Simulation of the QUENCH-06 experiment with MELCOR 1.8.5

International Nuclear Information System (INIS)

Stanojevic, M.; Leskovar, M.

2001-01-01

The MELCOR 1.8.5 code input model and simulation results of the OECD/NEA international standard problem No. 45 (ISP-45) are presented. ISP-45 was performed as QUENCH-06 experiment at Forschungszentrum Karlsruhe. The objectives of the QUENCH program are the analysis of the heat-up, oxidation and delayed reflood phases of a PWR type fuel rod bundle in the QUENCH facility and investigation of the thermal, mechanical, physical and chemical behavior of fuel rod claddings under transient cool-down conditions. The objectives of the OECD/NEA ISP program are the extension of the reflood database to identify key phenomena occurring during accident management measure procedures and code validation, i.e., reliability and accuracy of severe accident codes especially during the quench phase. The QUENCH test bundle is made up of 21 fuel rod simulators approximately 2.5 m long. The Zircaloy-4 rod cladding is identical to that used in pressurized water reactors with respect to material and dimensions. The bundle is heated electrically. The QUENCH-06 experiment had three phases: the pre-oxidation phase, the power transient phase and the reflood-quench phase. According to the ISP-45 specification, the MELCOR 1.8.5 simulation includes the events from the beginning of the pre-oxidation phase until the end of the reflood-quench phase and shut-off of electric power, steam and cooling water. Calculated results are discussed with respect to accuracy, plausibility and completeness. Shortcomings and limitations of the input model are described.(author)

CANFLEX fuel bundle cross-flow endurance test 2 (Test report)

Energy Technology Data Exchange (ETDEWEB)

Hong, Sung Deok; Chung, C. H.; Chang, S. K.; Kim, B. D. [Korea Atomic Energy Research Institute, Taejeon (Korea)

2000-02-01

This report describes cross-flow endurance test 2 that was conducted at the CANDU-Hot Test Loop. The test was completed on March 30, 1999 using a new CANFLEX bundle, built by KAERI. It was carried out for a total of 22 hours. After an initial period of ten hours, the test was stopped at the intervals of four hours for bundle inspection and inter-element gap measurement[7]. The test bundle end-plate to end-cap welds were inspected carefully for failure or crack propagation using liquid penetrant examination especially at the heat-affected zones. 12 refs., 4 figs., 10 tabs. (Author)

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

International Nuclear Information System (INIS)

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

2006-03-01

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

Large bundle BWR test CORA-18: Test results

International Nuclear Information System (INIS)

Hagen, S.; Hofmann, P.; Noack, V.; Sepold, L.; Schanz, G.; Schumacher, G.

1998-04-01

The CORA out-of-pile experiments are part of the international Severe Fuel Damage (SFD) Program. They were performed to provide information on the damage progression of Light Water Reactor (LWR) fuel elements in Loss-of-coolant Accidents in the temperature range 1200 C to 2400 C. CORA-18 was the large BWR bundle test corresponding to the PWR test CORA-7. It should investigate if there exists an influence of the BWR bundle size on the fuel damage behaviour. Therefore, the standard-type BWR CORA bundle with 18 fuel rod simulators was replaced by a large bundle with two additional surrounding rows of 30 rods (48 rods total). Power input and steam flow were increased proportionally to the number of fuel rod simulators to give the same initial heat-up rate of about 1 K/s as in the smaller bundles. Emphasis was put on the initial phase of the damage progression. More information on the chemical composition of initial and intermediate interaction products and their relocation behaviour should be obtained. Therefore, power and steam input were terminated after the onset of the temperature escalation. (orig.) [de

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

International Nuclear Information System (INIS)

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

2006-11-01

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

Evaluation of bundle duct interaction by out of pile compressive test of FBR bundles. FFTF type bundle

Energy Technology Data Exchange (ETDEWEB)

Tanaka, Kosuke; Yamamoto, Yuji; Nagamine, Tsuyoshi; Maeda, Koji [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center

2000-10-01

Bundle duct interaction (BDI) caused by expansion of fuel pin bundle becomes one of the main limiting factors for fuel life times. Then, it is important for the design of fast reactor fuel assembly to understand the BDI behavior in detail. In order to understand the BDI behavior, out of pile compressive tests were conducted for FFTF type bundle by use of X-ray CT equipment. In these compressive tests, two type bundles with different accuracy of initial wire position were conducted. The objective of this test is to evaluate the influence of the initial error from standard position of wire at the same axial position. The locations of the pins and the duct flats are analyzed from CT image data. Quantitative evaluation was performed at the CT image data and discussed the bundle deformation status under BDI condition. Following results are obtained. 1) The accuracy of initial wire position is strongly depends on the pin-to-duct contact behavior. In the case of bundle with large error from standard position, pin-to-duct contact is delayed. 2) The BDI mitigation of the bundle with small error from standard wire position is following: The elastic ovality is the dominant deformation in mild BDI condition, then the wire dispersion and pin dispersion are occurred in severe BDI condition. 3) The BDI mitigation of the bundle with large error from standard wire position is following: The elastic ovality and local bowing of pins with large error from standard wire position are occurred in mild BDI condition, then pin dispersion is occurred around pins with large error from standard wire position, finally wire dispersion is occurred in severe BDI condition. 4) The existence of pins with large error from standard wire position is effective to delay the pin-to-duct contact, but the existence of these pins is possible to contact of pin- to- pin. (author)

CANFLEX fuel bundle cross-flow endurance test 2 (test procedure)

Energy Technology Data Exchange (ETDEWEB)

Hong, Sung Deok; Chung, C. H.; Chang, S. K. [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-04-01

This report describes test procedure of cross-flow 2 test for CANFLEX fuel. In October 1996. a cross-flow test was successfully performed in the KAERI Hot Test Loop for four hours at a water flow rate of 31kg/s, temperature of 266 deg C and inlet pressure of 11MPa, but it is requested more extended time periods to determine a realistic operational margin for the CANFLEX bundle during abnormal refuelling operations. The test shall be conducted for twenty two hours under the reactor conditions. After an initial period of ten hours, the test shall be stopped at the intervals of four hours for bundle inspection and inspect the test bundle end-plate to end-cap welds for failure or crack propagation using liquid penetrant examination. 2 refs., 1 fig. (Author)

The QUENCH programme at Forschungszentrum Karlsruhe (FZK)

International Nuclear Information System (INIS)

Steinbrueck, M.; Schanz, G.; Sepold, L.; Stuckert, J.; Hering, W.; Homann, C.; Miassoedov, A.

2004-01-01

The QUENCH programme at FZK was launched to investigate the hydrogen source term during reflood of an overheated reactor core. It consists of large scale bundle experiments, separate-effects tests, modelling activities and application and validation of severe fuel damage (SFD) code systems. The paper describes the experimental part of the programme, namely the experimental facilities and test rigs as well as selected results obtained during the recent years. (author)

Bundle 13 position verification tool description and on-reactor use

Energy Technology Data Exchange (ETDEWEB)

Onderwater, T G [Canadian General Electric Co. Ltd., Peterborough, ON (Canada)

1997-12-31

To address the Power Pulse problem, Bruce B uses Gap: a comprehensive monitoring program by the station to maintain the gap between the fuel string and the upstream shield plug. The gap must be maintained within a band. The gap must not be so large as to allow excessive reactivity increases or cause high impact forces during reverse flow events. It should also not be so small as to cause crushed fuel during rapid, differential reactor/fuel string cool downs. Rapid cool downs are infrequent. The Bundle 13 Position Verification Tool (BPV tool) role is to independently measure the position of the upstream bundle of the fuel string. The measurements are made on-reactor, on-power and will allow verification of the Gap Management system`s calculated fuel string position. This paper reviews the reasons for developing the BPV tool. Design issues relevant to safe operation in the fuelling machine, fuel channel and fuel handling equipment are also reviewed. Tests ensuring no adverse effects on channel pressure losses are described and actual on-reactor, on-power results are discussed. (author). 4 figs.

Bundle 13 position verification tool description and on-reactor use

International Nuclear Information System (INIS)

Onderwater, T.G.

1996-01-01

To address the Power Pulse problem, Bruce B uses Gap: a comprehensive monitoring program by the station to maintain the gap between the fuel string and the upstream shield plug. The gap must be maintained within a band. The gap must not be so large as to allow excessive reactivity increases or cause high impact forces during reverse flow events. It should also not be so small as to cause crushed fuel during rapid, differential reactor/fuel string cool downs. Rapid cool downs are infrequent. The Bundle 13 Position Verification Tool (BPV tool) role is to independently measure the position of the upstream bundle of the fuel string. The measurements are made on-reactor, on-power and will allow verification of the Gap Management system's calculated fuel string position. This paper reviews the reasons for developing the BPV tool. Design issues relevant to safe operation in the fuelling machine, fuel channel and fuel handling equipment are also reviewed. Tests ensuring no adverse effects on channel pressure losses are described and actual on-reactor, on-power results are discussed. (author). 4 figs

Post test investigation of the bundle test ESBU-1

International Nuclear Information System (INIS)

Hagen, S.; Kapulla, H.; Malauschek, H.; Wallenfels, K.P.; Buescher, B.J.

1986-08-01

This KfK report describes the post test investigation of bundle experiment ESBU-1. ESBU-1 was the first of two bundle tests on the temperature escalation of Zircaloy clad fuel rods. The investigation of the temperature escalation is part of the program of out-of-pile experiments performed within the frame work of the PNS - Severe Fuel Damage program. The bundle was composed of a 3x3 fuel rod array of our fuel rod simulators (control tungsten heater, UO 2 -ring pellet and Zircaloy cladding). The length was 0.4 meter. After the test the bundle was embedded in epoxy and cut by a diamant saw. The cross sections are investigated by metallographic, SEM and EMP examinations. The results of these examinations are in good agreement with the seperate effects tests investigation of the PNS SFD-Program and inpile experiments of the Power Burst Facility. The investigations show that liquid Zircaloy dissolves UO 2 by taking away the oxygen from the oxide. Depending on the overall oxygen content the (U,Zr,O)-melt forms at refreezing a) three phases (low oxygen content): metallic α-Zry(U), a uranium-rich metallic (U,Zr)alloy, and a (U,Zr)O 2 mixed oxide, or b) two phases (high oxygen content): α-Zr(O) and the (U,Zr)O 2 mixed oxide. c) In melt regions where the local oxidation was very severe, such as in steam contact, only the (U,Zr)O 2 mixed oxide is formed already at test temperature. Also ZrO 2 formed during the initial time of the test is dissolved by the melt. (orig.) [de

Heat transfer to a dispersed two-phase flow and detailed quench front velocity research

International Nuclear Information System (INIS)

Boer, T.C. de; Molen, S.B. van der

1985-01-01

A programme to obtain a data base for 'Boildown and Reflood' computer code development and to obtain information on the influence of non-uniform temperature and/or power profile on the quench front velocity and prequench heat transfer, including unheated wall and grid effects, has been undertaken. It is in two parts. In the first (for the tube, annulus and a 4-rod bundle) an early wetting of the unheated shroud is shown. This leads to an increase in quench front velocity and in liquid transport downstream from the quench front. For the inverted annular flow regime the extended Bromley correlation gives good agreement with the experimental data. In the second part (36-rod bundle reflood test programme) the wall-temperature differences in the radial direction gives rise to heat transfer processes which are described and explained. (U.K.)

Comparison of ASTECV1.3.2 and ASTECV2 results for QUENCH 12 test

International Nuclear Information System (INIS)

Stefanova, A.

2010-01-01

This paper presents a comparison of QUENCH 12 test calculated results with ASTECv1.3R2 and ASTECv2 computer codes. The test was performed to investigate the behavior of VVER fuel assemblies. This investigation is a part of the 6th and 7th framework programs of the EC supported ISTC program. The test facility is located at Forschungszentrum in Karlsruhe. The structure of the test facility allows experimental studies under transient and accident conditions. The ASTEC1.3R2 and ASTECv2 computer codes have been used to simulate the investigated test. The base line input model for ASTEC was provided from Forschungszentrum, Karlsruhe. During the preparation of QUENCH - 12 experiment, the input deck was adapted to new initial and boundary conditions. The comparison show good agreement between measured data and ASTEC calculated results. (author)

Fast breeder fuel pin bundle tests in the KNK II-reactor

International Nuclear Information System (INIS)

Haefner, H.E.; Bojarsky, E.

1986-11-01

Three variants of ring elements with test bundles will be reported in this paper: In a first step a ring element was built with a permanently integrated test bundle (19 carbide pins of the Karlsruhe reference concept) while the proven fuel element components have been largely maintained. This irradiation will be completed in autumn 1986 after 380 full power days of operation. The central topic of this paper will be the technique of reloadable ring elements with replaceable test bundles. A first experiment, TOAST, is in preparation. For this experiment, above all the components of the fuel element head and foot had to be newly developed and tested. A special version of double-walled replaceable test bundles to be used in the TETRA temperature transient experiments will be briefly mentioned. It is envisaged in these experiments to vary in a defined manner the coolant flow at remotely assembled test bundles consisting of 19 KNK pins each having undergone a high burnup and to use a measuring and control plug placed on the test bundle so that a variety of fuel pin temperature programs can be realized. Finally, some additional aspects of bundle design will be indicated. (orig./GL) [de

Quench detection electronics testing protocol for SST-1 magnets

International Nuclear Information System (INIS)

Banaudha, Moni; Varmora, Pankaj; Parghi, Bhadresh; Prasad, Upendra

2017-01-01

Quench Detection (QD) system consisting 204 signal channels has been successfully installed and working well during plasma experiment of SST-1 Tokamak. QD system requires testing, validation and maintenance in every SST-1 campaign for better reliability and maintainability of the system. Standalone test of each channel of the system is essential for hard-ware validation. The standard Testing Protocol follow in every campaign which validate each section of QD electronics as well as voltage tap signal cables which are routed inside the cryostat and then extended outside of the SST-1 machine up-to the magnet control room. Fiber link for Quench signal transmission to the SST-1 magnet power supply is also test and validate before every plasma campaign. Precise instrument used as a dummy source of quench signal and for manual quench generation to test the each channel and Master Quench Logic. Each signal Integrated with the magnet DAQ system, signal observed at 1Hz and 50Hz configuration to validate the logging data, compare with actual and previous test data. This paper describes the testing protocol follow in every campaign to validate functionality of QD electronics, limitation of testing, test results and overall integration of the quench detection system for SST-1 magnet. (author)

Evaluation of bundle duct interaction by out-of-pile compression test of FBR fuel pin bundles

Energy Technology Data Exchange (ETDEWEB)

Tanaka, Kosuke; Yamamoto, Yuji; Nagamine, Tsuyoshi; Maeda, Koji [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center

2001-06-01

Bundle duct interaction (BDI) caused by expansion of fuel pin bundle is a main factor to limit the fuel lifetime. Therefore, it is important for the design of fast reactor fuel assembly to understand the fuel pin deformation behavior under BDI condition. In order to understand the fuel pin deformation behavior under BDI condition, out-of-pile compression tests were conducted for FBR fuel pin bundle by use of X-ray CT equipment. In these compression tests, two kinds of fuel pin bundles were conducted. One was the fuel pin bundle with the short wire-pitch and the other was the fuel pin bundle with the short wire-pitch and large diameter claddings. The general discussions were also performed based on the results of out-of-pile compression tests obtained by use of X-ray CT equipment in the previous work. Following results were obtained. 1) The occurrence of the pin-to-duct contact depends on the wire-pitch. In the fuel pin bundle with large wire-pitch, the pin-to-duct contact occurred at the early stage of BDI. The reason of this result is due to the low bowing rigidity of the fuel pins with long wire-pitch. 2) The value of the ovalation stiffness strongly depends on the geometry of cladding (diameter, thickness) and especially on wire-pitch. This result in this work revealed that the occurrence of the pin-to-duct contact depends on the value of the ovalation stiffness. 3) The occurrence of wire dispersion and dispersive displacement of pins depends on the wire-pitch strongly. In the fuel pin bundle with the long wire-pitch, the occurrence of the above-mentioned suppression mechanism to BDI is remarkable. 4) The suppression mechanism to BDI of the fuel pin bundle with the long wire-pitch is elastic oval deformation of cladding, wire dispersion and dispersive displacement of pins. On the other hand, the elastic and plastic oval deformation of cladding is the major suppression mechanism to BDI in the fuel pin bundle with the short wire-pitch. 5) The appearance of

Multi-bundle shashlik calorimeter prototypes beam-test results

International Nuclear Information System (INIS)

Badier, J.; Bloch, P.; Bityukov, S.; Bordalo, P.; Busson, P.; Charlot, C.; Dobrzynski, L.; Golutvin, I.; Guschin, E.; Issakov, V.; Ivanchenko, I.; Klimenko, V.; Marin, V.; Moissenz, P.; Obraztsov, V.; Ostankov, A.; Popov, V.; Puljak, I.; Ramos, S.; Seez, C.; Sergueev, S.; Soushkov, V.; Tanaka, R.; Varela, J.; Virdee, T.S.; Zaitchenko, A.; Zamiatin, N.

1995-01-01

The first beam-test results for two- and three-bundle shashlik tower prototypes are described. We found that the spatial resolution, the uniformity of energy response, the calorimeter reliability and hermeticity and also two showers separation are improved in multi-bundle design approach. ((orig.))

A comprehensive in-pile test of PWR fuel bundle

Energy Technology Data Exchange (ETDEWEB)

Kang Rixin; Zhang Shucheng; Chen Dianshan (Academia Sinica, Beijing (China). Inst. of Atomic Energy)

1991-02-01

An in-pile test of PWR fuel bundle has been conducted in HWRR at IAE of China. This paper describes the structure of the test bundle (3x3-2), fabrication process and quality control of the fuel rod, irradiation conditions and the main Post Irradiation Examination (PIE) results. The test fuel bundle was irradiated under the PWR operation and water chemistry conditions with an average linear power of 381 W/cm and reached an average burnup of 25010 MWd/tU of the fuel bundle. After the test, destructive and non-destructive examination of the fuel rods was conducted at hot laboratories. The fission gas release was 10.4-23%. The ridge height of cladding was 3 to 8 {mu}m. The hydrogen content of the cladding was 80 to 140 ppm. The fuel stack height was increased by 2.9 to 3.3 mm. The relative irradiation growth was about 0.11 to 0.17% of the fuel rod length. During the irradiation test, no fuel rod failure or other abnormal phenomena had been found by the on-line fuel failure monitoring system of the test loop and water sampling analysis. The structure of the test fuel assembly was left undamaged without twist and detectable deformation. (orig.).

Research reactor fuel bundle design review by means of hydrodynamic testing

International Nuclear Information System (INIS)

Pastorini, A.; Belinco, C.

1997-01-01

During the design steps of a fuel bundle for a nuclear reactor, some vibration tests are usually necessary to verify the prototype dynamical response characteristics and the structural integrity. To perform these tests, the known hydrodynamic loop facilities are used to evaluate the vibrational response of the bundle under the different flow conditions that may appear in the reactor. This paper describes the tests performed on a 19 plate fuel bundle prototype designed for a low power research reactor. The tests were done in order to know the dynamical characteristics of the plates and also of the whole bundle under different flow rate conditions. The paper includes a description of the test facilities and the results obtained during the dynamical characterization tests and some preliminary comments about the tests under flowing water are also presented. (author) [es

Quenching behaviour of hot zircaloy tube

International Nuclear Information System (INIS)

Chinchole, A.S.; Kulkarni, P.P.; Nayak, A.K.; Vijayan, P.K.

2015-01-01

The quenching process plays a very important role in case of safety of nuclear reactors. During large break Loss of Coolant Accident in a nuclear reactor, the cooling water from the system is lost. Under this condition, cold water is injected from emergency core cooling system. Quenching behaviour of such heated rod bundle is really complex. It is well known that nanofluids have better heat removal capability and high heat transfer coefficient owing to enhanced thermal properties. Alumina nano-particles result in better cooling abilities compared with the traditionally used quenching media. In this paper, the authors have carried out experiments on quenching behaviour of hot zircaloy tube with demineralized water and nanofluids. It was observed that, the tube got quenched within few seconds even with the presence of decay heat and shows slightly reduced quenching time compared with DM water. (author)

Verification of the FBR fuel bundle-duct interaction analysis code BAMBOO by the out-of-pile bundle compression test with large diameter pins

Science.gov (United States)

Uwaba, Tomoyuki; Ito, Masahiro; Nemoto, Junichi; Ichikawa, Shoichi; Katsuyama, Kozo

2014-09-01

The BAMBOO computer code was verified by results for the out-of-pile bundle compression test with large diameter pin bundle deformation under the bundle-duct interaction (BDI) condition. The pin diameters of the examined test bundles were 8.5 mm and 10.4 mm, which are targeted as preliminary fuel pin diameters for the upgraded core of the prototype fast breeder reactor (FBR) and for demonstration and commercial FBRs studied in the FaCT project. In the bundle compression test, bundle cross-sectional views were obtained from X-ray computer tomography (CT) images and local parameters of bundle deformation such as pin-to-duct and pin-to-pin clearances were measured by CT image analyses. In the verification, calculation results of bundle deformation obtained by the BAMBOO code analyses were compared with the experimental results from the CT image analyses. The comparison showed that the BAMBOO code reasonably predicts deformation of large diameter pin bundles under the BDI condition by assuming that pin bowing and cladding oval distortion are the major deformation mechanisms, the same as in the case of small diameter pin bundles. In addition, the BAMBOO analysis results confirmed that cladding oval distortion effectively suppresses BDI in large diameter pin bundles as well as in small diameter pin bundles.

A fluorescence quenching test for the detection of flavonoid transformation.

Science.gov (United States)

Schoefer, L; Braune, A; Blaut, M

2001-11-13

A novel fluorescence quenching test for the detection of flavonoid degradation by microorganisms was developed. The test is based on the ability of the flavonoids to quench the fluorescence of 1,6-diphenyl-1,3,5-hexatriene (DPH). Several members of the anthocyanidins, flavones, isoflavones, flavonols, flavanones, dihydroflavanones, chalcones, dihydrochalcones and catechins were tested with regard to their quenching properties. The anthocyanidins were the most potent quenchers of DPH fluorescence, while the flavanones, dihydroflavanones and dihydrochalcones, quenched the fluorescence only weakly. The catechins had no visible impact on DPH fluorescence. The developed test allows a quick and easy differentiation between flavonoid-degrading and flavonoid-non-degrading bacteria. The investigation of individual reactions of flavonoid transformation with the developed test system is also possible.

Quench testing of HTS sub-elements for 13 kA and 600 A leads designed to the specifications for the CERN Large Hadron Collider project

CERN Document Server

Cowey, L; Krischel, D; Bock, J J

2000-01-01

Ability to safely withstand and survive self quench conditions is an important consideration in the design and utilisation of HTS current leads. The provision of a non superconducting shunt path allows current to be diverted in the event of a transition to the normal state. This shunt should allow very rapid transfer of current out of the HTS material and be able to safely support the full load current for the time required to detect the fault and reduce the current to zero. However, the shunt should also be designed to minimise the increased heat load which will result from it's addition to the lead. Test of leads based on melt cast BSCCO 2212 utilising a fully integrated silver gold alloy sheath are described. The HTS sub- elements form part of a full 13 kA lead, designed to the specifications of CERN for the LHC project. The sub-elements proved able to fully comply with and exceed the quench performance required by CERN. The HTS module was quenched at the full design current and continued to maintain this ...

Successful magnet quench test for CAST.

CERN Multimedia

Brice Maximilien

2002-01-01

The CERN Axion Solar Telescope (CAST) consists of a prototype LHC dipole magnet with photon detectors at each end. It searches for very weakly interacting neutral particles called axions, which should originate in the core of the Sun. The telescope, located at Point 8, can move vertically within its wheeled platform, which travels horizontally along tracks in the floor. In this way, the telescope can view the Sun at sunrise through one end and at sunset through the other end. It has been cooled down to below 1.8 K and reached ~95% of its final magnetic field of 9 tesla before a quench was induced to test the whole cryogenic system under such conditions. The cryogenic system responded as expected to the magnet quench and CAST is now ready to start its three-year search for solar axions. Photos 01 & 02 : Members of the LHC cryogenics team pose in front of the axion telescope on the day of the first quench test, together with some of the CAST collaboration.

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

International Nuclear Information System (INIS)

Felde, D.K.

1982-04-01

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

Safety analysis report of the irradiation test of Type-B bundle

Energy Technology Data Exchange (ETDEWEB)

Lee, Choong Sung; Lim, I. C.; Lee, B. C.; Ryu, J. S.; Kim, H. R

2000-06-01

The HANARO fuel, U{sub 3}Si-A1, has been developed by AECL and tested in NRU reactor. In the course of the fuel qualification tests, only one case was performed under the higher power condition than maximum linear power which was expected in the design stage. The Korea regulatory body, KINS imposed that HANARO shall be operated at the power level less than 24MW which is 80% of the design full power until HANARO shows the repetitive performance of the fuel at the power condition abov e 112.8KW/m. To resolve this imposition, KAERI designed two types of special test bundles: two non-instrumented(Type-A) and one instrumented(Type-B) test bundles. Two Type-A bundles were irradiated in HANARO: one of them has finished PIE and the other is under PIE. Type-B bundle was loaded in the core during 1.32 day at 1996, but outstanding FIV(flow induced vibration) was observed at the pool top because of long guide tube attached to the top of the bundle. The successful installation of the chimney fastener to fix the guide tube resulted in conducting the irradiation test of Type-B bundle again. The test will start at mid- July, 2000. In order to safely do the Type-B irradiation test, the safety analysis for the nuclear, mechanical and thermal-hydraulic aspects was performed. The reactivity worth and the maximum 1 near power predicted by VENTURE are 6.3mk/k and 121.6kW/m, respectively. Thermal margins for normal and transient conditions using MATRA-h, are assessed to satisfy the safety criteria.

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.)

Temperature escalation in PWR fuel rod simulator bundles due to the zircaloy/steam reaction: Post test investigations of bundle test ESBU-2A

International Nuclear Information System (INIS)

Hagen, S.; Kapulla, H.; Malauschek, H.; Wallenfels, K.P.; Buescher, B.

1986-11-01

This KfK report describes the post test investigation of bundle experiment ESBU-2a. ESBU-2a was the second of two bundle tests on the temperature escalation of zircaloy clad fuel rods. The investigation of the temperature escalation is part of the program of out-of-pile experiments performed within the frame work of the PNS-Severe Fuel Damage program. The bundle was composed of a 3x3 fuel rod array of our fuel rod simulators (central tungsten heater, UO 2 -ring pellet and zircaloy cladding). The length was 0.4 meter. The bundle was heated to a maximum temperature of 2175 0 C. Molten cladding which dissolved part of the UO 2 pellets and slumped away from the already oxidized cladding formed a lump in the lower part of the bundle. After the test the bundle was embedded in epoxy and sectioned with a diamand saw, in the region of the refrozen melt. The cross sections were investigated by metallographic examination. The refrozen (U,Zr,O) melt consists variously of three phases with increasing oxygen content (metallic α-Zry, metallic (U,Zr) alloy and a (U,Zr)O 2 mixed oxide), two phases (α-Zry, (U,Zr)O 2 mixed oxide), or one phase ((U,Zr)O 2 mixed oxide). The cross sections show the increasing oxidation of the cladding with increasing elevation (temperature). A strong azimuthal dependency of the oxidation is found. In regions where the initial oxidized cladding is contacted by the melt one can recognize the interaction between the metallic melt and ZrO 2 of the cladding. Oxygen is taken away from the ZrO 2 . If the melt is in direct contact with steam a relatively well defined oxide layer is formed. (orig.) [de

Collimation quench test with 6.5 TeV proton beams

CERN Document Server

Salvachua Ferrando, Belen Maria; Bruce, Roderik; Hermes, Pascal Dominik; Holzer, Eva Barbara; Jacquet, Delphine; Kalliokoski, Matti; Mereghetti, Alessio; Mirarchi, Daniele; Redaelli, Stefano; Skordis, Eleftherios; Valentino, Gianluca; Valloni, Alessandra; Wollmann, Daniel; Zerlauth, Markus; CERN. Geneva. ATS Department

2016-01-01

We show here the analysis of the MD test that aimed to quench the superconducting magnets in the dispersion suppressor region downstream of the main betatron collimation system. In Run I there were several attempts to quench the magnets in the same region. This was done by exciting the Beam 2 in a controlled way using the transverse damper and generating losses leaking from the collimation cleaning. No quench was achieved in 2013 with a maximum of 1 MW of beam power loss absorbed by the collimation system at 4 TeV beam energy. In 2015 a new collimation quench test was done at 6.5 TeV aiming at similar power loss over longer period, 5-10 s. The main outcome of this test is reviewed.

Validation of ASTECV2.1 based on the QUENCH-08 experiment

Energy Technology Data Exchange (ETDEWEB)

Gómez-García-Toraño, Ignacio, E-mail: ignacio.torano@kit.edu [Karlsruhe Institute of Technology, Institute for Neutron Physics and Reactor Technology (INR), Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Sánchez-Espinoza, Víctor-Hugo; Stieglitz, Robert [Karlsruhe Institute of Technology, Institute for Neutron Physics and Reactor Technology (INR), Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Stuckert, Juri [Karlsruhe Institute of Technology, Institute for Applied Materials-Applied Materials Physics (IAM-AWP), Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Laborde, Laurent; Belon, Sébastien [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), Nuclear Safety Division/Safety Research/Severe Accident Department, Saint Paul Lez Durance 13115 (France)

2017-04-01

Highlights: • ASTECV2.1 can reproduce QUENCH-08 experimental trends e.g. hydrogen generation. • Radial temperature gradient and heat transfer through argon gap are underestimated. • Mesh sizes lower than 55 mm needed to capture the strong axial temperature gradient. • Minor variations of external electrical resistance strongly affect bundle heat-up. • Modelling of a bypass and inclusion of currents partially overcome discrepancies. - Abstract: The Fukushima accidents have shown that further improvements of Severe Accident Management Guidelines (SAMGs) are still necessary. Hence, the enhancement of severe accident codes and their validation based on integral experiments is pursued worldwide. In particular, the capabilities of the European integral severe accident ASTECV2.1 code are being extended within the CESAM project through the improvement of physical models, code numerics and an extensive code validation. Among the different strategies encompassed in the plant SAMGs, one of the most important ones to prevent core damage is the injection of water into the overheated core (reflooding). However, under certain conditions, reflooding may trigger a sharp hydrogen generation that may jeopardize the containment. Within this work, ASTECV2.1 models describing the early in-vessel phase of the severe accident and its termination by core reflooding are validated against data from the QUENCH test facility. The QUENCH-08, involving the injection of 15 g/s (about 0.6 g/s/rod) of saturated steam at a bundle temperature of 2073 K, has been selected for this comparison. Results show that ASTECV2.1 is able to reproduce the experimental temperatures and oxide thicknesses at representative bundle locations. The predicted total hydrogen generation (76 g) is similar to the experimental one (84 g). In addition, the choices of an axial mesh size lower than 55 mm and of an external electrical resistance of a 7 mΩ/rod have been justified with parametric analyses. Finally, new

High sensitive quench detection method using an integrated test wire

International Nuclear Information System (INIS)

Fevrier, A.; Tavergnier, J.P.; Nithart, H.; Kiblaire, M.; Duchateau, J.L.

1981-01-01

A high sensitive quench detection method which works even in the presence of an external perturbing magnetic field is reported. The quench signal is obtained from the difference in voltages at the superconducting winding terminals and at the terminals at a secondary winding strongly coupled to the primary. The secondary winding could consist of a ''zero-current strand'' of the superconducting cable not connected to one of the winding terminals or an integrated normal test wire inside the superconducting cable. Experimental results on quench detection obtained by this method are described. It is shown that the integrated test wire method leads to efficient and sensitive quench detection, especially in the presence of an external perturbing magnetic field

Lessons learned from the quench-11 training exercise

International Nuclear Information System (INIS)

Hohorst, J.K.; Allison, C.M.

2007-01-01

16 organizations in 12 countries are participating in a RELAP/SCDAPSIM training exercise based on the Quench 11 experiment performed at Karlsruhe (Germany) in 2005. This exercise is being conducted in parallel to an International Standard Problem (ISP). Both the ISP and the RELAP/SCDAPSIM training exercise included a 'semi-blind' portion that was completed in the fall of 2006 and an 'open' portion that is to be completed in the summer of 2007. The RELAP/SCDAPSIM training exercise is coordinated by Innovative Systems Software with support by the International SCDAP Development and Training Program (SDTP). The Quench-11 experiment is based on an electrically heated fuel rod bundle representative of a PWR design. The bundle was subjected to a boil down transient, heat-up, and quenching with peak temperatures exceeding the melting point of the Zircaloy cladding. This experiment was chosen by the European Union as an International Benchmark exercise to compare the effectiveness of quenching models in the severe accident computer codes used today for accident analysis. This paper briefly describes (a) RELAP/SCDAPSIM/MOD3.4, (b) the Quench facility and experiments used in the training exercise, and (c) the training guidelines provided to the participants followed by a more detailed description of the lessons learned from the initial 'semi-blind' portion. The representative results demonstrate that good analysts can still have a difficult time predicting the thermal hydraulic response of a relative simple transient in a complex system

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

International Nuclear Information System (INIS)

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

2005-01-01

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

Co-planar deformation and thermal propagation behavior in a bundle burst test

International Nuclear Information System (INIS)

Uetsuka, Hiroshi; Koizumi, Yasuo; Kawasaki, Satoru

1980-07-01

The probability of the suggested feedback mechanism which could lead to co-planar deformation in a bundle burst test was assessed by the data of test and the calculation based on simplified model. Following four points were evaluated. (1) The probability of local deformation during early heat up stage. (2) The relation between the characteristic of heater and the feedback mechanism. (3) Thermal propagation behavior between two adjacent rods during heat up stage. (4) The propagation of ballooning in a bundle. The probability of suggested feedback mechanism was denied in all the evaluation. The feedback mechanism suggested by Burman could not be a controlling mechanism in co-planar deformation in a bundle burst test. (author)

Quench propagation tests on the LHC superconducting magnet string

CERN Document Server

Coull, L; Krainz, G; Rodríguez-Mateos, F; Schmidt, R

1996-01-01

The installation and testing of a series connection of superconducting magnets (three 10 m long dipoles and one 3 m long quadrupole) has been a necessary step in the verification of the viability of the Large Hadron Collider at CERN. In the LHC machine, if one of the lattice dipoles or quadrupoles quenches, the current will be by-passed through cold diodes and the whole magnet chain will be de-excited by opening dump switches. In such a scenario it is very important to know whether the quench propagates from the initially quenching magnet to adjacent ones. A series of experiments have been performed with the LHC Test String powered at different current levels and at different de-excitation rates in order to understand possible mechanisms for such a propagation, and the time delays involved. Results of the tests and implications regarding the LHC machine operation are described in this paper.

