French diagnostic Judkins left 4 catheters. An ostial left main stenosis was seen; the right and left coronary trees otherwise had no significant stenoses. He had normal LV function. He underwent inpatient CABG 7 days later.

  1. Salmon Site Remedial Investigation Report, Main Body

    Energy Technology Data Exchange (ETDEWEB)

    US DOE/NV

    1999-09-01

    This Salmon Site Remedial Investigation Report provides the results of activities initiated by the U.S. Department of Energy (DOE) to determine if contamination at the Salmon Site poses a current or future risk to human health and the environment. These results were used to develop and evaluate a range of risk-based remedial alternatives. Located in Lamar County, Mississippi, the Salmon Site was used by the U.S. Atomic Energy Commission (predecessor to the DOE) between 1964 and 1970 for two nuclear and two gas explosions conducted deep underground in a salt dome. The testing resulted in the release of radionuclides into the salt dome. During reentry drilling and other site activities, liquid and solid wastes containing radioactivity were generated resulting in surface soil and groundwater contamination. Most of the waste and contaminated soil and water were disposed of in 1993 during site restoration either in the cavities left by the tests or in an injection well. Other radioactive wastes were transported to the Nevada Test Site for disposal. Nonradioactive wastes were disposed of in pits at the site and capped with clean soil and graded. The preliminary investigation showed residual contamination in the Surface Ground Zero mud pits below the water table. Remedial investigations results concluded the contaminant concentrations detected present no significant risk to existing and/or future land users, if surface institutional controls and subsurface restrictions are maintained. Recent sampling results determined no significant contamination in the surface or shallow subsurface. The test cavity resulting from the experiments is contaminated and cannot be economically remediated with existing technologies. The ecological sampling did not detect biological uptake of contaminants in the plants or animals sampled. Based on the current use of the Salmon Site, the following remedial actions were identified to protect both human health and the environment: (1) the

  2. Main attributes influencing spent nuclear fuel management

    International Nuclear Information System (INIS)

    Andreescu, N.; Ohai, D.

    1997-01-01

    All activities regarding nuclear fuel, following its discharge from the NPP, constitute the spent fuel management and are grouped in two possible back end variants, namely reprocessing (including HLW vitrification and geological disposal) and direct disposal of spent fuel. In order to select the appropriate variant it is necessary to analyse the aggregate fulfillment of the imposed requirements, particularly of the derived attributes, defined as distinguishing characteristics of the factors used in the decision making process. The main identified attributes are the following: - environmental impact, - availability of suitable sites, - non-proliferation degree, -strategy of energy, - technological complexity and technical maturity, -possible further technical improvements, - size of nuclear programme, - total costs, - public acceptance, - peculiarity of CANDU fuel. The significance of the attributes in the Romanian case, taking into consideration the present situation, as a low scenario and a high scenario corresponding to an important development of the nuclear power, after the year 2010, is presented. According to their importance the ranking of attributes is proposed . Subsequently, the ranking could be used for adequate weighing of attributes in order to realize a multi-criteria analysis and a relevant comparison of back end variants. (authors)

  3. CVD - main concepts, applications and restrictions

    International Nuclear Information System (INIS)

    Bliznakovska, B.; Milosevski, M.; Krawczynski, S.; Meixner, C.; Koetter, H.R.

    1993-01-01

    Despite of the fact that the existing literature covering the last two decades is plentiful with data related to CVD, this document is an attempt to provide to a reader a concise information about the nature of CVD technique at production of technologically important materials as well as to point at special references. The text is devided into three separate sections. The first section, The Main Features of CVD, is intended to give a complete comprehensive picture of the CVD technique through process description and characterization. The basic principles of thermodynamics, CVD chemical reactions classification, CVD chemical kinetics aspects and physics of CVD (with particular attention on the gas-flow phenomena) are included. As an additional aspect, in CVD unavoidable aspect however, the role of the coating/substrate compatibility on the overall process was outlined. The second section, CVD Equipment, concerns on the pecularities of the complete CVD unit pointing out the individual significances of the separate parts, i.e. pumping system, reactor chamber, control system. The aim of this section is to create to a reader a basic understanding of the arising problems but connected to be actual CVD performance. As a final goal of this review the reader's attention is turned upon the CVD applications for production of an up-to-date important class of coatings such as multilayer coatings. (orig.)

  4. Proposed Fermilab upgrade main injector project

    International Nuclear Information System (INIS)

    1992-04-01

    The US Department of Energy (DOE) proposes to construct and operate a ''Fermilab Main Injector'' (FMI), a 150 GeV proton injector accelerator, at the Fermi National Accelerator Laboratory (Fermilab) in Batavia, Illinois. The purpose and need for this action are given of this Environmental Assessment (EA). A description of the proposed FMI and construction activities are also given. The proposed FMI would be housed in an underground tunnel with a circumference of approximately 2.1 miles (3.4 kilometers), and the construction would affect approximately 135 acres of the 6,800 acre Fermilab site. The purpose of the proposed FMI is to construct and bring into operation a new 150 GeV proton injector accelerator. This addition to Fermilab's Tevatron would enable scientists to penetrate ever more deeply into the subatomic world through the detection of the super massive particles that can be created when a proton and antiproton collide head-on. The conversion of energy into matter in these collisions makes it possible to create particles that existed only an instant after the beginning of time. The proposed FMI would significantly extend the scientific reach of the Tevatron, the world's first superconducting accelerator and highest energy proton-antiproton collider

  5. Main prospects of EDF's nuclear program

    International Nuclear Information System (INIS)

    Pierre Bacher, M.; Jean Pierrard, M.

    1994-01-01

    Today, EDF is at a half way point in its third major standardized series, the N4 1400 MW series. The main objective agreed upon for this N4 series, was to improve the insertion of man in the control loop. After the TMI accident in 1979, selective improvements had already been introduced in the 900 series ; the 1300 series, which was the underway, was also the object of more significant improvements (for instance a digital control system) ; but it still seemed desirable that all the lessons to be learned from the accident be reflected in the third series : the fully computerized instrumentation and control system developed for the N4 series ?comparable to the one developed for Airbus ?is today the first of its kind in the world. The demand increase for electric energy dramatically stopped in 1993. However, the prospect for the coming years is more optimistic. Electricity consumption is expected to increase slowly by an average 2% per year. In this context, EDF is preparing its energy program for the nest century and takes into account that hydraulic generation capacity will remain stable and that the French thermal units fueled with coal are rather old and will have to be replaced.

  6. The two main theories on dental bruxism.

    Science.gov (United States)

    Behr, Michael; Hahnel, Sebastian; Faltermeier, Andreas; Bürgers, Ralf; Kolbeck, Carola; Handel, Gerhard; Proff, Peter

    2012-03-20

    Bruxism is characterized by non-functional contact of mandibular and maxillary teeth resulting in clenching or grating of teeth. Theories on factors causing bruxism are a matter of controversy in current literature. The dental profession has predominantly viewed peripheral local morphological disorders, such as malocclusion, as the cause of clenching and gnashing. This etiological model is based on the theory that occlusal maladjustment results in reduced masticatory muscle tone. In the absence of occlusal equilibration, motor neuron activity of masticatory muscles is triggered by periodontal receptors. The second theory assumes that central disturbances in the area of the basal ganglia are the main cause of bruxism. An imbalance in the circuit processing of the basal ganglia is supposed to be responsible for muscle hyperactivity during nocturnal dyskinesia such as bruxism. Some authors assume that bruxism constitutes sleep-related parafunctional activity (parasomnia). A recent model, which may explain the potential imbalance of the basal ganglia, is neuroplasticity. Neural plasticity is based on the ability of synapses to change the way they work. Activation of neural plasticity can change the relationship between inhibitory and excitatory neurons. It seems obvious that bruxism is not a symptom specific to just one disease. Many forms (and causes) of bruxism may exist simultaneously, as, for example, peripheral or central forms. Copyright © 2011 Elsevier GmbH. All rights reserved.

  7. Main physical problems of superhigh energy accelerators

    International Nuclear Information System (INIS)

    Lapidus, L.I.

    1979-01-01

    A survey is given of the state and prospects for the scientific researches to be carried out at the largest charged particle accelerators now under construction. The fundamental problems of the elementary particle physics are considered which can be solved on the base of experiments at high-energy accelerators. The problems to be solved involve development of the theory of various quark number, accurate determination of the charged and neutral intermediate vector boson masses in the Weinberg-Salam theory, the problem of production of t-quark, W -+ - and Z deg bosons, Higgs mesons and investigation of their interactions, examination of quark and lepton spectra, studies on the effects of strong interactions. As a result of the investigations on hadrons at maximum momentum transfers, the data on space-time structure at short distances can be obtained. It is emphasized that there are no engineering barriers to the construction of such accelerators. The main problem lies in financial investment. A conclusion is drawn that the next generation of accelerators will be developed on the base of cooperation between many countries [ru

  8. Brand new hall in the main building

    CERN Multimedia

    Corinne Pralavorio

    2014-01-01

    The renovation of the UNIQA and post office premises is getting under way, with their reopening scheduled for the spring.   The renovation of the large hall in the main building (Building 500) has finally reached the home straight. As of this week, building contractors will get to work on the last part – the offices of UNIQA and La Poste. In the last week of November, the two concessions moved their offices across Route Scherrer to the same part of Building 510 where UBS was temporarily housed during the bank’s refurbishment. Their services were therefore unavailable for one day. The renovation work will last until the spring, with the new offices expected to open in May 2015. Between now and then, the windows and insulation will be completely refitted, with a view to reducing heat loss considerably, and, above all, the premises will be modernised to improve customer reception and service. For example, UNIQA’s new premises will feature a confidential area, guarantee...

  9. Fusion reactor blanket-main design aspects

    International Nuclear Information System (INIS)

    Strebkov, Yu.; Sidorov, A.; Danilov, I.

    1994-01-01

    The main function of the fusion reactor blanket is ensuring tritium breeding and radiation shield. The blanket version depends on the reactor type (experimental, DEMO, commercial) and its parameters. Blanket operation conditions are defined with the heat flux, neutron load/fluence, cyclic operation, dynamic heating/force loading, MHD effects etc. DEMO/commercial blanket design is distinguished e.g. by rather high heat load and neutron fluence - up to 100 W/cm 2 and 7 MWa/m 2 accordingly. This conditions impose specific requirements for the materials, structure, maintenance of the blanket and its most loaded components - FW and limiter. The liquid Li-Pb eutectic is one of the possible breeder for different kinds of blanket in view of its advantages one of which is the blanket convertibility that allow to have shielding blanket (borated water) or breeding one (Li-Pb eutectic). Using Li-Pb eutectic for both ITER and DEMO blankets have been considered. In the conceptual ITER design the solid eutectic blanket was carried out. The liquid eutectic breeder/coolant is suggested also for the advanced (high parameter) blanket

  10. Geology of the Cupsuptic quadrangle, Maine

    Science.gov (United States)

    Harwood, David S.

