Critical heat flux of water in vertical tubes with an upper plenum and a closed bottom

International Nuclear Information System (INIS)

Kim, Hong Chae; Baek, Won Pil; Chang, Soon Heung

2000-01-01

An experimental study is conducted for vertical round tubes with an upper plenum and a closed bottom to investigate CHF behavior and CHF onset location under the counter-current condition. The measured CHF values are well predicted by general Wallis type flooding correlations. A 1-D steady state analytical flooding model for thermosyphon by El-Genk and Saber was assessed with the data and the liquid film thickness at the liquid entrance was calculated. The CHF onset position becomes different with L/D and D, and liquid entrance geometry affects only CHF values not CHF onset positions

Upper plenum break LOCA investigation in the ISB-VVER and PSB-VVER facilities

Energy Technology Data Exchange (ETDEWEB)

Blinkov, V. N.; Melikhov, O. I.; Lipatov, I. A.; Nikonov, S. M.; Dremin, G. I.; Galchanskaya, S. A.; Gashenko, M. P.; Rovnov, A. A.; Kapustin, A. V.; Elkin, I. V. [EREC, Moscow (Russian Federation)

2003-07-01

The capability to define the actual NPP transient/accident scenario depends to a great extent on facilities' scaling and reliability of the system thermalhydraulic codes which, in turn, are assessed against the experimental data taken in the same facilities. At the present time, it is received fact that the rigorous modeling of the cumulative set of all thermalhydraulic processes in the plant primary and secondary sides during accident is unfeasible. Therefore, the extrapolation of the facilities loops behavior to the actual systems constitutes a fundamental problem in this area. In the paper, some aspects for the problem have been discussed in the course of comparative analysis of the data derived from the 11 % upper plenum break LOCA tests performed in the PSB-VVER and ISB-VVER integral test facilities under the close scenarios. Both facilities, PSB-VVER and ISB-VVER, are modeled the same VVER-1000 reactor in different scales. The thermalhydraulic behavior of the primary systems in both facilities has been discussed in the paper, and shown to be similar. Also, the attention has been focused upon the discrepancies in the significant variables trends. The discrepancies are shown to be caused by influence of peculiarities of the facilities hardware and due to the scale factor. The scaling study is an important aspect of the thermalhydraulic codes verification procedure. Being qualified against the experimentally simulated accident sequence in two test facilities of different scales, the thermalhydraulic codes will be capable of evaluation of the prototype behavior to greater accuracy.

Experimental and CFD Studies of Coolant Flow Mixing within Scaled Models of the Upper and Lower Plenums of NGNP Gas-Cooled Reactors

Energy Technology Data Exchange (ETDEWEB)

Hassan, Yassin [Texas A & M Univ., College Station, TX (United States); Anand, Nk [Texas A & M Univ., College Station, TX (United States)

2016-03-30

A 1/16th scaled VHTR experimental model was constructed and the preliminary test was performed in this study. To produce benchmark data for CFD validation in the future, the facility was first run at partial operation with five pipes being heated. PIV was performed to extract the vector velocity field for three adjacent naturally convective jets at statistically steady state. A small recirculation zone was found between the pipes, and the jets entered the merging zone at 3 cm from the pipe outlet but diverged as the flow approached the top of the test geometry. Turbulence analysis shows the turbulence intensity peaked at 41-45% as the jets mixed. A sensitivity analysis confirmed that 1000 frames were sufficient to measure statistically steady state. The results were then validated by extracting the flow rate from the PIV jet velocity profile, and comparing it with an analytic flow rate and ultrasonic flowmeter; all flow rates lie within the uncertainty of the other two methods for Tests 1 and 2. This test facility can be used for further analysis of naturally convective mixing, and eventually produce benchmark data for CFD validation for the VHTR during a PCC or DCC accident scenario. Next, a PTV study of 3000 images (1500 image pairs) were used to quantify the velocity field in the upper plenum. A sensitivity analysis confirmed that 1500 frames were sufficient to precisely estimate the flow. Subsequently, three (3, 9, and 15 cm) Y-lines from the pipe output were extracted to consider the output differences between 50 to 1500 frames. The average velocity field and standard deviation error that accrued in the three different tests were calculated to assess repeatability. The error was varied, from 1 to 14%, depending on Y-elevation. The error decreased as the flow moved farther from the output pipe. In addition, turbulent intensity was calculated and found to be high near the output. Reynolds stresses and turbulent intensity were used to validate the data by

Experimental and CFD Studies of Coolant Flow Mixing within Scaled Models of the Upper and Lower Plenums of NGNP Gas-Cooled Reactors

International Nuclear Information System (INIS)

Hassan, Yassin; Anand, Nk

2016-01-01

A 1/16th scaled VHTR experimental model was constructed and the preliminary test was performed in this study. To produce benchmark data for CFD validation in the future, the facility was first run at partial operation with five pipes being heated. PIV was performed to extract the vector velocity field for three adjacent naturally convective jets at statistically steady state. A small recirculation zone was found between the pipes, and the jets entered the merging zone at 3 cm from the pipe outlet but diverged as the flow approached the top of the test geometry. Turbulence analysis shows the turbulence intensity peaked at 41-45% as the jets mixed. A sensitivity analysis confirmed that 1000 frames were sufficient to measure statistically steady state. The results were then validated by extracting the flow rate from the PIV jet velocity profile, and comparing it with an analytic flow rate and ultrasonic flowmeter; all flow rates lie within the uncertainty of the other two methods for Tests 1 and 2. This test facility can be used for further analysis of naturally convective mixing, and eventually produce benchmark data for CFD validation for the VHTR during a PCC or DCC accident scenario. Next, a PTV study of 3000 images (1500 image pairs) were used to quantify the velocity field in the upper plenum. A sensitivity analysis confirmed that 1500 frames were sufficient to precisely estimate the flow. Subsequently, three (3, 9, and 15 cm) Y-lines from the pipe output were extracted to consider the output differences between 50 to 1500 frames. The average velocity field and standard deviation error that accrued in the three different tests were calculated to assess repeatability. The error was varied, from 1 to 14%, depending on Y-elevation. The error decreased as the flow moved farther from the output pipe. In addition, turbulent intensity was calculated and found to be high near the output. Reynolds stresses and turbulent intensity were used to validate the data by

Improved plenum pressure gradient facemaps for PKL reactors

International Nuclear Information System (INIS)

Crowley, D.A.; Hamm, L.L.

1988-05-01

This report documents the development of improved plenum pressure gradient facemaps* for PKL Mark 16--31 and Mark 22 reactor charges. These new maps are based on the 1985 L-area AC flow tests. Use of the L-area data base for estimating C-area plenum pressure gradient maps is inappropriate because the nozzle geometry plays a major role in determining the shape of the plenum pressure profile. These plenum pressure gradient facemaps are used in the emergency cooling system (ECS) and in the flow instability (FI) loss of coolant accident (LOCA) limits calculations. For the ECS LOCA limits calculations, the maps are used as input to the FLOWZONE computer code to determine the average flow within a flowzone during normal operating conditions. For the FI LOCA limits calculations, the maps are used as plenum pressure boundary conditions in the FLOWTRAN computer code to determine the maximum pre-incident assembly flow within a flowzone. These maps will also be used for flowzoning and transient protection limits analyses

Program plan for correction of US instrument degradation or failure in the Upper Plenum Test Facility (UPTF) in the Federal Republic of Germany

International Nuclear Information System (INIS)

Rhee, G.S.; Chen, Y.S.; Shotkin, L.M.

1987-07-01

This report documents, as of September, 1986, the investigation of the failure or degradation of some of the advanced two-phase flow instruments supplied by the United States Nuclear Regulatory Commission (USNRC) to the German Upper Plenum Test Facility (UPTF). These instruments include Tie-Plate Drag Bodies (DBs), Breakthrough Detectors (BTDs), Loop Drag Disc (DD) paddles, Fluid Distribution Grid (FDG) sensors, and Liquid Level Detector (LLD) sensors. The exact causes for these instrument degradations or failures are not known, but several potential causes have been identified. For DBs and BTDs, the primary mechanism for the degradation appears to be a leakage in the Inconel 600 strain gage encapsulation and the subsequent burnout of the strain gage elements. Excessive loads appear to be the cause of the degradation or failure of the drag discs. The degradation cause for most of the FDGs and LLDs may be either steam/water erosion or mechanical abrasion of the sapphire sensor tips. However, some of the FDG tips were found to be cracked also. The corrective actions are being directed towards identification of the primary causes for the instrument degradation or failure and methods of preventing recurrance and toward minimizing the impact on the test program. All possible action items are being reviewed to arrange them in terms of priority and the likelihood of success so that the best results can be obtained under the constraints of a fixed amount of resources and limited time

Application of mesh free lattice Boltzmann method to the analysis of very high temperature reactor lower plenum

Energy Technology Data Exchange (ETDEWEB)

Park, Jong Woon [Dongguk Univ., Gyeongju (Korea, Republic of). Dept. of Energy and Environment

2011-11-15

Inside a helium-cooled very high temperature reactor (VHTR) lower plenum, hot gas jets from upper fuel channels with very high velocities and temperatures and is mixed before flowing out. One of the major concerns is local hot spots in the plenum due to inefficient mixing of the helium exiting from differentially heated fuel channels and it involves complex fluid flow physics. For this situation, mesh-free technique, especially Lattice Boltzmann Method (LBM), is thus of particular interest owing to its merit of no mesh generation. As an attempt to find efficiency of the method in such a problem, 3 dimensional flow field inside a scaled test model of the VHTR lower plenum is computed with commercial XFLOW code. Large eddy simulation (LES) and classical Smagorinsky eddy viscosity (EV) turbulence models are employed to investigate the capability of the LBM in capturing large scale vortex shedding. (orig.)

Analytical model for transient fluid mixing in upper outlet plenum of an LMFBR

International Nuclear Information System (INIS)

Yang, J.W.; Agrawal, A.K.

1976-01-01

A two-zone mixing model based on the lumped-parameter approach was developed for the analysis of transient thermal response in the outlet plenum of an LMFBR. The maximum penetration of core flow is used as the criterion for dividing the sodium region into two mixing zones. The model considers the transient sodium temperature affected by the thermal expansion of sodium, heat transfer with cover gas, heat capacity of different sections of metal and the addition of by-pass flow into the plenum. The results of numerical calculations indicate that effects of flow stratification, chimney height, metal heat capacity and by-pass flow are important for transient sodium temperature calculation. Thermal expansion of sodium and heat transfer with the cover gas do not play any significant role on sodium temperature

Experiment data report for Semiscale Mod-1 Test S-05-4 (alternate ECC injection test)

International Nuclear Information System (INIS)

Collins, B.L.; Feldman, E.M.

1977-03-01

Recorded test data are presented for Test S-05-4 of the Semiscale Mod-1 alternate emergency core coolant injection test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-05-4 was conducted from initial conditions of 2266 psia and 543 0 F to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the cold leg broken loop piping. During the test, cooling water was injected into the cold leg of each loop and into the vessel upper plenum to simulate emergency core coolant injection in a PWR. The upper plenum coolant injection was scaled according to the heat stored in the metal mass of the upper plenum

Determining Bond Sodium Remaining in Plenum Region of Spent Nuclear Driver Fuel

International Nuclear Information System (INIS)

Vaden, D.; Li, S.X.

2008-01-01

The Fuel Conditioning Facility (FCF) at the Idaho National Laboratory (INL) treats spent nuclear fuel using an electro-chemical process that separates the uranium from the fission products, sodium thermal bond, and cladding materials (REF 1). Upon immersion into the ER electrolyte, the sodium used to thermally bond the fuel to the clad jacket chemically reacts with the UCl3 in the electrolyte producing NaCl and uranium metal. The uranium in the spent fuel is separated from the cladding and fission products by taking advantage of the electro-chemical potential differences between uranium and the other fuel components. Assuming all the sodium in the thermal bond is converted to NaCl in the ER, the difference between the cumulative bond sodium mass in the fuel elements and the cumulative sodium mass found in the driver ER electrolyte inventory provides an upper mass limit for the sodium that migrated to the upper gas region, or plenum section, of the fuel element during irradiation in the reactor. The plenums are to be processed as metal waste via melting and metal consolidation operations. However, depending on the amount of sodium in the plenums, additional processing may be required to remove the sodium before metal waste processing

Fundamental validation of simulation method for thermal stratification in upper plenum of fast reactors. Analysis of sodium experiment

International Nuclear Information System (INIS)

Ohno, Shuji; Ohshima, Hiroyuki; Sugahara, Akihiro; Ohki, Hiroshi

2010-01-01

Three-dimensional thermal-hydraulic analyses have been carried out for a sodium experiment in a relatively simple axis-symmetric geometry using a commercial CFD code in order to validate simulating methods for thermal stratification behavior in an upper plenum of sodium-cooled fast reactor. Detailed comparison between simulated results and experimental measurement has demonstrated that the code reproduced fairly well the fundamental thermal stratification behaviors such as vertical temperature gradient and upward movement of a stratification interface when utilizing high-order discretization scheme and appropriate mesh size. Furthermore, the investigation has clarified the influence of RANS type turbulence models on phenomena predictability; i.e. the standard k-ε model, the RNG k-ε model and the Reynolds Stress Model. (author)

Study on mixed convective flow penetration into subassembly from reactor hot plenum in FBRs

Energy Technology Data Exchange (ETDEWEB)

Kobayashi, J.; Ohshima, H.; Kamide, H.; Ieda, Y. [Power Reactor and Nuclear Fuel Development Corporation, Ibaraki (Japan)

1995-09-01

Fundamental experiments using water were carried out in order to reveal the phenomenon of mixed convective flow penetration into subassemblies from a reactor`s upper plenum of fast breeder reactors. This phenomenon appears under a certain natural circulation conditions during the operation of the direct reactor auxiliary cooling system for decay heat removal and might influence the natural circulation head which determines the core flow rate and therefore affects the core coolability. In the experiment, a simplified model which simulates an upper plenum and a subassembly was used and the ultrasonic velocity profile monitor as well as thermocouples were applied for the simultaneous measurement of velocity and temperature distributions in the subassembly. From the measured data, empirical equations related to the penetration flow onset condition and the penetration depth were obtained using relevant parameters which were derived from dimensional analysis.

Hydraulics in the RPV lower-plenum of EPR

International Nuclear Information System (INIS)

Barois, G.; Goreaud, N.; Nicaise, N.

2001-01-01

The in-core instrumentation penetrations of the European Pressurised water Reactor (EPR) have been removed from RPV-bottom to RPV-head, leaving empty the lower plenum of the RPV (Reactor Pressure Vessel). In a lower plenum with no internal structure, huge vortices may appear, with negative consequences, such as high disturbance of the core inlet flow distribution, and high increase of the RPV pressure loss. FRAMATOME ANP developed a specific Flow Distribution Device (FDD), annular shaped, located in the RPV lower plenum below the core support plate, which prevents huge vortices from appearing and guarantees a satisfying flow distribution at core inlet in normal operating conditions. The design of the FDD has been optimised with a numerical approach, using the 3-D CFD-code STAR-CD, previously qualified on scale mockup tests. The model developed represents the EPR RPV from the cold leg to core inlet. Thus, the flow distribution at core inlet, the mixing between loop-flows upstream core inlet and the pressure loss in the lower plenum can be evaluated. The optimised FDD provides satisfying performances for all these relevant functional items. (author)

Design of a new SI engine intake manifold with variable length plenum

International Nuclear Information System (INIS)

Ceviz, M.A.; Akin, M.

2010-01-01

This paper investigates the effects of intake plenum length/volume on the performance characteristics of a spark-ignited engine with electronically controlled fuel injectors. Previous work was carried out mainly on the engine with carburetor producing a mixture desirable for combustion and dispatching the mixture to the intake manifold. The more stringent emission legislations have driven engine development towards concepts based on electronic-controlled fuel injection rather than the use of carburetors. In the engine with multipoint fuel injection system using electronically controlled fuel injectors has an intake manifold in which only the air flows and, the fuel is injected onto the intake valve. Since the intake manifolds transport mainly air, the supercharging effects of the variable length intake plenum will be different from carbureted engine. Engine tests have been carried out with the aim of constituting a base study to design a new variable length intake manifold plenum. Engine performance characteristics such as brake torque, brake power, thermal efficiency and specific fuel consumption were taken into consideration to evaluate the effects of the variation in the length of intake plenum. The results showed that the variation in the plenum length causes an improvement on the engine performance characteristics especially on the fuel consumption at high load and low engine speeds which are put forward the system using for urban roads. According to the test results, plenum length must be extended for low engine speeds and shortened as the engine speed increases. A system taking into account the results of the study was developed to adjust the intake plenum length.

Empirical method to calculate Clinch River Breeder Reactor (CRBR) inlet plenum transient temperatures

International Nuclear Information System (INIS)

Howarth, W.L.

1976-01-01

Sodium flow enters the CRBR inlet plenum via three loops or inlets. An empirical equation was developed to calculate transient temperatures in the CRBR inlet plenum from known loop flows and temperatures. The constants in the empirical equation were derived from 1/4 scale Inlet Plenum Model tests using water as the test fluid. The sodium temperature distribution was simulated by an electrolyte. Step electrolyte transients at 100 percent model flow were used to calculate the equation constants. Step electrolyte runs at 50 percent and 10 percent flow confirmed that the constants were independent of flow. Also, a transient was tested which varied simultaneously flow rate and electrolyte. Agreement of the test results with the empirical equation results was good which verifies the empirical equation

Development of thermal mixing enhancement method for lower plenum of the High Temperature Test Facility

Energy Technology Data Exchange (ETDEWEB)

Gradecka, Malwina Joanna, E-mail: malgrad@gmail.com; Woods, Brian G., E-mail: brian.woods@oregonstate.edu

2016-08-15

Highlights: • Coolant mixing in lower plenum might be insufficient and pose operational issues. • Two mixing methods were developed to lower the coolant temperature variation. • The methods resulted with reduction of the temperature variation by 60% and 71%. - Abstract: The High Temperature Gas-cooled Reactor (HTGR) is one of the most mature Gen IV reactor concepts under development today. The High Temperature Test Facility (HTTF) at Oregon State University is a test facility that supports the R&D needs for HTGRs. This study focuses on the issue of helium mixing after the core section in the HTTF, the results of which are generally applicable in HTGRs. In the HTTF, hot helium jets at different temperatures are supposed to uniformly mix in the lower plenum (LP) chamber. However, the level of mixing is not sufficient to reduce the peak helium temperature before the hot jet impinges the LP structure, which can cause issues with structural materials and operational issues in the heat exchanger downstream. The maximum allowable temperature variation in the outlet duct connected to the lower plenum is defined as 40 K (±20 K from the average temperature), while the CFD simulations of this study indicate that the reference design suffers temperature variations in the duct as high as 100 K. To solve this issue, the installation of mixing-enhancing structures within the outlet duct were proposed and analyzed using CFD modeling. We show that using either an optimized “Kwiat” structure (developed in this study) or a motionless mixer installed in the outlet duct, the temperature variations can be brought dramatically, with acceptable increases in pressure drop. The optimal solution appears to be to install double motionless mixers with long blades in the outlet duct, which brings the temperature variation into the acceptable range (from 100 K down to 18 K), with a resulting pressure drop increase in the HTTF loop of 0.73 kPa (6% of total pressure drop).

Development of thermal mixing enhancement method for lower plenum of the High Temperature Test Facility

International Nuclear Information System (INIS)

Gradecka, Malwina Joanna; Woods, Brian G.

2016-01-01

Highlights: • Coolant mixing in lower plenum might be insufficient and pose operational issues. • Two mixing methods were developed to lower the coolant temperature variation. • The methods resulted with reduction of the temperature variation by 60% and 71%. - Abstract: The High Temperature Gas-cooled Reactor (HTGR) is one of the most mature Gen IV reactor concepts under development today. The High Temperature Test Facility (HTTF) at Oregon State University is a test facility that supports the R&D needs for HTGRs. This study focuses on the issue of helium mixing after the core section in the HTTF, the results of which are generally applicable in HTGRs. In the HTTF, hot helium jets at different temperatures are supposed to uniformly mix in the lower plenum (LP) chamber. However, the level of mixing is not sufficient to reduce the peak helium temperature before the hot jet impinges the LP structure, which can cause issues with structural materials and operational issues in the heat exchanger downstream. The maximum allowable temperature variation in the outlet duct connected to the lower plenum is defined as 40 K (±20 K from the average temperature), while the CFD simulations of this study indicate that the reference design suffers temperature variations in the duct as high as 100 K. To solve this issue, the installation of mixing-enhancing structures within the outlet duct were proposed and analyzed using CFD modeling. We show that using either an optimized “Kwiat” structure (developed in this study) or a motionless mixer installed in the outlet duct, the temperature variations can be brought dramatically, with acceptable increases in pressure drop. The optimal solution appears to be to install double motionless mixers with long blades in the outlet duct, which brings the temperature variation into the acceptable range (from 100 K down to 18 K), with a resulting pressure drop increase in the HTTF loop of 0.73 kPa (6% of total pressure drop).

Comparison of TRAC-PF1/MOD1 to a no-failure UPI test in the Cylindrical Core Test Facility

International Nuclear Information System (INIS)

Cappiello, M.; Spore, J.

1986-01-01

TRAC-PF1/MOD1 is compared to a no-failure upper plenum injection reflood test in the Cylindrical Core Test Facility. The results show that TRAC can accurately predict the asymmetric channeling of fluid from upper plenum into the core and that a multidimensional modeling capability is required to do so. The rod temperature behavior is accurately predicted for both the peak cladding temperature and the quench time in the high- and low-power zones. Excessive downflow of liquid at the tie plate is predicted as a result of the interfacial drag model used in TRAC. 10 figs

Potential for HEPA filter damage from water spray systems in filter plenums

Energy Technology Data Exchange (ETDEWEB)

Bergman, W. [Lawrence Livermore National Lab., CA (United States); Fretthold, J.K. [Rocky Flats Safe Sites of Colorado, Golden, CO (United States); Slawski, J.W. [Department of Energy, Germantown, MD (United States)

1997-08-01

The water spray systems in high efficiency particulate air (HEPA) filter plenums that are used in nearly all Department of Energy (DOE) facilities for protection against fire was designed under the assumption that the HEPA filters would not be damaged by the water sprays. The most likely scenario for filter damage involves filter plugging by the water spray, followed by the fan blowing out the filter medium. A number of controlled laboratory tests that were previously conducted in the late 1980s are reviewed in this paper to provide a technical basis for the potential HEPA filter damage by the water spray system in HEPA filter plenums. In addition to the laboratory tests, the scenario for BEPA filter damage during fires has also occurred in the field. A fire in a four-stage, BEPA filter plenum at Rocky Flats in 1980 caused the first three stages of BEPA filters to blow out of their housing and the fourth stage to severely bow. Details of this recently declassified fire are presented in this paper. Although these previous findings suggest serious potential problems exist with the current water spray system in filter plenums, additional studies are required to confirm unequivocally that DOE`s critical facilities are at risk. 22 refs., 15 figs.

Benchmark Analyses of Sodium Natural Convection in the Upper Plenum of the Monju Reactor Vessel. Final Report of a Coordinated Research Project 2008-2012

International Nuclear Information System (INIS)

2014-11-01

The IAEA supports Member States in the area of advanced fast reactor technology development by providing a major fulcrum for information exchange and collaborative research programmes. The IAEA’s activities in this field are mainly carried out within the framework of the Technical Working Group on Fast Reactors (TWG-FR), which assists in the implementation of corresponding IAEA support, and ensures that all technical activities are in line with expressed needs of Member States. Among this broad range, the IAEA proposes and establishes coordinated research projects (CRPs), aimed at improving Member State capability in fast reactor design and analysis. An important opportunity to perform collaborative research activities was provided by the system startup tests carried out by the Japan Atomic Energy Agency (JAEA) in the prototype loop type sodium cooled fast reactor Monju, in particular a turbine trip test performed in December 1995. As the JAEA opened the experimental dataset to international collaboration in 2008, the IAEA launched the CRP on Benchmark Analyses of Sodium Natural Convection in the Upper Plenum of the Monju Reactor Vessel. The CRP, together with eight institutes from seven States, has contributed to improving capabilities in sodium cooled fast reactors simulation through code verification and validation, with particular emphasis on thermal stratification and natural circulation phenomena

Study of impact of the AP1000{sup Registered-Sign} reactor vessel upper internals design on fuel performance

Energy Technology Data Exchange (ETDEWEB)

Xu Yiban; Conner, Michael; Yuan Kun; Dzodzo, Milorad B.; Karoutas, Zeses; Beltz, Steven A.; Ray, Sumit; Bissett, Teresa A. [Westinghouse Electric Company, Cranberry Township, PA 16066 (United States); Chieng, Ching-Chang, E-mail: cchieng@ess.nthu.edu.tw [National Tsing Hua University, Hsinchu 30043, Taiwan (China); Kao, Min-Tsung; Wu, Chung-Yun [National Tsing Hua University, Hsinchu 30043, Taiwan (China)

2012-11-15

One aspect of the AP1000{sup Registered-Sign} reactor design is the reduction in the number of major components and simplification in manufacturing. One design change relative to current Westinghouse reactors of similar size is the reduction in the number of reactor vessel outlet nozzles/hot legs leaving the upper plenum from three to two. With regard to fuel performance, this design difference creates a different flow field in the AP1000 reactor vessel upper plenum (the region above the core). The flow exiting core and entering the upper plenum must turn 90 Degree-Sign , flow laterally through the upper plenum around support structures, and exit through one of the two outlet nozzles. While the flow in the top of the core is mostly axial, there is some lateral flow component as the core flow reacts to the flow field and pressure distribution in the upper plenum. The pressure distribution in the upper plenum varies laterally depending upon various factors including the proximity to the outlet nozzles. To determine how the lateral flow in the top of the AP1000 core compares to current Westinghouse reactors, a computational fluid dynamics (CFD) model of the flow in the upper portion of the AP1000 reactor vessel including the top region of the core, the upper plenum, the reactor vessel outlet nozzles, and a portion of the hot legs was created. Due to geometric symmetry, the computational domain was reduced to a quarter (from the top view) that includes Vulgar-Fraction-One-Quarter of the top of the core, Vulgar-Fraction-One-Quarter of the upper plenum, and Vulgar-Fraction-One-Half of an outlet nozzle. Results from this model include predicted velocity fields and pressure distributions throughout the model domain. The flow patterns inside and around guide tubes clearly demonstrate the influence of lateral flow due to the presence of the outlet nozzles. From these results, comparisons of AP1000 flow versus current Westinghouse plants were performed. Field performance

ECC delivery to lower plenum under downcomer injection part 2. RELAP5 assessment

International Nuclear Information System (INIS)

Bang, Young Seok; Shin, An Dong; Kim, Hho Jung

2000-01-01

In the present study, the capability of the thermal-hydraulic codes, RELAP5/MOD3.2.2 gamma, in predicting the steam-water interaction and the related ECC delivery to lower plenum under downcomer injection condition during refill phase is evaluated using the experimental data of the UPTF Test 21A. The facility is modeled in detail, and the test condition simulated for code calculations. The calculation result is compared with the applicable measurement data and discussed for the pressure response, ECC bypass behavior, lower plenum delivery, global water mass distribution, and local behavior in downcomer

Experimental Modeling of VHTR Plenum Flows during Normal Operation and Pressurized Conduction Cooldown

Energy Technology Data Exchange (ETDEWEB)

Glenn E McCreery; Keith G Condie

2006-09-01

The Very High Temperature Reactor (VHTR) is the leading candidate for the Next Generation Nuclear Power (NGNP) Project in the U.S. which has the goal of demonstrating the production of emissions free electricity and hydrogen by 2015. The present document addresses experimental modeling of flow and thermal mixing phenomena of importance during normal or reduced power operation and during a loss of forced reactor cooling (pressurized conduction cooldown) scenario. The objectives of the experiments are, 1), provide benchmark data for assessment and improvement of codes proposed for NGNP designs and safety studies, and, 2), obtain a better understanding of related phenomena, behavior and needs. Physical models of VHTR vessel upper and lower plenums which use various working fluids to scale phenomena of interest are described. The models may be used to both simulate natural convection conditions during pressurized conduction cooldown and turbulent lower plenum flow during normal or reduced power operation.

Plenum separator system for pool-type nuclear reactors

International Nuclear Information System (INIS)

Sharbaugh, J.E.

1983-01-01

This invention provides a plenum separator system for pool-type nuclear reactors which substantially lessens undesirable thermal effects on major components. A primary feature of the invention is the addition of one or more intermediate plena, containing substantially stagnant and stratified coolant, which separate the hot and cold plena and particularly the hot plena from critical reactor components. This plenum separator system also includes a plurality of components which together form a dual pass flow path annular region spaced from the reactor vessel wall by an annular gas space. The bypass flow through the flow path is relatively small and is drawn from the main coolant pumps and discharged to an intermediate plenum

Experiment data report for Semiscale Mod-1 Test S-05-3 (alternate ECC injection test)

International Nuclear Information System (INIS)

Feldman, E.M.; Patton, M.L. Jr.; Sackett, K.E.

1977-03-01

Recorded test data are presented for Test S-05-3 of the Semiscale Mod-1 alternate ECC injection test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-05-3 was conducted from initial conditions of 2263 psia and 545 0 F to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the cold leg broken loop piping. During the test, cooling water was injected into the cold leg sides of the intact and broken loops and into the vessel upper plenum to simulate emergency core coolant injection in a PWR. For Test S-05-3, specifically the effects of upper plenum coolant injection on core thermal and system response were being investigated

Experiment on performance of upper head injection system with ROSA-II

International Nuclear Information System (INIS)

1978-05-01

Of the total 10 ROSA-II/UHI performance tests, 6 were reported previously. The rest are presented and discussion is made on the effects of heat generation in the core and UHI injection and repeatability of experiments. In addition, the following are described: (1) Pressure spikes observed in the upper head after sudden stoppage of UHI injection, and (2) discharge flow oscillation possibly due to UHI water injection into the upper plenum. (auth.)

2D/3D program. Upper plenum test facility - UPTF. Test No. 1

International Nuclear Information System (INIS)

1987-01-01

Test No.1 was a quasi-steady state, separate effect test involving the UPTF-System with blocked break valves and blocked pump simulators. Initially the test vessel, the cold and hot leg nozzels as well as the pump seals were completely filled witht hot water in this test. This test was designed to investigate the fluid-fluid mixing phenomena and the development of the fluid and wall temperature fields in the cold leg and downcomer region of a PWR. The experiment was performed by injecting a cold water stream into one cold leg of UPTF while the system was initially filled with stagnant hot water. (orig.)

Optimization of inlet plenum of A PBMR using surrogate modeling

International Nuclear Information System (INIS)

Lee, Sang-Moon; Kim, Kwang-Yong

2009-01-01

The purpose of present work is to optimize the design of inlet plenum of PBMR type gas cooled nuclear reactor numerically using a combining of three-dimensional Reynolds-averaged Navier-Stokes (RANS) analysis and surrogate modeling technique. Shear stress transport (SST) turbulence model is used as a turbulence closure. Three geometric design variables are selected, namely, rising channel diameter to plenum height ratio, aspect ratio of the plenum cross section, and inlet port angle. The objective function is defined as a linear combination of uniformity of three-dimensional flow distribution term and pressure drop in the inlet plenum and rising channels of PBMR term with a weighting factor. Twenty design points are selected using Latin-hypercube method of design of experiment and objective function values are obtained at each design point using RANS solver. (author)

Modeling study of deposition locations in the 291-Z plenum

International Nuclear Information System (INIS)

Mahoney, L.A.; Glissmeyer, J.A.

1994-06-01

The TEMPEST (Trent and Eyler 1991) and PART5 computer codes were used to predict the probable locations of particle deposition in the suction-side plenum of the 291-Z building in the 200 Area of the Hanford Site, the exhaust fan building for the 234-5Z, 236-Z, and 232-Z buildings in the 200 Area of the Hanford Site. The Tempest code provided velocity fields for the airflow through the plenum. These velocity fields were then used with TEMPEST to provide modeling of near-floor particle concentrations without particle sticking (100% resuspension). The same velocity fields were also used with PART5 to provide modeling of particle deposition with sticking (0% resuspension). Some of the parameters whose importance was tested were particle size, point of injection and exhaust fan configuration

Analysis of transient thermal response in the outlet plenum of an LMFBR

International Nuclear Information System (INIS)

Yang, J.W.

1976-05-01

A two-zone mixing model based on the lumped-parameter approach was developed for the analysis of transient thermal response in the upper outlet plenum of an LMFBR. The one-dimensional turbulent jet flow equations were solved to determine the maximum penetration of the core flow. The maximum penetration is used as the criterion for dividing the sodium region into two mixing zones. The lumped-parameter model considers the transient sodium temperature affected by the thermal expansion of sodium, heat transfer with cover gas, heat capacity of different sections of metal and the addition of bypass flow into the plenum. Numerical calculations were performed for two cases. The first case corresponds to a normal scram followed by flow coast-down. The second case represents the double-ended pipe rupture at the inlet of cold leg followed by reactor scram. The results indicate that effects of flow stratification, chimney height, metal heat capacity and bypass flow are important for transient sodium temperature calculation. Thermal expansion of sodium and heat transfer with the cover gas does not play any significant role on sodium temperature. This two-zone mixing model will be a part of the thermohydraulic transient code SSC

Analysis research on mixing characteristics of lower plenum of Qinshan phase Ⅱ NPP by CFD method

International Nuclear Information System (INIS)

Mao Huihui; He Peifeng; Lu Chuan; Zhang Hongliang

2015-01-01

The flowing and mixing characteristics of the lower plenum of Qinshan Phase n NPP were analyzed by CFD method. The calculation results were compared with the results of the reactor hydraulic simulation test. On core inlet mass flow distributions, both upwind and high resolution advection schemes show good agreements with test results. While on lower plenum mixing characteristics, the calculation results from either upwind or high resolution advection schemes show relatively large differences to the test data. Relatively, upwind advection schemes predict better anticipations on maximum and minimum mixing factors. Furthermore, whether or not considering helix flow by main pump is the most possible key factor that leads to difference between CFD calculation and test results. (authors)

ROSA-II test data report, 13

International Nuclear Information System (INIS)

1978-07-01

Results of the ROSA-II test simulating a loss-of-coolant accident (LOCA) in a PWR are presented, including test conditions and interpretations of phenomena observed in test runs 502, 505, 506 and 507. Development tests were performed to find a more effective ECCS injection method than the existing one based on cold leg injection. A combined injection of hot water into upper plenum in early stage of blowdown and subsequent cold water into lower plenum is the most effective method for a cold leg break. A hot leg injection of a low pressure injection system is effective for direct core cooling and early reflooding. The generalization for actual reactors will require analyses with a reliable code. (auth.)

Evaluation report on CCTF core-II reflood test C2 - 18 (Run 78)

International Nuclear Information System (INIS)

Iguchi, Tadashi; Akimoto, Hajime; Okubo, Tsutomu; Murao, Yoshio; Sugimoto, Jun; Hojo, Tsuneyuki.

1987-03-01

This report presents the result of the upper plenum injection (UPI) test C2 - 18 (Run 78), which was conducted on November 13, 1984 with the Cylindrical Core Test Facility (CCTF) at Japan Atomic Energy Research Institute (JAERI). The CCTF is a 1/21.4 scale model of a 1,100 MWe PWR with four loop active components to provide information on the system and core thermo-hydrodynamics during reflood phase. The objectives of the test are to investigate the refill behavior with UPI condition and to investigate the reflood behavior with UPI Best-Estimate (BE) condition. The test was performed to simulate refill/reflood behavior with UPI and BE conditions (However, the LPCI flow rate was determined based on single failure of LPCI pumps.). The result of the test showed the followings. (1) Little special phenomena were recognized under UPI and BE conditions in comparison with those under UPI and Evaluation-Model (EM) conditions, although certain special phenoma (i.e. significant fluid oscillation) were recognized under Cold-Leg-Injection (CLI) and BE conditions in comparison with those under CLI and EM conditions. (2) Water inventory in lower plenum increased smoothly due to water injected into both upper plenum and cold leg during refill phase, similarly to that in refill-simulation test with CLI condition. Small difference in refill behavior with UPI condition is the existing of steam condensation in upper plenum, resulting in lower steam binding and higher core cooling during early reflood phase. This indicates the conservatism of UPI against CLI during early reflood phase. (3) The good core-cooling capability was confirmed under UPI and BE conditions. (author)

CATHARE2 analysis on the loss of residual heat removal system during mid-loop operation : pressurizer and SGI outlet plenum manways open

International Nuclear Information System (INIS)

Chung, Young Jong; Chang, Won Pyo.

1997-06-01

The present study is to analyze the BETHSY test 6.9c using CATHARE2 v1.3u. BETHSY test 6.9c simulates plant conditions following loss of residual heat removal system under mid-loop operation. The configuration is that the pressurizer and steam generator outlet plenum manways are opened as vent paths in order to protect the system from overpressurization by removing the steam generated in the core. Most of the important physical phenomena are observed in the experiment have been predicted reasonably by the CATHARE2 code. Since the differential pressure between the pressurizer and the surge line is overestimated, the peak pressure in the upper plenum is predicted higher than the experimental value by 11 kPa and occurrence is delayed by 210s. Also earlier core uncovery is predicted, mainly due to overprediction of the manway flows. The analysis results are demonstrated that opening of the pressurizer and the steam generator outlet plenum manways is effective to prevent the core uncovery by only gravity feed injection. Although some disagreements found in detailed phenomena, the prediction of the overall system behavior by the code does not deviate from the experimental results unacceptably. The core bypass flowrate is found to be very sensitive to mass distribution in the core and the system behaviors are strongly affected by phase separation modeling under low pressure and particularly stratified flow condition. the main purpose of the present study is to understand physical phenomena under the accident and to assess the capability of CATHARE2 prediction for enhancement of reliability in actual plant analyses. (author). 11 refs., 3 tabs., 41 figs

Analysis of the VVER Standard Problem INSC-PSBV1 '11% Coolant Leak from Upper Plenum' with RELAP5/MOD3.2

International Nuclear Information System (INIS)

Melikhov, O.; Melikhov, V.; Parfenov, Yu.; Gavritenkova, O.; Lipatov, I.; Elkin, I.; Bayless, P.

2004-01-01

Analyses of a loss-of-coolant experiment carried out at the PSB-VVER test facility with the RELAP5/MOD3.2 code have been performed independently by analysts at the Electrogorsk Research and Engineering Center (EREC) and the Idaho National Engineering and Environmental Laboratory (INEEL). The PSB-VVER facility is a full-height scale model of a VVER 1000 reactor that is approximately 1/300 scale in volume and power. VVER Standard Problem INSC-PSBV1 represents an 11% leak from the upper plenum of the PSB-VVER facility, simulating the rupture of one of the accumulator injection lines. The safety-significant thermalhydraulic phenomena occurring in VVER type reactors addressed by this experiment were identified in the test validation matrix. Most of the phenomena of the validation matrix were reasonably simulated by RELAP5/MOD3.2 in both calculations. The major differences between the test and the calculations were the timing of the core heatup, and the thermal response to the accumulator injection cycles in both calculations. The INEEL calculation had a more extensive axial heatup, with most of the core experiencing small heat-ups. The accumulator injection was more effective in quenching the core in the test than in the INEEL calculation. This difference is attributed to the liquid distribution in the core, rather than to the heat transfer models in the code. The code calculation had a more uniform axial distribution of the liquid in the core, and the accumulator injection did not have much impact on the core liquid inventory. In the EREC calculation, only one heatup of the cladding temperature was observed for upper and middle section of the fuel rods before the final heatup. The small heat-ups were not reproduced in EREC calculation. The difference could be attributed to differences in liquid distribution, namely the core region in the EREC calculation contains more liquid over most of the transient than in the experiment. The distribution of liquid in the core in

Coolant mixing in the LMFBR outlet plenum

International Nuclear Information System (INIS)

Chen, Y.B.; Golay, M.W.

1977-06-01

Small scale experiments involving water flows are used to provide mean flow and turbulence field data for LMFBR outlet plenum flows. Measurements are performed at Reynolds Number (Re) values of 33000 and 70000 in a 1/15-scale FFTF geometry and at Re = 35000 in a 3/80-scale CRBR geometry. The experimental behavior is predicted using two different turbulence model computer programs, TEACH-T and VARR-II. It is found that the qualitative nature of the flow field within the plenum depends strongly upon the distribution of the mean inlet velocity field, upon the degree of inlet turbulence, and upon the turbulence momentum exchange model used in the calculations. It is found in the FFTF geometry that the TEACH-T predictions are better than that of VARR-II, and in the CRBR geometry neither code provides a good prediction of the observed behavior. From the sensitivity analysis, it is found that the production and dissipation of turbulence are the dominant terms in the transport equations for turbulent kinetic energy and turbulent energy dissipation rate, and the diffusion terms are relatively small. From the same study a new set of empirical constants for the turbulence model is evolved for the prediction of plenum flows

Effects of lower plenum flow structure on core inlet flow of ABWR

International Nuclear Information System (INIS)

Watanabe, Shun; Abe, Yutaka; Kaneko, Akiko; Watanabe, Fumitoshi; Tezuka, Kenichi

2010-01-01

The evaluation of coolant flow structure at a lower plenum of an advanced boiling water reactor (ABWR) in which there are many structures is very important in order to improve generating power. Although the simulation results by CFD (Computational Fluid Dynamics) codes can predict such complicated flow in the lower plenum, it is required to establish the database of flow structure in lower plenum of ABWR experimentally for the benchmark of the CFD codes. In the model of the lower plenum, we measured velocity profiles with LDV and PIV. And differential pressure of constructed model is measured with differential pressure instrument. It was identified that the velocity and differential pressure profiles also showed the tendency to be flat in the core inlet. Moreover, vortexes were observed around side entry orifice by PIV measurement. (author)

Stratification in SNR-300 outlet plenum

International Nuclear Information System (INIS)

Reinders, R.

1983-01-01

In the inner outlet plenum of the SNR-300 under steady state conditions a large toroidal vortex is expected. The main flow passes through the gap between dipplate and shield vessel to the outer annular space. Only 3% of the flow pass the 24 emergency cooling holes, situated in the shield vessel. The sodium leaves the reactor tank through the 3 symmetrically arranged outlet nozzles. For a scram flow rates and temperatures are decreased simultaneously, so it is expected, that stratification occurs in the inner outlet plenum. A measure of stratification effects is the Archimedes Number Ar, which is the relation of buoyancy forces (negative) to kinetic energy. (The Archimedes Number is nearly identical with the Richardson Number). For values Ar>1 stratification can occur. Under the assumption of stratification the code TIRE was developed, which is only applicable for the period of time after some 50 sec after scram. This code serves for long term calculations. As the equations are very simple, it is a very fast code which gives the possibility to calculate transients for some hours real time. This code mainly has to take into account the pressure difference between inner plenum and outlet annulus caused by geodatic pressure. That force is in equilibrium with the pressure drop over the gap and holes in the shield vessel. For more detailed calculations of flow pattern and temperature distribution the code MIX and INKO 2T are applied. MIX was developed and validated at ANL, INKO 2T is a development of INTERATOM. INKO 2T is under validation. Mock up experiments were carried out with water to simulate the transient behavior of the SNR-300 outlet plenum. Calculations obtained by INKO 2T for steady state and the transient are shown for the flow pattern. Results of measurements also prove that stratification begins after about 30 sec. Measurements and detailed calculations show that it is admissible to use the code TIRE for the long term calculations. Calculations for a scram

TMI-2 reactor vessel plenum final lift

International Nuclear Information System (INIS)

Wilson, D.C.

1986-01-01

Removal of the plenum assembly from the TMI-2 reactor vessel was necessary to gain access to the core region for defueling. The plenum was lifted from the reactor vessel by the polar crane using three specially designed pendant assemblies. It was then transferred in air to the flooded deep end of the refueling canal and lowered onto a storage stand where it will remain throughout the defueling effort. The lift and transfer were successfully accomplished on May 15, 1985 in just under three hours by a lift team located in a shielded area within the reactor building. The success of the program is attributed to extensive mockup and training activities plus thorough preparations to address potential problems. 54 refs

CFD predictions of standby liquid control system mixing in lower plenum of a BWR

International Nuclear Information System (INIS)

Boyd, Christopher; Skarda, Raymond

2014-01-01

Highlights: • Computational fluid dynamics analysis of BWR lower plenum. • Mixing and stratification of the standby liquid control system injection. • Scoping study highlights the expected flow paths and limitations of experiments. - Abstract: During an anticipated transient without scram (ATWS) scenario in certain boiling water reactor (BWR) systems, a standby liquid control system (SLCS) is used to inject a sodium pentaborate solution into the reactor system in order to quickly shut down (scram) the reactor without the use of the control rods. Some BWR designs utilize a SLCS that injects through a set of nozzles on a vertical pipe in the peripheral region of the lower plenum of the reactor vessel. During the scenario, system water levels are reduced and natural circulation flow rates down through the jet pump nozzles and up into the core are a small fraction of the rated system flow. It is during this period that the SLCS flows are considered. This work outlines some initial scoping studies completed by the staff at the Nuclear Regulatory Commission (NRC). An attempt at benchmarking the computational fluid dynamics (CFD) approach using a set of available test data from a small facility is outlined. Due to our lack of information related to specific details of the facility geometry along with the limited data available from the test, the benchmark exercise produced only a qualitative basis for selecting turbulence models and mesh density. A CFD model simulating a full-scale reactor system is developed for the lower plenum of a representative BWR/4 design and SLCS flows and mixing are studied under a range of flow conditions. The full-scale BWR simulation builds upon the lessons learned from the benchmark exercise. One challenge for this work is the large size of the domain and the relatively small size of the geometric details such as flow passages and gaps. The geometry is simplified to make meshing feasible by eliminating some of the small features. The

Experiments on the lower plenum response during a severe accident

International Nuclear Information System (INIS)

Henry, Robert E.; Hammersley, Robert J.; Klopp, George T.; Merilo, Mati

2004-01-01

Severe accident evaluations for nuclear reactors consider the response when the core materials have been overheated sufficient to melt and change geometry. One possible consequence of this is that molten core debris could drain into the lower plenum, as occurred in the TMI-2 accident. Given this state, several physical processes need to be analyzed, i.e. the extent of debris particulation and cooling, the potential for thermal attack of lower plenum structures, the thermal transient of the RPV and the potential for external cooling of the RPV lower head. These are important and complex processes, the evaluations of which need to be guided by well founded experiments. To support the development of the MAAP codes, recent experiments have been performed on specific issues such as: 1. the response of lower head penetrations submerged in a high temperature melt, 2. the net steam generation rate when molten debris drains into the lower plenum, 3. the formation of a contact resistance when molten debris drains through water and contacts the RPV wall and 4. the potential for external cooling of the RPV lower head. This paper discusses these experiments and their results. More importantly, it discusses how these are used in formulating models to represent the lower plenum response in the MAAP codes. (author)

Intake plenum volume and its influence on the engine performance, cyclic variability and emissions

International Nuclear Information System (INIS)

Ceviz, M.A.

2007-01-01

Intake manifold connects the intake system to the intake valve of the engine and through which air or air-fuel mixture is drawn into the cylinder. Details of the flow in intake manifolds are extremely complex. Recently, most of engine companies are focused on variable intake manifold technology due to their improvement on engine performance. This paper investigates the effects of intake plenum volume variation on engine performance and emissions to constitute a base study for variable intake plenum. Brake and indicated engine performance characteristics, coefficient of variation in indicated mean effective pressure (COV imep ) as an indicator for cyclic variability, pulsating flow pressure in the intake manifold runner, and CO, CO 2 and HC emissions were taken into consideration to evaluate the effects of different plenum volumes. The results of this study showed that the variation in the plenum volume causes an improvement on the engine performance and the pollutant emissions. The brake torque and related performance characteristics improved pronouncedly about between 1700 and 2600 rpm by increasing plenum volume. Additionally, although the increase in the plenum volume caused the mixture leaner due to the increase in the intake runner pressure and lean mixtures inclined to increase the cyclic variability, a decrease was interestingly observed in the COV imep

UPTF/TEST10B/RUN081, Steam/Water Flow Phenomena Reflood PWR Cold Leg Break LOCA

International Nuclear Information System (INIS)

1998-01-01

1 - Description of test facility: The Upper Plenum Test Facility (UPTF) is a geometrical full-scale simulation of the primary system of the four-loop 1300 MWe Siemens/KWU pressurized water reactor (PWR) at Grafenrheinfeld. The test vessel, upper plenum and its internals, downcomer, primary loops, pressurizer and surge line are replicas of the reference plant. The core, coolant pumps, steam generators and containment of a PWR are replaced by simulators which simulate the boundary and initial conditions during end-of-blowdown, refill and reflood phase following a loss-of-coolant accident (LOCA) with a hot or cold leg break. The break size and location can be simulated in the broken loop. The emergency core coolant (ECC) injection systems at the UPTF are designed to simulate the various ECC injection modes, such as hot leg, upper plenum, cold leg, downcomer or combined hot and cold leg injection of different ECC systems of German and US/Japan PWRs. Moreover, eight vent valves are mounted in the core barrel above the hot leg nozzle elevation for simulation of ABB and B and W PWRs. The UPTF primary system is divided into the investigation and simulation areas. The investigation areas, which are the exact replicas of a GPWR, consist of the upper plenum with internals, hot legs, cold legs and downcomer. The realistic thermal-hydraulic behavior in the investigation areas is assured by appropriate initial and boundary conditions of the area interface. The boundary conditions are realized by above mentioned simulators, the setup and the operation of which are based on small-scale data and mathematical models. The simulation areas include core simulator, steam generator simulators, pump simulators and containment simulator. The steam production and entrainment in a real core during a LOCA are simulated by steam and water injection through the core simulator. 2 - Description of test: Investigation of steam/water flow phenomena at the upper tie plate and in the upper plenum and

Assessment of TRAC-PD2 reflood core thermo-hydraulic model by CCTF Test C1-16

International Nuclear Information System (INIS)

Sugimoto, Jun

1982-11-01

The TRAC-PD2 reflood core thermo-hydraulic model was assessed by CCTF Test C1-16. The measured data were utilized as core boundary conditions in the TRAC calculations. The results indicate that the core inlet liquid temperature and the core heater rod temperatures are in reasonable agreement with data, but the pressure distribution in the core and water pool formation in the upper plenum are not in good agreement. The parametric effects of the droplet critical Weber number, the material properties of the heater rod, the noding of the upper plenum, and the minimum stable film boiling temperature are also discussed. (author)

Interferometric investigation of turbulently fluctuating temperature in an LMFBR outlet plenum geometry

International Nuclear Information System (INIS)

Bennett, R.G.; Golay, M.W.

1975-01-01

A novel optical technique is described for the measurement of turbulently fluctuating temperature in a transparent fluid flow. The technique employs a Mach-Zehnder interferometer of extremely short field and a simple photoconductive diode detector. The system produces a nearly linear D.C. electrical analog of the turbulent temperature fluctuations in a small, 1 mm 3 volume. The frequency response extends well above 2500 Hz, and can be improved by the choice of a more sophisticated photodetector. The turbulent sodium mixing in the ANL 1 1 / 15 -scale FFTF outlet plenum is investigated with a scale model outlet mixing plenum, using flows of air. The scale design represents a cross section of the ANL outlet plenum, so that the average recirculating flow inside the test cell is two dimensional. The range of the instrument is 120 0 F above the ambient air temperature. The accuracy is generally +-5 0 F, with most of the error due to noise originating from building vibrations and room noise. The power spectral density of the fluctuating temperature has been observed experimentally at six different stations in the flow. A strong 300 Hz component is generated in the inlet region, which decays as the flow progresses along streamlines. The effect of the inlet Reynolds number and the temperature difference between the inlet flows on the power spectral density has also been investigated. Traces of the actual fluctuating temperature are included for the six stations

Experimental evaluation of blockage ratio and plenum evacuation system flow effects on pressure distribution for bodies of revolution in 0.1 scale model test section of NASA Lewis Research Center's proposed altitude wind tunnel

Science.gov (United States)

Burley, Richard R.; Harrington, Douglas E.

1987-01-01

An experimental investigation was conducted in the slotted test section of the 0.1-scale model of the proposed Altitude Wind Tunnel to evaluate wall interference effects at tunnel Mach numbers from 0.70 to 0.95 on bodies of revolution with blockage rates of 0.43, 3, 6, and 12 percent. The amount of flow that had to be removed from the plenum chamber (which surrounded the slotted test section) by the plenum evacuation system (PES) to eliminate wall interference effects was determined. The effectiveness of tunnel reentry flaps in removing flow from the plenum chamber was examined. The 0.43-percent blockage model was the only one free of wall interference effects with no PES flow. Surface pressures on the forward part of the other models were greater than interference-free results and were not influenced by PES flow. Interference-free results were achieved on the aft part of the 3- and 6-percent blockage models with the proper amount of PES flow. The required PES flow was substantially reduced by opening the reentry flaps.

FFTF thermal-hydraulic testing results affecting piping and vessel component design in LMFBR's

International Nuclear Information System (INIS)

Stover, R.L.; Beaver, T.R.; Chang, S.C.

1983-01-01

The Fast Flux Test Facility completed four years of pre-operational testing in April 1982. This paper describes thermal-hydraulic testing results from this period which impact piping and vessel component design in LMFBRs. Data discussed are piping flow oscillations, piping thermal stratification and vessel upper plenum stratification. Results from testing verified that plant design limits were met

Post-Test Analysis of 11% Break at PSB-VVER Experimental Facility using Cathare 2 Code

Science.gov (United States)

Sabotinov, Luben; Chevrier, Patrick

The best estimate French thermal-hydraulic computer code CATHARE 2 Version 2.5_1 was used for post-test analysis of the experiment “11% upper plenum break”, conducted at the large-scale test facility PSB-VVER in Russia. The PSB rig is 1:300 scaled model of VVER-1000 NPP. A computer model has been developed for CATHARE 2 V2.5_1, taking into account all important components of the PSB facility: reactor model (lower plenum, core, bypass, upper plenum, downcomer), 4 separated loops, pressurizer, horizontal multitube steam generators, break section. The secondary side is represented by recirculation model. A large number of sensitivity calculations has been performed regarding break modeling, reactor pressure vessel modeling, counter current flow modeling, hydraulic losses, heat losses. The comparison between calculated and experimental results shows good prediction of the basic thermal-hydraulic phenomena and parameters such as pressures, temperatures, void fractions, loop seal clearance, etc. The experimental and calculation results are very sensitive regarding the fuel cladding temperature, which show a periodical nature. With the applied CATHARE 1D modeling, the global thermal-hydraulic parameters and the core heat up have been reasonably predicted.

Study on in-vessel thermohydraulics phenomena of sodium-cooled fast reactors. 1. Numerical investigation for the rationalization of hydrodynamics in the upper plenum

International Nuclear Information System (INIS)

Muramatsu, Toshiharu; Yamaguchi, Akira

2002-02-01

A large-scale sodium-cooled fast breeder reactor in feasibility studies on commercialized fast reactors has a tendency of consideration of thorough simplified and compacted system designs to realize drastic economical improvements. Therefore, special attention should be paid to thermohydraulic designs for a gas entrainment behavior from free surfaces, a flow-induced vibration of in-vessel components, a thermal shock for various structures due to high-speed coolant flows, nonsymmetrical coolant flows, etc. in the reactor vessel. In-vessel thermohydraulic analyses were carried out using a multi-dimensional code AQUA to understand the thermohydraulic characteristics in the upper plenum, and to investigate suitable in-vessel structure for the elimination of gas entrainment possibility. From the analysis, the following results were obtained. (1) It is difficult to rationale in-vessel flow patterns through adjustments of porous ratio and pressure loss for a hold down plate and baffle plates installed in an upper core structure. (2) Dummy plug insertion to a slit of the upper core structure is one of effective measures to stabilize in-vessel flow patterns. (3) Flow guide devices such as a baffle ring and a partial inner barrel are also effective measures to eliminate impinging jet to a dipped plate (D/P) and to reduce horizontal flow velocity components at free surface. (4) Installations of labyrinth structures to a R/V - D/P gap is successful for decreasing of free surface horizontal flows. (5) Gap closing of an in-vessel fuel pot and two cold trap components has the effects of reductions for free surface horizontal flows and for the difference of free surface levels. Following future investigations are important preventive measures against the gas entrainment from the free surface. (1) Flattening of spatial axial velocity distributions at the R/V - D/P gap. (2) Alleviation measures of vortex concentration at free surface. (3) Separation measures of 3-dimensional vortex

Measurement of heat and momentum eddy diffusivities in recirculating LMFBR outlet plenum flows

International Nuclear Information System (INIS)

Manno, V.P.; Golay, M.W.

1978-06-01

An optical technique has been developed for the measurement of the eddy diffusivity of heat in a transparent flowing medium. The method uses a combination of two established measurement tools: a Mach-Zehnder interferometer for the monitoring of turbulently fluctuating temperature and a Laser Doppler Anemometer (LDA) for the measurement of turbulent velocity fluctuations. The technique is applied to the investigation of flow fields characteristic of the LMFBR outlet plenum. The study is accomplished using air as the working fluid in a small scale Plexiglas test section. Lows are introduced into both the 1 / 15 scale FFTF outlet plenum and the 3 / 80 scale CRBR geometry plenum at inlet Reynolds numbers of 22,000. Measurements of the eddy diffusivity of heat and the eddy diffusivity of momentum are performed at a total of 11 measurement stations. Significant differences of the turbulence parameters are found between the two geometries, and the higher chimney structure of the CRBR case is found to be the major cause of the distinction. Spectral intensity studies of the fluctuating electronic analog signals of velocity and temperature are also performed. Error analysis of the overall technique indicates an experimental error of 10% in the determination of the eddy diffusivity of heat and 6% in the evaluation of turbulent momentum viscosity. In general it is seen that the turbulence in the cases observed is not isotropic, and use of isotropic turbulent heat and momentum diffusivities in transport modelling would not be a valid procedure

Mitigation of thermal transients by tube bundle inlet plenum design

International Nuclear Information System (INIS)

Oras, J.J.; Kasza, K.E.

1984-06-01

A multiphase program aimed at investigating the importance of thermal buoyancy to LMFBR steam-generator and heat-exchanger thermal hydraulics under low-flow transient conditions is being conducted in the Argonne Mixing Components Test Facility (MCTF) on a 60 0 sector shell-side flow model of the Westinghouse straight-tube steam generator being developed under the US/DOE large-component development program. A series of shell-side constant-flow thermal-downramp transient tests have been conducted focusing on the phenomenon of thermal-buoyancy-induced-flow channeling. In addition, it was discovered that a shell-inlet flow-distribution plenum can play a significant role in mitigating the severity of a thermal transient entering a steam generator or heat exchanger

Validation of turbulence models for LMFBR outlet plenum flows

International Nuclear Information System (INIS)

Chen, Y.B.; Golay, M.W.

1977-01-01

Small scale experiments involving water flows are used to provide mean flow and turbulence field data for LMFBR outlet plenum flows. Measurements are performed at Reynolds number (Re) values of 33000 and 70000 in a 1/15 - scale FFTF geometry and at Re = 35000 in a 3/80-scale CRBR geometry. The experimental behavior is predicted using two different two-equation turbulence model computer programs, TEACH-T and VARR-II. It is found that the qualitative nature of the flow field within the plenum depends strongly upon the distribution of the mean inlet flow field, importantly also upon the degree of inlet turbulence, and also upon the turbulent momentum exchange model used in the calculations. In the FFTF geometry, the TEACH-T predictions agree well with the experiments. 7 refs

Deformation and collapse of zircaloy fuel rod cladding into plenum axial gaps

International Nuclear Information System (INIS)

Pfennigwerth, P.L.; Gorscak, D.A.; Selsley, I.A.

1983-01-01

To minimize support structure, blanket and reflector fuel rods of the thoria urania-fueled Light Water Breeder Reactor (LWBR) were designed with non-freestanding Zircaloy-4 cladding. An analytical model was developed to predict deformation of unirradiated cladding into axial gaps of fuel rod plenum regions where it is unsupported. This model uses the ACCEPT finite element computer program to calculate elastic-plastic deformation of cladding due to external pressure. The finite element is 20-node, triquadratic, isoparametric, and 3-dimensional. Its curved surface permits accurate modeling of the tube geometry, including geometric nonuniformities such as circumferential wall thickness variation and initial tube out-of-roundness. Progressive increases in axial gap length due to cladding elongation and fuel stack shrinkage are modeled, as are deformations of fuel pellets and stainless steel support sleeves which bound plenum axial gaps in LWBR type blanket fuel rods. Zircaloy-4 primary and secondary thermal creep representations were developed from uniaxial creep testing of fuel rod tubing. Creep response to multi-axial loading is modeled with a variation of Hill's formulation for anisotropic materials. Coefficients accounting for anisotropic thermal creep in Zircaloy-4 tubes were developed from creep testing of externally pressurized tubes having fixed axial gaps in the range 2.5 cm to 5.7 cm and radial clearances over simulated fuel pellets ranging from zero to 0.089 mm. (orig./RW)

Evaluation report on CCTF core-II reflood test C2 - 8 (Run 67)

International Nuclear Information System (INIS)

Akimoto, Hajime; Iguchi, Tadashi; Okubo, Tsutomu; Murao, Yoshio; Okabe, Kazuharu; Sugimoto, Jun.