Pre- and post- test calculation for the parameter-SF1 experiment with ATHLET-CD

Energy Technology Data Exchange (ETDEWEB)

Erdmann, W.; Trambauer, K.; Stuckert, J. [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (GRS), Koln (Germany)

2006-07-01

The main objective of the PARAMETER-SF1 experiment in the frame of the ISTC project 3194 is the experimental and analytical investigation of the Russian VVER-1000 fuel rod assemblies behavior under simulated conditions of a severe accident. The special feature is to study the effect of flooding a superheated test bundle from the top (top quenching) which has not yet been investigated at all. - Simulation of the PARAMETER test facility To calculate the special effects of the top quenching, some aspects are important: detailed simulation of the bundle top, top and bottom quench front, heat losses at top/bottom of bundle, electrical heater power. - Main initial and boundary conditions The proposed initial and boundary conditions for the double-blind pre-test calculation were quite different from the actual experimental data during the test e.g.: electric power, mass flow (water, steam, argon), temperature. - Conclusions: first experiment with top flooding proposed initial condition given in the specification could not be performed during the experiment, bundle parameters deviated from anticipated values, thus, the pre-calculations not comparable with the experiment, post-calculations with ATHLET-CD showed good agreement with experiment data, top flooding is well predicted, calculational results sensitive with respect to: boundary conditions, nodalization. (authors)

Application of thermal hydraulic and severe accident code SOCRAT/V3 to bottom water reflood experiment QUENCH-LOCA-0

International Nuclear Information System (INIS)

Vasiliev, A.D.; Stuckert, J.

2013-01-01

Highlights: ► QLOCA-0 test simulates a design basis LOCA NPP accident with maximum temperature 1300 K. ► Deep understanding of hydraulics and thermal mechanics under accident conditions is necessary. ► We model the test QLOCA-0 with bottom flooding using the Russian code SOCRAT/V3. ► Calculated and experimental data are in a good agreement. ► Experimental procedure is determined to reach a representative LOCA scenario in future tests. -- Abstract: The thermal hydraulic and SFD (severe fuel damage) best estimate computer modeling code SOCRAT/V3 has been used for the calculation of QUENCH-LOCA-0 experiment. The new QUENCH-LOCA bundle tests with different cladding materials will simulate a representative scenario of the LOCA (loss of coolant accident) nuclear power plant accident sequence in which the overheated up to 1300 K reactor core would be reflooded from the bottom by ECCS (emergency core cooling system). The first test QUENCH-LOCA-0 was successfully conducted at the KIT, Karlsruhe, Germany, in July 22, 2010, and was performed as the commissioning test for this series. The rod claddings are identical to that used in PWRs. The bundle was electrically heated in steam from 800 K to 1340 K with the heat-up rate of approximately 2.7 K/s. After cooling in the saturated steam the bottom flooding with water flow rate of about 100 g/s was initiated. The SOCRAT calculated results are in a good agreement with experimental data taking into account additional quenching due to water condensate entrainment at the steam cooling stage. SOCRAT/V3 has been used for estimation of further steps in experimental procedure to reach a representative LOCA scenario in future tests

Experimental investigation of the coolability of blocked hexagonal bundles

Energy Technology Data Exchange (ETDEWEB)

Hózer, Zoltán, E-mail: zoltan.hozer@energia.mta.hu; Nagy, Imre; Kunstár, Mihály; Szabó, Péter; Vér, Nóra; Farkas, Róbert; Trosztel, István; Vimi, András

2017-06-15

Highlights: • Experiments were performed with electrically heated hexagonal fuel bundles. • Coolability of ballooned VVER-440 type bundle was confirmed up to high blockage rate. • Pellet relocation effect causes delay in the cool-down of the bundle. • The bypass line does not prevent the reflood of ballooned fuel rods. - Abstract: The CODEX-COOL experimental series was carried out in order to evaluate the effect of ballooning and pellet relocation in hexagonal bundles on the coolability of fuel rods after a LOCA event. The effects of blockage geometry, coolant flowrate, initial temperature and axial profile were investigated. The experimental results confirmed that a VVER bundle up to 80% blockage rate remains coolable after a LOCA event under design basis conditions. The ballooned section creates some obstacles for the cooling water during reflood of the bundle, but this effect causes only a short delay in the cooling down of the hot fuel rods. The accumulation of fuel pellet debris in the ballooned volume results in a local power peak, which leads to further slowing down of quench front.

Beam-induced quench test of LHC main quadrupole

CERN Document Server

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

2011-01-01

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

The Comparison Analysis of Thermalhydraulic Behavior Between A Reference 37-element Bundle and A Modified 37-element Bundle

International Nuclear Information System (INIS)

Ryu, Eui-Seung; You, Sung-Chang

2014-01-01

As pressure tube diameter creep increase, the coolant flows through some of the interior subchannels of the fuel bundle are reduced and consequently reduces the Critical Heat Flux (CHF). For this reason, Canadian Utilities have performed the project that developing the new fuel design (modified 37-element bundle) to increase critical heat flux. The modified 37-element (37M) bundle has the same overall geometry as the reference 37-element (37R) bundle that is using in the Wolsong units now but the center element diameter has been reduced from 13.06mm to 11.5mm. The reduction in center element diameter of the 37M bundle design increase the flow of center areas to improve the cooling and thus to enhance CHF. The CHF experiments with 37M bundle string simulator in un-crept and crept (3.3%, 5.1% peak creep) flow channels were completed at Stern Laboratories in 2008. A substantially large increase in dryout-power was observed for the 37M bundle compared to the 37R bundle, particularly in the 5.1% crept channel. As a result of the experiments, Ontario Power Generation (OPG) and Bruce Power (BP) have increased the operational margin with this CHF correlation and has fully refueled the 37M fuel on some units or almost done on the other units. KHNP also has performed the project to refuel the 37M bundle which is the same design with OPG and BP recently. This paper summarizes the comparison assessment of Thermalhydraulic (T/H) behavior for 37M bundle and 37R bundle with their own correlations and geometry parameters. This analysis performed with the thermal hydraulic code (NUCIRC) and the site measured data at the Wolsong Unit2. Tests to evaluate the CHF performance with the 37M fuel bundle have been conducted in 2008 using the un-crept, 3.3% crept and 5.1% crept flow channels in the CHF Test facility at Stern Laboratories. In addition pressure drop tests have been performed at the same time. The changes of geometry from 37R bundle to 37M bundle reduced the center element

Evaluation of the linear power of HANARO test fuel bundles

Energy Technology Data Exchange (ETDEWEB)

Lee, Choong Sung; Seo, C. G.; Lee, B. C.; Kim, H. R

2001-02-01

The HANARO fuel was developed by AECL and it is configured in a bundle of rods containing uranium silicide. AECL has conducted a variety of tests using specimen in order to achieve its qualification and licensing and the highest linear power was evaluated to be 112.8kW/m. In design stage of HANARO, the best estimated maximum linear power at hot spot was found to occur in the transition core from the initial to the equilibrium and its value was 108kW/m, which exceeds 112.8kW/m if the physics uncertainty of the HANARO nuclear design model is taken into account. Consequently, the licensing body issued the conditional permit to operate HANARO and the fuel integrity at the linear power higher than 112.8kW/m was requested to be confirmed through irradiation tests by realizing its repeatability. Hereby, KAERI designed uninstrumented and instrumented test fuel bundles and conducted their burnup tests. In parallel with the tests, the nuclear design model has been revised and updated to enable us to pursue the pin-by-pin power history. This report describes the best estimated power history of the test fuel bundles using the revised model. In conclusion, HANARO fuel keeps its integrity at power condition greater than 120kW/m.

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)

Critical power experiment with a tight-lattice 37-rod bundle

International Nuclear Information System (INIS)

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

2006-01-01

Since most of critical power or CHF data have been collected in tube, annulus, or BWR geometries under BWR flow conditions, critical power data for highly tight and triangular lattice bundles under low mass velocity are indispensable for thermal-hydraulic design of Reduced-Moderation Water Reactor. Large-scale thermal-hydraulic experiments which use a basic 37-rod bundle test section (rod diameter: 13.0 mm, gap width between rods: 1.3 mm) were therefore carried out in this study within range of 2-9 MPa in pressure and 150-1,000 kg/(m 2 ·s) in mass velocity. Fundamental characteristics of boiling transition were investigated through effects of flow parameter on critical power and those of rod number. It was confirmed that the fundamental characteristics in 37-rod bundle are similar to those in 7-rod bundle and in case of the BWR geometry. The results of the transverse non-uniform power distribution test and subchannel analysis suggest that the critical power becomes higher when the transverse local quality distribution closes to uniform. (author)

Thermal-hydraulic design calculations for the annular fuel element with replaceable test bundles (TOAST) on the test zone position 205 of KNK II/3

International Nuclear Information System (INIS)

Norajitra, P.

1984-10-01

Annular fuel elements are foreseen in KNK II as carrier elements for irradiation inserts and test bundles. For the third core a reloadable annular element on position 205 is foreseen, in which replaceable 19-pin test bundles (TOAST) shall be irradiated. The present report deals with the thermal-hydraulic design of the annular carrier element and the test bundle, whereby the test bundle required additional optimization. The code CIA has been used for the calculations. Start of irradiation of the subassembly is planned at the beginning of the third core operation. After optimization of the pin-spacer geometry in the test bundle, design calculations for both bundles were performed, whereby thermal coupling between both was taken into account. The calculated mass-flows and temperature distributions are given for the nominal and the eccentric element configuration. The calculated bundle pressure losses have been corrected according to experimental results [de

Quench limits

International Nuclear Information System (INIS)

Sapinski, M.

2012-01-01

With thirteen beam induced quenches and numerous Machine Development tests, the current knowledge of LHC magnets quench limits still contains a lot of unknowns. Various approaches to determine the quench limits are reviewed and results of the tests are presented. Attempt to reconstruct a coherent picture emerging from these results is taken. The available methods of computation of the quench levels are presented together with dedicated particle shower simulations which are necessary to understand the tests. The future experiments, needed to reach better understanding of quench limits as well as limits for the machine operation are investigated. The possible strategies to set BLM (Beam Loss Monitor) thresholds are discussed. (author)

Fabrication of CANFLEX bundle kit for irradiation test in NRU

International Nuclear Information System (INIS)

Cho, Moon Sung; Kwon, Hyuk Il; Ji, Chul Goo; Chang, Ho Il; Sim, Ki Seob; Suk, Ho Chun.

1997-10-01

CANFLEX bundle kit was prepared at KAERI for the fabrication of complete bundle at AECL. Completed bundle will be used for irradiation test in NRU. Provisions in the 'Quality Assurance Manual for HWR Fuel Projects,' 'Manufacturing Plan' and 'Quality Verification, Inspection and Test Plan' were implemented as appropriately for the preparation of CANFLEX kit. A set of CANFLEX kit consist of 43 fuel sheath of two different sizes with spacers, bearing pads and buttons attached, 2 pieces of end plates and 86 pieces of end caps with two different sizes. All the documents utilized as references for the fabrication such as drawings, specifications, operating instructions, QC instructions and supplier's certificates are specified in this report. Especially, suppliers' certificates and inspection reports for the purchased material as well as KAERI's inspection report are integrated as attachments to this report. Attached to this report are supplier's certificates and KAERI inspection reports for the procured materials and KAERI QC inspection reports for tubes, pads, spacers, buttons, end caps, end plates and fuel sheath. (author). 37 refs

Design and preliminary test results of the quench detection system for IFSMTF

International Nuclear Information System (INIS)

Shen, S.S.; Walstrom, P.L.; Wilson, C.T.; Goddard, J.S.

1985-01-01

A unique quench detection system was designed for the International Fusion Superconducting Magnet Test Facility (IFSMTF), where a simultaneous test of six large superconducting toroidal field magnets will be carried out. The scheme was based on analog subtraction of self and neighboring pickup winding voltage from the coil voltage to yield a compensated signal proportional to a normal-zone voltage. The compensated signals were input to quench detection modules that give a quench output signal to discharge the coil if the compensated signals exceed preset thresholds for preset time durations. This paper summarizes the design and analysis of the system and presents the experimental results of the simulation tests, two-coil charging-discharging tests, and the normal-zone recovery tests

Visual observations of a degraded bundle of irradiated fuel: the Phebus FPT1 test

International Nuclear Information System (INIS)

Barrachin, M.; Bottomley, P.D.

1999-01-01

The international Phebus-FP (Fission Product) project is managed by the Institut de Protection et Surete Nucleaire in collaboration with Electricite de France (EDF), the European Commission (EC), the USNRC (USA), COG (Canada), NUPEC and JAERI (Japan), KAERI (South Korea), PSI and HSK (Switzerland). It is designed to measure the source-term and to study the degradation of irradiated UO 2 fuel in conditions typical of a severe loss of coolant accident in a pressurised water reactor (PWR). In the first test (FPT0), performed in December '93, a bundle of 20 fresh fuel rods and a central Ag-In-Cd control rod underwent a short 15-day irradiation to generate fission products before testing in the Phebus reactor in Cadarache. The second test (FPT1) was performed in July '96, in the same conditions and geometry, but using irradiated fuel (-23 GWd/tU). In the FPT1 test, the bundle was heated to an estimated 3000 K over a period of 30 minutes in order to induce a substantial liquefaction of the bundle. After the test, the bundle was embedded in epoxy and cut at different levels to investigate the mechanisms of the core degradation. This paper reports the visual observations of the degraded FPT1 bundle, very preliminary interpretations about the scenario of degradation and a comparison between the behaviour of the fuel in the FPT0 and FPT1 tests. (author)

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

International Nuclear Information System (INIS)

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

1995-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1995-08-01

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

Test and Simulation Results for Quenches Induced by Fast Losses on a LHC Quadrupole

CERN Document Server

Bracco, Ch; Bartmann, W; Bednarek, M; Lechner, A; Sapinski, M; Vittal Shetty, N; Schmidt, R; Solfaroli Camillocci, M; Verweij, A

2014-01-01

A test program for beam induced quenches was started in the LHC in 2011 in order to reduce as much as possible BLM-triggered beam dumps, without jeopardising the safety of the superconducting magnets. A first measurement was performed to asses the quench level of a quadrupole located in the LHC injection region in case of fast (ns) losses. It consisted in dumping single bunches onto an injection protection collimator located right upstream of the quadrupole, varying the bunch intensity up to 3×1010 protons and ramping the quadrupole current up to 2200 A. No quench was recorded at that time. The test was repeated in 2013 with increased bunch intensity (6.5×1010 protons); a quench occurred when powering the magnet at 2500 A. The comparison between measurements during beam induced and quench heaters induced quenches is shown. Results of FLUKA simulations on energy deposition, calculations on quench behaviour using the QP3 code and the respective estimates of quench levels are also presented.

Analytical support for the B4C control rod test QUENCH-07

International Nuclear Information System (INIS)

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

2003-04-01

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

Experimental observation of the droplet size change across a wet grid spacer in a 6 × 6 rod bundle

International Nuclear Information System (INIS)

Cho, Hyoung Kyu; Choi, Ki Yong; Cho, Seok; Song, Chul-Hwa

2011-01-01

Highlights: ► In this study, an experiment on the droplet behavior inside a heated rod bundle has been performed. ► The experiment was focused on the change of droplet size induced by a spacer grid in a rod bundle. ► The major measuring parameters of the experiment were the droplet size and velocity. ► This test provided the data on the change of the droplet size after collision with a wet grid spacer. - Abstract: During the reflood phase of a postulated loss of coolant accident in a nuclear reactor, entrainment of liquid droplets can occur at a quench front of reflooding water. It is widely recognized that the behavior of the entrained droplets crucially affects the reflood heat transfer phenomena by decreasing the superheated steam temperature and interacting with a rod bundle and spacer grids. For this reason, various experimental and numerical studies have been performed to examine droplet behavior such as the droplet size, velocity and droplet fraction inside a rod array. In this study, an experiment on the droplet behavior inside a heated rod bundle has been performed. The experiment was focused on the change of droplet size induced by a spacer grid in a rod bundle geometry, which results in the change of the interfacial heat transfer between droplets and superheated steam. A 6 × 6 rod bundle test facility in Korea Atomic Energy Research Institute was used for the experiment. Steam was supplied by an external boiler into the bottom of the test channel, and a droplet injection nozzle was equipped instead of simulating a quench front of reflooding water. The major measuring parameters of the experiment were the droplet size and velocity, which were measured by a high-speed camera and a digital image processing technique. A series of experiments were conducted with various flow conditions of a steam injection velocity, heater temperature, droplet size, and droplet flow rate. The experiments provided the data on the change of the Sauter mean diameter of

Design verification of the CANFLEX fuel bundle - quality assurance requirements for mechanical flow testing

International Nuclear Information System (INIS)

Alavi, P.; Oldaker, I.E.; Chung, C.H.; Suk, H.C.

1997-01-01

As part of the design verification program for the new fuel bundle, a series of out-reactor tests was conducted on the CANFLEX 43-element fuel bundle design. These tests simulated current CANDU 6 reactor normal operating conditions of flow, temperature and pressure. This paper describes the Quality Assurance (QA) Program implemented for the tests that were run at the testing laboratories of Atomic Energy of Canada Limited (AECL) and Korea Atomic energy Research Institute (KAERI). (author)

Testing beam-induced quench levels of LHC superconducting magnets

Directory of Open Access Journals (Sweden)

B. Auchmann

2015-06-01

Full Text Available In the years 2009–2013 the Large Hadron Collider (LHC has been operated with the top beam energies of 3.5 and 4 TeV per proton (from 2012 instead of the nominal 7 TeV. The currents in the superconducting magnets were reduced accordingly. To date only seventeen beam-induced quenches have occurred; eight of them during specially designed quench tests, the others during injection. There has not been a single beam-induced quench during normal collider operation with stored beam. The conditions, however, are expected to become much more challenging after the long LHC shutdown. The magnets will be operating at near nominal currents, and in the presence of high energy and high intensity beams with a stored energy of up to 362 MJ per beam. In this paper we summarize our efforts to understand the quench levels of LHC superconducting magnets. We describe beam-loss events and dedicated experiments with beam, as well as the simulation methods used to reproduce the observable signals. The simulated energy deposition in the coils is compared to the quench levels predicted by electrothermal models, thus allowing one to validate and improve the models which are used to set beam-dump thresholds on beam-loss monitors for run 2.

Testing beam-induced quench levels of LHC superconducting magnets

Science.gov (United States)

Auchmann, B.; Baer, T.; Bednarek, M.; Bellodi, G.; Bracco, C.; Bruce, R.; Cerutti, F.; Chetvertkova, V.; Dehning, B.; Granieri, P. P.; Hofle, W.; Holzer, E. B.; Lechner, A.; Nebot Del Busto, E.; Priebe, A.; Redaelli, S.; Salvachua, B.; Sapinski, M.; Schmidt, R.; Shetty, N.; Skordis, E.; Solfaroli, M.; Steckert, J.; Valuch, D.; Verweij, A.; Wenninger, J.; Wollmann, D.; Zerlauth, M.

2015-06-01

In the years 2009-2013 the Large Hadron Collider (LHC) has been operated with the top beam energies of 3.5 and 4 TeV per proton (from 2012) instead of the nominal 7 TeV. The currents in the superconducting magnets were reduced accordingly. To date only seventeen beam-induced quenches have occurred; eight of them during specially designed quench tests, the others during injection. There has not been a single beam-induced quench during normal collider operation with stored beam. The conditions, however, are expected to become much more challenging after the long LHC shutdown. The magnets will be operating at near nominal currents, and in the presence of high energy and high intensity beams with a stored energy of up to 362 MJ per beam. In this paper we summarize our efforts to understand the quench levels of LHC superconducting magnets. We describe beam-loss events and dedicated experiments with beam, as well as the simulation methods used to reproduce the observable signals. The simulated energy deposition in the coils is compared to the quench levels predicted by electrothermal models, thus allowing one to validate and improve the models which are used to set beam-dump thresholds on beam-loss monitors for run 2.

Testing beam-induced quench levels of LHC superconducting magnets

CERN Document Server

Auchmann, B.; Bednarek, M.; Bellodi, G.; Bracco, C.; Bruce, R.; Cerutti, F.; Chetvertkova, V.; Dehning, B.; Granieri, P.P.; Hofle, W.; Holzer, E.B.; Lechner, A.; Del Busto, E. Nebot; Priebe, A.; Redaelli, S.; Salvachua, B.; Sapinski, M.; Schmidt, R.; Shetty, N.; Skordis, E.; Solfaroli, M.; Steckert, J.; Valuch, D.; Verweij, A.; Wenninger, J.; Wollmann, D.; Zerlauth, M.

2015-06-25

In the years 2009-2013 the Large Hadron Collider (LHC) has been operated with the top beam energies of 3.5 TeV and 4 TeV per proton (from 2012) instead of the nominal 7 TeV. The currents in the superconducting magnets were reduced accordingly. To date only seventeen beam-induced quenches have occurred; eight of them during specially designed quench tests, the others during injection. There has not been a single beam- induced quench during normal collider operation with stored beam. The conditions, however, are expected to become much more challenging after the long LHC shutdown. The magnets will be operating at near nominal currents, and in the presence of high energy and high intensity beams with a stored energy of up to 362 MJ per beam. In this paper we summarize our efforts to understand the quench levels of LHC superconducting magnets. We describe beam-loss events and dedicated experiments with beam, as well as the simulation methods used to reproduce the observable signals. The simulated energy depositio...

Detailed analysis of the bundle damage scenario in the PHEBUS FPT0

International Nuclear Information System (INIS)

Park, Rae Joon; Kim, Sang Baik; Kim, Hee Dong; Yoo, Kun Joong

1998-03-01

The PHEBUS FP program and the test facility have been investigated, and the late phase melt progression in the PHEBUS FPT0 has been analyzed in the present study. The objectives of this program are to investigate fission product (FP) release and this program consists of six in-pile tests, which are FPT0, FPT1, FPT4, FPT2, FPT5, and FPT3, under different thermal hydraulic and fuel rod environment conditions. The first test, FPT0, was performed in December 1993, and the second test, FPT1, was performed in July 1996. The present study has been performed to evaluate a late phase damage scenario of the fuel bundle using the FPT0 test results, which are primarily a non-destructive Post Irradiation Examination (PIE) and a destructive PIE. The fuel bundle degradation scenario is summarized as follows: the fuel rod cladding failed at approximately 7,000 seconds; the control rod materials ruptured at 11,000 seconds; the stainless-steel reaction occurs at approximately 12,100 seconds; the upper fuel bundle materials melted and relocated to the elevation between 35 and 45 cm at the period between 14,750 and 15,200 seconds; the molten pool and the debris were formed at the elevation between 26 and 36 cm at the period between 15,200 and 18,100 seconds; the molten pool and the debris dropped the elevation between 15 and 25 cm from the bfc at approximately 18,100 seconds; the molten pool was finally quenched by the injected steam. (author). 45 refs., 10 tabs., 73 figs

Quench tests of Nb3Al small racetrack magnets

International Nuclear Information System (INIS)

Yamada, R.; Kikuchi, A.; Tartaglia, Michael Albert; Ambrosio, G.; Andreev, N.; Barzi, E.; Carcagno, R.; Feher, S.; Kashikhin, V.V.; Kotelnikov, S.; Lamm, Michael J.; Fermilab; NIMC, Tsukuba; KEK, Tsukuba

2007-01-01

Two Cu stabilized Nb3Al strands, F1 (Nb matrixed) and F3 (Ta matrixed), have been made at NIMS and their Rutherford cables were made at Fermilab in collaboration with NIMS. A Small Race-track magnet using F1 Rutherford cable, the first Nb3Al dipole magnet in the world, was constructed and tested to full current at Fermilab. This magnet was tested extensively to full short sample data and its quench characteristics were studied and reported. The 3-D magnetic field calculation was done with ANSYS to find the peak field. The quench characteristics of the magnet are explained with the characteristics of the Nb3Al strand and Rutherford cable. The other Small Race-track magnet using Ta matrixed F3 strand was constructed and will be tested in the near future. The advantages and disadvantages of these Nb3Al cables are discussed

Quench tests of Nb3Al small racetrack magnets

Energy Technology Data Exchange (ETDEWEB)

Yamada, R.; Kikuchi, A.; Tartaglia, Michael Albert; Ambrosio, G.; Andreev, N.; Barzi, E.; Carcagno, R.; Feher, S.; Kashikhin, V.V.; Kotelnikov, S.; Lamm, Michael J.; /Fermilab /NIMC, Tsukuba /KEK, Tsukuba

2007-08-01

Two Cu stabilized Nb3Al strands, F1 (Nb matrixed) and F3 (Ta matrixed), have been made at NIMS and their Rutherford cables were made at Fermilab in collaboration with NIMS. A Small Race-track magnet using F1 Rutherford cable, the first Nb3Al dipole magnet in the world, was constructed and tested to full current at Fermilab. This magnet was tested extensively to full short sample data and its quench characteristics were studied and reported. The 3-D magnetic field calculation was done with ANSYS to find the peak field. The quench characteristics of the magnet are explained with the characteristics of the Nb3Al strand and Rutherford cable. The other Small Race-track magnet using Ta matrixed F3 strand was constructed and will be tested in the near future. The advantages and disadvantages of these Nb3Al cables are discussed.

Fabrication of a CANFLEX-RU designed bundle for power ramp irradiation test in NRU

International Nuclear Information System (INIS)

Cho, Moon Sung

2000-11-01

The BDL-443 CANFLEX-RU bundle AKW was fabricated at Korea Atomic Energy Research Institute (KAERI) for power ramp irradiation testing in NRU reactor. The bundle was fabricated with IDR and ADU fuel pellets in adjacent elements and contains fuel pellets enriched to 1.65 wt% 235 U in the outer and intermediate rings and also contains pellets enriched to 2.00 wt% 235 U in the inner ring. This bundle does not have a center element to allow for insertion on a hanger bar. KAERI produced the IDR pellets with the IDR-source UO 2 powder supplied by BNFL. ADU pellets were fabricated and supplied by AECL. Bundle kits (Zircaloy-4 end plates, end plugs, and sheaths with brazed appendages) manufactured at KAERI earlier in 1996 were used for the fabrication of the bundle. The CANFLEX bundle was fabricated successfully at KAERI according to the QA provisions specified in references and as per relevant KAERI drawings and technical specification. This report covers the fabrication activities performed at KAERI. Fabrication processes performed at AECL will be documented in a separate report

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

Energy Technology Data Exchange (ETDEWEB)

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

2003-04-01

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

Determination of the hydrogen source term during the reflooding of an overheated core: Calculation results of the integral reflood test QUENCH-03 with PWR-type bundle

International Nuclear Information System (INIS)

Chikhi, Nourdine; Nguyen, Nam Giang; Fleurot, Joelle

2012-01-01

Highlights: ► Calculation of QUENCH-03 experiment with ASTEC/CATHARE. ► Validation of reflooding model in severe accidents conditions. ► Demonstration of a minimum flow rate for a successful reflood by using a system code. ► Effect of injection flow rate on hydrogen production. ► Effect of initial core temperature on hydrogen production. - Abstract: During a severe accident, one of the main accident management procedure consists of injecting water in the reactor core by means of various safety injection devices. Nevertheless, the success of a core reflood is not guaranteed because of possible negative effects: temperature escalation, enhanced hydrogen production, enhanced release of fission products, core degradation due to thermal shock, shattering, debris and melt formation. The QUENCH-03 experiment was carried out to investigate the behavior on reflooding at high temperature of LWR fuel rods with little oxidation. Posttest calculations with the ASTEC-CATHARE V2 code were made for code assessment and validation of the new reflooding model. This thermal–hydraulic model is used to detect the quench front position and to calculate the heat transfer between fuel and fluid in the transition boiling region. Comparisons between the calculational and experimental results are presented. Emphasis has been placed on clad temperature, hydrogen production and melt relocation. The effects of core state damage (initial temperature at reflooding onset) and the reflood mass flow rate on the hydrogen source term were investigated using the QUENCH-03 test as a base case. Calculations were made by varying both parameters in the input data deck. The results demonstrate (and confirm) the existence of a minimum flow rate for a successful reflood.

Comparison and Interpretation Report of the OECD International Standard Problem No. 45 - Exercise (QUENCH-06)

International Nuclear Information System (INIS)

Hering, W.; Homann, Ch.; Lamy, J.S.; Miassoedov, A.; Schanz, G.; Sepold, L.; Steinbrueck, M.

2002-10-01

The International Standard Problem (ISP) No. 45 is part of the overall ISP program of the OECD/NEA and is dedicated to the behavior of heat-up and delayed reflood of fuel elements in nuclear reactors during a hypothetical accident. ISP-45 is related to the out-of-pile bundle quench experiment QUENCH-06, performed at Forschungszentrum Karlsruhe (FZK), Germany, on December 13, 2000. Special attention was paid to hydrogen production. To assess the ability of severe accident codes to simulate processes during core heat-up and reflood at temperatures above 2000 K, the behavior of the bundle during the whole experiment should be calculated on the basis of the necessary experimental initial and boundary conditions, but without knowing further experimental details. In this so-called blind phase 21 participants from 15 nations contributed with 8 different code systems (ATHLET-CD, ICARE/CATHARE, IMPACT/SAMPSON, GENFLO, MAAP, MELCOR, SCDAPSIM, SCDAP-3D). Additionally, posttest calculations using the in-house version SCDAP/RELAP5 mod3.2.irs are used for comparison. After the end of the blind phase all measured data were made available and the participants were invited to deliver a second calculation, where this knowledge could be used (so-called open phase). In this report, results of the blind calculations are presented, analyzed, and compared to experimental data. During heat-up most results do not deviate significantly from one another, except as a consequence of some obvious user errors, so that a definition of a mainstream is justified. For the quench phase the lack of adequate hydraulic modeling becomes obvious: some participants could not match the observed cool-down rates, others had to use very fine meshes to compensate code deficiencies. To overcome this insufficiency some newly developed reflood models were used in MAAP and MELCOR. In QUENCH-06, oxide layers were thick enough to protect the cladding from melting and failure below 2200 K, so that no massive hydrogen

A quench detection/logging system for the SSCL Magnet Test Laboratory

International Nuclear Information System (INIS)

Kim, K.; Coles, M.; Dryer, J.; Lambert, D.

1993-05-01

The quench in a magnet describes a process which occurs while the superconductivity state goes to the normal resistive state. The consequence of a quench is the conversion of the stored electromagnetic energy into heat. During this process the initiating point will reach a high temperature, which will char the insulation or melt the conductor and thereby destroy the magnet. To prevent the magnet from being lost, it is standard practice to observe several resistance and/or inductance voltages across the magnet as quench signatures -- detection. When a quench symptom is detected, protection operations are initiated: proper shutdown of the magnet excitation systems and treatment to dilute the heat energy at a spot -- protection. The temperature rise is diluted by firing heaters along the length of the magnet to insure that the dissipated energy is spread. To develop a reliable quench detection system, two distinct approaches have been tried in the past: (i) Understanding of the Noise Mechanism and Sub-system Optimization, and (ii) Escaping from the Known Electromagnetic Noises by Observing Optical Waves or Acoustic Waves. The MTL of SSCL confronts a mass-measurement of about 10,000 production magnets. To meet the testing schedule, the false quench detection rate needs to be further optimized while the true quench detection rate remains secure for the magnet measurement safety. To meet these requirements, we followed an iterative top-down approach. First we defined the signal and noise characteristics of the quench phenomena by using existing software tools to build a rapid prototype system incorporating all proven functionality of the existing system. Then we further optimize the system through iterative upgrading based on our signal and noise character findings

A phenomenological model of thermal-hydraulics of convective boiling during the quenching of hot rod bundles

International Nuclear Information System (INIS)

Unal, C.; Nelson, R.