    1966-01-01

    The Cupsuptic quadrangle, in west-central Maine, lies in a relatively narrow belt of pre-Silurian rocks extending from the Connecticut River valley across northern New Hampshire to north-central Maine. The Albee Formation, composed of green, purple, and black phyllite with interbedded-quartzite, is exposed in the core of a regional anticlinorium overlain to the southeast by greenstone of the Oquossoc Formation which in turn is overlain by black slate of the Kamankeag Formation. In the northern part of the quadrangle the Albee Formation is overlain by black slate, feldspathic graywacke, and minor greenstone of the Dixville Formation. The Kamankeag Formation is dated as 1-ate Middle Ordovician by graptolites (zone 12) found near the base of the unit. The Dixville Formation is correlated with the Kamankeag Formation and Oquossoc Formation and is considered to be Middle Ordovician. The Albee Formation is considered to be Middle to Lower Ordovician from correlations with similar rocks in northeastern and southwestern Vermont. The Oquossoc and Kamankeag Formations are correlated with the Amonoosuc and Partridge Formations of northern New Hampshire. The pre-Silurian rocks are unconformably overlain by unnamed rocks of Silurian age in the southeast, west-central, and northwest ninths of the quadrangle. The basal Silurian units are boulder to cobble polymict conglomerate and quartz-pebble conglomerate of late Lower Silurian (Upper Llandovery) age. The overlying rocks are either well-bedded slate and quartzite, silty limestone, or arenaceous limestone. Thearenaceous limestone contains Upper Silurian (Lower Ludlow) brachiopods. The stratified rocks have been intruded by three stocks of biotite-muscovite quartz monzonite, a large body of metadiorite and associated serpentinite, smaller bodies of gabbro, granodiorite, and intrusive felsite, as well as numerous diabase and quartz monzonite dikes. The metadiorite and serpentinite, and possibly the gabbro and granodiorite are Late

  11. Main tendencies meeting future energy demands

    International Nuclear Information System (INIS)

    Flach, G.; Riesner, W.; Ufer, D.

    1989-09-01

    The economic development in the German Democratic Republic within the preceding 10 years has proved that future stable economic growth of about 4 to 4.5% per annum is only achievable by ways including methods of saving resources. This requires due to the close interdependences between the social development and the level of the development in the energy sector long-term growth rates of the national income of 4 to 4.5% per annum at primary energy growth rates of less than 1% per annum. It comprises three main tendencies: 1. Organization of a system with scientific-technical, technological, economic structural-political and educational measures ensuring in the long term less increase of the energy demand while keeping the economic growth at a constant level. 2. The long-term moderate extension and modernization of the GDR's energy basis is characterized by continuing use of the indigenous brown coal resources for the existing power plant capacities and for district heating. 3. The use of modern and safe nuclear power technologies defines a new and in future more and more important element of the energy basis. Currently about 10% of electricity in the GDR are covered by nuclear energy, in 2000 it will be one third, after 2000 the growth process will continue. The experience shows: If conditions of deepened scientific consideration of all technological processes and the use of modern diagnosis and computer technologies as well as permanent improvement of the safety-technological components and equipment are guaranteed an increasing use of such systems for the production of electricity and heat is socially acceptable. Ensuring a high level of education and technical training of everyone employed in the nuclear energy industry, strict safety restrictions and independent governmental control of these restrictions are important preconditions for the further development in this field. 3 refs, 5 tabs

  12. MAIN TRENDS OF MODERN ART EDUCATION

    Directory of Open Access Journals (Sweden)

    Alla Kozyr

    2016-04-01

    Full Text Available The article deals with the features of the development of art education in Ukraine from the standpoint of the philosophy of modern education. Requirements of educational training standards for future teachers are outlined. This requirements determine not only what teachers need to know, to be able to do, what skills have, but also to create a construction of teacher’s highly skilled work, focused on achieving acmeologic level with the fundamental principles of national identity, socio-cultural conformity, humanism, problematic, learner-centered approach and dialogization in the learning process. The importance of the problem of science-based strategy and tactics of the further development of teacher education as an integrated system of training highly qualified specialists are disclosed due to the modern issues of the educational process. The solution to this problem is aimed at improving the content of professional education of future teachers, which is associated with significant changes in the quality teachers training, which the present day requires. The possibility of improvement of this training can be realized thanks to the classification, validate and implement an integrated system of the formation of professional skills of future teachers, aimed at optimal achievement of the goal. Thus, the main trends of today’s art education are determined. Considering the professional skills of prospective music teachers can only be provided by methodological analysis of this concept, based on the provisions of the dialectical method, system approach and the theory of knowledge and research of the phenomenon of education as a condition of the formation of the perfect work of teachers, because it will justify the skill as a multifactorial and multifaceted phenomenon of education.

  13. Hotel Intercontinental, en Frankfurt-Main (Alemania

    Directory of Open Access Journals (Sweden)

    Apel, Otto

    1966-09-01

    Full Text Available This building has basements, and 21 storeys above the ground. The basements are occupied with the usual services, and a garage, for clients, capable of holding 500 cars. The concierge’s office, the reception office, main hall, restaurant and services are on the ground floor. On the first floor there is a large ball room, where 600 people can assemble, as well as other reception halls. On each of the 2nd to 18th floors there are thirty rooms of one type, and eighteen of another. A luxury apartment occupies the 19th floor, and a restaurant for 200 people as well as a bar is on the top floor. This hotel, which is slab shaped, fits well with its environment and is very pleasantly proportioned.Consta de: planta de sótanos, planta baja y 20 plantas de altura. La planta de sótano contiene las distintas instalaciones; dependencias del personal, etcétera, y un garaje, reservado para los viajeros, capaz de albergar 500 coches. La planta baja comprende: la conserjería, recepción, hall, restaurantes, etc. En la planta primera existe una gran sala de baile para 600 personas, salas de reuniones, etc. Las plantas, de la 2.ª a la 18.ª contienen treinta habitaciones de un tipo y dieciocho de otro, en cada planta. En la planta 19.ª hay un apartamento de lujo. Y en la última planta se distribuyen: un restaurante para 200 personas, bar, etc. El bloque, dentro de su configuración paralelepipédica, es de forma armónica y está perfectamente encajado en el paisaje urbano circundante.

  14. The AGS main magnet power supply upgrade

    International Nuclear Information System (INIS)

    Sandberg, J.N.; Casella, R.; Geller, J.; Marneris, I.; Soukas, A.; Schumburg, N.

    1995-01-01

    The AGS Main Magnet Power Supply consists of a group of thyristor controlled power converters that operate from full rectify to full invert. In order to minimize ripple during the critical periods of injection and extraction 24 pulse converters are used for these portions of the cycle. The maximum voltage available in this mode is nominally 2,000 volts. The converters that are functional during this portion of the cycle are called the flat-top bank or ''F'' bank modules. During acceleration and invert where voltages of up to 12,000 volts are needed and where the ripple requirements are less stringent, groups of twelve pulse converters are operational. These converters are called the Pulsed bank or ''P'' bank modules. The original controlled rectifier system consisted of 96 large mercury filled excitron tubes divided equally between the P bank and F bank converters. These devices were extremely durable and ran successfully for over twenty years. It was, decided to replace the excitron farm with multiple arrangements of three-phase, full-wave, bridge modules that utilize silicon controlled rectifiers (SCR's or thyristors) as the switching element. In order to match the existing transformer connections and buswork, eight identical modules were required; four for the P bank system and four for the F bank system. In order to reduce noise pickup and provide electrical isolation the high level SCR gate triggers are provided via fiberoptic cable. The status of various parameters such as water flow, auxiliary power supply performance, trigger circuitry failure, over voltage, overcurrent, and loss of phase reference are monitored via a programmable logic controller (PLCs). The PLCs use isolated input and output modules for various voltage levels from TTL to 150 Vdc to 125 Vac. These devices are extremely flexible and have allowed modifications and improvements that have enhanced the performance over any equivalent hard wired system

  15. CAA modeling of helicopter main rotor in hover

    Directory of Open Access Journals (Sweden)

    Kusyumov Alexander N.

    2017-01-01

    Full Text Available In this work rotor aeroacoustics in hover is considered. Farfield observers are used and the nearfield flow parameters are obtained using the in house HMB and commercial Fluent CFD codes (identical hexa-grids are used for both solvers. Farfield noise at a remote observer position is calculated at post processing stage using FW–H solver implemented in Fluent and HMB. The main rotor of the UH-1H helicopter is considered as a test case for comparison to experimental data. The sound pressure level is estimated for different rotor blade collectives and observation angles.

  16. Quench Heater Experiments on the LHC Main Superconducting Magnets

    OpenAIRE

    Rodríguez-Mateos, F; Pugnat, P; Sanfilippo, S; Schmidt, R; Siemko, A; Sonnemann, F

    2000-01-01

    In case of a quench in one of the main dipoles and quadrupoles of CERN's Large Hadron Collider (LHC), the magnet has to be protected against excessive temperatures and high voltages. In order to uniformly distribute the stored magnetic energy in the coils, heater strips installed in the magnet are fired after quench detection. Tests of different quench heater configurations were performed on various 1 m long model and 15 m long prototype dipole magnets, as well as on a 3 m long prototype quad...

  17. CAA modeling of helicopter main rotor in hover

    Science.gov (United States)

    Kusyumov, Alexander N.; Mikhailov, Sergey A.; Batrakov, Andrey S.; Kusyumov, Sergey A.; Barakos, George

    In this work rotor aeroacoustics in hover is considered. Farfield observers are used and the nearfield flow parameters are obtained using the in house HMB and commercial Fluent CFD codes (identical hexa-grids are used for both solvers). Farfield noise at a remote observer position is calculated at post processing stage using FW-H solver implemented in Fluent and HMB. The main rotor of the UH-1H helicopter is considered as a test case for comparison to experimental data. The sound pressure level is estimated for different rotor blade collectives and observation angles.

  18. PLEIADES SYSTEM ARCHITECTURE AND MAIN PERFORMANCES

    Directory of Open Access Journals (Sweden)

    M. A. Gleyzes

    2012-07-01

    Full Text Available France, under the leadership of the French Space Agency (CNES, has set up a cooperative program with Austria, Belgium, Spain, Sweden, in order to develop a space Earth Observation system called PLEIADES. PLEIADES is a dual system, this means that it is intended to fulfill an extended panel of both civilian and Defense user’s needs.. This paper reports the status of the satellite after its launch and the in orbit commissioning, the PLEIADES satellite first model has been launched at the end of year 2011, the second model will be launched about 12 months later. It describes the main mission characteristics and performances status. It exposes how the system, satellite and ground segment have been designed in order to be compliant with a dual exploitation between civilian and defense partners. The system is based on the use of a set of newly European developed technologies to feature the satellite. In order to maximize the agility of the satellite, weight and inertia have been reduced using a compact hexagonal shape for the satellite bus. The optical mission consists in Earth optical observation composed of 0.7 m nadir resolution for the panchromatic band and 2.8 m nadir resolution for the four multi-spectral bands. The image swath is about 20 km. PLEIADES delivers optical high resolution products consisting in a Panchromatic image, into which is merged a four multispectral bands image, orthorectified on a Digital Terrain Model (DTM. Thanks to the huge satellite agility obtained with control momentum gyros as actuators, the optical system delivers as well instantaneous stereo images, under different stereoscopic conditions and mosaic images, issued from along the track thus enlarging the field of view. The ground segment is composed of a dual ground center located in CNES Toulouse premises in charge of preparing the dual mission command plan and of the real time contacts with the satellite through a control center. The dual ground center

  19. Apparent Brecciation Gradient, Mount Desert Island, Maine

    Science.gov (United States)

    Hawkins, A. T.; Johnson, S. E.