1987-01-01

In order to study the system pressure effect of the core cooling and flow behavior during the reflood phase of a PWR LOCA, a test was performed with CCTF under the system pressure pf 0.15 MPa as a counterpart test of the CCTF test C2-1 (system pressure 0.42 MPa) and the CCTF test C2-4 (system pressure 0.20 MPa). Through the comparisons of results from these three tests, the following conclusions were obtained: (1) The higher system pressure resulted in the lower temperature rise, the shorter turnaround time and the shorter quench time as observed in the CCTF Core-I system pressure effect tests. (2) The higher system pressure resulted in higher core water head, higher upper plenum water head, higher mass flow rate through the primary loops. On the other hand, the higher system pressure resulted in lower downcomer water head and lower pressure drop through the primary loops and the broken cold leg. These system pressure effects on the flow behavior in the primary system are almost the same as observed in the system pressure effect tests in the CCTF Core-I test series. (3) Before the mixture level in the upper plenum reached the level of the hot leg nozzle, the loop flow resistance coefficient of the intact loops was nearly constant regardless of the system pressure. After the mixture level reached the level of the hot leg nozzle, the loop flow resistance coefficient was increased due to the water accumulation in the hot leg piping and the inlet plenum of the steam generator in these tests. (J.P.N.)

PBDOWN: A computer code for simulation of core material discharge and expansion in the upper coolant plenum in a hypothetical unprotected loss of flow accident in a LMFBR

International Nuclear Information System (INIS)

Royl, P.

1985-01-01

The report gives a description of the code PBDOWN (Pool Blow Down), its equations, input specifications and subroutines and it lists the input and output for some samples. Besides that some analysis results for the SNR-300 are discussed, that were obtained with this code. PBDOWN is an integral blow-down and expansion code, which simulates core material discharge and expansion into a sodium filled upper coolant plenum after build-up of vapour pressures in an unprotected loss of flow accident. The model includes the effect of sodium entrainment into an expending bubble of fuel or steel vapour with various assumptions for the heat transfer and vaporization of the entrained sodium droplets. The expanding vapour bubble is connected to the discharging pool via an orifice of a given size through which a time dependent ejection is simulated using quasi-stationary blow down correlations. The model allows bounding analysis of the possible influence of sodium vapour as a secondary working fluid, that is activated outside the pool on the overall expansion energy and discharge

Creys-Malville nuclear plant. Simulation of the cold plenum thermal-hydraulics. 12 zone model presentation

International Nuclear Information System (INIS)

Faulot, J.P.

1990-05-01

The CRUSIFI code has been developed by SEPTEN (Engineering and Construction Division) with SICLE software during 1983-1985 in order to study the CREYS-MALVILLE dynamic behavior. At the time, the version was based on project data (version 2.3). It includes a 2 zones model for the cold plenum thermal-hydraulics, modelling which does not allow to reproduce accurately dissymetries apt to occur as well in usual operating (hydraulic dissymetries bound to one or many systems out of order), as during incidentally operating (hydraulic dissymetries bound to primary pump working back or thermal dissymetries after a transient on one or many secondary loops). Moreover, a 2 zones model cannot simulate axial temperature gradients which appear during double stratification phenomenon (upper and lower part of the plenum) produced by alternating thermal shock. A 12 zones model (4 sectors with 3 axial zones each) such as model developed by R$DD (Research and Development Division) allows to satisfy correctly these problems. This report is a specification of the chosen modelling. This model is now operational after qualifying with experimental transients on mockup and reactor. It is to-day connected with the EDF general operating code CRUSIFI (calibrating version 3.0). It could be easily integrated in a four loops plant modelling such as the CREYS-MALVILLE simulator in a four loops plant modelling such as the CREYS-MALVILLE simulator under construction at the present time by THOMSON

COBRA/TRAC analysis of two-dimensional thermal-hydraulic behavior in SCTF reflood tests

International Nuclear Information System (INIS)

Iwamura, Takamichi; Ohnuki, Akira; Sobajima, Makoto; Adachi, Hiromichi

1987-01-01

The effects of radial power distribution and non-uniform upper plenum water accumulation on thermal-hydraulic behavior in the core were observed in the reflood tests with Slab Core Test Facility (SCTF). In order to examine the predictability of these two effects by a multi-dimensional analysis code, the COBRA/TRAC calculations were made. The calculated results indicated that the heat transfer enhancement in high power bundles above quench front was caused by high vapor flow rate in those bundles due to the radial power distribution. On the other hand, the heat transfer degradation in the peripheral bundles under the condition of non-uniform upper plenum water accumulation was caused by the lower flow rates of vapor and entrained liquid above the quench front in those bundles by the reason that vapor concentrated in the center bundles due to the cross flow induced by the horizontal pressure gradient in the core. The above-mentioned two-dimensional heat transfer behaviors calculated with the COBRA/TRAC code is similar to those observed in SCTF tests and therefore those calculations are useful to investigate the mechanism of the two-dimensional effects in SCTF reflood tests. (author)

Influence of graphite discs, chamfers, and plenums on temperature distributions in high burnup fuel

International Nuclear Information System (INIS)

Ranger, A.; Tayal, M.; Singh, P.

1990-04-01

Previous studies have demonstrated the desirability to increase the fuel burnups in CANDU reactors from 7-9 GW.d/t to 21 GW.d/t. At high burnups, one consideration in fuel integrity is fission gas pressure, which is predicted to reach about 13 MPa. The gas pressure can be kept below the coolant pressure (about 10 MPa) via a variety of options such as bigger chamfers, deeper dishes, central hole, and plenums. However, it is important to address the temperature perturbations produced by the bigger chamfers and plenums which in turn, affect the gas pressure. Another consideration in fuel integrity is to reduce the likelihood of fuel failures via environmentally assisted cracking. Insertion of graphite discs between neighbouring pellets will lower the pellet temperatures, hence, lower fission gas release and lower expansion of the pellet. Therefore, it is desired to quantify the effect of graphite discs on pellet temperatures. Thermal analyses of different fuel element geometries: with and without chamfers, graphite discs, and plenums were performed. The results indicate that the two-dimensional distributions of temperatures due to the presence of chamfers, graphite discs, or plenums can have a significant impact on the integrity of high burnup fuel as we have been able to quantify in this paper

Reduction of sound transmission across plenum windows by incorporating an array of rigid cylinders

Science.gov (United States)

Tang, S. K.

2018-02-01

The potential improvement of plenum window noise reduction by installing rigid circular cylinder arrays into the window cavity is investigated numerically using the finite-element method in this study. A two-dimensional approach is adopted. The sound transmission characteristics and propagation within the plenum window are also examined in detail. Results show that the installation of the cylinders in general gives rise to broadband improvement of noise reduction across a plenum window regardless of the direction of sound incidence. Such acoustical performance becomes better when more cylinder columns are installed, but it is suggested that the number of cylinder rows should not exceed two. Results also show that the cylinder positions relative to the nodal/anti-nodal planes of the acoustic modes are crucial in the noise reduction enhancement mechanisms. Noise reduction can further be enhanced by staggering the cylinder rows, such that each cylinder row supports the development of a different acoustic mode. For the simple cylinder arrangements considered in this study, the traffic noise reduction enhancement observed in this study can be as high as 4-5 dB, which is already comparable to or higher than the maximum achieved by installing sound absorption into a plenum window.

System pressure effects on reflooding phenomena observed in the SCTF Core-I forced flooding tests

International Nuclear Information System (INIS)

Adachi, Hiromichi; Sudo, Yukio; Sobajima, Makoto; Iwamura, Takamichi; Osakabe, Masahiro; Ohnuki, Akira; Abe, Yutaka

1983-06-01

The Slab Core Test Facility was constructed to investigate two-dimensional thermo-hydrodynamics in the core and the interaction in fluid behavior between the core and the upper plenum during the last part of blowdown, refill and reflood phases of a posturated loss-of-coolant accident (LOCA) of a pressurized water reactor (PWR). The present report described the analytical results on the effects of system pressure on reflooding phenomena observed in Tests Sl-SH2, Sl-01 and Sl-02 which are belonging to the SCTF Core-I forced-feed reflooding test series. Nominal system pressures in these tests are 0.4, 0.2 and 0.15 MPa, respectively. By comparison among the data of these three tests, the effects of system pressure on thermo-hydrodynamic behavior in the pressure vessel including the core and the primary coolant loops of the SCTF can be clarified under the forced flooding condition. Major items investigated in the present report are (1) overall temperature behaviors in the core, (2) change of heat transfer coefficient and heat flux at the rod surface before the quench, (3) two-dimensional thermo-hydrodynamic behaviors in the core and upper plenum and (4) hot leg carryover. (author)

Gratiae plenum: Latin, Greek and the Cominform

Directory of Open Access Journals (Sweden)

David Movrin

2010-12-01

Full Text Available The survival of classics in the People’s Republic of Slovenia after World War II was dominated by the long shadow of the Coryphaeus of the Sciences, Joseph Stalin. Since 1945, the profile of the discipline was determined by the Communist Party, which followed the Soviet example, well-nigh destroying the classical education in the process. Fran Bradač, head of Classics at the University of Ljubljana, was removed for political reasons; the classical gymnasium belonging to the Church was closed down; Greek was struck from the curriculum of the two remaining state classical gymnasia; Latin, previously a central subject at every gymnasium, was severely reduced in 1945, only to disappear entirely in 1946. The classicists who continued to teach were forced to take ‘reorientation courses’ which enabled them to teach Russian and other more suitable subjects. By 1949, only two out of the 42 classicists employed by the Ministry of Education were actually teaching Latin. The Classics department at the university, where only two students were studying in 1949, was on the brink of closure.  Paradoxically, the classical tradition was saved by Stalin’s attack on the same Party. The Cominform conflict in 1948 astonished the Yugoslav communists and pushed them towards a tactical détente with the West, prompting a revision of some of their policies, including education. The process was led by the top echelons of the Party — such as Milovan Djilas, head of the central Agitprop, Boris Kidrič, in charge of Yugoslav economy, and Edvard Kardelj, the Party’s chief ideologue — during the Third Plenum of the Central Committee Politburo in Belgrade in December 1949. Their newly discovered love of Latin and Greek, documented in the minutes of the Politburo Plenum, was overseen only by the discriminating eye of Josip Broz Tito. Classical gymnasia were revived, Latin was reintroduced to some of the other gymnasia, students returned to study classics at the

Summary report of incineration plenum fire: Building 771, July 2, 1980

International Nuclear Information System (INIS)

Fretthold, J.K.

1995-01-01

At about 1100 on July 2, 1980, a temperature rise above normal was recorded on charts monitoring operation of the incinerator in Room 149, Building 771. The plenum overheat alarm sounded at 1215, emergency actions initiated, and the fire was extinguished and mop-up began at about 1300. Investigation determined that the fire in the plenum was caused by a heat rise in the system, a deteriorated bypass valve on the No. 3 heat exchanger (KOH scrubber), nitration of the urethane seal on the HEPA filter media to the filter frame, and accumulation of metallic fines on the filter media. It was concluded that the management system responded properly, except for the ring- down system to activate the Emergency Operations Center

BMFT-UPTF densitometer system test report

International Nuclear Information System (INIS)

Menkhaus, D.E.

1985-11-01

This report documents acceptance test results performed on the five Upper Plenum Test Facility (UPTF) three-beam densitometer systems and spare parts. The five densitometer systems are used on the UPTF four hot legs and broken cold leg to measure average chordal-beam densities. The primary objectives of the tests performed were: to verify all assemblies fit as designed (mechanical fitup); to ensure radiation levels met the criteria (<2.5 mR/h); to verify that design accuracy requirements were met (performance tests); and to verify proper operation of the densitometer systems (functional checks). 15 figs., 11 tabs

Numerical investigation of flow characteristics in a prototypical lower plenum of a prismatic VHTR

International Nuclear Information System (INIS)

Ying, Alice; Narula, Manmeet; Abdou, Mohamed; Tsai, Peter; Ando, Yuya

2007-01-01

The aim of this study is to obtain insights into the flow behavior, as well as to develop predictive capability with regards to the flow and thermal mixing, that occurs in the lower plenum of a typical prismatic VHTR (Very High Temperature Reactor) concept. In this paper, numerical modeling has been used to capture qualitative phenomena observed during an experiment performed at INL, using a finite volume, thermo-fluid solver system, 'SC/Tetra' from CRADLE. The choice of the correct turbulence model is critical to accurately predict the flow in the VHTR lower plenum. Four different turbulence models have been used in this study and the flow predictions are significantly different. A trail of marker particles and fluid temperature as a passive scalar have been used to qualitatively study the flow characteristics, specifically the turbulent mixing of water jets. The quantitative experimental data, when available, will be used to compare and improve on the available turbulence models. Preliminary numerical modeling has been carried out to address the issue of hot streaking and buoyancy effects of hot helium jets in the lower plenum. (author)

Evaluation report on SCTF Core-III Test S3-22

International Nuclear Information System (INIS)

Okubo, Tsutomu; Iguchi, Tadashi; Iwamura, Takamichi; Akimoto, Hajime; Ohnuki, Akira; Abe, Yutaka; Murao, Yoshio; Adachi, Hiromichi.

1991-07-01

Two tests (Tests S3-20 and S3-22) were conducted with JAERI's Slab Core Test Facility (SCTF) Core-III in order to investigate water break-through and core cooling behaviors under the intermittent ECC water delivery from the hot legs to one location in the upper plenum and the alternate ECC water delivery to two locations in the upper plenum during reflooding, respectively. This report presents an analysis on Test S3-22 (the alternate case). Subcooled ECC water was injected alternately just above the upper core support plate above Bundles 7 and 8 and Bundles 3 and 4. The total injection rate from both injection ports was the same as that in SCTF Test S3-20 and Test S3-13. Analyzing the test data together with those of Tests S3-13 and S3-20 the following has been found: (1) Alternate break-through occurred immediately corresponding to the alternate ECC water injection except for one period, during which no break-through was observed. However, there observed a difference in break-through behavior that break-through was strong above the low power region, whereas weak above the high power region. (2) Although its break-through behavior was different, nearly the same core cooling as in the continuous or intermittent ECC water delivery case was observed except for the period around quench. (3) Around quench time, degraded core cooling comparing to the continuous or intermittent ECC water delivery case was observed. That is, quench time at the midplane level of the present test was 35 s later than in the continuous case. This is considered to result from decrease in core water inventory caused by water sealing at the cross-over leg. (J.P.N.)

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

International Nuclear Information System (INIS)

1992-01-01

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

Flow distribution in the inlet plenum of steam generator

International Nuclear Information System (INIS)

Khadamakar, H.P.; Patwardhan, A.W.; Padmakumar, G.; Vaidyanathan, G.

2011-01-01

Highlights: → Various flow distribution devices have been studied to make the flow distribution uniform in axial as well as tangential direction. → Experiments were performed using Ultrasonic Velocity Profiler (UVP) and Particle Image Velocimetry (PIV). → CFD modeling has been carried out to give more insights. → Various flow distribution devices have been compared. - Abstract: The flow distribution in a 1/5th and 1/8th scale models of inlet plenum of steam generator (SG) has been studied by a combination of experiments and Computational Fluid Dynamics (CFD) simulations. The distribution of liquid sodium in the inlet plenum of the SG strongly affects the thermal as well as mechanical performance of the steam generator. Various flow distribution devices have been used to make the flow distribution uniform in axial as well as tangential direction in the window region. Experiments have been conducted to measure the radial velocity distribution using Ultrasonic Velocity Profiler (UVP) and Particle Image Velocimetry (PIV) under a variety of conditions. CFD modeling has been carried out for various configurations to give more insight into the flow distribution phenomena. The various flow distribution devices have been compared on the basis of a non-uniformity index parameter.

nijotech vol. 6 no. 1 september 1982 iloeje 25 effects of parallel

African Journals Online (AJOL)

Dr Obe

upper plenum into the channels and lower plenum of Boiling Water Nuclear Power. Reactors during ... pressures varied from near atmospheric to a little over 1.7 bar. Test section ... The steam generator was an Electro-Magic (Model. 3100) unit ...

A CFD Study on Inlet Plenum Flow Field of Pebble Bed Reactor

International Nuclear Information System (INIS)

Kim, Min Hwan; Lee, Won Jae; Chang, Jong Hwa

2005-01-01

High temperature gas cooled reactor, largely divided into two types of PBR (Pebble Bed Reactor) and PMR (Prismatic Modular Reactor), has becomes great interest of researchers in connection with the hydrogen production. KAERI has started a project to develop the gas cooled reactor for the hydrogen production and has been doing in-depth study for selecting the reactor type between PBR and PMR. As a part of the study, PBMR (Pebble Bed Modular Reactor) was selected as a reference PBR reactor for the CFD analysis and the flow field of its inlet plenum was simulated with computational fluid dynamics program CFX5. Due to asymmetrical arrangement of pipes to the inlet plenum, non-uniform flow distribution has been expected to occur, giving rise to non-uniform power distribution at the core. Flow fields of different arrangement of inlet pipes were also investigated, as one of measures to reduce the non-uniformity

Two-dimensional thermal-hydraulic behavior in core in SCTF Core-II cold leg injection tests

International Nuclear Information System (INIS)

Iwamura, Takamichi; Sobajima, Makoto; Okubo, Tsutomu; Ohnuki, Akira; Abe, Yutaka; Adachi, Hiromichi

1985-07-01

Major purpose of the Slab Core Test Program is to investigate the two-dimensional thermal-hydraulic behavior in the core during the reflood phase in a PWR-LOCA. In order to investigate the effects of radial power profile, three cold leg injection tests with different radial power profiles under the same total heating power and core stored energy were performed by using the Slab Core Test Facility (SCTF) Core-II. It was revealed by comparing these three tests that the heat transfer was enhanced in the higher power bundles and degraded in the lower power bundles in the non-uniform radial power profile tests. The turnaround temperature in the high power bundles were evaluated to be reduced by about 40 to 120 K. On the other hand, a two-dimensional flow in the core was also induced by the non-uniform water accumulation in the upper plenum and the quench was delayed resultantly in the bundles corresponding to the peripheral bundles of a PWR. However, the effect of the non-uniform upper plenum water accumulation on the turnaround temperature was small because the effect dominated after the turnaround of the cladding temperature. Selected data from Tests S2-SH1, S2-SH2 and S2-O6 are also presented in this report. Some data from Tests S2-SH1 and S2-SH2 were compared with TRAC post-test calculations performed by the Los Alamos National Laboratory. (author)

Melt jet fragmentation and oxidation in the lower plenum

International Nuclear Information System (INIS)

Berthoud, G.

2001-01-01

During the late phases of a PWR Severe Accident, the core materials discharge into the lower plenum in which water is still present. In that case, we are then concerned by the possible occurrence of a Steam Explosion which may endanger the vessel structure and by the following cooling of the melt debris. So, we have two possible ways of vessel rupture: a mechanical one following an energetic Steam Explosion and a thermal one due to insufficient debris cooling. Both types of problems are linked with the degree of fragmentation of the core material during its penetration into the water of the lower plenum. One of the most likely mode of discharge consists in corium streams or jets. The fragmentation will build a corium-water mixture (the pre-mixing sequence) which, under certain circumstances, may undergo a fine fragmentation sequence leading to an energetic Steam Explosion (the explosion sequence). Whatever the occurrence of a Steam Explosion, the resulting debris will accumulate at the bottom of the Reactor Vessel and the cooling of such a ''debris bed'' is known to be highly dependant of the granulometry and build up of the debris bed which are linked with the previous sequence of corium fragmentation and dispersion. In CEA, the MC3D Code has been developed to deal with all these phenomena. (author)

Evaluation report on SCTF Core-II test S2-08

International Nuclear Information System (INIS)

Ohnuki, Akira; Iwamura, Takamichi; Abe, Yutaka; Murao, Yoshio; Adachi, Hiromichi.

1991-01-01

The present report investigates the effects of the difference of the core inlet subcooling during reflood in a PWR-LOCA on the thermal-hydraulic behaviors including two-dimensional behaviors in the pressure vessel in the Slab Core Test Facility (SCTF) Core-II tests under gravity feed mode. The following test results are examined: Tests S2-02 (Reference test) and Test S2-08 (High subcooling test). The degree of the difference of the subcooling between the two tests was about 20 to 35 K in the LPCI period. The following conclusions were obtained from this study: (1) Higher the subcooling gave larger amount of water accumulation in the core and gave better core cooling. These tendencies were also recognized in comparisons under the same distance from the quench front. Since the same tendencies can be predicted in the analyses with REFLA code because of the lower steam generation rate below quench front in the high subcooling test, the differences in the tests are supposed to be caused by the same reason. (2) Higher the subcooling gave larger amount of water accumulation in upper plenum. The carry-over liquid mass into hot leg became smaller in the later period in the higher subcooling test. These differences for carry-over and de-entrainment characteristics can be explained by the differences of quench velocity and of steam mass flow rate generated in the core. (3) No significant influence of the different degree of the subcooling was observed on the two-dimensional thermal-hydraulic behaviors in the pressure vessel. Namely, radial differences of sectional void fraction, heat transfer coefficient and the pressure among bundles at the same elevation were almost the same amount for the two tests. Radial differences of liquid levels in the upper plenum was also almost the same amount for the two tests. (J.P.N.)

Experimental study of core thermohydraulics in fast reactors during transition from forced to natural circulation. Influence of inter-wrapper flow

International Nuclear Information System (INIS)

Kamide, H.; Hayashi, K.; Momoi, K.

1997-01-01

The evaluation of core thermohydraulics under natural circulation conditions is important to utilize inherent safety features of Fast Reactors. When heat exchangers of a decay heat removal system are operated in an upper plenum of reactor vessel, cold sodium is provided by the heat exchangers. Core-plenum interactions will occur during a natural circulation condition due to this cold sodium in the upper plenum, e.g., it can penetrate into gap regions between fuel subassemblies (inter-wrapper flow, IWF) and the flow may reverse in low power core channels. These interactions will significantly modify the flow and temperature distributions in the core. Sodium experiments were carried out to study these phenomena. In a test section, seven subassemblies are housed and connected to an upper plenum. The influences of core-plenum interactions on the core thermohydraulics were investigated under steady state conditions and also in transitions from forced to natural circulation. Cooling effects of IWF on the fuel subassemblies were found in spite of natural circulation flow reduction in the primary loop due to temperature decreases in the upper non-heated section in the core. The inter-wrapper flow can effectively cool the core under extreme conditions of low flow rates through the core. (author)

Numerical simulation and geometry optimization of hot-gas mixing in lower plenum of high temperature gas-cooled reactor

International Nuclear Information System (INIS)

Wang Hang; Wang Jie; Laurien, E.

2010-01-01

The lower plenum in high temperature gas-cooled reactor was designed to mix the gas of different temperatures from the reactor core. Previous researches suggest the current geometry of the lower plenum to be improved for better mixing capability and lower pressure drop. In the presented work, a series of varied geometries were investigated with numerical simulation way. The choice of appropriate mesh type and size used in the geometry variation was discussed with the reference of experimental data. The original thin ribs in the current design were merged into thicker ones, and a junction located at the starting end of the outlet pipe was introduced. After comparing several potential optimization methods, an improved geometry was selected with the merged ribs increasing the pre-defined mixing coefficient and the junction reducing the pressure drop. Future work was discussed based on the simulation of real reactor case. The work shows a direction for design improvements of the lower plenum geometry. (authors)

Decontamination and demolition of a former plutonium processing facility's process exhaust system, firescreen, and filter plenum buildings

International Nuclear Information System (INIS)

LaFrate, P.J. Jr.; Stout, D.S.; Elliott, J.W.

1996-01-01

The Los Alamos National Laboratory (LANL) Decommissioning Project has decontaminated, demolished, and decommissioned a process exhaust system, two filter plenum buildings, and a firescreen plenum structure at Technical Area 21 (TA-2 1). The project began in August 1995 and was completed in January 1996. These high-efficiency particulate air (HEPA) filter plenums and associated ventilation ductwork provided process exhaust to fume hoods and glove boxes in TA-21 Buildings 2 through 5 when these buildings were active plutonium and uranium processing and research facilities. This paper summarizes the history of TA-21 plutonium and uranium processing and research activities and provides a detailed discussion of integrated work process controls, characterize-as-you-go methodology, unique engineering controls, decontamination techniques, demolition methodology, waste minimization, and volume reduction. Also presented in detail are the challenges facing the LANL Decommissioning Project to safely and economically decontaminate and demolish surplus facilities and the unique solutions to tough problems. This paper also shows the effectiveness of the integrated work package concept to control work through all phases

Decontamination and demolition of a former plutonium processing facility's process exhaust system, firescreen, and filter plenum buildings

International Nuclear Information System (INIS)

LaFrate, P.J. Jr.; Stout, D.S.; Elliott, J.W.

1996-01-01

The Los Alamos National Laboratory (LANL) Decommissioning Project has decontaminated, demolished, and decommissioned a process exhaust system, two filter plenum buildings, and a firescreen plenum structure at Technical Area 21 (TA-21). The project began in August 1995 and was completed in January 1996. These high-efficiency particulate air (HEPA) filter plenums and associated ventilation ductwork provided process exhaust to fume hoods and glove boxes in TA-21 Buildings 2 through 5 when these buildings were active plutonium and uranium processing and research facilities. This paper summarizes the history of TA-21 plutonium and uranium processing and research activities and provides a detailed discussion of integrated work process controls, characterize-as-you-go methodology, unique engineering controls, decontamination techniques, demolition methodology, waste minimization, and volume reduction. Also presented in detail are the challenges facing the LANL Decommissioning Project to safely and economically decontaminate and demolish surplus facilities and the unique solutions to tough problems. This paper also shows the effectiveness of the integrated work package concept to control work through all phases

Experiment data report for Semiscale Mod-1 Test S-05-5 (alternate ECC injection test)

International Nuclear Information System (INIS)

Collins, B.L.; Patton, M.L. Jr.; Sackett, K.E.

1977-04-01

Recorded test data are presented for Test S-05-5 of the Semiscale Mod-1 alternate ECC injection test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-05-5 was conducted from initial conditions of 2263 psia and 537 0 F to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the cold leg broken loop piping. During the test, cooling water was injected into the cold leg of the intact and broken loops to simulate emergency core coolant injection in a PWR. The upper plenum was vented through a reflood bypass line interconnecting the hot and cold legs of the broken loop

Development Of An Experiment For Measuring Flow Phenomena Occurring In A Lower Plenum For VHTR CFD Assessment

Energy Technology Data Exchange (ETDEWEB)

D. M. McEligot; K.G. Condie; G. E. Mc Creery; H. M. Mc Ilroy

2005-09-01

The objective of the present report is to document the design of our first experiment to measure generic flow phenomena expected to occur in the lower plenum of a typical prismatic VHTR (Very High Temperature Reactor) concept. In the process, fabrication sketches are provided for the use of CFD (computational fluid dynamics) analysts wishing to employ the data for assessment of their proposed codes. The general approach of the project is to develop new benchmark experiments for assessment in parallel with CFD and coupled CFD/systems code calculations for the same geometry. One aspect of the complex flow in a prismatic VHTR is being addressed: flow and thermal mixing in the lower plenum ("hot streaking" issue). Current prismatic VHTR concepts were examined to identify their proposed flow conditions and geometries over the range from normal operation to decay heat removal in a pressurized cooldown. Approximate analyses were applied to determine key non-dimensional parameters and their magnitudes over this operating range. The flow in the lower plenum can locally be considered to be a situation of multiple jets into a confined crossflow -- with obstructions. Flow is expected to be turbulent with momentum-dominated turbulent jets entering; buoyancy influences are estimated to be negligible in normal full power operation. Experiments are needed for the combined features of the lower plenum flows. Missing from the typical jet experiments available are interactions with nearby circular posts and with vertical posts in the vicinity of vertical walls - with near stagnant surroundings at one extreme and significant crossflow at the other.

Effects of Parallel Channel Interactions, Steam Flow, Liquid Subcool ...

African Journals Online (AJOL)

Tests were performed to examine the effects of parallel channel interactions, steam flow, liquid subcool and channel heat addition on the delivery of liquid from the upper plenum into the channels and lower plenum of Boiling Water Nuclear Power Reactors during reflood transients. Early liquid delivery into the channels, ...

Evaluation report on CCTF core-I reflood test C1-5 (Run 14)

International Nuclear Information System (INIS)

Murao, Yoshio; Akimoto, Hajime; Sudoh, Takashi; Okubo, Tsutomu

1983-02-01

A study of a cylindrical core test facility (CCTF) test was performed for modeling the system behavior during the reflood phase of a PWR-LOCA and the following conclusions were obtained: 1) With the exception of some points, the observed phenomena are similar to a model derived from an evaluation model for a PWR safety evaluation. 2) The different points are the water accumulation in the upper plenum, the ECC bypass in the downcomer, the reduction of the effective downcomer head and the pressure drop at the broken cold leg nozzle and in the interconnected pipes. (author)

Development of Lower Plenum Molten Pool Module of Severe Accident Analysis Code in Korea

International Nuclear Information System (INIS)

Son, Donggun; Kim, Dong-Ha; Park, Rae-Jun; Bae, Jun-Ho; Shim, Suk-Ku; Marigomen, Ralph

2014-01-01

in the program. Stand-alone SIMPLE program only uses limited input from the core, RCS and LHVF module from the COMPASS code. SIMPLE program calculates the behavior of the water and steam in the lower plenum, heat transfer mechanisms through the interfaces among the water, steam, corium jet, debris bed, metallic and oxidic pools as well as the crust above the oxidic pool. The case of uniform melt relocation was tested for the verification and limited validation of the SIMPLE program

Estimate of LOCA-FI plenum pressure uncertainty for a five-ring RELAP5 production reactor model

International Nuclear Information System (INIS)

Griggs, D.P.

1993-03-01

The RELAP5/MOD2.5 code (RELAP5) is used to perform best-estimate analyses of certain postulated Design Basis Accidents (DBAs) in SRS production reactors. Currently, the most limiting DBA in terms of reactor power level is an instantaneous double-ended guillotine break (DEGB) loss of coolant accident (LOCA). A six-loop RELAP5 K Reactor model is used to analyze the reactor system behavior dozing the Flow Instability (FI) phase of the LOCA, which comprises only the first 5 seconds following the DEGB. The RELAP5 K Reactor model includes tank and plenum nodalizations having five radial rings and six azimuthal sectors. The reactor system analysis provides time-dependent plenum and tank bottom pressures for use as boundary conditions in the FLOWTRAN code, which models a single fuel assembly in detail. RELAP5 also performs the system analysis for the latter phase of the LOCA, denoted the Emergency Cooling System (ECS) phase. Results from the RELAP analysis are used to provide boundary conditions to the FLOWTRAN-TF code, which is an advanced two-phase version of FLOWTRAN. The RELAP5 K Reactor model has been tested for LOCA-FI and Loss-of-Pumping Accident analyses and the results compared with equivalent analyses performed with the TRAC-PF1/MOD1 code (TRAC). An equivalent RELAP5 six-loop, five-ring, six-sector L Reactor model has been benchmarked against qualified single-phase system data from the 1989 L-Area In-Reactor Test Program. The RELAP5 K and L Reactor models have also been subjected to an independent Quality Assurance verification

Integral analysis of debris material and heat transport in reactor vessel lower plenum

International Nuclear Information System (INIS)

Suh, K.Y.; Henry, R.E.

1994-01-01

An integral, fast-running, two-region model has been developed to characterize the debris material and heat transport in the reactor lower plenum under severe accident conditions. The debris bed is segregated into the oxidic pool and an overlying metallic layer. Debris crusts can develop on three surfaces: the top of the molten pool, the RPV wall, and the internal structures. To account for the decay heat generation, the crust temperature profile is assumed to be parabolic. The oxidic debris pool is homogeneously mixed and has the same material composition, and hence the same thermophysical properties, as the crusts, while the metallic constituents are assumed to rise to the top of the debris pool. Steady-state relationships are used to describe the heat transfer rates, with the assessment of solid or liquid state, and the liquid superheat in the pool being based on the average debris temperature. Natural convection heat transfer from the molten debris pool to the upper, lower and embedded crusts is calculated based on the pool Rayleigh number with the conduction heat transfer from the crusts being determined by the crust temperature profile. The downward heat flux is transferred to the lowest part of the RPV lower head through a crust-to-RPV contact resistance. The sideward heat flux is transferred to the upper regions of the RPV lower head as well as to the internal structures. The upward heat flux goes to the metal layer, water, or available heat sink structures above. Quenching due to water ingression is modeled separately from the energy transfer through the crust. The RPV wall temperature distribution and the primary system pressure are utilized to estimate challenges to the RPV integrity. ((orig.))

Molten material relocation into the lower plenum: a status report

International Nuclear Information System (INIS)

1998-09-01

This report, prepared by the task group 'Degraded Core Cooling' (DCC) for the CSNI, summarizes the experimental and theoretical knowledge of molten material relocation from a degraded core to the lower plenum of the reactor vessel under the main severe accident scenarios envisaged for both PWRs and BWRs, and boundary conditions. Consequences of movement of material to the lower head are considered with respect to the potential for reactor pressure vessel failure. The following models are reviewed: SCDAP/RELAP5, ICARE/CATHARE, ATHLET-CD/KESS, MELCOR, MAAP4, ESCADRE, etc.

SCTF Core-I test results

International Nuclear Information System (INIS)

Adachi, Hiromichi; Sudo, Yukio; Iwamura, Takamichi; Osakabe, Masahiro; Ohnuki, Akira; Hirano, Kemmei

1982-07-01

The Slab Core Test Facility (SCTF) of Japan Atomic Energy Research Institute (JAERI) was constructed to investigate two-dimensional thermohydrodynamics in the core and the communication in fluid behavior between the core and the upper plenum during the last part of blowdown, refill and reflood phases of a posturated loss-of-coolant accident (LOCA) of a pressurized water reactor (PWR). In the present report, effects of system pressure on reflooding phenomena shall be discussed based on the data of Tests S1-SH2, S1-01 and S1-02 which are the parameteris tests for system pressure effects belonging to the SCTF Core-I forced flooding test series. Major items discussed in this report are (1) hydrodynamic behavior in the system, (2) core thermal behavior, (3) core heat transfer and (4) two-dimensional hydrodynamic behavior in the pressure vessel including the core. (author)

Fracture mechanics evaluation of LOFT lower plenum injection nozzle

International Nuclear Information System (INIS)

Nagata, P.K.; Reuter, W.G.

1977-01-01

An analysis to establish whether or not a leak-before-break concept would apply to the LOFT lower plenum injection nozzle is described. The analysis encompassed the structure from the inlet side of valve V-2170 to the lower plenum nozzle-to-reactor vessel weld on the left side of the emergency core cooling system (ECCS). The defect that was assumed to exist was of such a size that the probability of its being missed by the applicable inspection technique was near zero. The Inconel 600 nozzle forging with an initial assumed defect size of 0.64 cm (0.25 in.) deep would behave as follows: (1) the axially oriented defect would result in leak before rupture (the number of cycles to rupture was 11,000), (2) the circumferentially oriented defect would result in a rupture before leak. The number of cycles to failure would be in excess of 14,000. Based on the conservative assumption that the thermal stresses were membrane stresses as opposed to a bending stress, the following were found. For the Inconel 82 weld metal (thickness of 1.3 cm [0.53 in.]) and AISI 316 SST valve body, with an initial assumed defect of 0.25 cm (0.1 in.), the crack would grow through the thickness in a minimum of 3950 cycles and to a critical rupture crack length of 5.1 cm (2.0 in.) in an additional 80 cycles. The Inconel 82 weld metal at the shell body (thickness of 9.7 cm or 3.8 in.) with an assumed defect 1.3 cm (0.5 in.) deep would fail in 334 cycles. Calculations made assuming a linear stress gradient instead of the above-mentioned flat distribution through the wall indicated that the number of stress cycles increased to 2200

Study of thermal-hydraulic characteristics in an LMFBR intermediate plenum

International Nuclear Information System (INIS)

Uotani, M.; Naohara, N.; Kinoshita, I.

1985-01-01

Experimental studies using water and liquid metal were conducted in order to investigate the thermal-hydraulic characteristics of an LMFBR intermediate plenum. The present study is an attempt to evaluate the effect of natural convection on the temperature field and to validate the prediction method of temperature profile in a thermally stratified cavity. The experimental results indicated that the effect of the natural convection on flow velocity and heat transfer in the cavity is reduced with increasing the modified stratification parameter. The calculation by FEM code and a simple 1-D model are effective to predict the temperature profile in the cavity

PIV Experiments to Measure Flow Phenomena in a Scaled Model of a VHTR Lower Plenum

Energy Technology Data Exchange (ETDEWEB)

Hugh M. McIlroy, Jr.; Donald M. McEligot; Richard R. Schultz; Daniel Christensen; Robert J. Pink; Ryan C. Johnson

2006-09-01

A report of experimental data collected at the Matched-Index-of-Refraction (MIR) Laboratory in support of contract DE-AC07-05ID14517 and the INL Standard Problem on measurements of flow phenomena occurring in a lower plenum of a typical prismatic VHTR concept reactor to assess CFD code is presented. Background on the experimental setup and procedures is provided along with several samples of data obtained from the 3-D PIV system and an assessment of experimental uncertainty is provided. Data collected in this study include 3-dimensional velocity-field descriptions of the flow in all four inlet jets and the entire lower plenum with inlet jet Reynolds numbers (ReJet) of approximately 4300 and 12,400. These investigations have generated over 2 terabytes of data that has been processed to describe the various velocity components in formats suitable for external release and archived on removable hard disks. The processed data from both experimental studies are available in multi-column text format.

Investigation of the coolability of a continuous mass of relocated debris to a water-filled lower plenum. Technical report

International Nuclear Information System (INIS)

Rempe, J.L.; Wolf, J.R.; Chavez, S.A.; Condie, K.G.; Hagrman, D.L.; Carmack, W.J.

1994-09-01

This report documents work performed to support the development of an analytical and experimental program to investigate the coolability of a continuous mass of debris that relocates to a water-filled lower plenum. The objective of this program is to provide an adequate data base for developing and validating a model to predict the coolability of a continuous mass of debris relocating to a water-filled lower plenum. The model must address higher pressure scenarios, such as the TMI-2 accident, and lower pressure scenarios, which recent calculations indicate are more likely for most operating LWR plants. The model must also address a range of possible debris compositions

Validation of RETRAN-03 by simulating a peach bottom turbine trip and boiloff at the full integral simulation test facility

International Nuclear Information System (INIS)

Westacott, J.L.; Peterson, C.E.

1992-01-01

This paper reports that the RETRAN-03 computer code is validated by simulating two tests that were performed at the Full Integral Simulation Test (FIST) facility. The RETRAN-03 results of a turbine trip (test 4PTT1) and failure to maintain water level at decay power (test T1QUV) are compared with the FIST test data. The RETRAN-03 analysis of test 4PTT1 is compared with a previous TRAC-BWR analysis of the test. Sensitivity to various model nodalizations and RETRAN-03 slip options are studied by comparing results of test T1QUV. The predicted thermal-hydraulic responses of both tests agree well with the test data. The pressure response of test 4PTT1 and the boiloff rate for test T1QUV are accurately predicted. Core uncovery time is found to be sensitive to the upper downcomer and upper plenum nodalization. The RETRAN-03 algebraic and dynamic slip options produce similar results for test T1QUV

UPTF loop seal tests and their RELAP simulation

International Nuclear Information System (INIS)

Tuomainen, M.; Tuunanen, J.

1997-01-01

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

Post-test analysis of semiscale tests S-UT-6 and S-UT-7 using TRAC PF1

International Nuclear Information System (INIS)

Boyack, B.E.

1983-01-01

A posttest study of Semiscale Tests S-UT-6 and S-UT-7 has been completed to assess TRAC-PFl predictions of pressurized water-reactor (PWR) small-break transients. The comparisons of the TRAC calculations and experimental results show that the correct qualitative influence of upper-head injection (UHI) was predicted. The major phenomenological difference predicted was the mode of core voiding. The data show a slow boiloff from the top of the core resulting in a dryout near the top of the core only. TRAC predicted a more extensive voiding with fluid forced from the bottom of the core by a pressure increase in the upper vessel plenum. The pressure increase was the primary consequence of a failure to predict a complete clearance of the seal in the intact-loop pump-suction upflow leg. Further review of the interphasic drag correlations, entrainment correlations, and critical-flow model is recommended. 20 figures

Components inspection of Monju, a sodium bonded type control rod

International Nuclear Information System (INIS)

Harada, Kiyoshi; Matsushita, Yuichi; Lee, Chunchan; Abe, Hideaki; Watahiki, Naohisa

2002-03-01

This Report addresses a result of a sodium test conducted on components of a Double Poral Filter Sodium Bonded Type Control Rod that is expected to be a next generation, long life Control Rod. Upper and lower Poral Filter Sodium Bonded Type Control Rod components were mocked up to conduct a sodium test. During the test, sodium chargeability, formation of Gas Plenum at the upper part of the components, sodium drain-ability and NaOH clean-ability were recognized under actual plant condition. The following are results obtained: (1) Sodium Chargeability at Control Rod Insertion to EVST. Sodium was charged into the components when the mocked-up was inserted in sodium of 190degC, with insertion speed of 6 m/min which is an actual insertion speed to EVST. (2) Formation of Upper Gas Plenum by Helium Gas generated in Control Rod Components Gas Plenum formation within deviation of 9% was confirmed by releasing helium gas into the mocked-up which is immersed in sodium of 620degC and 190degC. Length of Gas Plenum is confirmed to be retained in certain length even if helium gas is further released into formed Gas Plenum. (3) Sodium Drain-ability of Control Rod Components when Drawing from EVST. Drain-ability was confirmed to be sufficient and no sodium residue was found in the mocked-up when the mocked-up was drawn out from sodium of 190degC, with drawing speed of 6 m/min which is an actual drawing speed from EVST. (4) Clean-ability of NaOH Solution against Sodium Residue in Control Rod Components. Sodium and NaOH solution reacted calmly, however, clean-ability was not sufficient. When Sodium fully remained in Control Rod Components, it made circulation of NaOH solution not enough. (author)

Summary report of NEPTUN investigations into transient thermal hydraulics of the passive decay heat removal

International Nuclear Information System (INIS)

Weinberg, D.; Hoffmann, H.; Rust, K.; Frey, H.H.; Hain, K.; Leiling, W.; Hayafune, H.

1995-12-01

The results corroborate the findings of tests with the RAMONA model. With the core power reduction at scram and the start of the decay heat exchangers operation cold fluid is delivered into the prevailing upper plenum. A temperature stratification develops with distinct large temperature gradients. The onset of natural convection is mainly influenced by two effects, namely, the temperature increase on the intermediate heat exchangers primary sides as a result of which the downward pressures are reduced, and the startup of the decay heat exchangers which leads to a decrease of the buoyancy forces in the core. The temperatures of the upper plenum are systematically reduced as soon as the decay heat exchangers are in operation. Then mixed fluid in the hot plenum reaches the intermediate heat exchangers inlet windows and causes an increase in the core flow rate. The primary pump coastdown curve influences the primary system thermal hydraulics only during the first thousand seconds after scram. The longer the pumps operate the more cold fluid is delivered via the core to the upper plenum. The delay of the start of the decay heat exchangers operation separates the two effects which influence the core mass flow, namely the heatup of the intermediate heat exchangers as well as the formation of the stratification in the upper plenum. Increasing the power as well as the operation of only half of the available decay heat exchangers increase the system temperatures. A permeable above core structure produces a temperature stratification along the total upper plenum, and therefore a lower temperature gradient in the region between core outlet and lower edge of the above core structure, in comparison to the impermeable design. A complete flow path blockage of the primary fluid through the intermediate heat exchangers leads to an enhanced cooling effect of the interstitial flow and gives rise to a thermosiphon effect inside the core elements. (orig./GL) [de

Thermophysical, hydrodynamic and mechanical aspects of molten core relocation to lower plenum

Energy Technology Data Exchange (ETDEWEB)

Suh, Kune Y; Huh, Chang Wook [Seoul National University, Seoul (Korea, Republic of)

1998-12-31

This paper presents the current state of knowledge on molten material relocation into the lower plenum. Consequences of movement of material to the lower head are considered with regard to the potential for reactor pressure vessel failure from both thermal hydraulic and mechanical standpoints. The models are applied to evaluating various in-vessel retention strategies for the Korean Standard Power Plant (KSNPP) reactor. The results are summarized in terms of thermal response of the reactor vessel from the very relevant severe accident management perspective. 10 refs., 1 fig., 1 tab. (Author)

Thermophysical, hydrodynamic and mechanical aspects of molten core relocation to lower plenum

Energy Technology Data Exchange (ETDEWEB)

Suh, Kune Y.; Huh, Chang Wook [Seoul National University, Seoul (Korea, Republic of)

1997-12-31

This paper presents the current state of knowledge on molten material relocation into the lower plenum. Consequences of movement of material to the lower head are considered with regard to the potential for reactor pressure vessel failure from both thermal hydraulic and mechanical standpoints. The models are applied to evaluating various in-vessel retention strategies for the Korean Standard Power Plant (KSNPP) reactor. The results are summarized in terms of thermal response of the reactor vessel from the very relevant severe accident management perspective. 10 refs., 1 fig., 1 tab. (Author)

UPTF test instrumentation. Measurement system identification, engineering units and computed parameters

International Nuclear Information System (INIS)

Sarkar, J.; Liebert, J.; Laeufer, R.

1992-11-01

This updated version of the previous report /1/ contains, besides additional instrumentation needed for 2D/3D Programme, the supplementary instrumentation in the inlet plenum of SG simulator and hot and cold leg of broken loop, the cold leg of intact loops and the upper plenum to meet the requirements (Test Phase A) of the UPTF Programme, TRAM, sponsored by the Federal Minister of Research and Technology (BMFT) of the Federal Republic of Germany. For understanding, the derivation and the description of the identification codes for the entire conventional and advanced measurement systems classifying the function, and the equipment unit, key, as adopted in the conventional power plants, have been included. Amendments have also been made to the appendices. In particular, the list of measurement systems covering the measurement identification code, instrument, measured quantity, measuring range, band width, uncertainty and sensor location has been updated and extended to include the supplementary instrumentation. Beyond these amendments, the uncertainties of measurements have been precisely specified. The measurement identification codes which also stand for the identification of the corresponding measured quantities in engineering units and the identification codes derived therefrom for the computed parameters have been adequately detailed. (orig.)

Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Progress report, December 1, 1975--February 29, 1976

International Nuclear Information System (INIS)

Todreas, N.E.; Golay, M.W.

1976-01-01

Progress is summarized in the following task areas: assessment of available data, experimental water mixing investigations, analytic model development, and analytical and experimental investigation of velocity and temperature fields in outlet plenum flow mixing

Reactor safety issues resolved by the 2D/3D Program

International Nuclear Information System (INIS)

Damerell, P.S.; Simons, J.W.

1993-07-01

The 2D/3D Program studied multidimensional thermal-hydraulics in a PWR core and primary system during the end-of-blowdown and post-blowdown phases of a large-break LOCA (LBLOCA), and during selected small-break LOCA (SBLOCA) transients. The program included tests at the Cylindrical Core Test Facility (CCTF), the Slab Core Test Facility (SCTF), and the Upper Plenum Test Facility (UPTF), and computer analyses using TRAC. Tests at CCTF investigated core thermal-hydraulics and overall system behavior while tests at SCTF concentrated on multidimensional core thermal-hydraulics. The UPTF tests investigated two-phase flow behavior in the downcomer, upper plenum, tie plate region, and primary loops. TRAC analyses evaluated thermal-hydraulic behavior throughout the primary system in tests as well as in PWRs. This report summarizes the test and analysis results in each of the main areas where improved information was obtained in the 2D/3D Program. The discussion is organized in terms of the reactor safety issues investigated

Prediction of corium debris characteristics in lower plenum of a nordic BWR in different accident scenarios using MELCOR code - 15367

International Nuclear Information System (INIS)

Phung, V.A.; Galushin, S.; Raub, S.; Goronovski, A.; Villanueva, W.; Koeoep, K; Grishchenko, D.; Kudinov, P.

2015-01-01

Severe accident management strategy in Nordic boiling water reactors (BWRs) relies on ex-vessel core debris coolability. The mode of corium melt release from the vessel determines conditions for ex-vessel accident progression and threats to containment integrity, e.g., formation of a non-coolable debris bed and possibility of energetic steam explosion. In-vessel core degradation and relocation is an important stage which determines characteristics of corium debris in the vessel lower plenum, such as mass, composition, thermal properties, timing of relocation, and decay heat. These properties affect debris reheating and remelting, melt interactions with the vessel structures, and possibly vessel failure and melt ejection mode. Core degradation and relocation is contingent upon the accident scenario parameters such as recovery time and capacity of safety systems. The goal of this work is to obtain a better understanding of the impact of the accident scenarios and timing of the events on core relocation phenomena and resulting properties of the debris bed in the vessel lower plenum of Nordic BWRs. In this study, severe accidents in a Nordic BWR reference plant are initiated by a station black out event, which is the main contributor to core damage frequency of the reactor. The work focuses on identifying ranges of debris bed characteristics in the lower plenum as functions of the accident scenario with different recovery timing and capacity of safety systems. The severe accident analysis code MELCOR coupled with GA-IDPSA is used in this work. GA-IDPSA is a Genetic Algorithm-based Integrated Deterministic Probabilistic Safety Analysis tool, which has been developed to search uncertain input parameter space. The search is guided by different target functions. Scenario grouping and clustering approach is applied in order to estimate the ranges of debris characteristics and identify scenario regions of core relocation that can lead to significantly different debris bed

Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Progress report, March 1, 1977--May 31, 1977

International Nuclear Information System (INIS)

Todreas, N.E.; Golay, M.W.; Wolf, L.

1977-01-01

Progress is summarized in the following tasks: (1) bundle flow studies (wrapped and bare rods); (2) subchannel flow studies (bare rods); (3) LMFBR outlet plenum flow mixing; and (4) theoretical determination of local temperature fields in LMFBR fuel rod bundles

TRIO a general computer code for reactor 3-D flows analysis. Application to a LMFBR hot plenum

International Nuclear Information System (INIS)

Magnaud, J.P.; Rouzaud, P.

1985-09-01

TRIO is a code developed at CEA to investigate general incompressible 2D and 3D viscous flows. Two calculations are presented: the lid driven cubic cavity at Re=400; steady state (velocity and temperature field) of a LMFBR hot plenum, carried out in order to prepare the calculation of a cold shock consecutive to a reactor scram. 8 refs., 26 figs.

Flow model study of 'Monju' reactor vessel

International Nuclear Information System (INIS)

Miyaguchi, Kimihide

1980-01-01

In the case of designing the structures in nuclear reactors, various problems to be considered regarding thermo-hydrodynamics exist, such as the distribution of flow quantity and the pressure loss in reactors and the thermal shock to inlet and outlet nozzles. In order to grasp the flow characteristics of coolant in reactors, the 1/2 scale model of the reactor structure of ''Monju'' was attached to the water flow testing facility in the Oarai Engineering Center, and the simulation experiment has been carried out. The flow characteristics in reactors clarified by experiment and analysis so far are the distribution of flow quantity between high and low pressure regions in reactors, the distribution of flow quantity among flow zones in respective regions of high and low pressure, the pressure loss in respective parts in reactors, the flow pattern and the mixing effect of coolant in upper and lower plenums, the effect of the twisting angle of inlet nozzles on the flow characteristics in lower plenums, the effect of internal cylinders on the flow characteristics in upper plenums and so on. On the basis of these test results, the improvement of the design of structures in reactors was made, and the confirmation test on the improved structures was carried out. The testing method, the calculation method, the test results and the reflection to the design of actual machines are described. (Kako, I.)

A Field Test for Upper Body Strength and Endurance.

Science.gov (United States)

Nelson, Jack K.; And Others

1991-01-01

Researchers studied the reliability of the modified push-up test in measuring upper body strength and endurance in elementary through college students. It also examined the accuracy of partner scoring. The test proved much easier to administer than the regular floor push-up. It was valid and reliable for all students and suitable for partner…

Thermal stratification of sodium in the BN 600 reactor

International Nuclear Information System (INIS)

Obmelukhin, J.A.; Obukhov, P.I.; Rinejskij, A.A.; Sobolev, V.A.; Sherbakov, S.I.

1983-01-01

The signs of thermal stratification of sodium in the BN 600 reactor upper plenum revealed by the analysis of standard temperature sensors' readings are defined. The initial conditions for existence of different temperature sodium layers are given. Two approaches for realizing on a computer of equations describing sodium motion in the upper plenum of the reactor are presented. (author)

Development and testing of new upper-limb prosthetic devices: research designs for usability testing.

Science.gov (United States)

Resnik, Linda

2011-01-01

The purposes of this article are to describe usability testing and introduce designs and methods of usability testing research as it relates to upper-limb prosthetics. This article defines usability, describes usability research, discusses research approaches to and designs for usability testing, and highlights a variety of methodological considerations, including sampling, sample size requirements, and usability metrics. Usability testing is compared with other types of study designs used in prosthetic research.

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-05-15

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-15

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

Experimental study on breakup and fragmentation behavior of molten material jet in complicated structure of BWR lower plenum

International Nuclear Information System (INIS)

Saito, Ryusuke; Abe, Yutaka; Yoshida, Hiroyuki

2014-01-01

To estimate the state of reactor pressure vessel of Fukushima Daiichi nuclear power plant, it is important to clarify the breakup and fragmentation of molten material jet in the lower plenum of boiling water reactor (BWR) by a numerical simulation. To clarify the effects of complicated structures on the jet behavior experimentally and validate the simulation code, we conduct the visualized experiments simulating the severe accident in the BWR lower plenum. In this study, jet breakup, fragmentation and surrounding velocity profiles of the jet were observed by the backlight method and the particle image velocimetry (PIV) method. From experimental results using the backlight method, it was clarified that jet tip velocity depends on the conditions whether complicated structures exist or not and also clarified that the structures prevent the core of the jet from expanding. From measurements by the PIV method, the surrounding velocity profiles of the jet in the complicated structures were relatively larger than the condition without structure. Finally, fragment diameters measured in the present study well agree with the theory suggested by Kataoka and Ishii by changing the coefficient term. Thus, it was suggested that the fragmentation mechanism was mainly controlled by shearing stress. (author)

Reactor safety issues resolved by the 2D/3D program

International Nuclear Information System (INIS)

1995-09-01

The 2D/3D Program studied multidimensional thermal-hydraulics in a PWR core and primary system during the end-of-blowdown and post-blowdown phases of a large-break LOCA (LBLOCA), and during selected small-break LOCA (SBLOCA) transients. The program included tests at the Cylindrical Core Test Facility (CCTF), the Slab Core Test Facility (SCTF), and the Upper Plenum Test Facility (UPTF), and computer analyses using TRAC. Tests at CCTF investigated core thermal-hydraulics and overall system behavior while tests at SCTF concentrated on multidimensional core thermal-hydraulics. The UPTF tests investigated two-phase flow behavior in the downcomer, upper plenum, tie plate region, and primary loops. TRAC analyses evaluated thermal-hydraulic behavior throughout the primary system in tests as well as in PWRs. This report summarizes the test and analysis results in each of the main areas where improved information was obtained in the 2D/3D Program. The discussion is organized in terms of the reactor safety issues investigated. This report was prepared in a coordination among US, Germany and Japan. US and Germany have published the report as NUREG/IA-0127 and GRS-101 respectively. (author)

Reactor safety issues resolved by the 2D/3D program

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-09-01

The 2D/3D Program studied multidimensional thermal-hydraulics in a PWR core and primary system during the end-of-blowdown and post-blowdown phases of a large-break LOCA (LBLOCA), and during selected small-break LOCA (SBLOCA) transients. The program included tests at the Cylindrical Core Test Facility (CCTF), the Slab Core Test Facility (SCTF), and the Upper Plenum Test Facility (UPTF), and computer analyses using TRAC. Tests at CCTF investigated core thermal-hydraulics and overall system behavior while tests at SCTF concentrated on multidimensional core thermal-hydraulics. The UPTF tests investigated two-phase flow behavior in the downcomer, upper plenum, tie plate region, and primary loops. TRAC analyses evaluated thermal-hydraulic behavior throughout the primary system in tests as well as in PWRs. This report summarizes the test and analysis results in each of the main areas where improved information was obtained in the 2D/3D Program. The discussion is organized in terms of the reactor safety issues investigated. This report was prepared in a coordination among US, Germany and Japan. US and Germany have published the report as NUREG/IA-0127 and GRS-101 respectively. (author).

IDAHO NATIONAL LABORATORY PROGRAM TO OBTAIN BENCHMARK DATA ON THE FLOW PHENOMENA IN A SCALED MODEL OF A PRISMATIC GAS-COOLED REACTOR LOWER PLENUM FOR THE VALIDATION OF CFD CODES

International Nuclear Information System (INIS)

Hugh M. McIlroy Jr.; Donald M. McEligot; Robert J. Pink

2008-01-01

The experimental program that is being conducted at the Matched Index-of-Refraction (MIR) Flow Facility at Idaho National Laboratory (INL) to obtain benchmark data on measurements of flow phenomena in a scaled model of a typical prismatic gas-cooled (GCR) reactor lower plenum using 3-D Particle Image Velocimetry (PIV) is presented. A detailed description of the model, scaling, the experimental facility, 3-D PIV system, measurement uncertainties and analysis, experimental procedures and samples of the data sets that have been obtained are included. Samples of the data set that are presented include mean-velocity-field and turbulence data in an approximately 1:7 scale model of a region of the lower plenum of a typical prismatic GCR design. This experiment has been selected as the first Standard Problem endorsed by the Generation IV International Forum. Results concentrate on the region of the lower plenum near its far reflector wall (away from the outlet duct). Inlet jet Reynolds numbers (based on the jet diameter and the time-mean average flow rate) are approximately 4,300 and 12,400. The measurements reveal undeveloped, non-uniform flow in the inlet jets and complicated flow patterns in the model lower plenum. Data include three-dimensional vector plots, data displays along the coordinate planes (slices) and charts that describe the component flows at specific regions in the model. Information on inlet flow is also presented

International Standard problem ISP 14: behaviour of a fuel bundle simulator during a specified heatup and flooding period (Rebeka experiment): results of post-test analyses: final comparison report

International Nuclear Information System (INIS)

Karwat, H.

1985-02-01

The test consisted in investigating the non-steady material behaviour of a bundle of electrically heated fuel rod simulators with respect to local fuel temperatures, cladding strain, time to burst and local strain at location of burst, together with the thermal hydraulic boundary conditions. The original aim has not been fully achievable. The applied codes for mechanical fuel behaviour largely demonstrated their capabilities for pretest predictions when certain local fluid dynamic parameters are well known to the code users. The difficulties expected with proper analysis of thermal hydraulics of the test were confirmed, caused by the coupling between pin cooling conditions, rod upper plenum calculations and the feedback to clad deformation and burst simulation

Test Specifications and the Design of the Wire Wrapped 37-Pin Fuel Assembly for Hydrodynamic Experiments

International Nuclear Information System (INIS)

Chang, S. K.; Euh, D. J.; Bae, H.; Lee, H. Y.; Choi, S. R.

2013-01-01

Most influencing parameters on uncertainties and sensitivities of the CFD analyses are the friction coefficient and the mixing coefficient. The friction coefficient is related to the flow distribution in reactor sub-channels. The mixing coefficient is defined with the cross flow between neighboring sub-channels. The eventual purpose of the thermal hydraulic design considering these parameters is to guarantee the fuel cladding integrity as the design limit parameter. At the moment, the experimental program is being undertaken to quantify these friction and mixing parameters which characterize the flow distribution in sub-channels, and the wire wrapped 37-pin rod assembly and its hexagonal test rig have been designed and fabricated. The quantified thermal hydraulic experimental data from this program are utilized primarily to estimate the accuracy of the safety analysis codes and their thermal hydraulic model. A wire wrapped 37 pin fuel assembly has been designed for the measurements of the flow distribution, where the measurements are utilized to quantify the friction coefficient and the mixing coefficient. The test rig of the wire wrapped 37 pin fuel assembly has been fabricated considering the geometric and flow dynamic similarities. It comprises four components i. e., the upper plenum, the fuel housing, the lower plenum, and the wire wrapped 37 pin fuel assembly. At further works, the quantified friction and mixing coefficients through the experiments are going to be utilized for insuring the reliability of the CFD analysis results

Design of the upper internals structure for the Clinch River Breeder Reactor Plant

International Nuclear Information System (INIS)

Thompson, D.C.; Novendstern, E.H.