1991-01-01

After completion of the thermal-hydraulic model developed in a companion paper, the authors performed developmental assessment calculation of the model using steady-state and transient post-critical heat flux (CHF) data. This paper discusses the results of those calculations. The overall interfacial drag model predicted reasonable drag coefficients for both the nucleate boiling and the inverted annular flow (IAF) regimes. The predicted pressure drops agreed reasonably well with the measured data of two transient experiments, CCTF Run 14 and a Lehigh reflood test. The thermal-hydraulic model for post-CHF convective heat transfer predicted the rewetting velocities reasonably well for both experiments. The predicted average slope of the wall temperature traces for these tests showed reasonable agreement with the measured data, indicating that the transient-calculated precursory cooling rates agreed with measured data. The hot-patch model, in conjunction with the other thermal-hydraulic models, was capable of modeling the Winfrith post-CHF hot-patch experiments. The hot-patch model kept the wall temperatures at the specified levels in the hot-patch regions and did not allow any quench-front propagation from either the bottom or the top of the test section. The interfacial heat-transfer model tended to slightly underpredict the vapor temperatures. The maximum difference between calculated and measured vapor temperatures was 20%, with a 10% difference for the remainder of the runs considered. The wall-to-fluid heat transfer was predicted reasonably well, and the predicted wall temperatures were in reasonable agreement with measured data with a maximum relative error of less than 13%

Characterizing Water Quenching Systems with a Quench Probe

Science.gov (United States)

Ferguson, B. Lynn; Li, Zhichao; Freborg, Andrew M.

2014-12-01

Quench probes have been used effectively to characterize the quality of quenchants for many years. For this purpose, a variety of commercial probes, as well as the necessary data acquisition system for determining the time-temperature data for a set of standardized test conditions, are available for purchase. The type of information obtained from such probes provides a good basis for comparing media, characterizing general cooling capabilities, and checking media condition over time. However, these data do not adequately characterize the actual production quenching process in terms of heat transfer behavior in many cases, especially when high temperature gradients are present. Faced with the need to characterize water quenching practices, including conventional and intensive practices, a quench probe was developed. This paper describes that probe, the data collection system, the data gathered for both intensive quenching and conventional water quenching, and the heat transfer coefficients determined for these processes. Process sensitivities are investigated and highlight some intricacies of quenching.

Fabrication of a CANFLEX-RU designed bundle for power ramp irradiation test in NRU

Energy Technology Data Exchange (ETDEWEB)

Cho, Moon Sung

2000-11-01

The BDL-443 CANFLEX-RU bundle AKW was fabricated at Korea Atomic Energy Research Institute (KAERI) for power ramp irradiation testing in NRU reactor. The bundle was fabricated with IDR and ADU fuel pellets in adjacent elements and contains fuel pellets enriched to 1.65 wt% {sup 235}U in the outer and intermediate rings and also contains pellets enriched to 2.00 wt% {sup 235}U in the inner ring. This bundle does not have a center element to allow for insertion on a hanger bar. KAERI produced the IDR pellets with the IDR-source UO{sub 2} powder supplied by BNFL. ADU pellets were fabricated and supplied by AECL. Bundle kits (Zircaloy-4 end plates, end plugs, and sheaths with brazed appendages) manufactured at KAERI earlier in 1996 were used for the fabrication of the bundle. The CANFLEX bundle was fabricated successfully at KAERI according to the QA provisions specified in references and as per relevant KAERI drawings and technical specification. This report covers the fabrication activities performed at KAERI. Fabrication processes performed at AECL will be documented in a separate report.

Sectional pipeline bundles. Design, fabrication and testing of a subsea pipeline connection system

Energy Technology Data Exchange (ETDEWEB)

NONE

1999-08-01

The tests of the prototype system indicated that the system is applicable for connecting pipeline bundle sections. The overall performance of the system is therefore concluded to be satisfactory. Some modifications are required though, for improving the reliability of the system to the level required for offshore North Sea application. The tests showed that connection of the pipeline bundle sections can be performed for alignment tolerances larger than those expected during a typical subsea installation. Pull-in of bundle end sections can be performed with pull-in wires deployed from surface. The offshore tests showed that handling of wires must be done with great care to avoid possibility for wire entanglement, especially if a fully diverless system is to be used. The flowline connection tool was found to be suitable for final alignment of the individual spool ends. It was demonstrated that face to face contact between the hub faces in the connector was obtained after tie-in. Pressure tests showed that the connector could be sealed by the tie-in force applied by the connection tool tie-in system. However, the standard connector clamp which was used, was found to be insuficient for maintaining the connector effectively sealed after removal of the pull-in force applied by the connection tool. Based on the results proposals for improvements of the system are included. Improvements are applicable to the current system for connection of bundle sections or for tie-in operations, relating to conventional pipelines. The improvements also includes a strong connection clamp suitable for subsea use. The connection clamp will replace the standard clamp devise used in this project. (au) EFP-96. 41 refs.

A quench detection/logging system for the SSCL Magnet Test Laboratory

International Nuclear Information System (INIS)

Kim, K.; Coles, M.; Dryer, J.; Lambert, D.

1994-01-01

The quench in a magnet describes a process which occurs while the superconductivity state goes to the normal resistive state. The consequence of a quench is the conversion of the stored electromagnetic energy into heat. During this process the initiating point will reach a high temperature, which will char the insulation or melt the conductor and thereby destroy the magnet. To prevent the magnet from being lost, it is standard practice to observe several resistance and/or inductance voltages across the magnet as quench signatures - Detection. When a quench symptom is detected, protection operations are initiated: proper shutdown of the magnet excitation systems and treatment to dilute the heat energy at a spot - Protection. The temperature rise is diluted by firing heaters along the length of the magnet to ensure that the dissipated energy is spread. It is interesting that there is not a significant amount of published research on detection. To afford a more reliable quench detection system, two distinct approaches have been tried in the past: (i) Understanding of the Noise Mechanism and Sub-system Optimization, and (ii) Escaping from the Known Electromagnetic Noises by Observing Optical Waves or Acoustic Waves. The MTL of SSCL confronts a mass-measurement of about 10,000 production magnets. To meet the testing schedule, the false quench detection rate needs to be further optimized while the true quench detection rate remains secure for the magnet measurement safety. To meet these requirements, the authors followed an iterative top-down approach. First they defend the signal and noise characteristics of the quench phenomena by using existing software tools to build a rapid prototype system incorporating all proven functionality of the existing system. Then they further optimize the system through iterative upgrading based on their signal and noise character findings

Design Report for a 19-pin carbide test-bundle in a ring-subassembly of the test zone of KNK II/2

International Nuclear Information System (INIS)

Haefner, H.E.

1982-03-01

This report describes a 19-rod carbide test bundle in an annular oxide ring element placed at the position 201 of the test zone in the second core of KNK II as well as its behavior during the period of operation. The selected fuel rod concept includes low pellet density and a relatively large gap width as well as helium bonding between fuel and cladding. Characteristic design and operation data are: rod diameter 8.5 mm, pellet diameter 7.0 mm, maximum nominal linear rating 800 W/cm, maximum nominal burnup 70 MWd/kgHM. This report exclusively deals with the carbide test bundle and its individual components; it describes methods, criteria and results concerning the design. The annular carrier element with its head and foot is treated in a separate report. The loadability of the test bundle and its individual components is demonstrated by generally valid standards for strength criteria [de

Quenching Combustible Dust Mixtures Using Electric Particulate Suspensions (EPS): A New Testing Method For Microgravity

Science.gov (United States)

Colver, Gerald M.; Greene, Nathanael; Shoemaker, David; Xu, Hua

2003-01-01

The Electric Particulate Suspension (EPS) is a combustion ignition system being developed at Iowa State University for evaluating quenching effects of powders in microgravity (quenching distance, ignition energy, flammability limits). Because of the high cloud uniformity possible and its simplicity, the EPS method has potential for "benchmark" design of quenching flames that would provide NASA and the scientific community with a new fire standard. Microgravity is expected to increase suspension uniformity even further and extend combustion testing to higher concentrations (rich fuel limit) than is possible at normal gravity. Two new combustion parameters are being investigated with this new method: (1) the particle velocity distribution and (2) particle-oxidant slip velocity. Both walls and (inert) particles can be tested as quenching media. The EPS method supports combustion modeling by providing accurate measurement of flame-quenching distance as a parameter in laminar flame theory as it closely relates to characteristic flame thickness and flame structure. Because of its design simplicity, EPS is suitable for testing on the International Space Station (ISS). Laser scans showing stratification effects at 1-g have been studied for different materials, aluminum, glass, and copper. PTV/PIV and a leak hole sampling rig give particle velocity distribution with particle slip velocity evaluated using LDA. Sample quenching and ignition energy curves are given for aluminum powder. Testing is planned for the KC-135 and NASA s two second drop tower. Only 1-g ground-based data have been reported to date.

The climb of dissociated dislocations in a quenched Cu-13.43 at.% Al alloy

International Nuclear Information System (INIS)

Decamps, B.; Cherns, D.; Condat, M.

1983-01-01

The weak-beam electron microscopy technique has been used to study the climb of dissociated dislocations in a Cu-13.43 at.% Al alloy under conditions of supersaturation of vacancies introduced by quenching. The results are similar to those obtained under electron irradiation (interstitial climb) in the same alloy (Cherns, Hirsch and Saka 1980) in that climb may proceed by the nucleation of prismatic loops on the individual partials. The nature of the loops is such as to minimize the total energy of the configuration (partial plus loop) and to maximize their edge component. Interaction with the other partial has been observed, causing the entire dislocation to climb. Additional features observed suggest that climb under quenching is initiated by the nucleation of Frank loops. The detailed configurations also enable climb by absorption of vacancies and interstitials to be distinguished. (author)

Quenches after LS1

International Nuclear Information System (INIS)

Verweij, A.P.

2012-01-01

In this paper I will give an overview of the different types of quenches that occur in the LHC, followed by an estimate of the number of quenches that we can expect after LS1. Beam-induced quenches and false triggering of the QPS will be the main cause of those quenches that cause a beam dump. Possibly in total up to 10-20 per year. After consolidation of the 13 kA joints, the approach for the BLM settings can be less conservative than in 2010-2012 in order to maximize beam time. This will cause some quenches but, anyhow, a beam.induced quench is not more risky than a quench provoked by false triggering. It is not easy to predict the number of BLM triggered beam dumps, needed to avoid magnet quenches, because it is not sure how to scale beam losses and UFO's from 3.5 TeV to 6.5 TeV, and it is not sure if the thresholds at 3.5 TeV are correct. Quench events will be much more massive (ex: RB quench at 6 kA → 2 MJ, RB quench at 11 kA → 6-20 MJ), and as a result cryo recuperation much longer. There will also be more ramp induced quenches after a FPA in other circuits due to higher ramp rates and smaller temperature margins (mutual coupling)

Quench and protection characteristics of the GEM test coil

International Nuclear Information System (INIS)

Chaniotakis, E.A.; Marston, P.G.

1994-01-01

The GEM test coil, will be wound from 70 m of conductor identical to that used in the full scale magnet. The coil configuration will duplicate the field distribution of the full scale magnet and current control will duplicate full scale current decay characteristics. Therefore, quench/protection analysis of this coil will reveal very important information about the behavior of the full scale model. Due to the uncertainty associated with the contact between the cable, the conduit and the sheath, a parametric analysis has been performed in order to determine and bracket the behavior. With no electrical contact the quench evolves normally until, due to heat transfer from the sheath into the cable, the superconductor temperature becomes critical and the entire length becomes normal

Three dimensional numeric quench simulation of Super-FRS dipole test coil for FAIR project

International Nuclear Information System (INIS)

Wu Wei; Ma Lizhen; He Yuan; Yuan Ping

2013-01-01

The prototype of superferric dipoles for Super-FRS of Facility for Antiprotons and Ion Research (FAIR) project was designed, fabricated, and tested in China. To investigate the performance of the superconducting coil, a so-called test coil was fabricated and tested in advance. A 3D model based on ANSYS and OPERA 3D was developed in parallel, not only to check if the design matches the numerical simulation, but also to study more details of quench phenomena. The model simplifies the epoxy impregnated coil into an anisotropic continuum medium. The simulation combines ANSYS solver routines for nonlinear transient thermal analysis, the OPERA 3D for magnetic field evaluation and the ANSYS script language for calculations of Joule heat and differential equations of the protection circuits. The time changes of temperature, voltage and current decay, and quench propagation during quench process were analyzed and illustrated. Finally, the test results of the test coil were demonstrated and compared with the results of simulation. (authors)

Photoluminescence quenching by OH in Er- and Pr-doped glasses for 1.5 and 1.3 μm optical amplifiers

Science.gov (United States)

Faber, Anne J.; Simons, Dennis R.; Yan, Yingchao; de Waal, Henk

1994-09-01

In this paper we report on the effect of hydroxyl (OH) groups on the photoluminescence in the near IR (1.5 and 1.3 micrometers ) in rare earth (Er, Pr)-doped glasses. The 1.5 micrometers emission of Er-doped phosphate glasses was studied, before and after a special heat treatment. The luminescent lifetime of the 1.5 micrometers emission increases substantially, typically from 3 ms up to 7.2 ms for a 2 mole% Er2O3-doped phosphate glass, due to the controlled heat treatment. The increase in lifetime is ascribed to a decrease in OH- concentration, which is confirmed by IR-absorption spectroscopy. The quenching by OH is described by a simplified quenching model, which predicts the 1.5 micrometers emission lifetime as a function of Er- concentration with the OH-concentration as parameter. It appears that the larger part of the OH groups is coupled to Er ions and thus acts as quenching center. Photoluminescence quenching by OH groups is also reported for the 1.3 micrometers emission of Pr in GeS2-glasses: In pure OH-free GeS2 glass the 1.3 micrometers emission lifetime is as high as 350 microsecond(s) , for a 400 ppm dopant level. In GeS2 glasses containing only small amounts of OH (approximately 100 ppm), this lifetime is less than 200 microsecond(s) . Both examples demonstrate that for the fabrication of efficient glass optical amplifiers at the telecommunication windows 1.3 and 1.5 micrometers , the OH-impurity level of the host glass must be kept as low as possible.

Quench origins

International Nuclear Information System (INIS)

Devred, A.

1990-03-01

Quenches can be divided into two categories; conductor-limited and energy-deposited quenches. A conductor-limited quench occurs when the current in the magnet exceeds the capacity of the superconductor; it is characterized by a strong correlation with temperature. An energy-deposited quench occurs when a disturbance releases enough energy to trigger a quench; the main disturbances during magnet energization are frictional movements of the conductor due to increasing Lorentz forces. The current level of the conductor-limited quenches defines the limit of the magnet performance, and can only be surpassed by lowering the operating temperature; the occurrence of a constant current at quench during the magnetic testing is called a plateau. Usually it takes a few energy-deposited quenches of increasing currents to reach the plateau; these first few steps are called the magnet's training. The goal in designing a magnet is to be able to energize it and to reliably operate it at the design current without training. This can be achieved by optimizing the magnet's operating margin, that is, by designing and building the magnet in such a way that the sizes of the mechanical disturbances needed to trigger a quench are much larger than the achievable mechanical tolerances. (N.K.) 112 refs

Dynamic behaviour of FBR fuel pin bundles

International Nuclear Information System (INIS)

Martin, P.H.; Van Dorsselaere, J.P.; Ravenet, A.

1990-01-01

A programme of shock tests on a fast neutron reactor subassembly model (SPX1 geometry) including a complete bundle of fuel pins (dummy elements) is being carried out in the BELIER test facility at Cadarache. The purpose of these tests is: to determine the distribution of dynamic forces applied to the fuel rod clads under the impact conditions encountered in a reactor during a earthquake; to reduce as much as possible the conservatism of the methods presently used for the calculation of those forces. The test programme, now being completed, consists of the following steps: impacts on the mock-up in air with an non-compact bundle (situation of the subassembly at beginning of life (BOL) with clearances within the bundle); impacts under the same conditions but with fluid (water) in the subassembly; impacts on the mock-up in air and with a compacted bundle (simulating the conditions of an end-of-life (EOL) bundle with no clearance within the bundle). The accelerations studied in these tests cover the range encountered in design calculations for the subassembly frequencies in beam mode. (author)

Evaluation of Single-Bundle versus Double-Bundle PCL Reconstructions with More Than 10-Year Follow-Up

Directory of Open Access Journals (Sweden)

Masataka Deie

2015-01-01

Full Text Available Background. Posterior cruciate ligament (PCL injuries are not rare in acute knee injuries, and several recent anatomical studies of the PCL and reconstructive surgical techniques have generated improved patient results. Now, we have evaluated PCL reconstructions performed by either the single-bundle or double-bundle technique in a patient group followed up retrospectively for more than 10 years. Methods. PCL reconstructions were conducted using the single-bundle (27 cases or double-bundle (13 cases method from 1999 to 2002. The mean age at surgery was 34 years in the single-bundle group and 32 years in the double-bundle group. The mean follow-up period was 12.5 years. Patients were evaluated by Lysholm scoring, the gravity sag view, and knee arthrometry. Results. The Lysholm score after surgery was 89.1±5.6 points for the single-bundle group and 91.9±4.5 points for the double-bundle group. There was no significant difference between the methods in the side-to-side differences by gravity sag view or knee arthrometer evaluation, although several cases in both groups showed a side-to-side difference exceeding 5 mm by the latter evaluation method. Conclusions. We found no significant difference between single- and double-bundle PCL reconstructions during more than 10 years of follow-up.

Verification of the FBR fuel bundle–duct interaction analysis code BAMBOO by the out-of-pile bundle compression test with large diameter pins

Energy Technology Data Exchange (ETDEWEB)

Uwaba, Tomoyuki, E-mail: uwaba.tomoyuki@jaea.go.jp [Japan Atomic Energy Agency, 4002, Narita-cho, Oarai-machi, Ibaraki 311-1393 (Japan); Ito, Masahiro; Nemoto, Junichi [Japan Atomic Energy Agency, 4002, Narita-cho, Oarai-machi, Ibaraki 311-1393 (Japan); Ichikawa, Shoichi [Japan Atomic Energy Agency, 2-1, Shiraki, Tsuruga-shi, Fukui 919-1279 (Japan); Katsuyama, Kozo [Japan Atomic Energy Agency, 4002, Narita-cho, Oarai-machi, Ibaraki 311-1393 (Japan)

2014-09-15

The BAMBOO computer code was verified by results for the out-of-pile bundle compression test with large diameter pin bundle deformation under the bundle–duct interaction (BDI) condition. The pin diameters of the examined test bundles were 8.5 mm and 10.4 mm, which are targeted as preliminary fuel pin diameters for the upgraded core of the prototype fast breeder reactor (FBR) and for demonstration and commercial FBRs studied in the FaCT project. In the bundle compression test, bundle cross-sectional views were obtained from X-ray computer tomography (CT) images and local parameters of bundle deformation such as pin-to-duct and pin-to-pin clearances were measured by CT image analyses. In the verification, calculation results of bundle deformation obtained by the BAMBOO code analyses were compared with the experimental results from the CT image analyses. The comparison showed that the BAMBOO code reasonably predicts deformation of large diameter pin bundles under the BDI condition by assuming that pin bowing and cladding oval distortion are the major deformation mechanisms, the same as in the case of small diameter pin bundles. In addition, the BAMBOO analysis results confirmed that cladding oval distortion effectively suppresses BDI in large diameter pin bundles as well as in small diameter pin bundles.

Quench protection test results and comparative simulations on the first 10 meter prototype dipoles for the Large Hadron Collider

International Nuclear Information System (INIS)

Rodriguez-Mateos, F.; Gerin, G.; Marquis, A.

1996-01-01

The first 10 meter long dipole prototypes made by European Industry within the framework of the R and D program for the Large Hadron Collider (LHC) have been tested at CERN. As a part of the test program, a series of quench protection tests have been carried out in order to qualify the basic protection scheme foreseen for the LHC dipoles (quench heaters and cold diodes). Results are presented on the quench heater performance, and on the maximum temperatures and voltages observed during quenches under the so-called machine conditions. Moreover, an update of the quench simulation package specially developed at CERN (QUABER 2) has been recently made. Details on this new version of QUABER are given. Simulation runs have been made specifically to validate the model with the results from the measurements on quench protection mentioned above

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

International Nuclear Information System (INIS)

Baker, S.P.

1980-06-01

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

Architecture of a software quench management system

International Nuclear Information System (INIS)

Jerzy M. Nogiec et al.

2001-01-01

Testing superconducting accelerator magnets is inherently coupled with the proper handling of quenches; i.e., protecting the magnet and characterizing the quench process. Therefore, software implementations must include elements of both data acquisition and real-time controls. The architecture of the quench management software developed at Fermilab's Magnet Test Facility is described. This system consists of quench detection, quench protection, and quench characterization components that execute concurrently in a distributed system. Collaboration between the elements of quench detection, quench characterization and current control are discussed, together with a schema of distributed saving of various quench-related data. Solutions to synchronization and reliability in such a distributed quench system are also presented

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

International Nuclear Information System (INIS)

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

1984-07-01

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

Evaluation of droplet deposition in rod bundle

International Nuclear Information System (INIS)

Ji, W.; Gu, C.Y.; Anglart, H.

1997-01-01

Deposition model for droplets in gas droplet two-phase flow in rod bundle is developed in this work using the Lagrangian method. The model is evaluated in a 9-rod bundle geometry. The deposition coefficient in the bundle geometry are compared with that in round tube. The influences of the droplet size and gas mass flow rate on deposition coefficient are investigated. Furthermore, the droplet motion is studied in more detail by dividing the bundle channel into sub-channels. The results show that the overall deposition coefficient in the bundle geometry is close to that in the round tube with the diameter equal to the bundle hydraulic diameter. The calculated deposition coefficient is found to be higher for higher gas mass flux and smaller droplets. The study in the sub-channels show that the ratio between the local deposition coefficient for a sub-channel and the averaged value for the whole bundle is close to a constant value, deviations from the mean value for all the calculated cases being within the range of ±13%. (author)

Influence of Bundle Diameter and Attachment Point on Kinematic Behavior in Double Bundle Anterior Cruciate Ligament Reconstruction Using Computational Model

Directory of Open Access Journals (Sweden)

Oh Soo Kwon

2014-01-01

Full Text Available A protocol to choose the graft diameter attachment point of each bundle has not yet been determined since they are usually dependent on a surgeon’s preference. Therefore, the influence of bundle diameters and attachment points on the kinematics of the knee joint needs to be quantitatively analyzed. A three-dimensional knee model was reconstructed with computed tomography images of a 26-year-old man. Based on the model, models of double bundle anterior cruciate ligament (ACL reconstruction were developed. The anterior tibial translations for the anterior drawer test and the internal tibial rotation for the pivot shift test were investigated according to variation of bundle diameters and attachment points. For the model in this study, the knee kinematics after the double bundle ACL reconstruction were dependent on the attachment point and not much influenced by the bundle diameter although larger sized anterior-medial bundles provided increased stability in the knee joint. Therefore, in the clinical setting, the bundle attachment point needs to be considered prior to the bundle diameter, and the current selection method of graft diameters for both bundles appears justified.

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

International Nuclear Information System (INIS)

1987-01-01

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

Bundling of harvesting residues and whole-trees and the treatment of bundles; Hakkuutaehteiden ja kokopuiden niputus ja nippujen kaesittely

Energy Technology Data Exchange (ETDEWEB)

Kaipainen, H; Seppaenen, V; Rinne, S

1997-12-31

The conditions on which the bundling of the harvesting residues from spruce regeneration fellings would become profitable were studied. The calculations showed that one of the most important features was sufficient compaction of the bundle, so that the portion of the wood in the unit volume of the bundle has to be more than 40 %. The tests showed that the timber grab loader of farm tractor was insufficient for production of dense bundles. The feeding and compression device of the prototype bundler was constructed in the research and with this device the required density was obtained.The rate of compaction of the dry spruce felling residues was about 40 % and that of the fresh residues was more than 50 %. The comparison between the bundles showed that the calorific value of the fresh bundle per unit volume was nearly 30 % higher than that of the dry bundle. This means that the treatment of the bundles should be done of fresh felling residues. Drying of the bundles succeeded well, and the crushing and chipping tests showed that the processing of the bundles at the plant is possible. The treatability of the bundles was also excellent. By using the prototype, developed in the research, it was possible to produce a bundle of the fresh spruce harvesting residues, the diameter of which was about 50 cm and the length about 3 m, and the rate of compaction over 50 %. By these values the reduction target of the costs is obtainable

Bundling of harvesting residues and whole-trees and the treatment of bundles; Hakkuutaehteiden ja kokopuiden niputus ja nippujen kaesittely

Energy Technology Data Exchange (ETDEWEB)

Kaipainen, H.; Seppaenen, V.; Rinne, S.

1996-12-31

The conditions on which the bundling of the harvesting residues from spruce regeneration fellings would become profitable were studied. The calculations showed that one of the most important features was sufficient compaction of the bundle, so that the portion of the wood in the unit volume of the bundle has to be more than 40 %. The tests showed that the timber grab loader of farm tractor was insufficient for production of dense bundles. The feeding and compression device of the prototype bundler was constructed in the research and with this device the required density was obtained.The rate of compaction of the dry spruce felling residues was about 40 % and that of the fresh residues was more than 50 %. The comparison between the bundles showed that the calorific value of the fresh bundle per unit volume was nearly 30 % higher than that of the dry bundle. This means that the treatment of the bundles should be done of fresh felling residues. Drying of the bundles succeeded well, and the crushing and chipping tests showed that the processing of the bundles at the plant is possible. The treatability of the bundles was also excellent. By using the prototype, developed in the research, it was possible to produce a bundle of the fresh spruce harvesting residues, the diameter of which was about 50 cm and the length about 3 m, and the rate of compaction over 50 %. By these values the reduction target of the costs is obtainable

FLECHT low flooding rate skewed test series data report

International Nuclear Information System (INIS)

Rosal, E.R.; Conway, C.E.; Krepinevich, M.C.

1977-05-01

The FLECHT Low Flooding Rate Tests were conducted in an improved original FLECHT Test Facility to provide heat transfer coefficient and entrainment data at forced flooding rates of 1 in./sec. and with electrically heated rod bundles which had cosine and top skewed axial power profiles. The top-skewed axial power profile test series has now been successfully completed and is here reported. For these tests the rod bundle was enclosed in a low mass cylindrical housing which would minimize the wall housing effects encountered in the cosine test series. These tests examined the effects of initial clad temperature, variable stepped and continuously variable flooding rates, housing heat release, rod peak power, constant low flooding rates, coolant subcooling, hot and cold channel entrainment, and bundle stored and generated power. Data obtained in runs which met the test specifications are reported here, and include rod clad temperatures, turn around and quench times, heat transfer coefficients, inlet flooding rates, overall mass balances, differential pressures and calculated void fractions in the test section, thimble wall and steam temperatures, and exhaust steam and liquid carryover rates

Experimental and numerical investigations of BWR fuel bundle inlet flow

International Nuclear Information System (INIS)

Hoashi, E; Morooka, S; Ishitori, T; Komita, H; Endo, T; Honda, H; Yamamoto, T; Kato, T; Kawamura, S

2009-01-01

We have been studying the mechanism of the flow pattern near the fuel bundle inlet of BWR using both flow visualization test and computational fluid dynamics (CFD) simulation. In the visualization test, both single- and multi-bundle test sections were used. The former test section includes only a corner orifice facing two support beams and the latter simulates 16 bundles surrounded by four beams. An observation window is set on the side of the walls imitating the support beams upstream of the orifices in both test sections. In the CFD simulation, as well as the visualization test, the single-bundle model is composed of one bundle with a corner orifice and the multi-bundle model is a 1/4 cut of the test section that includes 4 bundles with the following four orifices: a corner orifice facing the corner of the two neighboring support beams, a center orifice at the opposite side from the corner orifice, and two side orifices. Twin-vortices were observed just upstream of the corner orifice in the multi-bundle test as well as the single-bundle test. A single-vortex and a vortex filament were observed at the side orifice inlet and no vortex was observed at the center orifice. These flow patterns were also predicted in the CFD simulation using Reynolds Stress Model as a turbulent model and the results were in good agreement with the test results mentioned above. (author)

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

International Nuclear Information System (INIS)

Sugawara, S.

1995-01-01

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

CANFLEX fuel bundle junction pressure drop

International Nuclear Information System (INIS)

Chung, H. J.; Chung, C. H.; Jun, J. S.; Hong, S. D.; Chang, S. K.; Kim, B. D.

1996-11-01

This report describes the junction pressure drop test results which are to used to determine the alignment angle between bundles to achieve the most probable fuel string pressure drop for randomly aligned bundles for use in the fuel string total pressure drop test. (author). 4 tabs., 17 figs

CANFLEX fuel bundle junction pressure drop

Energy Technology Data Exchange (ETDEWEB)

Chung, H. J.; Chung, C. H.; Jun, J. S.; Hong, S. D.; Chang, S. K.; Kim, B. D.

1996-11-01

This report describes the junction pressure drop test results which are to used to determine the alignment angle between bundles to achieve the most probable fuel string pressure drop for randomly aligned bundles for use in the fuel string total pressure drop test. (author). 4 tabs., 17 figs.

Boiling heat transfer on horizontal tube bundles

International Nuclear Information System (INIS)

Anon.

1987-01-01

Nucleate boiling heat transfer characteristics for a tube in a bundle differ from that for a single tube in a pool and this difference is known as 'tube bundle effect.' There exist two bundle effects, positive and negative. The positive bundle effect enhances heat transfer due to convective flow induced by rising bubbles generated from the lower tubes, while the negative bundle effect deteriorates heat transfer due to vapor blanketing caused by accumulation of bubbles. Staggered tube bundles tested and found that the upper tubes in bundles have higher heat transfer coefficients than the lower tubes. The effects of various parameters such as pressure, tube geometry and oil contamination on heat transfer have been examined. Some workers attempted to clarify the mechanism of occurrence of 'bundle effect' by testing tube arrangements of small scale. All reported only enhancement in heat transfer but results showed the symptom of heat transfer deterioration at higher heat fluxes. As mentioned above, it has not been clarified so far even whether the 'tube bundle effect' should serve as enhancement or deterioration of heat transfer in nucleate boiling. In this study, experiments are performed in detail by using bundles of small scale, and effects of heat flux distribution, pressure and tube location are clarified. Furthermore, some consideration on the mechanisms of occurrence of 'tube bundle effect' is made and a method for prediction of heat transfer rate is proposed

Quench propagation in High Temperature Superconducting materials integrated in high current leads

CERN Document Server

Milani, D

2001-01-01

High temperature superconductors (HTS) have been integrated in the high current leads for the Large Hadron Collider (LHC), under construction at CERN, in order to reduce the heat leak into the liquid helium bath due to the joule effect. The use of the HTS technology in the lower part of the current leads allowed to significantly reduce the heat charge on the cryogenic system. Hybrid current leads have been designed to fulfill the LHC requirements with respect to thermal load; several tests have been performed to study the lead behavior especially during a quench transient. Quench experiments have been performed at CERN on 13 kA prototypes to determine the adequate design and protection. In all the tests it is possible to know the temperature profile of the HTS only with the help of quench simulations that model the thermo-hydraulic processes during quench. The development of a theoretical model for the simulation allows reducing the number of test to perform and to scale the experimental result to other curre...

IFPE/MT4-MT6A-LOCA, Large-break LOCA in-reactor fuel bundle materials tests at NRU

International Nuclear Information System (INIS)

Cunningham, Mitchel E.; Turnbull, J.A.

2003-01-01

Description - Objectives - Results: The U.S. Nuclear Regulatory Commission (NRC) conducted a series of thermal-hydraulic and cladding mechanical deformation tests in the National Research Universal (NRU) reactor at the Chalk River National Laboratory in Canada. The objective of these tests was to perform simulated loss-of-coolant-accident (LOCA) experiments using full-length light-water reactor fuel rods to study mechanical deformation, flow blockage, and coolability. Three phases of a LOCA (i.e., heat-up, reflood, and quench) were performed in situ using nuclear fissioning to simulate the low-level decay power during a LOCA after shutdown. All tests used PWR-type, non-irradiated fuel rods. Provided here is information on two materials tests, MT-6A and MT-4, which PNNL considers the better characterized for the purposes of setting up computer cases. The NRU reactor is a heterogeneous, thermal, tank-type research reactor. It has a power level of 135 MWth and is heavy-water moderated and cooled. The coolant has an inlet temperature of 310 K at a pressure of 0.65 MPa. The MT tests were conducted in a specially designed test train to supply the specified coolant conditions of flowing steam, stagnant steam, and then reflood. Typical instrumentation for the MT tests included fuel centerline thermocouples, cladding inner surface thermocouples, cladding outer surface thermocouples, rod internal gas pressure transducers or pressure switches, coolant channel steam probes, and self-powered neutron detectors. This instrumentation allowed for determining rupture times and cladding temperature. The test rods for the LOCA cases in the NRU reactor were irradiated in flowing steam prior to the transient, stagnant steam during the transient and prior to reflood, and then reflood conditions to complete the transient. Both cladding inner surface and outer surface temperatures were measured, in addition to coolant temperatures. However, only cladding inner surface temperatures were

Effect of tempering temperature on microstructure and sliding wear property of laser quenched 4Cr13 steel

NARCIS (Netherlands)

Ouyang, J.H.; Pei, Y.T.; Li, X.D.; Lei, T.C.

1994-01-01

4Cr13 martensite stainless steel was quenched by a CO2 laser and tempered for 2 h at different temperatures in the range 200 °C to 550 °C. The microstructure of treated layer was observed by SEM, XRD and TEM. Tempering leads to the decomposition of a large number of retained austenites in laser

Porous debris behavior modeling of QUENCH-02, QUENCH-03 and QUENCH-09 experiments

International Nuclear Information System (INIS)

Kisselev, A.E.; Kobelev, G.V.; Strizhov, V.F.; Vasiliev, A.D.