    2004-05-01

    Mount Desert Island, Maine, comprises a shallow level, Siluro-Devonian igneous complex surrounded by a distinctive breccia zone ("shatter zone" of Gilman and Chapman, 1988). The zone is very well exposed on the southern and eastern shores of the island and provides a unique opportunity to examine subvolcanic processes. The breccia of the Shatter Zone shows wide variation in percent matrix and clast, and may represent a spatial and temporal gradient in breccia formation due to a single eruptive or other catastrophic volcanic event. The shatter zone was divided into five developmental stages based on the extent of brecciation: Bar Harbor Formation, Sols Cliffs breccia, Seeley Road breccia, Dubois breccia, and Great Head breccia. A digital camera was employed to capture scale images of representative outcrops using a 0.5 m square Plexiglas frame. Individual images were joined in Adobe Photoshop to create a composite image of each outcrop. The composite photo was then exported to Adobe Illustrator, which was used to outline the clasts and produce a digital map of the outcrop for analysis. The fractal dimension (Fd) of each clast was calculated using NIH Image and a Euclidean distance mapping method described by Bérubé and Jébrak (1999) to quantify the morphology of the fragments, or the complexity of the outline. The more complex the fragment outline, the higher the fractal dimension, indicating that the fragment is less "mature" or has had less exposure to erosional processes, such as the injection of an igneous matrix. Sols Cliffs breccia has an average Fd of 1.125, whereas Great Head breccia has an average Fd of 1.040, with the stages between having intermediate values. The more complex clasts of the Sols Cliffs breccia with a small amount (26.38%) of matrix material suggests that it is the first stage in a sequence of brecciation ending at the more mature, matrix-supported (71.37%) breccia of Great Head. The results of this study will be used to guide isotopic

  20. Main Achievements 2003-2004 - Nuclear Physics

    International Nuclear Information System (INIS)

    2005-01-01

    Two Departments of our Institute are engaged in nuclear studies, in the following areas: studies of the nuclear reaction mechanism at low, intermediate and high energies, studies of nuclear structure by means of gamma spectroscopy, and theoretical research concerning nuclear structure and reaction mechanisms. Most of these studies are carried out in the form of international collaborations with the world-leading nuclear physics experimental facilities. Our physicists usually play an important role in these collaborative projects and often lead them. Nuclear structure experiments were performed mainly within the following European Large Scale Facilities: ALPI-INFN-Legnaro, VIVITRONIReS-Strasbourg, UNILAC/SIS-GSI-Darmstadt, K100-Cyclotron-Jyvaeskylea with the use of the GASP, GARFIELD, EUROBALL, ICARE, RISING + FRS, RITU+JUROGAM systems and with the application of RFD, HECTOR, DIAMANT, EUCLIDES ancillary detectors. Experimental data were also obtained at the Argonne National Laboratory, USA, with the GAMMASPHERE array and the ATLAS accelerator. In addition, we are involved in planning the experiments for the project of international accelerator facility of the next generation FAIR (Facility for Antiproton and Ion Research) at GSI. The nuclear reaction experiments were performed at the Joint Institute of Nuclear Physics in Dubna (collaborations FASA and COMBAS), in GANIL in Caen, in the Forschungszentrum Juelich at the accelerator COSY in the framework of collaboration PISA, as well as at the Warsaw Laboratory of Heavy Ions. The hadronic nuclear physics experiments were carried out exclusively at the Forschungszentrum Juelich where we have participated in international collaborations COSY11, GEM and HIRES. Recently, we have joined international detector project WASA planned at Forschungszentrum Juelich and plan to participate in the project PANDA, being constructed in GSI Darmstadt. Both detectors will be devoted to low and intermediate hadronic physics. We also

  1. MAINS: MULTI-AGENT INTELLIGENT SERVICE ARCHITECTURE FOR CLOUD COMPUTING

    Directory of Open Access Journals (Sweden)

    T. Joshva Devadas

    2014-04-01

    Full Text Available Computing has been transformed to a model having commoditized services. These services are modeled similar to the utility services water and electricity. The Internet has been stunningly successful over the course of past three decades in supporting multitude of distributed applications and a wide variety of network technologies. However, its popularity has become the biggest impediment to its further growth with the handheld devices mobile and laptops. Agents are intelligent software system that works on behalf of others. Agents are incorporated in many innovative applications in order to improve the performance of the system. Agent uses its possessed knowledge to react with the system and helps to improve the performance. Agents are introduced in the cloud computing is to minimize the response time when similar request is raised from an end user in the globe. In this paper, we have introduced a Multi Agent Intelligent system (MAINS prior to cloud service models and it was tested using sample dataset. Performance of the MAINS layer was analyzed in three aspects and the outcome of the analysis proves that MAINS Layer provides a flexible model to create cloud applications and deploying them in variety of applications.

  2. Leakage detection in underground gas mains with radioaktive argon

    International Nuclear Information System (INIS)

    Schmitz, J.

    1975-01-01

    In the field of gas supply, radionuclide techniques are suitable for the routine monitoring of transport mains by means of highly active tracer clouds and a measuring scraper as well as for exact leakage detection in local gas distribution systems. Very good results are obtained in the case of mains lying lower than 1.5 m if the length and alignment of the pipe section allow a towing probe to be pulled through. This was investigated systematically on a model stretch under practical conditions. The attempts to detect leakages were made with the aid of the radioactive isotope 41 Ar. Under conditions close to practice concerning pipe bedding, branching, pre-pressure, and leakage diameter, a leak with leakage rates as small as approx. 1 l/min could be measured with the aid of a towing probe with a precision of +-0.5 m. This accuracy is another advantage of this method. Branching and fittings with a big dead volume do not interfere with the evaluation. The investment for this method can be compared to other physical/technical investigations on mains, e.g. weld seam tests. (orig./LN) [de

  3. America Saves! Energizing Main Street's Small Businesses

    Energy Technology Data Exchange (ETDEWEB)

    Lindberg, James [National Trust for Historic Preservation, Washington, DC (United States)

    2016-09-30

    The America Saves! Energizing Main Street Small Businesses project engaged the 1,200-member National Main Street Center (NMSC) network of downtown organizations and other local, regional, and national partners to test a methodology for sharing customized energy efficiency information with owners of commercial buildings smaller than 50,000 square feet. Led by the National Trust for Historic Preservation’s Preservation Green Lab, the project marshalled local staff and volunteers to gather voluntarily-disclosed energy use information from participating businesses. This information was analyzed using a remote auditing tool (validated by the National Renewable Energy Lab) to assess energy savings opportunities and design retrofit strategies targeting seven building types (food service and sales, attached mixed-use, strip mall, retail, office, lodging, and schools). The original project design contemplated extensive leveraging of the Green Button protocol for sharing annualized utility data at a district scale. Due the lack of adoption of Green Button, the project partners developed customized approaches to data collection in each of twelve pilot communities. The project team encountered considerable challenges in gathering standardized annual utility data from local partners. After overcoming these issues, the data was uploaded to a data storehouse. Over 450 properties were benchmarked and the remote auditing tool was tested using full building profiles and utility records for more than 100 commercial properties in three of the pilot communities. The audit tool demonstrated potential for quickly capturing, analyzing, and communicating energy efficiency opportunities in small commercial buildings. However, the project team found that the unique physical characteristics and use patterns (partial vacancy, periodic intensive uses) of small commercial buildings required more trouble-shooting and data correction than was anticipated. In addition, the project revealed that

  4. Modern wind energy technology for Russian applications. Main report

    Energy Technology Data Exchange (ETDEWEB)

    Hauge Madsen, P.; Winther-Jensen, M., Bindner, H.W. [and others

    1999-05-01

    The general objective of the project is to establish a technical foundation for an intensified application of wind energy in Russia with medium to large wind turbines and transfer/adaptation of Danish and European wind turbine technology as a basis for future joint ventures and technology exports. More specifically, the objective is to develop and establish the basic knowledge and design criteria for adaptation and development of Danish wind turbine technology for application under Russian conditions. The research programme is envisaged to be carried out in three phases, the first phase being the project reported herein. The main purpose of phase 1 is to assess the needs for modifications and adaptations of established standard (in casu Danish) wind turbine designs for decentralised energy systems with a limited number of medium sized wind turbines and for grid connected wind turbines in cold climate and in-land sites of Russia. As part of this work it is necessary to clarify the types of operational conditions and requirements that are to be met by wind turbines operating in such conditions, and to outline suitable test procedures and test set-up is for verifications of such adapted and modified wind turbines. The reporting of this project is made in one main report and four topical reports, all of them issued as Risoe reports. This is the Main Report, (Risoe-R-1069), summing up the activities and findings of phase 1 and outlining a strategy for Russian-Danish cooperation in wind energy as agreed upon between the Russian and the Danish parties. (au)

  5. Improved Main Shaft Seal Life in Gas Turbines Using Laser Surface Texturing

    Science.gov (United States)

    McNickle, Alan D.; Etsion, Izhak

    2002-10-01

    This paper presents a general overview of the improved main shaft seal life in gas turbines using laser surface texturing (LST). The contents include: 1) Laser Surface Texturing System; 2) Seal Schematic with LST applied; 3) Dynamic Rig Tests; 4) Surface Finish Definitions; 5) Wear Test Rig; 6) Dynamic Test Rig; 7) Seal Cross Section-Rig Test; and 8) Typical Test Results. This paper is in viewgraph form.

  6. Sensitivity to the Main Allergens in Children with Allergic Diseases

    Directory of Open Access Journals (Sweden)

    O.D. Kuznietsova

    2015-11-01

    Objective of the research — to study hypersensitivity to the main allergens in children with allergic diseases based on the results of skin allergy testing, as well as to analyze the structure of diseases. Materials and methods. We have examined 228 children using skin prick testing, the estimation of results was conducted 25–40 minutes after performing the test. Associations between the results of skin prick test with various allergens were studied using cross-correlation analysis in the package of applied statistics Statistics 6.0. Results. 85.5 % of children were sensitized to the pollen allergens, domestic — 54 %, food — 21 %, fungal allergens — 35 %. Among pollen plants, there prevails sensitization to ambrosia — 47.8 %, sunflower — 49.5 %, cyclachaena — 38.5 %; among domestic allergens — to the tick species D.рteronyssinus and D.farinae — 24 %, cat hair — 19.7 %, among fungal — to Alternaria (23 %. Most often hyperergic reaction (papule diameter ≥ 8 mm was observed to cyclachaena (44 %, sunflower (46 %, ambrosia (50 %, cat hair (42 %, D.farinae (39 %. We have established significant (р < 0.05 correlations of mainly middle strength between positive prick-tests in pairs: ambrosia — cyclachaena (r = +0.43, ambrosia — sunflower (r = +0.43, acarus D.рteronyssinus — D.farinae (r = +0.66, mixture «birch, alder, oak, hazel» — ryegrass (r = +0.53, beef meat — egg yolk (r = +0.42, pork meat — chicken meat (r = +0.35, milk (r = +0.36, wool of sheep — pork (r = +0.36. Conclusion. Predominance of sensitization to pollen allergens represents the epidemiological situation in the South region of Ukraine. The presence of correlations between the different types of allergens indicates the cross reactions between them. In case of multiple positive results of skin allergen tests, the study using molecular allergy diagnostic method is recommended to establish genuine or cross allergy.

  7. Neuropsychological testing.

    Science.gov (United States)

    Zucchella, Chiara; Federico, Angela; Martini, Alice; Tinazzi, Michele; Bartolo, Michelangelo; Tamburin, Stefano

    2018-06-01

    Neuropsychological testing is a key diagnostic tool for assessing people with dementia and mild cognitive impairment, but can also help in other neurological conditions such as Parkinson's disease, stroke, multiple sclerosis, traumatic brain injury and epilepsy. While cognitive screening tests offer gross information, detailed neuropsychological evaluation can provide data on different cognitive domains (visuospatial function, memory, attention, executive function, language and praxis) as well as neuropsychiatric and behavioural features. We should regard neuropsychological testing as an extension of the neurological examination applied to higher order cortical function, since each cognitive domain has an anatomical substrate. Ideally, neurologists should discuss the indications and results of neuropsychological assessment with a clinical neuropsychologist. This paper summarises the rationale, indications, main features, most common tests and pitfalls in neuropsychological evaluation. © Article author(s) (or their employer(s) unless otherwise stated in the text of the article) 2018. All rights reserved. No commercial use is permitted unless otherwise expressly granted.

  8. The main chemical safety problems in main process of nuclear fuel reprocessing plant

    International Nuclear Information System (INIS)

    Song Fengli; Zhao Shangui; Liu Xinhua; Zhang Chunlong; Lu Dan; Liu Yuntao; Yang Xiaowei; Wang Shijun

    2014-01-01

    There are many chemical reactions in the aqueous process of nuclear fuel reprocessing. The reaction conditions and the products are different so that the chemical safety problems are different. In the paper the chemical reactions in the aqueous process of nuclear fuel reprocessing are described and the main chemical safety problems are analyzed. The reference is offered to the design and accident analysis of the nuclear fuel reprocessing plant. (authors)

  9. Development of a test bed for operator aid and advanced control concepts in nuclear power plants

    International Nuclear Information System (INIS)

    Turinsky, P.J.; Doster, J.M.; Kim, K.D.; Al-Chalabi, R.M.; Khedro, T.; Sues, R.H.; Yacout, A.M.