1977-01-01

The Upper Internals Structure (UIS) is located above the core and is supported from the head at four locations. It is designed to perform the following primary functions: provide secondary core holddown in the event of a malfunction of the core hydraulic holddown system; provide support for routing all in-vessel instrumentation to core assemblies; maintain alignment between the core assemblies, the UIS and the closure head; provide guidance and crossflow protection for the control rod drivelines; and mix/duct flow to the upper region of the vessel outlet plenum to minimize rapid temperature changes to components during a reactor trip transient. In accomplishing these functions, the UIS will experience a sodium environment with temperatures up to 1200 0 F (649 0 C), and as many as 7 x 10 8 cycles of fluid temperature fluctuations up to 250 0 F (121 0 C) at full power operation. It must be designed to survive these conditions in combination with seismic and flow-induced vibration loadings for its 30 year design life. The design program of designing to controlled functional requirements and design conditions is discussed. Included is a description of the significant parts of the design and the approach used to balance the requirement of tight joints. The thermal and hydraulic environment including the results of a comprehensive test program are discussed. The test program results establish the basis of the thermal boundary used in the structural evaluation, and the UIS vibration characteristics. A summary of the areas which have required design changes is included with a summary of the structural evaluation of these changes

CAN UPPER EXTREMITY FUNCTIONAL TESTS PREDICT THE SOFTBALL THROW FOR DISTANCE: A PREDICTIVE VALIDITY INVESTIGATION

Science.gov (United States)

Hanney, William J.; Kolber, Morey J.; Davies, George J.; Riemann, Bryan

2011-01-01

Introduction: Understanding the relationships between performance tests and sport activity is important to the rehabilitation specialist. The purpose of this study was two- fold: 1) To identify if relationships exist between tests of upper body strength and power (Single Arm Seated Shot Put, Timed Push-Up, Timed Modified Pull-Up, and The Davies Closed Kinetic Chain Upper Extremity Stability Test, and the softball throw for distance), 2) To determine which variable or group of variables best predicts the performance of a sport specific task (the softball throw for distance). Methods: One hundred eighty subjects (111 females and 69 males, aged 18-45 years) performed the 5 upper extremity tests. The Pearson product moment correlation and a stepwise regression were used to determine whether relationships existed between performance on the tests and which upper extremity test result best explained the performance on the softball throw for distance. Results: There were significant correlations (r=.33 to r=.70, p=0.001) between performance on all of the tests. The modified pull-up test was the best predictor of the performance on the softball throw for distance (r2= 48.7), explaining 48.7% of variation in performance. When weight, height, and age were added to the regression equation the r2 values increased to 64.5, 66.2, and 67.5 respectively. Conclusion: The results of this study indicate that several upper extremity tests demonstrate significant relationships with one another and with the softball throw for distance. The modified pull up test was the best predictor of performance on the softball throw for distance. PMID:21712942

The push-off test: development of a simple, reliable test of upper extremity weight-bearing capability.

Science.gov (United States)

Vincent, Joshua I; MacDermid, Joy C; Michlovitz, Susan L; Rafuse, Richard; Wells-Rowsell, Christina; Wong, Owen; Bisbee, Leslie

2014-01-01

Longitudinal clinical measurement study. The push-off test (POT) is a novel and simple measure of upper extremity weight-bearing that can be measured with a grip dynamometer. There are no published studies on the validity and reliability of the POT. The relationship between upper extremity self-report activity/participation and impairment measures remain an unexplored realm. The primary purpose of this study is to estimate the intra and inter-rater reliability and construct validity of the POT. The secondary purpose is to estimate the relationship between upper extremity self-report activity/participation questionnaires and impairment measures. A convenience sample of 22 patients with wrist or elbow injuries were tested for POT, wrist/elbow range of motion (ROM), isometric wrist extension strength (WES) and grip strength; and completed two self-report activity/participation questionnaires: Disability of the Arm, Shoulder and the Hand (DASH) and Work Limitations Questionnaire (WLQ-26). POT's inter and intra-rater reliability and construct validity was tested. Pearson's correlations were run between the impairment measures and self-report questionnaires to look into the relationship amongst them. The POT demonstrated high inter-rater reliability (ICC affected = 0.97; 95% C.I. 0.93-0.99; ICC unaffected = 0.85; 95% C.I. 0.68-0.94) and intra-rater reliability (ICC affected = 0.96; 95% C.I. 0.92-0.97; ICC unaffected = 0.92; 95% C.I. 0.85-0.97). The POT was correlated moderately with the DASH (r = -0.47; p = 0.03). While examining the relationship between upper extremity self-reported activity/participation questionnaires and impairment measures the strongest correlation was between the DASH and the POT (r = -0.47; p = 0.03) and none of the correlations with the other physical impairment measures reached significance. At-work disability demonstrated insignificant correlations with physical impairments. The POT test provides a reliable and easily

Testing of Selective Laser Melting Turbomachinery Applicable to Exploration Upper Stage

Science.gov (United States)

Calvert, Marty; Turpin, Jason; Nettles, Mindy

2015-01-01

This task is to design, fabricate, and spin test to failure a Ti6-4 hydrogen turbopump impeller that was built using the selective laser melting (SLM) fabrication process (fig. 1). The impeller is sized around upper stage engine requirements. In addition to the spin burst test, material testing will be performed on coupons that are built with the impeller.

INTERSESSION RELIABILITY OF UPPER EXTREMITY ISOKINETIC PUSH-PULL TESTING.

Science.gov (United States)

Riemann, Bryan L; Davis, Sarah E; Huet, Kevin; Davies, George J

2016-02-01

Based on the frequency pushing and pulling patterns are used in functional activities, there is a need to establish an objective method of quantifying the muscle performance characteristics associated with these motions, particularly during the later stages of rehabilitation as criteria for discharge. While isokinetic assessment offers an approach to quantifying muscle performance, little is known about closed kinetic chain (CKC) isokinetic testing of the upper extremity (UE). To determine the intersession reliability of isokinetic upper extremity measurement of pushing and pulling peak force and average power at slow (0.24 m/s), medium (0.43 m/s) and fast (0.61 m/s) velocities in healthy young adults. The secondary purpose was to compare pushing and pulling peak force (PF) and average power (AP) between the upper extremity limbs (dominant, non-dominant) across the three velocities. Twenty-four physically active men and women completed a test-retest (>96 hours) protocol in order to establish isokinetic UE CKC reliability of PF and AP during five maximal push and pull repetitions at three velocities. Both limb and speed orders were randomized between subjects. High test-retest relative reliability using intraclass correlation coefficients (ICC2, 1) were revealed for PF (.91-.97) and AP (.85-.95) across velocities, limbs and directions. PF typical error (% coefficient of variation) ranged from 6.1% to 11.3% while AP ranged from 9.9% to 26.7%. PF decreased significantly (p pushing were significantly greater than pulling at all velocities, however the push-pull differences in PF became less as velocity increased. There were no significant differences identified between the dominant and nondominant limbs. Isokinetically derived UE CKC push-pull PF and AP are reliable measures. The lack of limb differences in healthy normal participants suggests that clinicians can consider bilateral comparisons when interpreting test performance. The increase in pushing PF and

Design of pellet surface grooves for fission gas plenum

International Nuclear Information System (INIS)

Carter, T.J.; Jones, L.R.; Macici, N.; Miller, G.C.

1986-01-01

In the Canada deuterium uranium pressurized heavy water reactor, short (50-cm) Zircaloy-4 clad bundles are fueled on-power. Although internal void volume within the fuel rods is adequate for the present once-through natural uranium cycle, the authors have investigated methods for increasing the internal gas storage volume needed in high-power, high-burnup, experimental ceramic fuels. This present work sought to prove the methodology for design of gas storage volume within the fuel pellets - specifically the use of grooves pressed or machined into the relatively cool pellet/cladding interface. Preanalysis and design of pellet groove shape and volume was accomplished using the TRUMP heat transfer code. Postirradiation examination (PIE) was used to check the initial design and heat transfer assumptions. Fission gas release was found to be higher for the grooved pellet rods than for the comparison rods with hollow or unmodified pellets. This had been expected from the initial TRUMP thermal analyses. The ELESIM fuel modeling code was used to check in-reactor performance, but some modifications were necessary to accommodate the loss of heat transfer surface to the grooves. It was concluded that for plenum design purposes, circumferential pellet grooves could be adequately modeled by the codes TRUMP and ELESIM

Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Final report

International Nuclear Information System (INIS)

Todreas, N.E.; Cheng, S.K.; Basehore, K.

1984-08-01

This project principally undertook the investigation of the thermal hydraulic performance of wire wrapped fuel bundles of LMFBR configuration. Results obtained included phenomenological models for friction factors, flow split and mixing characteristics; correlations for predicting these characteristics suitable for insertion in design codes; numerical codes for analyzing bundle behavior both of the lumped subchannel and distributed parameter categories and experimental techniques for pressure velocity, flow split, salt conductivity and temperature measurement in water cooled mockups of bundles and subchannels. Flow regimes investigated included laminar, transition and turbulent flow under forced convection and mixed convection conditions. Forced convections conditions were emphasized. Continuing efforts are underway at MIT to complete the investigation of the mixed convection regime initiated here. A number of investigations on outlet plenum behavior were also made. The reports of these investigations are identified

Hydrostratigraphic interpretation of test-hole and geophysical data, Upper Loup River Basin, Nebraska, 2008-10

Science.gov (United States)

Hobza, Christopher M.; Asch, Theodore H.; Bedrosian, Paul A.

2011-01-01

Nebraska's Upper Loup Natural Resources District is currently (2011) participating in the Elkhorn-Loup Model to understand the effect of various groundwater-management scenarios on surface-water resources. During Phase 1 of the Elkhorn-Loup Model, a lack of subsurface geological information in the Upper Loup Natural Resources District, hereafter referred to as the upper Loup study area, was identified as a gap in current knowledge that needed to be addressed. To improve the understanding of the hydrogeology of the upper Loup study area, the U.S. Geological Survey, in cooperation with the Upper Loup Natural Resources District and the University of Nebraska Conservation and Survey Division, collected and described the lithology of drill cuttings from nine test holes, and concurrently collected borehole geophysical data to identify the base of the High Plains aquifer. Surface geophysical data also were collected using time-domain electromagnetic (TDEM) and audio-magnetotelluric (AMT) methods at test-hole locations and between test holes, as a quick, non-invasive means of identifying the base of the High Plains aquifer.

UPTF-TRAM test A3. Turn-over of the hot-leg injected ECC in the steam generator direction

International Nuclear Information System (INIS)

Tenckhoff; Brand, B.; Weiss, P.

1993-06-01

The UPTF TRAM test A3 was a separate effects test to investigate the interaction between the hot leg-injected ECC and the single-phase or two-phase natural circulation in the hot leg in the case of an SBLOCA in a PWR. The experimental investigation of 7 runs was mainly concentrated on the following phenomena: - Transport of hot leg injected ECC water to the upper plenum or in the direction of steam generator, depending on the loop mass flow, -Utilization of the condensation potential of ECC water, - Mixing of the saturated water with the ECC water, - Effect of hot leg injection on the flow phenomena in the hot leg, - Effect of pressure (3 and 15 bar) on the scaling and hence the verification of the scaling concept applied. A preliminary evaluation of the test is presented in the Quick Look Report. (orig.) [de

Studies of flow stratification in the hot plenum of an LMFBR

Energy Technology Data Exchange (ETDEWEB)

Jones, P; Hickmott, S [Central Electricity Generating Board, Berkeley Nuclear Laboratories, Berkeley, Gloucestershire (United Kingdom)

1983-07-01

The paper reviews work at Berkeley Nuclear Laboratories on the extent and effects of buoyancy in the hot plenum of an LMFBR. It summarizes the experimental, theoretical and numerical work has has been conducted to aid the understanding of the complex transient flows which occur following a reactor trip. The experimental work has been conducted in small-scale idealised geometries which isolate the essential features of the reactor flows and is not intended to provide detailed design data. An integral theory has been devised to describe the thermal hydraulics of negatively-buoyant jets. The predictions are shown to be in good agreement with the experimental results and emphasize the need to correctly represent the inlet velocity and temperature profiles. Some preliminary calculations with a transient, two-dimensional, finite-element code are compared with the experimental results. These calculations reproduce the overall features of the flows but not the details of the stratified interface. The development of turbulence models for stratified flows is seen as a fruitful area for further research. (author)

Experiment data report for Semiscale Mod-1 Test S-05-2 (alternate ECC injection test)

International Nuclear Information System (INIS)

Feldman, E.M.; Collins, B.L.; Sackett, K.E.

1977-02-01

Recorded test data are presented for Test S-05-2 of the Semiscale Mod-1 alternate emergency core coolant (ECC) injection test series. This test is one of several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-05-2 was conducted from an initial cold leg fluid temperature of 545 0 F and an initial pressure of 2263 psia. A simulated double-ended offset shear cold leg break was used to investigate core and system response to a depressurization and reflood transient with ECC injection at the intact loop pump suction and broken loop cold leg. A reduced lower plenum volume was used for this test to more accurately represent the lower plenum of a PWR, based on system volume scaling. System flow was set to achieve a core fluid temperature differential of 65 0 F at a core power level of 1.44 MW. The flow resistance of the intact loop was based on core area scaling. An electrically heated core with a slightly peaked radial power profile was used in the pressure vessel to simulate the predicted surface heat flux of nuclear fuel rods during a loss-of-coolant accident

Issues of Exercising the Right to Defence amid the Explanations of the Plenum of the Supreme Court of the Russian Federation

Directory of Open Access Journals (Sweden)

Oksana A. Voltornist

2016-04-01

Full Text Available The article analyzes the explanations of the Plenum of the Supreme Court No. 29 dated June 30, 2015 “On application of laws by the courts ensuring the right to defense in criminal proceedings”. The author details the applied aspects of certain provisions of the aforementioned document within the criminal procedure legislation and estimates their significance for the judicial and investigative practice

Three-dimensional calculation of the flow in the cold plenum of the Fast Breeder Reactor 1500 MW

International Nuclear Information System (INIS)

Chabard, J.P.; Daubert, O.; Gregoire, J.P.; Hemmerich, P.

1987-01-01

To solve thermalhydraulics problems which are rising for example on the various parts of nuclear reactors, several departments of the Direction des Etudes et Recherches are developing the N3S code, three-dimensional code using the finite element method. First, this paper presents the basic equations (Navies-Stokes with turbulence modelling and coupled with the thermal equation) and well suited algorithms to solve them. The industrial adequacy of the code is clearly demonstrated through the application to the computation of the flow in the cold plenum of the Fast Breeder Reactor 1500 MW on a mesh of about 20000 velocity nodes [fr

Validation and Reliability of a Novel Test of Upper Body Isometric Strength.

Science.gov (United States)

Bellar, David; Marcus, Lena; Judge, Lawrence W

2015-09-29

The purpose of the present investigation was to examine the association of a novel test of upper body isometric strength against a 1RM bench press measurement. Forty college age adults (n = 20 female, n = 20 male; age 22.8 ± 2.8 years; body height 171.6 ± 10.8 cm; body mass 73.5 ± 16.3 kg; body fat 23.1 ± 5.4%) volunteered for the present investigation. The participants reported to the lab on three occasions. The first visit included anthropometric measurements and familiarization with both the upper body isometric test and bench press exercise. The final visits were conducted in a randomized order, with one being a 1RM assessment on the bench press and the other consisting of three trials of the upper body isometric assessment. For the isometric test, participants were positioned in a "push-up" style position while tethered (stainless steel chain) to a load cell (high frequency) anchored to the ground. The peak isometric force was consistent across all three trials (ICC = 0.98) suggesting good reliability. Multiple regression analysis was completed with the predictors: peak isometric force, gender, against the outcome variable 1RM bench press. The analysis resulted in a significant model (r2 = 0.861, p≤0.001) with all predictor variables attaining significance in the model (pIsometric peak strength had the greatest effect on the model (Beta = 5.19, p≤0.001). Results from this study suggest that the described isometric upper body strength assessment is likely a valid and reliable tool to determine strength. Further research is warranted to gather a larger pool of data in regard to this assessment.

Cold leg injection reflood test results in the SCTF Core-I under constant system pressure

International Nuclear Information System (INIS)

Adachi, Hiromichi; Iwamura, Takamichi; Sobajima, Makoto; Osakabe, Masahiro; Ohnuki, Akira; Abe, Yutaka; Murao, Yoshio.

1990-08-01

The Slab Core Test Facility (SCTF) was constructed to investigate two-dimensional thermal-hydrodynamics in the core and the interaction in fluid behavior between the core and the upper plenum during the last part of blowdown, refill and reflood phases of a postulated loss-of-coolant accident (LOCA) of a pressurized water reactor (PWR). The present report describes the analytical results on the system behavior observed in the SCTF Core-I cold leg injection tests, S1-14 (Run 520), S1-15 (521), S1-16 (522), S1-17 (523), S1-20 (530), S1-21 (531), S1-23 (536) and S1-24 (537), performed under constant system pressure condition during transient. Major discussion items are: (1) steam binding, (2) U-tube oscillations, (3) bypass of ECC water (4) core cooling behavior, (5) effect of vent valve and (6) other parameter effects. These results give us very useful information and suggestion on reflood behavior. (author)

Evaluation report on CCTF CORE-I REFLOOD TEST Cl-4 (Run 13) and Cl-15 (Run 24)

International Nuclear Information System (INIS)

Sudoh, Takashi; Murao, Yoshio.

1983-08-01

The tests Cl-4 and Cl-15 were performed with the Cylindrical Core Test Facility (CCTF) to investigate the effects of the depressurization process to simulate the refill phase, and the effects of the nitrogen to be injected after the end of the accumulator injection on the thermo-hydraulic behavior in the core and primary loop system during refill and reflood phases. In these tests, after the lower plenum was filled to 0.9m level with saturated water at 0.6 MPa, the accumulator water was injected into three intact cold legs in the depressurization period from 0.6 MPa to 0.2 MPa. The water in the lower plenum voided during the depressurization and the significant steam condensation occurred in or near the intact cold legs. The condensation caused high steam flow rate in the intact loops and the lower plenum flashing resulted in suppressed core water accumulation. The slightly lower core heat transfer coefficient due to the less core water caused the higher turnaround temperature and the longer quench time than those of the normal reflood test without the depressurization process. The nitrogen injection followed the accumulator injection was allowed in the test Cl-15. However, significant effects of the nitrogen injection was not observed. (author)

gamma-Zr-Hydride Precipitate in Irradiated Massive delta- Zr-Hydride

DEFF Research Database (Denmark)

Warren, M. R.; Bhattacharya, D. K.

1975-01-01

During examination of A Zircaloy-2-clad fuel pin, which had been part of a test fuel assembly in a boiling water reactor, several regions of severe internal hydriding were noticed in the upper-plenum end of the pin. Examination of similar fuel pins has shown that hydride of this type is caused by...... to irradiation-induced swelling....

High-Reynolds Number Circulation Control Testing in the National Transonic Facility

Science.gov (United States)

Milholen, William E., II; Jones, Gregory S.; Chan, David T.; Goodliff, Scott L.

2012-01-01

A new capability to test active flow control concepts and propulsion simulations at high Reynolds numbers in the National Transonic Facility at the NASA Langley Research Center is being developed. The first active flow control experiment was completed using the new FAST-MAC semi-span model to study Reynolds number scaling effects for several circulation control concepts. Testing was conducted over a wide range of Mach numbers, up to chord Reynolds numbers of 30 million. The model was equipped with four onboard flow control valves allowing independent control of the circulation control plenums, which were directed over a 15% chord simple-hinged flap. Preliminary analysis of the uncorrected lift data showed that the circulation control increased the low-speed maximum lift coefficient by 33%. At transonic speeds, the circulation control was capable of positively altering the shockwave pattern on the upper wing surface and reducing flow separation. Furthermore, application of the technique to only the outboard portion of the wing demonstrated the feasibility of a pneumatic based roll control capability.

Coolant stratification and its thermohydrodynamic specificity under natural circulation in horizontal steam generator collectors

Energy Technology Data Exchange (ETDEWEB)

Blagovechtchenski, A.; Leontieva, V.; Mitriukhin, A. [Saint-Petersburg Technical Univ. (Russian Federation)

1997-12-31

The experiments and the test facilities for the study of the stratification phenomenon in the hot plenum of reactor and the upper parts of the steam generator collectors in a nuclear power plant are described. The aim of the experiments was to define the conditions of the stratification initiation, to study the temperature field in the upper part, the definition of the characteristics in the stratification layer, and also to study the factors which cause the intensity of the stagnant volume cooling.

Coolant stratification and its thermohydrodynamic specificity under natural circulation in horizontal steam generator collectors

Energy Technology Data Exchange (ETDEWEB)

Blagovechtchenski, A; Leontieva, V; Mitriukhin, A [Saint-Petersburg Technical Univ. (Russian Federation)

1998-12-31

The experiments and the test facilities for the study of the stratification phenomenon in the hot plenum of reactor and the upper parts of the steam generator collectors in a nuclear power plant are described. The aim of the experiments was to define the conditions of the stratification initiation, to study the temperature field in the upper part, the definition of the characteristics in the stratification layer, and also to study the factors which cause the intensity of the stagnant volume cooling.

Validation and Reliability of a Novel Test of Upper Body Isometric Strength

Directory of Open Access Journals (Sweden)

Bellar David

2015-09-01

Full Text Available The purpose of the present investigation was to examine the association of a novel test of upper body isometric strength against a 1RM bench press measurement. Forty college age adults (n = 20 female, n = 20 male; age 22.8 ± 2.8 years; body height 171.6 ± 10.8 cm; body mass 73.5 ± 16.3 kg; body fat 23.1 ± 5.4% volunteered for the present investigation. The participants reported to the lab on three occasions. The first visit included anthropometric measurements and familiarization with both the upper body isometric test and bench press exercise. The final visits were conducted in a randomized order, with one being a 1RM assessment on the bench press and the other consisting of three trials of the upper body isometric assessment. For the isometric test, participants were positioned in a “push-up” style position while tethered (stainless steel chain to a load cell (high frequency anchored to the ground. The peak isometric force was consistent across all three trials (ICC = 0.98 suggesting good reliability. Multiple regression analysis was completed with the predictors: peak isometric force, gender, against the outcome variable 1RM bench press. The analysis resulted in a significant model (r2 = 0.861, p≤0.001 with all predictor variables attaining significance in the model (p<0.05. Isometric peak strength had the greatest effect on the model (Beta = 5.19, p≤0.001. Results from this study suggest that the described isometric upper body strength assessment is likely a valid and reliable tool to determine strength. Further research is warranted to gather a larger pool of data in regard to this assessment.

Validation and Reliability of a Novel Test of Upper Body Isometric Strength

Science.gov (United States)

Bellar, David; Marcus, Lena; Judge, Lawrence W.

2015-01-01

The purpose of the present investigation was to examine the association of a novel test of upper body isometric strength against a 1RM bench press measurement. Forty college age adults (n = 20 female, n = 20 male; age 22.8 ± 2.8 years; body height 171.6 ± 10.8 cm; body mass 73.5 ± 16.3 kg; body fat 23.1 ± 5.4%) volunteered for the present investigation. The participants reported to the lab on three occasions. The first visit included anthropometric measurements and familiarization with both the upper body isometric test and bench press exercise. The final visits were conducted in a randomized order, with one being a 1RM assessment on the bench press and the other consisting of three trials of the upper body isometric assessment. For the isometric test, participants were positioned in a “push-up” style position while tethered (stainless steel chain) to a load cell (high frequency) anchored to the ground. The peak isometric force was consistent across all three trials (ICC = 0.98) suggesting good reliability. Multiple regression analysis was completed with the predictors: peak isometric force, gender, against the outcome variable 1RM bench press. The analysis resulted in a significant model (r2 = 0.861, p≤0.001) with all predictor variables attaining significance in the model (p<0.05). Isometric peak strength had the greatest effect on the model (Beta = 5.19, p≤0.001). Results from this study suggest that the described isometric upper body strength assessment is likely a valid and reliable tool to determine strength. Further research is warranted to gather a larger pool of data in regard to this assessment. PMID:26557203

Study on cooling model for debris in lower plenum and countermeasures for prevention of focusing effect

International Nuclear Information System (INIS)

Guan Zhonghua; Yu Hongxing; Jiang Guangming

2008-01-01

From the basic energy conservation equations and experimental or empirical correlations, an intact model is constructed for the thermal calculation of the core debris in the lower plenum. For verification of this model, the results of two calculations for AP600 and AP1000 plants are compared with those presented in relevant literature. The analysis highlights on the impact of the decay heat power density and the focusing effect. In order to mitigate the focusing effect, it is proposed in this paper to change the lower head profile from hemisphere to parabola. The results show that this change of lower head profile can change the heat flux distribution of the debris, and mitigate the focusing effect. (authors)

Waterhammer modeling for the Ares I Upper Stage Reaction Control System cold flow development test article

Science.gov (United States)

Williams, Jonathan Hunter

The Upper Stage Reaction Control System provides in-flight three-axis attitude control for the Ares I Upper Stage. The system design must accommodate rapid thruster firing to maintain proper launch trajectory and thus allow for the possibility to pulse multiple thrusters simultaneously. Rapid thruster valve closure creates an increase in static pressure, known as waterhammer, which propagates throughout the propellant system at pressures exceeding nominal design values. A series of development tests conducted at Marshall Space Flight Center in 2009 were performed using a water-flow test article to better understand fluid characteristics of the Upper Stage Reaction Control System. A subset of the tests examined the waterhammer pressure and frequency response in the flight-representative system and provided data to anchor numerical models. This thesis presents a comparison of waterhammer test results with numerical model and analytical results. An overview of the flight system, test article, modeling and analysis are also provided.

Waterhammer Modeling for the Ares I Upper Stage Reaction Control System Cold Flow Development Test Article

Science.gov (United States)

Williams, Jonathan H.

2010-01-01

The Upper Stage Reaction Control System provides three-axis attitude control for the Ares I launch vehicle during active Upper Stage flight. The system design must accommodate rapid thruster firing to maintain the proper launch trajectory and thus allow for the possibility to pulse multiple thrusters simultaneously. Rapid thruster valve closure creates an increase in static pressure, known as waterhammer, which propagates throughout the propellant system at pressures exceeding nominal design values. A series of development tests conducted in the fall of 2009 at Marshall Space Flight Center were performed using a water-flow test article to better understand fluid performance characteristics of the Upper Stage Reaction Control System. A subset of the tests examined waterhammer along with the subsequent pressure and frequency response in the flight-representative system and provided data to anchor numerical models. This thesis presents a comparison of waterhammer test results with numerical model and analytical results. An overview of the flight system, test article, modeling and analysis are also provided.

Indicial response test for the support post structure of VHTR

International Nuclear Information System (INIS)

Futakawa, Masatoshi; Kikuchi, Kenji; Tachibana, Katsumi; Muto, Yasushi

1985-11-01

Fuel blocks and removable reflector blocks, which constitute a core of VHTR, are supported by support posts. Each support post is in contact with a hot plenum block at the top end and with a lower plenum block at the bottom end through hemispherical seats to absorb a relative displacement generated by the lateral movement of both blocks by means of small inclination or rotation of support posts. Indicial response tests have been carried out by using a specified one-dimensional vibration model in order to estimate the effects of the support post length, the mass of hot plenum block and the hemispherical radii of both support and post seat on the vibrational characteristics in the support post structure. Futhermore the experimental results have been compared with the analytical ones obtained from the Lagrange's equation. The following are the conclusions derived. (1) The hemispherical radii of support post and post seat have a large effect on the frequency of vibration in the support post structure. (2) The frequency of vibration in the support post structure is predictable using the Lagrange's equation. (author)

Loop-type FBR reactor

International Nuclear Information System (INIS)

Ogura, Kenji; Kimura, Kimitaka; Jinbo, Masaichi; Hirayama, Hiroshi; Taguchi, Junzo; Hirata, Noriaki; Ozaki, Kenji; Maruyama, Shigeki.

1996-01-01

The inside of a vessel of an intermediate heat exchanger is divided vertically by a partition wall into a high temperature plenum region and a low temperature plenum region, a perforated horizontal plate is disposed in a horizontal direction at the upper portion and a flow shroud is disposed so as to surround the upper outside of the intermediate heat exchanger while passing through a lid from a perforated hole of the perforated horizontal plate. In addition, there is disposed a cylinder passing through the partition wall and the horizontal perforated plate for inserting a liquid surface penetrating equipment. The cylinder has an upper end opened above the liquid level of a liquid metal during normal operation and below the liquid level of the liquid metal during shut down of the reactor, and the lower end is opened in a lower plenum region. Vibrations of liquid level due to the high temperature liquid metal inflown from a hot leg pipeline to the inside of the vessel of the intermediate heat exchanger are suppressed by the perforated horizontal plate during reactor operation. On the other hand, upon shut down of the reactor, since the liquid level rises up to the upper portion of the cylinder, the liquid metal at low temperature inflows into the lower plenum region, and the liquid metal at high temperature above the horizontal perforated plate is eliminated in an early stage. (N.H.)

Theoretical study on flow-induced vibration of a cylindrical weir due to fluid discharge

International Nuclear Information System (INIS)

Fujita, Katsuhisa; Ito, Tomohiro; Hirota, Kazuo; Kodama, Tetsuhiko

1994-01-01

In a FBR, the inside of the reactor vessel is cooled by liquid sodium. Liquid sodium is supplied to the upper plenum from its bottom and discharges over the top of the cylindrical weir down to the lower plenum. The weir is so thin in order to decrease the thermal stress on it that the fluid--structure interaction becomes predominant. A fluidelastic vibration of the weir due to fluid discharge was discovered in a French FBR. In this study, a theoretical model was developed on the ''fluid--elastic mode'' instability of a cylindrical weir due to fluid discharge from the upper plenum to the lower plenum. In the analysis, the fluctuation of both the discharge flow rate over a weir due to the vibration of the cylindrical shell and the pressure in the lower plenum due to fluid discharge were formulated. Instability criteria was derived from the added damping ratio due to fluid discharge using modal analysis. The natural modes and modal mass of the weir were obtained by the analysis using the FEM code taking the fluid - structure interaction into consideration. The theoretical instability range in terms of the fall height and the flow rate is compared with the experimental results. The theoretical values showed a good agreement with the experimental ones

Test plan for In Situ Vitrification Engineering-Scale Test No. 6, EG ampersand G Idaho, Inc., Job Number 318230

International Nuclear Information System (INIS)

1991-03-01

The objectives of the test included the effects of in situ vitrification on containerized sludge contained in a simulated randomly-disposed array. From this arrangement, the test results obtained the following data applicable to Idaho National Engineering Laboratory Large Field Testing: canister burst pressure and temperature, canister depressurization rate, melt encapsulation rate of the canister and the hood area plenum temperatures, pressures, compositional analyses, and flows as affected by gas releases. 10 figs., 1 tab

Overview report of RAMONA-NEPTUN program on passive decay heat removal

International Nuclear Information System (INIS)

Weinberg, D.; Rust, K.; Hoffmann, H.

1996-03-01

The design of the advanced sodium-cooled European Fast Reactor provides a safety graded decay heat removal concept which ensures the coolability of the primary system by natural convection when forced cooling is lost. The findings of the RAMONA and NEPTUN experiments indicate that the decay heat can be safely removed by natural convection. The operation of the decay heat exchangers being installed in the upper plenum causes the formation of a thermal stratification associated with a pronounced temperature gradient. The vertical extent of the stratification and the qualitity of the gradient are depending on the fact whether a permeable or an impermeable shell covers the above core structure. A delayed startup time of the decay heat exchangers leads only to a slight increase of the temperatures in the upper plenum. A complete failure of half of the decay heat exchangers causes a higher temperature level in the primary system, but does not alter the global temperature distribution. The transient development of the temperatures is faster going on in a three-loop model than in a four-loop model due to the lower amount of heat stored in the compacter primary vessel. If no coolant reaches the core inlet side via the intermediate heat exchangers, the core remains coolable. In this case, cold water of the upper plenum penetrates into the subassemblies (thermosyphon effects) and the interwrapper spaces existing in the NEPTUN core. The core coolability from above is feasible without any difficulty though the temperatures increase to a minor degree at the top end of the core. The thermal hydraulic computer code FLUTAN was applied for the 3D numerical simulation of the majority of the steady state RAMONA and NEPTUN tests as well as for selected transient RAMONA tests. (orig./HP) [de

Study on natural convection in core barrel. Experimental and numerical results for band type spacer pads

International Nuclear Information System (INIS)

Hayashi, Kenji; Kawamata, Nobuhiro; Kamide, Hideki

2003-03-01

In a fast reactor an Inter-Wrapper Flow (IWF) is one of significant phenomena for decay heat removal under natural circulation condition, when a direct reactor auxiliary cooling system (DRACS) is adopted for decay heat removal system. Cold coolant provided by dipped heat exchangers (DHX) of DRACS can penetrate into the core barrel (region between the subassemblies) and it makes natural convection int he core barrel. Such IWF will depend on a spacer pad geometry of subassemblies. Water experiment, TRIF (Test Rig for Inter-wrapper Flow), was carried out for IWF in a reactor core. The test section modeled a 1/12th sector of the core and upper plenum of reactor vessel. Experimental parameters were the spacer pad geometry and flow path geometries connecting the upper plenum and core barrel. Numerical simulation using AQUA code was also performed to confirm applicability of a simulation method. An experimental series using a button type spacer pad had been carried out. Here a band type spacer pad was examined. Temperatures at subassembly wall were measured with parameter of the flow path geometries; one was a connection pipe between the upper plenum and core barrel and the other was flow hole in core former plates between the outermost subassemblies and the core barrel. It was found that these flow paths were effective to remove heat in the core in case of the band type spacer pad. A general purpose three dimensional analysis code, AQUA, was applied to the experimental analysis. Each subassembly and inter wrapper gap region were modeled by slab mesh geometry. Pressure loss coefficient at the pacer pad was set based on the geometry. The numerical simulation results were in good agreement with measured temperature profiles in the core. (author)

Functional Capacity Evaluation in Upper Limb Reduction Deficiency and Amputation : Development and Pilot Testing

NARCIS (Netherlands)

Postema, S G; Bongers, R M; Reneman, M F; van der Sluis, C K

Purpose To develop and pilot test a functional capacity evaluation (FCE) for individuals with upper limb absence (ULA) due to reduction deficiency or amputation, and to examine the relationship between FCE results and presence of musculoskeletal complaints (MSC). Method Five tests (overhead lifting,

Evaluation report on CCTF Core-II reflood test C2-16 (Run 76)

International Nuclear Information System (INIS)

Iguchi, Tadashi; Akimoto, Hajime; Okubo, Tsutomu; Hojo, Tsuneyuki; Murao, Yoshio; Sugimoto, Jun.

1987-03-01

This report presents the result of the upper plenum injection (UPI) test C2-16 (Run 76), which was conducted on October 23, 1984, with the Cylindrical Core Test Facility (CCTF) at Japan Atomic Energy Research Institute (JAERI). The CCTF is a 1/21.4 scale model of a 1,100 MWe PWR with four loop active components to provide information on the system and core thermo-hydrodynamics during reflood. The objectives of the test are to investigate the reflood phenomena with single failure UPI condition and to investigate the effect of the asymmetry of UPI on the reflood phenomena. The test was performed with an asymmetric UPI condition at the injection rate simulating single failure of LPCI pumps. It was observed that, (1) a UPI test simulating no LPCI pump failure gave the slightly lower peak clad temperature than a UPI test simulating single LPCI pump failure, indicating that single LPCI pump failure assumption is conserrative for UPI condition, and (2) an asymmetric UPI lead to a higher core water accumulation and then a higher heat transfer coefficient, resultantly a lower peak clad temperature than a symmetric UPI, indicating that asymmetric UPI does not lead to a poorer core cooling than symmetric UPI. (author)

Current collector design for closed-plenum polymer electrolyte membrane fuel cells

Science.gov (United States)

Daniels, F. A.; Attingre, C.; Kucernak, A. R.; Brett, D. J. L.

2014-03-01

This work presents a non-isothermal, single-phase, three-dimensional model of the effects of current collector geometry in a 5 cm2 closed-plenum polymer electrolyte membrane (PEM) fuel cell constructed using printed circuit boards (PCBs). Two geometries were considered in this study: parallel slot and circular hole designs. A computational fluid dynamics (CFD) package was used to account for species, momentum, charge and membrane water distribution within the cell for each design. The model shows that the cell can reach high current densities in the range of 0.8 A cm-2-1.2 A cm-2 at 0.45 V for both designs. The results indicate that the transport phenomena are significantly governed by the flow field plate design. A sensitivity analysis on the channel opening ratio shows that the parallel slot design with a 50% opening ratio shows the most promising performance due to better species, heat and charge distribution. Modelling and experimental analysis confirm that flooding inhibits performance, but the risk can be minimised by reducing the relative humidity of the cathode feed to 50%. Moreover, overheating is a potential problem due to the insulating effect of the PCB base layer and as such strategies should be implemented to combat its adverse effects.

Natural circulation under severe accident conditions

International Nuclear Information System (INIS)

Pafford, D.J.; Hanson, D.J.; Tung, V.X.; Chmielewski, S.V.

1992-01-01

Research is being conducted to better understand natural circulation phenomena in mixtures of steam and noncondensibles and its influence on the temperature of the vessel internals and the hot leg, pressurizer surge line, and steam generator tubes. The temperature of these structures is important because their failure prior to reactor vessel lower head failure could reduce the likelihood of containment failure as a result of direct containment heating. Computer code calculations (MELPROG, SCDAP/RELAP5/MOD3) predict high fluid temperatures in the upper plenum resulting from in-vessel natural circulation. Using a simple model for the guide tube phenomena, high upper plenum temperatures are shown to be consistent with the relatively low temperatures that were deduced metallurgically from leadscrews removed from the TMI-2 upper plenum. Evaluation of the capabilities of the RELAP5/MOD3 computer code to predict natural circulation behavior was also performed. The code was used to model the Westinghouse natural circulation experimental facility. Comparisons between code calculations and results from experiments show good agreement

Analysis of large break loss of coolant accident with simultaneous injection into cold leg and hot leg

International Nuclear Information System (INIS)

Luo Bangqi

1997-01-01

When a large break loss of coolant accident occurs, the most part of the safety injection water injected into the cold leg by the safety injection system will flow through the channel between the pressure vessel and the barrel out of the break into the containment, only a little part of the safety injection water can flow into the reactor core. If the safety injection can inject into both the cold leg and the hot leg simultaneously, the safety injection water injected from the cold leg will flow into the core more easily, because the safety injection water injected from the hot leg will carry out more heat from the upper plenum and the core, so the upper plenum and the core is depressed. In addition, a small part of the safety injection water injected from the hot leg will flow down in the core after impinging the guide tubes in the upper plenum, so the core will get more safety injection water than only cold leg injection, and the core will be much safer

Natural circulation in reactor coolant system

International Nuclear Information System (INIS)

Han, J.T.

1987-01-01

Reactor coolant system (RCS) natural circulation in a PWR is the buoyancy-driven coolant circulation between the core and the upper-plenum region (in-vessel circulation) with or without a countercurrent flow in the hot leg piping between the vessel and steam generators (ex-vessel circulation). This kind of multidimensional bouyancy-driven flow circulation serves as a means of transferring the heat from the core to the structures in the upper plenum, hot legs, and possibly steam generators. As a result, the RCS piping and other pressure boundaries may be heated to high temperatures at which the structural integrity is challenged. RCS natural circulation is likely to occur during the core uncovery period of the TMLB' accident in a PWR when the vessel upper plenum and hot leg are already drained and filled with steam and possibly other gaseous species. RCS natural circulation is being studied for the Surry plant during the TMLB' accident in which station blackout coincides with the loss of auxiliary feedwater and no operator actions. The effects of the multidimensional RCS natural circulation during the TMLB' accident are discussed

Evaluation report on CCTF Core-II reflood test C2-4 (Run 62)

International Nuclear Information System (INIS)

Okubo, Tsutomu; Iguchi, Tadashi; Sugimoto, Jun; Akimoto, Hajime; Murao, Yoshio; Okabe, Kazuharu.

1985-03-01

This report presents a data evaluation of the CCTF Core-II test C2-4 (Run 62), which was conducted on May 12, 1983. This test was conducted to investigate the reproducibility of tests in the CCTF Core-II test series. Therefore, the initial and boundary conditions of the present test were determined to be the same as those for the previously performed base case test (Test C2-SH1). Comparing the data of the present test with those of Test C2-SH1, the following results are obtained. (1) The initial and boundary conditions for the two tests were nearly identical except the temperature of the core barrel and the lower plenum fluid. The difference in the latter is considered to result in the difference in the core inlet subcooling of about 6 K at most. (2) The system behavior was almost identical. (3) The core cooling behavior was also nearly identical except a little difference in the rod surface temperature in the upper part of the high power region. (4) Taking account that the difference mentioned above in item (3) is small and can be explained qualitatively to be caused by the difference in the core inlet subcooling mentioned above in item (1), it is considered practically that there is the reproducibility of the thermo-hydrodynamic behavior in the CCTF Core-II tests. (author)

3D modeling of the primary circuit in the reactor pressure vessel of a PHWR

Energy Technology Data Exchange (ETDEWEB)

Ramajo, Damian, E-mail: dramajo@santafe-conicet.gov.ar; Corzo, Santiago; Schiliuk, Nicolas; Nigro, Norberto

2013-12-15

A computational fluid dynamics (CFD) simulation of the reactor pressure vessel (RPV) of the pressurized heavy water reactor (PHWR) of 745 electrical MW Atucha II nuclear power plant was carried out. A three dimensional (3D) detailed model was employed to simulate coolant circuit considering the upper and lower plenums, the downcomer and the hot and cold legs. Control rods and coolant channel tubes at the upper plenum were included to quantify the mixing flow with more realism. The whole set of 451 coolant channels were modeled by means of a zero dimensional methodology. That is, the effect of each coolant channel was modeled through the introduction of a source point at the upper plenum and a sink point at the lower plenum. For each coupled sink/source points (SSP) the mass, momentum and energy balance were solved considering the local pressure difference and the temperature between the corresponding points where sinks and sources were placed. Based on this strategy, three models with increasingly level of approximation were implemented. For the first model the 451 coolant channels were reduced to only 57 pairs of SSP to represent all the coolant channels, concentrating the effect of several coolant channels in a unique pair of sink and source while taking into account geometric design details. For the second model, 225 pairs of SSP were introduced. Finally, for the third model each one of the 451 coolant channels were modeled by means of one pair of SSP. Depending on the coolant channel location, the radial power distribution and the pressure loss caused by the corresponding flow restrictor present by design were considered. Simulations carried out give insight in the complexity of the flow. As expected, the greater the details of the model the better the accuracy reached in the representation of the RPV behavior. In addition, the flow distributor located at the lower plenum showed to be very efficient since, the mass flow at each channel was found to be fairly

Evaluation report on CCTF core-I reflood tests Cl-16 (Run 25), Cl-21 (Run 40) and Cl-22 (Run 41)

International Nuclear Information System (INIS)

Murao, Yoshio; Iguchi, Tadashi; Sudoh, Takashi.

1983-05-01

In order to confirm that the phenomena in the CCTF are consistent with and analogous to those in the other test facilities, three CCTF tests with the experimental conditions simulated the FLECHT-SET TESTS, 3105B, 2714B and 3420B were performed. The downcomer and the upper plenum water accumulations and the pressure drops in the intact loops of the CCTF and the FLECHT-SET were the same, however, the pressure drops in the broken loop and resultant hydrodynamic oscillation in the system and the core thermo-hydrodynamic behaviors were different from each other. It was found that the differences were mainly introduced from the pressure drops at the broken cold leg nozzle in the CCTF, while the pressure drops did not appear in the FLECHT-SET tests because of the different design and operation of the facility. Accordingly, under the consideration of the defferences of the designs and operation methods of the facilities, the phenomena observed in both facilities can be concluded to be analogous with each other

Full-length high-temperature severe fuel damage test No. 5

International Nuclear Information System (INIS)

Lanning, D.D.; Lombardo, N.J.; Hensley, W.K.; Fitzsimmons, D.E.; Panisko, F.E.; Hartwell, J.K.

1993-09-01

This report describes and presents data from a severe fuel damage test that was conducted in the National Research Universal (NRU) reactor at Chalk River Nuclear Laboratories (CRNL), Ontario, Canada. The test, designated FLHT-5, was the fourth in a series of full-length high-temperature (FLHT) tests on light-water reactor fuel. The tests were designed and performed by staff from the US Department of Energy's Pacific Northwest Laboratory (PNL), operated by Battelle Memorial Institute. The test operation and test results are described in this report. The fuel bundle in the FLHT-5 experiment included 10 unirradiated full-length pressurized-water reactor (PWR) rods, 1 irradiated PWR rod and 1 dummy gamma thermometer. The fuel rods were subjected to a very low coolant flow while operating at low fission power. This caused coolant boilaway, rod dryout and overheating to temperatures above 2600 K, severe fuel rod damage, hydrogen generation, and fission product release. The test assembly and its effluent path were extensively instrumented to record temperatures, pressures, flow rates, hydrogen evolution, and fission product release during the boilaway/heatup transient. Post-test gamma scanning of the upper plenum indicated significant iodine and cesium release and deposition. Both stack gas activity and on-line gamma spectrometer data indicated significant (∼50%) release of noble fission gases. Post-test visual examination of one side of the fuel bundle revealed no massive relocation and flow blockage; however, rundown of molten cladding was evident

Air-water tests in support of LLTR series II Test A-4

International Nuclear Information System (INIS)

Chen, K.

1980-07-01

A series of tests injecting air into a tank of stagnant water was conducted in June 1980 utilizing the GE Plenum Mixing Test Facility in San Jose, California. The test was concerned with investigating the behavior of air jets at a submerged orifice in water over a wide range of flow rates. The main objective was to improve the basic understanding of gas-liquid phenomena (e.g., leak dynamics, gas bubble agglomeration, etc.) in a simulated tube bundle through visualization. The experimental results from these air-water tests will be used as a guide to help select the leak size for LLTR Series II Test A-4 because air-water system is a good simulation of water-sodium mixture

Experimental optimization of temperature distribution in the hot-gas duct through the installation of internals in the hot-gas plenum of a high-temperature reactor

International Nuclear Information System (INIS)

Henssen, J.; Mauersberger, R.

1990-01-01

The flow conditions in the hot-gas plenum and in the adjacent hot-gas ducts and hot-gas pipes for the high-temperature reactor project PNP-1000 (nuclear process heat project for 1000 MW thermal output) have been examined experimentally. The experiments were performed in a closed loop in which the flow model to be analyzed, representing a 60deg sector of the core bottom of the PNP-1000 with connecting hot-gas piping and diverting arrangements, was installed. The model scale was approx. 1:5.6. The temperature and flow velocity distribution in the hot-gas duct was registered by means of 14 dual hot-wire flowmeters. Through structural changes and/or the installation of internals into the hot-gas plenum of the core bottom offering little flow resistance coolant gas temperature differentials produced in the core could be reduced to such an extent that a degree of mixture amounting to over 80% was achieved at the entrance of the connected heat exchanger systems. Thereby the desired goal of an adequate degree of mixture of the hot gas involving an acceptable pressure loss was reached. (orig.)

49 CFR 572.145 - Upper and lower torso assemblies and torso flexion test procedure.

Science.gov (United States)

2010-10-01

...) ANTHROPOMORPHIC TEST DEVICES 3-year-Old Child Crash Test Dummy, Alpha Version § 572.145 Upper and lower torso... lumbar spine and abdomen of a fully assembled dummy (drawing 210-0000) to flexion articulation between... in paragraph (c) of this section, the lumbar spine-abdomen assembly shall flex by an amount that...

Analysis of select Mod-1 semiscale blowdown heat transfer tests. Final report

International Nuclear Information System (INIS)

Irani, A.A.; Fujita, N.; Mecham, D.C.; Ching, J.T.; Gose, G.C.; Hentzen, R.D.; Sawtelle, G.R.; Moore, K.V.

1976-10-01

The report contains the RELAP4 analysis and sensitivity studies of Semiscale Tests S-02-2 and S-02-7. The Semiscale System is an electrically heated experiment designed to produce data on system performance typical of PWR thermal-hydraulic behavior. The RELAP4 program used for these analyses is a digital computer program developed to predict the thermal-hydraulic behavior of experimental systems and water-cooled nuclear reactors subjected to postulated transients. The results of the analysis for Test S-02-2 were in very good agreement with the data. Two parameters which required improvement were identified. These were the lower plenum density and the mass flow on the vessel side of the break. Subsequently, before analyzing Test S-02-7, the lower plenum was renodalized and the critical flow model at the vessel side break was modified. The results of the analysis of Test S-02-7 compared more favorably with the data than those of S-02-2. Additional sensitivity studies included time step studies, steam generator and downcomer modeling, and core nodalization

Computational Fluid Dynamic Analysis of the VHTR Lower Plenum Standard Problem

International Nuclear Information System (INIS)

Johnson, Richard W.; Schultz, Richard R.

2009-01-01

The United States Department of Energy is promoting the resurgence of nuclear power in the U. S. for both electrical power generation and production of process heat required for industrial processes such as the manufacture of hydrogen for use as a fuel in automobiles. The DOE project is called the next generation nuclear plant (NGNP) and is based on a Generation IV reactor concept called the very high temperature reactor (VHTR), which will use helium as the coolant at temperatures ranging from 450 C to perhaps 1000 C. While computational fluid dynamics (CFD) has not been used for past safety analysis for nuclear reactors in the U.S., it is being considered for safety analysis for existing and future reactors. It is fully recognized that CFD simulation codes will have to be validated for flow physics reasonably close to actual fluid dynamic conditions expected in normal and accident operational situations. To this end, experimental data have been obtained in a scaled model of a narrow slice of the lower plenum of a prismatic VHTR. The present report presents results of CFD examinations of these data to explore potential issues with the geometry, the initial conditions, the flow dynamics and the data needed to fully specify the inlet and boundary conditions; results for several turbulence models are examined. Issues are addressed and recommendations about the data are made

Experimental Measurement of Flow Phenomena in a VHTR Lower Plenum Model

Energy Technology Data Exchange (ETDEWEB)

Hugh M. McIlroy Jr.; Keith G. Condie; Glenn E. McCreery; Donald M. McEligot; Robert J. Pink

2006-06-01

The Very-High-Temperature Reactor (VHTR) is one of six reactor technologies chosen for further development by the Generation IV International Forum. In addition this system is the leading candidate for the Next Generation Nuclear Power (NGNP) Project in the U.S which has the goal of demonstrating the production of emissions free electricity and hydrogen by 2015. In preparation for the thermal-hydraulics and safety analyses that will be required to confirm the performance of the NGNP, work has begun on readying the computational tools that will be needed to predict the thermal-hydraulics conditions and safety margins of the reactor design. Meaningful feasibility studies for VHTR designs will require accurate, reliable predictions of material temperatures which depend upon the thermal convection in the coolant channels of the core and other components. Unfortunately, one-dimensional system codes for gas-cooled reactors typically underpredict these temperatures, particularly for reduced power operations and hypothesized accident scenarios. Likewise, most turbulence models in general-purpose CFD codes also underpredict these temperatures. Matched-Index-of-Refraction (MIR) fluid dynamics experiments have been designed and built to develop benchmark databases for the assessment of CFD solutions of the momentum equations, scalar mixing and turbulence models for typical VHTR plenum geometries in the limiting case of negligible buoyancy and constant fluid properties.

Fuel rods

International Nuclear Information System (INIS)

Hattori, Shinji; Kajiwara, Koichi.

1980-01-01

Purpose: To ensure the safety for the fuel rod failures by adapting plenum springs to function when small forces such as during transportation of fuel rods is exerted and not to function the resilient force when a relatively great force is exerted. Constitution: Between an upper end plug and a plenum spring in a fuel rod, is disposed an insertion member to the lower portion of which is mounted a pin. This pin is kept upright and causes the plenum spring to function resiliently to the pellets against the loads due to accelerations and mechanical vibrations exerted during transportation of the fuel rods. While on the other hand, if a compression force of a relatively high level is exerted to the plenum spring during reactor operation, the pin of the insertion member is buckled and the insertion member is inserted to the inside of the plenum spring, whereby the pellets are allowed to expand freely and the failures in the fuel elements can be prevented. (Moriyama, K.)

49 CFR 572.165 - Upper and lower torso assemblies and torso flexion test procedure.

Science.gov (United States)

2010-10-01

...) ANTHROPOMORPHIC TEST DEVICES Hybrid III Six-Year-Old Weighted Child Test Dummy § 572.165 Upper and lower torso... determine the stiffness effects of the lumbar spine (specified in 49 CFR 572.125(a)), including cable... bushing (specified in 49 CFR 572.125(a)), nut (specified in 49 CFR 572.125(a)), spine box weighting plates...

Evaluation report on CCTF Core-I reflood tests C1-5 (Run 14), C1-7 (Run 16) and C1-14 (Run 23)

International Nuclear Information System (INIS)

Sugimoto, Jun; Muurao, Yoshio

1983-02-01

The present report describes the effects of the initial clad temperature on the reflood phenomena observed in the Cylindrical Core Test Facility (CCTF) at Japan Atomic Energy Research Institute. The evaluation is based on the data of tests C1-5, C1-7 and C1-14 of the CCTF-Core I test series. Nominal initial peak clad temperatures in these tests are 600 0 C, 700 0 C and 800 0 C, respectively. With the higher initial clad temperature, the higher loop mass flow rate and the lower water accumulation in the core and the upper plenum were obtained in an early reflood transient. However, the core inlet flow conditions, which is sensitive to the core cooling, were not much affected by the higher initial clad temperature. The slower quench front propagation was observed with the higher initial clad temperature. However, the heat transfer coefficient was almost identical with each other before the turnaround time, which resulted in the lower temperature rise with the highest initial clad temperature. This qualitatively agreed with the results of the forced feed FLECHT experiment. (author)

Predictable anomalies of process parameters on failure mode of internal structures in RPV by transient thermal-hydraulic analysis

International Nuclear Information System (INIS)

Maki, Akira; Mori, Michitsugu; Kanemoto, Shigeru; Konishi, Hideo

1997-01-01

A study has been conducted to evaluate how process parameters will exhibit the change in the event of the troubles related to reactor internal by using transient thermal-hydraulic analysis codes (RETRAN3D-MOD002, etc.). In the present study, the following six events are analytically investigated: 1) a leak from the upper plenum; 2) a leak from the middle part of a shroud; 3) a leak from the lower plenum; 4) a leak from the riser pipe for the jet-pump; 5) the blockage of the jet-pump nozzle; and 6) a leak from the jet-pump diffuser. The results by analyses indicated that the leak from the upper plenum resulted in increasing in the inlet temperature of primary loop recirculation (PLR) and in the differential pressure at the core support plate, and decreasing in the neutron flux (reactor power). Similar analyses were made for the five other events to identify the pattern of relevant process parameter variation in each event. (author)

Uncertainty evaluation in the self-alignment test of the upper plate of a press

International Nuclear Information System (INIS)

Lourenço, Alexandre S; E Sousa, J Alves

2015-01-01

This paper describes a method to evaluate uncertainty of the self-alignment test of the upper plate of a press according to EN 12390-4:2000. The method, the algorithms and the sources of uncertainty are described

Evaluating Upper-Body Strength and Power From a Single Test: The Ballistic Push-up.

Science.gov (United States)

Wang, Ran; Hoffman, Jay R; Sadres, Eliahu; Bartolomei, Sandro; Muddle, Tyler W D; Fukuda, David H; Stout, Jeffrey R

2017-05-01

Wang, R, Hoffman, JR, Sadres, E, Bartolomei, S, Muddle, TWD, Fukuda, DH, and Stout, JR. Evaluating upper-body strength and power from a single test: the ballistic push-up. J Strength Cond Res 31(5): 1338-1345, 2017-The purpose of this study was to examine the reliability of the ballistic push-up (BPU) exercise and to develop a prediction model for both maximal strength (1 repetition maximum [1RM]) in the bench press exercise and upper-body power. Sixty recreationally active men completed a 1RM bench press and 2 BPU assessments in 3 separate testing sessions. Peak and mean force, peak and mean rate of force development, net impulse, peak velocity, flight time, and peak and mean power were determined. Intraclass correlation coefficients were used to examine the reliability of the BPU. Stepwise linear regression was used to develop 1RM bench press and power prediction equations. Intraclass correlation coefficient's ranged from 0.849 to 0.971 for the BPU measurements. Multiple regression analysis provided the following 1RM bench press prediction equation: 1RM = 0.31 × Mean Force - 1.64 × Body Mass + 0.70 (R = 0.837, standard error of the estimate [SEE] = 11 kg); time-based power prediction equation: Peak Power = 11.0 × Body Mass + 2012.3 × Flight Time - 338.0 (R = 0.658, SEE = 150 W), Mean Power = 6.7 × Body Mass + 1004.4 × Flight Time - 224.6 (R = 0.664, SEE = 82 W); and velocity-based power prediction equation: Peak Power = 8.1 × Body Mass + 818.6 × Peak Velocity - 762.0 (R = 0.797, SEE = 115 W); Mean Power = 5.2 × Body Mass + 435.9 × Peak Velocity - 467.7 (R = 0.838, SEE = 57 W). The BPU is a reliable test for both upper-body strength and power. Results indicate that the mean force generated from the BPU can be used to predict 1RM bench press, whereas peak velocity and flight time measured during the BPU can be used to predict upper-body power. These findings support the potential use of the BPU as a valid method to evaluate upper-body strength and power.

One-dimensional three-field model of condensation in horizontal countercurrent flow with supercritical liquid velocity

International Nuclear Information System (INIS)

Trewin, Richard R.

2011-01-01

Highlights: → CCFL in the hot leg of a PWR with ECC Injection. → Three-Field Model of counter flowing water film and entrained droplets. → Flow of steam can cause a hydraulic jump in the supercritical flow of water. → Condensation of steam on subcooled water increases the required flow for hydraulic jump. → Better agreement with UPTF experimental data than Wallis-type correlation. - Abstract: A one-dimensional three-field model was developed to predict the flow of liquid and vapor that results from countercurrent flow of water injected into the hot leg of a PWR and the oncoming steam flowing from the upper plenum. The model solves the conservation equations for mass, momentum, and energy in a continuous-vapor field, a continuous-liquid field, and a dispersed-liquid (entrained-droplet) field. Single-effect experiments performed in the upper plenum test facility (UPTF) of the former SIEMENS KWU (now AREVA) at Mannheim, Germany, were used to validate the countercurrent flow limitation (CCFL) model in case of emergency core cooling water injection into the hot legs. Subcooled water and saturated steam flowed countercurrent in a horizontal pipe with an inside diameter of 0.75 m. The flow of injected water was varied from 150 kg/s to 400 kg/s, and the flow of steam varied from 13 kg/s to 178 kg/s. The subcooling of the liquid ranged from 0 K to 104 K. The velocity of the water at the injection point was supercritical (greater than the celerity of a gravity wave) for all the experiments. The three-field model was successfully used to predict the experimental data, and the results from the model provide insight into the mechanisms that influence the flows of liquid and vapor during countercurrent flow in a hot leg. When the injected water was saturated and the flow of steam was small, all or most of the injected water flowed to the upper plenum. Because the velocity of the liquid remained supercritical, entrainment of droplets was suppressed. When the injected

The Reliability of Quality of Upper Extremity Skills Test in Children with Cerebral Palsy

Directory of Open Access Journals (Sweden)

Nazila Akbar-Fahimi

2012-01-01

Full Text Available Objective: The aim of this study was to survey the reliability of Intra-rater and Inter-rater with and without video camera assessment in children with spastic cerebral palsy. Materials & Methods: In this cross-sectional study, we validate the Quality of Upper Extremity Skill Test questionnaire. Fifty children with hemiplegia aged 19 to 95 months (mean age 61.31 ± 25.7 month were enrolled in our study using non random available approach. After obtaining parents’ consent, intra-rater assessment was performed in one session and intera rater assessment with camera after 10 days. Then, the third examiner did the reassessment using film observation of 46 children from 50. Spearman correlation for survey the reliability of intra-rater & inter rater with & without video recording assessment & gross motor function classification system 66 for determined functionality of child were used. Results: Intra-rater correlation was 0.774-0.996, Inter-rater correlation was 0.663-0.998 and correlation for video camera assessment was 0.710-0.974 for the first and third evaluation and 0.652-0.938 for second and third evaluation. P value for sub scales and total score was P<0.01. Conclusion: There is a high correlation in Intra rater and inter rater assessment with and without video recording in Quality of Upper Extremity Skill Test in children with cerebral palsy. So that it can be used as a reliable test to evaluate Quality of Upper Extremity Skills in these children.

Three-dimensional crust and upper mantle structure at the Nevada test site

International Nuclear Information System (INIS)

Taylor, S.R.

1983-01-01

The three-dimensional crust and upper mantle structure at the Nevada Test Site (NTS) is derived by combining teleseismic P wave travel time residuals with Pn source time terms. The NTS time terms and relative teleseismic residuals are calculated by treating the explosions as a network of 'receivers' which record 'shots' located at the surrounding stations. Utilization of the Pn time terms allows for better crustal resolution than is possible from teleseismic information alone. Average relative teleseismic P wave residuals show a consistent progression of positive (late arrivals) to negative residuals from east to west across the NTS. However, Pn time terms beneath Rainier Mesa are at least 0.3 and 0.5 s less than those beneath Pahute Mesa and Yucca Flat, respectively, indicating the presence of high-velocity crustal material or crustal thinning beneath Rainier Mesa. The time terms at Pahute Mesa are surprisingly uniform, and the largest time terms and residuals are observed in the northwest and southern parts of Yucca Flat. The Pn time terms show a slight correlation with the working-point velocity at the shot point for Pahute Mesa and Yucca Flat, indicating that part of the observed lateral variations are caused by shallow effects of the upper crust. Three-dimensional inversion of the travel time residuals suggests that Yucca Flat is characterized by low-velocity anomalies confined to the upper crust, Rainer Mesa by very high velocities in the upper and middle crust, and Pahute Mesa by a high-velocity anomaly extending through the crust and into the upper mantle. Relatively low velocities are observed in the lower crust beneath the Timber Mountain caldera south of Pahute Mesa with no expression in the upper mantle. These observed differences in velocity beneath the Tertiary Silent Canyon and Timber Mountain calderas may be related to their magma volume and mode of enrichment from a mantle-derived magma source

Laboratory test variables useful for distinguishing upper from lower gastrointestinal bleeding.