2006-01-01

The heat-up, melting, relocation, hydrogen generation phenomena, relevant for high-temperature stages both in a reactor case and small-scale integral tests like QUENCH, are governed in particular by heat and mass transfer in porous debris and molten pools which are formed in the core region. Porous debris formation and behavior in QUENCH experiments (QUENCH-02, QUENCH-03, QUENCH-09) plays a considerable role and its adequate modeling is important for thermal analysis. In particular, the analysis of QUENCH experiments shows that the major hydrogen release takes place in debris and melt regions formed in the upper part of the fuel assembly. The porous debris model was implemented in the Russian best estimate numerical code RATEG/SVECHA/HEFEST developed for modelling thermal hydraulics and severe accident phenomena in a reactor. The original approach for debris evolution is developed in the model from classical principles using a set of parameters including debris porosity; average particle diameter; temperatures and mass fractions of solid, liquid and gas phases; specific interface areas between different phases; effective thermal conductivity of each phase, including radiative heat conductivity; mass and energy fluxes through the interfaces. The debris model is based on the system of continuity, momentum and energy conservation equations, which consider the dynamics of volume-averaged velocities and temperatures of fluid, solid and gaseous phases of porous debris. The model is used for calculation of QUENCH experiments. The results obtained by the model are compared to experimental data concerning different aspects of thermal behavior: thermal hydraulics of porous debris, radiative heat transfer in a porous medium, the generalized melting and refreezing behavior of materials, hydrogen production. (author)

Post-test investigation result on the WWER-1000 fuel tested under severe accident conditions

International Nuclear Information System (INIS)

Goryachev, A.; Shtuckert, Yu.; Zwir, E.; Stupina, L.

1996-01-01

The model bundle of WWER-type were tested under SFD condition in the out-of-pile CORA installation. The objective of the test was to provide an information on the WWER-type fuel bundles behaviour under severe fuel damage accident conditions. Also it was assumed to compare the WWER-type bundle damage mechanisms with these experienced in the PWR-type bundle tests with aim to confirm a possibility to use the various code systems, worked our for PWR as applied to WWER. In order to ensure the possibility of the comparison of the calculated core degradation parameters with the real state of the tested bundle, some parameters have been measured on the bundle cross-sections under examination. Quantitative parameters of the bundle degradation have been evaluated by digital image processing of the bundle cross-sections. The obtained results are shown together with corresponding results obtained by the other participants of this investigation. (author). 3 refs, 13 figs

Annular burnout data from rod-bundle experiments

International Nuclear Information System (INIS)

Yoder, G.L.; Morris, D.G.; Mullins, C.B.

1983-01-01

Burnout data for annular flow in a rod bundle are presented for both transient and steady-state conditions. Tests were performed at the Oak Ridge National Laboratory in the Thermal Hydraulic Test Facility (THTF), a pressurized-water loop containing an electrically heated 64-rod bundle. The bundle configuration is typical of later generation pressurized-water reactors with 17 x 17 fuel arrays. Both axial and radial power profiles are flat. All experiments were carried out in upflow with subcooled inlet conditions, insuring accurate flow measurement. Conditions within the bundle were typical of those which could be encountered during a nuclear reactor loss-of-coolant accident

Comparison of computer codes (CE-THERM, FRAP-T5, GT3-FLECHT, and TRUMP-FLECHT) with data from the NRU-LOCA thermal hydraulic tests

International Nuclear Information System (INIS)

Mohr, C.L.; Rausch, W.N.; Hesson, G.M.

1981-07-01

The LOCA Simulation Program in the NRU reactor is the first set of experiments to provide data on the behavior of full-length, nuclear-heated PWR fuel bundles during the heatup, reflood, and quench phases of a loss-of-coolant accident (LOCA). This paper compares the temperature time histories of 4 experimental test cases with 4 computer codes: CE-THERM, FRAP-T5, GT3-FLECHT, and TRUMP-FLECHT. The preliminary comparisons between prediction and experiment show that the state-of-the art fuel codes have large uncertainties and are not necessarily conservative in predicting peak temperatures, turn around times, and bundle quench times

Quench and safety tests on a toroidal field coil of Tore Supra

International Nuclear Information System (INIS)

Ciazynski, D.; Cure, C.; Duchateau, J.L.

1987-01-01

As a part of the safety analysis of the magnet, three quenches have been initiated in one of the TF coils in the Saclay test facility. While transporting a given current, the coil is insulated from the refrigerator: the temperatures of the helium and of the coil increase slowly on account of thermal losses. At the current sharing temperature a quench rapidly propagates and the protection system makes the coil discharge in the dump resistor. At three levels of current, electrical, thermal and hydraulic measurements have been performed. All these results are taken into account for the safety design of TORE SUPRA

Which test for CAD should be used in patients with left bundle branch block?

Science.gov (United States)

Xu, Bo; Cremer, Paul; Jaber, Wael; Moir, Stuart; Harb, Serge C; Rodriguez, L Leonardo

2018-03-01

Exercise stress electrocardiography is unreliable as a test for obstructive coronary artery disease (CAD) if the patient has left bundle branch block. The authors provide an algorithm for using alternative tests: exercise stress echocardiography, dobutamine echocardiography, computed tomographic (CT) angiography, and nuclear myocardial perfusion imaging. Copyright © 2018 Cleveland Clinic.

An in vitro biomechanical comparison of anterior cruciate ligament reconstruction: single bundle versus anatomical double bundle techniques

Directory of Open Access Journals (Sweden)

Sandra Umeda Sasaki

2008-01-01

Full Text Available INTRODUCTION: Anterior cruciate ligament ruptures are frequent, especially in sports. Surgical reconstruction with autologous grafts is widely employed in the international literature. Controversies remain with respect to technique variations as continuous research for improvement takes place. One of these variations is the anatomical double bundle technique, which is performed instead of the conventional single bundle technique. More recently, there has been a tendency towards positioning the two bundles through double bone tunnels in the femur and tibia (anatomical reconstruction. OBJECTIVES: To compare, through biomechanical tests, the practice of anatomical double bundle anterior cruciate ligament reconstruction with a patellar graft to conventional single bundle reconstruction with the same amount of patellar graft in a paired experimental cadaver study. METHODS: Nine pairs of male cadaver knees ranging in age from 44 to 63 years were randomized into two groups: group A (single bundle and group B (anatomical reconstruction. Each knee was biomechanically tested under three conditions: intact anterior cruciate ligament, reconstructed anterior cruciate ligament, and injured anterior cruciate ligament. Maximum anterior dislocation, rigidity, and passive internal tibia rotation were recorded with knees submitted to a 100 N horizontal anterior dislocation force applied to the tibia with the knees at 30, 60 and 90 degrees of flexion. RESULTS: There were no differences between the two techniques for any of the measurements by ANOVA tests. CONCLUSION: The technique of anatomical double bundle reconstruction of the anterior cruciate ligament with bone-patellar tendon-bone graft has a similar biomechanical behavior with regard to anterior tibial dislocation, rigidity, and passive internal tibial rotation.

Design report for an annular fuel element for accommodation of a carbide test bundle on the ring position of the KNK II/2 test zone

International Nuclear Information System (INIS)

Haefner, H.E.

1982-03-01

This report describes an annular oxide element with Mark II rods for accommodation of a 19-pin carbide test bundle on position 201 in the test zone of the second core of KNK II as well as its behavior during the period of operation. The ring element comprises within a driver wrapper in three rows of pins 102 fuel pins of 7.6 mm diameter and six structural rods for fixing the spark eroded spacers. The report deals with the ring element with its individual components fuel rod, bundle, wrappers, head and foot and describes methods, criteria and results concerning the design. The carbide test bundle to be accommodated by the annular carrier element will be treated in a separate report. The loadability of the annular element with its components is demonstrated by generally valid standards for strength criteria

Quark contributions to baryon magnetic moments in full, quenched, and partially quenched QCD

International Nuclear Information System (INIS)

Leinweber, Derek B.

2004-01-01

The chiral nonanalytic behavior of quark-flavor contributions to the magnetic moments of octet baryons is determined in full, quenched and partially quenched QCD, using an intuitive and efficient diagrammatic formulation of quenched and partially quenched chiral perturbation theory. The technique provides a separation of quark-sector magnetic-moment contributions into direct sea-quark loop, valence-quark, indirect sea-quark loop and quenched valence contributions, the latter being the conventional view of the quenched approximation. Both meson and baryon mass violations of SU(3)-flavor symmetry are accounted for. Following a comprehensive examination of the individual quark-sector contributions to octet baryon magnetic moments, numerous opportunities to observe and test the underlying structure of baryons and the nature of chiral nonanalytic behavior in QCD and its quenched variants are discussed. In particular, the valence u-quark contribution to the proton magnetic moment provides the optimal opportunity to directly view nonanalytic behavior associated with the meson cloud of full QCD and the quenched meson cloud of quenched QCD. The u quark in Σ + provides the best opportunity to display the artifacts of the quenched approximation

Bundling harvester; Nippukorjausharvesteri

Energy Technology Data Exchange (ETDEWEB)

Koponen, K [Eko-Log Oy, Kuopio (Finland)

1997-12-31

The staring point of the project was to design and construct, by taking the silvicultural point of view into account, a harvesting and processing system especially for energy-wood, containing manually driven bundling harvester, automatizing of the harvester, and automatized loading. The equipment forms an ideal method for entrepreneur`s-line harvesting. The target is to apply the system also for owner`s-line harvesting. The profitability of the system promotes the utilization of the system in both cases. The objectives of the project were: to construct a test equipment and prototypes for all the project stages, to carry out terrain and strain tests in order to examine the usability and durability, as well as the capacity of the machine, to test the applicability of the Eko-Log system in simultaneous harvesting of energy and pulp woods, and to start the marketing and manufacturing of the products. The basic problems of the construction of the bundling harvester have been solved using terrain-tests. The prototype machine has been shown to be operable. Loading of the bundles to form sufficiently economically transportable loads has been studied, and simultaneously, the branch-biomass has been tried to be utilized without loosing the profitability of transportation. The results have been promising, and will promote the profitable utilization of wood-energy

Bundling harvester; Nippukorjausharvesteri

Energy Technology Data Exchange (ETDEWEB)

Koponen, K. [Eko-Log Oy, Kuopio (Finland)

1996-12-31

The staring point of the project was to design and construct, by taking the silvicultural point of view into account, a harvesting and processing system especially for energy-wood, containing manually driven bundling harvester, automatizing of the harvester, and automatized loading. The equipment forms an ideal method for entrepreneur`s-line harvesting. The target is to apply the system also for owner`s-line harvesting. The profitability of the system promotes the utilization of the system in both cases. The objectives of the project were: to construct a test equipment and prototypes for all the project stages, to carry out terrain and strain tests in order to examine the usability and durability, as well as the capacity of the machine, to test the applicability of the Eko-Log system in simultaneous harvesting of energy and pulp woods, and to start the marketing and manufacturing of the products. The basic problems of the construction of the bundling harvester have been solved using terrain-tests. The prototype machine has been shown to be operable. Loading of the bundles to form sufficiently economically transportable loads has been studied, and simultaneously, the branch-biomass has been tried to be utilized without loosing the profitability of transportation. The results have been promising, and will promote the profitable utilization of wood-energy

Thermal neutron measurement using the instrumented test bundle and assessment of maximum linear power in HANARO

Energy Technology Data Exchange (ETDEWEB)

Lee, C. S.; Seo, C. K.; Lee, B. C.; Kim, H. N.; Kang, B. W. [KAERI, Taejon (Korea, Republic of)

2000-10-01

The HANARO fuel, U{sub 3}Si-Al, has been developed by AECL and tested in NRU reactor. Due to the lack of the data performed under the high power, the repetitive conduct of the irradiation test was required under the power greater than 108kW/m, which is the estimated maximum linear power in the design stage. Accordingly, the instrumented test bundle with SPND(Self Powered Neutron Detector) was fabricated and its irradiation test was performed in IR2 of HANARO. The measured thermal neutron flux with SPND is compared with calculation results by HANAFMS(HANARO Fuel Management System). The difference in the measured and calculated thermal flux values are below {+-}11% and the accuracy of the linear power predicted by HANAFMS is consequently accompanied. Therefore, it is believed that the maximum linear power above 120kW/m is achieved during the irradiation test of the test bundle.

Bundled payment fails to gain a foothold In California: the experience of the IHA bundled payment demonstration.

Science.gov (United States)

Ridgely, M Susan; de Vries, David; Bozic, Kevin J; Hussey, Peter S

2014-08-01

To determine whether bundled payment could be an effective payment model for California, the Integrated Healthcare Association convened a group of stakeholders (health plans, hospitals, ambulatory surgery centers, physician organizations, and vendors) to develop, through a consensus process, the methods and means of implementing bundled payment. In spite of a high level of enthusiasm and effort, the pilot did not succeed in its goal to implement bundled payment for orthopedic procedures across multiple payers and hospital-physician partners. An evaluation of the pilot documented a number of barriers, such as administrative burden, state regulatory uncertainty, and disagreements about bundle definition and assumption of risk. Ultimately, few contracts were signed, which resulted in insufficient volume to test hypotheses about the impact of bundled payment on quality and costs. Although bundled payment failed to gain a foothold in California, the evaluation provides lessons for future bundled payment initiatives. Project HOPE—The People-to-People Health Foundation, Inc.

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

International Nuclear Information System (INIS)

Hassan, Y.A.

1987-01-01

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

Computation of 3D thermohydraulics in partially blocked bundles during the reflood phase of a LOCA

International Nuclear Information System (INIS)

Cicero, G.M.; Briere, E.; Fornaciari, G.

1994-06-01

In Pressurized Water Reactors (PWR), ballooning of the fuel rod claddings may occur during a LOCA, since the fuel rod claddings are heated up, and the system pressure is low. The severe blockages that may result induce cross-flow diversion and three-dimensional effects on thermohydraulics in the core bundle, during the reflood phase. To improve the knowledge of these phenomena and their physical modelling in the code CATHARE, 3D computer codes are needed. In 1990, EDF has started up a development and validation program of the 3D THYC computer code to analyze the thermohydraulics of the flow during the reflood phase, in partially blocked bundles. The main objective is to calculate the temperatures of the rods above the quench front, when they are cooled by superheated steam with saturated droplets. First, this paper introduces the THYC model developed for reflood studies. Secondly, we report the first qualification results on a Flooding Experiments with Blocked Array (FEBA) test. Thirdly, we analyze the model predictions on a large break LOCA transient, in a 900 MW PWR 11x11 core area with a 3x3 central blockage. THYC simulates the transient in the bundle around and above the blockage, until the quench front enters the computational domain. Previously, a 1D CATHARE simulation gives the boundary conditions and, in the reactor core case, the deformation of the blocked fuel rods. The results analysis focused on the time evolution of the clad temperatures in the blocked and in the bypass region. In the FEBA test simulation, the main observations are properly predicted within the blockage. Temperatures are lower in blocked rod sleeves than in unblocked rod claddings since the steam gap reduces the power transmitted by the heater rod to the sleeve. In the core case, the model predicts the opposite result. Within the blockage, ballooned rod temperatures are higher than non-ballooned rod ones. We show by sensitivity studies that these behaviour difference between FEBA rods

Laboratory manual for salt mixing test in rod bundles

International Nuclear Information System (INIS)

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

1978-10-01

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

Tokyo Guidelines 2018: management bundles for acute cholangitis and cholecystitis

NARCIS (Netherlands)

Mayumi, Toshihiko; Okamoto, Kohji; Takada, Tadahiro; Strasberg, Steven M.; Solomkin, Joseph S.; Schlossberg, David; Pitt, Henry A.; Yoshida, Masahiro; Gomi, Harumi; Miura, Fumihiko; Garden, O. James; Kiriyama, Seiki; Yokoe, Masamichi; Endo, Itaru; Asbun, Horacio J.; Iwashita, Yukio; Hibi, Taizo; Umezawa, Akiko; Suzuki, Kenji; Itoi, Takao; Hata, Jiro; Han, Ho-Seong; Hwang, Tsann-Long; Dervenis, Christos; Asai, Koji; Mori, Yasuhisa; Huang, Wayne Shih-Wei; Belli, Giulio; Mukai, Shuntaro; Jagannath, Palepu; Cherqui, Daniel; Kozaka, Kazuto; Baron, Todd H.; de Santibañes, Eduardo; Higuchi, Ryota; Wada, Keita; Gouma, Dirk J.; Deziel, Daniel J.; Liau, Kui-Hin; Wakabayashi, Go; Padbury, Robert; Jonas, Eduard; Supe, Avinash Nivritti; Singh, Harjit; Gabata, Toshifumi; Chan, Angus C. W.; Lau, Wan Yee; Fan, Sheung Tat; Chen, Miin-Fu; Ker, Chen-Guo; Yoon, Yoo-Seok; Choi, In-Seok; Kim, Myung-Hwan; Yoon, Dong-Sup; Kitano, Seigo; Inomata, Masafumi; Hirata, Koichi; Inui, Kazuo; Sumiyama, Yoshinobu; Yamamoto, Masakazu

2018-01-01

Management bundles that define items or procedures strongly recommended in clinical practice have been used in many guidelines in recent years. Application of these bundles facilitates the adaptation of guidelines and helps improve the prognosis of target diseases. In Tokyo Guidelines 2013 (TG13),

A subchannel and CFD analysis of void distribution for the BWR fuel bundle test benchmark

International Nuclear Information System (INIS)

In, Wang-Kee; Hwang, Dae-Hyun; Jeong, Jae Jun

2013-01-01

Highlights: ► We analyzed subchannel void distributions using subchannel, system and CFD codes. ► The mean error and standard deviation at steady states were compared. ► The deviation of the CFD simulation was greater than those of the others. ► The large deviation of the CFD prediction is due to interface model uncertainties. -- Abstract: The subchannel grade and microscopic void distributions in the NUPEC (Nuclear Power Engineering Corporation) BFBT (BWR Full-Size Fine-Mesh Bundle Tests) facility have been evaluated with a subchannel analysis code MATRA, a system code MARS and a CFD code CFX-10. Sixteen test series from five different test bundles were selected for the analysis of the steady-state subchannel void distributions. Four test cases for a high burn-up 8 × 8 fuel bundle with a single water rod were simulated using CFX-10 for the microscopic void distribution benchmark. Two transient cases, a turbine trip without a bypass as a typical power transient and a re-circulation pump trip as a flow transient, were also chosen for this analysis. It was found that the steady-state void distributions calculated by both the MATRA and MARS codes coincided well with the measured data in the range of thermodynamic qualities from 5 to 25%. The results of the transient calculations were also similar to each other and very reasonable. The CFD simulation reproduced the overall radial void distribution trend which produces less vapor in the central part of the bundle and more vapor in the periphery. However, the predicted variation of the void distribution inside the subchannels is small, while the measured one is large showing a very high concentration in the center of the subchannels. The variations of the void distribution between the center of the subchannels and the subchannel gap are estimated to be about 5–10% for the CFD prediction and more than 20% for the experiment

Fuel bundle examination techniques for the Phebus fission product test

International Nuclear Information System (INIS)

Blanc, J.Y.; Clement, B.; Hardt, P. von der

1996-01-01

The paper develops the non-destructive examinations, with a special emphasis on transmission tomography, performed in the Phebus facility, using a linear accelerator associated with a line scan camera based on PCD components. This particular technique enabled the high level of penetration to be obtained, necessary for this high density application. Spatial resolution is not far from the theoretical limit and the density resolution is often adequate. This technique permitted: 1) to define beforehand the cuts on a precise basis, avoiding a long step-by-step choice as in previous in-pile tests; 2) to determine, at an early stage, mass balance, material relocations (in association with axial gamma spectrometry), and FP distribution, as an input into re-calculations of the bundle events. However, classical cuttings, periscopic visual examinations, macrographies, micrographies and EPMA analyses remain essential to give oxidation levels (in the less degraded zones), phase aspect and composition, to distinguish between materials of identical density, and, if possible, to estimate temperatures. Oxidation resistance of sensors (thermocouples or ultrasonic thermometers) is also traced. The EPMA gives access to the molten material chemical analyses, especially in the molten fuel blockage area. The first results show that an important part of the fuel bundle melted (which was one of the objectives of this test) and that the degradation level is close to TIMI-2 with a molten plug under a cavity surrounded by an uranium-rich crust. In lower and upper areas fuel rods are less damaged. Complementaries between these examination techniques and between international teams involved will be major advantages in the Phebus FPT0 test comprehension. 3 refs, 9 figs

Design and operation of the quench protection system for the Fermilab tevatron

International Nuclear Information System (INIS)

Martin, P.S.

1989-01-01

The operation of a superconducting accelerator, in addition to cryogenic requirements, introduces a new complexity not present in a conventional accelerator. A method is required for protecting the magnets from possible overheating or overvoltage conditions in the event that some magnets quench, that is, are elevated in temperature so that they are no longer superconducting. The development of that system is the topic of this chapter. Any quench protection system has two very important ingredients. First, it must be designed with sufficient integrity to remain functional even under abnormal circumstances. The magnets must be protected during power failures, for example. Quenches involving a large number of components can also be hazardous because of the redistribution of voltages during the quench. Some of the system integrity can be achieved through redundancy. Frequent testing of critical elements of the system also assures the overall integrity. Second, the quench protection system must protect against damage from quenches regardless of their location or the excitation current at the time. It is not sufficient to protect just the magnet coils; either the leads between magnets must be fully stabilized or the quench protection system must protect them. The next section presents a brief discussion of the basic properties of superconductors and the phenomenon of quench propagation. 10 references, 13 figures

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

International Nuclear Information System (INIS)

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

2009-01-01

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

Quenches in the superconducting magnet CELLO

International Nuclear Information System (INIS)

Hassenzahl, W.V.

1979-01-01

The superconducting magnet CELLO was tested with currents up to 3200 A at Saclay and has been installed at DESY in Hamburg where it will be used for particle physics experiments requiring colliding beams of electrons and positrons. The testing of this unique, large, one-layer solenoid provides an excellent opportunity to evaluate the theory of quench propagation under adiabatic conditions, that is, in a coil in which the conductors are not in direct contact with helium. In an early test of this coil, quenches that occurred, gives the details of the damaged conductor, and includes an analysis of the quenches. Observed axial quench velocities are compared to the calculated values based on both measurements and calculations of the thermal conductivity of the fabricated coil

Characteristics of CANDU fuel bundles that caused pressure tube fretting at the bundle midplane

Energy Technology Data Exchange (ETDEWEB)

Dennier, D; Manzer, A M [Atomic Energy of Canada Ltd., Mississauga, ON (Canada); Koehn, E [Ontario Hydro, Toronto, ON (Canada)

1996-12-31

Detailed measurements on new bundles, and those that caused fretting during in- and out-reactor tests, have given insight into the factors responsible for fretting at the midplane of the inlet bundle. Bottom fuel elements that were attached near radial endplate spokes and had inboard bearing pads in the rolled joint cavity produced a significant portion of the observed fret marks. These elements are influenced by several driving forces that deflect the centre bearing pads towards the pressure tube surface. The evidence suggests that slight changes in bundle design may be possible to reduce pressure tube fretting. (author). 4 refs., 3 tabs., 8 figs.

Quench Tests of LHC Magnets with Beam: Studies on Beam Loss development and determination of Quench levels

CERN Document Server

Priebe, A; Sapinski, M

The application of superconducting materials in the field of high energy accelerator physics not only opens the doors to the generation of the magnetic fields unattainable to normal conductors but also demands facing new challenges. A transition fromthe superconducting state, which is characterized by a resistance-free flow of the electric current, to the normal conducting state is called quenching. This process might be extremely dangerous and even lead to destruction of amagnet superconducting coil if no protecting actions are taken. Therefore, the knowledge of a magnet quench level, i.e. amount of energy which causes the transition to the resistive state, is crucial for the safety and operational efficiency of the accelerator. Regarding that, specific thresholds are incorporated to dedicated quench prevention systems in order to suppress the origin of detected energy perturbation, for example beam losses, or mitigate the consequences of the quenching process by dissipating the energy stored in the magnetic...

Fluid-Elastic Instability Tests on Parallel Triangular Tube Bundles with Different Mass Ratio Values under Increasing and Decreasing Flow Velocities

Directory of Open Access Journals (Sweden)

Xu Zhang

2016-01-01

Full Text Available To study the effects of increasing and decreasing flow velocities on the fluid-elastic instability of tube bundles, the responses of an elastically mounted tube in a rigid parallel triangular tube bundle with a pitch-to-diameter ratio of 1.67 were tested in a water tunnel subjected to crossflow. Aluminum and stainless steel tubes were tested, respectively. In the in-line and transverse directions, the amplitudes, power spectrum density functions, response frequencies, added mass coefficients, and other results were obtained and compared. Results show that the nonlinear hysteresis phenomenon occurred in both tube bundle vibrations. When the flow velocity is decreasing, the tubes which have been in the state of fluid-elastic instability can keep on this state for a certain flow velocity range. During this process, the response frequencies of the tubes will decrease. Furthermore, the response frequencies of the aluminum tube can decrease much more than those of the stainless steel tube. The fluid-elastic instability constants fitted for these experiments were obtained from experimental data. A deeper insight into the fluid-elastic instability of tube bundles was also obtained by synthesizing the results. This study is beneficial for designing and operating equipment with tube bundles inside, as well as for further research on the fluid-elastic instability of tube bundles.

Hydraulic characteristics of HANARO fuel bundles

Energy Technology Data Exchange (ETDEWEB)

Cho, S; Chung, H J; Chun, S Y; Yang, S K; Chung, M K [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1998-12-31

This paper presents the hydraulic characteristics measured by using LDV (Laser Doppler Velocimetry) in subchannels of HANARO, KAERI research reactor, fuel bundle. The fuel bundle consists of 18 axially finned rods with 3 spacer grids, which are arranged in cylindrical configuration. The effects of the spacer grids on the turbulent flow were investigated by the experimental results. Pressure drops for each component of the fuel bundle were measured, and the friction factors of fuel bundle and loss coefficients for the spacer grids were estimated from the measured pressure drops. Implications regarding the turbulent thermal mixing were discussed. Vibration test results measured by using laser vibrometer were presented. 9 refs., 12 figs. (Author)

Hydraulic characteristics of HANARO fuel bundles

Energy Technology Data Exchange (ETDEWEB)

Cho, S.; Chung, H. J.; Chun, S. Y.; Yang, S. K.; Chung, M. K. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1997-12-31

This paper presents the hydraulic characteristics measured by using LDV (Laser Doppler Velocimetry) in subchannels of HANARO, KAERI research reactor, fuel bundle. The fuel bundle consists of 18 axially finned rods with 3 spacer grids, which are arranged in cylindrical configuration. The effects of the spacer grids on the turbulent flow were investigated by the experimental results. Pressure drops for each component of the fuel bundle were measured, and the friction factors of fuel bundle and loss coefficients for the spacer grids were estimated from the measured pressure drops. Implications regarding the turbulent thermal mixing were discussed. Vibration test results measured by using laser vibrometer were presented. 9 refs., 12 figs. (Author)

Design of electronic measurement and quench detection equipment for the Current Lead Test facility Karlsruhe (CuLTKa)

International Nuclear Information System (INIS)

Hollik, Markus; Fietz, Walter H.; Fink, Stefan; Gehrlein, Mirko; Heller, Reinhard; Lange, Christian; Möhring, Tobias

2013-01-01

The Current Lead Test facility Karlsruhe (CuLTKa) is under construction at the Karlsruhe Institute of Technology (KIT) to perform acceptance tests of high temperature superconductor (HTS) current leads (CL). CuLTKa is in progress and present planning expects the completion in 2013. The data acquisition system is based on a modular design with electronic measurement and monitoring equipment covering a test voltage of 50 kV DC against ground. It provides plug-in units which enable temperature and voltage measurement at high voltage potential and in addition quench detection units which detect a loss of superconductivity reliably and quickly to avoid damage of the superconducting device under test. Prototype units for quench detection, temperature and voltage measurement have been successfully tested. Six temperature measurement units are already in use in the KIT test facility TOSKA and operated reliably during the acceptance tests of the HTS current leads for Wendelstein 7-X (W7-X) in 2011/2012. CuLTKa will be used first for 26 current leads which will be built in KIT for the fusion experiment JT-60SA. The present paper gives an overview of the design of the electronic measurement and quench detection equipment

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

International Nuclear Information System (INIS)

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

1983-12-01

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

Simulation of quenches in SSC magnets with passive quench protection

International Nuclear Information System (INIS)

Koepke, K.

1985-06-01

The relative ease of protecting an SSC magnet following a quench and the implications of quench protection on magnet reliability and operation are necessary inputs in a rational magnet selection process. As it appears likely that the magnet selection will be made prior to full scale prototype testing, an alternative means is required to ascertain the surviveability of contending magnet types. This paper attempts to provide a basis for magnet selection by calculating the peak expected quench temperatures in the 3 T Design C magnet and the 6 T Design D magnet as a function of magnet length. A passive, ''cold diode'' protection system has been assumed. The relative merits of passive versus active protection systems have been discussed in a previous report. It is therefore assumed that - given the experience gained from the Tevatron system - that an active quench protection system can be employed to protect the magnets in the eventuality of unreliable cold diode function

CFD modeling of secondary flows in fuel rod bundles

International Nuclear Information System (INIS)

Baglietto, Emilio; Ninokata, Hisashi

2004-01-01

An optimized non-linear eddy viscosity model is introduced, for calculations of detailed coolant velocity distribution in a tight lattice fuel bundle. The low Reynolds formulation has been optimized based on DNS data for channel flow. The non-linear stress-strain relationship has been modified in the coefficients to model the flow anisotropy, which causes the formation of turbulence driven secondary flows inside the bundle subchannels. Predictions of the model are first compared to experimental measurements of secondary flows in a triangularly arrayed rod bundle with p/d=1.3. Subsequently wall shear stress and velocity predictions are compared with different experimental data for a rod bundle with p/d=1.17. The model shows to be able to correctly reproduce the scale of the secondary motion, and to accurately reproduce both wall shear stress and velocity distributions inside the rod bundle subchannels. (author)

40 CFR 1065.675 - CLD quench verification calculations.

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 32 2010-07-01 2010-07-01 false CLD quench verification calculations... POLLUTION CONTROLS ENGINE-TESTING PROCEDURES Calculations and Data Requirements § 1065.675 CLD quench verification calculations. Perform CLD quench-check calculations as follows: (a) Perform a CLD analyzer quench...

Higher order jet prolongations type gauge natural bundles over vector bundles

Directory of Open Access Journals (Sweden)

Jan Kurek

2004-05-01

Full Text Available Let $rgeq 3$ and $mgeq 2$ be natural numbers and $E$ be a vector bundle with $m$-dimensional basis. We find all gauge natural bundles ``similar" to the $r$-jet prolongation bundle $J^rE$ of $E$. We also find all gauge natural bundles ``similar" to the vector $r$-tangent bundle $(J^r_{fl}(E,R_0^*$ of $E$.

Investigation of the Phebus FPT0 bundle degradation with SCDAP/RELAP5

International Nuclear Information System (INIS)

Smit, S.O.; Sengpiel, W.; Hering, W.

1998-04-01

The in-pile experiment Phebus FPT0 provides an excellent data base reflecting the course and the consequences of a severe core melt accident starting from the core uncovery up to bundle degradation and molten pool formation. In the IRS post-test calculations of the Phebus FPT0 have been performed with SCDAP/RELAP5. A detailed parameter study has shown that some models used in the code still have to be improved and that some parametric models need to be substituted by more physical models. In the context of this parameter study, the heat transfer through the Phebus FPT0 shroud has been identified to be one of the most influential physical processes on the course of bundle degradation. Especially the gap behaviour and the heat transport through the gaps of the FPT0 shroud have shown to be insufficiently modeled by the original code version. Therefore, the shroud heat transfer model has been improved to consider dynamic gap closure by thermal expansion of the shroud materials and to take into account radiation heat transfer through open gaps. In this report, the results of the parameter study for FPT0 obtained with the original code are compared to the results of a reference calculation which includes the improved shroud model. It is shown that SCDAP/RELAP5 is now able to calculate the heat losses through a shroud containing gas-filled gaps like that of Phebus FPT0 quite accurately. Thus, SCDAP/RELAP5 now can also be used more successfully for test analyses of experiments like Phebus FPT1 and FPT2, and of the QUENCH test series. (orig./MM) [de

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

Application of Best Estimate Approach for Modelling of QUENCH-03 and QUENCH-06 Experiments

Directory of Open Access Journals (Sweden)

Tadas Kaliatka

2016-04-01

In this article, the QUENCH-03 and QUENCH-06 experiments are modelled using ASTEC and RELAP/SCDAPSIM codes. For the uncertainty and sensitivity analysis, SUSA3.5 and SUNSET tools were used. The article demonstrates that applying the best estimate approach, it is possible to develop basic QUENCH input deck and to develop the two sets of input parameters, covering maximal and minimal ranges of uncertainties. These allow simulating different (but with the same nature tests, receiving calculation results with the evaluated range of uncertainties.

Development and pilot testing of a patient-participatory pressure ulcer prevention care bundle.

Science.gov (United States)

Gillespie, Brigid M; Chaboyer, Wendy; Sykes, Mark; O'Brien, Jennifer; Brandis, Susan

2014-01-01

This study developed and piloted a patient-centered pressure ulcer prevention care bundle for adult hospitalized patients to promote patient participation in prevention. The care bundle had 3 core messages: (1) keep moving, (2) care for your skin, and (3) ensure a good diet. A brief video, combined brochure/checklist, and poster were developed as training resources. Patient evaluation identified benefits of the care bundle; however, the combined checklist/brochure was rarely used.