    1990-01-01

    A great amount of research and development is currently under way in the utilization of artificial intelligence (AI), expert system, and control theory advances in nuclear power plants as a basis for operator aids and automatic control systems. This activity requires access to the measured dynamic responses of the plant to malfunction, operator- or automatic-control-initiated actions. This can be achieved by either simulating plant behavior or by using an actual plant. The advantage of utilizing an actual plant versus a simulator is that the true behavior is assured of both the power generation system and instrumentation. Clearly, the disadvantages of using an actual plant are availability due to licensing, economic, and risk constraints and inability to address accident conditions. In this work the authors have decided to employ a functional one-ninth scale model of a pressurized water reactor (PWR). The scaled PWR (SPWR) facility is a two-loop representation of a Westinghouse PWR utilizing freon as the working fluid and electric heater rods for the core. The heater rods are driven by a neutron kinetics model accounting for measured thermal core conditions. A control valve in the main steam line takes the place of the turbine generator. A range of normal operating and accident situations can be addressed. The SPWR comes close to offering all the advantages of both a simulator and an actual physical plant in regard to research and development on AI, expert system, and control theory applications. The SPWR is being employed in the development of an expert-system-based operator aid system. The current status of this project is described

  10. Analysis of large scale tests for AP-600 passive containment cooling system

    International Nuclear Information System (INIS)

    Sha, W.T.; Chien, T.H.; Sun, J.G.; Chao, B.T.

    1997-01-01

    One unique feature of the AP-600 is its passive containment cooling system (PCCS), which is designed to maintain containment pressure below the design limit for 72 hours without action by the reactor operator. During a design-basis accident, i.e., either a loss-of-coolant or a main steam-line break accident, steam escapes and comes in contact with the much cooler containment vessel wall. Heat is transferred to the inside surface of the steel containment wall by convection and condensation of steam and through the containment steel wall by conduction. Heat is then transferred from the outside of the containment surface by heating and evaporation of a thin liquid film that is formed by applying water at the top of the containment vessel dome. Air in the annual space is heated by both convection and injection of steam from the evaporating liquid film. The heated air and vapor rise as a result of natural circulation and exit the shield building through the outlets above the containment shell. All of the analytical models that are developed for and used in the COMMIX-ID code for predicting performance of the PCCS will be described. These models cover governing conservation equations for multicomponents single phase flow, transport equations for the κ-ε two-equation turbulence model, auxiliary equations, liquid-film tracking model for both inside (condensate) and outside (evaporating liquid film) surfaces of the containment vessel wall, thermal coupling between flow domains inside and outside the containment vessel, and heat and mass transfer models. Various key parameters of the COMMIX-ID results and corresponding AP-600 PCCS experimental data are compared and the agreement is good. Significant findings from this study are summarized

  11. A study on the shell wall thinning causes identified through experiment, numerical analysis and ultrasonic test of high-pressure feedwater heater

    International Nuclear Information System (INIS)

    Hwang, Kyeong Mo; Woo, Lee; Jin, Tae Eun; Kim, Kyung Hoon

    2008-01-01

    Feedwater heaters of many nuclear power plants have recently experienced severe wall thinning damage, which accelerates as the operation progresses. Several nuclear power plants in Korea have undergone this damage around the impingement baffle - installed downstream of the high-pressure turbine extraction steam line - inside numbers 5A and 5B feedwater heaters. At that point, the extracted steam from the high-pressure turbine consists in the form of two-phase fluid at high temperature, high pressure and high velocity. Since it flows in reverse direction after impinging the impingement baffle, the shell wall of number 5 high-pressure feedwater heater may be affected by flow-accelerated corrosion. This paper describes the comparisons between the numerical analysis results using the FLUENT code and the downscaled experimental data in an effort to determine root causes of the shell wall thinning of the high-pressure feedwater heaters. The numerical analysis and experimental data were also confirmed by the actual wall thickness measured by ultrasonic tests. From the comparison of the results for the local velocity profiles and the wall thinning measurements, the local velocity component only in the y-direction flowing vertically to the shell wall, and not in the x- and z-directions, was analogous to the wall thinning data

  12. MAIN-testi kasutamine eesti laste jutustamisoskuse hindamiseks

    Directory of Open Access Journals (Sweden)

    Andra Kütt

    2018-04-01

    Full Text Available "Using the Multilingual Assessment Instrument for Narratives test for the assessment of Estonian children’s narrative skills" In the Estonian language context, the Multilingual Assessment Instrument for Narratives (MAIN has not been used for research purposes. A total of 18 children (9 boys and 9 girls between the ages of 4 and 8 took part in trial tests: 15 Estonian mother-tongue and 3 Estonian-Russian bilingual mother-tongue children. The article was an analysis of the suitability of the test and an analysis of the test’s preliminary results in assessing Estonian children’s narrative skills. The preliminary test showed that storytelling is difficult for children. It emerged that when producing a narrative, it was not natural for children to indicate a starting point (giving time or place, but rather storytelling began immediately with a conflict and goal as internal components. Challenges and goals also turned out to be the most common and therefore the simplest content in the children’s narratives. There was little complexity in the children’s narratives, and most presented individual goals without conflicts or outcomes. Least common was all three macrostructural components (goal-conflict-outcome in succession in the children’s stories. Stories that achieved higher scores in production were also more complex and these displayed components and sequencing of components that show the cohesion and completeness of the work. Stories that achieved lower scores rather contain individual goals of narrative production, but with conflicts and outcomes to a lesser extent. In the Estonian children’s narratives, there were few words referring to a person’s inner feelings or reactions. The results showed that Estonian children find it difficult to use emotion words, which is evidenced by their limited use of IST words (including emotion words as well as their null-rated understanding of the internal reactions and states of individuals. No

  13. Advanced Health Management System for the Space Shuttle Main Engine

    Science.gov (United States)

    Davidson, Matt; Stephens, John; Rodela, Chris

    2006-01-01

    Pratt & Whitney Rocketdyne, Inc., in cooperation with NASA-Marshall Space Flight Center (MSFC), has developed a new Advanced Health Management System (AHMS) controller for the Space Shuttle Main Engine (SSME) that will increase the probability of successfully placing the shuttle into the intended orbit and increase the safety of the Space Transportation System (STS) launches. The AHMS is an upgrade o the current Block II engine controller whose primary component is an improved vibration monitoring system called the Real-Time Vibration Monitoring System (RTVMS) that can effectively and reliably monitor the state of the high pressure turbomachinery and provide engine protection through a new synchronous vibration redline which enables engine shutdown if the vibration exceeds predetermined thresholds. The introduction of this system required improvements and modification to the Block II controller such as redesigning the Digital Computer Unit (DCU) memory and the Flight Accelerometer Safety Cut-Off System (FASCOS) circuitry, eliminating the existing memory retention batteries, installation of the Digital Signal Processor (DSP) technology, and installation of a High Speed Serial Interface (HSSI) with accompanying outside world connectors. Test stand hot-fire testing along with lab testing have verified successful implementation and is expected to reduce the probability of catastrophic engine failures during the shuttle ascent phase and improve safely by about 23% according to the Quantitative Risk Assessment System (QRAS), leading to a safer and more reliable SSME.

  14. Maine winter roads : salt, safety, environment and cost.

    Science.gov (United States)

    2010-02-01

    This report presents the results of a fourteen-month effort by a research team from the University of : Maine in cooperation with the Maine Department of Transportation (MaineDOT) to conduct : research, engage stakeholders, provide information, and f...

  15. Main directions of Research Institute of Experimental and Theoretic Physics

    International Nuclear Information System (INIS)

    Tazhibaeva, I.L.

    1997-01-01

    The characteristic of main directions of the Research Institute of Experimental and Theoretic Physics (RIETF) activity is given in the paper. It is noted, that Institute is headquarters organisation in 4 following scientific programs of Ministry of Science - Academy of Science of Republic of Kazakhstan: Physics and mechanics of gases, plasma and liquid; Theoretical physics; Nonlinear processes and structural self-organization of substance; Research works Comet. Since 1994 RIETF is one of executors on interstate scientific program ITER. There are following priorities in activity of the institute: - actual problems of relativity theory, gravitation and quantum mechanics; - research on combustion problems and heat-mass-transfer; - physics of gases, plasma and liquid; physics non-equilibrium processes in plasma an in plasma-similar media; - solid state physics and material testing problems; modification of materials properties; electrophysical, optical and structural researches of substance; - interactions of nuclear, electromagnet radiation and accelerated particles with substance; - theoretical and experimental nuclear physics and physics of cosmic rays

  16. Maine Tidal Power Initiative: Environmental Impact Protocols For Tidal Power

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, Michael Leroy [Univ. of Maine, Orono, ME; Zydlewski, Gayle Barbin [Univ. of Maine, Orono, ME; Xue, Huijie [Univ. of Maine, Orono, ME; Johnson, Teresa R. [Univ. of Maine, Orono, ME

    2014-02-02

    The Maine Tidal Power Initiative (MTPI), an interdisciplinary group of engineers, biologists, oceanographers, and social scientists, has been conducting research to evaluate tidal energy resources and better understand the potential effects and impacts of marine hydro-kinetic (MHK) development on the environment and local community. Project efforts include: 1) resource assessment, 2) development of initial device design parameters using scale model tests, 3) baseline environmental studies and monitoring, and 4) human and community responses. This work included in-situ measurement of the environmental and social response to the pre-commercial Turbine Generator Unit (TGU®) developed by Ocean Renewable Power Company (ORPC) as well as considering the path forward for smaller community scale projects.

  17. Short-term forecasting of turbidity in trunk main networks.

    Science.gov (United States)

    Meyers, Gregory; Kapelan, Zoran; Keedwell, Edward

    2017-11-01

    Water discolouration is an increasingly important and expensive issue due to rising customer expectations, tighter regulatory demands and ageing Water Distribution Systems (WDSs) in the UK and abroad. This paper presents a new turbidity forecasting methodology capable of aiding operational staff and enabling proactive management strategies. The turbidity forecasting methodology developed here is completely data-driven and does not require hydraulic or water quality network model that is expensive to build and maintain. The methodology is tested and verified on a real trunk main network with observed turbidity measurement data. Results obtained show that the methodology can detect if discolouration material is mobilised, estimate if sufficient turbidity will be generated to exceed a preselected threshold and approximate how long the material will take to reach the downstream meter. Classification based forecasts of turbidity can be reliably made up to 5 h ahead although at the expense of increased false alarm rates. The methodology presented here could be used as an early warning system that can enable a multitude of cost beneficial proactive management strategies to be implemented as an alternative to expensive trunk mains cleaning programs. Copyright © 2017 Elsevier Ltd. All rights reserved.

  18. Lubrication of Space Shuttle Main Engine Turbopump Bearings

    Science.gov (United States)

    Gibson, Howard; Munafo, Paul (Technical Monitor)

    2001-01-01

    The Space Shuttle has three main engines that are used for propulsion into orbit. These engines are fed propellants by four turbopumps on each engine. A main element in the turbopump is the bearings supporting the rotor that spins the turbine blades and the pump impeller. These bearings are required to spin at very high speeds, support radial and thrust loads, and have high wear resistance without the benefit of lubrication. The liquid hydrogen and oxygen propellants flow through the bearings to cool the surfaces. The volatile nature of the propellants excludes any conventional means of lubrication. Lubrication for these bearings is provided by the ball separator inside the bearing. The separator is a composite material that supplies a transfer film of lubrication to the rings and balls. New separator materials and lubrication schemes have been investigated at Marshall Space Flight Center in a bearing test rig with promising results. Hybrid bearings with silicon nitride balls have also been evaluated. The use of hybrid, silicon nitride ball bearings in conjunction -with better separator materials has shown excellent results. The work that Marshall has done is being utilized in turbopumps flying on the space shuttle fleet and will be utilized in future space travel. This result of this work is valuable for all aerospace and commercial applications where high-speed bearings are used.