Science.gov (United States)

Tomizawa, Minoru; Shinozaki, Fuminobu; Hasegawa, Rumiko; Shirai, Yoshinori; Motoyoshi, Yasufumi; Sugiyama, Takao; Yamamoto, Shigenori; Ishige, Naoki

2015-05-28

To distinguish upper from lower gastrointestinal (GI) bleeding. Patient records between April 2011 and March 2014 were analyzed retrospectively (3296 upper endoscopy, and 1520 colonoscopy). Seventy-six patients had upper GI bleeding (Upper group) and 65 had lower GI bleeding (Lower group). Variables were compared between the groups using one-way analysis of variance. Logistic regression was performed to identify variables significantly associated with the diagnosis of upper vs lower GI bleeding. Receiver-operator characteristic (ROC) analysis was performed to determine the threshold value that could distinguish upper from lower GI bleeding. Hemoglobin (P = 0.023), total protein (P = 0.0002), and lactate dehydrogenase (P = 0.009) were significantly lower in the Upper group than in the Lower group. Blood urea nitrogen (BUN) was higher in the Upper group than in the Lower group (P = 0.0065). Logistic regression analysis revealed that BUN was most strongly associated with the diagnosis of upper vs lower GI bleeding. ROC analysis revealed a threshold BUN value of 21.0 mg/dL, with a specificity of 93.0%. The threshold BUN value for distinguishing upper from lower GI bleeding was 21.0 mg/dL.

Experiments on the behaviour of thermite melt injected into sodium: Final report on the THINA test results

International Nuclear Information System (INIS)

Huber, F.; Kaiser, A.; Peppler, W.

1994-01-01

During hypothetical accidents of fast breeder reactors the core melts and part of the core material inventory is ejected into the upper coolant plenum. As a consequence, a fuel to coolant thermal interaction occurs between the melt and the sodium. A series of simulating experiments was carried out in KfK/IRS to improve the knowledge about the phenomenology of molten fuel/coolant interactions and to support theoretical work on the safety of fast breeder reactors. In the tests, a thermite melt of up to 3270 K is injected from below into a sodium pool the temperature of which is between 770 and 820 K. The masses of the melt and the sodium are about five and 150 kg, respectively. Thermal interactions have been observed to occur as a sequence of small local pressure events mainly during the melt injection. Large-scale vapour explosions have not been observed. Generally, the conversion ratios of thermal to mechanical energy have been low. (author)

Fifth in situ vitrification engineering-scale test of simulated INEL buried waste sites

International Nuclear Information System (INIS)

Bergsman, T.M.; Shade, J.W.; Farnsworth, R.K.

1992-06-01

In September 1990, an engineering-scale in situ vitrification (ISV) test was conducted on sealed canisters containing a combined mixture of buried waste materials expected to be present at the Idaho National Engineering Laboratory (INEL) Subsurface Disposal Area (SDA). The test was part of a Pacific Northwest Laboratory (PNL) program to assist INEL in treatability studies of the potential application of ISV to mixed transuranic wastes at the INEL SDA. The purpose of this test was to determine the effect of a close-packed layer of sealed containers on ISV processing performance. Specific objectives included determining (1) the effect of releases from sealed containers on hood plenum pressure and temperature, (2) the release pressure ad temperatures of the sealed canisters, (3) the relationships between canister depressurization and melt encapsulation, (4) the resulting glass and soil quality, (5) the potential effects of thermal transport due to a canister layer, (6) the effects on particle entrainment of differing angles of approach for the ISV melt front, and (7) the effects of these canisters on the volatilization of voltatile and semivolatile contaminants into the hood plenum

TEST-RETEST RELIABILITY OF THE CLOSED KINETIC CHAIN UPPER EXTREMITY STABILITY TEST (CKCUEST) IN ADOLESCENTS: RELIABILITY OF CKCUEST IN ADOLESCENTS.

Science.gov (United States)

de Oliveira, Valéria M A; Pitangui, Ana C R; Nascimento, Vinícius Y S; da Silva, Hítalo A; Dos Passos, Muana H P; de Araújo, Rodrigo C

2017-02-01

The Closed Kinetic Chain Upper Extremity Stability Test (CKCUEST) has been proposed as an option to assess upper limb function and stability; however, there are few studies that support the use of this test in adolescents. The purpose of the present study was to investigate the intersession reliability and agreement of three CKCUEST scores in adolescents and establish clinimetric values for this test. Test-retest reliability. Twenty-five healthy adolescents of both sexes were evaluated. The subjects performed two CKCUEST with an interval of one week between the tests. An intraclass correlation coefficient (ICC 3,3 ) two-way mixed model with a 95% interval of confidence was utilized to determine intersession reliability. A Bland-Altman graph was plotted to analyze the agreement between assessments. The presence of systematic error was evaluated by a one-sample t test. The difference between the evaluation and reevaluation was observed using a paired-sample t test. The level of significance was set at 0.05. Standard error of measurements and minimum detectable changes were calculated. The intersession reliability of the average touches score, normalized score, and power score were 0.68, 0.68 and 0.87, the standard error of measurement were 2.17, 1.35 and 6.49, and the minimal detectable change was 6.01, 3.74 and 17.98, respectively. The presence of systematic error (p test with moderate to excellent reliability when used with adolescents. The CKCUEST is a measurement with moderate to excellent reliability for adolescents. 2b.

Reflux condensation behavior in SBLOCA tests of ATLAS facility

International Nuclear Information System (INIS)

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

2017-01-01

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

FIX-II. Loca-blowdown heat transfer and pump trip experiments. Summary report of phase 1: Design of experiments

International Nuclear Information System (INIS)

Waaranperae, Y.; Nilsson, L.; Gustafsson, P.Aa.; Jonsson, N.O.

1979-06-01

FIX-II is a loss of coolant blowdown heat transfer experiment, performed under contract for The Swedish Nuclear Power Inspectorate, SKI. The purpose of the experiments is to provide measurements from simulations of a pipe rupture on an external recirculation line in a Swedish BWR. Pump trips in BWRs with internal recirculation pumps will also be simulated. The existing FIX-loop at the Thermal Engineering Laboratory of Studsvik Energiteknik AB will be modified and used for the experiments. Components are included to simulate the steam dome, downcomer, two recirculation lines with one pump each, lower plenum, core (36-rod full length bundle), control rod guide tubes, core bypass, upper plenum and steam separators. The results of the first phase of the project are reported here. The following tasks are included in Phase 1: reactor reference analysis, scaling calculations of the FIX loop, development of fuel rod simulators, design of test section and test loop layout and proposal for test program. Further details of the work and results obtained for the different sub-projects are published in a number ofdetailed reports. (author)

Experiment data report for Semiscale Mod-1 Test S-05-1 (alternate ECC injection test)

International Nuclear Information System (INIS)

Feldman, E.M.; Patton, M.L. Jr.; Sackett, K.E.

1977-02-01

Recorded test data are presented for Test S-05-1 of the Semiscale Mod-1 alternate ECC injection test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-05-1 was conducted from initial conditions of 2263 psia and 544 0 F to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the cold leg broken loop piping. During the test, cooling water was injected into the vessel lower plenum to simulate emergency core coolant injection in a PWR, with the flow rate based on system volume scaling

Summary report of NEPTUN investigations into the steady state thermal hydraulics of the passive decay heat removal

International Nuclear Information System (INIS)

Rust, K.; Weinberg, D.; Hoffmann, H.; Frey, H.H.; Baumann, W.; Hain, K.; Leiling, W.; Hayafune, H.; Ohira, H.

1995-12-01

During the course of steady state NEPTUN investigations, the effects of different design and operating parameters were studied; in particular: The shell design of the above core sturcture, the core power, the number of decay heat exchangers put in operation, the complete flow path blockage at the primary side of the intermediate heat exchangers, and the fluid level in the primary vessel. The findings of the NEPTUN experiments indicate that the decay heat can be safely removed by natural convection. The interwrapper flow makes an essential contribution to that behavior. The decay heat exchangers installed in the upper plenum cause a thermal stratification associated with a pronounced gradient. The vertical extent of the stratification and the quantity of the gradient are depending on the fact whether a permeable or an impermeable shell covers the above core structure. An increase of the core power or a reduction of the number of decay heat exchangers being in operation leads to a higher temperature level in the primary system but does not alter the global temperature distribution. In the case that no coolant enters the inlet windows at the primary side of the intermediate and decay heat exchangers, the core remains coolable as far as the primary vessel is filled with fluid up to a minimum level. Cold water penetrates from the upper plenum into the core and removes the decay heat. The thermal hydraulic computer code FLUTAN was applied for the three-dimensional numerical simulation of the majority of NEPTUN tests reported here. The comparison of computed against experimental data indicates a qualitatively and quantitatively satisfying agreement of the findings with respect to the field of isotherms as well as the temperature profiles in the upper plenum and within the core region of very complex geometry. (orig./HP) [de

Evaluation report on SCTF Core-III tests S3-7 and S3-8

International Nuclear Information System (INIS)

Okubo, Tsutomu; Iguchi, Tadashi; Iwamura, Takamichi

1990-03-01

It has been said that the Emergency Core Cooling (ECC) water injected into the hot legs flows into the upper plenum and then falls back to the core (i.e. break-through) during reflood phase in a German type Pressurized Water Reactor (GPWR) with the combined-injection-type ECCS, and that the break-through occurs where the water temperature at the tie plate area is lower and subcooled. Based on this information two tests were conducted with the Slab Core Test Facility (SCTF) Core-III in order to investigate the effects of the water temperature distribution at the tie plate area on the break-through and the core cooling. In these tests, the subcooled ECC water was injected just above the Upper Core Support Plate (UCSP) in order to establish the desired water temperature distribution at the tie plate area. In one test (Test S3-7) the ECC water injection above the UCSP was performed above Bundles 3 and 4, and in the other test (Test S3-8) above Bundles 7 and 8 during initial 60 s a and then was changed to above Bundles 3 and 4. The test data were compared with those of Test S3-SH1, in which the injection was performed above Bundles 7 and 8 and the other test conditions were the same as in Tests S3-7 and S3-8. Analyzing these test data, the following has been found: The break-through occurs where the water temperature at the tie plate area is subcooled and the core cooling is enhanced significantly in the break-through region. The break-through location changes, with some time lag, following the change of the water temperature distribution at the tie plate area. Furthermore, the core cooling in the non-break-through regions is almost the same regardless of the location of the break-through. (author)

LOX/LH2 propulsion system for launch vehicle upper stage, test results

Science.gov (United States)

Ikeda, T.; Imachi, U.; Yuzawa, Y.; Kondo, Y.; Miyoshi, K.; Higashino, K.

1984-01-01

The test results of small LOX/LH2 engines for two propulsion systems, a pump fed system and a pressure fed system are reported. The pump fed system has the advantages of higher performances and higher mass fraction. The pressure fed system has the advantages of higher reliability and relative simplicity. Adoption of these cryogenic propulsion systems for upper stage of launch vehicle increases the payload capability with low cost. The 1,000 kg thrust class engine was selected for this cryogenic stage. A thrust chamber assembly for the pressure fed propulsion system was tested. It is indicated that it has good performance to meet system requirements.

Free Surface Water Tunnel (FSWT)

Data.gov (United States)

Federal Laboratory Consortium — Description: The Free Surface Water Tunnel consists of the intake plenum, the test section and the exit plenum. The intake plenum starts with a perforated pipe that...

Steady-state irradiation testing of U-Pu-Zr fuel to >18% burnup

International Nuclear Information System (INIS)

Pahl, R.G.; Wisner, R.S.; Billone, M.C.; Hofman, G.L.

1990-01-01

Tests of austenitic stainless steel clad U-xP-10Zr fuel (x=o, 8, 19 wt. %) to peak burnups as high as 18.4 at. % have been completed in the EBR-II. Fuel swelling and fractional fission gas release are slowly increasing functions of burnup beyond 2 at. % burnup. Increasing plutonium content in the fuel reduces swelling and decreases the amount of fission gas which diffuses from fuel to plenum. LIFE-METAL code modelling of cladding strains is consistent with creep by fission gas loading and irradiation-induced swelling mechanisms. Fuel/cladding chemical interaction involves the ingress of rare-earth fission products. Constituent redistribution in the fuel had not limited steady-state performance. Cladding breach behavior at closure welds, in the gas plenum, and in the fuel column region have been benign events. 3 refs., 5 figs

Validation Studies for Numerical Simulations of Flow Phenomena Expected in the Lower Plenum of a Prismatic VHTR Reference Design

International Nuclear Information System (INIS)

Richard W. Johnson

2005-01-01

The final design of the very high temperature reactor (VHTR) of the fourth generation of nuclear power plants (Gen IV) has not yet been established. The VHTR may be either a prismatic (block) or pebble bed type. It may be either gas-cooled or cooled with an as yet unspecified molten salt. However, a conceptual design of a gas-cooled VHTR, based on the General Atomics GT-MHR, does exist and is called the prismatic VHTR reference design, MacDonald et al [2003], General Atomics [1996]. The present validation studies are based on the prismatic VHTR reference design. In the prismatic VHTR reference design, the flow in the lower plenum will be introduced by dozens of turbulent jets issuing into a large crossflow that must negotiate dozens of cylindrical support columns as it flows toward the exit duct of the reactor vessel. The jets will not all be at the same temperature due to the radial variation of power density expected in the core. However, it is important that the coolant be well mixed when it enters the power conversion unit to ensure proper operation and long life of the power conversion machinery. Hence, it is deemed important to be able to accurately model the flow and mixing of the variable temperature coolant in the lower plenum and exit duct. Accurate flow modeling involves determining modeling strategies including the fineness of the grid needed, iterative convergence tolerance, numerical discretization method used, whether the flow is steady or unsteady, and the turbulence model and wall treatment employed. It also involves validation of the computer code and turbulence model against a series of separate and combined flow phenomena and selection of the data used for the validation. The present report describes progress made to date for the task entitled ''CFD software validation of jets in crossflow'' which was designed to investigate the issues pertaining to the validation process

Investigation of water content in primary upper shield of high temperature engineering test reactor (HTTR)

International Nuclear Information System (INIS)

Sumita, Junya; Sawa, Kazuhiro; Mogi, Haruyoshi; Itahashi, Shuuji; Kitami, Toshiyuki; Akutu, Youichi; Fuchita, Yasuhiro; Kawaguchi, Toru; Moriya, Masahiro

1999-09-01

A primary upper shield of the High Temperature Engineering Test Reactor (HTTR) is composed of concrete (grout) which is packed into iron frames. The main function of the primary upper shield is to attenuate neutron and gamma ray from the core, that leads to satisfy dose equivalent rate limit of operating floor and stand-pipe room. Water content in the concrete is one of the most important things because it strongly affects neutron-shielding ability. Then, we carried out out-of-pile experiments to investigate relationship between temperature and water content in the concrete. Based on the experimental results, a hydrolysis-diffusion model was developed to investigate water release behavior from the concrete. The model showed that water content used for shielding design in the primary upper shield of the HTTR will be maintained if temperature during operating life is under 110degC. (author)

Demonstration tests for HTGR fuel elements and core components with test sections in HENDEL

Energy Technology Data Exchange (ETDEWEB)

Miyamoto, Yoshiaki; Hino, Ryutaro; Inagaki, Yoshiyuki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

1995-03-01

In the fuel stack test section (T{sub 1}) of the Helium Engineering Demonstration Loop (HENDEL), thermal and hydraulic performances of helium gas flows through a fuel rod channel and a fuel stack have been investigated for the High-Temperature Engineering Test Reactor (HTTR) core thermal design. The test data showed that the turbulent characteristics appearing in the Reynolds number above 2000: no typical behavior in the transition zone, and friction factors and heat transfer coefficients in the fuel channel were found to be higher than those in a smooth annular channel. Heat transfer behavior of gas flow in a fuel element channel with blockage and cross-flow through a gap between upper and lower fuel elements stacked was revealed using the mock-up models. On the other hand, demonstration tests have been performed to verify thermal and hydraulic characteristics and structural integrity related to the core bottom structure using a full-scale test facility named as the in-core structure test section (T{sub 2}). The sealing performance test revealed that the leakage of low-temperature helium gas through gaps between the permanent reflector blocks to the core was very low level compared with the HTTR design value and no change of the leakage flow rate were observed after a long term operation. The heat transfer tests including thermal transient at shutdown of gas circulators verified good insulating performance of core insulation structures in the core bottom structure and the hot gas duct; the temperature of the metal portion of these structure was below the design value. Examination of the thermal mixing characteristics indicated that the mixing of the hot helium gas started at a hot plenum and finished completely at downstream of the outlet hot gas duct. The present results obtained from these demonstration tests have been practically applied to the detailed design works and licensing procedures of the HTTR. (J.P.N.) 92 refs.

Bradykinesia-akinesia incoordination test: validating an online keyboard test of upper limb function.

Science.gov (United States)

Noyce, Alastair J; Nagy, Anna; Acharya, Shami; Hadavi, Shahrzad; Bestwick, Jonathan P; Fearnley, Julian; Lees, Andrew J; Giovannoni, Gavin

2014-01-01

The Bradykinesia Akinesia Incoordination (BRAIN) test is a computer keyboard-tapping task that was developed for use in assessing the effect of symptomatic treatment on motor function in Parkinson's disease (PD). An online version has now been designed for use in a wider clinical context and the research setting. Validation of the online BRAIN test was undertaken in 58 patients with Parkinson's disease (PD) and 93 age-matched, non-neurological controls. Kinesia scores (KS30, number of key taps in 30 seconds), akinesia times (AT30, mean dwell time on each key in milliseconds), incoordination scores (IS30, variance of travelling time between key presses) and dysmetria scores (DS30, accuracy of key presses) were compared between groups. These parameters were correlated against total motor scores and sub-scores from the Unified Parkinson's Disease Rating Scale (UPDRS). Mean KS30, AT30 and IS30 were significantly different between PD patients and controls (p≤0.0001). Sensitivity for 85% specificity was 50% for KS30, 40% for AT30 and 29% for IS30. KS30, AT30 and IS30 correlated significantly with UPDRS total motor scores (r = -0.53, r = 0.27 and r = 0.28 respectively) and motor UPDRS sub-scores. The reliability of KS30, AT30 and DS30 was good on repeated testing. The BRAIN test is a reliable, convenient test of upper limb motor function that can be used routinely in the outpatient clinic, at home and in clinical trials. In addition, it can be used as an objective longitudinal measurement of emerging motor dysfunction for the prediction of PD in at-risk cohorts.

Bradykinesia-akinesia incoordination test: validating an online keyboard test of upper limb function.

Directory of Open Access Journals (Sweden)

Alastair J Noyce

Full Text Available The Bradykinesia Akinesia Incoordination (BRAIN test is a computer keyboard-tapping task that was developed for use in assessing the effect of symptomatic treatment on motor function in Parkinson's disease (PD. An online version has now been designed for use in a wider clinical context and the research setting.Validation of the online BRAIN test was undertaken in 58 patients with Parkinson's disease (PD and 93 age-matched, non-neurological controls. Kinesia scores (KS30, number of key taps in 30 seconds, akinesia times (AT30, mean dwell time on each key in milliseconds, incoordination scores (IS30, variance of travelling time between key presses and dysmetria scores (DS30, accuracy of key presses were compared between groups. These parameters were correlated against total motor scores and sub-scores from the Unified Parkinson's Disease Rating Scale (UPDRS.Mean KS30, AT30 and IS30 were significantly different between PD patients and controls (p≤0.0001. Sensitivity for 85% specificity was 50% for KS30, 40% for AT30 and 29% for IS30. KS30, AT30 and IS30 correlated significantly with UPDRS total motor scores (r = -0.53, r = 0.27 and r = 0.28 respectively and motor UPDRS sub-scores. The reliability of KS30, AT30 and DS30 was good on repeated testing.The BRAIN test is a reliable, convenient test of upper limb motor function that can be used routinely in the outpatient clinic, at home and in clinical trials. In addition, it can be used as an objective longitudinal measurement of emerging motor dysfunction for the prediction of PD in at-risk cohorts.

Developing Tools to Test the Thermo-Mechanical Models, Examples at Crustal and Upper Mantle Scale

Science.gov (United States)

Le Pourhiet, L.; Yamato, P.; Burov, E.; Gurnis, M.

2005-12-01

Testing geodynamical model is never an easy task. Depending on the spatio-temporal scale of the model, different testable predictions are needed and no magic reciepe exist. This contribution first presents different methods that have been used to test themo-mechanical modeling results at upper crustal, lithospheric and upper mantle scale using three geodynamical examples : the Gulf of Corinth (Greece), the Western Alps, and the Sierra Nevada. At short spatio-temporal scale (e.g. Gulf of Corinth). The resolution of the numerical models is usually sufficient to catch the timing and kinematics of the faults precisely enough to be tested by tectono-stratigraphic arguments. In active deforming area, microseismicity can be compared to the effective rheology and P and T axes of the focal mechanism can be compared with local orientation of the major component of the stress tensor. At lithospheric scale the resolution of the models doesn't permit anymore to constrain the models by direct observations (i.e. structural data from field or seismic reflection). Instead, synthetic P-T-t path may be computed and compared to natural ones in term of rate of exhumation for ancient orogens. Topography may also help but on continent it mainly depends on erosion laws that are complicated to constrain. Deeper in the mantle, the only available constrain are long wave length topographic data and tomographic "data". The major problem to overcome now at lithospheric and upper mantle scale, is that the so called "data" results actually from inverse models of the real data and that those inverse model are based on synthetic models. Post processing P and S wave velocities is not sufficient to be able to make testable prediction at upper mantle scale. Instead of that, direct wave propagations model must be computed. This allows checking if the differences between two models constitute a testable prediction or not. On longer term, we may be able to use those synthetic models to reduce the residue

UPTF-TRAM experiments for SBLOCA: Evaluation of condensation processes in TRAM tests A6 and A7

Energy Technology Data Exchange (ETDEWEB)

Sonneburg, H.G.; Tuunanen, J.; Palazov, V.V. [Gesellschaft fuer Anlagen-und Reaktorsicherheit (GRS), Muenchen (Germany)

1995-09-01

The investigation of thermal-hydraulic phenomena related to reactor transients with accident management measures is the goal of the TRansient and accident Management (TRAM) experimental programme being carried out at the Upper Plenum Test Facility (UPTF) at Mannheim (Germany). These experimental investigations and test analyses are funded by the German Federal Minister for Research and Technology (BMFT). The UPTF simulates these phenomena in a 1:1 such relative to the dimension of a PWR. Condensation of steam during Emergency Core Cooling (ECC) water injection from accumulators into the primary system is one of the phenomena studied within the accumulators into the primary system is one of the phenomena studied within the TRAM programme. This phenomenon partly controls the efficiency of accumulator injection if the high pressure safety systems fail. Beside this, the condensation within the nitrogen inside the accumulator for a certain period controls the pressure development inside the accumulator. Thus, both condensation phenomena determine the ECC flow rate delivered to the primary system. Concerning the condensation inside the primary system, this is also of safety relevance in the case of Pressurized Thermal Shock (PTS) during cold leg injection.

A Comparison between Bench Press Throw and Ballistic Push-Up tests to assess upper-body power in trained individuals.

Science.gov (United States)

Bartolomei, Sandro; Nigro, Federico; Ruggeri, Sandro; Lanzoni, Ivan Malagoli; Ciacci, Simone; Merni, Franco; Sadres, Eliahu; Hoffman, Jay R; Semprini, Gabriele

2018-03-06

The purpose of the present study was to validate the ballistic push-up test performed with hands on a force plate (BPU) as a method to measure upper-body power. Twenty-eight experienced resistance trained men (age = 25.4 ± 5.2 y; body mass = 78.5 ± 9.0 kg; body height = 179.6 ± 7.8 cm) performed, two days apart, a bench press 1RM test and upper-body power tests. Mean power and peak power were assessed using the bench press throw test (BT) and the BPU test performed in randomized order. The area under the force/power curve (AUC) obtained at BT was also calculated. Power expressed at BPU was estimated using a time-based prediction equation. Mean force and the participant's body weight were used to predict the bench press 1RM. Pearson product moment correlations were used to examine relationships between the power assessment methods and between the predicted 1RM bench and the actual value. Large correlations (0.79; p bench and the 1RM predicted by the BPU. Results of the present study indicate that BPU represents a valid and reliable method to estimate the upper-body power in resistance-trained individuals.

Upper limb assessment using a Virtual Peg Insertion Test.

Science.gov (United States)

Fluet, Marie-Christine; Lambercy, Olivier; Gassert, Roger

2011-01-01

This paper presents the initial evaluation of a Virtual Peg Insertion Test developed to assess sensorimotor functions of arm and hand using an instrumented tool, virtual reality and haptic feedback. Nine performance parameters derived from kinematic and kinetic data were selected and compared between two groups of healthy subjects performing the task with the dominant and non-dominant hand, as well as with a group of chronic stroke subjects suffering from different levels of upper limb impairment. Results showed significantly smaller grasping forces applied by the stroke subjects compared to the healthy subjects. The grasping force profiles suggest a poor coordination between position and grasping for the stroke subjects, and the collision forces with the virtual board were found to be indicative of sensory deficits. These preliminary results suggest that the analyzed parameters could be valid indicators of impairment. © 2011 IEEE

About technical possibility to use VEERA facility for investigation of coolant stratification phenomenon in horizontal steam generators

International Nuclear Information System (INIS)

Mitioukov, V.; Mitrioukhine, A.; Korteniemi, V.

1997-01-01

The presentation gives a brief insight on possibility of using the VEERA facility in studying the stratification phenomenon. The idea for such experiments is to use the facility upper plenum part to simulate the conditions in upper part of horizontal steam generator hot collector. The upper part of steam generator hot collector is one of the locations where the stratification can take part during natural circulation mode

Experiment data report for LOFT nonnuclear test L1-3

International Nuclear Information System (INIS)

Millar, G.M.

1977-04-01

Test L1-3 was the third in a series of five nonnuclear isothermal blowdown tests conducted by the Loss of Fluid Test (LOFT) Program. For this test the LOFT Facility was configured to simulate a loss-of-coolant accident in a large pressurized water reactor resulting from a 200 percent double-ended shear break in a cold leg of the primary coolant system. A hydraulic core simulator assembly was installed in place of the nuclear core. The initial conditions in the primary coolant system intact loop were: temperature at 540 0 F, pressure at 2256 psig, and loop flow at 2.34 x 10 6 lbm/hr. During system depressurization, emergency core cooling water was specified to be injected into the lower plenum of the reactor vessel using an accumulator, a low-pressure injection system pump, and a high-pressure injection system pump to provide data on the effects of emergency core cooling on the system thermal-hydraulic response. Injection into the lower plenum was initiated from the high- and low-pressure injection systems. Injection from the accumulator, however, was not initiated because a valve was inadvertently left closed. The experiment, therefore, was not completely successful in that one of the objectives outlined in the experiment operating specification for this test was not accomplished. Test L1-3 was repeated at Test L1-3A to meet the experimental requirements. Despite these difficulties, Test L1-3 did provide very valuable data to verify experiment repeatability

Estimated Uncertainties in the Idaho National Laboratory Matched-Index-of-Refraction Lower Plenum Experiment

International Nuclear Information System (INIS)

Donald M. McEligot; Hugh M. McIlroy, Jr.; Ryan C. Johnson

2007-01-01

The purpose of the fluid dynamics experiments in the MIR (Matched-Index-of-Refraction) flow system at Idaho National Laboratory (INL) is to develop benchmark databases for the assessment of Computational Fluid Dynamics (CFD) solutions of the momentum equations, scalar mixing, and turbulence models for typical Very High Temperature Reactor (VHTR) plenum geometries in the limiting case of negligible buoyancy and constant fluid properties. The experiments use optical techniques, primarily particle image velocimetry (PIV) in the INL MIR flow system. The benefit of the MIR technique is that it permits optical measurements to determine flow characteristics in passages and around objects to be obtained without locating a disturbing transducer in the flow field and without distortion of the optical paths. The objective of the present report is to develop understanding of the magnitudes of experimental uncertainties in the results to be obtained in such experiments. Unheated MIR experiments are first steps when the geometry is complicated. One does not want to use a computational technique, which will not even handle constant properties properly. This report addresses the general background, requirements for benchmark databases, estimation of experimental uncertainties in mean velocities and turbulence quantities, the MIR experiment, PIV uncertainties, positioning uncertainties, and other contributing measurement uncertainties

Evaluation report on CCTF Core-II reflood tests C2-AC1 (run 51) and C2-4 (run 62)

International Nuclear Information System (INIS)

Sugimoto, Jun; Iguchi, Tadashi; Murao, Yoshio

1984-02-01

A reflood test program has been conducted at Japan Atomic Energy Research Institute (JAERI) using large scale test facilities named Cylindrical Core Test Facility (CCTF) and Slab Core Test Facility (SCTF). The present report describes the effect of the initial clad temperature i.e., the initial stored energy on reflood phenomena observed in CCTF Core-II tests C2-ACl and C2-4. The peak clad temperatures of tests C2-ACl and C2-4 were 863 K and 1069 K, respectively at reflood initiation. With higher initial clad temperature, obtained were lower water accumulation in the core and upper plenum, and higher loop mass flow rate in an early reflood transient due to larger heat release of the stored energy in the core. Core inlet flow conditions were only affected shortly after the reflood initiation, causing the suppressed flooding rate and the larger U-tube flow oscillation between the core and the downcomer. In the core, with higher initial clad temperature, slower quench front propagation and higher turnaround temperature were observed. Responses to a higher initial clad temperature were similar to those observed in CCTF Core-I and FLECHT tests. Thus, the lower temperature rise with higher initial clad temperature was experimentally confirmed. The importance of higher flooding rate at initial period was analytically shown for further decreasing the temperature rise. (author)

Fuel rod for use in BWR type reactor

International Nuclear Information System (INIS)

Takeuchi, Kiyoshi.

1989-01-01

A hollow intermediate end plug is disposed to a plenum portion of a fuel rod and a plenum spring is disposed between the end plug and the upper end of a fuel pellet. Then, a hollow portion is disposed between the intermediate end plug and an upper end plug. Thus, since a only a non exothermic portion is present from the intermediate end plug to the upper end plug, oxidation, corrosion, etc. to the fuel can are not caused so much as in the exothermic portion. Accordingly, the wall thickness of the fuel may be reduced to such a extent as only capable of withstanding the external pressure by coolants and the increasing inner pressure due to the release of FP gases and, accordingly, the wall thickness can be reduced as compared with that of the fuel portion in the fuel can. Further, since the power density per unit length of the fuel rod is reduced for fuels with increased number of fuel rods, it is possible to design so as to reduce the release amount of FP gases thereby decreasing the plenum volume. Further, since the surface area in the coolant phase stream portion is reduced, it can be expected for decreasing the pressure loss of fuels and accompanying effect for improving the channel stability. (T.M.)

Universal Point of Care Testing for Lynch Syndrome in Patients with Upper Tract Urothelial Carcinoma.

Science.gov (United States)

Metcalfe, Michael J; Petros, Firas G; Rao, Priya; Mork, Maureen E; Xiao, Lianchun; Broaddus, Russell R; Matin, Surena F

2018-01-01

Patients with Lynch syndrome are at risk for upper tract urothelial carcinoma. We sought to identify the incidence and most reliable means of point of care screening for Lynch syndrome in patients with upper tract urothelial carcinoma. A total of 115 consecutive patients with upper tract urothelial carcinoma without a history of Lynch syndrome were universally screened during followup from January 2013 through July 2016. We evaluated patient and family history using AMS (Amsterdam criteria) I and II, and tumor immunohistochemistry for mismatch repair proteins and microsatellite instability. Patients who were positive for AMS I/II, microsatellite instability or immunohistochemistry were classified as potentially having Lynch syndrome and referred for clinical genetic analysis and counseling. Patients with known Lynch syndrome served as positive controls. Of the 115 patients 16 (13.9%) screened positive for potential Lynch syndrome. Of these patients 7.0% met AMS II criteria, 11.3% had loss of at least 1 mismatch repair protein and 6.0% had high microsatellite instability. All 16 patients were referred for germline testing, 9 completed genetic analysis and counseling, and 6 were confirmed to have Lynch syndrome. All 7 patients with upper tract urothelial carcinoma who had a known history of Lynch syndrome were positive for AMS II criteria and at least a single mismatch repair protein loss while 5 of 6 had high microsatellite instability. We identified 13.9% of upper tract urothelial carcinoma cases as potential Lynch syndrome and 5.2% as confirmed Lynch syndrome at the point of care. These findings have important implications for universal screening of upper tract urothelial carcinoma, representing one of the highest rates of undiagnosed genetic disease in a urological cancer. Copyright © 2018 American Urological Association Education and Research, Inc. Published by Elsevier Inc. All rights reserved.

Experiment data report for semiscale MOD-1 test S-01-3 (isothermal blowdown with core resistance simulator)

International Nuclear Information System (INIS)

Zender, S.N.

1975-03-01

Recorded test data are presented for Test S-01-3 of the semiscale Mod-1 isothermal blowdown test series. Test S-01-3 is one of several semiscale Mod-1 experiments which are counterparts of the planned Loss-of-Fluid Test (LOFT) nonnuclear experiments. System hardware is of the LOFT design, selected using volumetric scaling methods, and initial conditions duplicate those identified for the LOFT nonnuclear tests. Test S-01-3 employed an intact loop resistance that was low relative to that of the first test in the series (Test S-01-2) to establish the importance of intact loop resistance on system response during blowdown. An orificed structure was used in the pressure vessel to simulate the LOFT core simulator. The test was initiated at isothermal conditions of 2245 psig and 538 0 F by a simulated offset shear of the cold-leg broken loop piping. During system depressurization, coolant was injected into the lower plenum of the pressure vessel to provide data on the effects of emergency core cooling on system response. Additionally, to aid in determination of the effects of accumulator gas on pressure suppression system response, the nitrogen used to charge the accumulator systems for Test S-01-3 was allowed to vent into the lower plenum following depletion of the coolant. (U.S.)

About technical possibility to use VEERA facility for investigation of coolant stratification phenomenon in horizontal steam generators

Energy Technology Data Exchange (ETDEWEB)

Mitioukov, V.; Mitrioukhine, A. [St. Petersburg State Technical Univ. (Russian Federation); Korteniemi, V. [Lappeenranta Univ. of Technology (Finland)

1997-12-31

The presentation gives a brief insight on possibility of using the VEERA facility in studying the stratification phenomenon. The idea for such experiments is to use the facility upper plenum part to simulate the conditions in upper part of horizontal steam generator hot collector. The upper part of steam generator hot collector is one of the locations where the stratification can take part during natural circulation mode. 4 refs.

About technical possibility to use VEERA facility for investigation of coolant stratification phenomenon in horizontal steam generators

Energy Technology Data Exchange (ETDEWEB)

Mitioukov, V; Mitrioukhine, A [St. Petersburg State Technical Univ. (Russian Federation); Korteniemi, V [Lappeenranta Univ. of Technology (Finland)

1998-12-31

The presentation gives a brief insight on possibility of using the VEERA facility in studying the stratification phenomenon. The idea for such experiments is to use the facility upper plenum part to simulate the conditions in upper part of horizontal steam generator hot collector. The upper part of steam generator hot collector is one of the locations where the stratification can take part during natural circulation mode. 4 refs.

Development of specimen size and test rate effects on the J-integral upper transition behavior of A533B steel

International Nuclear Information System (INIS)

Joyce, James A.

1988-01-01

During the past three years a test method has been developed for dynamic testing of fracture mechanics specimens which is specifically designed for application to the upper transition temperature range. The method uses drop tower loading rates of 2.5 m/sec and obtains a J IC or a J-R curve using an analytical key curve approach verified by initial and final crack length measurements obtained from the fracture surface. A J-R curve is obtained from each specimen and contains crack growth corrections so that it is directly comparable with static results obtained in accordance with the ASTM E1152 J-R curve test method. The test procedure has been applied to A106 steel, A533B steel and US Navy HY80 and HY100 steels at temperatures from -200F to 150F. Standard 1T three point bend specimens were used for the A533B and the HY100 steel. Static test results have shown that the J at cleavage initiation (which is presently an unstandardized quantity) is specimen a/W independent throughout the ductile to brittle transition but of course demonstrates considerable statistical scatter in the vicinity of the ductile upper shelf. Dynamic J-R tests have shown an increase in J IC with test rate for most, but not for all, materials. Separation of J into elastic and plastic components shows that the elastic J component increases with test rate in a fashion consistent with the materials tensile sensitivity to test rate but the plastic J component decreases with test rate - an apparent visco-plastic phenomena. For A106 steel the plastic J decrease exceeds the elastic J increase and the upper shelf toughness falls - while the other materials have demonstrated a relatively larger increase in the elastic J component and a smaller decrease in the plastic J component giving an overall increase in upper shelf toughness. Separation of the J integral into elastic and plastic components has demonstrated that J EL is specimen scale and geometry dependent while J PL is relatively scale and geometry

Ipsilesional upper limb performance in stroke individuals: relationship among outcomes of different tests used to assess hand function

Directory of Open Access Journals (Sweden)

Bianca Pinto Cunha

Full Text Available Abstract Introduction: Stroke individuals have sensorimotor repercussions on their ipsilesional upper limb. Therefore, it is important to use tests that allow an adequate assessment and follow-up of such deficits. Physical and occupational therapists commonly use maximal grip strength tests to assess the functional condition of stroke individuals. However, one could ask whether a single test is able to characterize the hand function in this population. Objective: The aim of this study was to investigate the relationship among outcomes of different tests frequently used to describe the function of the hand in the ipsilesional upper limb of stroke individuals. Methods: Twenty-two stroke individuals performed four hand function tests: maximal handgrip strength (HGSMax, maximal pinch grip strength (PGSMax, Jebsen-Taylor Hand Function Test (JTHFT and Nine Hole Peg Test (9-HPT. All tests were performed with the ipsilesional hand. Pearson's correlation analyses were performed. Results: the results indicated a moderate and positive relationship between HGSMax and JTHFT (r = 0.50 and between JTHFT and 9-HPT (r = 0.55. Conclusion: We conclude that the existence of only moderate relationships between test outcomes demonstrates the need to use at least two instruments to better describe the ipsilesional hand function of stroke individuals.

Quarterly technical progress report on water reactor safety programs sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research, April--June 1977

Energy Technology Data Exchange (ETDEWEB)

1977-08-01

Results from the previously conducted Semiscale Mod-1 ECC injection test series were analyzed. Testing in the LOFT counterpart test series was essentially completed, and the steam generator tube rupture test series was begun. Two tests in the alternate ECC injection test series were conducted which included injection of emergency core coolant into the upper plenum through use of the low pressure injection system. The Loss-of-Fluid Test Program successfully completed nonnuclear Loss-of-Coolant Experiment L1-4. A nuclear test, GC 2-3, in the Power Burst Facility Reactor was performed to evaluate the power oscillation method of determining gap conductance and to determine the effects of initial gap size, fill gas composition, and fuel density on the thermal performance of a light water reactor fuel rod. Additional test results were obtained relative to the behavior of irradiated fuel rods during a fast power increase and during a high power film boiling transient. Fuel model development and verification activities continued for the steady state and transient Fuel Rod Analysis Program, FRAP-S and FRAP-T. A computer code known as RELAP4/MOD7 is being developed to provide best-estimate modeling for reflood during a postulated loss-of-coolant accident (LOCA). A prediction of the fourth test in the boiling water reactor (BWR) Blowdown/Emergency Core Cooling Program was completed and an uncertainty analysis was completed of experimental steady state stable film boiling data for water flowing vertically upward in round tubes. A new multinational cooperative program to study the behavior of entrained liquid in the upper plenum and cross flow in the core during the reflood phase of a pressurized water reactor LOCA was defined.

Modeling for evaluation of debris coolability in lower plenum of reactor pressure vessel

International Nuclear Information System (INIS)

Maruyama, Yu; Moriyama, Kiyofumi; Nakamura, Hideo; Hirano, Masashi

2003-01-01

Effectiveness of debris cooling by water that fills a gap between the debris and the lower head wall was estimated through steady calculations in reactor scale. In those calculations, the maximum coolable debris depth was assessed as a function of gap width with combination of correlations for critical heat flux and turbulent natural convection of a volumetrically heated pool. The results indicated that the gap with a width of 1 to 2 mm was capable of cooling the debris under the conditions of the TMI-2 accident, and that a significantly larger gap width was needed to retain a larger amount of debris within the lower plenum. Transient models on gap growth and water penetration into the gap were developed and incorporated into CAMP code along with turbulent natural convection model developed by Yin, Nagano and Tsuji, categorized in low Reynolds number type two-equation model. The validation of the turbulent model was made with the UCLA experiment on natural convection of a volumetrically heated pool. It was confirmed that CAMP code predicted well the distribution of local heat transfer coefficients along the vessel inner surface. The gap cooling model was validated by analyzing the in-vessel debris coolability experiments at JAERI, where molten Al 2 O 3 was poured into a water-filled hemispherical vessel. The temperature history measured on the vessel outer surface was satisfactorily reproduced by CAMP code. (author)

A simple blowdown code for SUPER-SARA loop conditions

International Nuclear Information System (INIS)

Fritz, G.

1981-01-01

The Super Sara test programme (SSTP) is aimed to study in pile the fuel and cluster behaviour under two types of accident conditions: - the ''Large break loss of coolant'' condition (LB-Loca), - the ''Severe fuel damage'' (SFD) in a boildown caused by a small break. BIVOL was developed for the LB-Loca situation. This code is made for a loop where essentially two volumes define the thermohydraulics during the blowdown. In the SUPERSARA loop these two volumes are represented by the hot leg and cold leg pipings together with the respective upper and lower plenum of the test section

Event course analysis of core disruptive accidents

International Nuclear Information System (INIS)

Hering, W.; Homann, C.; Sengpiel, W.; Struwe, D.; Messainguiral, C.

1995-01-01

The theortical studies of the behavior of a PWR core in a meltdown accident are focused on hydrogen release, materials redistribution in the core area including forming of an oxide melt pool, quantity of melt and its composition, and temperatures attained by the RPV internals (esp. in the upper plenum) during the accident up to the time of melt relocation into the lower plenum. The calculations are done by the SCDAP/RELAP5 code. For its validation selected CORA results and Phebus FPTO results have been used. (orig.)

Event course analysis of core disruptive accidents; Ereignisablaufanalyse kernzerstoerender Unfaelle

Energy Technology Data Exchange (ETDEWEB)

Hering, W.; Homann, C.; Sengpiel, W.; Struwe, D.; Messainguiral, C.

1995-08-01

The theortical studies of the behavior of a PWR core in a meltdown accident are focused on hydrogen release, materials redistribution in the core area including forming of an oxide melt pool, quantity of melt and its composition, and temperatures attained by the RPV internals (esp. in the upper plenum) during the accident up to the time of melt relocation into the lower plenum. The calculations are done by the SCDAP/RELAP5 code. For its validation selected CORA results and Phebus FPTO results have been used. (orig.)

Top flooding modeling with MAAP4 code

International Nuclear Information System (INIS)

Brunet-Thibault, E.; Marguet, S.

2006-01-01

An engineering top flooding model was developed in MAAP4.04d.4, the severe accident code used in EDF, to simulate the thermal-hydraulic phenomena that should take place if emergency core cooling (ECC) water was injected in hot leg during quenching. In the framework of the ISTC (International Science and Technology Centre), a top flooding test was proposed in the PARAMETER facility (Podolsk, Russia). The MAAP calculation of the PARAMETER top flooding test is presented in this paper. A comparison between top and bottom flooding was made on the bundle test geometry. According to this study, top flooding appears to cool quickly and effectively the upper plenum internals. (author)

Multi-dimensional analysis of the ECC behavior in the UPI plant Kori Unit 1

International Nuclear Information System (INIS)

Bae, Sungwon; Chung, Bub-Dong; Bang, Young Seok

2008-01-01

A multi-dimensional transient analysis during the LBLOCA of the Kori Unit 1 has been performed by using the MARS code. Based on 1-D nodalization of the Kori Unit 1, the reactor vessel nodalizations have been replaced by the multi-dimensional component. The multi-dimensional component for the reactor vessel is designed as 5 radial, 8 peripheral, and 21 vertical grids. It is assumed that the fuel assemblies are homogeneously distributed in inner 3 radial grids. The outer 1 radial grid region is modeled as the core bypass. The outer-model 1 radial grid is used for the downcomer region. The corresponding heat structures and fuels are modified to fit for the multi-dimensional reactor vessel model. The form drag coefficients for the upper plenum and the core have been designated as 0.6 and 9.39, respectively. The form drag coefficients for the radial and peripheral directions are assigned to the same on the assumption of homogeneous distribution of the flow obstacles. After obtaining the 102% power steady operation condition, cold leg LOCA simulation is performed during 400 second period. The multi-dimensional steady run results show no severe differences compared to the traditional 1-D nodalization results. After the ECC injection starts, a liquid pool is maintained at the upper plenum because the ECCS water can not overcome the upward gas flow that comes from the reactor core through the upper tie plate. The depth of ECCS water pool is predicted as about 20% of the total height from the upper tie plate and the center line of the hot leg pipe. At the vicinity region of the active ECCS show higher depth of liquid pool. The accumulated water flow rate passing the upper tie plate is calculated by the transient result. Much downward water flow is obtained at the outer-most region of upper plenum space. The downward flow dominant region is about 32.3% of the total upper tie plate area. The accumulated ECCS bypass ratio is predicted as 27.64% at 300 second. It is calculated

Tellurium release and deposition during the TMI-2 accident

International Nuclear Information System (INIS)

Vinjamuri, K.; Osetek, D.J.; Hobbins, R.R.; Jessup, J.S.

1984-09-01

The estimated behavior of tellurium during and after the accident at the Three Mile Island Unit-2 is presented. The behavior is based on all available measurement data for /sup 129m/Te, 132 Te, stable tellurium ( 126 Te, 128 Te and 130 Te), and best estimate calculations of tellurium release and transport. The predicted release was calculated using current techniques that relate release rate to fuel temperature and holdup of tellurium in zircaloy until significant oxidation occurs. The calculated release fraction was low, approx. 7%, but the total measured release for samples analyzed to date is about 5.8%. Of the measured tellurium about 2.4, 1.8, 0.88, 0.42, 0.17 and 0.086% of core inventory were in the containment sump water, upper plenum assembly surfaces, containment solids in the sump water, makeup and purification demineralizer, containment inside surface, and the reactor primary coolant, respectively. A significant fraction (54%) of the tellurium calculated to be retained on the upper plenum surfaces (4.61% of the core inventory) was deposited during the high pressure injection of coolant at about 200 min after the reactor scram. Comparison of tellurium behavior with in-pile and out-of-pile tests strongly suggests that zircaloy holds tellurium until significant cladding oxidation occurs

Average structure of the upper earth mantle and crust between Albuquerque and the Nevada Test Site

International Nuclear Information System (INIS)

Garbin, H.D.

1979-08-01

Models of Earth structures were constructed by inverting seismic data obtained from nuclear events with a 1600-m-long laser strain meter. With these models the general structure of the earth's upper mantle and crust between Albuquerque and the Nevada Test Site was determined. 3 figures, 3 tables

Nondestructive testing of hardness of grey iron casts by upper harmonics method

International Nuclear Information System (INIS)

Ershov, R.E.; Ivanenko, T.G.; Kuznetsky, S.S.; Mutovin, V.A.

1982-01-01

The relation between microstructure of grey iron and his surface hardness on the one hand and magnetic characteristics on the other hand has been investigated on toroidal samples. It is shown that surface hardness of grey iron and parameter a in Froelich formula are dependent linearly on the total surface of graphite inclusions. The conclusion was made that the testing of the grey iron hardness better to make by using the phase of the upper harmonics of the output signal of the overlain transducer. The coefficient of the linear correlation between readings corresponding device and the hardness is 0.86. (orig.)

Hot-Fire Test of Liquid Oxygen/Hydrogen Space Launch Mission Injector Applicable to Exploration Upper Stage

Science.gov (United States)

Barnett, Greg; Turpin, Jason; Nettles, Mindy

2015-01-01

This task is to hot-fire test an existing Space Launch Mission (SLM) injector that is applicable for all expander cycle engines being considered for the exploration upper stage. The work leverages investment made in FY 2013 that was used to additively manufacture three injectors (fig. 1) all by different vendors..

Determination of fission gas release of spent nuclear fuel in puncturing test and in leaching experiments under anoxic conditions

Energy Technology Data Exchange (ETDEWEB)

González-Robles, E., E-mail: ernesto.gonzalez-robles@kit.edu [Karlsruhe Institute of Technology, Institute for Nuclear Waste Disposal (KIT-INE), P.O. Box 3640, D-76021, Karlsruhe (Germany); Metz, V. [Karlsruhe Institute of Technology, Institute for Nuclear Waste Disposal (KIT-INE), P.O. Box 3640, D-76021, Karlsruhe (Germany); Wegen, D.H. [European Commission, Joint Research Centre, Institute for Transuranium Elements (JRC-ITU), P.O. Box 2340, 76125, Karlsruhe (Germany); Herm, M. [Karlsruhe Institute of Technology, Institute for Nuclear Waste Disposal (KIT-INE), P.O. Box 3640, D-76021, Karlsruhe (Germany); Papaioannou, D. [European Commission, Joint Research Centre, Institute for Transuranium Elements (JRC-ITU), P.O. Box 2340, 76125, Karlsruhe (Germany); Bohnert, E. [Karlsruhe Institute of Technology, Institute for Nuclear Waste Disposal (KIT-INE), P.O. Box 3640, D-76021, Karlsruhe (Germany); Gretter, R. [European Commission, Joint Research Centre, Institute for Transuranium Elements (JRC-ITU), P.O. Box 2340, 76125, Karlsruhe (Germany); Müller, N. [Karlsruhe Institute of Technology, Institute for Nuclear Waste Disposal (KIT-INE), P.O. Box 3640, D-76021, Karlsruhe (Germany); Nasyrow, R.; Weerd, W. de; Wiss, T. [European Commission, Joint Research Centre, Institute for Transuranium Elements (JRC-ITU), P.O. Box 2340, 76125, Karlsruhe (Germany); Kienzler, B. [Karlsruhe Institute of Technology, Institute for Nuclear Waste Disposal (KIT-INE), P.O. Box 3640, D-76021, Karlsruhe (Germany)

2016-10-15

During reactor operation the fission gases Kr and Xe are formed within the UO{sub 2} matrix of nuclear fuel. Their quantification is important to evaluate their impact on critical parameters regarding the fuel behaviour during irradiation and (long-term) interim storage, such as internal pressure of the fuel rod and fuel swelling. Moreover the content of Kr and Xe in the plenum of a fuel rod and their content in the UO{sub 2} fuel itself are widely used as indicators for the release properties of {sup 129}I, {sup 137}Cs, and other safety relevant radionuclides with respect to final disposal of spent nuclear fuel. The present study deals with the fission gas release from spent nuclear fuel exposed to simulated groundwater in comparison with the fission gas previously released to the fuel rod plenum during irradiation in reactor. In a unique approach we determined both the Kr and Xe inventories in the plenum by means of a puncturing test and in leaching experiments with a cladded fuel pellet and fuel fragments in bicarbonate water under 3.2 bar H{sub 2} overpressure. The fractional inventory of the fission gases released during irradiation into the plenum was (8.3 ± 0.9) %. The fraction of inventory of fission gases released during the leaching experiments was (17 ± 2) % after 333 days of leaching of the cladded pellet and (25 ± 2) % after 447 days of leaching of the fuel fragments, respectively. The relatively high release of fission gases in the experiment with fuel fragments was caused by the increased accessibility of water to the Kr and Xe occluded in the fuel.

Investigation of analytical methods in thermal stratification analysis. Evaluation of flow rates through flow holes for normal and scram conditions of 40% power operation with AQUA code

International Nuclear Information System (INIS)

Doi, Yoshihiro; Muramatsu, Toshiharu

1997-08-01

Thermal stratification phenomena are observed in an upper plenum of liquid metal fast breeder reactors (LMFBRs) under reactor scram conditions, which give rise to thermal stress on structural components. Therefore it is important to evaluate characteristics of phenomena in the design of the internal structure in an LMFBR plenum. To evaluate flow rates through flow holes of the prototype fast breeder reactor, MONJU, numerical analyses were carried out with AQUA code for normal and scram conditions with 40% power operation. Through comparison of analysis results and measured temperature, thermal stratification phenomena in 300 second period after the scram was evaluated. Flow rate through the upper flow holes, the lower flow holes and annular gap between the inner barrel and the reactor vessel were evaluated with the measured temperature and the analysis results individually. (J.P.N.)

The Performance of the Upper Limb scores correlate with pulmonary function test measures and Egen Klassifikation scores in Duchenne muscular dystrophy.

Science.gov (United States)

Lee, Ha Neul; Sawnani, Hemant; Horn, Paul S; Rybalsky, Irina; Relucio, Lani; Wong, Brenda L

2016-01-01

The Performance of the Upper Limb scale was developed as an outcome measure specifically for ambulant and non-ambulant patients with Duchenne muscular dystrophy and is implemented in clinical trials needing longitudinal data. The aim of this study is to determine whether this novel tool correlates with functional ability using pulmonary function test, cardiac function test and Egen Klassifikation scale scores as clinical measures. In this cross-sectional study, 43 non-ambulatory Duchenne males from ages 10 to 30 years and on long-term glucocorticoid treatment were enrolled. Cardiac and pulmonary function test results were analyzed to assess cardiopulmonary function, and Egen Klassifikation scores were analyzed to assess functional ability. The Performance of the Upper Limb scores correlated with pulmonary function measures and had inverse correlation with Egen Klassifikation scores. There was no correlation with left ventricular ejection fraction and left ventricular dysfunction. Body mass index and decreased joint range of motion affected total Performance of the Upper Limb scores and should be considered in clinical trial designs. Copyright © 2016 Elsevier B.V. All rights reserved.

Large-Scale Liquid Hydrogen Tank Rapid Chill and Fill Testing for the Advanced Shuttle Upper Stage Concept

Science.gov (United States)

Flachbart, R. H.; Hedayat, A.; Holt, K. A.; Sims, J.; Johnson, E. F.; Hastings, L. J.; Lak, T.

2013-01-01

Cryogenic upper stages in the Space Shuttle program were prohibited primarily due to a safety risk of a 'return to launch site' abort. An upper stage concept addressed this concern by proposing that the stage be launched empty and filled using shuttle external tank residuals after the atmospheric pressure could no longer sustain an explosion. However, only about 5 minutes was allowed for tank fill. Liquid hydrogen testing was conducted within a near-ambient environment using the multipurpose hydrogen test bed 638.5 ft3 (18m3) cylindrical tank with a spray bar mounted longitudinally inside. Although the tank was filled within 5 minutes, chilldown of the tank structure was incomplete, and excessive tank pressures occurred upon vent valve closure. Elevated tank wall temperatures below the liquid level were clearly characteristic of film boiling. The test results have substantial implications for on-orbit cryogen transfer since the formation of a vapor film would be much less inhibited due to the reduced gravity. However, the heavy tank walls could become an asset in normal gravity testing for on-orbit transfer, i.e., if film boiling in a nonflight weight tank can be inhibited in normal gravity, then analytical modeling anchored with the data could be applied to reduced gravity environments with increased confidence.

Solar Thermal Upper Stage Cryogen System Engineering Checkout Test

Science.gov (United States)

Olsen, A. D; Cady, E. C.; Jenkins, D. S.

1999-01-01

The Solar Thermal Upper Stage technology (STUSTD) program is a solar thermal propulsion technology program cooperatively sponsored by a Boeing led team and by NASA MSFC. A key element of its technology program is development of a liquid hydrogen (LH2) storage and supply system which employs multi-layer insulation, liquid acquisition devices, active and passive thermodynamic vent systems, and variable 40W tank heaters to reliably provide near constant pressure H2 to a solar thermal engine in the low-gravity of space operation. The LH2 storage and supply system is designed to operate as a passive, pressure fed supply system at a constant pressure of about 45 psia. During operation of the solar thermal engine over a small portion of the orbit the LH2 storage and supply system propulsively vents through the enjoy at a controlled flowrate. During the long coast portion of the orbit, the LH2 tank is locked up (unvented). Thus, all of the vented H2 flow is used in the engine for thrust and none is wastefully vented overboard. The key to managing the tank pressure and therefore the H2 flow to the engine is to manage and balance the energy flow into the LH2 tank with the MLI and tank heaters with the energy flow out of the LH2 tank through the vented H2 flow. A moderate scale (71 cu ft) LH2 storage and supply system was installed and insulated at the NASA MSFC Test Area 300. The operation of the system is described in this paper. The test program for the LH2 system consisted of two parts: 1) a series of engineering tests to characterize the performance of the various components in the system: and 2) a 30-day simulation of a complete LEO and GEO transfer mission. This paper describes the results of the engineering tests, and correlates these results with analytical models used to design future advanced Solar Orbit Transfer Vehicles.

The validity of upper-limb neurodynamic tests for detecting peripheral neuropathic pain.

Science.gov (United States)

Nee, Robert J; Jull, Gwendolen A; Vicenzino, Bill; Coppieters, Michel W

2012-05-01

The validity of upper-limb neurodynamic tests (ULNTs) for detecting peripheral neuropathic pain (PNP) was assessed by reviewing the evidence on plausibility, the definition of a positive test, reliability, and concurrent validity. Evidence was identified by a structured search for peer-reviewed articles published in English before May 2011. The quality of concurrent validity studies was assessed with the Quality Assessment of Diagnostic Accuracy Studies tool, where appropriate. Biomechanical and experimental pain data support the plausibility of ULNTs. Evidence suggests that a positive ULNT should at least partially reproduce the patient's symptoms and that structural differentiation should change these symptoms. Data indicate that this definition of a positive ULNT is reliable when used clinically. Limited evidence suggests that the median nerve test, but not the radial nerve test, helps determine whether a patient has cervical radiculopathy. The median nerve test does not help diagnose carpal tunnel syndrome. These findings should be interpreted cautiously, because diagnostic accuracy might have been distorted by the investigators' definitions of a positive ULNT. Furthermore, patients with PNP who presented with increased nerve mechanosensitivity rather than conduction loss might have been incorrectly classified by electrophysiological reference standards as not having PNP. The only evidence for concurrent validity of the ulnar nerve test was a case study on cubital tunnel syndrome. We recommend that researchers develop more comprehensive reference standards for PNP to accurately assess the concurrent validity of ULNTs and continue investigating the predictive validity of ULNTs for prognosis or treatment response.

Predicting the Occurrence of Hypotension in Stable Patients With Nonvariceal Upper Gastrointestinal Bleeding: Point-of-Care Lactate Testing.

Science.gov (United States)

Ko, Byuk Sung; Kim, Won Young; Ryoo, Seung Mok; Ahn, Shin; Sohn, Chang Hwan; Seo, Dong Woo; Lee, Yoon-Seon; Lim, Kyoung Soo; Jung, Hwoon-Yong

2015-11-01

It is difficult to assess risk in normotensive patients with upper gastrointestinal bleeding. The aim of this study was to evaluate whether the initial lactate value can predict the in-hospital occurrence of hypotension in stable patients with acute nonvariceal upper gastrointestinal bleeding. Retrospective, observational, single-center study. Emergency department of a tertiary-care, university-affiliated hospital during a 5-year period. Medical records of 3,489 patients with acute upper gastrointestinal bleeding who were normotensive at presentation to the emergency department. We analyzed the ability of point-of-care testing of lactate at emergency department admission to predict hypotension development (defined as systolic blood pressure upper gastrointestinal bleeding, 157 patients experienced hypotension within 24 hours. Lactate was independently associated with hypotension development (odds ratio, 1.6; 95% CI, 1.4-1.7), and the risk of hypotension significantly increased as the lactate increased from 2.5-4.9 mmol/L (odds ratio, 2.2) to 5.0-7.4 mmol/L (odds ratio, 4.0) and to greater than or equal to 7.5 mmol/L (odds ratio, 39.2) (pupper gastrointestinal bleeding. However, subsequently, prospective validate research will be required to clarify this.

Influence of structure improvement of guide tubes and bundles in pressurized water reactor (PWR) on drop of control rods

International Nuclear Information System (INIS)

Shen Xiuzhong; Yu Pingan; Yang Guanyue

1996-01-01

In order to alleviate the cross hydraulic load on control rod guide tubes and bundles, some protective sleeves are added to those near the upper plenum outlet nozzles (4 symmetric bundles: 02-26, 03-25, 11-29, 12-28). In a 1/4 scale transparent model of the PWR upper plenum of Qinshan Nuclear Power Station, water was chosen as the fluid and hydraulic experiments with improved control rod guide tubes and bundles were carried out. The results were carefully compared with those of the experiments with unimproved control rod guide tubes and bundles. It is concluded that adding protective sleeves to the control rod guide tubes and bundles near the outlet nozzles will help to lighten the hydraulic load on them and make certain of the free movement and rapid dropping of control rods in the tubes and bundles in emergency by order

In-place HEPA filter penetration test

International Nuclear Information System (INIS)

Bergman, W.; Wilson, K.; Elliott, J.; Bettencourt, B.; Slawski, J.W.