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)

International Benchmark on Pressurised Water Reactor Sub-channel and Bundle Tests. Volume II: Benchmark Results of Phase I: Void Distribution

International Nuclear Information System (INIS)

Rubin, Adam; Avramova, Maria; Velazquez-Lozada, Alexander

2016-03-01

This report summarised the first phase of the Nuclear Energy Agency (NEA) and the US Nuclear Regulatory Commission Benchmark based on NUPEC PWR Sub-channel and Bundle Tests (PSBT), which was intended to provide data for the verification of void distribution models in participants' codes. This phase was composed of four exercises; Exercise 1: steady-state single sub-channel benchmark, Exercise 2: steady-state rod bundle benchmark, Exercise 3: transient rod bundle benchmark and Exercise 4: a pressure drop benchmark. The experimental data provided to the participants of this benchmark is from a series of void measurement tests using full-size mock-up tests for both Boiling Water Reactors (BWRs) and Pressurised Water Reactors (PWRs). These tests were performed from 1987 to 1995 by the Nuclear Power Engineering Corporation (NUPEC) in Japan and made available by the Japan Nuclear Energy Safety Organisation (JNES) for the purposes of this benchmark, which was organised by Pennsylvania State University. Twenty-one institutions from nine countries participated in this benchmark. Seventeen different computer codes were used in Exercises 1, 2, 3 and 4. Among the computer codes were porous media, sub-channel, systems thermal-hydraulic code and Computational Fluid Dynamics (CFD) codes. It was observed that the codes tended to overpredict the thermal equilibrium quality at lower elevations and under predict it at higher elevations. There was also a tendency to overpredict void fraction at lower elevations and underpredict it at high elevations for the bundle test cases. The overprediction of void fraction at low elevations is likely caused by the x-ray densitometer measurement method used. Under sub-cooled boiling conditions, the voids accumulate at heated surfaces (and are therefore not seen in the centre of the sub-channel, where the measurements are being taken), so the experimentally-determined void fractions will be lower than the actual void fraction. Some of the best

Critical power characteristics in 37-rod tight lattice bundles under transient conditions

International Nuclear Information System (INIS)

Liu, Wei; Kureta, Masatoshi; Tamai, Hidesada; Ohnuki, Akira; Akimoto, Hajime

2007-01-01

Critical power characteristics in the postulated abnormal transient processes that may be possibly met in the operation of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) were investigated for the design of the FLWR core. Transient Boiling Transition (BT) tests were carried out using two sets of 37-rod tight lattice rod bundles (rod diameter: 13 mm; rod clearance: 1.3 mm or 1.0 mm) at Japan Atomic Energy Agency (JAEA) under the conditions covering the FLWR operating condition (P ex =7.2 MPa, T in =556 K) for mass velocity G=400-800 kg/(m 2 s). For the postulated power increase and flow decrease transients, no obvious change of the critical power against the steady one was observed. The traditional quasi-steady characteristic was confirmed to be working for the postulated power increase and flow decrease transients. The experiments were analyzed with TRAC-BF1 code, where the JAEA newest critical power correlation for the tight lattice rod bundles was implemented for the BT judgment. The TRAC-BF1 code showed good prediction for the occurrence or the non occurrence of the BT and for the exact BT starting time. The tranditional quasi-steady state prediction of the BT in transient process was confirmed to be applicable for the postulated abnormal transient processes in the tight lattice rod bundles. (author)

intercritical heat treatments effects on low carbon steels quenched

African Journals Online (AJOL)

DR B. A. EZEKOYE

Department of Physics and Astronomy, University of Nigeria, Nsukka. 2. E-mail: benjamin.ezekoye@unn.edu.ng; bezekoye@yahoo.com. ABSTRACT. Six low carbon steels containing carbon in the range 0.13-0.18wt%C were studied after intercritical quenching, intercritical quenching with low temperature tempering, ...

Hydrated and Dehydrated Tertiary Interactions–Opening and Closing–of a Four-Helix Bundle Peptide

Science.gov (United States)

Lignell, Martin; Tegler, Lotta T.; Becker, Hans-Christian

2009-01-01

Abstract The structural heterogeneity and thermal denaturation of a dansyl-labeled four-helix bundle homodimeric peptide was studied with steady-state and time-resolved fluorescence spectroscopy and with circular dichroism (CD). At room temperature the fluorescence decay of the polarity-sensitive dansyl, located in the hydrophobic core region, can be described by a broad distribution of fluorescence lifetimes, reflecting the heterogeneous microenvironment. However, the lifetime distribution is nearly bimodal, which we ascribe to the presence of two major conformational subgroups. Since the fluorescence lifetime reflects the water content of the four-helix bundle conformations, we can use the lifetime analysis to monitor the change in hydration state of the hydrophobic core of the four-helix bundle. Increasing the temperature from 9°C to 23°C leads to an increased population of molten-globule-like conformations with a less ordered helical backbone structure. The fluorescence emission maximum remains constant in this temperature interval, and the hydrophobic core is not strongly affected. Above 30°C the structural dynamics involve transient openings of the four-helix bundle structure, as evidenced by the emergence of a water-quenched component and less negative CD. Above 60°C the homodimer starts to dissociate, as shown by the increasing loss of CD and narrow, short-lived fluorescence lifetime distributions. PMID:19619472

Development of nuclear fuel. Development of CANDU advanced fuel bundle

International Nuclear Information System (INIS)

Suk, Ho Chun; Hwang, Woan; Jeong, Young Hwan; Jung, Sung Hoon

1991-07-01

In order to develop CANDU advanced fuel, the agreement of the joint research between KAERI and AECL was made on February 19, 1991. AECL conceptual design of CANFLEX bundle for Bruce reactors was analyzed and then the reference design and design drawing of the advanced fuel bundle with natural uranium fuel for CANDU-6 reactor were completed. The CANFLEX fuel cladding was preliminarily investigated. The fabricability of the advanced fuel bundle was investigated. The design and purchase of the machinery tools for the bundle fabrication for hydraulic scoping tests were performed. As a result of CANFLEX tube examination, the tubes were found to be meet the criteria proposed in the technical specification. The dummy bundles for hydraulic scoping tests have been fabricated by using the process and tools, where the process parameters and tools have been newly established. (Author)

The Atiyah bundle and connections on a principal bundle

Indian Academy of Sciences (India)

be the fiber bundle constructed as in (1.1) for the universal principal G-bundle. In a work in progress, we hope to show that the universal G-connection can be realized as a fiber bundle over C(EG). Turning this ... a G-invariant vector field on EG|U . In other words, we get a bijective linear map between. A(EG)(U) (the space of ...

Demonstrating the compatibility of Canflex fuel bundles with a CANDU 6 fuelling machine

Energy Technology Data Exchange (ETDEWEB)

Alavi, P; Oldaker, I E [Atomic Energy of Canada Ltd., Mississauga, ON (Canada); Suk, H C; Choi, C B [Korea Atomic Energy Research Inst., Taejon (Korea, Republic of)

1997-12-31

CANFLEX is a new 43-element fuel bundle, designed for high operating margins. It has many small-diameter elements in its two outer rings, and large-diameter elements in its centre rings. By this means, the linear heat ratings are lower than those of standard 37-element bundles for similar power outputs. A necessary part of the out-reactor qualification program for the CANFLEX fuel bundle design, is a demonstration of the bundle`s compatibility with the mechanical components in a CANDU 6 Fuelling Machine (FM) under typical conditions of pressure, flow and temperature. The diameter of the CANFLEX bundle is the same as that of a 37-element bundle, but the smaller-diameter elements in the outer ring result in a slightly larger end-plate diameter. Therefore, to minimize any risk of unanticipated damage to the CANDU 6 FM sidestops, a series of measurements and static laboratory tests were undertaken prior to the fuelling machine tests. The tests and measurements showed that; a) the CANFLEX bundle end plate is compatible with the FM sidestops, b) all the dimensions of the CANFLEX fuel bundle are within the specified limits. (author). 3 tabs., 3 figs.

Quench Simulation Studies: Program documentation of SPQR

CERN Document Server

Sonnemann, F

2001-01-01

Quench experiments are being performed on prototypes of the superconducting magnets and busbars to determine the adequate design and protection. Many tests can only be understood correctly with the help of quench simulations that model the thermo-hydraulic and electrodynamic processes during a quench. In some cases simulations are the only method to scale the experimental results of prototype measurements to match the situation of quenching superconducting elements in the LHC. This note introduces the theoretical quench model and the use of the simulation program SPQR (Simulation Program for Quench Research), which has been developed to compute the quench process in superconducting magnets and busbars. The model approximates the heat balance equation with the finite difference method including the temperature dependence of the material parameters. SPQR allows the simulation of longitudinal quench propagation along a superconducting cable, the transverse propagation between adjacent conductors, heat transfer i...

Strain-based quench detection for a solenoid superconducting magnet

International Nuclear Information System (INIS)

Wang Xingzhe; Guan Mingzhi; Ma Lizhen

2012-01-01

In this paper, we present a non-electric quench detection method based on the strain gauge measurement of a superconducting solenoid magnet at cryogenic temperature under an intense magnetic field. Unlike the traditional voltage measurement of quench detection, the strain-based detection method utilizes low-temperature strain gauges, which evidently reduce electromagnetic noise and breakdown, to measure the magneto/thermo-mechanical behavior of the superconducting magnet during excitation. The magnet excitation, quench tests and trainings were performed on a prototype 5 T superconducting solenoid magnet. The transient strains and their abrupt changes were compared with the current, magnetic field and temperature signals collected during excitation and quench tests to indicate that the strain gauge measurements can detect the quench feature of the superconducting magnet. The proposed method is expected to be able to detect the quench of a superconducting coil independently or utilized together with other electrical methods. In addition, the axial quench propagation velocity of the solenoid is evaluated by the quench time lags among different localized strains. The propagation velocity is enhanced after repeated quench trainings. (paper)

Pressure drop redistribution experimental analysis in axial flow along the bundles

International Nuclear Information System (INIS)

Bastos Franco, C. de; Carajilescov, P.

1992-01-01

Fuel elements of PWR type nuclear reactors are composed of rod bundles, arranged in square arrays, held by grid type spacers. The coolant flows axially along the bundle. Although such elements are laterally open, pressure drop experiments are performed in closed type test sections, originating the appearance of subchannels of different geometries. Utilizing a test section of two bundles of 4 x 4 pins and performing experiments with and without separation between the bundles, the flow redistribution factors, the friction, and the grid drag coefficients were determined for the interior subchannels. 03 refs, 06 figs, 02 tabs. (B.C.A.)

Development of CANDU advanced fuel bundle

International Nuclear Information System (INIS)

Suk, H. C.; Hwang, W.; Rhee, B. W.; Jung, S. H.; Chung, C. H.

1992-05-01

This research project is underway in cooperation with AECL to develop the CANDU advanced fuel bundle (so-called, CANFLEX) which can enhance reactor safety and fuel economy in comparison with the current CANDU fuel and which can be used with natural uranium, slightly enriched uranium and other advanced fuel cycle. As the final schedule, the advanced fuel will be verified by carrying out a large scale demonstration of the bundle irradiation in a commercial CANDU reactor for 1996 and 1997, and consequently will be used in the existing and future CANDU reactors in Korea. The research activities during this year include the detail design of CANFLEX fuel with natural enriched uranium (CANFLEX-NU). Based on this design, CANFLEX fuel was mocked up. Out-of-pile hydraulic scoping tests were conducted with the fuel in the CANDU Cold Test Loop to investigate the condition under which maximum pressure drop occurs and the maximum value of the bundle pressure drop. (Author)

Rapid Quench in an Electrostatic Levitator

Science.gov (United States)

SanSoucie, Michael P.; Rogers, Jan R.; Matson, Douglas M.

2016-01-01

The Electrostatic Levitation (ESL) Laboratory at the NASA Marshall Space Flight Center (MSFC) is a unique facility for investigators studying high-temperature materials. The ESL laboratory's main chamber has been upgraded with the addition of a rapid quench system. This system allows samples to be dropped into a quench vessel that can be filled with a low melting point material, such as a gallium or indium alloy, as a quench medium. Thereby allowing rapid quenching of undercooled liquid metals. Up to eight quench vessels can be loaded into a wheel inside the chamber that is indexed with control software. The system has been tested successfully with samples of zirconium, iron-cobalt alloys, titanium-zirconium-nickel alloys, and a silicon-cobalt alloy. This new rapid quench system will allow materials science studies of undercooled materials and new materials development. In this presentation, the system is described and some initial results are presented.

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

International Nuclear Information System (INIS)

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

2001-09-01

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

Quench detection, protection and simulation studies on SST-1 magnets

International Nuclear Information System (INIS)

Sharma, Aashoo N.; Khristi, Yohan; Pradhan, Subrata; Doshi, Kalpesh; Prasad, Upendra; Banaudha, Moni; Varmora, Pankaj; Praghi, Bhadresh R.

2015-01-01

Steady-state Superconducting Tokamak-1 (SST-1) is India's first tokamak with superconducting toroidal field (TF) and Poloidal Field (PF) magnets. These magnets are made with NbTi based Cable-In-Conduit-Conductors. The quench characteristic of SST-1 CICC has been extensively studied both analytically and using simulation codes. Dedicated experiments like model coil test program, TF coil test program and laboratory experiments were conducted to fully characterize the performance of the CICC and the magnets made using this CICC. Results of quench experiments performed during these tests have been used to design the SST-1 quench detection and protection system. Simulation results of TF coil quenches and slow propagation quench of TF busbars have been used to further optimize these systems during the SST-1 tokamak operation. Redundant hydraulic based quench detection is also proposed for the TF coil quench detection. This paper will give the overview of these development and simulation activities. (author)

MSFC Electrostatic Levitator (ESL) Rapid Quench System

Science.gov (United States)

SanSoucie, Michael P.; Craven, Paul D.; Rogers, Jan R.

2014-01-01

The NASA Marshall Space Flight Center (MSFC) Electrostatic Levitator (ESL) Laboratory is a unique facility for investigators studying high-temperature materials. The laboratory boasts two levitators in which samples can be levitated, heated, melted, undercooled, and resolidified, all without the interference of a container or data-gathering instrument. The ESL main chamber has been upgraded with the addition of a rapid quench system. This system allows samples to be dropped into a quench vessel that can be filled with a low melting point material, such as a gallium or indium alloy. Thereby allowing rapid quenching of undercooled liquid metals. Up to 8 quench vessels can be loaded into the quench wheel, which is indexed with LabVIEW control software. This allows up to 8 samples to be rapidly quenched before having to open the chamber. The system has been tested successfully on several zirconium samples. Future work will be done with other materials using different quench mediums. Microstructural analysis will also be done on successfully quench samples.

Polyelectrolyte bundles

Energy Technology Data Exchange (ETDEWEB)

Limbach, H J; Sayar, M; Holm, C [Max-Planck-Institut fuer Polymerforschung, Ackermannweg 10, 55128 Mainz (Germany)

2004-06-09

Using extensive molecular dynamics simulations we study the behaviour of polyelectrolytes with hydrophobic side chains, which are known to form cylindrical micelles in aqueous solution. We investigate the stability of such bundles with respect to hydrophobicity, the strength of the electrostatic interaction and the bundle size. We show that for the parameter range relevant for sulfonated poly(para-phenylenes) (PPP) one finds a stable finite bundle size. In a more generic model we also show the influence of the length of the precursor oligomer on the stability of the bundles. We also point out that our model has close similarities to DNA solutions with added condensing agents, hinting at the possibility that the size of DNA aggregates is, under certain circumstances, thermodynamically limited.

Polyelectrolyte bundles

International Nuclear Information System (INIS)

Limbach, H J; Sayar, M; Holm, C

2004-01-01

Using extensive molecular dynamics simulations we study the behaviour of polyelectrolytes with hydrophobic side chains, which are known to form cylindrical micelles in aqueous solution. We investigate the stability of such bundles with respect to hydrophobicity, the strength of the electrostatic interaction and the bundle size. We show that for the parameter range relevant for sulfonated poly(para-phenylenes) (PPP) one finds a stable finite bundle size. In a more generic model we also show the influence of the length of the precursor oligomer on the stability of the bundles. We also point out that our model has close similarities to DNA solutions with added condensing agents, hinting at the possibility that the size of DNA aggregates is, under certain circumstances, thermodynamically limited

Polyelectrolyte bundles

Science.gov (United States)

Limbach, H. J.; Sayar, M.; Holm, C.

2004-06-01

Using extensive Molecular Dynamics simulations we study the behavior of polyelectrolytes with hydrophobic side chains, which are known to form cylindrical micelles in aqueous solution. We investigate the stability of such bundles with respect to hydrophobicity, the strength of the electrostatic interaction, and the bundle size. We show that for the parameter range relevant for sulfonated poly-para-phenylenes (PPP) one finds a stable finite bundle size. In a more generic model we also show the influence of the length of the precursor oligomer on the stability of the bundles. We also point out that our model has close similarities to DNA solutions with added condensing agents, hinting to the possibility that the size of DNA aggregates is under certain circumstances thermodynamically limited.

Criterion for the onset of quench for low-flow reflood

International Nuclear Information System (INIS)

Hsu, Y.Y.; Young, M.W.

1982-07-01

This study provides a criterion for the onset of quench for low flow reflood. The criterion is a combination of two conditions: T/sub clad/ < T/sub limiting quench/ where T = Temperature, and α < 0.95 where α = void fraction. This criterion was obtained by examining temperature data from tests simulating PWR reflood, such as FLECHT, THTF, PBF, CCTF, and FEBA tests, with void fraction data from CCTF, FEBA, and FLECHT low flood tests. The data show that quenching initiated at α = 0.95 and that the majority of quench occurred at void fractions near 0.85. The results show that rods can be completely quenched by entrained droplets even if the collapsed liquid level does not advance. A thorough discussion of the analysis which supports this quench criterion is given in the text of this report

Strategic Aspects of Bundling

International Nuclear Information System (INIS)

Podesta, Marion

2008-01-01

The increase of bundle supply has become widespread in several sectors (for instance in telecommunications and energy fields). This paper review relates strategic aspects of bundling. The main purpose of this paper is to analyze profitability of bundling strategies according to the degree of competition and the characteristics of goods. Moreover, bundling can be used as price discrimination tool, screening device or entry barriers. In monopoly case bundling strategy is efficient to sort consumers in different categories in order to capture a maximum of surplus. However, when competition increases, the profitability on bundling strategies depends on correlation of consumers' reservations values. (author)

Nonabelian bundle 2-gerbes

OpenAIRE

Jurco, Branislav

2009-01-01

We define 2-crossed module bundle 2-gerbes related to general Lie 2-crossed modules and discuss their properties. A 2-crossed module bundle 2-gerbe over a manifold is defined in terms of a so called 2-crossed module bundle gerbe, which is a crossed module bundle gerbe equipped with an extra sructure. It is shown that string structures can be described and classified using 2-crossed module bundle 2-gerbes.

Polycation induced actin bundles.

Science.gov (United States)

Muhlrad, Andras; Grintsevich, Elena E; Reisler, Emil

2011-04-01

Three polycations, polylysine, the polyamine spermine and the polycationic protein lysozyme were used to study the formation, structure, ionic strength sensitivity and dissociation of polycation-induced actin bundles. Bundles form fast, simultaneously with the polymerization of MgATP-G-actins, upon the addition of polycations to solutions of actins at low ionic strength conditions. This indicates that nuclei and/or nascent filaments bundle due to attractive, electrostatic effect of polycations and the neutralization of repulsive interactions of negative charges on actin. The attractive forces between the filaments are strong, as shown by the low (in nanomolar range) critical concentration of their bundling at low ionic strength. These bundles are sensitive to ionic strength and disassemble partially in 100 mM NaCl, but both the dissociation and ionic strength sensitivity can be countered by higher polycation concentrations. Cys374 residues of actin monomers residing on neighboring filaments in the bundles can be cross-linked by the short span (5.4Å) MTS-1 (1,1-methanedyl bismethanethiosulfonate) cross-linker, which indicates a tight packing of filaments in the bundles. The interfilament cross-links, which connect monomers located on oppositely oriented filaments, prevent disassembly of bundles at high ionic strength. Cofilin and the polysaccharide polyanion heparin disassemble lysozyme induced actin bundles more effectively than the polylysine-induced bundles. The actin-lysozyme bundles are pathologically significant as both proteins are found in the pulmonary airways of cystic fibrosis patients. Their bundles contribute to the formation of viscous mucus, which is the main cause of breathing difficulties and eventual death in this disorder. Copyright © 2011 Elsevier B.V. All rights reserved.

Ab initio density functional theory investigation of crystalline bundles of polygonized single-walled silicon carbide nanotubes

Energy Technology Data Exchange (ETDEWEB)

Moradian, Rostam; Behzad, Somayeh; Chegel, Raad [Physics Department, Faculty of Science, Razi University, Kermanshah (Iran, Islamic Republic of)], E-mail: moradian.rostam@gmail.com

2008-11-19

By using ab initio density functional theory, the structural characterizations and electronic properties of two large-diameter (13, 13) and (14, 14) armchair silicon carbide nanotube (SiCNT) bundles are investigated. Full structural optimizations show that the cross sections of these large-diameter SiCNTs in the bundles have a nearly hexagonal shape. The effects of inter-tube coupling on the electronic dispersions of large-diameter SiCNT bundles are demonstrated. By comparing the band structures of the triangular lattices of (14, 14) SiCNTs with nearly hexagonal and circular cross sections we found that the polygonization of the tubes in the bundle leads to a further dispersion of the occupied bands and an increase in the bandgap by 0.18 eV.

Ab initio density functional theory investigation of crystalline bundles of polygonized single-walled silicon carbide nanotubes

International Nuclear Information System (INIS)

Moradian, Rostam; Behzad, Somayeh; Chegel, Raad

2008-01-01

By using ab initio density functional theory, the structural characterizations and electronic properties of two large-diameter (13, 13) and (14, 14) armchair silicon carbide nanotube (SiCNT) bundles are investigated. Full structural optimizations show that the cross sections of these large-diameter SiCNTs in the bundles have a nearly hexagonal shape. The effects of inter-tube coupling on the electronic dispersions of large-diameter SiCNT bundles are demonstrated. By comparing the band structures of the triangular lattices of (14, 14) SiCNTs with nearly hexagonal and circular cross sections we found that the polygonization of the tubes in the bundle leads to a further dispersion of the occupied bands and an increase in the bandgap by 0.18 eV.

Ab initio density functional theory investigation of crystalline bundles of polygonized single-walled silicon carbide nanotubes

Science.gov (United States)

Moradian, Rostam; Behzad, Somayeh; Chegel, Raad

2008-11-01

By using ab initio density functional theory, the structural characterizations and electronic properties of two large-diameter (13, 13) and (14, 14) armchair silicon carbide nanotube (SiCNT) bundles are investigated. Full structural optimizations show that the cross sections of these large-diameter SiCNTs in the bundles have a nearly hexagonal shape. The effects of inter-tube coupling on the electronic dispersions of large-diameter SiCNT bundles are demonstrated. By comparing the band structures of the triangular lattices of (14, 14) SiCNTs with nearly hexagonal and circular cross sections we found that the polygonization of the tubes in the bundle leads to a further dispersion of the occupied bands and an increase in the bandgap by 0.18 eV.

Group quenching and galactic conformity at low redshift

Science.gov (United States)

Treyer, M.; Kraljic, K.; Arnouts, S.; de la Torre, S.; Pichon, C.; Dubois, Y.; Vibert, D.; Milliard, B.; Laigle, C.; Seibert, M.; Brown, M. J. I.; Grootes, M. W.; Wright, A. H.; Liske, J.; Lara-Lopez, M. A.; Bland-Hawthorn, J.

2018-06-01

We quantify the quenching impact of the group environment using the spectroscopic survey Galaxy and Mass Assembly to z ˜ 0.2. The fraction of red (quiescent) galaxies, whether in groups or isolated, increases with both stellar mass and large-scale (5 Mpc) density. At fixed stellar mass, the red fraction is on average higher for satellites of red centrals than of blue (star-forming) centrals, a galactic conformity effect that increases with density. Most of the signal originates from groups that have the highest stellar mass, reside in the densest environments, and have massive, red only centrals. Assuming a colour-dependent halo-to-stellar-mass ratio, whereby red central galaxies inhabit significantly more massive haloes than blue ones of the same stellar mass, two regimes emerge more distinctly: at log (Mhalo/M⊙) ≲ 13, central quenching is still ongoing, conformity is no longer existent, and satellites and group centrals exhibit the same quenching excess over field galaxies at all mass and density, in agreement with the concept of `group quenching'; at log (Mh/M⊙) ≳ 13, a cut-off that sets apart massive (log (M⋆/M⊙) > 11), fully quenched group centrals, conformity is meaningless, and satellites undergo significantly more quenching than their counterparts in smaller haloes. The latter effect strongly increases with density, giving rise to the density-dependent conformity signal when both regimes are mixed. The star formation of blue satellites in massive haloes is also suppressed compared to blue field galaxies, while blue group centrals and the majority of blue satellites, which reside in low-mass haloes, show no deviation from the colour-stellar mass relation of blue field galaxies.

Group quenching and galactic conformity at low redshift

Science.gov (United States)

Treyer, M.; Kraljic, K.; Arnouts, S.; de la Torre, S.; Pichon, C.; Dubois, Y.; Vibert, D.; Milliard, B.; Laigle, C.; Seibert, M.; Brown, M. J. I.; Grootes, M. W.; Wright, A. H.; Liske, J.; Lara-Lopez, M. A.; Bland-Hawthorn, J.

2018-03-01

We quantify the quenching impact of the group environment using the spectroscopic survey Galaxy and Mass Assembly (GAMA) to z ˜ 0.2. The fraction of red (quiescent) galaxies, whether in groups or isolated, increases with both stellar mass and large-scale (5 Mpc) density. At fixed stellar mass, the red fraction is on average higher for satellites of red centrals than of blue (star-forming) centrals, a galactic conformity effect that increases with density. Most of the signal originates from groups that have the highest stellar mass, reside in the densest environments, and have massive, red only centrals. Assuming a color-dependent halo-to-stellar-mass ratio, whereby red central galaxies inhabit significantly more massive halos than blue ones of the same stellar mass, two regimes emerge more distinctly: at log (Mhalo/M⊙) ≲ 13, central quenching is still ongoing, conformity is no longer existent, and satellites and group centrals exhibit the same quenching excess over field galaxies at all mass and density, in agreement with the concept of "group quenching"; at log (Mh/M⊙) ≳ 13, a cutoff that sets apart massive (log (M⋆/M⊙) > 11), fully quenched group centrals, conformity is meaningless, and satellites undergo significantly more quenching than their counterparts in smaller halos. The latter effect strongly increases with density, giving rise to the density-dependent conformity signal when both regimes are mixed. The star-formation of blue satellites in massive halos is also suppressed compared to blue field galaxies, while blue group centrals and the majority of blue satellites, which reside in low mass halos, show no deviation from the color-stellar mass relation of blue field galaxies.

MELCOR 1.8.1 calculations of ISP31: The CORA-13 experiment

International Nuclear Information System (INIS)

Gross, R.J.; Thompson, S.L.; Martinez, G.M.

1993-06-01

The MELCOR code was used to simulate one of GRS's (a reactor research group in Germany) core degradation experiments conducted in the CORA out-of-pile test facility. This test, designated CORA-13, was selected as one of the International Standard Problems, Number ISP31, by the Organization for Economic Cooperation and Development. In this blind calculation, only initial and boundary conditions were provided. The experiment consisted of a small core bundle of twenty-five PWR fuel elements that was electrically heated to temperatures greater than 2,800 K. The experiment composed three phases: a 3,000 second gas preheat phase, an 1,870 second transient phase, and a 180 second water quench phase. MELCOR predictions are compared both to the experimental data and to eight other ISP31 submittals. Temperatures of various components, energy balance, zircaloy oxidation, and core blockage are examined. Up to the point where oxidation was significant, MELCOR temperatures agreed very well with the experiment -- usually to within 50 K. MELCOR predicted oxidation to occur about 100 seconds earlier and at a faster rate than experimental data. The large oxidation spike that occurred during quench was not predicted. However, the experiment produced 210 grams of hydrogen, while MELCOR predicted 184 grams, which was one of the closest integral predictions of the nine submittals. Core blockage was of the right magnitude; however, material collected on the lower grid spacer in the experiment at an axial location of 450 mm, while in MELCOR the material collected at the 50 to 150 mm location. In general, compared to the other submittals, the MELCOR calculation was superior

MELCOR 1.8.1 calculations of ISP31: The CORA-13 experiment

Energy Technology Data Exchange (ETDEWEB)

Gross, R.J.; Thompson, S.L.; Martinez, G.M.

1993-06-01

The MELCOR code was used to simulate one of GRS`s (a reactor research group in Germany) core degradation experiments conducted in the CORA out-of-pile test facility. This test, designated CORA-13, was selected as one of the International Standard Problems, Number ISP31, by the Organization for Economic Cooperation and Development. In this blind calculation, only initial and boundary conditions were provided. The experiment consisted of a small core bundle of twenty-five PWR fuel elements that was electrically heated to temperatures greater than 2,800 K. The experiment composed three phases: a 3,000 second gas preheat phase, an 1,870 second transient phase, and a 180 second water quench phase. MELCOR predictions are compared both to the experimental data and to eight other ISP31 submittals. Temperatures of various components, energy balance, zircaloy oxidation, and core blockage are examined. Up to the point where oxidation was significant, MELCOR temperatures agreed very well with the experiment -- usually to within 50 K. MELCOR predicted oxidation to occur about 100 seconds earlier and at a faster rate than experimental data. The large oxidation spike that occurred during quench was not predicted. However, the experiment produced 210 grams of hydrogen, while MELCOR predicted 184 grams, which was one of the closest integral predictions of the nine submittals. Core blockage was of the right magnitude; however, material collected on the lower grid spacer in the experiment at an axial location of 450 mm, while in MELCOR the material collected at the 50 to 150 mm location. In general, compared to the other submittals, the MELCOR calculation was superior.

Laboratory manual for salt-mixing test in 37- and 217-pin bundles

International Nuclear Information System (INIS)

Chan, Y.N.; Todreas, N.E.

1980-08-01

This laboratory manual deals with the procedure employed during salt tracer experiments used in evaluating the hydraulic characteristics of a rod bundle. A description of the standard equipment used is given together with the details of manufacture of probes used for detecting the salt concentration. Details of the bundle construction have been excluded as they are availble in the reference cited. An attempt has been made to point out potential trouble areas and procedures

Nefness of adjoint bundles for ample vector bundles

Directory of Open Access Journals (Sweden)

Hidetoshi Maeda

1995-11-01

Full Text Available Let E be an ample vector bundle of rank >1 on a smooth complex projective variety X of dimension n. This paper gives a classification of pairs (X,E whose adjoint bundles K_X+det E are not nef in the case when  r=n-2.

Quenching of Einstein-coefficients by photons

International Nuclear Information System (INIS)

Aumayr, F.; Skinner, C.H.; Suckewer, S.; Princeton Univ., NJ; Lee, W.

1991-02-01

Experimental evidence is presented for the change of Einstein's A-coefficients for spontaneous transitions from the upper laser level of an argon ion laser discharge due to the presence of the high-intensity laser flux. To demonstrate that this quenching effect cannot be attributed to a reduction in self-absorption of the strong spontaneous emission line, absorption and line profile measurements have been performed. Computer modelling of the reduction of self absorption due to Rabi splitting also indicated that this effect is too small to explain the observed quenching of spontaneous line emissions. 13 refs., 11 figs

Quenching of Einstein-coefficients by photons

International Nuclear Information System (INIS)

Aumayr, F.; Lee, W.; Skinner, C.H.; Suckewer, S.

1991-03-01

Experimental evidence is presented for the change of Einstein's A- coefficients for spontaneous transitions from the upper laser level of argon ion laser discharge due to the presence of the high- intensity laser flux. To demonstrate that this quenching effect cannot be attributed to a reduction in self-absorption of the strong spontaneous emission line, absorption and line profile measurements have been performed. Computer modelling of the reduction of self absorption due to Rabi splitting also indicated that this effect is too small to explain the observed quenching of spontaneous line emissions. 13 refs., 11 figs

Testing of high current by-pass diodes for the LHC magnet quench protection

International Nuclear Information System (INIS)

Berland, V.; Hagedorn, D.; Rodriguez-Mateos, F.

1996-01-01

Within the framework of the Large Hadron Collider (LHC) R and D program, CERN is performing experiments to establish the current carrying capability of irradiated diodes at liquid Helium temperatures for the superconducting magnet protection. Even if the diodes are degraded by radiation dose and neutron fluence, they must be able to support the by-pass current during a magnet quench and the de-excitation of the superconducting magnet ring. During this discharge, the current in the diode reaches a maximum value up to 13 kA and decreased with an exponential time constant of 100 s. Two sets of 75 mm wafer diameter epitaxial diodes, one irradiated and one non-irradiated, were submitted to this experiment. The irradiated diodes have been exposed to radiation in the accelerator environment up to 20 kGy and then annealed at room temperature. After the radiation exposure the diodes had shown a degradation of forward voltage of 50% which reduced to about 14% after the thermal annealing. During the long duration high current tests, one of the diodes was destroyed and the other two irradiated diodes showed a different behavior compared with non-irradiated diodes

First experience with the new Coupling Loss Induced Quench system

CERN Document Server

Ravaioli, E; Dudarev, A V; Kirby, G; Sperin, K A; ten Kate, H H J; Verweij, A P

2014-01-01

New-generation high-field superconducting magnets pose a challenge relating to the protection of the coil winding pack in the case of a quench. The high stored energy per unit volume calls for a very efficient quench detection and fast quench propagation in order to avoid damage due to overheating. A new protection system called Coupling-Loss Induced Quench (CLIQ) was recently, developed and tested at CERN. This method provokes a fast change in the magnet transport current by means of a capacitive discharge. The resulting change in the local magnetic field induces inter-filament and inter-strand coupling losses which heat up the superconductor and eventually initiate a quench in a large fraction of the coil winding pack. The method is extensively tested on a Nb-Ti single-wire test solenoid magnet in the CERN Cryogenic Laboratory in order to assess its performance, optimize its operating parameters, and study new electrical configurations. Each parameter is thoroughly analyzed and its impact on the quench effi...