  19. DISSECTING THE QUASAR MAIN SEQUENCE: INSIGHT FROM HOST GALAXY PROPERTIES

    International Nuclear Information System (INIS)

    Sun, Jiayi; Shen, Yue

    2015-01-01

    The diverse properties of broad-line quasars appear to follow a well-defined main sequence along which the optical Fe ii strength increases. It has been suggested that this sequence is mainly driven by the Eddington ratio (L/L Edd ) of the black hole (BH) accretion. Shen and Ho demonstrated with quasar clustering analysis that the average BH mass decreases with increasing Fe ii strength when quasar luminosity is fixed, consistent with this suggestion. Here we perform an independent test by measuring the stellar velocity dispersion σ * (hence, the BH mass via the M–σ * relation) from decomposed host spectra in low-redshift Sloan Digital Sky Survey quasars. We found that at fixed quasar luminosity, σ * systematically decreases with increasing Fe ii strength, confirming that the Eddington ratio increases with Fe ii strength. We also found that at fixed luminosity and Fe ii strength, there is little dependence of σ * on the broad Hβ FWHM. These new results reinforce the framework that the Eddington ratio and orientation govern most of the diversity seen in broad-line quasar properties

  20. Frontiers in Materials Science and Technology

    Indian Academy of Sciences (India)

    R. Narasimhan (Krishtel eMaging) 1461 1996 Oct 15 13:05:22

    Fundamental physical, mechanical, magnetic and biological properties of materials are ... valve bodies, turbine casing and main steam lines. ... through orthopedics and fracture healing, to magneto-motive artificial hearts and pioneering.

  1. 78 FR 57180 - Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses...

    Science.gov (United States)

    2013-09-17

    ... submitted in a timely fashion based on the availability of the subsequent information. For further details... certain Main Steam Line (MSL) radiation monitors from the reference initiating conditions to address...

  2. Learning software testing with Test Studio

    CERN Document Server

    Madi, Rawane

    2013-01-01

    Learning Software Testing with Test Studio is a practical, hands-on guide that will help you get started with Test Studio to design your automated solution and tests. All through the book, there are best practices and tips and tricks inside Test Studio which can be employed to improve your solution just like an experienced QA.If you are a beginner or a professional QA who is seeking a fast, clear, and direct to the point start in automated software testing inside Test Studio, this book is for you. You should be familiar with the .NET framework, mainly Visual Studio, C#, and SQL, as the book's

  3. Arsenic evolution in fractured bedrock wells in central Maine, USA

    Science.gov (United States)

    Yang, Q.; Zheng, Y.; Culbertson, C.; Schalk, C.; Nielsen, M. G.; Marvinney, R.

    2010-12-01

    Elevated arsenic concentration in fractured bedrock wells has emerged as an important and challenging health problem, especially in rural areas without public water supply and mandatory monitoring of private wells. This has posed risks of skin, bladder, prostate diseases and cancers to private well users. In central Maine, including the study site, 31% of bedrock wells in meta-sedimentary formations have been reported of elevated arsenic concentrations of > 10 µg/L. Geophysical logging and fracture specific water sampling in high arsenic wells have been conducted to understand how water flowing through the aquifers enters the boreholes and how arsenic evolves in the fracture bedrock wells. Two domestic wells in Manchester, Maine, located 50 meter apart with 38 µg/L and 73 µg/L of arsenic in unfiltered water, were investigated to characterize fractures by geophysical logging and to determine flow rates by pumping test. Water samples, representing the bore hole and the fractures, were collected and analyzed for arsenic under ambient and pumping conditions. Transmissivity of the fractures was estimated at 0.23-10.6 m2/day. Water with high dissolved arsenic was supplied primarily by high yielding fractures near the bottom of the borehole. Dissolved arsenic concentrations in borehole water increased as fracture water with high arsenic was replacing borehole water with initially low dissolved arsenic in response to pumping. The precipitation of iron particulates enriched in arsenic was common during and after pumping. Laboratory experiment on well water samples over a period of 16 days suggested that in the borehole arsenic was mainly settled with iron enriched particles, likely amorphous ferric oxyhydroxides, with possibly minor adsorption on the iron minerals. Another bedrock well in Litchfield, Maine, with 478 µg/L of arsenic in the unfiltered well water, is being investigated to quantify and reconstruct of the groundwater flow under ambient and pumping conditions

  4. 30 CFR 75.302 - Main mine fans.

    Science.gov (United States)

    2010-07-01

    ... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Main mine fans. 75.302 Section 75.302 Mineral... SAFETY STANDARDS-UNDERGROUND COAL MINES Ventilation § 75.302 Main mine fans. Each coal mine shall be ventilated by one or more main mine fans. Booster fans shall not be installed underground to assist main mine...

  5. Retraining of the 1232 Main Dipole Magnets in the LHC

    Energy Technology Data Exchange (ETDEWEB)

    Verweij, A. [CERN; Auchmann, B.; Bednarek, M.; Bottura, L.; Charifoulline, Z.; Feher, S. [Fermilab; Hagen, P.; Modena, M.; Le Naour, S.; Romera, I.; Siemko, A.; Steckert, J.; Tock, J. Ph; Todesco, E.; Willering, G.; Wollmann, D.

    2016-01-05

    The Large Hadron Collider (LHC) contains eight main dipole circuits, each of them with 154 dipole magnets powered in series. These 15-m-long magnets are wound from Nb-Ti superconducting Rutherford cables, and have active quench detection triggering heaters to quickly force the transition of the coil to the normal conducting state in case of a quench, and hence reduce the hot spot temperature. During the reception tests in 2002-2007, all these magnets have been trained up to at least 12 kA, corresponding to a beam energy of 7.1 TeV. After installation in the accelerator, the circuits have been operated at reduced currents of up to 6.8 kA, from 2010 to 2013, corresponding to a beam energy of 4 TeV. After the first long shutdown of 2013-2014, the LHC runs at 6.5 TeV, requiring a dipole magnet current of 11.0 kA. A significant number of training quenches were needed to bring the 1232 magnets up to this current. In this paper, the circuit behavior in case of a quench is presented, as well as the quench training as compared to the initial training during the reception tests of the individual magnets.

  6. PWR-related integral safety experiments in the PKL 111 test facility SBLOCA under beyond-design-basis accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Weber, P.; Umminger, K.J.; Schoen, B. [Siemens AG Power Generation Group (KWU), Erlangen (France)

    1995-09-01

    The thermal hydraulic behavior of a PWR during beyond-design-basis accident scenarios is of vital interest for the verification and optimization of accident management procedures. Within the scope of the German reactor safety research program experiments were performed in the volumetrically scaled PKL 111 test facility by Siemens/KWU. This highly instrumented test rig simulates a KWU-design PWR (1300 MWe). In particular, the latest tests performed related to a SBLOCA with additional system failures, e.g. nitrogen entering the primary system. In the case of a SBLOCA, it is the goal of the operator to put the plant in a condition where the decay heat can be removed first using the low pressure emergency core cooling system and then the residual heat removal system. The experimental investigation presented assumed the following beyond-design-basis accident conditions: 0.5% break in a cold leg, 2 of 4 steam generators (SGs) isolated on the secondary side (feedwater- and steam line-valves closed), filled with steam on the primary side, cooldown of the primary system using the remaining two steam generators, high pressure injection system only in the two loops with intact steam generators, if possible no operator actions to reach the conditions for residual heat removal system activation. Furthermore, it was postulated that 2 of the 4 hot leg accumulators had a reduced initial water inventory (increased nitrogen inventory), allowing nitrogen to enter the primary system at a pressure of 15 bar and nearly preventing the heat transfer in the SGs ({open_quotes}passivating{close_quotes} U-tubes). Due to this the heat transfer regime in the intact steam generators changed remarkably. The primary system showed self-regulating system effects and heat transfer improved again (reflux-condenser mode in the U-tube inlet region).

  7. 33 CFR 334.20 - Gulf of Maine off Cape Small, Maine; naval aircraft practice mining range area.

    Science.gov (United States)

    2010-07-01

    ... REGULATIONS § 334.20 Gulf of Maine off Cape Small, Maine; naval aircraft practice mining range area. (a) The... 33 Navigation and Navigable Waters 3 2010-07-01 2010-07-01 false Gulf of Maine off Cape Small, Maine; naval aircraft practice mining range area. 334.20 Section 334.20 Navigation and Navigable Waters...

  8. Politics and Politicians – Main Topic and Main Characters on Television News

    Directory of Open Access Journals (Sweden)

    Viktorija Car

    2010-01-01

    Full Text Available The paper examines the relationship between television as a medium, and politics and politicians as the content of television news in Croatia. The first part of the paper explains the models of ‘media logic’, ‘partisan logic’ and ‘party logic’. The second part of the paper presents the results of the research conducted on the representative sample of primetime news HTV Dnevnik for the period 1991-2009, and Nova TV Dnevnik and RTL Vijesti for the period 2005-2009. The goal of the research was to examine the presence of political topics on primetime news, as well to what extent politicians are presented as main characters. The results show a sustained decline of politics on the news and their simultaneous replacement by news on disasters and lifestyle. Further, citizens and their opinion become more important than opinions of politicians, experts and scientists. Comparing the news on public service television and on commercial televisions, the author elaborates on the internal processes and changes of the television medium and finally introduces the concept of ‘power logic’ to explain news selection and news editing on the Croatian TV channels.

  9. Forecasting the Seasonal Timing of Maine's Lobster Fishery

    Directory of Open Access Journals (Sweden)

    Katherine E. Mills

    2017-11-01

    Full Text Available The fishery for American lobster is currently the highest-valued commercial fishery in the United States, worth over US$620 million in dockside value in 2015. During a marine heat wave in 2012, the fishery was disrupted by the early warming of spring ocean temperatures and subsequent influx of lobster landings. This situation resulted in a price collapse, as the supply chain was not prepared for the early and abundant landings of lobsters. Motivated by this series of events, we have developed a forecast of when the Maine (USA lobster fishery will shift into its high volume summer landings period. The forecast uses a regression approach to relate spring ocean temperatures derived from four NERACOOS buoys along the coast of Maine to the start day of the high landings period of the fishery. Tested against conditions in past years, the forecast is able to predict the start day to within 1 week of the actual start, and the forecast can be issued 3–4 months prior to the onset of the high-landings period, providing valuable lead-time for the fishery and its associated supply chain to prepare for the upcoming season. Forecast results are conveyed in a probabilistic manner and are updated weekly over a 6-week forecasting period so that users can assess the certainty and consistency of the forecast and factor the uncertainty into their use of the information in a given year. By focusing on the timing of events, this type of seasonal forecast provides climate-relevant information to users at time scales that are meaningful for operational decisions. As climate change alters seasonal phenology and reduces the reliability of past experience as a guide for future expectations, this type of forecast can enable fishing industry participants to better adjust to and prepare for operating in the context of climate change.