1997-01-01

We have demonstrated the feasibility of conducting penetration tests on high efficiency particulate air (HEPA) filters as installed in nuclear ventilation systems. The in-place penetration test, which is designed to yield equivalent penetration measurements as the standard DOP efficiency test, is based on measuring the aerosol penetration of the filter installation as a function of particle size using a portable laser particle counter. This in-place penetration test is compared to the current in-place leak test using light scattering photometers for single HEPA filter installations and for HEPA filter plenums using the shroud method. Test results show the in-place penetration test is more sensitive than the in-place leak test, has a similar operating procedure, but takes longer to conduct. Additional tests are required to confirm that the in-place penetration test yields identical results as the standard dioctyl phthalate (DOP) penetration test for HEPA filters with controlled leaks in the filter and gasket and duct by-pass leaks. Further development of the procedure is also required to reduce the test time before the in- place penetration test is practical

Unsteady Reynolds Averaged Navier-Stokes and Large Eddy Simulations of Flows across Staggered Tube Bundle for a VHTR Lower Plenum Design

International Nuclear Information System (INIS)

Choi, Hyeon Kyeong; Park, Jong Woon

2013-01-01

In this work, behavior of unsteady and oscillating flow through a typical tube bundle array are analyzed by unsteady computations: 2D unsteady Reynolds averaged Navier-Stokes (URANS) and 3D Large Eddy Simulation (LES) and the results are compared with existing experimental data. In order to confirm appropriateness and limitations of CFD applications in the Korean VHTR design, two types of unsteady computations are performed such as 2D unsteady Reynolds averaged Navier-Stokes (URANS) and 3D Large Eddy Simulation (LES) for the existing tube bundle array. The velocity component profiles are compared with the experimental data and it is concluded that the URANS with the standard k-ω model is reasonably appropriate for cost-effective VHTR lower plenum analysis. Nevertheless, if more accurate results are needed, the LES-Smagorinsky computation is recommended considering limitations in the time averaged RANS in capturing small eddies

Classical test theory and Rasch analysis validation of the Upper Limb Functional Index in subjects with upper limb musculoskeletal disorders.

Science.gov (United States)

Bravini, Elisabetta; Franchignoni, Franco; Giordano, Andrea; Sartorio, Francesco; Ferriero, Giorgio; Vercelli, Stefano; Foti, Calogero

2015-01-01

To perform a comprehensive analysis of the psychometric properties and dimensionality of the Upper Limb Functional Index (ULFI) using both classical test theory and Rasch analysis (RA). Prospective, single-group observational design. Freestanding rehabilitation center. Convenience sample of Italian-speaking subjects with upper limb musculoskeletal disorders (N=174). Not applicable. The Italian version of the ULFI. Data were analyzed using parallel analysis, exploratory factor analysis, and RA for evaluating dimensionality, functioning of rating scale categories, item fit, hierarchy of item difficulties, and reliability indices. Parallel analysis revealed 2 factors explaining 32.5% and 10.7% of the response variance. RA confirmed the failure of the unidimensionality assumption, and 6 items out of the 25 misfitted the Rasch model. When the analysis was rerun excluding the misfitting items, the scale showed acceptable fit values, loading meaningfully to a single factor. Item separation reliability and person separation reliability were .98 and .89, respectively. Cronbach alpha was .92. RA revealed weakness of the scale concerning dimensionality and internal construct validity. However, a set of 19 ULFI items defined through the statistical process demonstrated a unidimensional structure, good psychometric properties, and clinical meaningfulness. These findings represent a useful starting point for further analyses of the tool (based on modern psychometric approaches and confirmatory factor analysis) in larger samples, including different patient populations and nationalities. Copyright © 2015 American Congress of Rehabilitation Medicine. Published by Elsevier Inc. All rights reserved.

RELAP5/MOD 3.2 Analysis of the Loss of RHR System Experiment Scaled to NPP Krsko

International Nuclear Information System (INIS)

Bencik, V.; Bajs, T.; Prah, M.

1998-01-01

In the paper the RELAP5/MOD 3.2 analysis of the loss of Residual Heat Removal (RHR) system during midloop operation experiment performed at the Rig of Safety Assessment (ROSA)-IV/Large Scale Test Facility (LSTF) together with the analysis of the same test scenario scaled to NPP Krsko are presented. The experiment consisted in a loss of the RHR system at cold shutdown conditions along with a 5% cold leg break in the loop without pressurizer. The Safety Injection (SI) system was disable in the calculation. The aims of the work were to study the physical phenomena encountered under low power and low system pressure conditions while the upper part of the Reactor Coolant System (RCS) is filled with noncondensable. The impact of the bypass flow between upper plenum and downcomer inlet on transient responses was investigated. The transient was simulated for 6000 s. (author)

USB environment measurements based on full-scale static engine ground tests. [Upper Surface Blowing for YC-14

Science.gov (United States)

Sussman, M. B.; Harkonen, D. L.; Reed, J. B.

1976-01-01

Flow turning parameters, static pressures, surface temperatures, surface fluctuating pressures and acceleration levels were measured in the environment of a full-scale upper surface blowing (USB) propulsive-lift test configuration. The test components included a flightworthy CF6-50D engine, nacelle and USB flap assembly utilized in conjunction with ground verification testing of the USAF YC-14 Advanced Medium STOL Transport propulsion system. Results, based on a preliminary analysis of the data, generally show reasonable agreement with predicted levels based on model data. However, additional detailed analysis is required to confirm the preliminary evaluation, to help delineate certain discrepancies with model data and to establish a basis for future flight test comparisons.

Design considerations and experimental observations for the TAMU air-cooled reactor cavity cooling system for the VHTR

Energy Technology Data Exchange (ETDEWEB)

Sulaiman, S. A., E-mail: shamsulamri@tamu.edu; Dominguez-Ontiveros, E. E., E-mail: elvisdom@tamu.edu; Alhashimi, T., E-mail: jbudd123@tamu.edu; Budd, J. L., E-mail: dubaiboy@tamu.edu; Matos, M. D., E-mail: mailgoeshere@gmail.com; Hassan, Y. A., E-mail: yhasssan@tamu.edu [Department of Nuclear Engineering, Texas A and M University, College Station, TX, 77843-3133 (United States)

2015-04-29

The Reactor Cavity Cooling System (RCCS) is a promising passive decay heat removal system for the Very High Temperature Reactor (VHTR) to ensure reliability of the transfer of the core residual and decay heat to the environment under all off-normal circumstances. A small scale experimental test facility was constructed at Texas A and M University (TAMU) to study pertinent multifaceted thermal hydraulic phenomena in the air-cooled reactor cavity cooling system (RCCS) design based on the General Atomics (GA) concept for the Modular High Temperature Gas-Cooled Reactor (MHTGR). The TAMU Air-Cooled Experimental Test Facility is ⅛ scale from the proposed GA-MHTGR design. Groundwork for experimental investigations focusing into the complex turbulence mixing flow behavior inside the upper plenum is currently underway. The following paper illustrates some of the chief design considerations used in construction of the experimental test facility, complete with an outline of the planned instrumentation and data acquisition methods. Computational Fluid Dynamics (CFD) simulations were carried out to furnish some insights on the overall behavior of the air flow in the system. CFD simulations assisted the placement of the flow measurement sensors location. Preliminary experimental observations of experiments at 120oC inlet temperature suggested the presence of flow reversal for cases involving single active riser at both 5 m/s and 2.25 m/s, respectively and four active risers at 2.25 m/s. Flow reversal may lead to thermal stratification inside the upper plenum by means of steady state temperature measurements. A Particle Image Velocimetry (PIV) experiment was carried out to furnish some insight on flow patterns and directions.

2D/3D Program work summary report, [January 1988--December 1992

International Nuclear Information System (INIS)

Damerell, P.S.; Simons, J.W.

1993-06-01

The 2D/3D Program was carried out by Germany, Japan and the United States to investigate the thermal-hydraulics of a PWR large-break LOCA. A contributory approach was utilized in which each country contributed significant effort to the program and all three countries shared the research results. Germany constructed and operated the Upper Plenum Test Facility (UPTF), and Japan constructed and operated the Cylindrical Core Test Facility (CCTF) and the Slab Core Test Facility (SCTF). The US contribution consisted of provision of advanced instrumentation to each of the three test facilities, and assessment of the TRAC computer code against the test results. Evaluations of the test results were carried out in all three countries. This report summarizes the 2D/3D Program in terms of the contributing efforts of the participants

Contributions to and expectations from the CRP - Argonne National Laboratory (USA)

International Nuclear Information System (INIS)

Cahalan, J.E.

2007-01-01

For us, the chief benefit of the CRP will be validation of multidimensional fluid dynamics capabilities for analysis of outlet plenum temperature distributions. As reactor designers seek new fuel handling features to reduce costs, upper internal structure configurations are becoming more compact, and higher fidelity analysis techniques are required to assess thermal stresses. Argonne currently has 1) a reactor systems analysis code with an experimentally-based model for plenum stratification, 2) the COMMIX code (parent of the JAEA AQUA code), and 3) commercial fluid dynamics analysis codes. It is anticipated that all or some combination of these capabilities will be employed to perform the CRP analysis

Hydrologic test results for the upper Cohassett flow interior at borehole RRL-2, Hanford Site, Washington State

International Nuclear Information System (INIS)

Strait, S.R.; Spane, F.A. Jr.

1984-03-01

The results and description of hydrologic test activities for the upper Cohassett flow interior at borehole RRL-2 over the depth interval 3,057 to 3,172 feet are presented in this report. Hydrologic tests conducted include an over-pressure pulse test and a constant head injection test. Preliminary results from hydrologic tests performed indicate transmissivity values ranging from 1.8 x 10 -6 to 1.7 x 10 -4 square feet per day, with an assigned best estimate of 1.7 x 10 -4 square feet per day. The best estimates of equivalent hydraulic conductivity, based on a thickness for the effective test interval of 115 feet, is 1.5 x 10 -6 feet per day. Best-estimate values obtained from testing are consistent with results previously reported for similar Grande Ronde Basalt horizons. 12 refs., 6 figs., 3 tabs

Reliability and validity of a low load endurance strength test for upper and lower extremities in patients with fibromyalgia.

Science.gov (United States)

Munguía-Izquierdo, Diego; Legaz-Arrese, Alejandro

2012-11-01

To evaluate the reliability, standard error of the mean (SEM), clinical significant change, and known group validity of 2 assessments of endurance strength to low loads in patients with fibromyalgia syndrome (FS). Cross-sectional reliability and comparative study. University Pablo de Olavide, Seville, Spain. Middle-aged women with FS (n=95) and healthy women (n=64) matched for age, weight, and body mass index (BMI) were recruited for the study. Not applicable. The endurance strength to low loads tests of the upper and lower extremities and anthropometric measures (BMI) were used for the evaluations. The differences between the readings (tests 1 and 2) and the SDs of the differences, intraclass correlation coefficient (ICC) model (2,1), 95% confidence interval for the ICC, coefficient of repeatability, intrapatient SD, SEM, Wilcoxon signed-rank test, and Bland-Altman plots were used to examine reliability. A Mann-Whitney U test was used to analyze the differences in test values between the patient group and the control group. We hypothesized that patients with FS would have an endurance strength to low loads performance in lower and upper extremities at least twice as low as that of the healthy controls. Satisfactory test-retest reliability and SEMs were found for the lower extremity, dominant arm, and nondominant arm tests (ICC=.973-.979; P.05 for all). The Bland-Altman plots showed 95% limits of agreement for the lower extremity (4.7 to -4.5), dominant arm (3.8 to -4.4), and nondominant arm (3.9 to -4.1) tests. The endurance strength to low loads test scores for the patients with FS were 4-fold lower than for the controls in all performed tests (P<.001 for all). The endurance strength to low loads tests showed good reliability and known group validity and can be recommended for evaluating endurance strength to low loads in patients with FS. For individual evaluation, however, an improved score of at least 4 and 5 repetitions for the upper and lower extremities

Detailed flow analysis for the Three Mile Island unit 2 reactor accident

International Nuclear Information System (INIS)

Lillington, J.N.; Lyons, A.J.

1990-01-01

Some particular characteristics of the steam flow in the accident at the Three Mile Island unit 2 pressurized water reactor are investigated using the AEA Technology Flow3D code. Natural circulation flows with heat removal from the core and deposition in the upper plenum are predicted during the primary heating phase. The structure of the upper plenum cylinder and core blockage, owing to material relocation, are shown to force the flow into a complex three-dimensional pattern. The flows and temperature distributions from the calculations are shown to be consistent with the observed damage pattern above the core. Despite high core temperatures, damage was limited by the operation of one of the pumps at the end of the initial heating phase. Flow3D calculations are also carried out to demonstrate that the three-dimensional buoyancy driven flows are completely destroyed by the high steam generation rates arising from the pump operation. (author)

Diagnosis and treatment of upper limb apraxia

OpenAIRE

Dovern, A.; Fink, G. R.; Weiss, P. H.

2012-01-01

Upper limb apraxia, a disorder of higher motor cognition, is a common consequence of left-hemispheric stroke. Contrary to common assumption, apraxic deficits not only manifest themselves during clinical testing but also have delirious effects on the patients’ everyday life and rehabilitation. Thus, a reliable diagnosis and efficient treatment of upper limb apraxia is important to improve the patients’ prognosis after stroke. Nevertheless, to date, upper limb apraxia is still an underdiagnosed...

In-place HEPA filter penetration test

Energy Technology Data Exchange (ETDEWEB)

Bergman, W.; Wilson, K.; Elliott, J. [Lawrence Livermore National Lab., CA (United States)] [and others

1997-08-01

We have demonstrated the feasibility of conducting penetration tests on high efficiency particulate air (HEPA) filters as installed in nuclear ventilation systems. The in-place penetration test, which is designed to yield equivalent penetration measurements as the standard DOP efficiency test, is based on measuring the aerosol penetration of the filter installation as a function of particle size using a portable laser particle counter. This in-place penetration test is compared to the current in-place leak test using light scattering photometers for single HEPA filter installations and for HEPA filter plenums using the shroud method. Test results show the in-place penetration test is more sensitive than the in-place leak test, has a similar operating procedure, but takes longer to conduct. Additional tests are required to confirm that the in-place penetration test yields identical results as the standard dioctyl phthalate (DOP) penetration test for HEPA filters with controlled leaks in the filter and gasket and duct by-pass leaks. Further development of the procedure is also required to reduce the test time before the in-place penetration test is practical. 14 refs., 14 figs., 3 tabs.

Temperature Trend Detection in Upper Indus Basin by Using Mann-Kendall Test

Directory of Open Access Journals (Sweden)

Ateeq Ur Rauf

2016-10-01

Full Text Available Global warming and Climate change are commonly acknowledged as the most noteworthy environmental quandary the world is undergoing today. Contemporary studies have revealed that the Earth’s surface air temperature has augmented by 0.6°C – 0.8°C in the course of the 20th century, together with alterations in the hydrological cycle. This study focuses on detecting trends in seasonal temperature for the five selected stations in the Upper Indus Basin. The Mann-Kendall test was run at 5% significance level on time series data for each of the five stations during the time period, 1985 to 2014. The Standard Test Statistic (Zs indicates the presence of trend and whether it is increasing or decreasing. The analysis showed an increasing trend in mean monthly temperature at Astore, Gilgit and Gupiz in March and a decreasing trend for Astore, Drosh, Gilgit and Skardu in September. Gilgit and Gupiz showed unexpected increasing trend in October. This study concludes that the temperature starts increasing in March and stays elevated till the month of June and starts rising again in October thus resulting in expansion of summer season and prolonged glacial melting.

Upper High School Students' Understanding of Electromagnetism

Science.gov (United States)

Saglam, Murat; Millar, Robin

2006-01-01

Although electromagnetism is an important component of upper secondary school physics syllabuses in many countries, there has been relatively little research on students' understanding of the topic. A written test consisting of 16 diagnostic questions was developed and used to survey the understanding of electromagnetism of upper secondary school…

Simulation and uncertainties of the heat transfer from a heat-generating DEBRIS bed in the lower plenum

International Nuclear Information System (INIS)

Schaaf, K.; Trambauer, K.

1999-01-01

The findings of the TMI-2 post-accident analyses indicated that internal cooling mechanisms may have a considerable potential to sustain the vessel integrity after a relocation of core material to the lower plenum, provided that water is continuously available in the RPV. Numerous analytical and experimental research activities are currently underway in this respect. This paper illustrates some major findings of the experimental work on internal cooling mechanisms and describes the limitations and the uncertainties in the simulation of the heat transfer processes. Reference is made especially to the joint German DEBRIS/ RPV research program, which encompasses the experimental investigation of the thermal-hydraulics in gaps, of the heat transfer within a particulate debris bed, and of the high temperature performance of vessel steel, as well as the development of simulation models for the heat transfer in the lower head and the structural response of the RPV. In particular, the results of uncertainty and sensitivity analyses are presented, which have been carried out at GRS using an integral model that describes the major phenomena governing the long-term integrity of the reactor vessel. The investigation of a large-scale relocation indicated that the verification of a gap cooling mechanism as an inherent mechanism is questionable in terms of a stringent probabilistic uncertainty criterion, as long as the formation of a large molten pool cannot be excluded. (author)

Investigation of reflood models by coupling REFLA-1D and multi-loop system model

International Nuclear Information System (INIS)

Sugimoto, Jun; Murao, Yoshio

1983-09-01

A system analysis code REFLA-1DS was developed by coupling reflood analysis code REFLA-1D and a multi-loop primary system model. The reflood models in the code were investigated for the development of the integral system analysis code. The REFLA-1D, which was developed with the small scale reflood experiment at JAERI, consists of one-dimensional core model and a primary system model with a constant loop resistance. The multi-loop primary system model was developed with the Cylindrical Core Test Facility of JAERI's large scale reflood tests. The components modeled in the code are the upper plenum, the steam generator, the coolant pump, the ECC injection port, the downcomer and the broken cold leg nozzle. The coupling between the two models in REFLA-1DS is accomplished by applying the equivalent flow resistance calculated with the multiloop model to the REFLA-1D. The characteristics of the code is its simplicity of the system model and the solution method which enables the fast running and the easy reflood analysis for the further model development. A fairly good agreement was obtained with the results of the Cylindrical Core Test Facility for the calculated water levels in the downcomer, the core and the upper plenum. A qualitatively good agreement was obtained concerning the parametric effects of the system pressure, the ECC flow rate and the initial clad temperature. Needs for further code improvements of the models, however, were pointed out. These include the problem concerning the generation rate of the steam and water droplets in the core in an early period, the effect of the flow oscillation on the core cooling, the heat release from the downcomer wall, and the stable system calculation. (author)

Model validation using CFD-grade experimental database for NGNP Reactor Cavity Cooling Systems with water and air

Energy Technology Data Exchange (ETDEWEB)

Manera, Annalisa [Univ. of Michigan, Ann Arbor, MI (United States); Corradini, Michael [Univ. of Wisconsin, Madison, WI (United States); Petrov, Victor [Univ. of Michigan, Ann Arbor, MI (United States); Anderson, Mark [Univ. of Wisconsin, Madison, WI (United States); Tompkins, Casey [Univ. of Wisconsin, Madison, WI (United States); Nunez, Daniel [Univ. of Michigan, Ann Arbor, MI (United States)

2018-02-13

This project has been focused on the experimental and numerical investigations of the water-cooled and air-cooled Reactor Cavity Cooling System (RCCS) designs. At this aim, we have leveraged an existing experimental facility at the University of Wisconsin-Madison (UW), and we have designed and built a separate effect test facility at the University of Michigan. The experimental facility at UW has underwent several upgrades, including the installation of advanced instrumentation (i.e. wire-mesh sensors) built at the University of Michigan. These provides highresolution time-resolved measurements of the void-fraction distribution in the risers of the water-cooled RCCS facility. A phenomenological model has been developed to assess the water cooled RCCS system stability and determine the root cause behind the oscillatory behavior that occurs under normal two-phase operation. Testing under various perturbations to the water-cooled RCCS facility have resulted in changes in the stability of the integral system. In particular, the effects on stability of inlet orifices, water tank volume have and system pressure been investigated. MELCOR was used as a predictive tool when performing inlet orificing tests and was able to capture the Density Wave Oscillations (DWOs) that occurred upon reaching saturation in the risers. The experimental and numerical results have then been used to provide RCCS design recommendations. The experimental facility built at the University of Michigan was aimed at the investigation of mixing in the upper plenum of the air-cooled RCCS design. The facility has been equipped with state-of-theart high-resolution instrumentation to achieve so-called CFD grade experiments, that can be used for the validation of Computational Fluid Dynanmics (CFD) models, both RANS (Reynold-Averaged) and LES (Large Eddy Simulations). The effect of risers penetration in the upper plenum has been investigated as well.

Reliability of the Quality of Upper Extremity Skills Test for Children with Cerebral Palsy Aged 2 to 12 Years

Science.gov (United States)

Thorley, Megan; Lannin, Natasha; Cusick, Anne; Novak, Iona; Boyd, Roslyn

2012-01-01

Aim: To investigate reliability of the Quality of Upper Extremity Skills Test (QUEST) scores for children with cerebral palsy (CP) aged 2-12 years. Method: Thirty-one QUESTs from 24 children with CP were rated once by two raters and twice by one rater. Internal consistency of total scores, inter- and intra-rater reliability findings for total,…

The influence of a real job on upper limb performance in motor skill tests: which abilities are transferred?

Science.gov (United States)

Giangiardi, Vivian Farahte; Alouche, Sandra Regina; de Freitas, Sandra Maria Sbeghen Ferreira; Pires, Raquel Simoni; Padula, Rosimeire Simprini

2018-06-01

To investigate whether the specificities of real jobs create distinctions in the performance of workers in different motor tests for the upper limbs, 24 participants were divided into two groups according to their specific job: fine and repetitive tasks and general tasks. Both groups reproduced tasks related to aiming movements, handling and strength of the upper limbs. There were no significant differences between groups in the dexterity and performance of aiming movements. However, the general tasks group had higher grip strength than the repetitive tasks group, demonstrating differences according to job specificity. The results suggest that a particular motor skill in a specific job cannot improve performance in other tasks with the same motor requirements. The transfer of the fine and gross motor skills from previous experience in a job-specific task is the basis for allocating training and guidance to workers.

Nodalization qualification process of the PSBVVER facility for the Cathare2 thermal-hydraulic code

International Nuclear Information System (INIS)

Del Nevo, A.; Araneo, D.; D'Auria, F.; Galassi, G.

2004-01-01

The present document deals with the nodalization qualification process of the PSB-VVER test facility for Cathare2 code. PSB-VVER facility is a 1/300 volume scale model of a VVER-1000, reactor installed at Electrogorsk Research and Engineering Centre in 1998. The version V1.5b of the Cathare2 code has been used. In order to evaluate the nodalization performance, the qualifying procedure set up at the DIMNP of Pisa University (UNIPI) has been applied that foresees two qualification levels: a 'steady state' level and an 'on transient' level. After the steady state behavior check of the nodalization, it has been preformed the on transient qualification the PSB-VVER test 2. It is a 11% equivalent break in Upper Plenum with the actuation of one high pressure injection system, connected to the hot leg of the loop 4, and 4 passive systems (ECCS hydro-accumulators), connected to the outlet plenum and to the inlet chamber of the downcomer. The low-pressure injection system is not available in the test. The goal of this paper is to demonstrate that the first step of the nodalization qualification adopted for the PSB test analyses is achieved and the PSB facility input deck is available and ready to use. The quantitative accuracy of the performed calculation has been evaluated by using the FFT-BM tool developed at the University of Pisa.(author)

Falcon II seminar, Winfrith Technology Centre, 13-14 March 1991

International Nuclear Information System (INIS)

Bennett, P.J.; Bowsher, B.R.

1991-05-01

Falcon was designed to study the transport of fission products released from both simulant and trace-irradiated fuel through a circuit simulating the upper plenum, hot-leg structures and the containment. Various sophisticated analytical techniques were used to provide information on the chemical species and physical forms of the released material. Twenty integral experiments of increasing complexity were successfully completed. The initial experiments were designed to characterise the thermal-hydraulics of the system, while the final tests included: (i) aerosols from Ag-In-Cd control rod and boric acid, (ii) a source of fission products, (iii) a painted surface and aqueous sump in the containment vessel to study iodine chemistry. (author)

Reflood behavior at low initial clad temperature in Slab Core Test Facility Core-II

International Nuclear Information System (INIS)

Akimoto, Hajime; Sobajima, Makoto; Abe, Yutaka; Iwamura, Takamichi; Ohnuki, Akira; Okubo, Tsutomu; Murao, Yoshio; Okabe, Kazuharu; Adachi, Hiromichi.

1990-07-01

In order to study the reflood behavior with low initial clad temperature, a reflood test was performed using the Slab Core Test Facility (SCTF) with initial clad temperature of 573 K. The test conditions of the test are identical with those of SCTF base case test S2-SH1 (initial clad temperature 1073 K) except the initial clad temperature. Through the comparison of results from these two tests, the following conclusions were obtained. (1) The low initial clad temperature resulted in the low differential pressures through the primary loops due to smaller steam generation in the core. (2) The low initial clad temperature caused the accumulated mass in the core to be increased and the accumulated mass in the downcomer to be decreased in the period of the lower plenum injection with accumulator (before 50s). In the later period of the cold leg injection with LPCI (after 100s), the water accumulation rates in the core and the downcomer were almost the same between both tests. (3) The low initial clad temperature resulted in the increase of the core inlet mass flow rate in the lower plenum injection period. However, the core inlet mass flow rate was almost the same regardless of the initial clad temperature in the later period of the cold leg injection period. (4) The low initial clad temperature resulted in the low turnaround temperature, high temperature rise and fast bottom quench front propagation. (5) In the region apart from the quench front, low initial clad temperature resulted in the lower heat transfer. In the region near the quench front, almost the same heat transfer coefficient was observed between both tests. (6) No flow oscillation with a long period was observed in the SCTF test with low initial clad temperature of 573 K, while it was remarkable in the Cylindrical Core Test Facility (CCTF) test which was performed with the same initial clad temperature. (J.P.N.)

Investigation on in-vessel thermal transients in a fast breeder reactor

International Nuclear Information System (INIS)

Muramatsu, Toshiharu; Kasahara, Naoto

1999-01-01

Thermal stratification phenomena are observed in an upper plenum of liquid metal fast breeder reactors (LMFBRs) under reactor scram conditions, which give rise to thermal stress on structural components. Therefore it is important to evaluate characteristics of the phenomena in the design of the internal structures in an LMFBR plenum. To evaluate thermal stress characteristics for the inner barrel in a typical LMFBR upper plenum, numerical analysis was carried out with a multi-dimensional thermohydraulics code AQUA for a scram condition from full power operation conditions. Thereafter, thermal stress conditions for the inner barrel were evaluated by the use of a structural analysis code FINAS with the thermohydraulic results calculated by the AQUA code as boundary conditions. From the thermohydraulic analysis and the thermal stress analysis, the following results have been obtained. (1) A large axial temperature gradient was calculated at the region between the upper and lower flow holes located on the inner barrel. The axial position of the thermal stratification interface was fixed in the various circumferential directions. As for the comparison with a 40% operation condition, maximum temperature gradients at the lower flow hole region indicated a 2 times value of that in the 40% operation condition. (2) Transient thermal stratification phenomena were observed after 120 sec from the reactor scram in the numerical results. These tendencies on thermal stratification phenomena were sameness with the transient results from the 40% operation condition. (3) During the reactor trip from full power operation, large temperature gradient in both vertical and sectional direction are enforced around the lower flow hole, since there exists flow pass of low temperature sodium through this hole. As a result, the maximum thermal stress within 32.6 kg/mm 2 was predicted at the lower flow hole when considering stress concentration at the hole edge. (J.P.N.)

TMI-2 core damage: a summary of present knowledge

International Nuclear Information System (INIS)

Owen, D.E.; Mason, R.E.; Meininger, R.D.; Franz, W.A.

1983-01-01

Extensive fuel damage (oxidation and fragmentation) has occurred and the top approx. 1.5 m of the center portion of the TMI-2 core has relocated. The fuel fragmentation extends outward to slightly beyond one-half the core radius in the direction examined by the CCTV camera. While the radial extent of core fragmentation in other directions was not directly observed, control and spider drop data and in-core instrument data suggest that the core void is roughly symmetrical, although there are a few indications of severe fuel damage extending to the core periphery. The core material fragmented into a broad range of particle sizes, extending down to a few microns. APSR movement data, the observation of damaged fuel assemblies hanging unsupported from the bottom of the reactor upper plenum structure, and the observation of once-molten stainless steel immediately above the active core indicate high temperatures (up to at least 1720 K) extended to the very top of the core. The relative lack of damage to the underside of the plenum structure implies a sharp temperature demarcation at the core/plenum interface. Filter debris and leadscrew deposit analyses indicate extensive high temperature core materials interaction, melting of the Ag-In-Cd control material, and transport of particulate control material to the plenum and out of the vessel

Multi-bundle sodium experiments for thermohydraulics in core subassemblies during natural circulation decay heat removal operation

International Nuclear Information System (INIS)

Kamide, H.; Ieda, Y.; Toda, S.; Isozaki, T.; Sugawara, S.

1993-01-01

Two types of multi-subassembly sodium experiments, CCTL-CFR tests and PLANDTL-DHX tests, have been carried out in order to investigate thermohydraulics in a fast reactor core during natural circulation. Basic experiments are carried out in CCTL-CFR test rig without inter-wrapper gap and under steady state. Integral experiments are performed in PLANDTL-DHX test rig with the inter-wrapper gap and a dip cooler in an upper plenum under steady state and transient conditions. The first series of the experiments and post analyses showed that inter-subassembly heat transfer had significant effects on the transverse temperature distribution in the subassembly and was strongly coupled with intra-subassembly flow redistribution. And the cold sodium provided by the dip cooler could reduce the hot spot temperature in the pin bundle mainly via the inter-wrapper gap. (author)

Multi-bundle sodium experiments for thermohydraulics in core subassemblies during natural circulation decay heat removal operation

Energy Technology Data Exchange (ETDEWEB)

Kamide, H; Ieda, Y; Toda, S; Isozaki, T; Sugawara, S [Reactor Engineering Section, O-arai Engineering Center, Power Reactor and Nuclear Fuel Development Corporation, Narita, O-arai, Ibaraki-ken (Japan)

1993-02-01

Two types of multi-subassembly sodium experiments, CCTL-CFR tests and PLANDTL-DHX tests, have been carried out in order to investigate thermohydraulics in a fast reactor coreduring natural circulation. Basic experiments are carried out in CCTL-CFR test rig without inter-wrapper gap and under steady state. Integral experiments are performed in PLANDTL-DHX test rig with the inter-wrapper gap and a dip cooler in an upper plenum under steady state and transient conditions. The first series of the experiments and post analyses showed that inter-subassembly heat transfer had significant effects on the transverse temperature distribution in the subassembly and was strongly coupled with intra-subassembly flow redistribution. And the cold sodium provided by the dip cooler could reduce the hot spot temperature in the pin bundle mainly via the inter-wrapper gap. (author)

Loss-of-coolant accident test series TC-1 experiment operating specifications

International Nuclear Information System (INIS)

Yackle, T.R.

1979-09-01

The purpose of this document is to specify the experiment operating procedure for the test series TC-1. The effects of externally mounted cladding thermocouples on the fuel rod thermal behavior during LOCA blowdown and reflood cycles will be investigated in the test. Potential thermocouple effects include: (a) delayed DNB, (b) momentary cladding rewets following DNB, (c) premature cladding rewet during a blowdown two-phase slug period, and (d) early cladding rewet during reflood. The two-phase slug period will be controlled by momentarily opening the hot leg valve. The slug will consist of lower plenum liquid that is sent through the flow shrouds and will be designed to quench the fuel rods at a rate that is similar to the slug experienced early in the LOFT L2-2 and L2-3 tests

Enhancing load-following and/or spectral shift capability in single-sparger natural circulation boiling water reactors

International Nuclear Information System (INIS)

Oosterkamp, W.J.

1992-01-01

This patent describes a method for obtaining load-following capability in a coiling water reactor (BWR) wherein housed within a reactor pressure vessel (RPV) is a nuclear core disposed within a shroud having a shroud head and which with the RPV defines an annulus region disposed beneath the nuclear core, an upper steam dome connected to a steam outlet in the RPV, a core upper plenum formed within the shroud head and disposed atop the nuclear core, a chimney mounted atop the shroud head and in fluid communication with the core upper plenum and with a steam separator having a skirt which is in fluid communication with the steam dome, the region outside of the chimney defining a downcomer region, there being a water level established therein under normal operation of the BWR, and the RPV containing a feedwater inlet. It comprises: disposing a single sparger connected to the feedwater inlet above the steam separator skirt bottom about the interior circumference of the RPV at an elevation at approximately the water level established during normal operation of the BWR; and adjusting the feedwater flow through the inlet and into the sparger to vary the water level to be above, at or below the elevational location of the sparger in response to load-following need

ITER ECRH Upper Launcher: Test plan for qualification of the Diamond Torus Window Prototype III

Energy Technology Data Exchange (ETDEWEB)

Schreck, Sabine, E-mail: sabine.schreck@kit.edu [Karlsruhe Institute of Technology, Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Aiello, Gaetano; Meier, Andreas; Strauss, Dirk [Karlsruhe Institute of Technology, Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Gagliardi, Mario; Saibene, Gabriella [F4E, Antennas and Plasma Engineering, Josep Pla 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Scherer, Theo [Karlsruhe Institute of Technology, Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

2016-11-01

Highlights: • A qualification program for the ITER diamond torus window is being developed. • The testing program for the qualification of the bare diamond disk is defined. • First qualification tests show a very good quality of the diamond disk prototypes. - Abstract: The diamond window is part of the electron cyclotron heating upper launcher system for ITER. Together with the isolation valve it constitutes the primary vacuum boundary and it also acts as first tritium barrier. Therefore the window is classified as Safety/Protection Important Component (SIC/PIC) with the nuclear safety function “confinement”. As the diamond window unit is not entirely covered by standard codes, an ad-hoc qualification program needs to be defined, including analysis, prototyping and testing. In the framework of a contract with F4E, the test program for a diamond window prototype is being developed with the aim to prove its operability for normal, accidental and incidental conditions as identified in the ITER load specifications. Tests range from dielectric loss measurements for the bare Chemical Vapour Deposition (CVD) diamond disk up to mechanical and vacuum tests for the complete window assembly. Finally mm-wave properties have to be characterized for the complete window. A clear definition of the testing requirements and of the acceptance criteria is necessary as well as a complete documentation of the process. This paper will present the development of the test plan for a window prototype, which is currently under manufacturing. First tests are directed to the characterization of the bare diamond disk with a focus on its dielectric properties.

Upper cervical and upper thoracic thrust manipulation versus nonthrust mobilization in patients with mechanical neck pain: a multicenter randomized clinical trial.

Science.gov (United States)

Dunning, James R; Cleland, Joshua A; Waldrop, Mark A; Arnot, Cathy F; Young, Ian A; Turner, Michael; Sigurdsson, Gisli

2012-01-01

Randomized clinical trial. To compare the short-term effects of upper cervical and upper thoracic high-velocity low-amplitude (HVLA) thrust manipulation to nonthrust mobilization in patients with neck pain. Although upper cervical and upper thoracic HVLA thrust manipulation and nonthrust mobilization are common interventions for the management of neck pain, no studies have directly compared the effects of both upper cervical and upper thoracic HVLA thrust manipulation to nonthrust mobilization in patients with neck pain. Patients completed the Neck Disability Index, the numeric pain rating scale, the flexion-rotation test for measurement of C1-2 passive rotation range of motion, and the craniocervical flexion test for measurement of deep cervical flexor motor performance. Following the baseline evaluation, patients were randomized to receive either HVLA thrust manipulation or nonthrust mobilization to the upper cervical (C1-2) and upper thoracic (T1-2) spines. Patients were reexamined 48-hours after the initial examination and again completed the outcome measures. The effects of treatment on disability, pain, C1-2 passive rotation range of motion, and motor performance of the deep cervical flexors were examined with a 2-by-2 mixed-model analysis of variance (ANOVA). One hundred seven patients satisfied the eligibility criteria, agreed to participate, and were randomized into the HVLA thrust manipulation (n = 56) and nonthrust mobilization (n = 51) groups. The 2-by-2 ANOVA demonstrated that patients with mechanical neck pain who received the combination of upper cervical and upper thoracic HVLA thrust manipulation experienced significantly (Ppain (58.5%) than those of the nonthrust mobilization group (12.8% and 12.6%, respectively) following treatment. In addition, the HVLA thrust manipulation group had significantly (Pcervical flexor muscles as compared to the group that received nonthrust mobilization. The number needed to treat to avoid an unsuccessful outcome

Experiment data of ROSA-III integral test Run 7341

International Nuclear Information System (INIS)

Anoda, Yoshinari; Soda, Kunihisa; Tasaka, Kanji

1983-02-01

This report presents the test data of Run 7341 in the single failure test series of the ROSA-III program to conduct the system effect test concerning the response of a BWR during a LOCA with the ECC injection. The ROSA-III test facility is a volumetrically scaled (1/424) system of the BWR/6 with an electrically heated core and the scaled ECCS. Run 7341 is a double-ended break test at the recirculation pump inlet with the assumption that all ECCS function as designed. The test is initiated with the steam dome pressure of 7.28 MPa, the lower plenum subcooling of 11.0 K, the core inlet flow rate of 15.3 kg/s, and the core heat generation rate of 3.55 MW and proceeded as planned. The whole core is quenched after the ECCS actuation and the maximum fuel cladding temperature is 810 K. The effectiveness of ECC injection has been clarified from the comparison of the test results with those of the other test in the series without ECCS. (author)

Low upper-shelf toughness, high transition temperature test insert in HSST [Heavy Section Steel Technology] PTSE-2 [Pressurized Thermal Shock Experiment-2] vessel and wide plate test specimens: Final report

International Nuclear Information System (INIS)

Domian, H.A.

1987-02-01

A piece of A387, Grade 22 Class 2 (2-1/4 Cr - 1 Mo) steel plate specially heat treated to produce low upper-shelf (LUS) toughness and high transition temperature was installed in the side wall of Heavy Section Steel Technology (HHST) vessel V-8. This vessel is to be tested by the Oak Ridge National Laboratory (ORNL) in the Pressurized Thermal Shock Experiment-2 (PTSE-2) project of the HSST program. Comparable pieces of the plate were made into six wide plate specimens and other samples. These samples underwent tensile tests, Charpy tests, and J-integral tests. The results of these tests are given in this report

Natural convection as the way of heat removal from fast reactor core at cooldown regimes

International Nuclear Information System (INIS)

Zhukov, A.V.; Kuzina, J.A.; Uhov, V.A.; Sorokin, G.A.

2000-01-01

The problems of thermohydraulics in fast reactors at cooldown regimes at heat removal by natural convection are considered The results of experiments and calculations obtained in various countries in this area are presented. The special attention is given to heat removal through inter-assembly space in the core and also to problems of thermohydraulics in the upper plenum. (author)

Diagnosis and treatment of upper limb apraxia.

Science.gov (United States)

Dovern, A; Fink, G R; Weiss, P H

2012-07-01

Upper limb apraxia, a disorder of higher motor cognition, is a common consequence of left-hemispheric stroke. Contrary to common assumption, apraxic deficits not only manifest themselves during clinical testing but also have delirious effects on the patients' everyday life and rehabilitation. Thus, a reliable diagnosis and efficient treatment of upper limb apraxia is important to improve the patients' prognosis after stroke. Nevertheless, to date, upper limb apraxia is still an underdiagnosed and ill-treated entity. Based on a systematic literature search, this review summarizes the current tools of diagnosis and treatment strategies for upper limb apraxia. It furthermore provides clinicians with graded recommendations. In particular, a short screening test for apraxia, and a more comprehensive diagnostic apraxia test for clinical use are recommended. Although currently only a few randomized controlled studies investigate the efficacy of different apraxia treatments, the gesture training suggested by Smania and colleagues can be recommended for the therapy of apraxia, the effects of which were shown to extend to activities of daily living and to persist for at least 2 months after completion of the training. This review aims at directing the reader's attention to the ecological relevance of apraxia. Moreover, it provides clinicians with appropriate tools for the reliable diagnosis and effective treatment of apraxia. Nevertheless, this review also highlights the need for further research into how to improve diagnosis of apraxia based on neuropsychological models and to develop new therapeutic strategies.

Validity of Alternative Fitnessgram Upper Body Tests of Muscular Strength and Endurance among Seventh and Eighth Grade Males and Females

Science.gov (United States)

Hobayan, Kalani; Patterson, Debra; Sherman, Clay; Wiersma, Lenny

2014-01-01

In a society in which obesity levels have tripled in the past 30 years, the importance of increased fitness levels within the academic setting has become even more critical. The purpose of this study was to investigate the validity of alternative Fitnessgram upper body tests of muscular strength and endurance among seventh and eighth grade males…

Application of UPTF data for modeling liquid draindown in the downcomer region of a PWR using RELAP5/MOD2-B&W

Energy Technology Data Exchange (ETDEWEB)

Wissinger, G.; Klingenfus, J. [B & W Nuclear Technologies, Lynchburg, VA (United States)

1995-09-01

B&W Nuclear Technologies (BWNT) currently uses an evaluation model that analyzes large break loss-of-coolant accidents in pressurized water reactors using several computer codes. These codes separately calculate the system performance during the blowdown, refill, and reflooding phases of the transient. Multiple codes are used, in part, because a single code has been unable to effectively model the transition from blowdown to reflood, particularly in the downcomer region where high steam velocities do not allow the injected emergency core cooling (ECC) liquid to penetrate and begin to refill the vessel lower plenum until after the end of blowdown. BWNT is developing a method using the RELAP5/MOD2-B&W computer code that can correctly predict the liquid draindown behavior in the downcomer during the late blowdown and refill phases. Benchmarks of this method have been performed against Upper Plenum Test Facility (UPTF) data for ECC liquid penetration and valves using both cold leg and downcomer ECC injection. The use of this new method in plant applications should result in the calculation of a shorter refill period, leading to lower peak clad temperature predictions and increased core peaking. This paper identifies changes made to the RELAP/MOD2-B&W code to improve its predictive capabilities with respect to the data obtained in the UPTF tests.

Seven pin bundle fast top tests L01 and L02

International Nuclear Information System (INIS)

Davies, A.L.; Bowen, G.R.; Herbert, R.; Kear, K.L.; Tylka, J.P.; Holland, J.W.

1984-01-01

Tests L01 and L02 were the first two seven pin bundle tests in the PFR/TREAT program of fuel failure tests carried out jointly by the US and the UK. The two tests were on bottom plenum annular pellet mixed oxide fuel clad in 316 stainless steel. L01 used fresh fuel, while L02 used PFR irradiated 4% burn-up fuel, to determine any differences in the failure mechanism and subsequent fuel behavior due to irradiation. They were performed in flowing sodium in the Mark IIIA version of a TREAT integral loop. Both were fast transient overpower (TOP) tests intended to simulate 5 $/s reactivity ramp hypothetical accidents in a large fast reactor. The test objectives were to obtain information on fuel motion in the central hole before failure, the time and location of cladding failures, and material motion in the channel after failure, having particular regard to the effect of irradiation

Construction of a 2- by 2-foot transonic adaptive-wall test section at the NASA Ames Research Center

Science.gov (United States)

Morgan, Daniel G.; Lee, George

1986-01-01

The development of a new production-size, two-dimensional, adaptive-wall test section with ventilated walls at the NASA Ames Research Center is described. The new facility incorporates rapid closed-loop operation, computer/sensor integration, and on-line interference assessment and wall corrections. Air flow through the test section is controlled by a series of plenum compartments and three-way slide vales. A fast-scan laser velocimeter was built to measure velocity boundary conditions for the interference assessment scheme. A 15.2-cm- (6.0-in.-) chord NACA 0012 airfoil model will be used in the first experiments during calibration of the facility.

Protocol: Testing the Relevance of Acupuncture Theory in the Treatment of Myofascial Pain in the Upper Trapezius Muscle.

Science.gov (United States)

Elsdon, Dale S; Spanswick, Selina; Zaslawski, Chris; Meier, Peter C

2017-01-01

A protocol for a prospective single-blind parallel four-arm randomized placebo-controlled trial with repeated measures was designed to test the effects of various acupuncture methods compared with sham. Eighty self-selected participants with myofascial pain in the upper trapezius muscle were randomized into four groups. Group 1 received acupuncture to a myofascial trigger point (MTrP) in the upper trapezius. Group 2 received acupuncture to the MTrP in addition to relevant distal points. Group 3 received acupuncture to the relevant distal points only. Group 4 received a sham treatment to both the MTrP and distal points using a deactivated acupuncture laser device. Treatment was applied four times within 2 weeks with outcomes measured throughout the trial and at 2 weeks and 4 weeks posttreatment. Outcome measurements were a 100-mm visual analog pain scale, SF-36, pressure pain threshold, Neck Disability Index, the Upper Extremity Functional Index, lateral flexion in the neck, McGill Pain Questionnaire, Massachusetts General Hospital Acupuncture Sensation Scale, Working Alliance Inventory (short form), and the Credibility Expectance Questionnaire. Two-way analysis of variance (ANOVA) with repeated measures were used to assess the differences between groups. Copyright © 2017 Medical Association of Pharmacopuncture Institute. Published by Elsevier B.V. All rights reserved.

Measurement of two-phase flow momentum with force transducers

International Nuclear Information System (INIS)

Hardy, J.E.; Smith, J.E.

1990-01-01

Two strain-gage-based drag transducers were developed to measure two-phase flow in simulated pressurized water reactor (PWR) test facilities. One transducer, a drag body (DB), was designed to measure the bidirectional average momentum flux passing through an end box. The second drag sensor, a break through detector (BTD), was designed to sense liquid downflow from the upper plenum to the core region. After prototype sensors passed numerous acceptance tests, transducers were fabricated and installed in two experimental test facilities, one in Japan and one in West Germany. High-quality data were extracted from both the DBs and BTDs for a variety of loss-of-coolant accident (LOCA) scenarios. The information collected from these sensors has added to the understanding of the thermohydraulic phenomena that occur during the refill/reflood stage of a LOCA in a PWR. 9 refs., 15 figs

Upper gastrointestinal bleeding in patients with CKD.

Science.gov (United States)

Liang, Chih-Chia; Wang, Su-Ming; Kuo, Huey-Liang; Chang, Chiz-Tzung; Liu, Jiung-Hsiun; Lin, Hsin-Hung; Wang, I-Kuan; Yang, Ya-Fei; Lu, Yueh-Ju; Chou, Che-Yi; Huang, Chiu-Ching

2014-08-07

Patients with CKD receiving maintenance dialysis are at risk for upper gastrointestinal bleeding. However, the risk of upper gastrointestinal bleeding in patients with early CKD who are not receiving dialysis is unknown. The hypothesis was that their risk of upper gastrointestinal bleeding is negatively linked to renal function. To test this hypothesis, the association between eGFR and risk of upper gastrointestinal bleeding in patients with stages 3-5 CKD who were not receiving dialysis was analyzed. Patients with stages 3-5 CKD in the CKD program from 2003 to 2009 were enrolled and prospectively followed until December of 2012 to monitor the development of upper gastrointestinal bleeding. The risk of upper gastrointestinal bleeding was analyzed using competing-risks regression with time-varying covariates. In total, 2968 patients with stages 3-5 CKD who were not receiving dialysis were followed for a median of 1.9 years. The incidence of upper gastrointestinal bleeding per 100 patient-years was 3.7 (95% confidence interval, 3.5 to 3.9) in patients with stage 3 CKD, 5.0 (95% confidence interval, 4.8 to 5.3) in patients with stage 4 CKD, and 13.9 (95% confidence interval, 13.1 to 14.8) in patients with stage 5 CKD. Higher eGFR was associated with a lower risk of upper gastrointestinal bleeding (P=0.03), with a subdistribution hazard ratio of 0.93 (95% confidence interval, 0.87 to 0.99) for every 5 ml/min per 1.73 m(2) higher eGFR. A history of upper gastrointestinal bleeding (Pupper gastrointestinal bleeding risk. In patients with CKD who are not receiving dialysis, lower renal function is associated with higher risk for upper gastrointestinal bleeding. The risk is higher in patients with previous upper gastrointestinal bleeding history and low serum albumin. Copyright © 2014 by the American Society of Nephrology.

Performance test for the compressor of 100kW APU

International Nuclear Information System (INIS)

Lim, Byeung Jun; Cha, Bong Jun; Yang, Soo Seok; Lee, Kyoung Jin; Baik, Ki Young

2001-01-01

The performance test of a centrifugal compressor for APU(Auxiliary Power Unit) which is developed by the collaborative research of KARI and Samsung TechWin has been conducted. The investigated compressor consists of a curved inlet, a centrifugal impeller, a channel diffuser and a plenum chamber. The experiments were carried out in an open-loop centrifugal compressor test rig driven by a turbine. For three different diffusers, overall performance data were obtained at 80%, 90% and 97% of design speed. For the initially designed wedge-type diffuser, test results showed that the compressor was operated at a higher mass flow rate than the design requirement. By reducing the diffuser throat area, the compressor operating range was shifted to lower mass flow rate range. The test result of redesigned wedge-type diffuser showed high pressure loss. To reduce the diffuser loss, diffuser inlet radius was increased and airfoil-type of diffuser was adopted. This airfoil-type diffuser showed reasonal results in terms of design requirement

Construction, Geology, and Aquifer Testing of the Maalo Road, Aahoaka Hill, and Upper Eleele Tank Monitor Wells, Kauai, Hawaii

Science.gov (United States)

Izuka, Scot K.

2005-01-01

The Maalo Road, Aahoaka Hill, and Upper Eleele Tank monitor wells were constructed using rotary drilling methods between July 1998 and August 2002 as part of a program of exploratory drilling, aquifer testing, and hydrologic analysis on Kauai. Aquifer tests were conducted in the uncased boreholes of the wells. The Maalo Road monitor well in the Lihue Basin penetrated 915 feet, mostly through mafic lava flows. Most of the rock samples from this well had chemical compositions similar to the Koloa Volcanics, but the deepest sample analyzed had a composition similar to the Waimea Canyon Basalt. Water temperature ranged from 25.6 to 27.4 degrees Celsius and specific conductance ranged from 303 to 627 microsiemens per centimeter during aquifer testing. Discharge rate ranged from 174 to 220 gallons per minute and maximum drawdown was 138.25 ft during a 7-day sustained-discharge test, but the test was affected by pump and generator problems. The Aahoaka Hill monitor well in the Lihue Basin penetrated 804 feet, mostly through mafic lava flows and possibly dikes. The well penetrated rocks having chemical compositions similar to the Waimea Canyon Basalt. During the first three hours of a sustained-discharge aquifer test in which the discharge rate varied between 92 and 117 gallons per minute, water temperature was 24.6 to 25.6 degrees Celsius, and specific conductance was 212 to 238 microsiemens per centimeter; this test was halted after a short period because drawdown was high. In a subsequent 7-day test, discharge was 8 to 23 gallons per minute, and maximum drawdown was 37.71 feet after 1,515 minutes of testing. The Upper Eleele Tank monitor well is near the Hanapepe River Valley. The well penetrated 740 feet through soil, sediment, mafic lava flows, volcanic ash, and scoria. Rocks above a depth of 345 feet had compositions similar to the Koloa Volcanics, but a sample from 720 to 725 feet had a composition similar to rocks of the Waimea Canyon Basalt. During a 7-day aquifer

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.

Analysis of a simulated small break in the semiscale system under loss-of-coolant accident conditions

International Nuclear Information System (INIS)

Cartmill, C.E.

1978-01-01

The Semiscale Mod-1 experimental program conducted by EG and G Idaho, Inc., is part of the overall U.S. Nuclear Regulatory Commission (NRC) and Department of Energy (DOE) sponsored research and development program to investigate the behavior of the pressurized water reactor (PWR) system during an hypothesized loss-of-coolant accident (LOCA). The Semiscale Mod-1 program is intended to provide transient thermal-hydraulic data from a simulated LOCA using a small-scale experimental nonnuclear system. The Semiscale Mod-1 program is a major contributor of experimental data that provide a means of evaluating the adequacy of overall system analytical models as well as the models of the individual system components. Selected experimental data produced by this program will also be used to aid other DOE and NRC sponsored experimental programs, such as the Loss-of-Fluid Test (LOFT) program in optimizing test series, selecting test parameters, and evaluating test results. The Semiscale Mod-1 tests are performed with an experimental system which simulates the principal features of a nuclear plant but which is smaller in volume. Nuclear heating is simulated in the tests by a core composed of an array of electrically heated rods. The core is contained in a pressure vessel which also includes a downcomer, lower plenum, and upper plenum. The Semiscale system piping is arranged such that the intact loop represents three loops of a four-loop nuclear plant, and the broken loop represents the fourth loop. In the present configuration the intact loop contains an active steam generator and pump, and the broken loop contains passive simulators for the steam generator and pump

2D/3D Program work summary report

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-09-01

The 2D/3D Program was carried out by Germany, Japan and the United States to investigate the thermal-hydraulics of a PWR large-break LOCA. A contributory approach was utilized in which each country contributed significant effort to the program and all three countries shared the research results. Germany constructed and operated the Upper Plenum Test Facility (UPTF), and Japan constructed and operated the Cylindrical Core Test Facility (CCTF) and the Slab Core Test Facility (SCTF). The US contribution consisted of provision of advanced instrumentation to each of the three test facilities, and assessment of the TRAC computer code against the test results. Evaluations of the test results were carried out in all three countries. This report summarizes the 2D/3D Program in terms of the contributing efforts of the participants, and was prepared in a coordination among three countries. US and Germany have published the report as NUREG/IA-0126 and GRS-100, respectively. (author).

2D/3D Program work summary report

International Nuclear Information System (INIS)

1995-09-01

The 2D/3D Program was carried out by Germany, Japan and the United States to investigate the thermal-hydraulics of a PWR large-break LOCA. A contributory approach was utilized in which each country contributed significant effort to the program and all three countries shared the research results. Germany constructed and operated the Upper Plenum Test Facility (UPTF), and Japan constructed and operated the Cylindrical Core Test Facility (CCTF) and the Slab Core Test Facility (SCTF). The US contribution consisted of provision of advanced instrumentation to each of the three test facilities, and assessment of the TRAC computer code against the test results. Evaluations of the test results were carried out in all three countries. This report summarizes the 2D/3D Program in terms of the contributing efforts of the participants, and was prepared in a coordination among three countries. US and Germany have published the report as NUREG/IA-0126 and GRS-100, respectively. (author)

Post-test analysis with RELAP5/MOD2 of ROSA-IV/LSTF natural circulation test ST-NC-02

International Nuclear Information System (INIS)

Chauliac, C.; Kukita, Yutaka; Kawaji, Masahiro; Nakamura, Hideo; Tasaka, Kanji.

1988-10-01

Results of post-test analysis for the ROSA-IV/LSTF natural circulation experiment ST-NC-02 are presented. The experiment consisted of many steady-state stages registered for different primary inventories. The calculation was done with RELAP5/MOD2 CYCLE 36.00. Discrepancies between the calculation and the experiment are observed: the core flow rate is overestimated at inventories between 80 % and 95 %; the inventory at which dryout occurs in the core is also much overestimated. The causes of these discrepancies are studies through sensitivity calculations and the following key parameters are pointed out: the interfacial friction and the form loss coefficients in the vessel riser, the SG U-tube multidimensional behaviour, the interfacial friction in the SG inlet plenum and in the pipe located underneath. (author)

Thermal-hydraulic mixing in the split-core ANS reactor design

International Nuclear Information System (INIS)

Dorning, R.J.J.

1988-01-01

A design has been proposed for the advanced neutron source (ANS) reactor that incorporates a split core, one purpose of which is to create a mixing plenum between the upper and lower cores. It was hoped that in addition to introducing various desirable neutronics features, such as decreasing the fast neutron flux contamination of thermal and cold neutron beams located in the reactor midplane, this mixing plenum would make possible higher operating powers by lowering the maximum core temperature. This lower temperature was to be achieved as a result of the mixing, of the hot D 2 O coolant exiting the upper-core channels, and the cold D 2 O leaving the large upper core bypass. It was expected that this mixing would bring about a significantly reduced lower core maximum coolant inlet temperature. The authors have carried out large-scale computer calculations to determine the extent to which this mixing occurs in current split-core design geometry, which does not incorporate baffles, mixing devices, or other design features introduced to enhance mixing. The large-scale self-consistent calculations summarized here indicate that innovative design ideas to enhance mixing will be necessary if the split-core concept is to achieve the amount of thermal mixing needed to make possible significantly higher power operation and corresponding higher flux sources

Reliability and Sensitivity of the Power Push-up Test for Upper-Body Strength and Power in 6-15-Year-Old Male Athletes.

Science.gov (United States)

Gillen, Zachary M; Miramonti, Amelia A; McKay, Brianna D; Jenkins, Nathaniel D M; Leutzinger, Todd J; Cramer, Joel T

2018-01-01

Gillen, ZM, Miramonti, AA, McKay, BD, Jenkins, NDM, Leutzinger, TJ, and Cramer, JT. Reliability and sensitivity of the power push-up test for upper-body strength and power in 6-15-year-old male athletes. J Strength Cond Res 32(1): 83-96, 2018-The power push-up (PPU) test is an explosive upper-body test performed on a force plate and is currently being used in high school football combines throughout the United States. The purpose of this study was to quantify the reliability of the PPU test based on age and starting position (knees vs. toes) in young athletes. Sixty-eight boys (mean ± SD; age = 10.8 ± 2.0 years) were tested twice over 5 days. Boys were separated by age as 6-9 years (n = 16), 10-11 years (n = 26), and 12-15 years (n = 26). The PPU test was performed on a force plate while rotating from the knees vs. the toes. Measurements were peak force (PF, N), peak rate of force development (pRFD, N·s), average power (AP, W), and peak power (PP, W). Intraclass correlation coefficients (ICC2,1), SEMs, coefficients of variation (CVs), and minimum detectable changes (MDCs) were calculated to quantify reliability and sensitivity. Peak force from the knees in 10-15-year-olds, PF from the toes in 12-15-year-olds, and pRFD from the knees and toes in 12-15-year-olds were comparably reliable (ICC ≥ 0.84). Neither power measurements (AP or PP) for any age group, nor any measurements (PF, pRFD, AP, or PP) for the 6-9-year-olds were comparably reliable (ICC ≤ 0.74). When considering the reliable variables, PF was greater in the 12-15-year-olds than in 10-11-year-olds (p ≤ 0.05). In addition, in 12-15-year-olds, PF and pRFD were greater from the knees than from the toes (p ≤ 0.05). For reasons largely attributable to growth and development, the PPU test may be a reliable (ICC ≥ 0.80) and sensitive (CV ≤ 19%) measure of upper-body strength (PF), whereas pRFD was also reliable (ICC ≥ 0.80), but less sensitive (CV = 30-38%) in 10-15-year-old male athletes.

Emergency core cooling system

International Nuclear Information System (INIS)

Kato, Ken.

1989-01-01

In PWR type reactors, a cooling water spray portion of emergency core cooling pipelines incorporated into pipelines on high temperature side is protruded to the inside of an upper plenum. Upon rupture of primary pipelines, pressure in a pressure vessel is abruptly reduced to generate a great amount of steams in the reactor core, which are discharged at a high flow rate into the primary pipelines on high temperature side. However, since the inside of the upper plenum has a larger area and the steam flow is slow, as compared with that of the pipelines on the high temperature side, ECCS water can surely be supplied into the reactor core to promote the re-flooding of the reactor core and effectively cool the reactor. Since the nuclear reactor can effectively be cooled to enable the promotion of pressure reduction and effective supply of coolants during the period of pressure reduction upon LOCA, the capacity of the pressure accumulation vessel can be decreased. Further, the re-flooding time for the reactor is shortened to provide an effect contributing to the improvement of the safety and the reduction of the cost. (N.H.)

Measurement on the effect of sound wave in upper plenum of boiling water reactor

International Nuclear Information System (INIS)

Kumagai, Kosuke; Someya, Satoshi; Okamoto, Koji

2009-01-01

In recent years, the power uprate of Boiling Water Reactors have been conducted at several existing power plants as a way to improve plant economy. In one of the power uprated plants (117.8% uprates) in the United States, the steam dryer breakages due to fatigue fracture occurred. It is conceivable that the increased steam flow passing through the branches caused a self-induced vibration with the propagation of sound wave into the steam-dome. The resonance among the structure, flow and the pressure fluctuation resulted in the breakages. To understand the basic mechanism of the resonance, previous researches were done by a point measurement of the pressure and by a phase averaged measurement of the flow, while it was difficult to detect the interaction among them by the conventional method. In this study, Dynamic Particle Image Velocimetry (PIV) System was applied to investigate the effect of sound on natural convection and forced convection. Especially, when the phases of acoustic sources were different, various acoustic wave effects were checked. (author)

Study of development of non-destructive method for determining FGR from high burned PWR type fuel rod

International Nuclear Information System (INIS)

Yanagisawa, Kazuaki; Miyanishi, Hideyuki; Kitagawa, Isamu; Iida, Shozo; Ito, Tadaharu; Amano, Hidetoshi.