The MIMIC Model as a Tool for Differential Bundle Functioning Detection

Science.gov (United States)

Finch, W. Holmes

2012-01-01

Increasingly, researchers interested in identifying potentially biased test items are encouraged to use a confirmatory, rather than exploratory, approach. One such method for confirmatory testing is rooted in differential bundle functioning (DBF), where hypotheses regarding potential differential item functioning (DIF) for sets of items (bundles)…

ISP42 (PANDA Tests) - Blind Phase Comparison Report

International Nuclear Information System (INIS)

Luebbesmeyer, D.; Aksan, S.N.

2003-05-01

The International Standard Problem (ISP) No. 45 is part of the overall ISP program of the OECD/NEA and is dedicated to the behavior of heat-up and delayed reflood of fuel elements in nuclear reactors during a hypothetical accident. ISP-45 is related to the out-of-pile bundle quench experiment QUENCH-06, performed at Forschungszentrum Karlsruhe (FZK), Germany, on December 13, 2000. Special attention was paid to hydrogen production. To assess the ability of severe accident codes to simulate processes during core heat-up and reflood at temperatures above 2000 K, the behavior of the bundle during the whole experiment should be calculated on the basis of the necessary experimental initial and boundary conditions, but without knowing further experimental details. In this so-called blind phase 21 participants from 15 nations contributed with 8 different code systems (ATHLET-CD, ICARE/CATHARE, IMPACT/SAMPSON, GENFLO, MAAP, MELCOR, SCDAPSIM, SCDAP-3D). Additionally, posttest calculations using the in-house version SCDAP/RELAP5 mod3.2.irs are used for comparison. After the end of the blind phase all measured data were made available and the participants were invited to deliver a second calculation, where this knowledge could be used (so called open phase). In this report, results of the blind calculations are presented, analyzed, and compared to experimental data. During heat-up most results do not deviate significantly from one another, except as a consequence of some obvious user errors, so that a definition of a mainstream is justified. For the quench phase the lack of adequate hydraulic modeling becomes obvious: some participants could not match the observed cool-down rates, others had to use very fine meshes to compensate code deficiencies. To overcome this insufficiency some newly developed reflood models were used in MAAP and MELCOR. In QUENCH-06, oxide layers were thick enough to protect the cladding from melting and failure below 2200 K, so that no massive hydrogen

Effects of die quench forming on sheet thinning and 3-point bend testing of AA7075-T6

Science.gov (United States)

Kim, Samuel; Omer, Kaab; Rahmaan, Taamjeed; Butcher, Clifford; Worswick, Michael

2017-10-01

Lab-scaled AA7075 aluminum side impact beams were manufactured using the die quenching technique in which the sheet was solutionized and then quenched in-die during forming to a super saturated solid state. Sheet thinning measurements were taken at various locations throughout the length of the part and the effect of lubricant on surface scoring and material pick-up on the die was evaluated. The as-formed beams were subjected to a T6 aging treatment and then tested in three-point bending. Simulations were performed of the forming and mechanical testing experiments using the LS-DYNA finite element code. The thinning and mechanical response was predicted well.

Bundling harvester; Harvennuspuun automaattisen nippukorjausharvesterin kehittaeminen

Energy Technology Data Exchange (ETDEWEB)

Koponen, K [Eko-Log Oy, Kuopio (Finland)

1997-12-01

The starting point of the project was to design and construct, by taking the silvicultural point of view into account, a harvesting and processing system especially for energy-wood, containing manually driven bundling harvester, automating of the harvester, and automated loading. The equipment forms an ideal method for entrepreneur`s-line harvesting. The target is to apply the system also for owner`s-line harvesting. The profitability of the system promotes the utilisation of the system in both cases. The objectives of the project were: to construct a test equipment and prototypes for all the project stages, to carry out terrain and strain tests in order to examine the usability and durability, as well as the capacity of the machine, to test the applicability of the Eko-Log system in simultaneous harvesting of energy and pulp woods, and to start the marketing and manufacturing of the products. The basic problems of the construction of the bundling harvester have been solved using terrain-tests. The prototype machine has been shown to be operable. Loading of the bundles to form sufficiently economically transportable loads has been studied, and simultaneously, the branch-biomass has been tried to be utilised without loosing the profitability of transportation. The results have been promising, and will promote the profitable utilisation of wood-energy. (orig.)

Superconducting synchrotron power supply and quench protection scheme

International Nuclear Information System (INIS)

Stiening, R.; Flora, R.; Lauckner, R.; Tool, G.

1978-01-01

The power supply and quench protection scheme for the proposed Fermilab 6 km circumference superconducting synchrotron is described. Specifically, the following points are discussed: (1) the 46 MW thyristor power supply; (2) the 3 x 10 8 J emergency energy dump; (3) the distributed microprocessing system for the detection of quenches; (4) the thyristor network for shunting current around quenched magnets; and (5) the heaters internal to the magnets which cause rapid propagation of quenches. Test results on prototype systems are given

Quantum quenches with integrable pre-quench dynamics

OpenAIRE

Delfino, Gesualdo

2014-01-01

We consider the unitary time evolution of a one-dimensional quantum system which is in a stationary state for negative times and then undergoes a sudden change (quench) of a parameter of its Hamiltonian at t=0. For systems possessing a continuum limit described by a massive quantum field theory we investigate in general perturbative quenches for the case in which the theory is integrable before the quench.

Quantum quenches with integrable pre-quench dynamics

International Nuclear Information System (INIS)

Delfino, Gesualdo

2014-01-01

We consider the unitary time evolution of a one-dimensional quantum system which is in a stationary state for negative times and then undergoes a sudden change (quench) of a parameter of its Hamiltonian at t = 0. For systems possessing a continuum limit described by a massive quantum field theory we investigate in general perturbative quenches for the case in which the theory is integrable before the quench. (fast track communication)

Stability and quench of dual cooling channel cable-in-conduct superconductors

International Nuclear Information System (INIS)

Blau, Bertrand

1999-11-01

Presently, the most ambitious experimental project towards controlled thermonuclear fusion is the International Thermonuclear Experimental Reactor ITER. All coils of its magnet system will be superconducting since for magnetic fields in the range between 6 - 13 T high current densities are required. During recent years, in particular for fusion applications, a special configuration of superconductor was favoured: the so-called Cable-In-Conduit Conductor (CICC). The CICCs for ITER consist of a superconducting cable made of a large number of superconducting wires (NbTi or Nb 3 Sn) twisted around a central cooling channel, which are tightly jacketed in a metal conduit, providing the desired mechanical stiffness of the conductor against magnetic forces. Pressurized supercritical helium is pumped through the cable interstices and the central channel. The direct contact between the coolant and the cable provides good thermal stability of the conductor against sudden energy inputs. These disturbances can lead to a transition into the normal state (quench) if the released energy is sufficiently high, so that the temperature of the superconductor exceeds locally its critical temperature and if the energy cannot be absorbed efficiently by the surrounding helium. Stability of superconductors against quenches is one of the most important issues in applied superconductivity. The recovery capabilities of a CICC after thermal disturbances are governed by the heat transfer rate from the strands to the helium. The heat transfer is greatly affected by the flow velocity of the coolant. It has been shown theoretically that a temporal thermal disturbance in a CICC can induce an additional strong helium flow, which enhances the heat transfer rate and, hence, the stability. This self-stabilizing effect is believed to play an important role for the recovery capabilities of a CICC. The scope of this thesis is the experimental assessment of the quench and stability behaviour of dual cooling

Stability and quench of dual cooling channel cable-in-conduct superconductors

Energy Technology Data Exchange (ETDEWEB)

Blau, Bertrand [Ecole Polytechnique Federale de Lausanne, Centre de Recherches en Physique des Plasmas (CRPP), CH-1015 Lausanne (Switzerland)

1999-11-01

Presently, the most ambitious experimental project towards controlled thermonuclear fusion is the International Thermonuclear Experimental Reactor ITER. All coils of its magnet system will be superconducting since for magnetic fields in the range between 6 - 13 T high current densities are required. During recent years, in particular for fusion applications, a special configuration of superconductor was favoured: the so-called Cable-In-Conduit Conductor (CICC). The CICCs for ITER consist of a superconducting cable made of a large number of superconducting wires (NbTi or Nb{sub 3}Sn) twisted around a central cooling channel, which are tightly jacketed in a metal conduit, providing the desired mechanical stiffness of the conductor against magnetic forces. Pressurized supercritical helium is pumped through the cable interstices and the central channel. The direct contact between the coolant and the cable provides good thermal stability of the conductor against sudden energy inputs. These disturbances can lead to a transition into the normal state (quench) if the released energy is sufficiently high, so that the temperature of the superconductor exceeds locally its critical temperature and if the energy cannot be absorbed efficiently by the surrounding helium. Stability of superconductors against quenches is one of the most important issues in applied superconductivity. The recovery capabilities of a CICC after thermal disturbances are governed by the heat transfer rate from the strands to the helium. The heat transfer is greatly affected by the flow velocity of the coolant. It has been shown theoretically that a temporal thermal disturbance in a CICC can induce an additional strong helium flow, which enhances the heat transfer rate and, hence, the stability. This self-stabilizing effect is believed to play an important role for the recovery capabilities of a CICC. The scope of this thesis is the experimental assessment of the quench and stability behaviour of dual

Corrosion fatigue in nitrocarburized quenched and tempered steels

Science.gov (United States)

Khani, M. Karim; Dengel, D.

1996-05-01

In order to investigate the fatigue strength and fracture mechanism of salt bath nitrocarburized steels, specimens of the steels SAE 4135 and SAE 4140, in a quenched and tempered state, and additionally in a salt bath nitrocarburized and oxidizing cooled state as well as in a polished (after the oxidizing cooling) and renewed oxidized state, were subjected to comparative rotating bending fatigue tests in inert oil and 5 pct NaCl solution. In addition, some of the quenched and tempered specimens of SAE 4135 material were provided with an approximately 50-μm-thick electroless Ni-P layer, in order to compare corrosion fatigue behavior between the Ni-P layer and the nitride layers. Long-life corrosion fatigue tests of SAE 4135 material were carried out under small stresses in the long-life range up to 108 cycles with a test frequency of 100 Hz. Fatigue tests of SAE 4140 material were carried out in the range of finite life (low-cycle range) with a test frequency of 13 Hz. The results show that the 5 pct NaCl environment drastically reduced fatigue life, but nitrocarburizing plus oxidation treatment was found to improve the corrosion fatigue life over that of untreated and Ni-P coated specimens. The beneficial effect of nitrocarburizing followed by oxidation treatment on cor-rosion fatigue life results from the protection rendered by the compound layer by means of a well-sealed oxide layer, whereby the pores present in the compound layer fill up with oxides. The role of inclusions in initiating fatigue cracks was investigated. It was found that under corrosion fatigue conditions, the fatigue cracks started at cavities along the interfaces of MnS inclusions and matrix in the case of quenched and tempered specimens. The nitrocarburized specimens, however, showed a superposition of pitting corrosion and corrosion fatigue in which pores and nonmetallic inclusions in the compound layer play a predominant role concerning the formation of pits in the substrate.

Heat Transfer Analysis in Wire Bundles for Aerospace Vehicles

Science.gov (United States)

Rickman, S. L.; Iamello, C. J.

2016-01-01

Design of wiring for aerospace vehicles relies on an understanding of "ampacity" which refers to the current carrying capacity of wires, either, individually or in wire bundles. Designers rely on standards to derate allowable current flow to prevent exceedance of wire temperature limits due to resistive heat dissipation within the wires or wire bundles. These standards often add considerable margin and are based on empirical data. Commercial providers are taking an aggressive approach to wire sizing which challenges the conventional wisdom of the established standards. Thermal modelling of wire bundles may offer significant mass reduction in a system if the technique can be generalized to produce reliable temperature predictions for arbitrary bundle configurations. Thermal analysis has been applied to the problem of wire bundles wherein any or all of the wires within the bundle may carry current. Wire bundles present analytical challenges because the heat transfer path from conductors internal to the bundle is tortuous, relying on internal radiation and thermal interface conductance to move the heat from within the bundle to the external jacket where it can be carried away by convective and radiative heat transfer. The problem is further complicated by the dependence of wire electrical resistivity on temperature. Reduced heat transfer out of the bundle leads to higher conductor temperatures and, hence, increased resistive heat dissipation. Development of a generalized wire bundle thermal model is presented and compared with test data. The steady state heat balance for a single wire is derived and extended to the bundle configuration. The generalized model includes the effects of temperature varying resistance, internal radiation and thermal interface conductance, external radiation and temperature varying convective relief from the free surface. The sensitivity of the response to uncertainties in key model parameters is explored using Monte Carlo analysis.

Quench antenna for superconducting particle accelerator magnets

International Nuclear Information System (INIS)

Ogitsu, T.; Devred, A.; Kim, K.

1993-10-01

We report on the design, fabrication, and test of an assembly of stationary pickup coils which can be used to localize quench origins. After describing the pickup coils configuration, we develop a simple model of current redistribution which allows interpretation of the measured voltages and determination of the turn of the magnet coil in which the quench started. The technique is illustrated by analyzing the data from a quench of a 5-cm-aperture, 15-m-long SSC dipole magnet prototype

Quenched chiral logarithms

International Nuclear Information System (INIS)

Sharpe, S.R.

1992-04-01

I develop a diagrammatic method for calculating chiral logarithms in the quenched approximation. While not rigorous, the method is based on physically reasonable assumptions, which can be tested by numerical simulations. The main results are that, at leading order in the chiral expansion, (a) there are no chiral logarithms in quenched f π m u = m d ; (b) the chiral logarithms in B K and related kaon B-parameters are, for m d = m s the same in the quenched approximation as in the full theory (c) for m π and the condensate, there are extra chiral logarithms due to loops containing the η', which lead to a peculiar non-analytic dependence of these quantities on the bare quark mass. Following the work of Gasser and Leutwyler, I discuss how there is a predictable finite volume dependence associated with each chiral logarithm. I compare the resulting predictions with numerical results: for most quantities the expected volume dependence is smaller than the errors. but for B V and B A there is an observed dependence which is consistent with the predictions

Quench origins

International Nuclear Information System (INIS)

Devred, A.

1990-03-01

In this paper, I shall discuss the quench origins. I shall first establish a method of classification and introduce the notions of conductor-limited and energy-deposited quenches. Next the paper will be devoted to the study of conductor-limited quenches, and I shall introduce the notions of plateau and of fraction of short sample. Also the paper will be devoted to the study of energy-deposited quenches, and I shall introduce the notions of training and of minimum energy deposit; I shall then review the possible causes of energy release. Lastly, I shall introduce the notion of operating margin, and I shall indicate how to optimize the operating margin in order to limit the risk of premature quenching. 112 refs., 14 figs

Evaluation of CHF experimental data for non-square lattice 7-rod bundles

Energy Technology Data Exchange (ETDEWEB)

Hwang, Dae Hyun; Yoo, Y. J.; Kim, K. K.; Zee, S. Q

2001-01-01

A series of CHF experiments are conducted for 7-rod hexagonal test bundles in order to investigate the CHF characteristics of self-sustained square finned (SSF) rod bundles. The experiments are performed in the freon-loop and water-loop located at IPPE in Russia, and 609 data of freon-12 and 229 data of water are obtained from 7 kinds of test bundles classified by the combination of heated length and axial/radial power distributions. As the result of the evaluation of four representative CHF correlations, the EPRI-1 correlation reveals good prediction capability for SSF test bundles. The inlet parameter CHF correlation suggested by IPPE calculates the mean and the standard deviation of P/M for uniformly heated test bundles as 1.002 and 0.049, respectively. In spite of its excellent accuracy, the correlation has a discontinuity point at the boundary between the low velocity and high velocity conditions. KAERI's inlet parameter correlation eliminates this defect by introducing the complete evaporation model at low velocity condition, and calculates the mean and standard deviation of P/M as 0.095 and 0.062 for uniformly heated 496 data points, respectively. The mean/standard deviation of local parameter CHF correlations suggested by IPPE and KAERI are evaluated as 1.023/0.178 and 1.002/0.158, respectively. The inlet parameter correlation developed from uniformly heated test bundles tends to under-predict CHF about 3% for axially non-uniformly heated test bundles. On the other hand, the local parameter correlation reveals large scattering of P/M, and requires re-optimization of the correlation for non-uniform axial power distributions. As the result of the analysis of experimental data, it reveals that the correction model of axial power shapes suggested by IPPE is applicable to the inlet parameter correlations. For the test bundle of radial non-uniform power distribution, the physically unexpected results are obtained at some experimental conditions. In addition

Out-of-pile bundle temperature escalation under severe fuel damage conditions

International Nuclear Information System (INIS)

Hagen, S.; Peck, S.O.

1983-08-01

This report provides an overview of the test conduct, results, and posttest appearance of bundle test ESBU-1. The purpose of the test was to investigate fuel rod temperature escalation due to the exothermal zircaloy/steam reaction in a bundle geometry. The 3x3 bundle was surrounded by a zircaloy shroud and 6 mm of fiber ceramic insulation. The center rod escalated to a maximum of 2,250 0 C. Runoff of the melt apparently limited the escalation. Posttest visual examination of the bundle showed that cladding from every rod had melted, liquefied some fuel, flowed down the rod, and frozen in a solid mass that substantially blocked all flow channels. A large amount of powdery rubble, probably fuel that fractured during cooldown, was found on top of the blockage. Metallographic, EMP, and SEM examinations showed that the melt had dissolved both fuel and oxidized cladding, and had itself been oxidized by steam. (orig.) [de

Establishment and assessment of CHF data base for square-lattice rod bundles

International Nuclear Information System (INIS)

Hwang, Dae Hyun; Seo, K. W.; Kim, K. K.; Zee, S. Q.

2002-02-01

A CHF data base is constructed for square-lattice rod bundles, and assessed with various existing CHF prediction models. The CHF data base consists of 10725 data points obtained from 147 test bundles with uniform axial power distributions and 29 test bundles with non-uniform axial power distributions. The local thermal-hydraulic conditions in the subchannels are calculated by employing a subchannel analysis code MATRA. The influence of turbulent mixing parameter on CHF is evaluated quantitatively for selected test bundles with representative cross sectional configurations. The performance of various CHF prediction models including empirical correlations for round tubes or rod bundles, theoretical DNB models such as sublayer dryout model and bubble crowding model, and CHF lookup table for round tubes, are assessed for the localized rod bundle CHF data base. In view of the analysis result, it reveals that the 1995 AECL-IPPE CHF lookup table method is one of promising models in the aspect of the prediction accuracy and the applicable range. As the result of analysis employing the CHF lookup table for 9113 data points with uniform axial heat profile, the mean and the standard deviation of P/M are calculated as 1.003 and 0.115 by HBM, 1.022 and 0.319 by DSM respectively

Experimental investigation of 150-KG-scale corium melt jet quenching in water

Energy Technology Data Exchange (ETDEWEB)

Magallon, D.; Hohmann, H.

1995-09-01

This paper compares and discusses the results of two large scale FARO quenching tests known as L-11 and L-14, which involved, respectively, 151 kg of W% 76.7 UO{sub 2} + 19.2 ZrO{sub 2} + 4.1 Zr and 125 kg of W% 80 UO{sub 2} + 20 ZrO{sub 2} melts poured into 600-kg, 2-m-depth water at saturation at 5.0 MPa. The results are further compared with those of two previous tests performed using a pure oxidic melt, respectively 18 and 44 kg of W% 80 UO{sub 2} + 20 ZrO{sub 2} melt quenched in 1-m-depth water at saturation at 5.0 MPa. In all the tests, significant breakup and quenching took place during the melt fall through the water. No steam explosion occurred. In the tests performed with a pure oxide UO{sub 2}-ZrO{sub 2} melt, part of the corium (from 1/6 to 1/3) did not breakup and reached the bottom plate still molten whatever the water depth was. Test L-11 data suggest that full oxidation and complete breakup of the melt occurred during the melt fall through the water. A proportion of 64% of the total energy content of the melt was released to the water during this phase ({approximately}1.5 s), against 44% for L-14. The maximum temperature increase of the bottom plate was 330 K (L-14). The mean particle size of the debris ranged between 2.5 and 4.8mm.

Electrical and Quench Performance of the First MICE Coupling Coil

International Nuclear Information System (INIS)

Tartaglia, M. A.; Carcagno, R.; Makulski, A.; Nogiec, Jerzy; Orris, D.; Pilipenko, R.; Sylvester, C.; Caspi, S.; Pan, H.; Prestemon, S.; Virostek, S.

2014-01-01

The first MICE Coupling Coil has been tested in a conduction-cooled environment in the new Solenoid Test Facility at Fermilab. We present an overview of the power and quench protection scheme, and report on the electrical and quench performance results obtained during cold power tests of the magnet

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

International Nuclear Information System (INIS)

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

1978-06-01

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

Passive quench arrest by a chimney induced deluge at every quench front

International Nuclear Information System (INIS)

Sydoriak, S.G.

1984-01-01

This chapter describes a magnet in which a growing quench stops itself spontaneously within a fraction of one winding turn because vapor in quench-heated channels generates a progressively increasing downflow of liquid in advance of each of the quench fronts. The downflow eventually becomes a deluge as the quench grows. The design of the multiple arrested quench magnet is discussed. It is shown how to construct a magnet so that if an arrested quench arises when it is at its highest operating current, peak nucleate boiling will exist in all quenching channels

Rod bundle burnout data and correlation comparisons

International Nuclear Information System (INIS)

Yoder, G.L.; Morris, D.G.; Mullins, C.B.

1985-01-01

Rod bundle burnout data from 30 steady-state and 3 transient tests were obtained from experiments performed in the Thermal Hydraulic Test Facility at the Oak Ridge National Laboratory. The tests covered a parameter range relevant to intact core reactor accidents ranging from large break to small break loss-ofcoolant conditions. Instrumentation within the 64-rod test section indicated that burnout occurred over an axial range within the bundle. The distance from the point where the first dry rod was detected to the point where all rods were dry was up to 60 cm in some of the tests. The burnout data should prove useful in developing new correlations for use in reactor thermalhydraulic codes. Evaluation of several existing critical heat flux correlations using the data show that three correlations, the Barnett, Bowring, and Katto correlations, perform similarly and correlate the data better than the Biasi correlation

Heat transfer in tube bundles subjected to blockages. Pt. 1

International Nuclear Information System (INIS)

Khattab, M.; Mariy, A.; Habib, M.

1983-01-01

The present work is carried out on unblocked test section bundle, half blocked, single ballooning and four ballooning blockages. The hydro-thermal performance of the bundle, (4x4) stainless steel, under each of the previous cases are studied. It is found that the existance of blockages increases the eddies and swirling flow streams. Furthermore, the average heat transfer in a bundle without blockages is superior than that with blockages. The percentage decrease of the average heat transfer coefficient with blockages depends on the position and shape of the blockage. Correlations describing average heat transfer, pressure drop and friction factor are established. All experimental tests are carried out under non-boiling region. (orig.) [de

Pressure loss in two-phase flow through a microchannel rod bundle

International Nuclear Information System (INIS)

Smith, A.C.; Hamm, L.L.; Qureshi, Z.; Steeper, T.J.

1998-01-01

The purpose of the microchannel rod bundle two-phase flow test described here was to provide data for benchmarking safety analyses for the accelerator production of tritium (APT). The objective was to obtain pressure loss data for a typical accelerator target rod bundle over a wide range of two-phase flow conditions. The test rod bundle assembly was fabricated for single-phase pressure drop tests conducted at Los Alamos National Laboratory (LANL) and subsequently used for the two-phase flow testing described here. The results for a typical case are given. These results fall generally in the slug flow regime for the horizontal flow results of Fukano and Kariyasaki for a 1.0-mm circular channel. Fukano and Kariyasaki found that surface tension effects were dominant in the 1-mm channel and report no churn regime. The results were also compared with the flow regime maps given by Triplett et al. for flow in discrete microchannels. Triplett employed both circular and trapezoidal channels, the latter to approximate the rod bundle interstitial flow channel shape. It was found that the rod bundle flow fell across the slug-to-churn flow regime transition reported by Triplett. This is consistent with the expectation that cross flow among channels would result in turbulent mixing and would suppress the formation of large discrete bubbles

International Benchmark based on Pressurised Water Reactor Sub-channel and Bundle Tests. Volume III: Departure from Nucleate Boiling

International Nuclear Information System (INIS)

Rubin, Adam; Avramova, Maria; Velazquez-Lozada, Alexander

2016-03-01

This report summarised the second phase of the Nuclear Energy Agency (NEA) and the Nuclear Regulatory Commission (NRC) Benchmark Based on NUPEC PWR Sub-channel and Bundle Tests (PSBT), which was intended to provide data for the verification of Departure from Nucleate Boiling (DNB) prediction in existing thermal-hydraulics codes and provide direction in the development of future methods. This phase was composed of three exercises; Exercise 1: fluid temperature benchmark, Exercise 2: steady-state rod bundle benchmark and Exercise 3: transient rod bundle benchmark. The experimental data provided to the participants of this benchmark is from a series of void measurement tests using full-size mock-up tests for both BWRs and PWRs. These tests were performed from 1987 to 1995 by the Nuclear Power Engineering Corporation (NUPEC) in Japan and made available by the Japan Nuclear Energy Safety Organisation (JNES) for the purposes of this benchmark, which was organised by Pennsylvania State University. Nine institutions from seven countries participated in this benchmark. Nine different computer codes were used in Exercise 1, 2 and 3. Among the computer codes were porous media, sub-channel and systems thermal-hydraulic code. The improvement between FLICA-OVAP (sub-channel) and FLICA (sub-channel) was noticeable. The main difference between the two was that FLICA-OVAP implicitly assigned flow regime based on drift flux, while FLICA assumes single phase flows. In Exercises 2 and 3, the codes were generally able to predict the Departure from Nucleate Boiling (DNB) power as well as the axial location of the onset of DNB (for the steady-state cases) and the time of DNB (for the transient cases). It was noted that the codes that used the Electric-Power-Research- Institute (EPRI) Critical-Heat-Flux (CHF) correlation had the lowest mean error in Exercise 2 for the predicted DNB power

Statistical flaw strength distributions for glass fibres: Correlation between bundle test and AFM-derived flaw size density functions

International Nuclear Information System (INIS)

Foray, G.; Descamps-Mandine, A.; R’Mili, M.; Lamon, J.

2012-01-01

The present paper investigates glass fibre flaw size distributions. Two commercial fibre grades (HP and HD) mainly used in cement-based composite reinforcement were studied. Glass fibre fractography is a difficult and time consuming exercise, and thus is seldom carried out. An approach based on tensile tests on multifilament bundles and examination of the fibre surface by atomic force microscopy (AFM) was used. Bundles of more than 500 single filaments each were tested. Thus a statistically significant database of failure data was built up for the HP and HD glass fibres. Gaussian flaw distributions were derived from the filament tensile strength data or extracted from the AFM images. The two distributions were compared. Defect sizes computed from raw AFM images agreed reasonably well with those derived from tensile strength data. Finally, the pertinence of a Gaussian distribution was discussed. The alternative Pareto distribution provided a fair approximation when dealing with AFM flaw size.

Cobra-IE Evaluation by Simulation of the NUPEC BWR Full-Size Fine-Mesh Bundle Test (BFBT)

International Nuclear Information System (INIS)

Burns, C. J.; Aumiler, D.L.

2006-01-01

The COBRA-IE computer code is a thermal-hydraulic subchannel analysis program capable of simulating phenomena present in both PWRs and BWRs. As part of ongoing COBRA-IE assessment efforts, the code has been evaluated against experimental data from the NUPEC BWR Full-Size Fine-Mesh Bundle Tests (BFBT). The BFBT experiments utilized an 8 x 8 rod bundle to simulate BWR operating conditions and power profiles, providing an excellent database for investigation of the capabilities of the code. Benchmarks performed included steady-state and transient void distribution, single-phase and two-phase pressure drop, and steady-state and transient critical power measurements. COBRA-IE effectively captured the trends seen in the experimental data with acceptable prediction error. Future sensitivity studies are planned to investigate the effects of enabling and/or modifying optional code models dealing with void drift, turbulent mixing, rewetting, and CHF

Experimental study on the effect of heat flux tilt on rod bundle dryout limitation

Energy Technology Data Exchange (ETDEWEB)

Sugawara, S; Terunuma, K; Kamoshida, H [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

1996-12-31

The effect of heat flux tilt on rod bundle dryout limitation was studied experimentally using a full-scale mock-up test facility and simulated 36-rod fuel bundles in which heater pins have azimuthal nonuniform heat flux distribution (i.e., heat flux tilt). Experimental results for typical lateral power distribution in the bundle indicate that the bundle dryout power with azimuthal heat flux tilt is higher than that without azimuthal heat flux tilt in the entire experimental range. Consequently, it is concluded that the dryout experiment using the test bundle with heater pins which has circumferentially uniform heat flux distribution gives conservative results for the usual lateral power distribution in a bundle in which the relative power of outermost-circle fuel rods is higher than those of middle- and inner-circle ones. (author). 15 refs., 2 tabs., 8 figs.

Experimental study on the effect of heat flux tilt on rod bundle dryout limitation

International Nuclear Information System (INIS)

Sugawara, S.; Terunuma, K.; Kamoshida, H.

1995-01-01

The effect of heat flux tilt on rod bundle dryout limitation was studied experimentally using a full-scale mock-up test facility and simulated 36-rod fuel bundles in which heater pins have azimuthal nonuniform heat flux distribution (i.e., heat flux tilt). Experimental results for typical lateral power distribution in the bundle indicate that the bundle dryout power with azimuthal heat flux tilt is higher than that without azimuthal heat flux tilt in the entire experimental range. Consequently, it is concluded that the dryout experiment using the test bundle with heater pins which has circumferentially uniform heat flux distribution gives conservative results for the usual lateral power distribution in a bundle in which the relative power of outermost-circle fuel rods is higher than those of middle- and inner-circle ones. (author). 15 refs., 2 tabs., 8 figs

Bundle Branch Block

Science.gov (United States)

... known cause. Causes can include: Left bundle branch block Heart attacks (myocardial infarction) Thickened, stiffened or weakened ... myocarditis) High blood pressure (hypertension) Right bundle branch block A heart abnormality that's present at birth (congenital) — ...

A model for dispersed flow heat transfer in rod bundles during reflood

International Nuclear Information System (INIS)

Wong, S.

1980-01-01

The present model calculates the heat transfer characteristics of the non-equilibrium dispersed droplet flow regime above the quench front during reflood by solving simultaneously the wall-to-vapor interactions, wall-to-droplet interactions and vapor-to-droplet interactions by an iterative numerical method. The unique feature in the present study is various heat transfer mechanisms are combined in an overall energy balance equation, and the convective heat transfer to vapor is obtained by calculating the vapor temperature distributions at the heated walls. The reactor rod bundle geometry, axial variations of vapor temperature and flow properties, radiative heat transfers, and enhancement of heat transfer due to turbulence are considered carefully, so that the present model could be used to predict PWR (Pressurized Water Reactor) reflood heat transfers, and hence the fuel cladding wall temperature transients. In order to achieve closure of the problem formulations, the droplet size and its motion are determined from the FLECHT (Full Length Emergency Cooling Heat Transfer Program) low flooding rate series consine axial power shape test data. The model is then verified by comparing the heat transfer predictions with FLECHT low flooding rate series skewed axial power shape test data. Comparisons of predictions with data show satisfactory agreements

The impact of Lean bundles on hospital performance: does size matter?

Science.gov (United States)

Al-Hyari, Khalil; Abu Hammour, Sewar; Abu Zaid, Mohammad Khair Saleem; Haffar, Mohamed

2016-10-10

Purpose The purpose of this paper is to study the effect of the implementation of Lean bundles on hospital performance in private hospitals in Jordan and evaluate how much the size of organization can affect the relationship between Lean bundles implementation and hospital performance. Design/methodology/approach The research is considered as quantitative method (descriptive and hypothesis testing). Three statistical techniques were adopted to analyse the data. Structural equation modeling techniques and multi-group analysis were used to examine the research's hypothesis, and to perform the required statistical analysis of the data from the survey. Reliability analysis and confirmatory factor analysis were used to test the construct validity, reliability and measurement loadings that were performed. Findings Lean bundles have been identified as an effective approach that can dramatically improve the organizational performance of private hospitals in Jordan. Main Lean bundles - just in time, human resource management, and total quality management are applicable to large, small and medium hospitals without significant differences in advantages that depend on size. Originality/value According to the researchers' best knowledge, this is the first research that studies the impact of Lean bundles implementation in healthcare sector in Jordan. This research also makes a significant contribution for decision makers in healthcare to increase their awareness of Lean bundles.