  10. Testing the Main Determinants of Teachers' Professional Well-Being by Using a Mixed Method

    Science.gov (United States)

    Yildirim, Kamil

    2015-01-01

    Measuring the perception of professional well-being appears as a problematic issue in past studies. Literature reviews show that previous studies have not yet reached an agreement on measures of teachers' professional well-being. Furthermore, when we consider the cultural context, a research gap becomes clearer. The aim of this study was to…

  11. Relationship between Main Civilian Occupation and Army General Classification Test Standard Score. Part 2

    Science.gov (United States)

    1945-03-07

    Picture (285) ....... •"■*’ Cameraman, Motion Picture (043) 115 Canvas Cover Renairuan (OhU) ■ * Car Carpenter, Railway (046) i<" Car Mechanic...Film Editor, Motion Picture (l3l) .,,,.,♦ * 15 Filter Operator, ^ tor Supply (O83). # 10 Fingerprinter (307) ’. . * 30 Fire Fighter (383) ,. 128...Mechanic (322) .... Registered Nurse (225) ....... Repairman, Camera (042) Repairman, Canvas Cover (044) . . . Repairman, Central. Of fice (095

  12. Bioactivity tests of calcium phosphates with variant molar ratios of main components.

    Science.gov (United States)

    Pluta, Klaudia; Sobczak-Kupiec, Agnieszka; Półtorak, Olga; Malina, Dagmara; Tyliszczak, Bożena

    2018-03-09

    Calcium phosphates constitute attractive materials of biomedical applications. Among them particular attention is devoted to bioactive hydroxyapatite (HAp) and bioresorbable tricalcium phosphate (TCP) that possess ability to bind to living bones and can be used clinically as important bone substitutes. Notably, in vivo bone bioactivity can be predicted from apatite formation of bone immersed in SBF fluids. Thus, analyses of behavior of calcium phosphates immersed in various bio fluids are of great importance. Recently, stoichiometric HAp and TCP structures have been widely studied, whereas only limited number of publications have been devoted to analyses of nonstoichiometric calcium phosphates. Here, we report physicochemical analysis of natural and synthetic phosphates with variable Ca/P molar ratios. Subsequently attained structures were subjected to incubation in either artificial saliva or Ringer's fluids. Both pH and conductivity of such fluids were determined before and after incubation. Furthermore, the influence of the Ca/P values on such parameters was exemplified. Physicochemical analysis of received materials was performed by XRD and FT-IR characterization techniques. Their potential antibacterial activity and behavior in the presence of infectious microorganisms as Escherichia coli and Staphylococcus aureus was also evaluated. © 2018 Wiley Periodicals, Inc. J Biomed Mater Res Part A, 2018. © 2018 Wiley Periodicals, Inc.

  13. B-1 Aircraft Main Hydraulic Pump Tests With MIL-H-87257 Hydraulic Fluid

    National Research Council Canada - National Science Library

    Sharma, Shashi

    1998-01-01

    In an effort to convert the B-1 aircraft from MIL-H-5606 to M1-H-87257, the Air Force sponsored a study conducted by Rockwell International from April 1991 through June 1992, under contract F34601-89-C-0401...

  14. The main tests for quality control in X-ray equipment of radiodiagnosis

    International Nuclear Information System (INIS)

    Ferreira, R.S.

    1988-01-01

    All aspects of the relation between patient, examination and diagnosis for controling the quality in radiodiagnosis are showed. The bundle collimation for decreasing the scattered radiation in patient and the systems for measuring the exposure time are described. The yield valuation and the tension for X-rays tube are also cited. (C.G.C.) [pt

  15. A kaon physics program at the Fermilab Main Injector

    International Nuclear Information System (INIS)

    Cooper, Peter

    1997-11-01

    In this paper we describe a triad of kaon experiments which will form the foundation of a kaon physics program at Fermilab in the Main Injector era. These three experiments; KAMI, CKM and CPT, span the range of experiment types discussed above. KAMI will use the existing neutral kaon beam and the KTeV detector as the basis of a search for the Standard Model ultra rare decay K L → π 0 ν anti ν decay mode is by far the theoretically cleanest measurement of the Standard Model parameter responsible for CP violation. CKM will measure the analogous charged kaon decay mode. Together these two experiments will determine the Standard Model contribution to CP violation independent of the B meson sector. The Standard Model parameters controlling CP violation must be observed to be the same in the K and B meson sectors in order to confirm the Standard Model as the sole source of CP violation in nature. CPT is a hybrid beam experiment using a high purity K + beam to produce a pure K 0 beam in order to search for violation of CPT symmetry at a mass scale up to the Planck mass. CPT also will measure new CP violation parameters to test the Standard Model and search for rare K S decays. The Fermilab infrastructure for such a physics program largely already exists. The Main Injector will be an existing accelerator by late 1998 with beam properties comparable to any of the previous ''kaon factory'' proposals. The KTeV detector and neutral kaon beamline are unsurpassed in the world and were originally designed to also operate with the 120 GeV Main Injector beam as KAMI. The Fermilab Meson laboratory was originally designed as an area for fixed target experiments using 200 GeV proton beams. The charged kaon beam experiments will naturally find a home there. Both charged kaon experiments, CKM and CPT, will share a new high purity RF separated charged kaon beam based on superconducting RF technology which will provide the highest intensity and purity charged kaon beam in the world

  16. Oahu Sewer Main Lines, Oahu County HI, 2016, Honolulu GIS

    Data.gov (United States)

    U.S. Environmental Protection Agency — Linear features representing sewer main lines as maintained by Honolulu ENV Department of Environmental Services. Includes an inventory of sewer mains used for...

  17. SMALL MAIN-BELT ASTEROID SPECTROSCOPIC SURVEY, PHASE II

    Data.gov (United States)

    National Aeronautics and Space Administration — This data set contains visible-wavelength (0.435-0.925 micron) spectra for 1341 main-belt asteroids observed during the second phase of the Small Main-belt Asteroid...

  18. 46 CFR 108.419 - Fire main capacity.

    Science.gov (United States)

    2010-10-01

    ... COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) A-MOBILE OFFSHORE DRILLING UNITS DESIGN AND EQUIPMENT Fire Extinguishing Systems Fire Main System § 108.419 Fire main capacity. The diameter of the fire... pumps operating simultaneously. ...

  19. Space Shuttle Main Propulsion System Anomaly Detection: A Case Study

    Data.gov (United States)

    National Aeronautics and Space Administration — The space shuttle main engine (SSME) is part of the Main Propnlsion System (MPS) which is an extremely complex system containing several sub-systems and components,...

  20. Automation of radiometric testing

    International Nuclear Information System (INIS)

    Chekalin, A.S.; Temnik, A.K.; Butakova, G.E.; Goncharov, V.I.

    1983-01-01

    The main prerequisites for creation of automatic systems of radiometric testing as the means to increase the testing objectivity and quality have been considered, principles of their design being developed. The operating system is described for testing complex configuration products using RD-10R gamma flow detector as a sensor of initial information

  1. 2017 Census Test

    Science.gov (United States)

    /Programs Latest Information Are you in a Survey? 2020 Census 2018 Census Test 2010 Census American Information Surveys/Programs Main Are you in a Survey? 2020 Census 2018 Census Test If you have received a Census. Latest Information The 2018 Census Test will take place in Pierce County, Wash.; Providence

  2. 14 CFR 23.753 - Main float design.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Main float design. 23.753 Section 23.753... STANDARDS: NORMAL, UTILITY, ACROBATIC, AND COMMUTER CATEGORY AIRPLANES Design and Construction Floats and Hulls § 23.753 Main float design. Each seaplane main float must meet the requirements of § 23.521. [Doc...

  3. Maine Migrant Program: 1997-1998 Program Evaluation.

    Science.gov (United States)

    Bazinet, Suzanne C., Ed.

    The Maine Department of Education contracts with local educational agencies to administer the Maine Migrant Education Program. The program's overall mission is to provide the support necessary for migrant children to achieve Maine's academic standards. In 1997-98, 73 local migrant programs served 9,838 students, and 63 summer programs served 1,769…

  4. 30 CFR 57.8525 - Main fan maintenance.

    Science.gov (United States)

    2010-07-01

    ... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Main fan maintenance. 57.8525 Section 57.8525 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR METAL AND NONMETAL MINE... Underground Only § 57.8525 Main fan maintenance. Main fans shall be maintained according to either the...

  5. 30 CFR 57.8519 - Underground main fan controls.

    Science.gov (United States)

    2010-07-01

    ... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Underground main fan controls. 57.8519 Section... Ventilation Surface and Underground § 57.8519 Underground main fan controls. All underground main fans shall have controls placed at a suitable protected location remote from the fan and preferably on the surface...

  6. 30 CFR 75.311 - Main mine fan operation.

    Science.gov (United States)

    2010-07-01

    ... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Main mine fan operation. 75.311 Section 75.311... MANDATORY SAFETY STANDARDS-UNDERGROUND COAL MINES Ventilation § 75.311 Main mine fan operation. (a) Main mine fans shall be continuously operated, except as otherwise approved in the ventilation plan, or when...

  7. Supply and demand of timber for wood turning in Maine

    Science.gov (United States)

    Eric H. Wharton; Robert L., Jr. Nevel; Douglas S. Powell; Douglas S. Powell

    1987-01-01

    An analytical report on the volume of wood used by the wood-turning industry in Maine, and the volume of timber from the state's timberlands that may be suitable for turnstock. Findings are based on the third forest resource survey of Maine timberlands, and an industry canvass of primary manufacturing mills using wood from Maine timberlands, both conducted in 1982...

  8. 14 CFR 27.547 - Main rotor structure.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Main rotor structure. 27.547 Section 27.547... structure. (a) Each main rotor assembly (including rotor hubs and blades) must be designed as prescribed in this section. (b) [Reserved] (c) The main rotor structure must be designed to withstand the following...

  9. 30 CFR 57.8518 - Main and booster fans.

    Science.gov (United States)

    2010-07-01

    ... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Main and booster fans. 57.8518 Section 57.8518... and Underground § 57.8518 Main and booster fans. (a) All mine main and booster fans installed and used...-cycle shutdowns or planned or scheduled fan maintenance or fan adjustments where air quality is...

  10. 46 CFR 182.610 - Main steering gear.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 7 2010-10-01 2010-10-01 false Main steering gear. 182.610 Section 182.610 Shipping...) MACHINERY INSTALLATION Steering Systems § 182.610 Main steering gear. (a) A vessel must be provided with a main steering gear that is: (1) Of adequate strength and capable of steering the vessel at all service...

  11. Tracing Technological Development Trajectories: A Genetic Knowledge Persistence-Based Main Path Approach.

    Directory of Open Access Journals (Sweden)

    Hyunseok Park

    Full Text Available The aim of this paper is to propose a new method to identify main paths in a technological domain using patent citations. Previous approaches for using main path analysis have greatly improved our understanding of actual technological trajectories but nonetheless have some limitations. They have high potential to miss some dominant patents from the identified main paths; nonetheless, the high network complexity of their main paths makes qualitative tracing of trajectories problematic. The proposed method searches backward and forward paths from the high-persistence patents which are identified based on a standard genetic knowledge persistence algorithm. We tested the new method by applying it to the desalination and the solar photovoltaic domains and compared the results to output from the same domains using a prior method. The empirical results show that the proposed method can dramatically reduce network complexity without missing any dominantly important patents. The main paths identified by our approach for two test cases are almost 10x less complex than the main paths identified by the existing approach. The proposed approach identifies all dominantly important patents on the main paths, but the main paths identified by the existing approach miss about 20% of dominantly important patents.

  12. Retraining of the 1232 Main Dipole Magnets in the LHC

    CERN Document Server

    Verweij, A; Bednarek, M; Bottura, L; Charifoulline, Z; Feher, S; Hagen, P; Modena, M; Le Naour, S; Romera, I; Siemko, A; Steckert, J; Tock, J Ph; Todesco, E; Willering, G; Wollmann, D

    2016-01-01

    The Large Hadron Collider (LHC) contains eight main dipole circuits, each of them with 154 dipole magnets powered in series. These 15-m-long magnets are wound from Nb-Ti superconducting Rutherford cables, and have active quench detection triggering heaters to quickly force the transition of the coil to the normal conducting state in case of a quench, and hence reduce the hot spot temperature. During the reception tests in 2002-2007, all these magnets have been trained up to at least 12 kA, corresponding to a beam energy of 7.1 TeV. After installation in the accelerator, the circuits have been operated at reduced currents of up to 6.8 kA, from 2010 to 2013, corresponding to a beam energy of 4 TeV. After the first long shutdown of 2013-2014, the LHC runs at 6.5 TeV, requiring a dipole magnet current of 11.0 kA. A significant number of training quenches were needed to bring the 1232 magnets up to this current. In this paper, the circuit behavior in case of a quench is presented, as well as the quench training as...