1991-11-01

Experimental study was made to evaluate the FGR (Fission Product Gas Release) from high burned PWR type fuel rods by means of non-destructive method through measurement of the gamma activity of 85 Kr isotope which was accumulated in the fuel top plenum. Experimental result shows that it is possible to know the amounts of FGR at fuel plenum by the equations given in the followings. FGR = 0.28C/V f or FGR = 0.07C where, FGR (%) is the amounts of Xe and Kr released from UO 2 fuel, C (counts/h) the radioactivity of 85 Kr at plenum of the tested fuel rod and V f (ml) the plenum volume of the tested fuel rod, respectively. The present study was made by using 14 x 14 PWR type fuel rods preirradiated up to the burn-up of 42.1 MWd/kgU, followed by the pulse irradiation at Nuclear Safety Research Reactor of Japan Atomic Energy Research Institute (JAERI). The FGR of the tested segmented fuel rods were measured by puncturing and found to range from 0.6% to 12% according to the magnitude of the deposited energy given by pulse. Estimated experimental error bands against the above equations were within plus minus 30%. (author)

Analysis of turbulent natural convection heat transfer in a lower plenum during external cooling using the COSMO code

Energy Technology Data Exchange (ETDEWEB)

Noguchi, H. [Nuclear Power Engineering Corp., Tokyo (Japan); Sawatari, Y.; Imada, T. [Fuji Research Institute Corporation, Tokyo (Japan)

2000-11-01

The behavior of a large volumetrically heated melt pool is important to evaluate the feasibility of in-vessel retention by external flooding as an accident management. The COSMO (Coolability Simulation of Molten corium during severe accident) code has been developed at NUPEC to simulate turbulent natural convection heat transfer with internal heat source. The COSMO code solves thermal hydraulic conservation equations with turbulent model and can simulate melting and solidification process. The standard k-{epsilon} model has a limitation to describe the turbulent natural convection in the very high Rayleigh number condition (10{sup 16}-10{sup 17}) assumed to occur in a lower plenum of RPV during a severe accident. This limitation results from the assumption of an analogy of momentum and energy transfer phenomena in the standard model. In this paper the modified turbulent model in which the turbulent number is treated, as a function of the flux Richardson number derived from the experiment, has been incorporated and verified by using the BALI experiments. It was found that the prediction of averaged Nusselt number became better than that of the standard model. In order to extend the COSMO code to the actual scale analysis under the external flooding conditions, more realistic boundary condition derived from the experiments should be treated. In this work the CHF correlation from ULPU experiment or the heat transfer coefficient correlation from CYBL experiment have been applied. The preliminary analysis of an actual scale analysis has been carried out under the condition of the TMI-2 accident. (author)

Analysis of turbulent natural convection heat transfer in a lower plenum during external cooling using the COSMO code

International Nuclear Information System (INIS)

Noguchi, H.; Sawatari, Y.; Imada, T.

2000-01-01

The behavior of a large volumetrically heated melt pool is important to evaluate the feasibility of in-vessel retention by external flooding as an accident management. The COSMO (Coolability Simulation of Molten corium during severe accident) code has been developed at NUPEC to simulate turbulent natural convection heat transfer with internal heat source. The COSMO code solves thermal hydraulic conservation equations with turbulent model and can simulate melting and solidification process. The standard k-ε model has a limitation to describe the turbulent natural convection in the very high Rayleigh number condition (10 16 -10 17 ) assumed to occur in a lower plenum of RPV during a severe accident. This limitation results from the assumption of an analogy of momentum and energy transfer phenomena in the standard model. In this paper the modified turbulent model in which the turbulent number is treated, as a function of the flux Richardson number derived from the experiment, has been incorporated and verified by using the BALI experiments. It was found that the prediction of averaged Nusselt number became better than that of the standard model. In order to extend the COSMO code to the actual scale analysis under the external flooding conditions, more realistic boundary condition derived from the experiments should be treated. In this work the CHF correlation from ULPU experiment or the heat transfer coefficient correlation from CYBL experiment have been applied. The preliminary analysis of an actual scale analysis has been carried out under the condition of the TMI-2 accident. (author)

Evaluation of aseismic integrity in HTTR core-bottom structure. Pt. 1. Aseismic test for core-bottom structure

International Nuclear Information System (INIS)

Iyoku, T.; Futakawa, M.; Ishihara, M.

1994-01-01

The aseismic tests were carried out using (1)/(5)-scale and (1)/(3)-scale models of the core-bottom structure of the HTTR to quantitatively evaluate the response of acceleration, strain, impact load etc. The following conclusions are obtained. (i) The frequency response of the keyway strain is correlative with that of the impact acceleration on the hot plenum block. (ii) It was confirmed through (1)/(5)-scale and (1)/(3)-scale model tests that the applied similarity law is valid to evaluate the seismic response characteristics of the core-bottom structure. (ii) The stress of graphite components estimated from the scale model test using S 2 -earthquake excitation was sufficiently lower than the allowable stress used as the design criterion. ((orig.))

Revisiting large break LOCA with the CATHARE-3 three-field model

International Nuclear Information System (INIS)

Valette, Michel; Pouvreau, Jérôme; Bestion, Dominique; Emonot, Philippe

2011-01-01

Highlights: ► CATHARE 3 enables a three-field analysis of a LB LOCA. ► Reflooding experiments in isolated rod bundles are satisfactory predicted. ► A BETHSY integral test simulation supports the CATHARE 3 3-field assessment. - Abstract: Some aspects of large break LOCA analysis (steam binding, oscillatory reflooding, top-down reflooding) are expected to be improved in advanced system codes from more detailed description of flows by adding a third field for droplets. The future system code CATHARE-3 is under development by CEA and supported by EDF, AREVA-NP and IRSN in the frame of the NEPTUNE project and this paper shows some preliminary results obtained in reflooding conditions. A three-field model has been implemented, including vapor, continuous liquid and liquid droplet fields. This model features a set of nine equations of mass, momentum and energy balance. Such a model allows a more detailed description of the droplet transportation from core to steam generator, while countercurrent flow of continuous liquid is allowed. Code assessment against reflooding experiments in a rod bundle is presented, using 1D meshing of the bundle. Comparisons of CATHARE-3 simulations against data series from PERICLES and RBHT full scale experiments show satisfactory results. Quench front motions are well predicted, as well as clad temperatures in most of the tested runs. The BETHSY 6.7C Integral Effect Test simulating the gravity driven reflooding process in a scaled PWR circuit is then compared to CATHARE-3 simulation. The three-field model is applied in several parts of the circuit: core, upper plenum, hot leg and steam generator, represented by either 1D or 3D modules, while the classic six-equation model is used in the other parts of the loop. An analysis of these first results is presented and future work is defined for improving the droplet behavior simulation in both the upper plenum and the hot legs.

Experimental investigations of heat transfer during sodium boiling in fuel assembly model in justification of advanced fast reactor safety

International Nuclear Information System (INIS)

Khafizov, R.R.; Poplavskij, V.M.; Rachkov, V.I.; Sorokin, A.P.; Ashurko, Yu.M.; Volkov, A.V.; Ivanov, E.F.; Privezentsev, V.V.

2015-01-01

The experimental facility is built up and investigation of heat exchange during sodium boiling in simulated fast reactor core assembly in conditions of natural and forced circulation with sodium plenum and upper end shield model are conducted. It is shown that in the presence of sodium plenum there is possibility to provide long-term cooling of fuel assembly when heat flux density on the surface of fuel element simulator up to 140 and 170 kW/m 2 in conditions of natural and forced circulation, respectively. The obtained data is used for improving calculational model of sodium boiling process in fuel assembly and calculational code COREMELT verification. It is pointed out that heat transfer coefficients in the case of liquid metal boiling in fuel assemblies are slightly over the ones in the case of liquid metals boiling in pipes and pool boiling [ru

Advanced divertor experiments on DIII-D

International Nuclear Information System (INIS)

Schaffer, M.J.; Mahdavi, M.A.; Osborne, T.; Petrie, T.W.; Stambaugh, R.D.; Buchenauer, D.; Hill, D.N.; Klepper, C.C.

1991-04-01

The poloidal divertor is presently favored for next-step, high-power tokamaks. The DIII-D Advanced Divertor Program (ADP) aims to gain increased control over the divertor plasma and tokamak boundary conditions. This paper reports experiments done in the first phase of the ADP. The DIII-D lower divertor was modified by the addition of a toroidally symmetric, graphite-armoured, water-cooled divertor-biasing ring electrode at the entrance to a gas plenum. The plenum will eventually contain a He cryogenic loop for active divertor pumping. The separatrix ''strike'' position is controlled by the lower poloidal field shaping coils and can be varied smoothly from the ring electrode upper surface to the divertor floor far from the entrance aperture. External power, at up to 550 V and 8 kA separately, has been applied to the electrode to date. 5 refs., 5 figs

Experiment data of ROSA-III integral test RUN 710

International Nuclear Information System (INIS)

Koizumi, Yasuo; Tasaka, Kanji; Adachi, Hiromichi; Anoda, Yoshinari; Soda, Kunihisa

1981-01-01

The report presents data of RUN 710 at ROSA-III test facility. RUN 710 simulates a 200% double-ended break at the inlest side of a recirculation pump of a BWR. All ECCS are activated and electric power to simulated fuel rods in one core channel among four is not supplied in RUN 710. The primary initial conditions are steam dome pressure 7.35 MPa, lower plenum subcooling 10.8 K, core inlet flow rate 31.3 kg/s and core heat generation 2.42 MW. Peak cladding temperature is 609 K at Position 3, 352.5 mm above the mid plane of the core. All heater rods are quenched after ECCS actuation and the effectiveness of ECCS is confirmed. (author)

Evaluation report on CCTF core-I reflood tests Cl-5 (Run 14), Cl-10 (Run 19) and Cl-12 (Run 21)

International Nuclear Information System (INIS)

Akimoto, Hajime; Murao, Yoshio

1983-06-01

Three tests Cl-5 (Run 14), Cl-10 (Run 19) and Cl-12 (Run 21) were performed using the Cylindrical Core Test Facility to study the effect of the containment pressure on the core cooling and the system behaviors during the reflood phase of a PWR-LOCA. The containment pressures of these tests were 0.15, 0.20 and 0.30 MPa for the tests Cl-10, Cl-5 and Cl-12, respectively. Through the comparison of the test results from these three tests, the following results were obtained. (1) The higher containment pressure gave the higher heat transfer coefficient in the core. This resulted in the lower turnaround temperature, the shorter turnaround time and the shorter quench time at the higher containment pressure. (2) In the higher containment pressure test, the higher core water head, the higher upper plenum water head, the higher downcomer water head in the early period and the lower downcomer water head in the later period were observed than those in the lower containment pressure test. This resulted in the higher pressure drop through the intact loop in the early period of the tests and the lower pressure drop in the later period of the test with the containment pressure. (3) The pressure drop through the broken cold leg pressurized the primary system. The pressure drop through the broken cold leg was decreased with the containment pressure. (4) The core inlet mass flow rate was increased with the containment pressure as observed in the FLECHT-SET phase B1 test. In quantity, however, the effect of the containment pressure on the increase of the core inlet mass flow rate was less in the CCTF than that in the FLECHT-SET. The less sensitivity in the CCTF was attributed mainly to the great pressure drop through the broken cold leg, which was not observed in the FLECHT-SET with big broken cold leg. (5) The system effect of the containment pressure was explained quantitatively. (author)

Tc-99m DTPA renal function tests and diuretic renogram in the dilated upper urinary tract

Energy Technology Data Exchange (ETDEWEB)

Sakagami, Yoshinari; Yamaguchi, Osamu; Suzuki, Takayuki; Kameoka, Hiroshi; Shiraiwa, Yasuo; Suzuki, Akira (Fukushima Medical Coll. (Japan))

1992-09-01

The authors studied patients with dilated upper urinary tracts (16 patients) using radioisotopic split renal function tests and diuretic renogram with [sup 99]mTc-diethylenetriaminepentaacetic acid (Tc-99m DTPA). The etiology was ureteropelvic function stenosis, 11; primary megaureter, 3; ureteral stenosis, 1; and aberrant vessel, 1. Response to diuresis was classified into 3 groups: i.e., obstructive pattern, non-obstructive pattern and partial obstructive pattern. In the non-obstructive group, split renal function was good, and thus an operation was not indicated. In the obstructive group, split renal function was significantly decreased and these patients underwent surgery. In the partial obstructive group, if function of the obstructed kidney was approximately the same as that of the non-obstructed side, then surgery was considered unnecessary. (author).

Tc-99m DTPA renal function tests and diuretic renogram in the dilated upper urinary tract

International Nuclear Information System (INIS)

Sakagami, Yoshinari; Yamaguchi, Osamu; Suzuki, Takayuki; Kameoka, Hiroshi; Shiraiwa, Yasuo; Suzuki, Akira

1992-01-01

The authors studied patients with dilated upper urinary tracts (16 patients) using radioisotopic split renal function tests and diuretic renogram with 99 mTc-diethylenetriaminepentaacetic acid (Tc-99m DTPA). The etiology was ureteropelvic function stenosis, 11; primary megaureter, 3; ureteral stenosis, 1; and aberrant vessel, 1. Response to diuresis was classified into 3 groups: i.e., obstructive pattern, non-obstructive pattern and partial obstructive pattern. In the non-obstructive group, split renal function was good, and thus an operation was not indicated. In the obstructive group, split renal function was significantly decreased and these patients underwent surgery. In the partial obstructive group, if function of the obstructed kidney was approximately the same as that of the non-obstructed side, then surgery was considered unnecessary. (author)

ITER ECRH upper launcher torus diamond window – Prototyping, testing and qualification

Energy Technology Data Exchange (ETDEWEB)

Schreck, Sabine, E-mail: sabine.schreck@kit.edu [Karlsruhe Institute of Technology, Institute for Applied Materials, Association KIT-EURATOM, P.O. Box 3640, D-76021 Karlsruhe (Germany); Aiello, Gaetano; Meier, Andreas; Strauss, Dirk [Karlsruhe Institute of Technology, Institute for Applied Materials, Association KIT-EURATOM, P.O. Box 3640, D-76021 Karlsruhe (Germany); Ikeda, Ryosuke; Oda, Yasuhisa; Sakamoto, Keishi; Takahashi, Koji [Japan Atomic Energy Agency (JAEA), Plasma Heating Technology Group, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); Scherer, Theo [Karlsruhe Institute of Technology, Institute for Applied Materials, Association KIT-EURATOM, P.O. Box 3640, D-76021 Karlsruhe (Germany)

2015-10-15

Highlights: • The diamond window prototype shows a very good transmission capability during high power RF experiments. • An ad-hoc qualification programme for the diamond torus window is being developed (contract between KIT and F4E). • The window design has been updated focused on its mechanical integrity and manufacturing aspects. - Abstract: The diamond window assembly is part of the ITER primary vacuum boundary and acts as the first tritium barrier and therefore it is classified as Safety/Protection Important Component (SIC/PIC). It consists of an ultra-low loss CVD diamond disk mounted in a system of metallic parts (copper/steel) and has to fulfil adequate transmission capability for high power mm-waves. High power RF experiments with a 1st window prototype had shown parasitic heating due to small gaps in the housing. After a design optimization directed to the mm-wave properties, the parasitic excitations of oscillations have been avoided in a 2nd prototype. This one is equipped with inserted waveguide structures, which cover gaps in the metallic structure of the window housing. From high power RF-measurements with a 0.86 MW/100 s pulse a loss tangent of 7.1 × 10{sup −6} could be estimated, corresponding to an increase of temperature of only 120 mK between inlet and outlet of the cooling system. The diamond window assemblies cannot be entirely covered by codes and standards. To comply with the French safety regulations, instead an ad-hoc qualification programme is required, being developed in the framework of a contract between KIT and F4E. A new prototype (3rd) will be built, which is designed to fit to the single HELICOFLEX sealed waveguide structures of the ex-vessel mm-system of the EC upper launcher (UL). The testing programme ranges from mechanical to vacuum tests up to dielectric loss measurements at low and high power. A clear definition of the testing requirements and of the acceptance criteria is necessary as well as a complete

Analysis results from the Los Alamos 2D/3D program

International Nuclear Information System (INIS)

Boyack, B.E.; Cappiello, M.W.; Harmony, S.C.; Shire, P.R.; Siebe, D.A.

1987-01-01

Los Alamos National Laboratory is a participant in the 2D/3D program. Activities conducted at Los Alamos National Laboratory in support of 2D/3D program goals include analysis support of facility design, construction, and operation; provision of boundary and initial conditions for test-facility operations based on analysis of pressurized water reactors; performance of pretest and posttest predictions and analyses; and use of experimental results to validate and assess the single- and multi-dimensional, nonequilibrium features in the Transient Reactor Analysis Code (TRAC). During fiscal year 1987, Los Alamos conducted analytical assessment activities using data from the Slab Core Test Facility, The Cylindrical Core Test Facility, and the Upper Plenum Test Facility. Finally, Los Alamos continued work to provide TRAC improvements. In this paper, Los Alamos activities during fiscal year 1987 will be summarized; several significant accomplishments will be described in more detail to illustrate the work activities at Los Alamos

CCFL in hot legs and steam generators and its prediction with the CATHARE code

International Nuclear Information System (INIS)

Geffraye, G.; Bazin, P.; Pichon, P.

1995-01-01

This paper presents a study about the Counter-Current Flow Limitation (CCFL) prediction in hot legs and steam generators (SG) in both system test facilities and pressurized water reactors. Experimental data are analyzed, particularly the recent MHYRESA test data. Geometrical and scale effects on the flooding behavior are shown. The CATHARE code modelling problems concerning the CCFL prediction are discussed. A method which gives the user the possibility of controlling the flooding limit at a given location is developed. In order to minimize the user effect, a methodology is proposed to the user in case of a calculation with a counter-current flow between the upper plenum and the SF U-tubes. The following questions have to be made clear for the user: when to use the CATHARE CCFL option, which correlation to use, and where to locate the flooding limit

Impact of 2D/3D-project on LOCA-licensing analysis and reactor safety of PWRs

International Nuclear Information System (INIS)

Winkler, F.; Krebs, W.D.

1989-01-01

In the past LOCA-licensing analysis has included large conservatisms to compensate for the lack of detailed two phase flow and full scale experimental data. The 2D/3D-project was established to improve the data base in order to minimize the conservatisms required. The significant results and findings of the full scale Upper Plenum Test Facility (UPTF) and from the electrically heated Slab Core Test Facility (SCTF) were particularly useful for understanding the multidimensional phenomena in the primary system and in the core of a PWR. UPTF results were used to verify the TRAC-PF1 analysis of a PWR with combined ECC-Injection during the reflood phase of a large break-LOCA. Comparison of these results with results from classic licensing calculations quantifies the large safety margin in earlier licensing procedures and in reactor systems. (orig.)

CCFL in hot legs and steam generators and its prediction with the CATHARE code

Energy Technology Data Exchange (ETDEWEB)

Geffraye, G.; Bazin, P.; Pichon, P. [CEA/DRN/STR, Grenoble (France)

1995-09-01

This paper presents a study about the Counter-Current Flow Limitation (CCFL) prediction in hot legs and steam generators (SG) in both system test facilities and pressurized water reactors. Experimental data are analyzed, particularly the recent MHYRESA test data. Geometrical and scale effects on the flooding behavior are shown. The CATHARE code modelling problems concerning the CCFL prediction are discussed. A method which gives the user the possibility of controlling the flooding limit at a given location is developed. In order to minimize the user effect, a methodology is proposed to the user in case of a calculation with a counter-current flow between the upper plenum and the SF U-tubes. The following questions have to be made clear for the user: when to use the CATHARE CCFL option, which correlation to use, and where to locate the flooding limit.

A feasibility experiment for assessing the efficacy of ex-vessel cooling through the external gap structure

International Nuclear Information System (INIS)

Kang, K. H.; Kim, J. H.; Park, L. J.; Kim, S. B.; Hwang, I. S.

1999-01-01

This paper presents the results of a feasibility experiment for assessing the efficacy of ex-vessel cooling through the external gap structure during a severe accident. In this study, a 1/8 linear scale mockup of a lower plenum was used with Al2O3/Fe thermite melt as a corium simulant. The results show that in dry case test conducted without cooling the outside of the vessel, after about thirty second from the thermite ignition the vessel was heated to cause a complete melt penetration at about 30 degree upper position from the bottom. Whereas in wet case test conducted cooling the outside of the vessel with 0.85 kg/s of water flow rate using 2.5 cm of uniform gap structure, the vessel effectively cooled down with 23.7 K/s of cooling rate by nucleate boiling at the surface of the vessel. The results of two-dimensional analyses using FLUENT code show a similar trend of vessel thermal behavior presented in the tests. Synthesized the results of the tests and analyses work, a natural convection of the melt pool could cause the formation of hot spot at the upper portion of the vessel, but the vessel could effectively cool down by heat removal with ex-vessel cooling

Upper gastrointestinal alterations in kidney transplant candidates.

Science.gov (United States)

Homse Netto, João Pedro; Pinheiro, João Pedro Sant'Anna; Ferrari, Mariana Lopes; Soares, Mirella Tizziani; Silveira, Rogério Augusto Gomes; Maioli, Mariana Espiga; Delfino, Vinicius Daher Alvares

2018-05-14

The incidence of gastrointestinal disorders among patients with chronic kidney disease (CKD) is high, despite the lack of a good correlation between endoscopic findings and symptoms. Many services thus perform upper gastrointestinal (UGI) endoscopy on kidney transplant candidates. This study aims to describe the alterations seen on the upper endoscopies of 96 kidney-transplant candidates seen from 2014 to 2015. Ninety-six CKD patients underwent upper endoscopic examination as part of the preparation to receive kidney grafts. The data collected from the patients' medical records were charted on Microsoft Office Excel 2016 and presented descriptively. Mean values, medians, interquartile ranges and 95% confidence intervals of the clinic and epidemiological variables were calculated. Possible associations between endoscopic findings and infection by H. pylori were studied. Males accounted for 54.17% of the 96 patients included in the study. Median age and time on dialysis were 50 years and 50 months, respectively. The most frequent upper endoscopy finding was enanthematous pangastritis (57.30%), followed by erosive esophagitis (30.20%). Gastric intestinal metaplasia and peptic ulcer were found in 8.33% and 7.30% of the patients, respectively. H. pylori tests were positive in 49 patients, and H. pylori infection was correlated only with non-erosive esophagitis (P = 0.046). Abnormal upper endoscopy findings were detected in all studied patients. This study suggested that upper endoscopy is a valid procedure for kidney transplant candidates. However, prospective studies are needed to shed more light on this matter.

UPPER LIMB FUNCTIONAL ASSESSMENT USING HAPTIC INTERFACE

Directory of Open Access Journals (Sweden)

Aleš Bardorfer

2004-12-01

Full Text Available A new method for the assessment of the upper limb (UL functional state, using a haptic interface is presented. A haptic interface is used as a measuring device, capable of providing objective, repeatable and quantitative data of the UL motion. A patient is presented with a virtual environment, both graphically via a computer screen and haptically via the Phantom Premium 1.5 haptic interface. The setup allows the patient to explore and feel the virtual environment with three of his/her senses; sight, hearing, and most important, touch. Specially designed virtual environments are used to assess the patient’s UL movement capabilities. The tests range from tracking tasks – to assess the accuracy of movement – tracking tasks with added disturbances in a form of random forces – to assess the patient’s control abilities, a labyrinth test – to assess both speed and accuracy, to the last test for measuring the maximal force capacity of the UL.A new method for the assessment of the upper limb (UL functional state, using a haptic interface is presented. A haptic interface is used as a measuring device, capable of providing objective, repeatable and quantitative data of the UL motion. A patient is presented with a virtual environment, both graphically via a computer screen and haptically via the Phantom Premium 1.5 haptic interface. The setup allows the patient to explore and feel the virtual environment with three of his/her senses; sight, hearing, and most important, touch. Specially designed virtual environments are used to assess the patient’s UL movement capabilities. The tests range from tracking tasks–to assess the accuracy of movement-tracking tasks with added disturbances in a form of random forces-to assess the patient’s control abilities, a labyrinth test-to assess both speed and accuracy, to the last test for measuring the maximal force capacity of the UL.A comprehensive study, using the developed measurement setup within the

Data on loss of off-site electric power simulation tests of the high temperature engineering test reactor

International Nuclear Information System (INIS)

Takeda, Takeshi; Nakagawa, Shigeaki; Fujimoto, Nozomu; Tachibana, Yukio; Iyoku, Tatsuo

2002-07-01

The high temperature engineering test reactor (HTTR), the first high temperature gas-cooled reactor (HTGR) in Japan, achieved the first full power of 30 MW on December 7 in 2001. In the rise-to-power test of the HTTR, simulation tests on loss of off-site electric power from 15 and 30 MW operations were carried out by manual shutdown of off-site electric power. Because helium circulators and water pumps coasted down immediately after the loss of off-site electric power, flow rates of helium and water decreased to the scram points. To shut down the reactor safely, the subcriticality should be kept by the insertion of control rods and the auxiliary cooling system should cool the core continuously avoiding excessive cold shock to core graphite components. About 50 s later from the loss of off-site electric power, the auxiliary cooling system started up by supplying electricity from emergency power feeders. Temperature of hot plenum block among core graphite structures decreased continuously after the startup of the auxiliary cooling system. This report describes sequences of dynamic components and transient behaviors of the reactor and its cooling system during the simulation tests from 15 and 30 MW operations. (author)

The Effect of High Intensity Intermittent Exercise on Power Output for the Upper Body

Directory of Open Access Journals (Sweden)

Leonie Harvey

2015-06-01

Full Text Available The aim of the present study was to examine and measure high intensity, intermittent upper body performance, in addition to identifying areas of the body that affect the variance in total work done during the 5 × 6 s sprint test. Fifteen males completed an upper body 5 × 6 s sprint test on a modified electro-magnetically braked cycle ergometer, which consisted of five maximal effort sprints, each 6 s in duration, separated by 24 s of passive recovery. A fly wheel braking force corresponding to 5% of the participants’ body weight was used as the implemented resistance level. Body composition was measured using dual-energy X-ray absorptiometry (DEXA. Percent (% decrement was calculated as 100 − (Total work/ideal work × 100. Significant (P < 0.05 differences were found between sprints for both absolute and relative (W, W·kg−1, W·kg−1 Lean body mass (LBM and W·kg−1 Upper body lean body mass (UBLBM peak (PP and mean (MP power. The % decrement in total work done over the five sprints was 11.4%. Stepwise multiple linear regression analysis revealed that UBLBM accounts for 87% of the variance in total work done during the upper body 5 × 6 s sprint test. These results provide a descriptive analysis of upper body, high intensity intermittent exercise, demonstrating that PP and MP output decreased significantly during the upper body 5 × 6 s sprint test.

Upper Elementary Grades Bear the Brunt of Accountability

Science.gov (United States)

Anderson, Lorin W.

2009-01-01

Upper elementary teachers won't be surprised to learn that in every state, students enrolled in grades 3 through 8 bear the brunt of educational accountability. All states test all students at these grade levels in English/language arts and mathematics. Furthermore, an increasing number of states are testing students at selected elementary and…

Robot-Aided Upper-Limb Rehabilitation Based on Motor Imagery EEG

Directory of Open Access Journals (Sweden)

Baoguo Xu

2011-09-01

Full Text Available Stroke is a leading cause of disability worldwide. In this paper, a novel robot-assisted rehabilitation system based on motor imagery electroencephalography (EEG is developed for regular training of neurological rehabilitation for upper limb stroke patients. Firstly, three-dimensional animation was used to guide the patient image the upper limb movement and EEG signals were acquired by EEG amplifier. Secondly, eigenvectors were extracted by harmonic wavelet transform (HWT and linear discriminant analysis (LDA classifier was utilized to classify the pattern of the left and right upper limb motor imagery EEG signals. Finally, PC triggered the upper limb rehabilitation robot to perform motor therapy and gave the virtual feedback. Using this robot-assisted upper limb rehabilitation system, the patient's EEG of upper limb movement imagination is translated to control rehabilitation robot directly. Consequently, the proposed rehabilitation system can fully explore the patient's motivation and attention and directly facilitate upper limb post-stroke rehabilitation therapy. Experimental results on unimpaired participants were presented to demonstrate the feasibility of the rehabilitation system. Combining robot-assisted training with motor imagery-based BCI will make future rehabilitation therapy more effective. Clinical testing is still required for further proving this assumption.

Interventional studies of the upper gastrointestinal tract

International Nuclear Information System (INIS)

Shapiro, B.; Gross, M.D.

1985-01-01

Nuclear Medicine studies of the upper gastrointestinal (GI) tract provide a means whereby physiologic and pathophysiologic features can be observed from a unique and noninvasive perspective. While nuclear medicine studies by their very nature lack the high spatial resolution of the radiographic approach, the data derived are readily quantitated and presented in numerical fashion to provide functional and dynamic information in which the influences of interventions may be observed. This chapter outlines the scope of such interventions in studies of the upper GI tract with emphasis on examinations for gastroesophageal reflux and gastric emptying. The interactions of nutrients, physical maneuvers of pharmacologic agents on nuclear medicine studies of the upper GI tract may be intentional to render a test more sensitive or to evaluate the effect of therapy, or may represent an unintentional side effect that must be taken into account if misinterpretation is to be avoided

Thermomechanical simulation of WEST actively cooled upper divertor

International Nuclear Information System (INIS)

Batal, T.; Richou, M.; Guilhem, D.; Firdaouss, M.; Larroque, S.; Ferlay, F.; Missirlian, M.; Bucalossi, J.

2016-01-01

The Tore Supra tokamak is being transformed in an x-point divertor fusion device in the frame of the WEST (W-for tungsten-Environment in Steady-state Tokamak) project, launched in support to the ITER tungsten divertor strategy. The WEST project aims to test ITER-like W monoblock Plasma Facing Units (PFU). This ITER-like divertor will be tested under long plasma discharge up to 1000 s, with high heat flux density up to 20 MW/m 2 . This paper presents the results of ANSYS thermal-structural simulations of the WEST upper divertor. The upper divertor is made of twelve 30° sectors, each one composed of 38 PFU. The PFUs are actively cooled CuCrZr heat sinks and the incidence surface is coated with a thin tungsten layer. The fixing system is made of pins engaged in slotted holes. Besides, the fixing system of the sector assembly is the same as WEST lower divertor, so one upper divertor sector can be used indifferently in upper or Lower position during transitional operation phases in WEST. The total surface of the upper divertor is 8 m 2 , and it has to be able to extract up to 4 MW in steady-state, with peak heat flux values up to 8 MW/m 2 . The fixing system was designed to handle structural loads such as forces and torques resulting from halo and eddy current, respectively, especially during disruptions and Vertical Displacement Event (VDE). The torque resulting from eddy current is first calculated thanks to an internal CEA ANSYS APDL routine. Then the ANSYS structural and thermal-structural simulations of the PFU are presented, and its design is validated thanks to A-level RCC-MRx criteria. Finally, the most conservative load case is determined in order to validate the design of the pins and the support structure.

Thermomechanical simulation of WEST actively cooled upper divertor

Energy Technology Data Exchange (ETDEWEB)

Batal, T., E-mail: tristan.batal@cea.fr; Richou, M.; Guilhem, D.; Firdaouss, M.; Larroque, S.; Ferlay, F.; Missirlian, M.; Bucalossi, J.

2016-11-15

The Tore Supra tokamak is being transformed in an x-point divertor fusion device in the frame of the WEST (W-for tungsten-Environment in Steady-state Tokamak) project, launched in support to the ITER tungsten divertor strategy. The WEST project aims to test ITER-like W monoblock Plasma Facing Units (PFU). This ITER-like divertor will be tested under long plasma discharge up to 1000 s, with high heat flux density up to 20 MW/m{sup 2}. This paper presents the results of ANSYS thermal-structural simulations of the WEST upper divertor. The upper divertor is made of twelve 30° sectors, each one composed of 38 PFU. The PFUs are actively cooled CuCrZr heat sinks and the incidence surface is coated with a thin tungsten layer. The fixing system is made of pins engaged in slotted holes. Besides, the fixing system of the sector assembly is the same as WEST lower divertor, so one upper divertor sector can be used indifferently in upper or Lower position during transitional operation phases in WEST. The total surface of the upper divertor is 8 m{sup 2}, and it has to be able to extract up to 4 MW in steady-state, with peak heat flux values up to 8 MW/m{sup 2}. The fixing system was designed to handle structural loads such as forces and torques resulting from halo and eddy current, respectively, especially during disruptions and Vertical Displacement Event (VDE). The torque resulting from eddy current is first calculated thanks to an internal CEA ANSYS APDL routine. Then the ANSYS structural and thermal-structural simulations of the PFU are presented, and its design is validated thanks to A-level RCC-MRx criteria. Finally, the most conservative load case is determined in order to validate the design of the pins and the support structure.

An Investigation of Loop Seal Clearings in ATLAS SBLOCA Tests

Energy Technology Data Exchange (ETDEWEB)

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

2013-05-15

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

The Unsupported Upper Limb Exercise Test in People Without Disabilities: Assessing the Within-Day Test-Retest Reliability and the Effects of Age and Gender.

Science.gov (United States)

Oliveira, Ana; Cruz, Joana; Jácome, Cristina; Marques, Alda

2018-01-01

Purpose: To estimate the within-day test-retest reliability and standard error of measurement (SEM) of the unsupported upper limb exercise test (UULEX) in adults without disabilities and to determine the effects of age and gender on performance of the UULEX. Method: A cross-sectional study was conducted with 100 adults without disabilities (44 men, mean age 44.2 [SD 26] y; 56 women, mean age 38.1 [SD 24.1] y). Participants performed three UULEX tests to establish within-day reliability, measured using an intra-class correlation coefficient (ICC) model 2 (two-way random effects) with a single rater (ICC[2,1]) and SEM. The effects of age and gender were examined using two-factor mixed-design analysis of variance (ANOVA) and one-way repeated-measures ANOVA. For analysis purposes, four sub-groups were created: younger adults, older adults, men, and women. Results: Excellent within-day reliability and a small SEM were found in the four sub-groups (younger adults: ICC[2,1]=0.88; 95% CI: 0.82, 0.92; SEM∼40 s; older adults: ICC[2,1]=0.82; 95% CI: 0.72, 0.90; SEM∼50 s; men: ICC[2,1]=0.93; 95% CI: 0.88, 0.96; SEM∼30 s; women: ICC[2,1]=0.85; 95% CI: 0.78, 0.91; SEM∼45 s). Younger adults took, on average, 308.24 seconds longer than older adults to perform the test; older adults performed significantly better on the third test ( p 0.05). Conclusion: The within-day test-retest reliability and SEM values of the UULEX may be used to define the magnitude of the error obtained with repeated measures. One UULEX test seems to be adequate for younger adults to achieve reliable results, whereas three tests seem to be needed for older adults.

The use of computer adaptive tests in outcome assessments following upper limb trauma.

Science.gov (United States)

Jayakumar, P; Overbeek, C; Vranceanu, A-M; Williams, M; Lamb, S; Ring, D; Gwilym, S

2018-06-01

Aims Outcome measures quantifying aspects of health in a precise, efficient, and user-friendly manner are in demand. Computer adaptive tests (CATs) may overcome the limitations of established fixed scales and be more adept at measuring outcomes in trauma. The primary objective of this review was to gain a comprehensive understanding of the psychometric properties of CATs compared with fixed-length scales in the assessment of outcome in patients who have suffered trauma of the upper limb. Study designs, outcome measures and methodological quality are defined, along with trends in investigation. Materials and Methods A search of multiple electronic databases was undertaken on 1 January 2017 with terms related to "CATs", "orthopaedics", "trauma", and "anatomical regions". Studies involving adults suffering trauma to the upper limb, and undergoing any intervention, were eligible. Those involving the measurement of outcome with any CATs were included. Identification, screening, and eligibility were undertaken, followed by the extraction of data and quality assessment using the Consensus-Based Standards for the Selection of Health Measurement Instruments (COSMIN) criteria. The review is reported according to the Preferred Reporting Items for Systematic Reviews and Meta-Analyses (PRISMA) criteria and reg istered (PROSPERO: CRD42016053886). Results A total of 31 studies reported trauma conditions alone, or in combination with non-traumatic conditions using CATs. Most were cross-sectional with varying level of evidence, number of patients, type of study, range of conditions and methodological quality. CATs correlated well with fixed scales and had minimal or no floor-ceiling effects. They required significantly fewer questions and/or less time for completion. Patient-Reported Outcomes Measurement Information System (PROMIS) CATs were the most frequently used, and the use of CATs is increasing. Conclusion Early studies show valid and reliable outcome measurement with CATs

Oncoplastic Surgery for Upper/Upper Inner Quadrant Breast Cancer.

Science.gov (United States)

Lin, Joseph; Chen, Dar-Ren; Wang, Yu-Fen; Lai, Hung-Wen

2016-01-01

Tumors located in the upper/upper inner quadrant of the breast warrant more attention. A small lesion relative to the size of breast in this location may be resolved by performing a level I oncoplastic technique. However, a wide excision may significantly reduce the overall quality of the breast shape by distorting the visible breast line. From June 2012 to April 2015, 36 patients with breast cancer located in the upper/upper inner quadrant underwent breast-conservation surgery with matrix rotation mammoplasty. According to the size and location of the tumor relative to the nipple-areola complex, 11 patients underwent matrix rotation with periareolar de-epithelialization (donut group) and the other 25 underwent matrix rotation only (non-donut group). The cosmetic results were self-assessed by questionnaires. The average weights of the excised breast lumps in the donut and non-donut groups were 104.1 and 84.5 g, respectively. During the 3-year follow-up period, local recurrence was observed in one case and was managed with nipple-sparing mastectomy followed by breast reconstruction with prosthetic implants. In total, 31 patients (88.6%) ranked their postoperative result as either acceptable or satisfactory. The treated breasts were also self-evaluated by 27 patients (77.1%) to be nearly identical to or just slightly different from the untreated side. Matrix rotation is an easy breast-preserving technique for treating breast cancer located in the upper/upper inner quadrant of the breast that requires a relatively wide excision. With this technique, a larger breast tumor could be removed without compromising the breast appearance.

Difference in Functional Performance on the Upper-Quarter Y-Balance Test Between High School Baseball Players and Wrestlers.

Science.gov (United States)

Myers, Heather; Poletti, Mary; Butler, Robert J

2017-05-01

The Upper Quarter Y-Balance Test (YBT-UQ) is a unique movement test where individuals perform at the limits of their stability, requiring the coordination of balance, proprioception, range of motion, and stabilization. It is not yet clear if performance on the YBT-UQ differs between sports with dissimilar emphasis on upper-extremity performance. To compare performance on the YBT-UQ between wrestlers, whose sport requires some degree of closed-chain activity, and baseball players, whose sport is primarily open kinetic chain in nature. Cross-sectional. High school preparticipation physical assessment. 24 healthy high school male wrestlers (mean age 16.12 ± 1.24 y) and 24 healthy high school male baseball players (mean age 15.79 ± 1.25 y). All subjects performed the YBT-UQ, which requires reaching in 3 directions while maintaining a push-up position. The variables of interest include the maximum reach in each direction, as well as the composite score. In addition, asymmetries between limbs for each reach direction were compared. Wrestlers performed significantly better than baseball players in the medial direction, inferolateral direction, and in composite scores. In the medial direction, wrestlers exhibited greater scores (P < .01) on both left and right limbs, 10.5 ± 10.2%LL and 9.95 ± 10.2%LL, respectively. Significant differences (P < .01) were also observed in the inferolateral direction, with a difference of 11.3 ± 12.0%LL on the left and 8.7 ± 11.0%LL on the right. Composite scores were higher (P < .01) for the wrestlers, with a difference of 7.0% on the left and 7.1% on the right. This study suggests that wrestlers perform better on the YBT-UQ than baseball players. The findings may suggest sport-specific normative data for the YBT-UQ in high school athletes.

Robot-Aided Upper-Limb Rehabilitation Based on Motor Imagery EEG

Directory of Open Access Journals (Sweden)

Baoguo Xu

2011-09-01

Full Text Available Stroke is a leading cause of disability worldwide. In this paper, a novel robot‐assisted rehabilitation system based on motor imagery electroencephalography (EEG is developed for regular training of neurological rehabilitation for upper limb stroke patients. Firstly, three‐dimensional animation was used to guide the patient image the upper limb movement and EEG signals were acquired by EEG amplifier. Secondly, eigenvectors were extracted by harmonic wavelet transform (HWT and linear discriminant analysis (LDA classifier was utilized to classify the pattern of the left and right upper limb motor imagery EEG signals. Finally, PC triggered the upper limb rehabilitation robot to perform motor therapy and gave the virtual feedback. Using this robot‐assisted upper limb rehabilitation system, the patientʹs EEG of upper limb movement imagination is translated to control rehabilitation robot directly. Consequently, the proposed rehabilitation system can fully explore the patientʹs motivation and attention and directly facilitate upper limb post‐stroke rehabilitation therapy. Experimental results on unimpaired participants were presented to demonstrate the feasibility of the rehabilitation system. Combining robot‐assisted training with motor imagery‐ based BCI will make future rehabilitation therapy more effective. Clinical testing is still required for further proving this assumption.

An objective assessment of safety to drive in an upper limb cast.

Science.gov (United States)

Stevenson, H L; Peterson, N; Talbot, C; Dalal, S; Watts, A C; Trail, I A

2013-03-01

Patients managed with upper limb cast immobilization often seek advice about driving. There is very little published data to assist in decision making, and advice given varies between healthcare professionals. There are no specific guidelines available from the UK Drivers and Vehicles Licensing Agency, police, or insurance companies. Evidence-based guidelines would enable clinicians to standardize the advice given to patients. Six individuals (three male, three female; mean age 36 years, range 27-43 years) were assessed by a mobility occupational therapist and driving standards agency examiner while completing a formal driving test in six different types of upper limb casts (above-elbow, below-elbow neutral, and below-elbow cast incorporating the thumb [Bennett's cast]) on both left and right sides. Of the 36 tests, participants passed 31 tests, suggesting that most people were able to safely drive with upper limb cast immobilization. However, driving in a left above-elbow cast was considered unsafe.

Analysis of multi-dimensional and countercurrent effects in a BWR loss-of-coolant accident

International Nuclear Information System (INIS)

Shiralkar, B.S.; Dix, G.E.; Alamgir, M.

1989-01-01

The presence of parallel enclosed channels in a BWR provides opportunities for multiple flow regimes in co-current and countercurrent flow under Loss-of-Coolant Accident (LOCA) conditions. To address and understand these phenomena, an integrated experimental and analytical study has been conducted. The primary experimental facility was the Steam Sector Test Facility (SSTF) which simulated a full scale 30deg sector of a BWR/6 reactor vessel. Both steady-state separate effects tests and integral transients with vessel blowdown and refill were performed. The present of multi-dimensional and parallel channel effects was found to be very beneficial to BWR LOCA performance. The best estimate TRAC-BWR computer code was extended as part of this study by incorporation of a phenomenological upper plenum mixing model. TRAC-BWR was applied to the analysis of these full scale experiments. Excellent predictions of phenomena and experimental trends were achieved. (orig.)

Analysis of multidimensional and countercurrent effects in a BWR loss-of-coolant accident

International Nuclear Information System (INIS)

Shiralkar, B.S.; Dix, G.E.; Alamgir, M.

1991-01-01

The presence of parallel enclosed channels in a boiling water reactor (BWR) provides opportunities for multiple flow regimes in cocurrent and countercurrent flow under loss-of-coolant accident (LOCA) conditions. To address and understand these phenomena, an integrated experimental and analytical study has been conducted. The primary experimental facility was the steam sector test facility (SSFT), which simulated a full scale 30deg sector of a BWR/6 reactor vessel. Both steady-state separate effects tests an integral transients with vessel vlowdown and refill were performed. The presence of multidimensional and parallel-channel effects was found to be very beneficial to BWR LOCA performance. The best estimate TRAC-BWR computer code was extended as part of this study by incorporation of a phenomenological upper plenum mixing model. TRAC-BWR was applied to the analysis of these full scale experiments. Excellent predictions of phenomena and experimental trends were achieved. (orig.)

EVALUATION OF SHEAR STRENGTH FOR UPPER SLABS OF CAISSON FOUNDATION BASED ON LOAD CARRYING MECHANISM

Science.gov (United States)

Hattori, Hisamichi; Tadokoro, Toshiya; Tanimura, Yukihiro; Nishioka, Hidetoshi; Watanabe, Tadatomo; Maruyama, Osamu

In upper slabs of caisson foundation, a seismic desi gn is difficult with an incr ease in earthquake load. So we carried out loading tests and FEM analysis for upper slabs of caisson foundation. As a result, we proposed a new design method which takes into co nsideration the effective width on the pull out side based on crack pattern of test specimens, which is not considered in the existing design method. Moreover, we proposed a rational design method based on load carrying mechanism for upper slabs of caisson foundation.

Use of remifentanil to reduce propofol injection pain and the required propofol dose in upper digestive tract endoscopy diagnostic tests.

Science.gov (United States)

Uliana, Gustavo Nadal; Tambara, Elizabeth Milla; Baretta, Giorgio Alfredo Pedroso

2015-01-01

The introduction of propofol (2,6-diisopropylphenol) as a sedative agent has transformed the area of sedation for endoscopic procedures. However, a major drawback of sedation with the use of propofol is its high incidence of injection pain. The most widely used technique in reducing propofol injection pain is through the association of other drugs. The aim of this study was to evaluate the effect of remifentanil-propofol combination on the incidence of propofol injection pain and its influence on the total dose of propofol required for sedation in upper digestive tract endoscopy (UDE) diagnostic tests. One hundred and five patients undergoing upper digestive tract endoscopy were evaluated and randomly divided into 3 groups of 35 patients each. The Control Group received propofol alone; Study-group 1 received remifentanil at a fixed dose of 0.2mg/kg combined with propofol; Study-group 2 received remifentanil at a fixed dose of 0.3mg/kg combined with propofol. The incidence of propofol injection pain and the total dose of propofol required for the test were evaluated. The sample was very similar regarding age, weight, height, sex, and physical status. Statistical analysis was performed according to the nature of the evaluated data. Student's t-test was used to compare the mean of age, weight, height (cm), and dose (mg/kg) variables between groups. The χ(2) test was used to compare sex, physical status, and propofol injection pain between groups. The significance level was αpain and total dose of propofol (mg/kg) used. However, there were no statistical differences between the two study groups for these parameters. We conclude that the use of remifentanil at doses of 0.2mg/kg and 0.3mg/kg was effective for reducing both the propofol injection pain and the total dose of propofol used. Copyright © 2015 Sociedade Brasileira de Anestesiologia. Published by Elsevier Editora Ltda. All rights reserved.

In-pile TREAT Test L04: simulating a lead sub-assembly in an unprotected LMFBR loss-of-coolant accident

International Nuclear Information System (INIS)

Tylka, J.P.; Bauer, T.H.; Wright, A.E.; Davies, A.L.; Herbert, R.; Woods, W.J.

1983-01-01

Test L04 in the PFR/TREAT series is the first multi-pin, in-pile simulation of a LMFBR transient undercooling/overpower (TUCOP) accident using full length prototypic fuel irradiated in a fast reactor. L04 is a gridded 7-pin bundle test performed in the ANL Mk-III integral loop in a flowing sodium environment and uses prototypic, bottom plenum, UK reactor fuel, preirradiated in the PFR to an axial peak burn-up of 4.2 a/o. The objective of L04 was the study, by simulation, of coolant voiding and fuel motion during the initiating phase of a hypothetical TUCOP accident in a large LMFBR. Test L04 is intended to study the behavior of a centrally located, lead subassembly with the highest power-to-flow ratio

Analysis of a postulated accident scenario involving loss of forced flow in a LMFBR

International Nuclear Information System (INIS)

Moreira, M.L.

1985-01-01

A model to analyse a postulated accident scenario involving loss of forced flow in the reactor vessel of a LMFBR is used. Five phases of the accident are analysed: Natural Circulation, Subcooled Boiling, Nucleate Boiling, Core Dryout and Cladding melt. The heat conduction in the fuel cladding, coolant and lower and upper plenum are calculated by a lump-parameter model. Physical data of a prototype LMFBR reactor were used for the calculation. (author)

Effect of STS space suit on astronaut dominant upper limb EVA work performance

Science.gov (United States)

Greenisen, Michael C.

1987-01-01

The STS Space Suited and unsuited dominant upper limb performance was evaluated in order to quantify future EVA astronaut skeletal muscle upper limb performance expectations. Testing was performed with subjects standing in EVA STS foot restraints. Data was collected with a CYBEX Dynamometer enclosed in a waterproof container. Control data was taken in one g. During one g testing, weight of the Space Suit was relieved from the subject via an overhead crane with a special connection to the PLSS of the suit. Experimental data was acquired during simulated zero g, accomplished by neutral buoyancy in the Weightless Environment Training Facility. Unsuited subjects became neutrally buoyant via SCUBA BC vests. Actual zero g experimental data was collected during parabolic arc flights on board NASA's modified KC-135 aircraft. During all test conditions, subjects performed five EVA work tasks requiring dominant upper limb performance and ten individual joint articulation movements. Dynamometer velocities for each tested movement were 0 deg/sec, 30 or 60 deg/sec and 120 or 180 deg/sec, depending on the test, with three repetitions per test. Performance was measured in foot pounds of torque.

Oncoplastic Surgery for Upper/Upper Inner Quadrant Breast Cancer.

Directory of Open Access Journals (Sweden)

Joseph Lin

Full Text Available Tumors located in the upper/upper inner quadrant of the breast warrant more attention. A small lesion relative to the size of breast in this location may be resolved by performing a level I oncoplastic technique. However, a wide excision may significantly reduce the overall quality of the breast shape by distorting the visible breast line. From June 2012 to April 2015, 36 patients with breast cancer located in the upper/upper inner quadrant underwent breast-conservation surgery with matrix rotation mammoplasty. According to the size and location of the tumor relative to the nipple-areola complex, 11 patients underwent matrix rotation with periareolar de-epithelialization (donut group and the other 25 underwent matrix rotation only (non-donut group. The cosmetic results were self-assessed by questionnaires. The average weights of the excised breast lumps in the donut and non-donut groups were 104.1 and 84.5 g, respectively. During the 3-year follow-up period, local recurrence was observed in one case and was managed with nipple-sparing mastectomy followed by breast reconstruction with prosthetic implants. In total, 31 patients (88.6% ranked their postoperative result as either acceptable or satisfactory. The treated breasts were also self-evaluated by 27 patients (77.1% to be nearly identical to or just slightly different from the untreated side. Matrix rotation is an easy breast-preserving technique for treating breast cancer located in the upper/upper inner quadrant of the breast that requires a relatively wide excision. With this technique, a larger breast tumor could be removed without compromising the breast appearance.

Recent irradiation tests of uranium-plutonium-zirconium metal fuel elements

International Nuclear Information System (INIS)

Pahl, R.G.; Lahm, C.E.; Villarreal, R.; Hofman, G.L.; Beck, W.N.

1986-09-01

Uranium-Plutonium-Zirconium metal fuel irradiation tests to support the ANL Integral Fast Reactor concept are discussed. Satisfactory performance has been demonstrated to 2.9 at.% peak burnup in three alloys having 0, 8, and 19 wt % plutonium. Fuel swelling measurements at low burnup in alloys to 26 wt % plutonium show that fuel deformation is primarily radial in direction. Increasing the plutonium content in the fuel diminishes the rate of fuel-cladding gap closure and axial fuel column growth. Chemical redistribution occurs by 2.1 at.% peak burnup and generally involves the inward migration of zirconium and outward migration of uranium. Fission gas release to the plenum ranges from 46% to 56% in the alloys irradiated to 2.9 at.% peak burnup. No evidence of deleterious fuel-cladding chemical or mechanical interaction was observed

Upper bounds for Neyman-Pearson cooperative spectrum sensing

KAUST Repository

Zahabi, Sayed Jalal; Tadaion, Ali Akbar; Aissa, Sonia

2011-01-01

We consider a cooperative spectrum sensing scenario where the local sensors at the secondary users are viewed as one-level quantizers, and the quantized data are to be fused under Neyman-Pearson (N-P) criterion. We demonstrate how the N-P fusion results in a randomized test, which represents the total performance of our spectrum sensing scheme. We further introduce an upper performance bound for the overall primary user signal detection. An analytical procedure towards the upper bound and its relevant quantization setup at the local sensors are proposed and examined through simulations. © 2011 IEEE.

Upper bounds for Neyman-Pearson cooperative spectrum sensing

KAUST Repository

Zahabi, Sayed Jalal

2011-06-01

We consider a cooperative spectrum sensing scenario where the local sensors at the secondary users are viewed as one-level quantizers, and the quantized data are to be fused under Neyman-Pearson (N-P) criterion. We demonstrate how the N-P fusion results in a randomized test, which represents the total performance of our spectrum sensing scheme. We further introduce an upper performance bound for the overall primary user signal detection. An analytical procedure towards the upper bound and its relevant quantization setup at the local sensors are proposed and examined through simulations. © 2011 IEEE.

Upper limit of peak area

International Nuclear Information System (INIS)

Helene, O.A.M.

1982-08-01

The determination of the upper limit of peak area in a multi-channel spectra, with a known significance level is discussed. This problem is specially important when the peak area is masked by the background statistical fluctuations. The problem is exactly solved and, thus, the results are valid in experiments with small number of events. The results are submitted to a Monte Carlo test and applied to the 92 Nb beta decay. (Author) [pt

Survey of upper extremity injuries among martial arts participants.

Science.gov (United States)

Diesselhorst, Matthew M; Rayan, Ghazi M; Pasque, Charles B; Peyton Holder, R

2013-01-01

To survey participants at various experience levels of different martial arts (MA) about upper extremity injuries sustained during training and fighting. A 21-s question survey was designed and utilised. The survey was divided into four groups (Demographics, Injury Description, Injury Mechanism, and Miscellaneous information) to gain knowledge about upper extremity injuries sustained during martial arts participation. Chi-square testing was utilised to assess for significant associations. Males comprised 81% of respondents. Involvement in multiple forms of MA was the most prevalent (38%). The hand/wrist was the most common area injured (53%), followed by the shoulder/upper arm (27%) and the forearm/elbow (19%). Joint sprains/muscle strains were the most frequent injuries reported overall (47%), followed by abrasions/bruises (26%). Dislocations of the upper extremity were reported by 47% of participants while fractures occurred in 39%. Surgeries were required for 30% of participants. Females were less likely to require surgery and more likely to have shoulder and elbow injuries. Males were more likely to have hand injuries. Participants of Karate and Tae Kwon Do were more likely to have injuries to their hands, while participants of multiple forms were more likely to sustain injuries to their shoulders/upper arms and more likely to develop chronic upper extremity symptoms. With advanced level of training the likelihood of developing chronic upper extremity symptoms increases, and multiple surgeries were required. Hand protection was associated with a lower risk of hand injuries. Martial arts can be associated with substantial upper extremity injuries that may require surgery and extended time away from participation. Injuries may result in chronic upper extremity symptoms. Hand protection is important for reducing injuries to the hand and wrist.

PWR upper/lower internals shield

Energy Technology Data Exchange (ETDEWEB)

Homyk, W.A. [Indian Point Station, Buchanan, NY (United States)

1995-03-01

During refueling of a nuclear power plant, the reactor upper internals must be removed from the reactor vessel to permit transfer of the fuel. The upper internals are stored in the flooded reactor cavity. Refueling personnel working in containment at a number of nuclear stations typically receive radiation exposure from a portion of the highly contaminated upper intervals package which extends above the normal water level of the refueling pool. This same issue exists with reactor lower internals withdrawn for inservice inspection activities. One solution to this problem is to provide adequate shielding of the unimmersed portion. The use of lead sheets or blankets for shielding of the protruding components would be time consuming and require more effort for installation since the shielding mass would need to be transported to a support structure over the refueling pool. A preferable approach is to use the existing shielding mass of the refueling pool water. A method of shielding was devised which would use a vacuum pump to draw refueling pool water into an inverted canister suspended over the upper internals to provide shielding from the normally exposed components. During the Spring 1993 refueling of Indian Point 2 (IP2), a prototype shield device was demonstrated. This shield consists of a cylindrical tank open at the bottom that is suspended over the refueling pool with I-beams. The lower lip of the tank is two feet below normal pool level. After installation, the air width of the natural shielding provided by the existing pool water. This paper describes the design, development, testing and demonstration of the prototype device.

CATHARE-2 prediction of large primary to secondary leakage (PRISE) at PSB-VVER experimental facility

Energy Technology Data Exchange (ETDEWEB)

Sabotinov, L.; Chevrier, P. [Institut de Radioprotection et de Surete Nucleaire, 92 - Fontenay aux Roses (France)

2007-07-01

The large primary to secondary leakage (PRISE) is a specific loss-of-coolant accident in VVER reactors, related to the break of the steam generator collector cover, leading to loss of primary mass inventory and possible direct radioactive release to atmosphere. The best estimate thermal-hydraulic computer code CATHARE-2 Version 2.5-1 was used for post-test analysis of a PRISE experiment, conducted at the large scale test facility PSB-VVER in Russia. The PSB rig is 1:300 scaled model of VVER-1000 NPP. The accident is calculated with a 1.4% break size, which corresponds to 100 mm leak from primary to secondary side in the real NPP. A computer model has been developed for CATHARE-2 V2.5-1, taking into account all important components of the PSB facility: reactor model (lower plenum, core, bypass, upper plenum, downcomer), 4 separate loops, pressurizer, horizontal multi-tube steam generators, break section. The secondary side is presented by recirculation model. A large number of sensitivity calculations has been performed regarding break modeling, reactor pressure vessel modeling, counter current flow modeling, hydraulic losses, heat losses, steam generator level regulation. Comparison between calculated and experimental results shows good prediction of the basic thermal-hydraulic phenomena and parameters such as primary and secondary pressures, temperatures, loop flows, etc. Some discrepancies were observed in the calculations of primary mass inventory and loop seal clearance. Nevertheless the final core heat up, which is one of the most important safety criteria, was correctly predicted. (authors)

Investigation of hand function among children diagnosed with autism spectrum disorder with upper extremity trauma history.

Science.gov (United States)

Huri, Meral; Şahin, Sedef; Kayıhan, Hülya

2016-11-01

The present study was designed to compare hand function in autistic children with history of upper extremity trauma with that of autistic children those who do not have history of trauma. The study group included total of 65 children diagnosed with autism spectrum disorder (ASD) and was divided into 2 groups: children with trauma history (Group I) and control group (Group II) (Group I: n=28; Group II: n=37). Hand function was evaluated with 9-Hole Peg Test and Jebsen Hand Function Test. Somatosensory function was evaluated using somatosensory subtests of Sensory Integration and Praxis Test. Results were analyzed with Student's t-test and Mann-Whitney U test using SPSS version 20 software. Hand function and somatosensory perception test scores were statistically significantly better in children without upper extremity trauma history (pManual Form Perception and Localization of Tactile Stimuli Test results (p<0.05). Autistic children with upper extremity trauma history had poor somatosensory perception and hand function. It is important to raise awareness among emergency service staff and inform them about strong relationship between somatosensory perception, hand function, and upper extremity trauma in children with ASD in order to develop appropriate rehabilitation process and prevent further trauma.

Experiment on performance of upper head injection system with ROSA-II

International Nuclear Information System (INIS)

1976-09-01

Thermo-hydraulic behavior in the primary cooling system of a pressurized water reactor with an upper head injection system (UHI) in a postulated loss-of-coolant accident (LOCA) has been studied with ROSA-II test facility. Simulated UHI and internal structures of the pressure vessel were installed to the facility for the experiment. Nine maximum-sized double-ended break tests and one medium-sized split break test were performed for the cold-leg break condition. The results are as follows: (1) Fluid mixing in the upper head is not perfect. (2) Cold water injection into the steam or two-phase fluid causes violent depressurization due to the condensation. Flow pattern in the primary cooling system is largely influenced by the above two. (auth.)

Design and test of a Microsoft Kinect-based system for delivering adaptive visual feedback to stroke patients during training of upper limb movement.

Science.gov (United States)

Simonsen, Daniel; Popovic, Mirjana B; Spaich, Erika G; Andersen, Ole Kæseler

2017-11-01

The present paper describes the design and test of a low-cost Microsoft Kinect-based system for delivering adaptive visual feedback to stroke patients during the execution of an upper limb exercise. Eleven sub-acute stroke patients with varying degrees of upper limb function were recruited. Each subject participated in a control session (repeated twice) and a feedback session (repeated twice). In each session, the subjects were presented with a rectangular pattern displayed on a vertical mounted monitor embedded in the table in front of the patient. The subjects were asked to move a marker inside the rectangular pattern by using their most affected hand. During the feedback session, the thickness of the rectangular pattern was changed according to the performance of the subject, and the color of the marker changed according to its position, thereby guiding the subject's movements. In the control session, the thickness of the rectangular pattern and the color of the marker did not change. The results showed that the movement similarity and smoothness was higher in the feedback session than in the control session while the duration of the movement was longer. The present study showed that adaptive visual feedback delivered by use of the Kinect sensor can increase the similarity and smoothness of upper limb movement in stroke patients.