Single-Phase Bundle Flows Including Macroscopic Turbulence Model

Energy Technology Data Exchange (ETDEWEB)

Lee, Seung Jun; Yoon, Han Young [KAERI, Daejeon (Korea, Republic of); Yoon, Seok Jong; Cho, Hyoung Kyu [Seoul National University, Seoul (Korea, Republic of)

2016-05-15

To deal with various thermal hydraulic phenomena due to rapid change of fluid properties when an accident happens, securing mechanistic approaches as much as possible may reduce the uncertainty arising from improper applications of the experimental models. In this study, the turbulence mixing model, which is well defined in the subchannel analysis code such as VIPRE, COBRA, and MATRA by experiments, is replaced by a macroscopic k-e turbulence model, which represents the aspect of mathematical derivation. The performance of CUPID with macroscopic turbulence model is validated against several bundle experiments: CNEN 4x4 and PNL 7x7 rod bundle tests. In this study, the macroscopic k-e model has been validated for the application to subchannel analysis. It has been implemented in the CUPID code and validated against CNEN 4x4 and PNL 7x7 rod bundle tests. The results showed that the macroscopic k-e turbulence model can estimate the experiments properly.

FPGA-based quench detection system for super-FRS super-ferric dipole prototype

International Nuclear Information System (INIS)

Yang Tongjun; Wu Wei; Yao Qinggao; Yuan Ping; He Yuan; Han Shaofei; Ma Lizhen

2011-01-01

The quench detection system for Super-FRS super-ferric dipole prototype magnet of FAIR has been designed and built. The balance bridge was used to detect quench signal. In order to avoid blind zone of quench detection, two independent bridges were used. NI PXI-7830R FPGA was used to implement filter to quench signal and algorithm of quench decision and to produce quench trigger signal. Pre-sample technique was used in quench data acquisition. The data before and after quench could be recorded for analysis later. The test result indicated that the quench of the dipole's superconducting coil could be reliably detected by the quench detection module. (authors)

Quench protection and safety of the ATLAS central solenoid

CERN Document Server

Makida, Y; Haruyama, T; ten Kate, H H J; Kawai, M; Kobayashi, T; Kondo, T; Kondo, Y; Mizumaki, S; Olesen, G; Sbrissa, E; Yamamoto, A; Yamaoka, H

2002-01-01

Fabrication of the ATLAS central solenoid was completed and the performance test has been carried out. The solenoid was successfully charged up to 8.4 kA, which is 10% higher than the normal operational current of 7.6 kA. Two methods for quench protection, pure aluminum strips accelerating quench propagation and quench protection heaters distributing normal zones, are applied in order to safely dissipate the stored energy. In this paper, quench characteristics and protection methods of the ATLAS central solenoid are described. (14 refs).

ZnSe quantum dots based fluorescence quenching method for determination of paeoniflorin

International Nuclear Information System (INIS)

Chen, Zhi; Chen, Jiayi; Liang, Qiaowen; Wu, Dudu; Zeng, Yuaner; Jiang, Bin

2014-01-01

Water soluble ZnSe quantum dots (QDs) modified by mercaptoacetic acid (MAA) were used to determinate paeoniflorin in aqueous solutions by the fluorescence spectroscopic technique. The results showed that the fluorescence of the modified ZnSe QDs could be quenched by paeoniflorin effectively in physiological buffer solution. The optimum fluorescence intensity was found to be at incubation time 10 min, pH 7.0 and temperature 25 °C. Under the optimal conditions, the detection limit of paeoniflorin was 7.30×10 −7 mol L −1 . Moreover, the quenching mechanism was discussed to be a static quenching procedure, which was proved by quenching rate constant K q (1.02×10 13 L mol −1 s −1 ). -- Highlights: • The fluorescence intensity of ZnSe QDs could be quenched by paeoniflorin. • Foreign substance showed insignificant effect for determination of paeoniflorin. • The quenching mechanism was discussed to be a static quenching procedure

ZnSe quantum dots based fluorescence quenching method for determination of paeoniflorin

Energy Technology Data Exchange (ETDEWEB)

Chen, Zhi [Center of Analysis, Guangdong Medical College, Dongguan 523808 (China); School of Chinese Herbal Medicine, Guangzhou University of Chinese Medicine, Guangzhou 510006 (China); Chen, Jiayi; Liang, Qiaowen [School of Chinese Herbal Medicine, Guangzhou University of Chinese Medicine, Guangzhou 510006 (China); Wu, Dudu [Center of Analysis, Guangdong Medical College, Dongguan 523808 (China); Zeng, Yuaner, E-mail: zengyuaner@126.com [School of Chinese Herbal Medicine, Guangzhou University of Chinese Medicine, Guangzhou 510006 (China); Jiang, Bin, E-mail: gzjiangbin@hotmail.com [School of Chinese Herbal Medicine, Guangzhou University of Chinese Medicine, Guangzhou 510006 (China)

2014-01-15

Water soluble ZnSe quantum dots (QDs) modified by mercaptoacetic acid (MAA) were used to determinate paeoniflorin in aqueous solutions by the fluorescence spectroscopic technique. The results showed that the fluorescence of the modified ZnSe QDs could be quenched by paeoniflorin effectively in physiological buffer solution. The optimum fluorescence intensity was found to be at incubation time 10 min, pH 7.0 and temperature 25 °C. Under the optimal conditions, the detection limit of paeoniflorin was 7.30×10{sup −7} mol L{sup −1}. Moreover, the quenching mechanism was discussed to be a static quenching procedure, which was proved by quenching rate constant K{sub q} (1.02×10{sup 13} L mol{sup −1} s{sup −1}). -- Highlights: • The fluorescence intensity of ZnSe QDs could be quenched by paeoniflorin. • Foreign substance showed insignificant effect for determination of paeoniflorin. • The quenching mechanism was discussed to be a static quenching procedure.

Exposure Control Using Adaptive Multi-Stage Item Bundles.

Science.gov (United States)

Luecht, Richard M.

This paper presents a multistage adaptive testing test development paradigm that promises to handle content balancing and other test development needs, psychometric reliability concerns, and item exposure. The bundled multistage adaptive testing (BMAT) framework is a modification of the computer-adaptive sequential testing framework introduced by…

The bundles of algebraic and Dirac-Hestenes spinor fields

International Nuclear Information System (INIS)

Mosna, Ricardo A.; Rodrigues, Waldyr A. Jr.

2004-01-01

Our main objective in this paper is to clarify the ontology of Dirac-Hestenes spinor fields (DHSF) and its relationship with even multivector fields, on a Riemann-Cartan spacetime (RCST) M=(M,g,∇,τ g ,↑) admitting a spin structure, and to give a mathematically rigorous derivation of the so-called Dirac-Hestenes equation (DHE) in the case where M is a Lorentzian spacetime (the general case when M is a RCST will be discussed in another publication). To this aim we introduce the Clifford bundle of multivector fields (Cl(M,g)) and the left (Cl Spin 1,3 e l (M)) and right (Cl Spin 1,3 e r (M)) spin-Clifford bundles on the spin manifold (M,g). The relation between left ideal algebraic spinor fields (LIASF) and Dirac-Hestenes spinor fields (both fields are sections of Cl Spin 1,3 e l (M)) is clarified. We study in detail the theory of covariant derivatives of Clifford fields as well as that of left and right spin-Clifford fields. A consistent Dirac equation for a DHSF Ψ is a member of sec Cl Spin 1,3 e l (M) (denoted DECl l ) on a Lorentzian spacetime is found. We also obtain a representation of the DECl l in the Clifford bundle Cl(M,g). It is such equation that we call the DHE and it is satisfied by Clifford fields ψ Ξ is a member of sec Cl(M,g). This means that to each DHSF Ψ is a member of sec Cl Spin 1,3 e l (M) and spin frame Ξ is a member of sec P Spin 1,3 e (M), there is a well-defined sum of even multivector fields ψ Ξ isa member of sec Cl(M,g) (EMFS) associated with Ψ. Such an EMFS is called a representative of the DHSF on the given spin frame. And, of course, such a EMFS (the representative of the DHSF) is not a spinor field. With this crucial distinction between a DHSF and its representatives on the Clifford bundle, we provide a consistent theory for the covariant derivatives of Clifford and spinor fields of all kinds. We emphasize that the DECl l and the DHE, although related, are equations of different mathematical natures. We study also the

Safety assessment for the CANFLEX-NU fuel bundles with respect to the 37-element fuel bundles

Energy Technology Data Exchange (ETDEWEB)

Suk, H. C.; Lim, H. S. [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-11-01

The KAERI and AECL have jointly developed an advanced CANDU fuel, called CANFLEX-NU fuel bundle. CANFLEX 43-element bundle has some improved features of increased operating margin and enhanced safety compared to the existing 37-element bundle. Since CANFLEX fuel bundle is designed to be compatible with the CANDU-6 reactor design, the behaviour in the thermalhydraulic system will be nearly identical with 37-element bundle. But due to different element design and linear element power distribution between the two bundles, it is expected that CANFLEX fuel behaviour would be different from the behaviour of the 37-element fuel. Therefore, safety assessments on the design basis accidents which result if fuel failures are performed. For all accidents selected, it is observed that the loading of CANFLEX bundle in an existing CANDU-6 reactor would not worsen the reactor safety. It is also predicted that fission product release for CANFLEX fuel bundle generally is lower than that for 37-element bundle. 3 refs., 2 figs., 2 tabs. (Author)

BDI behavior evaluation of an upgraded Monju core and a demonstration core. (1) Plans for the out of pile bundle compressive tests for large diameter pins

International Nuclear Information System (INIS)

Ichikawa, Shoichi; Haga, Hiroyuki; Katsuyama, Kozo; Uwaba, Tomoyuki; Maeda, Koji; Nishinoiri, Kenji

2012-07-01

The life of FBR (Fast Breeder Reactor) fuel assembly is restricted by BDI (Bundle-Duct Interaction). Therefore, it is very important to carry out the out pile bundle compressive tests which can imitate BDI, in order to evaluate BDI behavior. The target of the conventional BDI behavior was small diameter pins (φ6.5mm) for fuel pellets which were used with the assembly of Monju (the Monju prototype fast breeder reactor) etc. Furthermore by an upgraded Monju core and a demonstration core, adoption of large diameter pins for the holler annular pellets is planned. Therefore, it was necessary to carry out BDI evaluation of a large diameter pin. Then, the plans for out of pile bundle compressive test for large diameter pins were are reported. (author)

TASK 2: QUENCH ZONE SIMULATION

Energy Technology Data Exchange (ETDEWEB)

Fusselman, Steve

2015-09-30

Aerojet Rocketdyne (AR) has developed an innovative gasifier concept incorporating advanced technologies in ultra-dense phase dry feed system, rapid mix injector, and advanced component cooling to significantly improve gasifier performance, life, and cost compared to commercially available state-of-the-art systems. A key feature of the AR gasifier design is the transition from the gasifier outlet into the quench zone, where the raw syngas is cooled to ~ 400°C by injection and vaporization of atomized water. Earlier pilot plant testing revealed a propensity for the original gasifier outlet design to accumulate slag in the outlet, leading to erratic syngas flow from the outlet. Subsequent design modifications successfully resolved this issue in the pilot plant gasifier. In order to gain greater insight into the physical phenomena occurring within this zone, AR developed a cold flow simulation apparatus with Coanda Research & Development with a high degree of similitude to hot fire conditions with the pilot scale gasifier design, and capable of accommodating a scaled-down quench zone for a demonstration-scale gasifier. The objective of this task was to validate similitude of the cold flow simulation model by comparison of pilot-scale outlet design performance, and to assess demonstration scale gasifier design feasibility from testing of a scaled-down outlet design. Test results did exhibit a strong correspondence with the two pilot scale outlet designs, indicating credible similitude for the cold flow simulation device. Testing of the scaled-down outlet revealed important considerations in the design and operation of the demonstration scale gasifier, in particular pertaining to the relative momentum between the downcoming raw syngas and the sprayed quench water and associated impacts on flow patterns within the quench zone. This report describes key findings from the test program, including assessment of pilot plant configuration simulations relative to actual

Numerical simulation of quench protection for a 1.5 T persistent mode MgB2 conduction-cooled MRI magnet

Science.gov (United States)

Deissler, Robert J.; Baig, Tanvir; Poole, Charles; Amin, Abdullah; Doll, David; Tomsic, Michael; Martens, Michael

2017-02-01

The active quench protection of a 1.5 T MgB2 conduction-cooled MRI magnet operating in persistent current mode is considered. An active quench protection system relies on the detection of the resistive voltage developed in the magnet, which is used to trigger the external energizing of quench heaters located on the surfaces of all ten coil bundles. A numerical integration of the heat equation is used to determine the development of the temperature profile and the maximum temperature in the coil at the origin, or ‘hot spot’, of the quench. Both n-value of the superconductor and magnetoresistance of the wire are included in the simulations. An MgB2 wire manufactured by Hyper Tech Research, Inc. was used as the basis to model the wire for the simulations. With the proposed active quench protection system, the maximum temperature was limited to 200 K or less, which is considered low enough to prevent damage to the magnet. By substituting Glidcop for the Monel in the wire sheath or by increasing the thermal conductivity of the insulation, the margin for safe operation was further increased, the maximum temperature decreasing by more than 40 K. The strain on the MgB2 filaments is calculated using ANSYS, verifying that the stress and strain limits in the MgB2 superconductor and epoxy insulation are not exceeded.

Electrochemical characteristics of bundle-type silicon nanorods as an anode material for lithium ion batteries

International Nuclear Information System (INIS)

Nguyen, Si Hieu; Lim, Jong Choo; Lee, Joong Kee

2012-01-01

Highlights: ► A metal-assisted chemical etching technique was performed on Si thin films. ► The etching process resulted in the formation of bundle-type Si nanorods. ► The morphology of Si electrodes closely relate to electrochemical characteristics. - Abstract: In order to prepare bundle-type silicon nanorods, a silver-assisted chemical etching technique was used to modify a 1.6 μm silicon thin film, which was deposited on Cu foil by Electron Cyclotron Resonance Plasma Enhanced Chemical Vapor Deposition. The bundle-type silicon nanorods on Cu foil were employed as anodes for a lithium secondary battery, without further treatment. The electrochemical characteristics of the pristine silicon thin film anodes and the bundle-type silicon nanorod anodes are different from one another. The electrochemical performance of the bundle-type silicon nanorod anodes exceeded that of the pristine Si thin film anodes. The specific capacity of the bundle-type silicon nanorod anodes is much higher than 3000 mAh g −1 at the first charge (Li insertion) cycle. The coulombic efficiency of bundle-type silicon anodes was stable at more than 97%, and the charge capacity remained at 1420 mAh g −1 , even after 100 cycles of charging and discharging. The results from the differential voltage analysis showed a side reaction at around 0.44–0.5 V, and the specific potential of this side reaction decreased after each cycle. The apparent diffusion coefficients of the two anode types were in the range of 10 −13 –10 −16 cm 2 s −1 in the first cycle. In subsequent charge cycles, these values for the silicon thin film anodes and the silicon nanorod bundle anode were approximately 10 −12 –10 −14 and 10 −13 –10 −15 cm 2 s −1 , respectively.

Results of heater induced quenches on a 1-m SSC model dipole

International Nuclear Information System (INIS)

Hassenzahl, W.V.

1985-10-01

This report describes the results of a series of heater induced quenches on the 1-m long SSC model dipole D-12C-7 constructed at LBL. Test results of the following types are described: quench propagation velocities - axial; quench propagation velocities - transverse; and rate of temperature rise in the conductor. The primary purpose of these tests was to measure quench velocities at a variety of locations and for several currents/fields which can be used to refine the quench predictions for longer magnets. Because of limited data in the low field region of this magnet, it is recommended that it be retested with additional voltage taps. 20 figs., 6 tabs

SIKAP KONSUMEN TERHADAP PRODUK BUNDLING AGRIBISNIS

Directory of Open Access Journals (Sweden)

Didi Junaedi

2017-04-01

implementation to Dekalb brand hybrid corn and Round-up brand herbicide. By analyzes how consumer attitudes toward buying intention in this regard farmers as buyer and retailers as products services. The data used is primary data. Primary data is obtained using 2 kind of respondents are retailers and farmers. The data obtained by distributed 30 questionnaires for retailers and 110 farmers in Grobogan. The descriptive statistic employed to analyzed data by using multiple linear regressions with t test, F test and coefficient of determination. The result showed that on retailers respondents attribute the product bundling has no significant influence to consumer buying intention but consumer attitudes significantly influence the buying intention. On the farmers respondents showed that attributes of the product bundling and consumer attitudes positive and significant influence to buying intention.

Constructing co-Higgs bundles on CP^2

OpenAIRE

Rayan, Steven

2013-01-01

On a complex manifold, a co-Higgs bundle is a holomorphic vector bundle with an endomorphism twisted by the tangent bundle. The notion of generalized holomorphic bundle in Hitchin's generalized geometry coincides with that of co-Higgs bundle when the generalized complex manifold is ordinary complex. Schwarzenberger's rank-2 vector bundle on the projective plane, constructed from a line bundle on the double cover CP^1 \\times CP^1 \\to CP^2, is naturally a co-Higgs bundle, with the twisted endom...

Quench observation using quench antennas on RHIC IR quadrupole magnets

International Nuclear Information System (INIS)

Ogitsu, T.; Terashima, A.; Tsuchiya, K.; Ganetis, G.; Muratore, J.; Wanderer, P.

1995-01-01

Quench observation using quench antennas is now being performed routinely on RHIC dipole and quadrupole magnets. Recently, a quench antenna was used on a RHIC IR magnet which is heavily instrumented with voltage taps. It was confirmed that the signals detected in the antenna coils do not contradict the voltage tap signals. The antenna also detects a sign of mechanical disturbance which could be related to a training quench. This paper summarizes signals detected in the antenna and discusses possible causes of these signals

Quench observation using quench antennas on RHIC IR quadrupole magnets

International Nuclear Information System (INIS)

Ogitsu, T.; Terashima, A.; Tsuchiya, K.; Ganetis, G.; Muratore, J.; Wanderer, P.

1996-01-01

Quench observation using quench antennas is now being performed routinely on RHIC dipole and quadrupole magnets. Recently, a quench antenna was used on a RHIC IR magnet which is heavily instrumented with voltage taps. It was confirmed that the signals detected in the antenna coils do not contradict the voltage tap signals. The antenna also detects a sign of mechanical disturbance which could be related to a training quench. This paper summarizes signals detected in the antenna and discusses possible causes of these signals

Corium quench in deep pool mixing experiments

International Nuclear Information System (INIS)

Spencer, B.W.; McUmber, L.; Gregorash, D.; Aeschlimann, R.; Sienicki, J.J.

1985-01-01

The results of two recent corium-water thermal interaction (CWTI) tests are described in which a stream of molten corium was poured into a deep pool of water in order to determine the mixing behavior, the corium-to-water heat transfer rates, and the characteristic sizes of the quenched debris. The corium composition was 60% UO 2 , 16% ZrO 2 , and 24% stainless steel by weight; its initial temperature was 3080 K, approx.160 K above the oxide phase liquidus temperature. The corium pour stream was a single-phase 2.2 cm dia liquid column which entered the water pool in film boiling at approx.4 m/s. The water subcooling was 6 and 75C in the two tests. Test results showed that with low subcooling, rapid steam generation caused the pool to boil up into a high void fraction regime. In contrast, with large subcooling no net steam generation occurred, and the pool remained relatively quiescent. Breakup of the jet appeared to occur by surface stripping. In neither test was the breakup complete during transit through the 32 cm deep water pool, and molten corium channeled to the base where it formed a melt layer. The characteristic heat transfer rates measured 3.5 MJ/s and 2.7 MJ/s during the fall stage for small and large subcooling, respectively; during the initial stage of bed quench, the surface heat fluxes measured 2.4 MW/m 2 and 3.7 MW/m 2 , respectively. A small mass of particles was formed in each test, measuring typically 0.1 to 1 mm and 1 to 5 mm dia for the large and small subcooling conditions, respectively. 9 refs., 13 figs., 1 tab

Calculating Quench Propagation with ANSYS(regsign)

International Nuclear Information System (INIS)

Caspi, S.; Chiesa, L.; Ferracin, P.; Gourlay, S.A.; Hafalia, R.; Hinkins, R.; Lietzke, A.F.; Prestemon, S.

2002-01-01

A commercial Finite-Element-Analysis program, ANSYS(reg s ign), is widely used in structural and thermal analysis. With the program's ability to include non-linear material properties and import complex CAD files, one can generate coil geometries and simulate quench propagation in superconducting magnets. A 'proof-of-principle' finite element model was developed assuming a resistivity that increases linearly from zero to its normal value at a temperature consistent with the assumed B magnetic field. More sophisticated models could easily include finer-grained coil, cable, structural, and circuit details. A quench is provoked by raising the temperature of an arbitrary superconducting element above its T c . The time response to this perturbation is calculated using small time-steps to allow convergence between steps. Snapshots of the temperature and voltage distributions allow examination of longitudinal and turn-to-turn quench propagation, quench-front annihilation, and cryo-stability. Modeling details are discussed, and a computed voltage history was compared with measurements from a recent magnet test.

Quench start localization in full-length SSC R ampersand D dipoles

International Nuclear Information System (INIS)

Devred, A.; Chapman, M.; Cortella, J.; Desportes, A.; Kaugerts, J.; Kirk, T.; Mirk, K.; Schermer, R.; Tompkins, J.C.; Turner, J.; Bleadon, M.; Brown, B.C.; Hanft, R.; Kuchnir, M.; Lamm, M.; Mantsch, P.; Mazur, P.O.; Orris, D.; Peoples, J.; Strait, J.; Tool, G.; Caspi, S.; Gilbert, W.; Meuser, R.; Peters, C.; Rechen, J.; Royet, J.; Scanlan, R.; Taylor, C.; Zbasnik, J.

1989-04-01

Full-length SSC R ampersand D dipole magnets instrumented with four voltage taps on each turn of the inner quarter coils have been tested. These voltage taps enable accurate location of the point at which the quenches start and detailed studies of quench development in the coil. Attention here is focused on localizing the quench source. After recalling the basic mechanism of a quench (why it occurs and how it propagates), the method of quench origin analysis is described: the quench propagation velocity on the turn where the quench occurs is calculated, and the quench location is then verified by reiterating the analysis on the adjacent turns. Last, the velocity value, which appears to be higher than previously measured, is discussed

Optimization of a quench detection system for superconducting magnets

International Nuclear Information System (INIS)

Borlein, M.

2004-12-01

Subject of this report is the detection of a quench in a superconducting magnet. For the safe operation of superconducting magnets one of the most important issues is the quench detection system which controls the superconducting state of the magnet and triggers a safety discharge if necessary. If it comes to a breakdown of the superconductivity (quench), the magnet has to be discharged very quickly to avoid any damage or danger for the magnet or its environment. First an introducing overview is given. Next different methods of quench detection will be presented, partially on the basis of existing quench detection systems and the applicability of these methods in different states of the magnet operation will be shown. The different quench detection methods are compared and evaluated partially by using test experiments described in the appendix. As an application example this report contains a proposal for the quench detection system for the Wendelstein 7-X facility, actually built by the Institute for Plasma Physics, Garching [de

Analyses of HANARO bundle experiment data using MATRA-h: revision

Energy Technology Data Exchange (ETDEWEB)

Lim, In Cheol; Park, Cheol; Chae, Hee Taek; Lee, Choong Sung

1999-08-01

When the construction and operation license for HANARO was renewed in 1995, imposed was a condition that the safety limit CHFR should have the margin of 25 percent. The reason for this were that the number of bundle CHF experiment data was not enough for the validation of the prediction of CHF in bundle geometry and that the ability of COBRA/KMRR to prediction the local coolant condition was not fully validated. For the resolution of this imposition, more bundle CHF data were gathered and the subchannel exit temperature distribution was obtained during the in-core irradiation test of instrumented bundle (Type-B bundle). also, for these experimental data, subchannel analyses were performed by using MATRA-h code which is the modified version of MATRA-a which is a modified version of KAERI's MATRA-a for the application to HANARO. By comparing the analysis results with the experimental results, it was found that the HANARO subchannel analysis method would give the conservative or best-estimated predictions for the CHF in bundle geometry. This report is the revision of KAERI/TR-1090/98 on the analysis of bundle experiment data using MATRA-h. (Author). 16 refs., 16 tabs., 25 figs.

Development of a Fast Breeder Reactor Fuel Bundle Deformation Analysis Code - BAMBOO: Development of a Pin Dispersion Model and Verification by the Out-of-Pile Compression Test

International Nuclear Information System (INIS)

Uwaba, Tomoyuki; Ito, Masahiro; Ukai, Shigeharu

2004-01-01

To analyze the wire-wrapped fast breeder reactor fuel pin bundle deformation under bundle/duct interaction conditions, the Japan Nuclear Cycle Development Institute has developed the BAMBOO computer code. This code uses the three-dimensional beam element to calculate fuel pin bowing and cladding oval distortion as the primary deformation mechanisms in a fuel pin bundle. The pin dispersion, which is disarrangement of pins in a bundle and would occur during irradiation, was modeled in this code to evaluate its effect on bundle deformation. By applying the contact analysis method commonly used in the finite element method, this model considers the contact conditions at various axial positions as well as the nodal points and can analyze the irregular arrangement of fuel pins with the deviation of the wire configuration.The dispersion model was introduced in the BAMBOO code and verified by using the results of the out-of-pile compression test of the bundle, where the dispersion was caused by the deviation of the wire position. And the effect of the dispersion on the bundle deformation was evaluated based on the analysis results of the code

Polycation induced actin bundles

OpenAIRE

Muhlrad, Andras; Grintsevich, Elena E.; Reisler, Emil

2011-01-01

Three polycations, polylysine, the polyamine spermine and the polycationic protein lysozyme were used to study the formation, structure, ionic strength sensitivity and dissociation of polycation-induced actin bundles. Bundles form fast, simultaneously with the polymerization of MgATP-G-actins, upon addition of polycations to solutions of actins at low ionic strength conditions. This indicates that nuclei and/or nascent filaments bundle due to attractive, electrostatic effect of polycations an...

Single and two-phase flow pressure drop for CANFLEX bundle

Energy Technology Data Exchange (ETDEWEB)

Park, Joo Hwan; Jun, Ji Su; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Dimmick, G R; Bullock, D E [Atomic Energy of Canada Limited, Ontario (Canada)

1999-12-31

Friction factor and two-phase flow frictional multiplier for a CANFLEX bundle are newly developed and presented in this paper. CANFLEX as a 43-element fuel bundle has been developed jointly by AECL/KAERI to provide greater operational flexibility for CANDU reactor operators and designers. Friction factor and two-phase flow frictional multiplier have been developed by using the experimental data of pressure drops obtained from two series of Freon-134a (R-134a) CHF tests with a string of simulated CANFLEX bundles in a single phase and a two-phase flow conditions. The friction factor for a CANFLEX bundle is found to be about 20% higher than that of Blasius for a smooth circular pipe. The pressure drop predicted by using the new correlations of friction factor and two-phase frictional multiplier are well agreed with the experimental pressure drop data of CANFLEX bundle within {+-} 5% error. 11 refs., 5 figs. (Author)

Single and two-phase flow pressure drop for CANFLEX bundle

Energy Technology Data Exchange (ETDEWEB)

Park, Joo Hwan; Jun, Ji Su; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Dimmick, G. R.; Bullock, D. E. [Atomic Energy of Canada Limited, Ontario (Canada)

1998-12-31

Friction factor and two-phase flow frictional multiplier for a CANFLEX bundle are newly developed and presented in this paper. CANFLEX as a 43-element fuel bundle has been developed jointly by AECL/KAERI to provide greater operational flexibility for CANDU reactor operators and designers. Friction factor and two-phase flow frictional multiplier have been developed by using the experimental data of pressure drops obtained from two series of Freon-134a (R-134a) CHF tests with a string of simulated CANFLEX bundles in a single phase and a two-phase flow conditions. The friction factor for a CANFLEX bundle is found to be about 20% higher than that of Blasius for a smooth circular pipe. The pressure drop predicted by using the new correlations of friction factor and two-phase frictional multiplier are well agreed with the experimental pressure drop data of CANFLEX bundle within {+-} 5% error. 11 refs., 5 figs. (Author)

CFD analysis of flow and heat transfer in Canadian supercritical water reactor bundle

International Nuclear Information System (INIS)

Podila, K.; Rao, Y.F.

2015-01-01

Highlights: • Flow and heat transfer in SCWR fuel bundle design by AECL is studied using CFD. • Bare-rod bundle geometry is tested at 23.5, 25 and 28 MPa using STAR-CCM+ code. • SST k–ω low-Re model was used to study occurrence of heat transfer deterioration. - Abstract: Within the Gen-IV International Forum, AECL is leading the effort in developing a conceptual design for the Canadian SCWR. AECL proposed a new fuel bundle design with two rings of fuel elements placed between central flow tube and the pressure tube. In line with the scope of the conceptual design, the objective of the present CFD work is to aid in developing a bundle heat transfer correlation for the Canadian SCWR fuel bundle design. This paper presents results from an ongoing effort in determining the conditions favorable for occurrence of HTD in the supercritical bundle flows. In the current investigation, bare-rod bundle geometry was tested for the proposed fuel bundle design at 23.5, 25 and 28 MPa using STAR-CCM+ CFD code. Taking advantage of the design symmetry of the fuel bundle, only 1/32 of the computational domain was simulated. The low-Reynolds number modification of SST k–ω turbulence model along with y + < 1 was used in the simulations. For lower mass flow simulations, the increase of inlet temperature and operational pressure was found effective in reducing the occurrence of HTD. For higher mass flow simulations, normal heat transfer behaviour was observed except for the lower pressure range (23.5 MPa)

Evaluation report on SCTF Core-II test S2-19

International Nuclear Information System (INIS)

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

1991-03-01

Experimental studies using Slab Core Test Facility (SCTF) have revealed that the heat transfer enhancement in higher power bundles is mainly governed by the radial power ratio in core during the reflood in PWR-LOCA. As a physical mechanism for the heat transfer enhancement, it can be considered from the experimental evidence that the increase of upward steam flow rate in a higher power bundle which is caused by the higher steam production rate in the bundle gives the higher upward liquid flow rate in the bundle and the increase of the liquid flow rate gives the heat transfer enhancement. In order to develop a mechanistic model for the heat transfer enhancement based on this idea, the following relations should be identified quantitatively: (1) Relation between the steam production rate and the upward liquid flow rate, (2) Cross flow rate above the quench front and (3) Relation between the degree of heat transfer enhancement due to radial power ratio and the amount of increase of upward liquid flow rate. In this report, the above relation (3) was investigated experimentally as a step to develop the mechanistic model using the SCTF where the relation between the radial power ratio and the heat transfer enhancement has been made clear quantitatively. The degree of increase of heat transfer between two forced feed tests with the different flow rate in LPCI period was compared with the degree of heat transfer enhancement under a radial power ratio in the previous SCTF tests. The two forced feed tests were performed under the condition without any significant two-dimensional hydraulic behavior in core. The ratio of the mass flow rate between the two tests was about double. (author)

MYOCARDIAL DEFORMATION AND COMPLETE LEFT BUNDLE BRANCH BLOCK

Directory of Open Access Journals (Sweden)

E. N. Pavlyukova

2015-12-01

Full Text Available Tissue Doppler imaging is evolving as a useful echocardiographic tool for quantitative assessment of left ventricular systolic and diastolic function. Over the last 10 years, myocardial deformation imaging has become possible initially with tissue Doppler , and more recently with myocardial speckle-tracking using 2D echocardiography. Unlike simple tissue velocity measurements, deformation measurements are specific for the region of interest. Strain rate or strain measurements have been used as sensitive indicators for subclinical diseases, and it is the most widely used tool to assess mechanical dyssynchrony. Left bundle branch block is a frequent, etiologically heterogeneous, clinically hostile and diagnostically challenging entity. About 2% of patients underwent cardiac stress testing show stable or intermittent left bundle branch block. Presence of left bundle branch block is associated with a lower and slower diastolic coronary flow velocity especially during hyperemia. Stress echocardiography is the best option for the diagnosis of ischemic heart disease, albeit specificity and sensitivity reduce in patients with left bundle branch block in the territory of left anterior descending artery in presence of initial septum dyskinesia.

Determination of quenching coefficients by time resolved emission spectroscopy

International Nuclear Information System (INIS)

Gans, T.; Schulz-von der Gathen, V.; Doebele, H.F.

2001-01-01

Capacitively coupled RF discharges (CCRF discharges) at 13.56 MHz in hydrogen exhibit a field reversal phase of about 10 ns during which an intense electron current provides collisional excitation, within the sheath region. After this strongly dominant short pulsed electron impact excitation, it is possible to determine quenching coefficients from the lifetime of the fluorescence at various pressures by time resolved OES even for high energy levels and without any restrictions of optical selection rules. This novel technique allows the measurement of quenching coefficients for atomic and molecular emission lines of hydrogen itself, as well as for emission lines of small admixtures (e.g. noble gases) to the hydrogen discharge, since with a fast gate-able ICCD camera operating at 13.56 MHz it is possible to measure even faint emission lines temporally resolved

Determination of quenching coefficients by time resolved emission spectroscopy

Energy Technology Data Exchange (ETDEWEB)

Gans, T.; Schulz-von der Gathen, V.; Doebele, H.F. [Essen Univ. (Gesamthochschule) (Germany). Inst. fuer Laser- und Plasmaphysik

2001-07-01

Capacitively coupled RF discharges (CCRF discharges) at 13.56 MHz in hydrogen exhibit a field reversal phase of about 10 ns during which an intense electron current provides collisional excitation, within the sheath region. After this strongly dominant short pulsed electron impact excitation, it is possible to determine quenching coefficients from the lifetime of the fluorescence at various pressures by time resolved OES even for high energy levels and without any restrictions of optical selection rules. This novel technique allows the measurement of quenching coefficients for atomic and molecular emission lines of hydrogen itself, as well as for emission lines of small admixtures (e.g. noble gases) to the hydrogen discharge, since with a fast gate-able ICCD camera operating at 13.56 MHz it is possible to measure even faint emission lines temporally resolved.