  13. On the Statistical Properties of the Lower Main Sequence

    International Nuclear Information System (INIS)

    Angelou, George C.; Bellinger, Earl P.; Hekker, Saskia; Basu, Sarbani

    2017-01-01

    Astronomy is in an era where all-sky surveys are mapping the Galaxy. The plethora of photometric, spectroscopic, asteroseismic, and astrometric data allows us to characterize the comprising stars in detail. Here we quantify to what extent precise stellar observations reveal information about the properties of a star, including properties that are unobserved, or even unobservable. We analyze the diagnostic potential of classical and asteroseismic observations for inferring stellar parameters such as age, mass, and radius from evolutionary tracks of solar-like oscillators on the lower main sequence. We perform rank correlation tests in order to determine the capacity of each observable quantity to probe structural components of stars and infer their evolutionary histories. We also analyze the principal components of classic and asteroseismic observables to highlight the degree of redundancy present in the measured quantities and demonstrate the extent to which information of the model parameters can be extracted. We perform multiple regression using combinations of observable quantities in a grid of evolutionary simulations and appraise the predictive utility of each combination in determining the properties of stars. We identify the combinations that are useful and provide limits to where each type of observable quantity can reveal information about a star. We investigate the accuracy with which targets in the upcoming TESS and PLATO missions can be characterized. We demonstrate that the combination of observations from GAIA and PLATO will allow us to tightly constrain stellar masses, ages, and radii with machine learning for the purposes of Galactic and planetary studies.

  14. Reconstruction of Twist Torque in Main Parachute Risers

    Science.gov (United States)

    Day, Joshua D.

    2015-01-01

    The reconstruction of twist torque in the Main Parachute Risers of the Capsule Parachute Assembly System (CPAS) has been successfully used to validate CPAS Model Memo conservative twist torque equations. Reconstruction of basic, one degree of freedom drop tests was used to create a functional process for the evaluation of more complex, rigid body simulation. The roll, pitch, and yaw of the body, the fly-out angles of the parachutes, and the relative location of the parachutes to the body are inputs to the torque simulation. The data collected by the Inertial Measurement Unit (IMU) was used to calculate the true torque. The simulation then used photogrammetric and IMU data as inputs into the Model Memo equations. The results were then compared to the true torque results to validate the Model Memo equations. The Model Memo parameters were based off of steel risers and the parameters will need to be re-evaluated for different materials. Photogrammetric data was found to be more accurate than the inertial data in accounting for the relative rotation between payload and cluster. The Model Memo equations were generally a good match and when not matching were generally conservative.

  15. Main considerations for modelling a station blackout scenario with trace

    International Nuclear Information System (INIS)

    Querol, Andrea; Turégano, Jara; Lorduy, María; Gallardo, Sergio; Verdú, Gumersindo

    2017-01-01

    In the nuclear safety field, the thermal hydraulic phenomena that take place during an accident in a nuclear power plant is of special importance. One of the most studied accidents is the Station BlackOut (SBO). The aim of the present work is the analysis of the PKL integral test facility nodalization using the thermal-hydraulic code TRACE5 to reproduce a SBO accidental scenario. The PKL facility reproduces the main components of the primary and secondary systems of its reference nuclear power plant (Philippsburg II). The results obtained with different nodalization have been compared: 3D vessel vs 1D vessel, Steam Generator (SG) modelling using PIPE or TEE components and pressurizer modelling with PIPE or PRIZER components. Both vessel nodalization (1D vessel and 3D vessel) reproduce the physical phenomena of the experiment. However, there are significant discrepancies between them. The appropriate modelling of the SG is also relevant in the results. Regarding the other nodalization (PIPE or TEE components for SG and PIPE or PRIZER components for pressurizer), do not produce relevant differences in the results. (author)

  16. On the Statistical Properties of the Lower Main Sequence

    Energy Technology Data Exchange (ETDEWEB)

    Angelou, George C.; Bellinger, Earl P.; Hekker, Saskia [Max-Planck-Institut für Sonnensystemforschung, Justus-von-Liebig-Weg 3, D-37077 Göttingen (Germany); Basu, Sarbani [Department of Astronomy, Yale University, New Haven, CT 06520 (United States)

    2017-04-20

    Astronomy is in an era where all-sky surveys are mapping the Galaxy. The plethora of photometric, spectroscopic, asteroseismic, and astrometric data allows us to characterize the comprising stars in detail. Here we quantify to what extent precise stellar observations reveal information about the properties of a star, including properties that are unobserved, or even unobservable. We analyze the diagnostic potential of classical and asteroseismic observations for inferring stellar parameters such as age, mass, and radius from evolutionary tracks of solar-like oscillators on the lower main sequence. We perform rank correlation tests in order to determine the capacity of each observable quantity to probe structural components of stars and infer their evolutionary histories. We also analyze the principal components of classic and asteroseismic observables to highlight the degree of redundancy present in the measured quantities and demonstrate the extent to which information of the model parameters can be extracted. We perform multiple regression using combinations of observable quantities in a grid of evolutionary simulations and appraise the predictive utility of each combination in determining the properties of stars. We identify the combinations that are useful and provide limits to where each type of observable quantity can reveal information about a star. We investigate the accuracy with which targets in the upcoming TESS and PLATO missions can be characterized. We demonstrate that the combination of observations from GAIA and PLATO will allow us to tightly constrain stellar masses, ages, and radii with machine learning for the purposes of Galactic and planetary studies.

  17. Main trends of upgrading the 1000 MW steam turbine

    International Nuclear Information System (INIS)

    Drahy, J.

    1990-01-01

    Parameters are compared for the 1000 MW steam turbine manufactured by the Skoda Works, Czechoslovakia, and turbines in the same power range by other manufacturers, viz. ABB, Siemens/KWU, GEC and LMZ. The Skoda turbine compares well with the other turbines with respect to all design parameters, and moreover, enables the most extensive heat extraction for district heating purposes. The main trends in upgrading this turbine are outlined; in particular, they include an additional increase in the heat extraction, which is made possible by a new design of the low-pressure section or by using a ''satellite'' turbine. The studies performed also indicate that the output of the full-speed saturated steam turbine can be increased to 1300 MW. An experimental turbine representing one flow of the high-pressure part of the 1000 MW turbine is being built on the 1:1 scale. It will serve to verify the methods of calculation of the wet steam flow and to experimentally test the high-pressure part over a wide span of the parameters. (Z.M.). 1 tab., 3 figs., 7 refs

  18. Mutual fund flows: an analysis of the main macroeconomic factors

    Directory of Open Access Journals (Sweden)

    Raphael Moses Roquete

    2015-03-01

    Full Text Available This paper analyzes whether some macroeconomic factors (country risk, IBrX volatility and Interbank Certificate of Deposit are related to mutual fund flows for the period between January 2005 and August 2014. In order to investigate whether the flow series behaved differently during this period, the Chow test was conducted for September 2008 (the month in which the Lehman Brothers investment bank collapsed. The regressions were performed and the parameters were estimated through the OLS method for both periods, the first running from January 2005 to August 2008 and the second from September 2008 to August 2014. For the period between January 2005 and August 2008, all the variables, except for the Interbank Certificate of Deposit, proved significant, at a significance level of 10%. For the subsequent period, none of the variables proved significant and the R² was very low, which may merely indicate that investors failed to analyze the main macroeconomic variables for mutual fund allocations or redemptions and simply considered other aspects, such as manager performance.

  19. Main considerations for modelling a station blackout scenario with trace

    Energy Technology Data Exchange (ETDEWEB)

    Querol, Andrea; Turégano, Jara; Lorduy, María; Gallardo, Sergio; Verdú, Gumersindo, E-mail: anquevi@upv.es, E-mail: jaturna@upv.es, E-mail: maloral@upv.es, E-mail: sergalbe@iqn.upv.es, E-mail: gverdu@iqn.upv.es [Instituto Universitario de Seguridad Industrial, Radiofísica y Medioambiental (ISIRYM), Universitat Politècnica de València (Spain)

    2017-07-01

    In the nuclear safety field, the thermal hydraulic phenomena that take place during an accident in a nuclear power plant is of special importance. One of the most studied accidents is the Station BlackOut (SBO). The aim of the present work is the analysis of the PKL integral test facility nodalization using the thermal-hydraulic code TRACE5 to reproduce a SBO accidental scenario. The PKL facility reproduces the main components of the primary and secondary systems of its reference nuclear power plant (Philippsburg II). The results obtained with different nodalization have been compared: 3D vessel vs 1D vessel, Steam Generator (SG) modelling using PIPE or TEE components and pressurizer modelling with PIPE or PRIZER components. Both vessel nodalization (1D vessel and 3D vessel) reproduce the physical phenomena of the experiment. However, there are significant discrepancies between them. The appropriate modelling of the SG is also relevant in the results. Regarding the other nodalization (PIPE or TEE components for SG and PIPE or PRIZER components for pressurizer), do not produce relevant differences in the results. (author)

  20. Organizational Communication: An Analysis of the Main Perspectives, Main Concepts and Future Directions of the Field

    Directory of Open Access Journals (Sweden)

    Yusuf Yüksel

    2013-09-01

    Full Text Available There is a scholarly debate since the 1980s regarding the content, theory, methodology and applications that define the scope of organizational communication and separate it from other related disciplines. This debate is critical in the sense that it enables to identity organizational communication in a rich manner and helps us define the scope of the field and its unique characteristics. Based on this main assumption, this study addressed the major theoretical/methodological dimensions of the field (functional, interpretive, and critical, conceptualization of the most critical concepts (organization, communication, culture, voice/control in these dimensions, and current gaps and future directions of the field. This study revealed that the field of organizational communication has made great improvements since the field emerged in the last three decade with its own content, methodology, and applications and generated an adequate body of research within these different perspectives. It is shown that representation of the field by different perspectives provides richness to the field compared with the time when organizational communication was solely dominated by functional, positivist research. Key words: Functional/interpretive/critical perspectives, communication, organization, culture, control, effectiveness. Örgütsel İletişim: Alanın Ana Yaklaşımları, Ana Kavramları ve Gelecek Yönelimlerinin AnaliziÖzÖrgütsel iletişim alanının kapsamı ve bu alanı ilgili displinlerden ayıracak içerik, teori, yöntem, ve uygulamalar üzerine akademik tartışmalar 1980’li yıllardan beri devam etmektedir. Bu tartışmalar, örgütsel iletişim alanının derinlemesine anlaşılması, sınırlarının belirlenmesi ve diğer disiplinlerden ayrılan özelliklerinin anlaşılması noktasında hayati öneme sahiptir. Bu temel varsayımdan hareketle, bu çalışma alandaki temel teorik/yöntemsel yaklaşımları (işlevsel, yorumlayıcı, ele

  1. 14 CFR 119.47 - Maintaining a principal base of operations, main operations base, and main maintenance base...

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 3 2010-01-01 2010-01-01 false Maintaining a principal base of operations, main operations base, and main maintenance base; change of address. 119.47 Section 119.47 Aeronautics... Under Part 121 or Part 135 of This Chapter § 119.47 Maintaining a principal base of operations, main...

  2. Containment Response Analysis for Equipment Qualification of Kori Nuclear Power Plant Unit 3 and 4

    Energy Technology Data Exchange (ETDEWEB)

    Byun, Choong Sup; Song, Dong Soo; Hwang, Yong Jun [Korea Electric Power Research Institute, Taejon (Korea, Republic of); Seo, Kwi Hyun; Song, Wan Jung [ENERGEO Inc., Sungnam (Korea, Republic of)

    2006-07-01

    Equipment that is used to perform a necessary safety function must be capable of maintaining functional operability under all service condition postulated to occur during the installed life for the time it is required. The pressure and temperature analyses for loss of coolant accident and main steam line break accident provide the bounding test envelope inside containment for the operability evaluation of safety equipment in harsh environmental. This paper describes the results of the containment pressure and temperature analysis for the equipment qualification (EQ) envelopes of Kori unit 3 and 4.