Natural circulation in a VVER reactor geometry: Experiments with the PACTEL facility and Cathare simulations

Energy Technology Data Exchange (ETDEWEB)

Raussi, P.; Kainulainen, S. [Lappeenranta Univ. of Technology, Lappeenranta (Finland); Kouhia, J. [VTT Energy, Lappeenranta (Finland)

1995-09-01

There are some 40 reactors based on the VVER design in use. Database available for computer code assessment for VVER reactors is rather limited. Experiments were conducted to study natural circulation behaviour in the PACTEL facility, a medium-scale integral test loop patterned after VVER pressurized water reactors. Flow behaviour over a range of coolant inventories was studied with a small-break experiment. In the small-break experiments, flow stagnation and system repressurization were observed when the water level in the upper plenum fell below the entrances to the hot legs. The cause was attributed to the hot leg loop seals, which are a unique feature of the VVER geometry. At low primary inventories, core cooling was achieved through the boiler-condenser mode. The experiment was simulated using French thermalhydraulic system code CATHARE.

Control room inleakage testing using tracer gases at Zion Generating station

International Nuclear Information System (INIS)

Lagus, P.L.; Brown, J.H.; Dubois, L.J.; Fleming, K.M.

1993-01-01

In order to assess the amount of air inleakage into the Control Room Envelope at Zion Generating Station (ZGS), a series of tracer gas tests using sulfur hexafluoride (SF 6 ) were performed on the Control Room ventilation system (PV system) and the Computer Room/Miscellaneous Area ventilation system (OV system) during February, 1991. Two redundant trains, denoted A and B comprise the PV system. Inleakage was measured for each train. An OV supply duct passes through the Control Room Envelope. Leakage from this duct into the Control Room would constitute air leakage into the Control room Envelope and hence any potential leakage had to be quantified. Each test attempted to measure the contribution (if any) of a particular section of PV return duct or OV supply duct to the total air inleakage into the Control Room. This paper reviews the tracer gas tests. Described here are the control room inleakage testing, HVAC equipment room duct inleakage, purge plenum inleakage, OV duct leakage into the control room envelope, vestibule PV return inleakage, TSC duct inleakage, and cable spreading room inleakage. Conclusions from the testing are presented. 5 refs., 4 figs., 7 tabs

Summary and evaluation of hydraulic property data available for the Hanford Site upper basalt confined aquifer system

International Nuclear Information System (INIS)

Spane, F.A. Jr.; Vermeul, V.R.

1994-09-01

Pacific Northwest Laboratory, as part of the Hanford Site Ground-Water Surveillance Project, examines the potential for offsite migration of contamination within the upper basalt confined aquifer system. For the past 40 years, hydrologic testing of the upper basalt confined aquifer has been conducted by a number of Hanford Site programs. Hydraulic property estimates are important for evaluating aquifer flow characteristics (i.e., ground-water flow patterns, flow velocity, transport travel time). Presented are the first comprehensive Hanford Site-wide summary of hydraulic properties for the upper basalt confined aquifer system (i.e., the upper Saddle Mountains Basalt). Available hydrologic test data were reevaluated using recently developed diagnostic test analysis methods. A comparison of calculated transmissivity estimates indicates that, for most test results, a general correspondence within a factor of two between reanalysis and previously reported test values was obtained. For a majority of the tests, previously reported values are greater than reanalysis estimates. This overestimation is attributed to a number of factors, including, in many cases, a misapplication of nonleaky confined aquifer analysis methods in previous analysis reports to tests that exhibit leaky confined aquifer response behavior. Results of the test analyses indicate a similar range for transmissivity values for the various hydro-geologic units making up the upper basalt confined aquifer. Approximately 90% of the calculated transmissivity values for upper basalt confined aquifer hydrogeologic units occur within the range of 10 0 to 10 2 m 2 /d, with 65% of the calculated estimate values occurring between 10 1 to 10 2 m 2 d. These summary findings are consistent with the general range of values previously reported for basalt interflow contact zones and sedimentary interbeds within the Saddle Mountains Basalt

ON COMPUTING UPPER LIMITS TO SOURCE INTENSITIES

International Nuclear Information System (INIS)

Kashyap, Vinay L.; Siemiginowska, Aneta; Van Dyk, David A.; Xu Jin; Connors, Alanna; Freeman, Peter E.; Zezas, Andreas

2010-01-01

A common problem in astrophysics is determining how bright a source could be and still not be detected in an observation. Despite the simplicity with which the problem can be stated, the solution involves complicated statistical issues that require careful analysis. In contrast to the more familiar confidence bound, this concept has never been formally analyzed, leading to a great variety of often ad hoc solutions. Here we formulate and describe the problem in a self-consistent manner. Detection significance is usually defined by the acceptable proportion of false positives (background fluctuations that are claimed as detections, or Type I error), and we invoke the complementary concept of false negatives (real sources that go undetected, or Type II error), based on the statistical power of a test, to compute an upper limit to the detectable source intensity. To determine the minimum intensity that a source must have for it to be detected, we first define a detection threshold and then compute the probabilities of detecting sources of various intensities at the given threshold. The intensity that corresponds to the specified Type II error probability defines that minimum intensity and is identified as the upper limit. Thus, an upper limit is a characteristic of the detection procedure rather than the strength of any particular source. It should not be confused with confidence intervals or other estimates of source intensity. This is particularly important given the large number of catalogs that are being generated from increasingly sensitive surveys. We discuss, with examples, the differences between these upper limits and confidence bounds. Both measures are useful quantities that should be reported in order to extract the most science from catalogs, though they answer different statistical questions: an upper bound describes an inference range on the source intensity, while an upper limit calibrates the detection process. We provide a recipe for computing upper

Advanced divertor experiments on DIII-D

International Nuclear Information System (INIS)

Schaffer, M.J.; Mahdavi, M.A.; Osborne, T.; Petrie, T.W.; Stambaugh, R.D.; Buchenauer, D.; Hill, D.N.; Klepper, C.C.

1991-01-01

The poloidal divertor is presently favored for next-step, high-power tokamaks. The DIII-D Advanced Divertor Program (ADP) aims to gain increased control over the divertor plasma and tokamak boundary conditions. This paper reports experiments done in the first phase of the ADP. The DIII-D lower divertor was modified by the addition of a toroidally symmetric, graphite-armoured, water-cooled divertor-biasing ring electrode at the entrance to a gas plenum. (In the past DIII-D operated with an open divertor.) The plenum will eventually contain a He cryogenic loop for active divertor pumping. The separatrix 'strike' position is controlled by the lower poloidal field shaping coils and can be varied smoothly from the ring electrode upper surface to the divertor floor far from the entrance aperture. External power, at up to 550 V and 8 kA separately, has been applied to the electrode to date. (author) 5 refs., 5 figs

Upper extremity deep venous thrombosis after port insertion: What are the risk factors?

Science.gov (United States)

Tabatabaie, Omidreza; Kasumova, Gyulnara G; Kent, Tara S; Eskander, Mariam F; Fadayomi, Ayotunde B; Ng, Sing Chau; Critchlow, Jonathan F; Tawa, Nicholas E; Tseng, Jennifer F

2017-08-01

Totally implantable venous access devices (ports) are widely used, especially for cancer chemotherapy. Although their use has been associated with upper extremity deep venous thrombosis, the risk factors of upper extremity deep venous thrombosis in patients with a port are not studied adequately. The Healthcare Cost and Utilization Project's Florida State Ambulatory Surgery and Services Database was queried between 2007 and 2011 for patients who underwent outpatient port insertion, identified by Current Procedural Terminology code. Patients were followed in the State Ambulatory Surgery and Services Database, State Inpatient Database, and State Emergency Department Database for upper extremity deep venous thrombosis occurrence. The cohort was divided into a test cohort and a validation cohort based on the year of port placement. A multivariable logistic regression model was developed to identify risk factors for upper extremity deep venous thrombosis in patients with a port. The model then was tested on the validation cohort. Of the 51,049 patients in the derivation cohort, 926 (1.81%) developed an upper extremity deep venous thrombosis. On multivariate analysis, independently significant predictors of upper extremity deep venous thrombosis included age deep venous thrombosis (odds ratio = 1.77), all-cause 30-day revisit (odds ratio = 2.36), African American race (versus white; odds ratio = 1.86), and other nonwhite races (odds ratio = 1.35). Additionally, compared with genitourinary malignancies, patients with gastrointestinal (odds ratio = 1.55), metastatic (odds ratio = 1.76), and lung cancers (odds ratio = 1.68) had greater risks of developing an upper extremity deep venous thrombosis. This study identified major risk factors of upper extremity deep venous thrombosis. Further studies are needed to evaluate the appropriateness of thromboprophylaxis in patients at greater risk of upper extremity deep venous thrombosis. Copyright © 2017 Elsevier Inc

Nonvariceal upper gastrointestinal bleeding

International Nuclear Information System (INIS)

Burke, Stephen J.; Weldon, Derik; Sun, Shiliang; Golzarian, Jafar

2007-01-01

Nonvariceal upper gastrointestinal bleeding (NUGB) remains a major medical problem even after advances in medical therapy with gastric acid suppression and cyclooxygenase (COX-2) inhibitors. Although the incidence of upper gastrointestinal bleeding presenting to the emergency room has slightly decreased, similar decreases in overall mortality and rebleeding rate have not been experienced over the last few decades. Many causes of upper gastrointestinal bleeding have been identified and will be reviewed. Endoscopic, radiographic and angiographic modalities continue to form the basis of the diagnosis of upper gastrointestinal bleeding with new research in the field of CT angiography to diagnose gastrointestinal bleeding. Endoscopic and angiographic treatment modalities will be highlighted, emphasizing a multi-modality treatment plan for upper gastrointestinal bleeding. (orig.)

Nonvariceal upper gastrointestinal bleeding

Energy Technology Data Exchange (ETDEWEB)

Burke, Stephen J.; Weldon, Derik; Sun, Shiliang [University of Iowa, Department of Radiology, Iowa, IA (United States); Golzarian, Jafar [University of Iowa, Department of Radiology, Iowa, IA (United States); University of Iowa, Department of Radiology, Carver College of Medicine, Iowa, IA (United States)

2007-07-15

Nonvariceal upper gastrointestinal bleeding (NUGB) remains a major medical problem even after advances in medical therapy with gastric acid suppression and cyclooxygenase (COX-2) inhibitors. Although the incidence of upper gastrointestinal bleeding presenting to the emergency room has slightly decreased, similar decreases in overall mortality and rebleeding rate have not been experienced over the last few decades. Many causes of upper gastrointestinal bleeding have been identified and will be reviewed. Endoscopic, radiographic and angiographic modalities continue to form the basis of the diagnosis of upper gastrointestinal bleeding with new research in the field of CT angiography to diagnose gastrointestinal bleeding. Endoscopic and angiographic treatment modalities will be highlighted, emphasizing a multi-modality treatment plan for upper gastrointestinal bleeding. (orig.)

Upper Gastrointestinal (GI) Series

Science.gov (United States)

... standard barium upper GI series, which uses only barium a double-contrast upper GI series, which uses both air and ... evenly coat your upper GI tract with the barium. If you are having a double-contrast study, you will swallow gas-forming crystals that ...

LOFT advanced fuel rod instrumentation development

International Nuclear Information System (INIS)

Billeter, T.R.; Brown, R.L.; Chan, A.I.Y.; Day, C.K.; Meyers, S.C.; Sheen, E.M.; Stringer, J.L.

1978-01-01

Advanced fuel rod instrumentation for the Loss of Fluid Test (LOFT) reactor is being developed by the Hanford Engineering Development Laboratory for the Nuclear Regulatory Commission. This effort calls for development of sensors to measure fuel rod axial motion, fuel centerline temperature (to 2200 0 C), fuel rod plenum gas pressure (to 2500 psig), and plenum gas temperature (to 1500 0 F). A parallel test and evaluation of several modified commercial sensors was undertaken and will result in commercial availability of the final qualified sensors. Necessary test facilities were prepared for the development and evaluation effort. Tests to date indicate a three coil Linear Variable Differential Transformer (LVDT), operated from temperature compensating signal source and processing electronics, will meet the desired requirements

Subscale Carbon-Carbon Nozzle Extension Development and Hot Fire Testing in Support of Upper Stage Liquid Rocket Engines

Science.gov (United States)

Gradl, Paul; Valentine, Peter; Crisanti, Matthew; Greene, Sandy Elam

2016-01-01

Upper stage and in-space liquid rocket engines are optimized for performance through the use of high area ratio nozzles to fully expand combustion gases to low exit pressures increasing exhaust velocities. Due to the large size of such nozzles and the related engine performance requirements, carbon-carbon (C/C) composite nozzle extensions are being considered for use in order to reduce weight impacts. NASA and industry partner Carbon-Carbon Advanced Technologies (C-CAT) are working towards advancing the technology readiness level of large-scale, domestically-fabricated, C/C nozzle extensions. These C/C extensions have the ability to reduce the overall costs of extensions relative to heritage metallic and composite extensions and to decrease weight by 50%. Material process and coating developments have advanced over the last several years, but hot fire testing to fully evaluate C/C nozzle extensions in relevant environments has been very limited. NASA and C-CAT have designed, fabricated and hot fire tested multiple subscale nozzle extension test articles of various C/C material systems, with the goal of assessing and advancing the manufacturability of these domestically producible materials as well as characterizing their performance when subjected to the typical environments found in a variety of liquid rocket and scramjet engines. Testing at the MSFC Test Stand 115 evaluated heritage and state-of-the-art C/C materials and coatings, demonstrating the capabilities of the high temperature materials and their fabrication methods. This paper discusses the design and fabrication of the 1.2k-lbf sized carbon-carbon nozzle extensions, provides an overview of the test campaign, presents results of the hot fire testing, and discusses potential follow-on development work.

Functional studies in 79-year-olds. II. Upper extremity function.

Science.gov (United States)

Lundgren-Lindquist, B; Sperling, L

1983-01-01

As part of the Gerontological and Geriatric Population Study of 79-year-old people in Göteborg, a representative subsample comprising 112 women and 93 men took part in a study of upper extremity function. Thirty-eight per cent of the women and 37% of the men had disorders in the upper extremities. The investigation included tests of co-ordination, static strength in the key-grip and the transversal volar grip, power capacity in opening jars and a bottle, basal movements in the upper extremities in personal hygiene and dressing activities, function in the kitchen e.g. reaching shelves, manual tasks including tests of pronation and supination of the forearm. In the key-grip as well as in the transversal volar grip men showed a generally larger decrease in strength with age than women compared to 70-year-olds in a previous population study. Significant correlations were found between strength in the key-grip and the performance time in the test of co-ordination. Women produced about 66% of the muscular force of the men when opening jars. Significant correlations were found between strength in the transversal volar grip and the maximal torque for opening the jars. Female and male subjects who were not capable of handling the electric plug in the manual ability test had significantly weaker strength in the key-grip. The importance of designing products and adapting the environment so as to correspond to the functional capacity of the elderly, is emphasized.

Americium/Curium Melter 2A Pilot Tests

International Nuclear Information System (INIS)

Smith, M.E.; Fellinger, A.P.; Jones, T.M.; Miller, C.B.; Miller, D.H.; Snyder, T.K.; Stone, M.E.; Witt, D.C.

1998-05-01

Isotopes of americium (Am) and curium (Cm) were produced in the past at the Savannah River Site (SRS) for research, medical, and radiological applications. These highly radioactive and valuable isotopes have been stored in an SRS reprocessing facility for a number of years. Vitrification of this solution will allow the material to be more safely stored until it is transported to the DOE Oak Ridge Reservation for use in research and medical applications. To this end, the Am/Cm Melter 2A pilot system, a full-scale non- radioactive pilot plant of the system to be installed at the reprocessing facility, was designed, constructed and tested. The full- scale pilot system has a frit and aqueous feed delivery system, a dual zone bushing melter, and an off-gas treatment system. The main items which were tested included the dual zone bushing melter, the drain tube with dual heating and cooling zones, glass compositions, and the off-gas system which used for the first time a film cooler/lower melter plenum. Most of the process and equipment were proven to function properly, but several problems were found which will need further work. A system description and a discussion of test results will be given

Effect of caffeine on upper-body anaerobic performance in wrestlers in simulated competition-day conditions.

Science.gov (United States)

Aedma, Martin; Timpmann, Saima; Ööpik, Vahur

2013-12-01

Peak power (PP) and mean power (MP) attained in upper body sprint performance test are considered important factors for competitive success in wrestling. This study aimed to determine whether acute caffeine ingestion would better maintain PP and MP across a simulated competition day in wrestling. In a double-blind, counterbalanced, crossover study, 14 trained wrestlers ingested either placebo or 5 mg/kg caffeine and completed four 6-min upper body intermittent sprint performance tests with 30-min recovery periods between consecutive tests. PP and MP were recorded during and blood lactate concentration was measured before and after each test. Ratings of perceived fatigue (RPF) and exertion (RPE) were recorded before and after each test, respectively. Heart rate (HR) was monitored across the whole testing period. Mean power decreased across four tests in both trials (p caffeine trial. Both pretest blood lactate concentration and HR were higher in caffeine than in placebo trial (p caffeine ingestion has a partially detrimental effect on upper body intermittent sprint performance in trained wrestlers. Elevated HR and blood lactate levels observed between tests after caffeine ingestion suggest that caffeine may impair recovery between consecutive maximal efforts.

Starting manufacturing phase of ITER upper ports

Energy Technology Data Exchange (ETDEWEB)

Utin, Yuri, E-mail: yuri.utin@iter.org [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Alekseev, Alexander; Sborchia, Carlo; Choi, Changho; Albin, Vincent; Barabash, Vladimir; Davis, James [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Fabritsiev, Sergey [NTC Sintez, Efremov Inst., 189631 Metallostroy, St. Petersburg (Russian Federation); Giraud, Benoit; Guirao, Julio [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Koenig, Werner [MAN Diesel & Turbo SE, Werftstrasse 17, Deggendorf (Germany); Kedrov, Igor; Kuzmin, Evgeny [NTC Sintez, Efremov Inst., 189631 Metallostroy, St. Petersburg (Russian Federation); Levesy, Bruno; Martinez, Jean-Marc [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Prebeck, Markus [MAN Diesel & Turbo SE, Werftstrasse 17, Deggendorf (Germany); Privalova, Elena [NTC Sintez, Efremov Inst., 189631 Metallostroy, St. Petersburg (Russian Federation); Ranzinger, Franz [MAN Diesel & Turbo SE, Werftstrasse 17, Deggendorf (Germany); Savrukhin, Petr [Russian Federation ITER Domestic Agency, Kurchatov sq.1, 123182 Moscow (Russian Federation); Schiller, Thomas [MAN Diesel & Turbo SE, Werftstrasse 17, Deggendorf (Germany); and others

2015-10-15

Highlights: • The port plugs are attached to the ports with high-strength fasteners. • Tightening of the fasteners via inductive heating was tested. • A concept for the port/plug sealing with metal-type gaskets has progressed. • Manufacturing design of the Upper Ports is in progress. • A full-scale mock-up of double-wall part of the port stub extension is in manufacturing process – acceptable final tolerances are expected. - Abstract: The ITER Vacuum Vessel (VV) features upper, equatorial and lower ports. The upper and regular equatorial ports are occupied by the port plugs. Although the port design has been overall completed in the past, the design of some remaining interfaces was still in progress: in particular, the Sealing Flange package, which includes the high-vacuum seals and the plug fasteners. As the ITER construction phase has started, the procurement of the VV ports has been launched. The VV upper ports will be procured by the Russian Federation Domestic Agency. The main suppliers were selected and the manufacturing design of the first parts is in full progress now. Since the VV is classified at nuclear level N2, the design and manufacture of its components are to be compliant with the French RCC-MR code and regulations for nuclear pressure equipment in France. These regulations make a strong impact to the port design and manufacturing process.

Starting manufacturing phase of ITER upper ports

International Nuclear Information System (INIS)

Utin, Yuri; Alekseev, Alexander; Sborchia, Carlo; Choi, Changho; Albin, Vincent; Barabash, Vladimir; Davis, James; Fabritsiev, Sergey; Giraud, Benoit; Guirao, Julio; Koenig, Werner; Kedrov, Igor; Kuzmin, Evgeny; Levesy, Bruno; Martinez, Jean-Marc; Prebeck, Markus; Privalova, Elena; Ranzinger, Franz; Savrukhin, Petr; Schiller, Thomas

2015-01-01

Highlights: • The port plugs are attached to the ports with high-strength fasteners. • Tightening of the fasteners via inductive heating was tested. • A concept for the port/plug sealing with metal-type gaskets has progressed. • Manufacturing design of the Upper Ports is in progress. • A full-scale mock-up of double-wall part of the port stub extension is in manufacturing process – acceptable final tolerances are expected. - Abstract: The ITER Vacuum Vessel (VV) features upper, equatorial and lower ports. The upper and regular equatorial ports are occupied by the port plugs. Although the port design has been overall completed in the past, the design of some remaining interfaces was still in progress: in particular, the Sealing Flange package, which includes the high-vacuum seals and the plug fasteners. As the ITER construction phase has started, the procurement of the VV ports has been launched. The VV upper ports will be procured by the Russian Federation Domestic Agency. The main suppliers were selected and the manufacturing design of the first parts is in full progress now. Since the VV is classified at nuclear level N2, the design and manufacture of its components are to be compliant with the French RCC-MR code and regulations for nuclear pressure equipment in France. These regulations make a strong impact to the port design and manufacturing process.

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)

Analysis results from the Los Alamos 2D/3D program

International Nuclear Information System (INIS)

Boyack, B.E.; Cappiello, M.W.; Stumpf, H.; Shire, P.; Gilbert, J.; Hedstrom, J.

1986-01-01

Los Alamos National Laboratory is a participant in the 2D/3D program. Activities conducted at Los Alamos National Laboratory in support of 2D/3D program goals include analysis support of facility design, construction, and operation; provision of boundary and initial conditions for test facility operations based on analysis of pressurized water reactors; performance of pretest and posttest predictions and analyses; and use of experimental results to validate and assess the single- and multidimensional nonequilibrium features in the Transient Reactor Analysis Code (TRAC). During Fiscal Year 1986, Los Alamos conducted analytical assessment activities using data from the Cylindrical Core Test Facility and the Slab Core Test Facility. Los Alamos also continued to provide support analysis for the planning of Upper Plenum Test Facility experiments. Finally, Los Alamos either completed or is currently working on three areas of TRAC modeling improvement. In this paper, Los Alamos activities during Fiscal Year 1986 are summarized; several significant accomplishments are described in more detail to illustrate the work activities at Los Alamos

Uprated OMS engine for upper stage propulsion

Science.gov (United States)

Boyd, William C.

1986-01-01

The results of a pre-development component demonstration program on the use of a gas generator-driven turbopump that increases the Space Shuttle's Orbital Maneuvering Engine (OME) operating pressure are given. Tests and analysis confirm the the capability of the concept to meet or exceed performance and life requirements. Storable propellant upper stage concepts are also discussed.

Flow instability tests for a particle bed reactor nuclear thermal rocket fuel element

Science.gov (United States)

Lawrence, Timothy J.

1993-05-01

Recent analyses have focused on the flow stability characteristics of a particle bed reactor (PBR). These laminar flow instabilities may exist in reactors with parallel paths and are caused by the heating of the gas at low Reynolds numbers. This phenomena can be described as follows: several parallel channels are connected at the plenum regions and are stabilized by some inlet temperature and pressure; a perturbation in one channel causes the temperature to rise and increases the gas viscosity and reduces the gas density; the pressure drop is fixed by the plenum regions, therefore, the mass flow rate in the channel would decrease; the decrease in flow reduces the ability to remove the energy added and the temperature increases; and finally, this process could continue until the fuel element fails. Several analyses based on different methods have derived similar curves to show that these instabilities may exist at low Reynolds numbers and high phi's ((Tfinal Tinitial)/Tinitial). These analyses need to be experimentally verified.

Benefits of glucocorticoids in non-ambulant boys/men with Duchenne muscular dystrophy: A multicentric longitudinal study using the Performance of Upper Limb test.

Science.gov (United States)

Pane, Marika; Fanelli, Lavinia; Mazzone, Elena Stacy; Olivieri, Giorgia; D'Amico, Adele; Messina, Sonia; Scutifero, Marianna; Battini, Roberta; Petillo, Roberta; Frosini, Silvia; Sivo, Serena; Vita, Gian Luca; Bruno, Claudio; Mongini, Tiziana; Pegoraro, Elena; De Sanctis, Roberto; Gardani, Alice; Berardinelli, Angela; Lanzillotta, Valentina; Carlesi, Adelina; Viggiano, Emanuela; Cavallaro, Filippo; Sframeli, Maria; Bello, Luca; Barp, Andrea; Bianco, Flaviana; Bonfiglio, Serena; Rolle, Enrica; Palermo, Concetta; D'Angelo, Grazia; Pini, Antonella; Iotti, Elena; Gorni, Ksenija; Baranello, Giovanni; Bertini, Enrico; Politano, Luisa; Sormani, Maria Pia; Mercuri, Eugenio

2015-10-01

The aim of this study was to establish the possible effect of glucocorticoid treatment on upper limb function in a cohort of 91 non-ambulant DMD boys and adults of age between 11 and 26 years. All 91 were assessed using the Performance of Upper Limb test. Forty-eight were still on glucocorticoid after loss of ambulation, 25 stopped steroids at the time they lost ambulation and 18 were GC naïve or had steroids while ambulant for less than a year. At baseline the total scores ranged between 0 and 74 (mean 41.20). The mean total scores were 47.92 in the glucocorticoid group, 36 in those who stopped at loss of ambulation and 30.5 in the naïve group (p < 0.001). The 12-month changes ranged between -20 and 4 (mean -4.4). The mean changes were -3.79 in the glucocorticoid group, -5.52 in those who stopped at loss of ambulation and -4.44 in the naïve group. This was more obvious in the patients between 12 and 18 years and at shoulder and elbow levels. Our findings suggest that continuing glucocorticoids throughout teenage years and adulthood after loss of ambulation appears to have a beneficial effect on upper limb function. Copyright © 2015 The Authors. Published by Elsevier B.V. All rights reserved.

Measurements of bundle end flux peaking effects in 37-element CANDU PHW fuel

International Nuclear Information System (INIS)

French, P.M.

1977-10-01

Thermal neutron bundle end flux peaking factors have been measured in fresh 37-element Bruce reactor natural UO 2 clusters in heavy water moderator, both with and without staggered plenums at the fuel stack ends, in representative elements throughout the clusters. The measurements were made at a square lattice pitch of 28.58 cm with heavy water coolant. The results indicate that outer element peaking factors are 1.142 +- 0.009 for bundles containing no plenums, and 1.155 +- 0.006 and 1.177 +- 0.006 at the non-plenum and plenum element ends respectively, for bundles containing staggered plenums, irrespective of the azimuthal orientation between pairs of bundles. Measurements are also reported for bundles containing plenums in every outer element, for bundles separated by a stainless steel flux suppressor, for longer graphite plenums, and for changes in plenum and bundle gap lengths. Some theoretical comparisons with the results, reported by other authors, have been summarized. (author)

Concrete Math Manipulatives in Upper Elementary Mathematics Classrooms

Science.gov (United States)

Graham, Janina Maria

2013-01-01

Today's mathematics standards require teachers to use concrete math manipulatives (CMM) to increase the proficiency of students, but many upper elementary teachers fail to use these resources. The effects of this resource disuse may decrease student learning potential and impede successful standardized test results. This case study allows leaders…

A New Orthodontic Appliance with a Mini Screw for Upper Molar Distalization.

Science.gov (United States)

Ozkalayci, Nurhat; Yetmez, Mehmet

2016-01-01

The aim of this study is to present a new upper molar distalization appliance called Cise distalizer designed as intraoral device supported with orthodontic mini screw for upper permanent molar distalization. The new appliance consists of eight main components. In order to understand the optimum force level, the appliance under static loading is tested by using strain gage measurement techniques. Results show that one of the open coils produces approximately 300 gr distalization force. Cise distalizer can provide totally 600 gr distalization force. This range of force level is enough for distalization of upper first and second molar teeth.

The characterization and monitoring of metallic fuel breaches in EBR-2

International Nuclear Information System (INIS)

Pahl, R.G.; Batte, G.L.; Mikaili, R.; Lambert, J.D.B.; Hofman, G.L.

1991-01-01

This paper discusses the characterization and monitoring of metallic fuel breaches which is now a significant part of the Integral Fast Reactor fuel testing program at Argonne National Laboratory. Irradiation experience with failed metallic fuel now includes natural breaches in the plenum and fuel column regions in lead ''endurance'' tests as well as fuel column breaches in artificially-defected fuel which have operated for months in the run-beyond-cladding breach (RBCB) mode. Analyses of the fission gas (FG) release-to-birth (R/B) ratios of selected historical breaches have been completed and have proven to be very useful in differentiating between plenum and fuel column breaches

Seal Analysis for the Ares-I Upper Stage Fuel Tank Manhole Cover

Science.gov (United States)

Phillips, Dawn R.; Wingate, Robert J.

2010-01-01

Techniques for studying the performance of Naflex pressure-assisted seals in the Ares-I Upper Stage liquid hydrogen tank manhole cover seal joint are explored. To assess the feasibility of using the identical seal design for the Upper Stage as was used for the Space Shuttle External Tank manhole covers, a preliminary seal deflection analysis using the ABAQUS commercial finite element software is employed. The ABAQUS analyses are performed using three-dimensional symmetric wedge finite element models. This analysis technique is validated by first modeling a heritage External Tank liquid hydrogen tank manhole cover joint and correlating the results to heritage test data. Once the technique is validated, the Upper Stage configuration is modeled. The Upper Stage analyses are performed at 1.4 times the expected pressure to comply with the Constellation Program factor of safety requirement on joint separation. Results from the analyses performed with the External Tank and Upper Stage models demonstrate the effects of several modeling assumptions on the seal deflection. The analyses for Upper Stage show that the integrity of the seal is successfully maintained.

A study on fission product retention capability in a sodium coolant system

International Nuclear Information System (INIS)

Satoh, K.; Kubo, S.; Hashiguchi, Y.; Itooka, S.; Akatsu, Y.; Miyagi, K.; Wakamatsu, M.; Endo, H.; Tachino, T.

1992-01-01

Three kinds of separate model tests have been performed using water and air, focusing on the transport behavior of FP gas bubbles from subassembly outlets into a cover gas region, to study the dominant processes regarding the retention for volatiles ejected with inert gas into sodium after fuel failures. In the case that whole fuel pin failures occurring coherently in a subassembly were assumed, a periodic formation of globules was observed at the subassembly outlet. The globules rapidly broke up into small bubbles of less than 10 mm in mean diameter. The small bubbles at the top region had a tendency to be coalesced during rising through the upper plenum. As the coolant flow rate increased, bubble deformation and breakup were accelerated, but the bubble transport time did not vary remarkably. It is expected that bubbles in sodium would play in a similar way as in the water test, and the importance of the bubble behavior for the retention capability of volatiles has been confirmed. (author)

Flow Induced Vibration Program at Argonne National Laboratory

Science.gov (United States)

1984-01-01

The Argonne National Laboratory's Flow Induced Vibration Program, currently residing in the Laboratory's Components Technology Division is discussed. Throughout its existence, the overall objective of the program was to develop and apply new and/or improved methods of analysis and testing for the design evaluation of nuclear reactor plant components and heat exchange equipment from the standpoint of flow induced vibration. Historically, the majority of the program activities were funded by the US Atomic Energy Commission, the Energy Research and Development Administration, and the Department of Energy. Current DOE funding is from the Breeder Mechanical Component Development Division, Office of Breeder Technology Projects; Energy Conversion and Utilization Technology Program, Office of Energy Systems Research; and Division of Engineering, Mathematical and Geosciences, office of Basic Energy Sciences. Testing of Clinch River Breeder Reactor upper plenum components was funded by the Clinch River Breeder Reactor Plant Project Office. Work was also performed under contract with Foster Wheeler, General Electric, Duke Power Company, US Nuclear Regulatory Commission, and Westinghouse.

PBDOWN - a computer code for simulating core material discharge and thermal to mechanical energy conversion in LMFBR hypothetical accidents

International Nuclear Information System (INIS)

Royl, P.

1981-01-01

PBDOWN is a computer code that simulates the blowdown of confined boiling materials ('pools') into a colder upper coolant plenum as time dependent ejection and expansion with consideration of a few selected exchange processes. Its application is restricted to situations resulting from hypothetical loss of flow (LOF) accidents in LMFBR's, where enough voiding has occured, that in core sodium vapor pressures become negligible. PBDOWN considers one working fluid for the discharge process (either fuel or steel) and a maximum of two working fluids (either fuel and sodium or steel and sodium) for the expansion process in the upper coolant plenum. Entrainment of sodium at the accelerated bubble liquid interfaces is mechanistically calculated by a Taylor instability entrainment model. Simulation of a hemispherical expansion form together with this mechanistic entrainment model gives a new integrated calculation of the time dependent sodium mass in the bubble. The paper summarizes the basic equations and assumptions of this computer model. Sample results compare different heat transfer and Na entrainment models during steel and fuel driven discharge processes. Mechanistic sodium entrainment simulation for SNR-type reactors coupled with a realistic heat transfer model is shown to reduce the integral mechanical work potential by a factor of 1.3 to 2.0 over the isentropic energy of the discharge working fluids. (orig.)

Experimental Study of the Twin Turbulent Water Jets Using Laser Doppler Anemometry for Validating Numerical Models

International Nuclear Information System (INIS)

Wang Huhu; Lee Saya; Hassan, Yassin A.; Ruggles, Arthur E.

2014-01-01

The design of next generation (Gen. IV) high-temperature nuclear reactors including gas-cooled and sodium-cooled ones involves massive numerical works especially the Computational Fluid Dynamics (CFD) simulations. The high cost of large-scale experiments and the inherent uncertainties existing in the turbulent models and wall functions of any CFD codes solving Reynolds-averaged Navier-Stokes (RANS) equations necessitate the high-spacial experimental data sets for benchmarking the simulation results. In Gen. IV conceptual reactors, the high- temperature flows mix in the upper plenum before entering the secondary cooling system. The mixing condition should be accurately estimated and fully understood as it is related to the thermal stresses induced in the upper plenum and the magnitudes of output power oscillations due to any changes of primary coolant temperature. The purpose of this study is to use Laser Doppler Anemometry (LDA) technique to measure the flow field of two submerged parallel jets issuing from two rectangular channels. The LDA data sets can be used to validate the corresponding simulation results. The jets studied in this work were at room temperature. The turbulent characteristics including the distributions of mean velocities, turbulence intensities, Reynolds stresses were studied. Uncertainty analysis was also performed to study the errors involved in this experiment. The experimental results in this work are valid for benchmarking any steady-state numerical simulations using turbulence models to solve RANS equations. (author)

Analysis of the ATLAS Cold Leg Top-Slot Break Experiment Using the MARS Code

Energy Technology Data Exchange (ETDEWEB)

Ha, T. W.; Jeong, J. J. [Pusan National University, Busan (Korea, Republic of)

2016-10-15

During a small-break loss of coolant accident (SBLOCA) or intermediate-break loss of coolant accident (IBLOCA) in a PWR, such as the APR1400, the steam volume in the reactor vessel upper plenum may continue to expand until the liquid phase in the horizontal intermediate legs is released, called loop seal clearing (LSC), due to the increase of the pressure in the upper plenum. A domestic standard problem (DSP) exercise using the ATLAS facility was promoted in order to transfer the database to domestic nuclear industries. For 4th DSP (DSP-04), the ATLAS cold leg top-slot break experiment was postulated. For the DSP-04, main concerns are to predict the LSC and LSR having a significantly effect on the behavior of the system under long term cooling. In this study, we simulated the ATLAS cold leg top-slot break experiment using the MARS code and the predicted LSC and LSR are compared to experimental results. The LTC-CL-04R was simulated using the MARS code. Most of the predicted results agree well with the experimental data. However, the timing of LSC and LSR is slightly different from each other and, thus, the behavior of the primary system is slightly different. The core heat up was not observed in the experiment and the calculation.

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

International Nuclear Information System (INIS)

Ozawa, Takayuki; Abe, Tomoyuki

2006-03-01

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

Upper respiratory tract infection, heterologous immunisation and meningococcal disease

NARCIS (Netherlands)

Scholten, R. J.; Bijlmer, H. A.; Tobi, H.; Dankert, J.; Bouter, L. M.

1999-01-01

To test the hypothesis that an episode of upper respiratory tract infection or heterologous immunisation is a predisposing factor for the occurrence of meningococcal disease, data from 377 cases of meningococcal disease and their household contacts (n = 1124) were analysed by conditional logistic

Upper extremity sensorimotor control among collegiate football players.

Science.gov (United States)

Laudner, Kevin G

2012-03-01

Injuries stemming from shoulder instability are very common among athletes participating in contact sports, such as football. Previous research has shown that increased laxity negatively affects the function of the sensorimotor system potentially leading to a pathological cycle of shoulder dysfunction. Currently, there are no data detailing such effects among football players. Therefore, the purpose of this study was to examine the differences in upper extremity sensorimotor control among football players compared with that of a control group. Forty-five collegiate football players and 70 male control subjects with no previous experience in contact sports participated. All the subjects had no recent history of upper extremity injury. Each subject performed three 30-second upper extremity balance trials on each arm. The balance trials were conducted in a single-arm push-up position with the test arm in the center of a force platform and the subjects' feet on a labile device. The trials were averaged, and the differences in radial area deviation between groups were analyzed using separate 1-way analyses of variance (p football players showed significantly more radial area deviation of the dominant (0.41 ± 1.23 cm2, p = 0.02) and nondominant arms (0.47 ± 1.63 cm2, p = 0.03) when compared with the control group. These results suggest that football players may have decreased sensorimotor control of the upper extremity compared with individuals with no contact sport experience. The decreased upper extremity sensorimotor control among the football players may be because of the frequent impacts accumulated during football participation. Football players may benefit from exercises that target the sensorimotor system. These findings may also be beneficial in the evaluation and treatment of various upper extremity injuries among football players.

The modeling and analysis of in-vessel corium/structure interaction in boiling water reactors

International Nuclear Information System (INIS)

Podowski, M.Z.; Kurul, N.; Kim, S.-W.; Baltyn, W.; Frid, W.

1997-01-01

A complete stand-alone state-of-the-art model has been developed of the interaction between corium debris in the lower plenum and the RPV walls and internal structures, including the vessel failure mechanisms. This new model has been formulated as a set of consistent computer modules which could be linked with other existing models and/or computer codes. The combined lower head and lower plenum modules were parametrically tested and applied to predict the consequences of a hypothetical station blackout in a Swedish BWR. (author)

The thermal-hydraulic fundamental experiment for the development of characteristic droplet behavior model

Energy Technology Data Exchange (ETDEWEB)

No, Hee Cheon; Moon, Young Min; Lee, Kyung Won; Lee, Sang Ik; Kim, Eung Su; Kim, Ji Hwan [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

2005-04-15

The main objective of this study are to develop the droplet behavior model in the upper plenum during the reflood phase of LBLOCA. We investigated the water accumulation and CCFL phenomena at UCSP, liquid carry-over rate into a hot-leg, the de-entrainment efficiency of vertical rods, and the de-entrainment rate on the inside wall of horizontal pipe. In addition, we investigated the parametric effects of various experimental conditions on the water accumulation and CCFL, and on the de-entrainment phenomena.

Computer vision for shoe upper profile measurement via upper and sole conformal matching

Science.gov (United States)

Hu, Zhongxu; Bicker, Robert; Taylor, Paul; Marshall, Chris

2007-01-01

This paper describes a structured light computer vision system applied to the measurement of the 3D profile of shoe uppers. The trajectory obtained is used to guide an industrial robot for automatic edge roughing around the contour of the shoe upper so that the bonding strength can be improved. Due to the specific contour and unevenness of the shoe upper, even if the 3D profile is obtained using computer vision, it is still difficult to reliably define the roughing path around the shape. However, the shape of the corresponding shoe sole is better defined, and it is much easier to measure the edge using computer vision. Therefore, a feasible strategy is to measure both the upper and sole profiles, and then align and fit the sole contour to the upper, in order to obtain the best fit. The trajectory of the edge of the desired roughing path is calculated and is then smoothed and interpolated using NURBS curves to guide an industrial robot for shoe upper surface removal; experiments show robust and consistent results. An outline description of the structured light vision system is given here, along with the calibration techniques used.

A summary of the low upper shelf toughness safety margin issue

International Nuclear Information System (INIS)

Merkle, J.G.

1991-01-01

The low upper shelf toughness issue has a long history, beginning with the choice of materials for the submerged arc welding process, but also potentially involving the use of A302-B plate. Criteria for vessels containing low upper shelf materials have usually been expressed in terms of the Charpy upper shelf impact energy. Although these criteria have had several different bases, the range of limiting values for wall thicknesses approaching 229 mm (9 in.) has remained between 54 to 68J (40 to 50 ft lbs). Allowable values for vessels with thinner walls and/or only circumferential low upper shelf welds might conceivably be less. A decision on criteria to be incorporated into the ASME Code is now being made. Choices to be made concern the method for estimating the decrease in upper shelf impact energy, flaw geometry for circumferential welds, statistical significance of toughness values, the choice between J D and J M , reference pressure, safety factors and the inclusion of tearing stability calculations by means of R curve extrapolation. NRC research programs have contributed significantly to the resolution of the low upper shelf issue. These programs embrace all aspects of the issue, including material characterization, large scale testing, analysis and criteria development. 52 refs., 5 figs

Dual-focus Magnification, High-Definition Endoscopy Improves Pathology Detection in Direct-to-Test Diagnostic Upper Gastrointestinal Endoscopy.

Science.gov (United States)

Bond, Ashley; Burkitt, Michael D; Cox, Trevor; Smart, Howard L; Probert, Chris; Haslam, Neil; Sarkar, Sanchoy

2017-03-01

In the UK, the majority of diagnostic upper gastrointestinal (UGI) endoscopies are a result of direct-to-test referral from the primary care physician. The diagnostic yield of these tests is relatively low, and the burden high on endoscopy services. Dual-focus magnification, high-definition endoscopy is expected to improve detection and classification of UGI mucosal lesions and also help minimize biopsies by allowing better targeting. This is a retrospective study of patients attending for direct-to-test UGI endoscopy from January 2015 to June 2015. The primary outcome of interest was the identification of significant pathology. Detection of significant pathology was modelled using logistic regression. 500 procedures were included. The mean age of patients was 61.5 (±15.6) years; 60.8% of patients were female. Ninety-four gastroscopies were performed using dual-focus magnification high-definition endoscopy. Increasing age, male gender, type of endoscope, and type of operator were all identified as significant factors influencing the odds of detecting significant mucosal pathology. Use of dual-focus magnification, high-definition endoscopy was associated with an odds ratio of 1.87 (95%CI 1.11-3.12) favouring the detection of significant pathology. Subsequent analysis suggested that the increased detection of pathology during dual-focus magnification, high-definition endoscopy also influenced patient follow-up and led to a 3.0 fold (p=0.04) increase in the proportion of patients entered into an UGI endoscopic surveillance program. Dual-focus magnification, high-definition endoscopy improved the diagnostic yield for significant mucosal pathology in patients referred for direct-to-test endoscopy. If this finding is recapitulated elsewhere it will have substantial impact on the provision of UGI endoscopic services.

Upper GI Bleeding in Children

Science.gov (United States)

Upper GI Bleeding in Children What is upper GI Bleeding? Irritation and ulcers of the lining of the esophagus, stomach or duodenum can result in upper GI bleeding. When this occurs the child may vomit blood ...

A New Orthodontic Appliance with a Mini Screw for Upper Molar Distalization

Directory of Open Access Journals (Sweden)

Nurhat Ozkalayci

2016-01-01

Full Text Available The aim of this study is to present a new upper molar distalization appliance called Cise distalizer designed as intraoral device supported with orthodontic mini screw for upper permanent molar distalization. The new appliance consists of eight main components. In order to understand the optimum force level, the appliance under static loading is tested by using strain gage measurement techniques. Results show that one of the open coils produces approximately 300 gr distalization force. Cise distalizer can provide totally 600 gr distalization force. This range of force level is enough for distalization of upper first and second molar teeth.

"Ballistic Six" Upper-Extremity Plyometric Training for the Pediatric Volleyball Players.

Science.gov (United States)

Turgut, Elif; Cinar-Medeni, Ozge; Colakoglu, Filiz F; Baltaci, Gul

2017-09-19

The Ballistic Six exercise program includes commonly used upper-body exercises, and the program is recommended for overhead throwing athletes. The purpose of the current study was to investigate the effects of a 12-week the Ballistic Six upper-extremity plyometric training program on upper-body explosive power, endurance, and reaction time in pediatric overhead athletes. Twenty-eight female pediatric volleyball players participated in the study. The participants were randomly divided into 2 study groups: an intervention group (upper-extremity plyometric training in addition to the volleyball training; n = 14) and a control group (the volleyball training only; n = 14). All the participants were assessed before and after a 12-week training program for upper-body power, strength and endurance, and reaction time. Statistical comparison was performed using an analysis of variance test. Comparisons showed that after a 12-week training program, the Ballistic Six upper-body plyometric training program resulted in more improvements in an overhead medicine ball throwing distance and a push-up performance, as well as greater improvements in the reaction time in the nonthrowing arm when compared with control training. In addition, a 12-week training program was found to be effective in achieving improvements in the reaction time in the throwing arm for both groups similarly. Compared with regular training, upper-body plyometric training resulted in additional improvements in upper-body power and strength and endurance among pediatric volleyball players. The findings of the study provide a basis for developing training protocols for pediatric volleyball players.

Dynamic PIV measurement on the effect of sound wave in upper plenum of boiling water reactor

International Nuclear Information System (INIS)

Kumagai, Kosuke; Someya, Satoshi; Okamoto, Koji

2008-01-01

In one of the power uprated plants in the United States, the steam dryer breakages due to fatigue fracture occurred. It is conceivable that the increased steam flow passing through the branches caused a self-induced vibration with the propagation of sound wave into the steam-dome. The resonance among the structure, flow and the pressure fluctuation resulted in the breakages. To understand the basic mechanism of the resonance, previous researches were done by a point measurement of the pressure and by a phase averaged measurement of the flow, while it was difficult to detect the interaction among them by the conventional method. In the preliminary study, Dynamic Particle Image Velocimetry (PIV) System was applied to investigate the effect of sound on the flow. (author)

Primed Physical Therapy Enhances Recovery of Upper Limb Function in Chronic Stroke Patients.

Science.gov (United States)

Ackerley, Suzanne J; Byblow, Winston D; Barber, P Alan; MacDonald, Hayley; McIntyre-Robinson, Andrew; Stinear, Cathy M

2016-05-01

Recovery of upper limb function is important for regaining independence after stroke. To test the effects of priming upper limb physical therapy with intermittent theta burst stimulation (iTBS), a form of noninvasive brain stimulation. Eighteen adults with first-ever chronic monohemispheric subcortical stroke participated in this randomized, controlled, triple-blinded trial. Intervention consisted of priming with real or sham iTBS to the ipsilesional primary motor cortex immediately before 45 minutes of upper limb physical therapy, daily for 10 days. Changes in upper limb function (Action Research Arm Test [ARAT]), upper limb impairment (Fugl-Meyer Scale), and corticomotor excitability, were assessed before, during, and immediately, 1 month and 3 months after the intervention. Functional magnetic resonance images were acquired before and at one month after the intervention. Improvements in ARAT were observed after the intervention period when therapy was primed with real iTBS, but not sham, and were maintained at 1 month. These improvements were not apparent halfway through the intervention, indicating a dose effect. Improvements in ARAT at 1 month were related to balancing of corticomotor excitability and an increase in ipsilesional premotor cortex activation during paretic hand grip. Two weeks of iTBS-primed therapy improves upper limb function at the chronic stage of stroke, for at least 1 month postintervention, whereas therapy alone may not be sufficient to alter function. This indicates a potential role for iTBS as an adjuvant to therapy delivered at the chronic stage. © The Author(s) 2015.

A new simple three-dimensional method to characterize upper airway in orthognathic surgery patient

DEFF Research Database (Denmark)

Di Carlo, Gabriele; Fernandez Gurani, Sirwan; Pinholt, Else Marie

2017-01-01

.2% for cross-sectional measurements, and 0.3 to 2.5% for linear measurements. No systematic errors were detected. CONCLUSIONS: This new proposed definition of upper airway boundaries was shown to be technical feasible and tested to be reliable in measuring upper airway in patients undergoing orthognathic......OBJECTIVES: To develop and validate a new reproducible 3D upper airway analysis based on skeletal structures not involved in the modification, which occur during orthognathic surgery. METHODS: From retrospective cohort of orthognathic surgically treated patients, pre- and postsurgical CBCT...

Usability testing of gaming and social media applications for stroke and cerebral palsy upper limb rehabilitation.

Science.gov (United States)

Valdés, Bulmaro A; Hilderman, Courtney G E; Hung, Chai-Ting; Shirzad, Navid; Van der Loos, H F Machiel

2014-01-01

As part of the FEATHERS (Functional Engagement in Assisted Therapy Through Exercise Robotics) project, two motion tracking and one social networking applications were developed for upper limb rehabilitation of stroke survivors and teenagers with cerebral palsy. The project aims to improve the engagement of clients during therapy by using video games and a social media platform. The applications allow users to control a cursor on a personal computer through bimanual motions, and to interact with their peers and therapists through the social media. The tracking applications use either a Microsoft Kinect or a PlayStation Eye camera, and the social media application was developed on Facebook. This paper presents a usability testing of these applications that was conducted with therapists from two rehabilitation clinics. The "Cognitive Walkthrough" and "Think Aloud" methods were used. The objectives of the study were to investigate the ease of use and potential issues or improvements of the applications, as well as the factors that facilitate and impede the adoption of technology in current rehabilitation programs.

Experimental investigation of boiling-water nuclear-reactor parallel-channel effects during a postulated loss-of-coolant accident

International Nuclear Information System (INIS)

Conlon, W.M.; Lahey, R.T. Jr.

1982-12-01

This report describes an experimental study of the influence of parallel channel effects (PCE) on the distribution of emergency core spray cooling water in a Boiling Water Nuclear Reactor (BWR) following a postulated design basis loss of coolant accident (LCA). The experiments were conducted in a scaled test section in which the reactor coolant was simulated by Freon-114 at conditions similar to those postulated to occur in the reactor vessel shortly after a LOCA. A BWR/4 was simulated by a (PCE) test section which contained three parallel heated channels to simulate fuel assemblies; a core bypass channel, and a jet pump channel. The test section also inlcuded scaled regions to simulate the lower and upper plena, downcomer, and steam separation regions of a BWR. A series of nine transient experiments were conducted, in which the lower plenum vaporization rate and heater rod power were varied while the core spray flow rate was held constant to simulate that of a BWR/4. During these experiments the flow distribution and heat transfer phenomena were observed and measured

Immediate effects of plantar inputs on the upper half muscles and upright posture: a preliminary study.

Science.gov (United States)

Ciuffolo, Fabio; Ferritto, Anna L; Muratore, Filippo; Tecco, Simona; Testa, Mauro; D'Attilio, Michele; Festa, Felice

2006-01-01

This purpose of this study was to investigate the immediate effects of plantar inputs on both the upper half muscle activity (anterior temporal, masseter, digastric, sternocleidomastoid, upper and lower trapezius, cervical) and the body posture, by means of electromyography (EMG) and vertical force platform, respectively. Twenty four (24) healthy adults, between the ages of 24 and 31 years (25.3 +/- 1.9), with no history of craniomandibular disorder or systemic musculoskeletal dysfunction, were randomly divided into two groups: test group (fourteen subjects) and control group (ten subjects). A first recording session (TO) measured the baseline EMG and postural patterns of both groups. After this session, the test group wore test shoes with insoles that stimulated the plantar surfaces, while the control group wore placebo shoes. After one hour, a second set of measurements (T1) were performed. Significant differences between the groups at baseline were observed in the left anterior temporal, left cervical, and left upper trapezius, as well as at T1 in the left anterior temporal and right upper trapezius (p postural blindness in the test group compared to the control group was observed. Further studies are warranted to investigate the short and long-term effects of this type of insole, in patients with both craniomandibular-cervical and lower extremity disorders.

49 CFR 572.125 - Upper and lower torso assemblies and torso flexion test procedure.

Science.gov (United States)

2010-10-01

... of the dummy, the pull cable, and the load cell as shown in Figure N5. (7) Apply a tension force in the midsagittal plane to the pull cable as shown in Figure N5 at any upper torso deflection rate... determine the stiffness effects of the lumbar spine (drawing 127-3002), including cable (drawing 127-8095...

THE INFLUENCE OF LOWER LIMB MOVEMENT ON UPPER LIMB MOVEMENT SYMMETRY WHILE SWIMMING THE BREASTSTROKE

OpenAIRE

M. Jaszczak

2011-01-01

This study 1) examined the influence of lower limb movement on upper limb movement symmetry, 2) determined the part of the propulsion phase displaying the greatest hand movement asymmetry, 3) diagnosed the range of upper limb propulsion phase which is the most prone to the influence of the lower limbs while swimming the breaststroke. Twenty-four participants took part in two tests. Half of them performed an asymmetrical leg movement. The propulsion in the first test was generated by four limb...

Is the 6-minute pegboard and ring test valid to evaluate upper limb function in hospitalized patients with acute exacerbation of COPD?

Directory of Open Access Journals (Sweden)

Felisberto RM

2018-05-01

Full Text Available Rosimeire Marcos Felisberto,1 Cassia Fabiane de Barros,1 Kelly Cristina Albanezi Nucci,1 Andre Luis Pereira de Albuquerque,1 Elaine Paulin,2 Christina May Moran de Brito,1 Wellington Pereira Yamaguti1 1Hospital Sírio-Libanês, São Paulo, SP, Brazil; 2Universidade do Estado de Santa Catarina (UDESC, Florianópolis, SC, Brazil Background: The 6-minute pegboard and ring test (6-PBRT is a useful test for assessing the functional capacity of upper limbs in patients with stable COPD. Although 6-PBRT has been validated in stable patients, the possibility of a high floor effect could compromise the validity of the test in the hospital setting. The aim of this study was to verify the convergent validity of 6-PBRT in hospitalized patients with acute exacerbation of COPD (AECOPD. Methods: A cross-sectional study was conducted in a tertiary hospital. Patients who were hospitalized due to AECOPD and healthy elderly participants, voluntarily recruited from the community, were considered for inclusion. All participants underwent a 6-PBRT. Isokinetic evaluation to measure the strength and endurance of elbow flexors and extensors, handgrip strength (HGS, spirometry testing, the modified Pulmonary Functional Status Dyspnea Questionnaire (PFSDQ-M, the COPD assessment test (CAT, and symptoms of dyspnea and fatigue were all measured as comparisons for convergent validity. Good convergent validity was considered if >75% of these hypotheses could be confirmed (correlation coefficient>0.50. Results: A total of 17 patients with AECOPD (70.9±5.1 years and forced expiratory volume in 1 second [FEV1] of 41.8%±17.9% of predicted and 11 healthy elderly subjects were included. The HGS showed a significant strong correlation with 6-PBRT performance (r=0.70; p=0.002. The performance in 6-PBRT presented a significant moderate correlation with elbow flexor torque peak (r=0.52; p=0.03 and elbow extensor torque peak (r=0.61; p=0.01. The total muscular work of the 15

Reduction of upper shelf energy of highly irradiated RPV steels

Energy Technology Data Exchange (ETDEWEB)

Otaka, M.; Osaki, T. [Japan Nuclear Energy Safety Organization (Japan)

2004-07-01

It is well known that as the embrittlement due to neutron irradiation of reactor pressure vessel (RPV) steels, there is the tendency of the decrease in Charpy absorbed energy at upper shelf region (USE), in addition to the shift of ductile-brittle transition temperature. Concerning to the regulation of the upper shelf region, no method is provided to evaluate integrity for RPV steels with USE of less than 68J in Japanese codes. Under the circumstance, the reduction tendency of USE using simulated Japanese RPV steels, irradiated by fast neutron up to 1 x 10{sup 24} n/m{sup 2}, E>1 MeV in the OECD Halden test reactor, was investigated to establish the basis of the USE prediction after 60 year plant operation for the integrity assessment of the RPVs. This paper describes the results of an atom probe tomography characterization of irradiated steels. A new form of USE prediction equation was developed based on the atom probe tomography characterization and the Charpy impact test results of the irradiated steels. And, the USE prediction equations have been determined through the regression analysis of the test reactor data combined with Japanese surveillance test data. (orig.)

Lithofacies and sequence stratigraphic description of the upper part of the Avon Park Formation and the Arcadia Formation in U.S. Geological Survey G–2984 test corehole, Broward County, Florida

Science.gov (United States)

Cunningham, Kevin J.; Robinson, Edward

2017-07-18

Rock core and sediment from U.S. Geological Survey test corehole G–2984 completed in 2011 in Broward County, Florida, provide an opportunity to improve the understanding of the lithostratigraphic, sequence stratigraphic, and hydrogeologic framework of the intermediate confining unit and Floridan aquifer system in southeastern Florida. A multidisciplinary approach including characterization of sequence stratigraphy, lithofacies, ichnology, foraminiferal paleontology, depositional environments, porosity, and permeability was used to describe the geologic samples from this test corehole. This information has produced a detailed characterization of the lithofacies and sequence stratigraphy of the upper part of the middle Eocene Avon Park Formation and Oligocene to middle Miocene Arcadia Formation. This enhancement of the knowledge of the sequence stratigraphic framework is especially important, because subaerial karst unconformities at the upper boundary of depositional cycles at various hierarchical scales are commonly associated with secondary porosity and enhanced permeability in the Floridan aquifer system.

Playing piano can improve upper extremity function after stroke: case studies.

Science.gov (United States)

Villeneuve, Myriam; Lamontagne, Anouk

2013-01-01

Music-supported therapy (MST) is an innovative approach that was shown to improve manual dexterity in acute stroke survivors. The feasibility of such intervention in chronic stroke survivors and its longer-term benefits, however, remain unknown. The objective of this pilot study was to estimate the short- and long-term effects of a 3-week piano training program on upper extremity function in persons with chronic stroke. A multiple pre-post sequential design was used, with measurements taken at baseline (week0, week3), prior to (week6) and after the intervention (week9), and at 3-week follow-up (week12). Three persons with stroke participated in the 3-week piano training program that combined structured piano lessons to home practice program. The songs, played on an electronic keyboard, involved all 5 digits of the affected hand and were displayed using a user-friendly MIDI program. After intervention, all the three participants showed improvements in their fine (nine hole peg test) and gross (box and block test) manual dexterity, as well as in the functional use of the upper extremity (Jebsen hand function test). Improvements were maintained at follow-up. These preliminary results support the feasibility of using an MST approach that combines structured lessons to home practice to improve upper extremity function in chronic stroke.

Playing Piano Can Improve Upper Extremity Function after Stroke: Case Studies

Directory of Open Access Journals (Sweden)

Myriam Villeneuve

2013-01-01

Full Text Available Music-supported therapy (MST is an innovative approach that was shown to improve manual dexterity in acute stroke survivors. The feasibility of such intervention in chronic stroke survivors and its longer-term benefits, however, remain unknown. The objective of this pilot study was to estimate the short- and long-term effects of a 3-week piano training program on upper extremity function in persons with chronic stroke. A multiple pre-post sequential design was used, with measurements taken at baseline (week0, week3, prior to (week6 and after the intervention (week9, and at 3-week follow-up (week12. Three persons with stroke participated in the 3-week piano training program that combined structured piano lessons to home practice program. The songs, played on an electronic keyboard, involved all 5 digits of the affected hand and were displayed using a user-friendly MIDI program. After intervention, all the three participants showed improvements in their fine (nine hole peg test and gross (box and block test manual dexterity, as well as in the functional use of the upper extremity (Jebsen hand function test. Improvements were maintained at follow-up. These preliminary results support the feasibility of using an MST approach that combines structured lessons to home practice to improve upper extremity function in chronic stroke.

Geno-toxicity assay of sediment and water samples from the Upper Silesia post-mining areas, Poland by means of Allium-test

Energy Technology Data Exchange (ETDEWEB)

Geras' kin, S.; Oudalova, A.; Michalik, B.; Dikareva, N.; Dikarev, V. [Russian Institute of Agricultural Radiology & Agroecology RAAS, Obninsk (Russian Federation)

2011-05-15

Genotoxic potential of two environmental compartments (water and sediment) from the Upper Silesia Coal Basin (USCB), Poland were evaluated and compared by employing root meristem cells of Allium cepa. The clear genotoxic effect of water and sediment sampled was shown, with an important contribution of severe types of cytogenetic abnormalities. The most biologically relevant pollutants were revealed through multivariate statistical analysis of relationships between biological effects registered and the environment contamination. Overall, results of simultaneous use of conventional monitoring methods and biological tests suggested that contemporary levels of persistent pollutants in post-mining areas of the USCB may enhance the risk both for human health and biological components of natural ecosystems.

Testing and analyses of a high temperature thermal barrier for gas-cooled reactors

International Nuclear Information System (INIS)

Black, W.E.; Betts, W.S.; Felten, P.

1979-01-01

A full size, multi-panel section of a thermal barrier system was fabricated from a nickel-base superalloy and a combination of fibrous blanket insulation materials for specific application in a steam cycle gas-cooled nuclear reactor. The 2.4 m square array was representative of the sidewall of the lower core outlet plenum and included coverplates, attachments, seals, and a simulated water-cooled liner. Testing was conducted in a reactor grade, helium-filled chamber at 816 0 C for 100 hours, which established a normal (baseline) condition; 982 0 C for 10 hours, which satisfied an emergency condition; 1093 0 C for 1 hour, which simulated a faulted condition; and 1260 0 C, which was a non-design condition test to demonstrate the temperature overshoot capability of the system. Post-test examination indicated: (1) an acceptable performance by the anti-friction chromium carbide (Cr 3 C 2 ) coating; (2) no significant galling between non-coated surfaces; (3) no distortion of attachment fixtures; (4) predictable coverplate deflection during the design conditions testing (normal, emergency, and faulted); and (5) considerable plastic deformation resulting from the near-incipient melting temperature. (orig.)

Development and reliability of the rating of compensatory movements in upper limb prosthesis wearers during work-related tasks.

Science.gov (United States)

van der Laan, Tallie M J; Postema, Sietke G; Reneman, Michiel F; Bongers, Raoul M; van der Sluis, Corry K

2018-02-10

Reliability study. Quantifying compensatory movements during work-related tasks may help to prevent musculoskeletal complaints in individuals with upper limb absence. (1) To develop a qualitative scoring system for rating compensatory shoulder and trunk movements in upper limb prosthesis wearers during the performance of functional capacity evaluation tests adjusted for use by 1-handed individuals (functional capacity evaluation-one handed [FCE-OH]); (2) to examine the interrater and intrarater reliability of the scoring system; and (3) to assess its feasibility. Movement patterns of 12 videotaped upper limb prosthesis wearers and 20 controls were analyzed. Compensatory movements were defined for each FCE-OH test, and a scoring system was developed, pilot tested, and adjusted. During reliability testing, 18 raters (12 FCE experts and 6 physiotherapists/gait analysts) scored videotapes of upper limb prosthesis wearers performing 4 FCE-OH tests 2 times (2 weeks apart). Agreement was expressed in % and kappa value. Feasibility (focus area's "acceptability", "demand," and "implementation") was determined by using a questionnaire. After 2 rounds of pilot testing and adjusting, reliability of a third version was tested. The interrater reliability for the first and second rating sessions were к = 0.54 (confidence interval [CI]: 0.52-0.57) and к = 0.64 (CI: 0.61-0.66), respectively. The intrarater reliability was к = 0.77 (CI: 0.72-0.82). The feasibility was good but could be improved by a training program. It seems possible to identify compensatory movements in upper limb prosthesis wearers during the performance of FCE-OH tests reliably by observation using the developed observational scoring system. Interrater reliability was satisfactory in most instances; intrarater reliability was good. Feasibility was established. Copyright © 2018 Hanley & Belfus. Published by Elsevier Inc. All rights reserved.

Final Report Integrated DM1200 Melter Testing Of Bubbler Configurations Using HLW AZ-101 Simulants VSL-04R4800-4, Rev. 0, 10/5/04

International Nuclear Information System (INIS)

Kruger, A.A.; Matlack, K.S.; Gong, W.; Bardakci, T.; D'Angelo, N.A.; Lutze, W.; Callow, R.A.; Brandys, M.; Kot, W.K.; Pegg, I.L.

2011-01-01

This report documents melter and off-gas performance results obtained on the DM1200 HLW Pilot Melter during processing of AZ-101 HLW simulants. The tests reported herein are a subset of six tests from a larger series of tests described in the Test Plan for the work; results from the other tests have been reported separately. The solids contents of the melter feeds were based on the WTP baseline value for the solids content of the feeds from pretreatment which changed during these tests from 20% to 15% undissolved solids resulting in tests conducted at two feed solids contents. Based on the results of earlier tests with single outlet 'J' bubblers, initial tests were performed with a total bubbling rate of 651 pm. The first set of tests (Tests 1A-1E) addressed the effects of skewing this total air flow rate back and forth between the two installed bubblers in comparison to a fixed equal division of flow between them. The second set of tests (2A-2D) addressed the effects of bubbler depth. Subsequently, as the location, type and number of bubbling outlets were varied, the optimum bubbling rate for each was determined. A third (3A-3C) and fourth (8A-8C) set of tests evaluated the effects of alternative bubbler designs with two gas outlets per bubbler instead of one by placing four bubblers in positions simulating multiple-outlet bubblers. Data from the simulated multiple outlet bubblers were used to design bubblers with two outlets for an additional set of tests (9A-9C). Test 9 was also used to determine the effect of small sugar additions to the feed on ruthenium volatility. Another set of tests (10A-10D) evaluated the effects on production rate of spiking the feed with chloride and sulfate. Variables held constant to the extent possible included melt temperature, plenum temperature, cold cap coverage, the waste simulant composition, and the target glass composition. The feed rate was increased to the point that a constant, essentially complete, cold cap was achieved

Procalcitonin Testing to Guide Antibiotic Therapy in Acute Upper and Lower Respiratory Tract Infections.

Science.gov (United States)

Schuetz, Philipp; Wirz, Yannick; Mueller, Beat

2018-03-06

Is the use of procalcitonin for guiding antibiotic decisions in patients with acute upper and lower respiratory tract infections associated with improved clinical outcomes compared with usual care? Among patients with varying types and severity of acute respiratory infection, using procalcitonin to guide decisions about antibiotics is associated with lower rates of antibiotic exposure, antibiotic-related adverse effects, and mortality.

Current position on severe accident phenomena

Energy Technology Data Exchange (ETDEWEB)

Henry, Robert E [Fauske and Associates, Inc., Burr Ridge, IL (United States)

2004-07-01

The phenomena addressed in this lecture are: in-vessel and ex-vessel hydrogen generation; in-vessel and in-containment natural circulation, steam explosions, direct containment heating, core-concrete interaction; debris coolability, containment strength/failure. The following events were modeled: axial and radial power distribution, two-phase level in the core, steam generation in covered section, decay heat generation, convection to gas, cladding oxidation, cold ballooning and rupture, natural circulation between the core and upper plenum, hydrogen generation, core meltdown, reflooding. Differences between PWR and BWR type reactors.

Current position on severe accident phenomena

International Nuclear Information System (INIS)

Henry, Robert E.

2004-01-01

The phenomena addressed in this lecture are: in-vessel and ex-vessel hydrogen generation; in-vessel and in-containment natural circulation, steam explosions, direct containment heating, core-concrete interaction; debris coolability, containment strength/failure. The following events were modeled: axial and radial power distribution, two-phase level in the core, steam generation in covered section, decay heat generation, convection to gas, cladding oxidation, cold ballooning and rupture, natural circulation between the core and upper plenum, hydrogen generation, core meltdown, reflooding. Differences between PWR and BWR type reactors

The sensitivity and the specifity of rapid antigen test in streptococcal upper respiratory tract infections.

Science.gov (United States)

Gurol, Yesim; Akan, Hulya; Izbirak, Guldal; Tekkanat, Zuhal Tazegun; Gunduz, Tehlile Silem; Hayran, Osman; Yilmaz, Gulden

2010-06-01

It is aimed to detect the sensitivity and specificity of rapid antigen detection of group A beta hemolytic streptococci from throat specimen compared with throat culture. The other goal of the study is to help in giving clinical decisions in upper respiratory tract infections according to the age group, by detection of sensitivity and positive predictive values of the rapid tests and throat cultures. Rapid antigen detection and throat culture results for group A beta hemolytic streptococci from outpatients attending to our university hospital between the first of November 2005 and 31st of December 2008 were evaluated retrospectively. Throat samples were obtained by swabs from the throat and transported in the Stuart medium and Quickvue Strep A [Quidel, San Diego, USA] cassette test was applied and for culture, specimen was inoculated on 5% blood sheep agar and identified according to bacitracin and trimethoprim-sulphametaxazole susceptibility from beta hemolytic colonies. During the dates between the first of November 2005 and 31st of December 2008, from 453 patients both rapid antigen detection and throat culture were evaluated. Rapid antigen detection sensitivity and specificity were found to be 64.6% and 96.79%, respectively. The positive predictive value was 80.95% whereas negative predictive value was 92.82%. Kappa index was 0.91. When the results were evaluated according to the age groups, the sensitivity and the positive predictive value of rapid antigen detection in children were 70%, 90.3% and in adults 59.4%, 70.4%. When bacterial infection is concerned to prevent unnecessary antibiotic use, rapid streptococcal antigen test (RSAT) is a reliable method to begin immediate treatment. To get the maximum sensitivity of RSAT, the specimen collection technique used and education of the health care workers is important. While giving clinical decision, it must be taken into consideration that the sensitivity and the positive predictive value of the RSAT is quite

Radiation dose estimates for oral agents used in upper gastrointestinal disease

International Nuclear Information System (INIS)

Siegel, J.A.; Wu, R.K.; Knight, L.C.; Zelac, R.E.; Stern, H.S.; Malmud, L.S.

1983-01-01

Radiation dosimetry was calculated for a number of orally administered radiopharmaceuticals used for study of upper gastrointestinal function. These include: Tc-99m sulfur colloid in water, in a cooked egg, and in chicken liver labeled in vivo; In-111 DTPA; Tc-99m DTPA; In-113m DTPA; Tc-99m ovalbumin in cooked egg; and In-111 colloid in chicken liver labeled in vivo. Radiation burdens to the stomach, small intestine, upper and lower large intestine, ovaries, testes, and total body are calculated for each preparation

A study on self-excited sloshing due to the fluid discharge over a flexible weir

International Nuclear Information System (INIS)

Nagakura, Hiroshi; Kaneko, Shigehiko.

1995-01-01

An analytical model for the fluid-elastic instability as observed in Super-Phenix-1 LMFBR is proposed. This fluid-structure system is constituted by the flexible weir and adjoining fluid plenums, and the fluid is discharged from the upstream plenum to the downstream plenum over a flexible weir. The characteristic equation of the system is derived for the case in which the weir vibrates at the frequency of the downstream plenum sloshing. The effects of the fluid level difference between the upstream and the downstream plenum and weir rigidity are examined, and the mechanism for instability is discussed. (author)

FBR type reactors

International Nuclear Information System (INIS)

Maemoto, Junko.

1985-01-01

Purpose: To moderate abrupt temperature change near the inner walls of a suspended body thereby prevent thermal shocks and thermal deformations to structural materials. Constitution: High temperature coolants during ordinary operation of the nuclear reactor flow from the reactor core through the flow holes of the suspended body and from the upper plenum into an intermediate heat exchanger. The temperature of the coolants is lowered with heat exchanging effect with secondary coolants in the heat exchange and the coolants are then flow through the lower plenum into the reactor core and heated again. Upon generation of reactor scram, the temperature of the coolants at the exit of the reactor core is reduced abruptly and the flow rate is lowered due to the pump coast down. However, mixing of the coolants in the suspended body is accelerated by the coolants at high temperature flowing out of the flow holes and the coolants at the low temperature flowing from the flow hole group, to reduce the temperature difference and moderate the stratification flow forming an abrupt temperature slope. (Yoshihara, H.)

Systems and methods for preventing flashback in a combustor assembly

Science.gov (United States)

Johnson, Thomas Edward; Ziminsky, Willy Steve; Stevenson, Christian Xavier

2016-04-05

Embodiments of the present application include a combustor assembly. The combustor assembly may include a combustion chamber, a first plenum, a second plenum, and one or more elongate air/fuel premixing injection tubes. Each of the elongate air/fuel premixing injection tubes may include a first length at least partially disposed within the first plenum and configured to receive a first fluid from the first plenum. Moreover, each of the elongate air/fuel premixing injection tubes may include a second length disposed downstream of the first length and at least partially disposed within the second plenum. The second length may be formed of a porous wall configured to allow a second fluid from the second plenum to enter the second length and create a boundary layer about the porous wall.

Integrated System Test Approaches for the NASA Ares I Crew Launch Vehicle

Science.gov (United States)

Cockrell, Charles

2008-01-01

NASA is maturing test and evaluation plans leading to flight readiness of the Ares I crew launch vehicle. Key development, qualification, and verification tests are planned . Upper stage engine sea-level and altitude testing. First stage development and qualification motors. Upper stage structural and thermal development and qualification test articles. Main Propulsion Test Article (MPTA). Upper stage green run testing. Integrated Vehicle Ground Vibration Testing (IVGVT). Aerodynamic characterization testing. Test and evaluation supports initial validation flights (Ares I-Y and Orion 1) and design certification.

Experimental evaluation of wall Mach number distributions of the octagonal test section proposed for NASA Lewis Research Center's altitude wind tunnel

Science.gov (United States)

Harrington, Douglas E.; Burley, Richard R.; Corban, Robert R.

1986-01-01

Wall Mach number distributions were determined over a range of test-section free-stream Mach numbers from 0.2 to 0.92. The test section was slotted and had a nominal porosity of 11 percent. Reentry flaps located at the test-section exit were varied from 0 (fully closed) to 9 (fully open) degrees. Flow was bled through the test-section slots by means of a plenum evacuation system (PES) and varied from 0 to 3 percent of tunnel flow. Variations in reentry flap angle or PES flow rate had little or no effect on the Mach number distributions in the first 70 percent of the test section. However, in the aft region of the test section, flap angle and PES flow rate had a major impact on the Mach number distributions. Optimum PES flow rates were nominally 2 to 2.5 percent wtih the flaps fully closed and less than 1 percent when the flaps were fully open. The standard deviation of the test-section wall Mach numbers at the optimum PES flow rates was 0.003 or less.

The Benefits of Departmentalization in Upper Elementary Grades for Students and Teachers

Science.gov (United States)

Johnson, Malissa Lee

2013-01-01

This study addressed the benefits of departmentalization in upper elementary grades for students and teachers. The variables of gender and classroom structure (departmentalized versus self-contained) were considered for student participants (n = 125). Results for students were evaluated on pre-test and post-test data using the following measures:…

Development of a qualified nodalization for small-break LOCA transient analysis in PSB-VVER integral test facility by RELAP5 system code

Energy Technology Data Exchange (ETDEWEB)

Shahedi, S. [Department of Energy Engineering, Sharif University of Technology, Azadi Street, Tehran (Iran, Islamic Republic of); Jafari, J., E-mail: jalil_jafari@yahoo.co [Reactors and Accelerators R and D School, Nuclear Science and Technology Research Institute, North Kargar Street, Tehran (Iran, Islamic Republic of); Boroushaki, M. [Department of Energy Engineering, Sharif University of Technology, Azadi Street, Tehran (Iran, Islamic Republic of); D' Auria, F. [DIMNP, University of Pisa, Via Diotisalvi 2, 56126 Pisa (Italy)

2010-10-15

This paper deals with development and qualification of a nodalization for modeling of the PSB-VVER integral test facility (ITF) by RELAP5/MOD3.2 code and prediction of its primary and secondary systems behaviors at steady state and transient conditions. The PSB-VVER is a full-height, 1/300 volume and power scale representation of a VVER-1000 NPP. A RELAP5 nodalization has been developed for PSB-VVER modeling and a nodalization qualification process has been applied for the developed nodalization at steady state and transient levels and a qualified nodalization has been proposed for modeling of the PSB ITF. The 11% small-break loss-of-coolant-accident (SBLOCA), i.e. rupture of one of the hydroaccumulators (HA) injection lines in the upper plenum (UP) region of reactor pressure vessel (RPV) below the hot legs (HL), inlets has been considered for nodalization qualification process. The influence of the different steam generator (SG) nodalizations on the RELAP5 results and on the nodalization qualification process has been examined. The 'steady state' qualification level includes checking the correctness of the initial and boundary conditions and geometrical fidelity. In the 'transient' qualification level, the time dependent results of the code calculation are compared with the experimental time trends from both the qualitative and quantitative point of view. For quantitative assessment of the results, a Fast Fourier Transform Based Method (FFTBM) has been used. The FFTBM was used to establish a range in which the steam generators nodalizations can vary.

Cellular Stress Response Gene Expression During Upper and Lower Body High Intensity Exercises.

Science.gov (United States)

Kochanowicz, Andrzej; Sawczyn, Stanisław; Niespodziński, Bartłomiej; Mieszkowski, Jan; Kochanowicz, Kazimierz; Żychowska, Małgorzata

2017-01-01

The aim was to compare the effect of upper and lower body high-intensity exercise on chosen genes expression in athletes and non-athletes. Fourteen elite male artistic gymnasts (EAG) aged 20.6 ± 3.3 years and 14 physically active men (PAM) aged 19.9 ± 1.0 years performed lower and upper body 30 s Wingate Tests. Blood samples were collected before, 5 and 30 minutes after each effort to assess gene expression via PCR. Significantly higher mechanical parameters after lower body exercise was observed in both groups, for relative power (8.7 ± 1.2 W/kg in gymnasts, 7.2 ± 1.2 W/kg in controls, p = 0.01) and mean power (6.7 ± 0.7 W/kg in gymnasts, 5.4 ± 0.8 W/kg in controls, p = 0.01). No differences in lower versus upper body gene expression were detected for all tested genes as well as between gymnasts and physical active man. For IL-6 m-RNA time-dependent effect was observed. Because of no significant differences in expression of genes associated with cellular stress response the similar adaptive effect to exercise may be obtained so by lower and upper body exercise.

S-22: Upper Extremity Plyometric Training for the Pediatric Overhead Athletes; Randomized Controled Trial

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Elif Turgut

2017-03-01

Full Text Available INTRODUCTION/ PURPOSE: The purpose of the current study was to investigate the effects of a 12-week upper extremity plyometric training program on upper body explosive power, strength and endurance in pediatric overhead athletes.MATERIALS-METHOD: Twenty-eight female pediatric volleyball players participated in the study. The participants were randomly divided into two study groups: an intervention group (upper extremity plyometric training, n = 14 and a control group (n = 14. All of the participants were assessed before and after a 12-week training program for upper body explosive power, strength and endurance. Statistical comparison was performed using an analysis of variance test. FINDINGS: Comparisons showed that after a 12-week training program, the upper body plyometric training program resulted in more improvements in an overhead medicine-ball throwing distance and a push-up performance when compared to control training. DISCUSSION / CONCLUSION: Compared to regular training, upper body plyometric training resulted in additional improvements in upper body power and strength and endurance among female pediatric volleyball players. The findings of the study provide a basis for developing training protocols for pediatric volleyball players.

Advanced upper limb prosthetic devices: implications for upper limb prosthetic rehabilitation.

Science.gov (United States)

Resnik, Linda; Meucci, Marissa R; Lieberman-Klinger, Shana; Fantini, Christopher; Kelty, Debra L; Disla, Roxanne; Sasson, Nicole

2012-04-01

The number of catastrophic injuries caused by improvised explosive devices in the Afghanistan and Iraq Wars has increased public, legislative, and research attention to upper limb amputation. The Department of Veterans Affairs (VA) has partnered with the Defense Advanced Research Projects Agency and DEKA Integrated Solutions to optimize the function of an advanced prosthetic arm system that will enable greater independence and function. In this special communication, we examine current practices in prosthetic rehabilitation including trends in adoption and use of prosthetic devices, financial considerations, and the role of rehabilitation team members in light of our experiences with a prototype advanced upper limb prosthesis during a VA study to optimize the device. We discuss key challenges in the adoption of advanced prosthetic technology and make recommendations for service provision and use of advanced upper limb prosthetics. Rates of prosthetic rejection are high among upper limb amputees. However, these rates may be reduced with sufficient training by a highly specialized, multidisciplinary team of clinicians, and a focus on patient education and empowerment throughout the rehabilitation process. There are significant challenges emerging that are unique to implementing the use of advanced upper limb prosthetic technology, and a lack of evidence to establish clinical guidelines regarding prosthetic prescription and treatment. Finally, we make recommendations for future research to aid in the identification of best practices and development of policy decisions regarding insurance coverage of prosthetic rehabilitation. Copyright © 2012 American Congress of Rehabilitation Medicine. Published by Elsevier Inc. All rights reserved.

The Effect of Upper Body Mass and Initial Knee Flexion on the Injury Outcome of Post Mortem Human Subject Pedestrian Isolated Legs.

Science.gov (United States)

Petit, Philippe; Trosseille, Xavier; Dufaure, Nicolas; Dubois, Denis; Potier, Pascal; Vallancien, Guy

2014-11-01

In the ECE 127 Regulation on pedestrian leg protection, as well as in the Euro NCAP test protocol, a legform impactor hits the vehicle at the speed of 40 kph. In these tests, the knee is fully extended and the leg is not coupled to the upper body. However, the typical configuration of a pedestrian impact differs since the knee is flexed during most of the gait cycle and the hip joint applies an unknown force to the femur. This study aimed at investigating the influence of the inertia of the upper body (modelled using an upper body mass fixed at the proximal end of the femur) and the initial knee flexion angle on the lower limb injury outcome. In total, 18 tests were conducted on 18 legs from 9 Post Mortem Human Subjects (PMHS). The principle of these tests was to impact the leg at 40 kph using a sled equipped with 3 crushing steel tubes, the stiffness of which were representative of the front face of a European sedan (bonnet leading edge, bumper and spoiler). The mass of the equipped sled was 74.5 kg. The test matrix was designed to perform 4 tests in 4 configurations combining two upper body masses (either 0 or 3 kg) and two knee angles (0 or 20 degrees) at 40 kph (11 m/s) plus 2 tests at 9 m/s. Autopsies were performed on the lower limbs and an injury assessment was established. The findings of this study were first that the increase of the upper body mass resulted in more severe injuries, second that an initial flexion of the knee, corresponding to its natural position during the gait cycle, decreased the severity of the injuries, and third that based on the injury outcome, a test conducted with no upper body mass and the knee fully extended was as severe as a test conducted with a 3 kg upper body mass and an initial knee flexion of 20°.

Upper gastrointestinal bleeding.

Science.gov (United States)

Feinman, Marcie; Haut, Elliott R

2014-02-01

Upper gastrointestinal (GI) bleeding remains a commonly encountered diagnosis for acute care surgeons. Initial stabilization and resuscitation of patients is imperative. Stable patients can have initiation of medical therapy and localization of the bleeding, whereas persistently unstable patients require emergent endoscopic or operative intervention. Minimally invasive techniques have surpassed surgery as the treatment of choice for most upper GI bleeding. Copyright © 2014 Elsevier Inc. All rights reserved.

Rethinking Functional Outcome Measures: The Development of a Novel Upper Limb Token Transfer Test to Assess Basal Ganglia Dysfunction

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Susanne P. Clinch

2018-05-01

Full Text Available The basal ganglia are implicated in a wide range of motor, cognitive and behavioral activities required for normal function. This region is predominantly affected in Huntington's disease (HD, meaning that functional ability progressively worsens. However, functional outcome measures for HD, particularly those for the upper limb, are limited meaning there is an imperative for well-defined, quantitative measures. Here we describe the development and evaluation of the Moneybox test (MBT. This novel, functional upper limb assessment was developed in accordance with translational neuroscience and physiological principles for people with a broad disease manifestation, such as HD. Participants with HD (n = 64 and healthy controls (n = 21 performed the MBT, which required subjects to transfer tokens into a container in order of size (Baseline Transfer, value (Complex Transfer with and without reciting the alphabet (Dual Transfer. Disease specific measures of motor, cognition, behavior, and function were collected. HD patients were grouped into disease stage, from which, discriminative and convergent validity was assessed using Analysis of Variance and Pearson's correlation respectively. Manifest HD participants were slower than pre-manifest and control participants, and achieved significantly lower MBT total scores. Performance in the Complex Transfer and Dual Transfer tasks were significantly different between pre-manifest and stage 1 HD. All MBT performance variables significantly correlated with routinely used measures of motor, cognition, behavior, and function. The MBT provides a valid, sensitive, and affordable functional outcome measure. Unlike current assessments, MBT performance significantly distinguished the subtle differences between the earliest disease stages of HD, which are the populations typically targeted in clinical trials.

Comparison the Serum STREM1 Levels Between Children with Upper and Lower UTI.

Science.gov (United States)

Ehsanipour, Fahime; Noorbakhsh, Samileh; Zarabi, Vida; Movahedi, Zahra; Rahimzadeh, Nahid

2017-01-01

Pyelonephritis is the most common and important infection among Iranian pediatric population. Differentiation between upper and lower Urinary Tract Infection (UTI) is often difficult based on clinical data. Therefore, definite diagnosis is helpful for choosing appropriate antibiotic and decision for hospital admission. The main purpose of this study was todetermine the diagnostic value of serum STREM-1 level in children suspicious to UTI and differentiation of upper UTI and lower UTI. This prospective cross sectional study (2010-2011) was performed to evaluate and compare the serum level of STREM- 1 (pg. /ml) in 36 diagnosed UTI patients (24 upper and 12 lower UTI) with 25 normal children (without UTI) in Rasoul Akram hospital, Tehran, Iran. The mean age of studied children was 3.64 years; 24 male and 37 female. Urinary analysis and urine culture were performed for all UTI cases and only the positive cultured cases with the same microorganism were enrolled in the study. Distinguishing the upper from lower UTI was done on the basis of clinical manifestation and laboratory tests and confirmed by Imaging studies (ultra sonography /or DMSA scan). Blood sampling was taken from all children and centrifuged .The level of STREM-1 (pg /ml) in all sera was determined by Enzyme immunoassay technique (Human TREM-1 immunoassay Sandwich test, Quantikine, R&D systems, Minneapolis; USA). Cut-off levels for STREM-1 were illustrated by ROC curve. The pUTI (427.72pg/ml) and controls (124.24 pg. /ml; P =0.000) ; with cutoff point 111.5 pg./ml ; it had 83.3% sensitivity; and 60 % specificity to distinguish UTI from control. Serum STREM -1 level had no significantly difference between the upper and lower UTI (500pg/ml vs. 283 pg. /ml, P value=0.1) with cutoff point 132 pg./ml it had 83.3% sensitivity ; and 60 % specificity to distinguish upper UTI from lower UTI. Our study demonstrates that even low amount of serum STREM-1 (111.5 pg./ml) has 83.3% sensitivity ; and 60 % specificity to

Crack-arrest behavior in SEN wide plates of low-upper-shelf base metal tested under nonisothermal conditions: WP-2 series

International Nuclear Information System (INIS)

Naus, D.J.; Keeney-Walker, J.; Bass, B.R.; Robinson, G.C. Jr.; Iskander, S.K.; Alexander, D.J.; Fields, R.J.; deWit, R.; Low, S.R.; Schwartz, C.W.

1990-08-01

The Heavy-Section Steel Technology (HSST) Program at the Oak Ridge National Laboratory under the sponsorship of the Nuclear Regulatory Commission is conducting analytical and experimental studies aimed at understanding the circumstances that would initiate the growth of an existing crack in a reactor pressure vessel (RPV) and the conditions leading to arrest of a propagating crack. Objectives of these studies are to determine (1) if the material will exhibit crack-arrest behavior when the driving force on a crack exceeds the ASME limit, (2) the relationship between K Ia and temperature, and (3) the interaction of fracture modes (arrest, stable crack growth, unstable crack growth, and tensile instability) when arrest occurs at high temperatures. In meeting these objectives, crack-arrest data are being developed over an expanded temperature range through tests involving large thermally shocked cylinders, pressurized thermally shocked vessels, and wide-plate specimens. The wide-plate specimens provide the opportunity for a significant number of data points to be obtained at relatively affordable costs. These tests are designed to provide fracture-toughness measurements approaching or above the onset of the Charpy upper-shelf regime in a rising toughness region and with an increasing driving force. This document discusses test methodology and results. 23 refs., 92 figs., 25 tabs

The application of accelerometers to measure movements of upper limbs: Pilot study

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Patrik Kutilek

2017-03-01

Full Text Available Background: Even though inertial measurement units (IMU are already being used experimentally for evaluating movements of segment of the axial skeleton, no studies have been found which have used IMUs to measure the behavior of the segments of upper limbs during quiet stance. Objective: The objective is to design a suitable application of IMUs to measure movements of the upper extremities in Romberg's test and analyze spontaneous arm movements. Second aim is to identify possible discrepancies between the dominant and non-dominant arm movements. Methods: The dominant and non-dominant upper limb of each participant was identified. Then, the movements of both upper limbs were measured by the Xsens system equipped with MTx motion trackers during the quiet stance on a firm surface with eyes open (EO and eyes closed (EC. The measured data was used to calculate the medians and maximums of the superior-inferior, medio-lateral and anterior-posterior acceleration. Also, tremor intensity was calculated to quantitatively evaluate the measured data. Results: The comparison of values of maximal accelerations of the dominant and non-dominant arms showed significant difference between the arms during EC conditions. The comparison of values of median accelerations of the dominant and non-dominant arms showed significant differences between the acceleration of arms in medio-lateral direction during EO and EC conditions. In all cases, values of maximal and median accelerations and values of tremor intensity of the dominant limb strongly correlated with values on the non-dominant limb. Conclusions: Findings suggest possible usefulness of the designed application of IMUs and evaluation methods for their use in Romberg's test in clinical practice for evaluation of upper limb movements.

Evaluation report on SCTF Core-III tests S3-14, S3-15 and S3-16

International Nuclear Information System (INIS)

Iwamura, Takamichi; Iguchi, Tadashi; Akimoto, Hajime; Okubo, Tsutomu; Ohnuki, Akira; Sakaki, Isao; Adachi, Hiromichi; Murao, Yoshio

1988-03-01

It was revealed from previous reflood tests using Slab Core Test Facility (SCTF) that the heat transfer was enhanced in high power bundles and degraded in low power bundles due to the radial power distribution. In the present study, the effects of local power ratio itself and the shape of radial power distribution were separately investigated by comparing three tests: S3-14 (flat power distribution), S3-15 (inclined power distribution) and S3-16 (steep power distribution). Those three tests were performed under the same boundary conditions, total power and initial stored energy. The emergency core cooling (ECC) water was injected into the lower plenum. The test results indicated that the heat transfer enhancement and degradation was governed mainly by the bundlewise radial power ratio and less dependent on the shape of radial power distribution when the difference in power ratios between adjacent bundles is less than 0.44. The degree of heat transfer enhancement at high power bundles was increased with the radial peak power ratio. Even at an average power bundle, the heat transfer was enhanced in the non-uniform radial power distribution tests. (author)

Individualized 3D printing navigation template for pedicle screw fixation in upper cervical spine.

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Fei Guo

Full Text Available Pedicle screw fixation in the upper cervical spine is a difficult and high-risk procedure. The screw is difficult to place rapidly and accurately, and can lead to serious injury of spinal cord or vertebral artery. The aim of this study was to design an individualized 3D printing navigation template for pedicle screw fixation in the upper cervical spine.Using CT thin slices data, we employed computer software to design the navigation template for pedicle screw fixation in the upper cervical spine (atlas and axis. The upper cervical spine models and navigation templates were produced by 3D printer with equal proportion, two sets for each case. In one set (Test group, pedicle screws fixation were guided by the navigation template; in the second set (Control group, the screws were fixed under fluoroscopy. According to the degree of pedicle cortex perforation and whether the screw needed to be refitted, the fixation effects were divided into 3 types: Type I, screw is fully located within the vertebral pedicle; Type II, degree of pedicle cortex perforation is 1 mm or with the poor internal fixation stability and in need of renovation. Type I and Type II were acceptable placements; Type III placements were unacceptable.A total of 19 upper cervical spine and 19 navigation templates were printed, and 37 pedicle screws were fixed in each group. Type I screw-placements in the test group totaled 32; Type II totaled 3; and Type III totaled 2; with an acceptable rate of 94.60%. Type I screw placements in the control group totaled 23; Type II totaled 3; and Type III totaled 11, with an acceptable rate of 70.27%. The acceptability rate in test group was higher than the rate in control group. The operation time and fluoroscopic frequency for each screw were decreased, compared with control group.The individualized 3D printing navigation template for pedicle screw fixation is easy and safe, with a high success rate in the upper cervical spine surgery.

Fuel injection assembly for use in turbine engines and method of assembling same

Science.gov (United States)

Berry, Jonathan Dwight; Johnson, Thomas Edward; York, William David; Uhm, Jong Ho

2015-12-15

A fuel injection assembly for use in a turbine engine is provided. The fuel injection assembly includes an end cover, an endcap assembly, a fluid supply chamber, and a plurality of tube assemblies positioned at the endcap assembly. Each of the tube assemblies includes housing having a fuel plenum and a cooling fluid plenum. The cooling fluid plenum is positioned downstream from the fuel plenum and separated from the fuel plenum by an intermediate wall. The plurality of tube assemblies also include a plurality of tubes that extends through the housing. Each of the plurality of tubes is coupled in flow communication with the fluid supply chamber and a combustion chamber positioned downstream from the tube assembly. The plurality of tube assemblies further includes an aft plate at a downstream end of the cooling fluid plenum. The plate includes at least one aperture.

Tracking upper limbs fatigue by means of electronic dynamometry

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Fernando Max Lima

2015-06-01

Full Text Available This study aimed to identify useful electronic grip dynamometry parameters to track differences between trained (TR and untrained (UT participants, and between dominant (DO and non-dominant (ND limbs as a consequence of upper limbs muscle fatigue following 10 RM tests of the brachial biceps. This experimental study with transversal design involved 18 young adult males, of whom 9 were untrained and 9 were experienced in resistance training.Isometric grip force was evaluated (30 seconds long previous and after 10RM tests by means of a G200 Model grip dynamometer with precision load cell (Biometrics(r. Significant differences between initial and final measurements were found only for trained participants: Peak force for TR-DO (67.1 vs 55.5 kgf, p = .0277; Raw average for TR-DO (46.96 vs 42.22 kgf, p = .0464, and for TR-ND (40.34 vs 36.13 kgf, p = .0277. Electronic grip dynamometry efficiently identified upper limbs fatigue in trained participants, being raw average measurements the best parameter.

Upper and lower bounds for disadvantage factors as a test of an algorithm used in a synthesis method

International Nuclear Information System (INIS)

Ackroyd, R.T.; Nanneh, M.M.

1988-01-01

A lower bound for the disadvantage factor of a lattice cell of arbitrary configuration is obtained using a finite element method which is based on a variational principle for the even-parity angular flux. An upper bound for the disadvantage factor is given by a finite element method using the complementary variational principle for the odd-parity angular flux. These theoretical results are illustrated by calculations for urnaium/graphite and uranium/water lattices. As the approximations are refined the fluxes obtained by the first method tend towards the actual flux from below in the moderator, and from above in the fuel. These trends are reversed for the second method. This derivation of benchmarks for disadvantage factors has been undertaken primarily as a test of an important algorithm used by the authors in a method of synthesising transport solutions starting with a diffusion theory approximation. The algorithm is used to convert odd-parity approximations for the angular flux into even-parity approximations and vice versa. (author)

Upper and lower bounds for disadvantage factors as a test of algorithm used in a synthesis method

International Nuclear Information System (INIS)

Nanneh, M.M.; Ackroyd, R.T.

1991-01-01

A lower bound for the disadvantage factor of a lattice cell of arbitrary configuration is obtained using a finite element method which is based on a variational principle for the even-parity angular flux. An upper bound for the disadvantage factor is given by a finite element method using the complementary variational principle for the odd-parity angular flux. These theoretical results are illustrated by calculations for uranium/graphite and uranium/water lattices. As the approximations are refined the fluxes obtained by the first method tend towards the actual flux from below in the moderator, and from above in the fuel. These trends are reversed for the second method. This derivation of benchmarks for disadvantage factors has been undertaken primarily as a test of an important algorithm used by the authors in a method of synthesising transport solutions starting with a diffusion theory approximation. The algorithm is used to convert odd-parity approximations for the angular flux into even-parity approximations and vice versa. (author). 15 refs., 8 tabs., 9 figs

Upper thoracic-spine disc degeneration in patients with cervical pain.

Science.gov (United States)

Arana, Estanislao; Martí-Bonmatí, Luis; Mollá, Enrique; Costa, Salvador

2004-01-01

To study the relationship of upper thoracic spine degenerative disc contour changes on MR imaging in patients with neck pain. The relation between upper thoracic and cervical spine degenerative disc disease is not well established. One hundred and fifty-six patients referred with cervical pain were studied. There were 73 women and 77 men with a mean age of 48.6 +/- 14.6 years (range, 19 to 83 years). All MR studies were performed with a large 23-cm FOV covering at least from the body of T4 to the clivus. Discs were coded as normal, protrusion/bulge or extrusion. Degenerative thoracic disc contour changes were observed in 13.4% of patients with cervical pain. T2-3 was the most commonly affected level of the upper thoracic spine, with 15 bulge/protrusions and one extrusion. Upper degenerative thoracic disc contour changes presented in older patients than the cervical levels (Student-Newman-Keuls test, P < 0.001). Degenerative disc contour changes at the C7-T1, T1-2, T2-3 and T3-4 levels were significantly correlated ( P = 0.001), but unrelated to any other disc disease, patient's gender or age. Degenerative cervical disc disease was closely related together ( P < 0.001), but not with any thoracic disc. A statistically significant relation was found within the upper thoracic discs, reflecting common pathoanatomical changes. The absence of relation to cervical segments is probably due to differences in their pathomechanisms.

Upper thoracic-spine disc degeneration in patients with cervical pain

Energy Technology Data Exchange (ETDEWEB)

Arana, Estanislao; Marti-Bonmati, Luis; Costa, Salvador [Department of Radiology, Clinica Quiron, Avda Blasco Ibanez 14, 46010, Valencia (Spain); Molla, Enrique [Department of Radiology, Clinica Quiron, Avda Blasco Ibanez 14, 46010, Valencia (Spain); Department of Morphological Sciences, University of Valencia, Valencia (Spain)

2004-01-01

To study the relationship of upper thoracic spine degenerative disc contour changes on MR imaging in patients with neck pain. The relation between upper thoracic and cervical spine degenerative disc disease is not well established. One hundred and fifty-six patients referred with cervical pain were studied. There were 73 women and 77 men with a mean age of 48.6{+-}14.6 years (range, 19 to 83 years). All MR studies were performed with a large 23-cm FOV covering at least from the body of T4 to the clivus. Discs were coded as normal, protrusion/bulge or extrusion. Degenerative thoracic disc contour changes were observed in 13.4% of patients with cervical pain. T2-3 was the most commonly affected level of the upper thoracic spine, with 15 bulge/protrusions and one extrusion. Upper degenerative thoracic disc contour changes presented in older patients than the cervical levels (Student-Newman-Keuls test, P<0.001). Degenerative disc contour changes at the C7-T1, T1-2, T2-3 and T3-4 levels were significantly correlated (P=0.001), but unrelated to any other disc disease, patient's gender or age. Degenerative cervical disc disease was closely related together (P<0.001), but not with any thoracic disc. A statistically significant relation was found within the upper thoracic discs, reflecting common pathoanatomical changes. The absence of relation to cervical segments is probably due to differences in their pathomechanisms. (orig.)

Upper thoracic-spine disc degeneration in patients with cervical pain

International Nuclear Information System (INIS)

Arana, Estanislao; Marti-Bonmati, Luis; Costa, Salvador; Molla, Enrique

2004-01-01

To study the relationship of upper thoracic spine degenerative disc contour changes on MR imaging in patients with neck pain. The relation between upper thoracic and cervical spine degenerative disc disease is not well established. One hundred and fifty-six patients referred with cervical pain were studied. There were 73 women and 77 men with a mean age of 48.6±14.6 years (range, 19 to 83 years). All MR studies were performed with a large 23-cm FOV covering at least from the body of T4 to the clivus. Discs were coded as normal, protrusion/bulge or extrusion. Degenerative thoracic disc contour changes were observed in 13.4% of patients with cervical pain. T2-3 was the most commonly affected level of the upper thoracic spine, with 15 bulge/protrusions and one extrusion. Upper degenerative thoracic disc contour changes presented in older patients than the cervical levels (Student-Newman-Keuls test, P<0.001). Degenerative disc contour changes at the C7-T1, T1-2, T2-3 and T3-4 levels were significantly correlated (P=0.001), but unrelated to any other disc disease, patient's gender or age. Degenerative cervical disc disease was closely related together (P<0.001), but not with any thoracic disc. A statistically significant relation was found within the upper thoracic discs, reflecting common pathoanatomical changes. The absence of relation to cervical segments is probably due to differences in their pathomechanisms. (orig.)

Validation of CATHARE 3D code against UPTF TRAM C3 transients

International Nuclear Information System (INIS)

Glantz, Tony; Freitas, Roberto

2007-01-01

Within the nuclear reactor safety analysis, one of the events that could potentially lead to a recriticality accident in case of a Small Break LOCA (SBLOCA) in a pressurized water reactor (PWR) is a boron dilution scenario followed by a coolant mixing transient. Some UPTF experiments can be interpreted as generic boron dilution experiments. In fact, the UPTF experiments were originally designed to conduct separate effects studies focused on multi-dimensional thermal hydraulic phenomena. But, in the case of experimental program TRAM, some studies are realized on the boron mixing: tests C3. Some of these tests have been used for the validation and assessment of the 3D module of CATHARE code. Results are very satisfying; CATHARE 3D code is able to reproduce correctly the main features of the UPTF TRAM C3 tests, the temperature mixing in the cold leg, the formation of a strong stratification in the upper downcomer, the perfect mixing temperature in the lower downcomer and the strong stratification in the lower plenum. These results are also compared with the CFX-5 and TRIO-U codes results on these tests. (author)

A Framework to Automate Assessment of Upper-Limb Motor Function Impairment: A Feasibility Study

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Paul Otten

2015-08-01

Full Text Available Standard upper-limb motor function impairment assessments, such as the Fugl-Meyer Assessment (FMA, are a critical aspect of rehabilitation after neurological disorders. These assessments typically take a long time (about 30 min for the FMA for a clinician to perform on a patient, which is a severe burden in a clinical environment. In this paper, we propose a framework for automating upper-limb motor assessments that uses low-cost sensors to collect movement data. The sensor data is then processed through a machine learning algorithm to determine a score for a patient’s upper-limb functionality. To demonstrate the feasibility of the proposed approach, we implemented a system based on the proposed framework that can automate most of the FMA. Our experiment shows that the system provides similar FMA scores to clinician scores, and reduces the time spent evaluating each patient by 82%. Moreover, the proposed framework can be used to implement customized tests or tests specified in other existing standard assessment methods.

Setup of a test facility for the ECH upper launcher in ITER; Aufbau eines Versuchsstandes fuer den ECH Upper Launcher in ITER. Schlussbericht

Energy Technology Data Exchange (ETDEWEB)

Scherer, T.; Aiello, G.; Meier, A.; Schreck, S.; Strauss, D.; Spaeh, P.; Vaccaro, A.

2015-07-01

In order to treat plasma instabilities in four of the upper ports in the ITER vacuum vessel electron-cyclotron launchers are installed. These consist essentially of a trapezoidally shaped steel construction, which contains the microwave components (essentially mirrors and waveguides). In the construction of such a launcher as essential given points the mechanical strength, the sufficient cooling of the system, effective screening of sensitive components against neutronic loads, as well as precise mountability must be regarded. The present reference design of the EC launcher is based on the so-called ''front-steering configuration'', in which the microwaves are fed through waveguides and through the diamond torus-window backward into the launcher and extend until around a third of the whole length of the launcher. From here the microwaves are further guided as quasi-optical beams via a mirror system forward to the plasma side of the launcher. There they are focused and fed via a set of adjustable mirrors to definite positions in the outerregions of the plasma. In order to assure that the microwave components are both protected against tha plasma and precisely positioned, their mounting in a stable, precise, and accessible structure is required. The extremely high requirements for such a structure, which exceed partly above hitherto typical and already industrially manufactured applications, make fabrication tests by means of certain sample objects as well as the construction of selected prototypes, which can also be used in view of their thermohydraulic suitability, indispensable.

Evaluation of HEPA filter service life

International Nuclear Information System (INIS)

Fretthold, J.K.; Stithem, A.R.

1997-01-01

Rocky Flats Environmental Technology Site (RFETS), has approximately 10,000 High Efficiency Particulate Air (HEPA) Filters installed in a variety of filter plenums. These ventilation/filtration plenum systems are used to control the release of airborne particulate contaminates to the environment during normal operations and potential accidents. This report summarizes the results of destructive and non-destructive tests on HEPA filters obtained from a wide variety of ages and service conditions. These tests were performed to determine an acceptable service life criteria for HEPA filters used at Rocky Flats Environmental Technology Site (RFETS). A total of 140 filters of various ages (1972 to 1996) and service history (new, aged unused, used) were tested. For the purpose of this report, filter age from manufacture date/initial test date to the current sample date was used, as opposed to the actual time a filter was installed in an operating system

Smartphone supported upper limb prosthesis

Directory of Open Access Journals (Sweden)

Hepp D.

2015-09-01

Full Text Available State of the art upper limb prostheses offer up to six active DoFs (degrees of freedom and are controlled using different grip patterns. This low number of DoFs combined with a machine-human-interface which does not provide control over all DoFs separately result in a lack of usability for the patient. The aim of this novel upper limb prosthesis is both offering simplified control possibilities for changing grip patterns depending on the patients’ priorities and the improvement of grasp capability. Design development followed the design process requirements given by the European Medical Device Directive 93/42 ECC and was structured into the topics mechanics, software and drive technology. First user needs were identified by literature research and by patient feedback. Consequently, concepts were evaluated against technical and usability requirements. A first evaluation prototype with one active DoF per finger was manufactured. In a second step a test setup with two active DoF per finger was designed. The prototype is connected to an Android based smartphone application. Two main grip patterns can be preselected in the software application and afterwards changed and used by the EMG signal. Three different control algorithms can be selected: “all-day”, “fine” and “tired muscle”. Further parameters can be adjusted to customize the prosthesis to the patients’ needs. First patient feedback certified the prosthesis an improved level of handling compared to the existing devices. Using the two DoF test setup, the possibilities of finger control with a neural network are evaluated at the moment. In a first user feedback test, the smartphone based software application increased the device usability, e.g. the change within preselected grip patterns and the “tired muscle” algorithm. Although the overall software application was positively rated, the handling of the prosthesis itself needs to be proven within a patient study to be

Chemistry aspects of the Falcon programme

International Nuclear Information System (INIS)

Beard, A.M.; Bennett, P.J.; Benson, C.G.; Sabathier, F.

1990-12-01

Experiments have been conducted in the Falcon facility to study the interaction of fission product vapours released from simulant and trace-irradiated fuel samples with bulk-materials aerosol such as control rod alloy and boric acid. These experiments were designed in collaboration with computer code specialists to represent, as closely as possible, the conditions pertaining within a severe reactor accident. Small segments of fuel were heated to ∼2000 K and the resulting vapour-aerosol release was studied along a pathway simulating the upper plenum, hot-leg structures and containment. (author)

Assessment of Body-Powered Upper Limb Prostheses by Able-Bodied Subjects, using the Box and Blocks Test and the Nine Hole Peg Test

NARCIS (Netherlands)

Haverkate, L.; Smit, G.; Plettenburg, D.H.

2014-01-01

Study Design: Experimental trial. Background: The functional performance of currently available body-powered prostheses is unknown. Objective: The goal of this study was to objectively assess and compare the functional performance of three commonly used body-powered upper limb terminal devices.

Analysis of a large break LOCA in the cold leg of the WWER-440/W-213 plant Griefswald, Unit 5

International Nuclear Information System (INIS)

Horche, W.

1993-01-01

The Gessellschaft fur Anlagen und Reaktorsicherheit (GRS) has performed a safety evaluation of the nuclear power plant (NPP) Greifswald, unit 5, of the Soviet type WWER-440/W-213, in cooperation with the French Institute de Protection of de Surete Nucleaire (IPSN) and other partners. Within this project an independent accident analysis is performed by GRS in order to assess the results of existing analysis and to supplement them. In this paper the analysis of the double-ended guillotine break (DEGB) of one cold leg of the main circulation pipe is described. The major objective of the calculation was the investigation of the accident sequence with reduced availability of the emergency core cooling system (single failure criterion). In addition, the simultaneous loss of onsite and offsite power and the failure of scram were assumed. The thermal-hydraulic system code ATHLET/FLUT, developed at GRS and already applied for the safety analysis of several WWER plants, was chosen again. The pressure in the confinement, the back pressure for the discharge model, was calculated as a function of time for this accident separately with GRS-Code RALOC. Furthermore, it was necessary to model the local concentration of direct accumulator injection into the reactor vessel with the help of a special two-channel model of the core and upper plenum. For this model, results were considered obtained from the 1:1 scaled test facility UPTF. It was assumed that only 25% of the upper plenum and core volume is directly penetrated by the injected water. The DEGB was defined in that loop, which is connected with one of three low-pressure injection subsystems. This means that this injected water flows towards the leak without passing the core. As single failure the failure of one of three diesel generators was assumed. The full paper will contain nodalization schemes, which are generated by the ATHLET-Input-Grafic

The Effect of Cultural Integration on the Development of Listening Comprehension among Iranian Upper-Intermediate EFL Learners

Directory of Open Access Journals (Sweden)

Mohammad Ali Fatemi

2014-12-01

Full Text Available Cultural integration can be used as an effective learning practice in contexts of English as Foreign Language (EFL classrooms. The present study aimed at investigating the effect of cultural integration on the development of Iranian EFL upper-intermediate learners' listening comprehension.  To this end, fifty-two upper-intermediate EFL learners were selected based on the Quick Placement Test, developed by Oxford University Press and University of Cambridge Local Examinations Syndicate (2012. These participants were randomly assigned into experimental (N=26 and control (N=26 groups. T-test analysis indicated significant effects of cultural integration on the development of listening comprehension on upper-intermediate EFL learners. The findings offer pedagogical implications for integrating First Language (L1 culture in EFL listening comprehension classrooms.

RANS simulation of the thermal mixing in HTTF LP during normal operation conditions – High Temperature Test Facility at Oregon State University

International Nuclear Information System (INIS)

Gradecka, Malwina J.; Woods, Brian

2014-01-01

Since High Temperature Gas-cooled Reactors are being considered as the most promising design of upcoming IV Gen reactors, key research areas were identified to address safety aspects of this design. A number of simulations and experiments need to be conducted in this field. In this paper, thermal-hydraulics aspects of coolant flow through Lower Plenum (LP) of HTGR were considered, specifically flow characteristics to identify the risk of temperature stratification in LP and hot spotting on LP floor. Local temperature gradients can cause material degradation. As the power profile is non-uniform across the core, jets of coolant exit the core region at different temperatures and enter the LP impinging on LP floor causing hot spots at LP structure and temperature stratification. To address those issues numerical simulation and an experiment are being developed. The numerical simulation provides coolant flow velocity and temperature fields. The purpose of this study is to investigate the mixing phenomenon in the LP due to risk of the hot streaking and thermal stratification phenomena during normal operation of HTTF. The following aspect are being examined: identification of gas flow behavior in lower plenum of HTTF based on CFD simulations, identification of hot streaking issue in the HTTF lower plenum using CFD tools, and computational investigation of gas mixing efficiency. This paper includes a description of experimental setup of HTTF, guidance for LP CFD modeling, and the results and analysis of CFD simulation. (author)

The use of Virtual Reality for upper limb rehabilitation of hemiparetic Stroke patients

Directory of Open Access Journals (Sweden)

Antonio Vinicius Soares

Full Text Available Introduction The Stroke is a neurologic disturbs that leads to a serious impact to the functionality and the quality of life of the survivors. It is necessary to develop new tools with rehabilitation objectives, where the Virtual Reality (VR is introduced as a useful therapeutic resource to the motor recovery, in an attractive and efficient way, restoring functions through adapted games. Objective Analyzing the therapeutic effects of the Virtual Reality (Serious Game in the recovery of the upper limb in hemiparetic Stroke patients. Methods Quasi-experimental research type time series, there are three pre and three post-tests already accomplished around 20 VR sessions. In the assessments the following measurement instruments were used: Fugl-Meyer Scale – session of the upper limb (FMS - UL; Range of Motion (ROM for flexion and abduction shoulder; Box and Block Test (BBT; Nine Holes and Peg Test (9HPT; the Nottingham Health Profile (NHP; and the Modified Ashworth Scale (MAS. Results Significant gains were observed in the FMS-UL tests, with increase of 25.6%; increase ROM of shoulder with 34.0% for abduction and 19% for flexion; BBT 25.0%; also reported improvement in quality of life by NHP; it did not occurred significant alterations for 9HPT nor in MAS. Conclusion Although the results found in this research are preliminary, they are indicative that the VR can contribute for the recovery of the upper limb in hemiparetic Stroke patients.

BEMUSE Phase III Report - Uncertainty and Sensitivity Analysis of the LOFT L2-5 Test

International Nuclear Information System (INIS)

Bazin, P.; Crecy, A. de; Glaeser, H.; Skorek, T.; Joucla, J.; Probst, P.; Chung, B.; Oh, D.Y.; Kyncl, M.; Pernica, R.; Macek, J.; Meca, R.; Macian, R.; D'Auria, F.; Petruzzi, A.; Perez, M.; Reventos, F.; Fujioka, K.

2007-02-01

This report summarises the various contributions (ten participants) for phase 3 of BEMUSE: Uncertainty and Sensitivity Analyses of the LOFT L2-5 experiment, a Large-Break Loss-of-Coolant-Accident (LB-LOCA). For this phase, precise requirements step by step were provided to the participants. Four main parts are defined, which are: 1. List and uncertainties of the input uncertain parameters. 2. Uncertainty analysis results. 3. Sensitivity analysis results. 4. Improved methods, assessment of the methods (optional). 5% and 95% percentiles have to be estimated for 6 output parameters, which are of two kinds: 1. Scalar output parameters (First Peak Cladding Temperature (PCT), Second Peak Cladding Temperature, Time of accumulator injection, Time of complete quenching); 2. Time trends output parameters (Maximum cladding temperature, Upper plenum pressure). The main lessons learnt from phase 3 of the BEMUSE programme are the following: - for uncertainty analysis, all the participants use a probabilistic method associated with the use of Wilks' formula, except for UNIPI with its CIAU method (Code with the Capability of Internal Assessment of Uncertainty). Use of both methods has been successfully mastered. - Compared with the experiment, the results of uncertainty analysis are good on the whole. For example, for the cladding temperature-type output parameters (1. PCT, 2. PCT, time of complete quenching, maximum cladding temperature), 8 participants out of 10 find upper and lower bounds which envelop the experimental data. - Sensitivity analysis has been successfully performed by all the participants using the probabilistic method. All the used influence measures include the range of variation of the input parameters. Synthesis tables of the most influential phenomena and parameters have been plotted and participants will be able to use them for the continuation of the BEMUSE programme

Spacesuit Soft Upper Torso Sizing Systems

Science.gov (United States)

Graziosi, David; Splawn, Keith

2011-01-01

The passive sizing system consists of a series of low-profile pulleys attached to the front and back of the shoulder bearings on a spacesuit soft upper torso (SUT), textile cord or stainless steel cable, and a modified commercial ratchet mechanism. The cord/cable is routed through the pulleys and attached to the ratchet mechanism mounted on the front of the spacesuit within reach of the suited subject. Upon actuating the ratchet mechanism, the shoulder bearing breadth is changed, providing variable upper torso sizing. The active system consists of a series of pressurizable nastic cells embedded into the fabric layers of a spacesuit SUT. These cells are integrated to the front and back of the SUT and are connected to an air source with a variable regulator. When inflated, the nastic cells provide a change in the overall shoulder bearing breadth of the spacesuit and thus, torso sizing. The research focused on the development of a high-performance sizing and actuation system. This technology has application as a suit-sizing mechanism to allow easier suit entry and more accurate suit fit with fewer torso sizes than the existing EMU (Extravehicular Mobility Unit) suit system. This advanced SUT will support NASA s Advanced EMU Evolutionary Concept of a two-sizes-fit-all upper torso for replacement of the current EMU hard upper torso (HUT). Both the passive and nastic sizing system approaches provide astronauts with real-time upper torso sizing, which translates into a more comfortable suit, providing enhanced fit resulting in improved crewmember performance during extravehicular activity. These systems will also benefit NASA by reducing flight logistics as well as overall suit system cost. The nastic sizing system approach provides additional structural redundancy over existing SUT designs by embedding additional coated fabric and uncoated fabric layers. Two sizing systems were selected to build into a prototype SUT: one active and one passive. From manned testing, it

Comparison of facility characteristics between SCTF Core-I and Core-II

International Nuclear Information System (INIS)

Adachi, Hiromichi; Iwamura, Takamichi; Sobajima, Makoto; Ohnuki, Akira; Abe, Yutaka; Murao, Yoshio.

1990-08-01

The Slab Core Test Facility (SCTF) was constructed to investigate two-dimensional thermal-hydraulics in the core and fluid behavior of carryover water out of the core including its feed-back effect to the core behavior mainly during the reflood phase of a large break loss-of-coolant accident (LOCA) of a pressurized water reactor (PWR). Since three simulated cores are used in the SCTF Test Program and the design of these three cores are slightly different one by one, repeatability test is required to justify a direct comparison of data obtained with different cores. In the present report, data of Test S2-13 (Run 618) obtained with SCTF Core-II were compared with those of Test S1-05 (Run 511) obtained with the Core-I, which were performed under the forced-flooding condition. Thermal-hydraulic behaviors in these two tests showed quite similar characteristics of both system behavior and two-dimensional core behaviors. Therefore, the test data obtained from the two cores can be compared directly with each other. After the turnaround of clad temperatures, however, some differences were found in upper plenum water accumulation and resultant two-dimensional core cooling behaviors such as quench front propagation from bottom to top of the core. (author)

Workability of coal seams in the Upper Silesian Coal Basin

Energy Technology Data Exchange (ETDEWEB)

Sikora, W; Fels, M; Soltysik, K

1978-04-01

This paper presents results of an investigation on workability of coal seams of stratigraphic groups from 100 to 700 in the: Upper Silesian Coal Basin. Analyzed are 2900 petrographic logs taken in the longwall workings and in narrow openings as well as about 9000 individual samples. Workability of coal seams, floors and partings is determined. Workability is described by the indicator f, (according to the Protodyakonov shatter method) and the indicator U, (compression strength of the unshaped test samples). The mean percentage content of indivi dual petrographic groups of coal as well as the mean workability indicator, f, of coals in the stratigraphic groups of coal seams in Upper Silesia are also determined.

Upper body push and pull strength ratio in recreationally active adults.

Science.gov (United States)

Negrete, Rodney J; Hanney, William J; Pabian, Patrick; Kolber, Morey J

2013-04-01

Agonist to antagonist strength data is commonly analyzed due to its association with injury and performance. The purpose of this study was to examine the agonist to antagonist ratio of upper body strength using two simple field tests (timed push up/timed modified pull up) in recreationally active adults and to establish the basis for reference standards. One hundred eighty (180) healthy recreationally active adults (111 females and 69 males, aged 18-45 years) performed two tests of upper body strength in random order: 1. Push-ups completed during 3 sets of 15 seconds with a 45 second rest period between each set and 2. Modified pull-ups completed during 3 sets of 15 seconds with a 45 second rest period between each set. The push-up to modified pull-up ratio for the males was 1.57:1, whereas females demonstrated a ratio of 2.72:1. The results suggest that for our group of healthy recreationally active subjects, the upper body "pushing" musculature is approximately 1.5-2.7 times stronger than the musculature involved for pulling. In this study, these recreationally active adults displayed greater strength during the timed push-ups than the modified pull-ups. The relationship of these imbalances to one's performance and or injury risk requires further investigation. The reference values, however, may serve the basis for future comparison and prospective investigations. The field tests in this study can be easily implemented by clinicians and an agonist/antagonist ratio can be determined and compared to our findings. 2b.

Modeling study of droplet behavior during blowdown period of large break LOCA based on experimental data

International Nuclear Information System (INIS)

Sakaba, Hiroshi; Umezawa, Shigemitsu; Teramae, Tetsuya; Furukawa, Yuji

2004-01-01

During LOCA (Loss Of Coolant Accident) in PWR, droplets behavior during blowdown period is one of the important phenomena. For example, the spattering from falling liquid film that flows from upper plenum generates those droplets in core region. The behavior of droplets in such flow has strong effect for cladding temperature behavior because these droplets are able to remove heat from a reactor core by its direct contact on fuel rods and its evaporation at the surface. For safety analysis of LOCA in PWR, it is necessary to evaluate droplet diameter precisely in order to predict fuel cladding temperature changing by the calculation code. Based on the test results, a new droplet behavior model was developed for the MCOBRA/TRC code that predicts the droplet behavior during such LOCA events. Furthermore, the verification calculations that simulated some blowdown tests were performed using by the MCOBRA/TRAC code. These results indicated the validity of this droplet model during blow down cooling period. The experiment was focused on investigating the Weber number of steady droplet in the blow down phenomenon of large break LOCA. (author)

Flow Induced Vibration Program at Argonne National Laboratory