  3. Method to detect steam generator tube leakage

    International Nuclear Information System (INIS)

    Watabe, Kiyomi

    1994-01-01

    It is important for plant operation to detect minor leakages from the steam generator tube at an early stage, thus, leakage detection has been performed using a condenser air ejector gas monitor and a steam generator blow down monitor, etc. In this study highly-sensitive main steam line monitors have been developed in order to identify leakages in the steam generator more quickly and accurately. The performance of the monitors was verified and the demonstration test at the actual plant was conducted for their intended application to the plants. (author)

  4. Test Architecture, Test Retrofit

    Science.gov (United States)

    Fulcher, Glenn; Davidson, Fred

    2009-01-01

    Just like buildings, tests are designed and built for specific purposes, people, and uses. However, both buildings and tests grow and change over time as the needs of their users change. Sometimes, they are also both used for purposes other than those intended in the original designs. This paper explores architecture as a metaphor for language…

  5. Castalia: A European Mission to a Main Belt Comet

    Science.gov (United States)

    Snodgrass, Colin; Castalia mission science Team

    2013-10-01

    Main Belt Comets (MBCs) are a newly identified population, with stable asteroid-like orbits in the outer main belt and a comet-like appearance. It is believed that they survived the age of the solar system in a dormant state and that their activity occurred only recently. Water ice is the only volatile expected to survive, and only when buried under an insulating surface. Excavation by impact could bring the water ice (closer) to the surface and trigger the start of MBC activity. The specific science goals of the Castalia mission are: 1. Characterize a new Solar System family, the MBCs, by in-situ investigation 2. Understand the physics of activity on MBCs 3. Directly detect water in the asteroid belt 4. Test if MBCs are a viable source for Earth’s water 5. Use MBCs as tracers of planetary system formation and evolution These goals can be achieved by a spacecraft designed to rendezvous with and orbit an MBC for some months, arriving before the active period begins for mapping before directly sampling the gas and dust released during the active phase. Given the low level of activity of MBCs, and the expectation that their activity comes from only a localized patch on the surface, the orbiting spacecraft will have to be able to maintain a very close orbit over extended periods - the Castalia plan envisages an orbiter capable of ‘hovering’ autonomously at distances of only a few km from the surface of the MBC. The straw-man instrument payload is made up of: - Visible and near-infrared spectral imager - Thermal infrared imager - Radio science - Dust impact detector - Dust composition analyzer - Neutral/ion mass spectrometer - Magnetometer - Plasma package In addition to this, the option of a surface science package is being considered. At the moment MBC 133P/Elst-Pizarro is the best-known target for such a mission. A design study for the Castalia mission has been carried out in partnership between the science team, DLR and OHB Systems. This study looked at

  6. Prenatal Genetic Diagnostic Tests

    Science.gov (United States)

    ... are available for many inherited disorders. The main disadvantage is that diagnostic testing carries a very small ... chromosomes, arranged in order of size. Microarray: A technology that examines all of a person’s genes to ...

  7. IPv6 Testing

    National Research Council Canada - National Science Library

    Landis, Christopher B

    2006-01-01

    .... The DoD is also forming working relationships to conduct testing and share information. IPv6 was developed to resolve the issues of IPv4, mainly the limited amount of addresses and lack of security...

  8. Lactic acid test

    Science.gov (United States)

    ... this page: //medlineplus.gov/ency/article/003507.htm Lactic acid test To use the sharing features on this page, please enable JavaScript. Lactic acid is mainly produced in muscle cells and red ...

  9. Airborne Test Bed Facility

    Data.gov (United States)

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

  10. Ares I Integrated Test Approach

    Science.gov (United States)

    Taylor, Jim

    2008-01-01

    This slide presentation reviews the testing approach that NASA is developing for the Ares I launch vehicle. NASA is planning a complete series of development, qualification and verification tests. These include: (1) Upper stage engine sea-level and altitude testing (2) First stage development and qualification motors (3) Upper stage structural and thermal development and qualification test articles (4) Main Propulsion Test Article (MPTA) (5) Upper stage green run testing (6) Integrated Vehicle Ground Vibration Testing (IVGVT) and (7) Aerodynamic characterization testing.

  11. Operating experience with an on-line vibration control system for PWR main coolant pumps

    International Nuclear Information System (INIS)

    Runkel, J.; Stegemann, D.; Vortriede, A.

    1996-01-01

    The main circulation pumps are key components of nuclear power plants with pressurized water reactors, because the availability of the main circulation pumps has a direct influence on the availability and electrical output of the entire plant. The on-line automatic vibration control system ASMAS was developed for early failure detection during the normal operation of the main circulation pumps in order to avoid unexpected outages and to establish the possibility of preventive maintenance of the pumps. This system is permanently and successfully operating in three German 1300 MW el NPP's with PWR and has been successfully tested in a 350 MW el NPP with a PWR. (orig.)

  12. Operating experience with an on-line vibration control system for PWR main coolant pumps

    International Nuclear Information System (INIS)

    Runkel, J.; Stegemann, D.; Vortriede, A.

    1998-01-01

    The main circulation pumps are key components of nuclear power plants with pressurized water reactors (PWRs), because the availability of the main circulation pumps has a direct influence on the availability and electrical output of the entire plant. The on-line automatic vibration control system ASMAS was developed for early failure detection during the normal operation of the main circulation pumps in order to avoid unexpected outages and to establish the possibility of preventive maintenance of the pumps. This system is permanently and successfully operating in three German 1300 MW e1 NPP's with PWR and has been successfully tested in a 350 MW e1 NPP with a PWR. (orig.)

  13. To test or not to test

    DEFF Research Database (Denmark)

    Rochon, Justine; Gondan, Matthias; Kieser, Meinhard

    2012-01-01

    Background: Student's two-sample t test is generally used for comparing the means of two independent samples, for example, two treatment arms. Under the null hypothesis, the t test assumes that the two samples arise from the same normally distributed population with unknown variance. Adequate...... control of the Type I error requires that the normality assumption holds, which is often examined by means of a preliminary Shapiro-Wilk test. The following two-stage procedure is widely accepted: If the preliminary test for normality is not significant, the t test is used; if the preliminary test rejects...... the null hypothesis of normality, a nonparametric test is applied in the main analysis. Methods: Equally sized samples were drawn from exponential, uniform, and normal distributions. The two-sample t test was conducted if either both samples (Strategy I) or the collapsed set of residuals from both samples...

  14. Functional conjugated pyridines via main-group element tuning.

    Science.gov (United States)

    Stolar, Monika; Baumgartner, Thomas

    2018-03-29

    Pyridine-based materials have seen widespread attention for the development of n-type organic materials. In recent years, the incorporation of main-group elements has also explored significant advantages for the development and tunability of organic conjugated materials. The unique chemical and electronic structure of main-group elements has led to several enhancements in conventional organic materials. This Feature article highlights recent main-group based pyridine materials by discussing property enhancements and application in organic electronics.

  15. Main - RPSD | LSDB Archive [Life Science Database Archive metadata

    Lifescience Database Archive (English)

    Full Text Available switchLanguage; BLAST Search Image Search Home About Archive Update History Data ... as rice. Data file File name: rpsd_main_sjis.zip File URL: ftp://ftp.biosciencedbc.jp/archive/rpsd/LATEST/r.../ftp.biosciencedbc.jp/archive/rpsd/LATEST/rpsd_main_utf8.zip File size: 120 KB Simple search URL http://togo...nse Update History of This Database Site Policy | Contact Us Main - RPSD | LSDB Archive ...

  16. Shock resistance testing

    International Nuclear Information System (INIS)

    Pouard, M.

    1984-03-01

    In the framework of mechanical tests and to answer the different requests for tests, the T.C.R (Transport Conditionnement et Retraitement) laboratory got test facilities. These installations allow to carry out tests of resistance to shocks, mainly at the safety level of components of nuclear power plants, mockups of transport casks for fuel elements and transport containers for radioactive materials. They include a tower and a catapult. This paper give a decription of the facilities and explain their operation way [fr

  17. Training of professional interpreters in Cuba: Its main historic backgrounds

    Directory of Open Access Journals (Sweden)

    Diana Oliveros-Domínguez

    2017-06-01

    Full Text Available The formation of interpreters in Cuba has as its main goal to train a professional who is able to mediate among Spanish speakers and not-Spanish speakers in our historic situation. This research article aims to analyze the main backgrounds of interpretation teaching at Universidad de Oriente, taking into account its dynamics and the didactic treatment of the cognitive factors involved in the interpretation process. This enables to deepen into the main characteristics of the process, through a study of short-term memory training and the didactic devices used for its improvement; and establish three main stages in the evolution.

  18. DIARRHEA IN CHILDREN: MAIN CAUSES AND WAYS OF TREATMENT

    Directory of Open Access Journals (Sweden)

    S.V. Bel’mer

    2010-01-01

    Full Text Available The article discusses main questions of diagnostics of diarrhea in children. Main cause of acute diarrhea is infection, mainly viral (rotavirus, etc.. Chronic diarrhea frequently has non-infectious origin. The need of multi-aspect diagnostics of diarrhea cause in children is related to the significance of treatment of main disease. Besides, treatment of chronic and acute diarrhea include major component: adsorbents based on smectite. In total treatment of diarrhea has to be complex with the use of dietotherapy and medications: mucocytoprotectors, regulators of motoric, pre- and probiotics.Key words: children, diarrhea, treatment.(Voprosy sovremennoi pediatrii — Current Pediatrics. 2010;9(6:135-138

  19. Analysis of main artifacts in scanning probe microscopy (1)

    International Nuclear Information System (INIS)

    Alekperov, S.D.; Alekperov, S.D.

    2012-01-01

    The analysis of experiment carrying methodology in the scanning probe microscopy (SPM) region is carried out, the main parameters influencing on image quality are revealed. In order to reveal the artifact reason the main components of SPM signal which are divided on 5 groups : the useful signal; noises connected with external influences and temperature drift; distortions connected with piezoceramics and piezo-scanner non-ideality; probe geometry influence; apparatus noises are considered. The main methods of removal and minimization of the given artifacts are considered. The second and third groups of main components of SPM signal are considered in the articles first part

  20. Depression in Main Caregivers of Dementia Patients: Prevalence and Predictors

    Directory of Open Access Journals (Sweden)

    Victoria Omranifard

    2018-01-01

    Full Text Available Background: The most common neurodegenerative disease is dementia. Family of dementia patients says that their lives have been changed extensively after happening of dementia to their patients. One of the problems of family and caregivers is depression of the caregiver. In this study, we aimed to find the prevalence of depression and factors can affect depression in the dementia caregivers. Materials and Methods: This study was cross-sectional study with convenient sampling method. Our society was 96 main caregivers of dementia patients in the year 2015 in Iran. We had two questionnaires, a demographic and Beck Depression Inventory (BDI. BDI Cronbach's alpha is 0.86 for psychiatric patients and 0.81 for nonpsychiatric persons, and Beck's scores are between 0 and 64. We used SPSS version 22 for statistical analysis. Results: According to Beck depression test, 69.8% (n = 67 out of 96 of all caregivers had scores in the range of depression. In bivariate analysis, we found higher dementia severity and lower support of other family members from the caregiver can predict higher depression in the caregiver. As well, in regression analysis using GLM model, we found higher age and lower educational level of the caregiver can predict higher depression in the caregiver. Moreover, regression analysis approved findings about severity and support of other family members in bivariate analysis. Conclusion: High-level depression is found in caregivers of dementia patients. It needs special attention from healthcare managers, clinicians and all of health-care personnel who deals with dementia patients and their caregivers.