Effect of Quenching Media on Mechanical Properties of Medium Carbon Steel 1030

Directory of Open Access Journals (Sweden)

Khansaa Dawood Salman

2018-01-01

Full Text Available This investigation aims to study the effect of quenching media (water, oil, Poly Vinyl Chloride PVC on mechanical properties of 1030 steel. The applications of this steel include machinery parts where strength and hardness are requisites. The steel is heated to about 950  and soaked for 1hr in electrical furnace and then quenched in different quenching medium such as water, oil and poly vinyl chloride. After heat treatment by quenching, the specimens are tempered at 250  for 1hr and then cooling in air. The mechanical properties of the specimens are determined by using universal tensile testing machine for tensile test, Vickers hardness apparatus for hardness testing, measuring the grain size of the phases and examine the microstructure of the specimens before and after heat-treatment. The results of this work showed that improving the mechanical properties of medium carbon 1030 steel, which is quenching by water gives the preferred results as the following: Quenching by water leads to increase σy, σu.t.s, K and hardness, but at the same time quenching by water leads to decrease E and n. Also the quenching by water and followed by tempering leads to improve the microstructure and decreasing (refining of the grain size of ferrite and pearlite phases of the steel used in this work.

Spectral analysis of colour-quenched and chemically quenched C 14 samples

International Nuclear Information System (INIS)

Grau Malonda, A.; Scott Guillearrd, P.E.

1987-01-01

Pairs of pulse height distribution curves, of C-14 samples, colour quenched and chemically quenched were obtained. The possibility to choose a counting window in order to obtain the counting efficiency curves, for both type of quenching was studied. (author). 7 figs., 7 refs

Spectral analysis of colour-quenched and chemically quenched C-14 samples

International Nuclear Information System (INIS)

Scott, P. E.; Grau, A.

1987-01-01

In this paper pairs of pulse height distribution curves, of C-14 samples, colour-quenched and chemically quenched was obtained. The possibility to choose a counting window in order to obtain the counting efficiency curves, for both type of quenching was studied. (Author) 7 refs

Finite element modelling of different CANDU fuel bundle types in various refuelling conditions

International Nuclear Information System (INIS)

Roman, M. R.; Ionescu, D. V.; Olteanu, G.; Florea, S.; Radut, A. C.

2016-01-01

The objective of this paper is to develop a finite element model for static strength analysis of the CANDU standard with 37 elements fuel bundle and the SEU43 with 43 elements fuel bundle design for various refuelling conditions. The computer code, ANSYS7.1, is used to simulate the axial compression in CANDU type fuel bundles subject to hydraulic drag loads, deflection of fuel elements, stresses and displacements in the end plates. Two possible situations for the fuelling machine side stops are considered in our analyses, as follows: the last fuel bundle is supported by the two side stops and a side stop can be blocked therefore, the last fuel bundle is supported by only one side stop. The results of the analyses performed are briefly presented and also illustrated in a graphical form. The finite element model developed in present study is verified against test results for endplate displacement and element bowing obtained from strength tests with fuel bundle string and fuelling machine side-stop simulators. Comparison of ANSYS model predictions with these experimental results led to a very good agreement. Despite the difference in hydraulic load between SEU43 and CANDU standard fuel bundles strings, the maximum stress in the SEU43 endplate is about the same with the maximum stress in the CANDU standard endplate. The comparative assessment reveals that SEU43 fuel bundle is able to withstand high flow rate without showing a significant geometric instability. (authors)

Real-time wavelet-based inline banknote-in-bundle counting for cut-and-bundle machines

Science.gov (United States)

Petker, Denis; Lohweg, Volker; Gillich, Eugen; Türke, Thomas; Willeke, Harald; Lochmüller, Jens; Schaede, Johannes

2011-03-01

Automatic banknote sheet cut-and-bundle machines are widely used within the scope of banknote production. Beside the cutting-and-bundling, which is a mature technology, image-processing-based quality inspection for this type of machine is attractive. We present in this work a new real-time Touchless Counting and perspective cutting blade quality insurance system, based on a Color-CCD-Camera and a dual-core Computer, for cut-and-bundle applications in banknote production. The system, which applies Wavelet-based multi-scale filtering is able to count banknotes inside a 100-bundle within 200-300 ms depending on the window size.

Evaluation of the Effect of Tube Pitch and Surface Alterations on Temperature Field at Sprinkled Tube Bundle

Directory of Open Access Journals (Sweden)

Kracík Petr

2015-01-01

Full Text Available Water flowing on a sprinkled tube bundle forms three basic modes: It is the Droplet mode (liquid drips from one tube to another, the Jet mode (with an increasing flow rate droplets merge into a column and the Membrane (Sheet mode (with further increasing of falling film liquid flow rate columns merge and create sheets between the tubes. With sufficient flow rate sheets merge at this state and the tube bundle is completely covered by a thin liquid film. There are several factors influencing the individual mode types as well as heat transfer. Beside the above mentioned falling film liquid flow rate they are for instance tube diameters, tube pitches in a tube bundle or a physical condition of a falling film liquid. This paper presents a summary of data measured at atmospheric pressure at a tube bundle consisting of copper tubes of 12 milimeters diameter and of the studied tube length one meter. The tubes are positioned horizontally one above another with the tested pitches of 15, 20, 25 and 30 mm and there is a distribution tube placed above them with water flowing out. The thermal gradient of 15–40 has been tested with all pitches where the falling film liquid’s temperature at the inlet of the distribution tube was 15 °C. The liquid was heated during the flow through the exchanger and the temperature of the sprinkled (heater liquid at the inlet of the exchanger with a constant flow rate about 7.2 litres per minute was 40 °C. The tested flow of the falling film liquid ranged from 1.0 to 13.0 litres per minute. Sequences of 180 exposures have been recorded in partial flow rate stages by thermographic camera with record frequency of 30 Hz which were consequently assessed using the Matlab programme. This paper presents results achieved at the above mentioned pitches and at three types of tube bundle surfaces.

Fluorescence Endoscopy in vivo based on Fiber-bundle Measurements

Energy Technology Data Exchange (ETDEWEB)

Zufiria, B.; Gomez-Garcia, P.; Stamatakis, K.; Vaquero, J.J.; Fresno, M.; Desco, M.; Ripoll, J.; Arranz, A.

2016-07-01

High-resolution imaging techniques have become important for the determination of the cellular organization that is coupled to organ function. In many cases the organ can be viewed without the need of ionizing radiation techniques in an easier way. This is the case of the gastrointestinal tract, an organ that can be directly accessed with endoscopy avoiding any invasive procedure. Here we describe the design, assembly and testing of a fluorescence high-resolution endoscope intended for the study of the cellular organization of the colon in an experimental mouse model of colon carcinoma. Access to the colon of the mouse took place using a fiber-optic bundle that redirects the light coming from a LED to produce fluorescence and detect it back through the fiber bundle. Results from in vivo and ex-vivo test using our fluorescence fiber bundle endoscope show altered tissue structure and destruction of the intestinal crypts in tumor-bearing areas compared with healthy tissue. (Author)

Thermo hydraulic and quench propagation characteristics of SST-1 TF coil

Energy Technology Data Exchange (ETDEWEB)

Sharma, A.N., E-mail: ansharma@ipr.res.in [Institute for Plasma Research, Gandhinagar (India); Pradhan, S. [Institute for Plasma Research, Gandhinagar (India); Duchateau, J.L. [CEA Cadarache, 13108 St Paul lez Durance Cedex (France); Khristi, Y.; Prasad, U.; Doshi, K.; Varmora, P.; Patel, D.; Tanna, V.L. [Institute for Plasma Research, Gandhinagar (India)

2014-02-15

Highlights: • Details of SST-1 TF coils, CICC. • Details of SST-1 TF coil cold test. • Quench analysis of TF magnet. • Flow changes following quench. • Predictive analysis of assembled magnet system. - Abstract: SST-1 toroidal field (TF) magnet system is comprising of sixteen superconducting modified ‘D’ shaped TF coils. During single coil test campaigns spanning from June 10, 2010 till January 24, 2011; the electromagnetic, thermal hydraulic and mechanical performances of each TF magnet have been qualified at its respective nominal operating current of 10,000 A in either two-phase or supercritical helium cooling conditions. During the current charging experiments, few quenches have initiated either as a consequence of irrecoverable normal zones or being induced in some of the TF magnets. Quench evolution in the TF coils have been analyzed in detail in order to understand the thermal hydraulic and quench propagation characteristics of the SST-1 TF magnets. The same were also simulated using 1D code Gandalf. This paper elaborates the details of the analyses and the quench simulation results. A predictive quench propagation analysis of 16 assembled TF magnets system has also been reported in this paper.

Co-Higgs bundles on P^1

OpenAIRE

Rayan, Steven

2010-01-01

Co-Higgs bundles are Higgs bundles in the sense of Simpson, but with Higgs fields that take values in the tangent bundle instead of the cotangent bundle. Given a vector bundle on P^1, we find necessary and sufficient conditions on its Grothendieck splitting for it to admit a stable Higgs field. We characterize the rank-2, odd-degree moduli space as a universal elliptic curve with a globally-defined equation. For ranks r=2,3,4, we explicitly verify the conjectural Betti numbers emerging from t...

Thermal conductivity of multi-walled carbon nanotube sheets: radiation losses and quenching of phonon modes

Energy Technology Data Exchange (ETDEWEB)

Aliev, Ali E; Lima, Marcio H; Baughman, Ray H [Alan G MacDiarmid NanoTech Institute, University of Texas at Dallas, Richardson, TX 75083 (United States); Silverman, Edward M, E-mail: Ali.Aliev@utdallas.edu [Northrop Grumman Space Technology, Redondo Beach, CA 90278 (United States)

2010-01-22

The extremely high thermal conductivity of individual carbon nanotubes, predicted theoretically and observed experimentally, has not yet been achieved for large nanotube assemblies. Resistances at tube-tube interconnections and tube-electrode interfaces have been considered the main obstacles for effective electronic and heat transport. Here we show that, even for infinitely long and perfect nanotubes with well-designed tube-electrode interfaces, excessive radial heat radiation from nanotube surfaces and quenching of phonon modes in large bundles are additional processes that substantially reduce thermal transport along nanotubes. Equivalent circuit simulations and an experimental self-heating 3{omega} technique were used to determine the peculiarities of anisotropic heat flow and thermal conductivity of single MWNTs, bundled MWNTs and aligned, free-standing MWNT sheets. The thermal conductivity of individual MWNTs grown by chemical vapor deposition and normalized to the density of graphite is much lower ({kappa}{sub MWNT} = 600 {+-} 100 W m{sup -1} K{sup -1}) than theoretically predicted. Coupling within MWNT bundles decreases this thermal conductivity to 150 W m{sup -1} K{sup -1}. Further decrease of the effective thermal conductivity in MWNT sheets to 50 W m{sup -1} K{sup -1} comes from tube-tube interconnections and sheet imperfections like dangling fiber ends, loops and misalignment of nanotubes. Optimal structures for enhancing thermal conductivity are discussed.

Impact of bundle deformation on CHF: ASSERT-PV assessment of extended burnup Bruce B bundle G85159W

International Nuclear Information System (INIS)

Rao, Y.F.; Manzer, A.M.

2005-01-01

This paper presents a subchannel thermalhydraulic analysis of the effect on critical heat flux (CHF) of bundle deformation such as element bow and diametral creep. The bundle geometry is based on the post-irradiation examination (PIE) data of a single bundle from the Bruce B Nuclear Generating Station, Bruce B bundle G85159W, which was irradiated for more than two years in the core during reactor commissioning. The subchannel code ASSERT-PV IST is used to assess changes in CHF and dryout power due to bundle deformation, compared to the reference, undeformed bundle. (author)

Fluid structure interaction in tube bundles

International Nuclear Information System (INIS)

Brochard, D.; Jedrzejewski, F.; Gibert, R.J.

1995-01-01

A lot of industrial components contain tube bundles immersed in a fluid. The mechanical analysis of such systems requires the study of the fluid structure interaction in the tube bundle. Simplified methods, based on homogenization methods, have been developed to analyse such phenomenon and have been validated through experimental results. Generally, these methods consider only the fluid motion in a plan normal to the bundle axis. This paper will analyse, in a first part, the fluid structure interaction in a tube bundle through a 2D finite element model representing the bundle cross section. The influence of various parameters like the bundle size, and the bundle confinement will be studied. These results will be then compared with results from homogenization methods. Finally, the influence of the 3D fluid motion will be investigated, in using simplified methods. (authors). 11 refs., 12 figs., 2 tabs

Quench Protection of SC Quadrupole Magnets

Science.gov (United States)

Feher, S.; Bossert, R.; Dimarco, J.; Mitchell, D.; Lamm, M. J.; Limon, P. J.; Mazur, P.; Nobrega, F.; Orris, D.; Ozelis, J. P.; Strait, J. B.; Tompkins, J. C.; Zlobin, A. V.; McInturff, A. D.

1997-05-01

The energy stored in a superconducting accelerator magnet is dissipated after a quench in the coil normal zones, heating the coil and generating a turn to turn and coil to ground voltage drop. Quench heaters are used to protect the superconducting magnet by greatly increasing the coil normal zone thus allowing the energy to be dissipated over a larger conductor volume. Such heaters will be required for the Fermilab/LBNL design of the high gradient quads (HGQ) designed for the LHC interaction regions. As a first step, heaters were installed and tested in several Tevatron low-β superconducting quadrupoles. Experimental studies in normal and superfluid helium are presented which show the heater-induced quench response as a function of magnet excitation current, magnet temperature and peak heater energy density.

New models of droplet deposition and entrainment for prediction of CHF in cylindrical rod bundles

Energy Technology Data Exchange (ETDEWEB)

Zhang, Haibin, E-mail: hb-zhang@xjtu.edu.cn [School of Chemical Engineering and Technology, Xi’an Jiaotong University, Xi’an 710049 (China); Department of Chemical Engineering, Imperial College, London SW7 2BY (United Kingdom); Hewitt, G.F. [Department of Chemical Engineering, Imperial College, London SW7 2BY (United Kingdom)

2016-08-15

Highlights: • New models of droplet deposition and entrainment in rod bundles is developed. • A new phenomenological model to predict the CHF in rod bundles is described. • The present model is well able to predict CHF in rod bundles. - Abstract: In this paper, we present a new set of model of droplet deposition and entrainment in cylindrical rod bundles based on the previously proposed model for annuli (effectively a “one-rod” bundle) (2016a). These models make it possible to evaluate the differences of the rates of droplet deposition and entrainment for the respective rods and for the outer tube by taking into account the geometrical characteristics of the rod bundles. Using these models, a phenomenological model to predict the CHF (critical heat flux) for upward annular flow in vertical rod bundles is described. The performance of the model is tested against the experimental data of Becker et al. (1964) for CHF in 3-rod and 7-rod bundles. These data include tests in which only the rods were heated and data for simultaneous uniform and non-uniform heating of the rods and the outer tube. It was shown that the predicted CHFs by the present model agree well with the experimental data and with the experimental observation that dryout occurred first on the outer rods in 7-rod bundles. It is expected that the methodology used will be generally applicable in the prediction of CHF in rod bundles.

Effect of bundle junction face and misalignment on the pressure drops across a randomly loaded and aligned 12 bundles in CANDU fuel channel

Energy Technology Data Exchange (ETDEWEB)

Suk, H. C.; Sim, K. S.; Chang, C. H.; Lee, Y. O. [Korea Atomic Energy Reaearch Institute, Taejon (Korea, Republic of)

1996-06-01

The pressure drop of twelve fuel bundle string in the CANDU-6 fuel channel is equal to the sum of the eleven junction pressure losses, the bundle string entrance and exit pressure losses, the skin friction pressure loss, and other appendage pressure losses, where the junction loss is dependent on the bundle and faces and angular alignments of the junctions. The results of the single junction pressure drop tests in a short rig show that the most probable pressure drop of the eleven junction was analytically equal to the eleven times of average pressure drop of all the possible single junction pressure drops, and also that the largest and smallest junction pressure drops across the eleven junctions probably occurred only with BA and BB type junctions, respectively, where A and B denote the bundle end sides with an end-plates on which a company monogram is stamped and unstamped, respectively. 5 refs., 7 figs., 1 tab. (author).

Steady-state thermohydraulic studies in seven-pin bundle out-of-pile experiments: nominal and distorted geometry tests

International Nuclear Information System (INIS)

Falzetti, L.; Meneghello, S.; Pezzilli, M.

1979-01-01

Two sets of experiments have been performed in sodium with two seven pin electrically heated bundles: the first with a nominal arrangement, the second with one dummy pin enlarged 20% in diameter in peripheral position. In this paper a rapid review of experimental results and theoretical works, related to the temperature distribution in these geometries, is presented together with a short description of the developed test section technology

Posttest examination of the VVER-1000 fuel rod bundle CORA-W2

International Nuclear Information System (INIS)

Sepold, L.

1995-06-01

The bundle meltdown experiment CORA-W2, representing the behavior of a Russian type VVER-1000 fuel element, with one B 4 C/stainless steel absorber rod was selected by the OECD/CSNI as International Standard Problem (ISP-36). The experimental results of CORA-W2 serve as data base for comparison with analytical predictions of the high-temperature material behavior by various code systems. The first part of the experimental results is described in KfK 5363 (1994), the second part is documented in this report which contains the destructive post-test examination results. The metallographical and analytical (SEM/EDX) post-test examinations were performed in Germany and Russia and are summarized in five individual contributions. The upper half of the bundle is completely oxidized, the lower half has kept the fuel rods relatively intact. The post-test examination results show the strong impact of the B 4 C absorber rod and the stainless steel grid spacers on the ''low-temperature'' bundle damage initiation and progression. The B 4 C absorber rod completely disappeared in the upper half of the bundle. The multicomponent melts relocated and formed coolant channel blockages on solidification with a maximum extent of about 30% in the lower part of the bundle. At temperatures above the melting point of the ZrNb1 cladding extensive fuel dissolution occurred. (orig.) [de

Fuel bundle to pressure tube fretting in Bruce and Darlington

Energy Technology Data Exchange (ETDEWEB)

Norsworthy, A G; Ditschun, A [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)

1996-12-31

As the fuel channel elongates due to creep, the fuel string moves relative to the inlet until the fuel pads at the inboard end eventually separate from the spacer sleeve, and the fuel resides on the burnish mark of the pressure tube. The bundle is then supported in a fashion which contributes to increased levels of vibration. Those pads which (due to geometric variation) have contact loads with the pressure tube within a certain range, vibrate, and cause significant fretting on the burnish mark, and further along at the midplane of the bundle. Inspection of the pressure tubes in Bruce A, Bruce B, and Darlington has revealed fret damage up to 0.55 mm at the burnish mark and slightly lower than this at the inlet bundle midplane. To date, all fret marks have been dealt with successfully without the need for tube replacement, but a program of work has been initiated to understand the mechanism and reduce the fretting. Such understanding is necessary to guide future design changes to the fuel bundle, to guide future inspection programs, to guide maintenance programs, and for longer term strategic planning. This paper discusses how the understanding of fretting has evolved and outlines a current hypothesis for the mechanism of fretting. The role of bundle geometry, excitation forces, and reactor conditions are reviewed, along with options under consideration to mitigate damage. (author). 4 refs., 2 tabs., 13 figs.

Fuel bundle to pressure tube fretting in Bruce and Darlington

International Nuclear Information System (INIS)

Norsworthy, A.G.; Ditschun, A.

1995-01-01

As the fuel channel elongates due to creep, the fuel string moves relative to the inlet until the fuel pads at the inboard end eventually separate from the spacer sleeve, and the fuel resides on the burnish mark of the pressure tube. The bundle is then supported in a fashion which contributes to increased levels of vibration. Those pads which (due to geometric variation) have contact loads with the pressure tube within a certain range, vibrate, and cause significant fretting on the burnish mark, and further along at the midplane of the bundle. Inspection of the pressure tubes in Bruce A, Bruce B, and Darlington has revealed fret damage up to 0.55 mm at the burnish mark and slightly lower than this at the inlet bundle midplane. To date, all fret marks have been dealt with successfully without the need for tube replacement, but a program of work has been initiated to understand the mechanism and reduce the fretting. Such understanding is necessary to guide future design changes to the fuel bundle, to guide future inspection programs, to guide maintenance programs, and for longer term strategic planning. This paper discusses how the understanding of fretting has evolved and outlines a current hypothesis for the mechanism of fretting. The role of bundle geometry, excitation forces, and reactor conditions are reviewed, along with options under consideration to mitigate damage. (author). 4 refs., 2 tabs., 13 figs

Entropy for frame bundle systems and Grassmann bundle systems induced by a diffeomorphism

Institute of Scientific and Technical Information of China (English)

SUN; Weniang(孙文祥)

2002-01-01

ALiao hyperbolic diffeomorphism has equal measure entropy and topological entropy to that ofits induced systems on frame bundles and Grassmann bundles. This solves a problem Liao posed in 1996 forLiao hyperbolic diffeomorphisms.

Lessons learnt from FARO/TERMOS corium melt quenching experiments

Energy Technology Data Exchange (ETDEWEB)

Magallon, D.; Huhtiniemi, I.; Hohmann, H. [Commission of the European Communities, Ispra (Italy). Joint Research Center

1998-01-01

The influence of melt quantity, melt composition, water depth and initial pressure on quenching is assessed on the basis of seven tests performed in various conditions in the TERMOS vessel of the FARO facility at JRC-Ispra. Tests involved UO{sub 2}-based melt quantities in the range 18-176 kg at a temperature of approximately 3000 K poured into saturated water. The results suggest that erosion of the melt jet column is an efficient contributor to the amount of break-up, and thus quenching, for large pours of corium melt. The presence of Zr metal in the melt induced a much more efficient quenching than in a similar test with no Zr metal, attributed to the oxidation of the Zr. Significant amounts of H{sub 2} were produced also in tests with pure oxidic melts (e.g. about 300 g for 157 kg melt). In the tests at 5.0 and 2.0 MPa good mixing with significant melt break-up and quenching was obtained during the penetration in the water. At 0.5 MPa, good penetration of the melt into the water could still be achieved, but a jump in the vessel pressurisation occurred when the melt contacted the bottom and part (5 kg) of the debris was re-ejected from the water. (author)

Tube bundle vibrations in transversal flow

International Nuclear Information System (INIS)

Gibert, R.J.; Sagner, M.

1978-01-01

This study gives important information concerning characteristic parameters about lock-in and whirling instability phenomena, in the case of tube arrays. The work is mainly an experimental one though models are also developed: 1) an equilateral pitch bundle (p=1,5 D with D=tube diameter) is tested. Tube damping (epsilon) and first eigenfrequency (f), flow velocity are explored in a large domain. Vibratory level of the tubes are measured and critical points are ploted on the fluidelastic parameters diagram. Several bundles with various usual pitches and arrangements (in line or staggered) are tested. Critical velocities are measured and the whirling instability characteristic coefficient is tabulated. A complementary experiment is made on tube rows with various pitches. This gives valuable informations concerning the look-in domain in VR and A'R diagram. Furthermore this puts in evidence the important effect of a frequency difference between two adjacent tubes on the whirling critical velocity

Development of CANFLEX fuel bundle

International Nuclear Information System (INIS)

Suk, Ho Chun; Hwang, Woan; Jeong, Young Hwan

1991-12-01

This research project is underway in cooperation with AECL to develop the CANDU advanced fuel bundle(so-called CANFLEX) which can enhance reactor safety and fuel economy in comparison with the current CANDU fuel and which can be used with natural uranium, slightly enriched uranium and other advanced fuel cycle. As the final schedule, the advanced fuel will be verified by carrying out a large scale demonstration of the bundle irradiation in a commercial CANDU reactors for 1996 and 1997, and consequently will be used in the existing and future reactors in Korea. The research activities during this year include the basic design of CANFLEX fuel with slightly enriched uranium(CANFLEX-SEU), with emphasis on the extension of fuel operation limit. Based on this basic design, CANFLEX fuel was mocked up. Out-of-pile hydraulic scoping tests were conducted with the fuel. (Author)

Performance of the MAGCOOL-subcooler cryogenic system after SSC quadrupole quenches

International Nuclear Information System (INIS)

Wu, K.C.

1993-01-01

The subcooler assembly installed in the MAGCOOL magnet test area at Brookhaven National Laboratory has been used for testing SSC dipoles, quadrupoles and a spool piece since 1989. A detailed description of the system, its steady state capacity and the performance after quenches of a 50 mm SSC dipole were given. Subsequent studies on low current quenches of the SSC dipoles and quenches of the RHIC dipoles were also carried out. In this paper, the performance of the subcooler after quenches of the SSC quadrupole QCC404 is presented. Pressures, temperatures and flow rates in the magnet cooling loop after magnet quenches are given as a function of time. The cooling rates and total energy removed by cooling during quench recovery have been calculated for quench currents between 2000 and 7952 amperes. Because the inductance of the quadrupole is about one tenth that of a SSC dipole, the stored energy released is small and the impact on the system is mild. The cooling loop pressure never exceeds 12 atmospheres and the cryogenic system recovers in less than 15 minutes. As in all past studies, the peak pressure and temperature in the magnet cooling loop are linearly proportional to the energy released during a quench and excellent agreement between the total cooling provided and the magnetic stored energy is found

Signal detection by active, noisy hair bundles

Science.gov (United States)

O'Maoiléidigh, Dáibhid; Salvi, Joshua D.; Hudspeth, A. J.

2018-05-01

Vertebrate ears employ hair bundles to transduce mechanical movements into electrical signals, but their performance is limited by noise. Hair bundles are substantially more sensitive to periodic stimulation when they are mechanically active, however, than when they are passive. We developed a model of active hair-bundle mechanics that predicts the conditions under which a bundle is most sensitive to periodic stimulation. The model relies only on the existence of mechanotransduction channels and an active adaptation mechanism that recloses the channels. For a frequency-detuned stimulus, a noisy hair bundle's phase-locked response and degree of entrainment as well as its detection bandwidth are maximized when the bundle exhibits low-amplitude spontaneous oscillations. The phase-locked response and entrainment of a bundle are predicted to peak as functions of the noise level. We confirmed several of these predictions experimentally by periodically forcing hair bundles held near the onset of self-oscillation. A hair bundle's active process amplifies the stimulus preferentially over the noise, allowing the bundle to detect periodic forces less than 1 pN in amplitude. Moreover, the addition of noise can improve a bundle's ability to detect the stimulus. Although, mechanical activity has not yet been observed in mammalian hair bundles, a related model predicts that active but quiescent bundles can oscillate spontaneously when they are loaded by a sufficiently massive object such as the tectorial membrane. Overall, this work indicates that auditory systems rely on active elements, composed of hair cells and their mechanical environment, that operate on the brink of self-oscillation.

Quench Modeling in High-field Nb3Sn Accelerator Magnets

Science.gov (United States)

Bermudez, S. Izquierdo; Bajas, H.; Bottura, L.

The development of high-field magnets is on-going in the framework of the LHC luminosity upgrade. The resulting peak field, in the range of 12 T to 13 T, requires the use Nb3Sn as superconductor. Due to the high stored energy density (compact winding for cost reduction) and the low stabilizer fraction (to achieve the desired margins), quench protection becomes a challenging problem. Accurate simulation of quench transientsin these magnets is hence crucial to the design choices, the definition of priority R&D and to prove that the magnets are fit for operation. In this paper we focus on the modelling of quench initiation and propagation, we describe approaches that are suitable for magnet simulation, and we compare numerical results with available experimental data.

Observations of environmental quenching in groups in the 11 Gyr since z = 2.5: Different quenching for central and satellite galaxies

International Nuclear Information System (INIS)

Tal, Tomer; Illingworth, Garth D.; Magee, Daniel; Dekel, Avishai; Oesch, Pascal; Van Dokkum, Pieter G.; Leja, Joel; Momcheva, Ivelina; Nelson, Erica J.; Muzzin, Adam; Franx, Marijn; Brammer, Gabriel B.; Marchesini, Danilo; Patel, Shannon G.; Quadri, Ryan F.; Rix, Hans-Walter; Skelton, Rosalind E.; Wake, David A.; Whitaker, Katherine E.

2014-01-01

We present direct observational evidence for star formation quenching in galaxy groups in the redshift range 0 < z < 2.5. We utilize a large sample of nearly 6000 groups, selected by fixed cumulative number density from three photometric catalogs, to follow the evolving quiescent fractions of central and satellite galaxies over roughly 11 Gyr. At z ∼ 0, central galaxies in our sample range in stellar mass from Milky Way/M31 analogs (M * /M ☉ = 6.5 × 10 10 ) to nearby massive ellipticals (M * /M ☉ = 1.5 × 10 11 ). Satellite galaxies in the same groups reach masses as low as twice that of the Large Magellanic Cloud (M * /M ☉ = 6.5 × 10 9 ). Using statistical background subtraction, we measure the average rest-frame colors of galaxies in our groups and calculate the evolving quiescent fractions of centrals and satellites over seven redshift bins. Our analysis shows clear evidence for star formation quenching in group halos, with a different quenching onset for centrals and their satellite galaxies. Using halo mass estimates for our central galaxies, we find that star formation shuts off in centrals when typical halo masses reach between 10 12 and 10 13 M ☉ , consistent with predictions from the halo quenching model. In contrast, satellite galaxies in the same groups most likely undergo quenching by environmental processes, whose onset is delayed with respect to their central galaxy. Although star formation is suppressed in all galaxies over time, the processes that govern quenching are different for centrals and satellites. While mass plays an important role in determining the star formation activity of central galaxies, quenching in satellite galaxies is dominated by the environment in which they reside.

Observations of environmental quenching in groups in the 11 Gyr since z = 2.5: Different quenching for central and satellite galaxies

Energy Technology Data Exchange (ETDEWEB)

Tal, Tomer; Illingworth, Garth D.; Magee, Daniel [UCO/Lick Observatory, University of California, Santa Cruz, CA 95064 (United States); Dekel, Avishai [Racah Institute of Physics, The Hebrew University, Jerusalem 91904 (Israel); Oesch, Pascal; Van Dokkum, Pieter G.; Leja, Joel; Momcheva, Ivelina; Nelson, Erica J. [Yale University Astronomy Department, P.O. Box 208101, New Haven, CT 06520-8101 (United States); Muzzin, Adam; Franx, Marijn [Leiden Observatory, Leiden University, NL-2300 RA Leiden (Netherlands); Brammer, Gabriel B. [Space Telescope Science Institute, 3700 San Martin Drive, Baltimore, MD 21218 (United States); Marchesini, Danilo [Department of Physics and Astronomy, Tufts University, Medford, MA 02155 (United States); Patel, Shannon G.; Quadri, Ryan F. [Carnegie Observatories, Pasadena, CA 91101 (United States); Rix, Hans-Walter [Max-Planck-Institut für Astronomie, Königstuhl 17, D-69117 Heidelberg (Germany); Skelton, Rosalind E. [South African Astronomical Observatory, Observatory Road, Cape Town (South Africa); Wake, David A. [Department of Astronomy, University of Wisconsin-Madison, Madison, WI 53706 (United States); Whitaker, Katherine E., E-mail: tal@ucolick.org [Astrophysics Science Division, Goddard Space Flight Center, Greenbelt, MD 20771 (United States)

2014-07-10

We present direct observational evidence for star formation quenching in galaxy groups in the redshift range 0 < z < 2.5. We utilize a large sample of nearly 6000 groups, selected by fixed cumulative number density from three photometric catalogs, to follow the evolving quiescent fractions of central and satellite galaxies over roughly 11 Gyr. At z ∼ 0, central galaxies in our sample range in stellar mass from Milky Way/M31 analogs (M{sub *}/M{sub ☉} = 6.5 × 10{sup 10}) to nearby massive ellipticals (M{sub *}/M{sub ☉} = 1.5 × 10{sup 11}). Satellite galaxies in the same groups reach masses as low as twice that of the Large Magellanic Cloud (M{sub *}/M{sub ☉} = 6.5 × 10{sup 9}). Using statistical background subtraction, we measure the average rest-frame colors of galaxies in our groups and calculate the evolving quiescent fractions of centrals and satellites over seven redshift bins. Our analysis shows clear evidence for star formation quenching in group halos, with a different quenching onset for centrals and their satellite galaxies. Using halo mass estimates for our central galaxies, we find that star formation shuts off in centrals when typical halo masses reach between 10{sup 12} and 10{sup 13} M{sub ☉}, consistent with predictions from the halo quenching model. In contrast, satellite galaxies in the same groups most likely undergo quenching by environmental processes, whose onset is delayed with respect to their central galaxy. Although star formation is suppressed in all galaxies over time, the processes that govern quenching are different for centrals and satellites. While mass plays an important role in determining the star formation activity of central galaxies, quenching in satellite galaxies is dominated by the environment in which they reside.

Assessment of boiling transition analysis code against data from NUPEC BWR full-size fine-mesh bundle tests

International Nuclear Information System (INIS)

Utsuno, Hideaki; Ishida, Naoyuki; Masuhara, Yasuhiro; Kasahara, Fumio

2004-01-01

Transient BT analysis code TCAPE based on mechanistic methods coupled with subchannel analysis has been developed for the evaluation on fuel integrity under abnormal operations in BWR. TCAPE consisted mainly of the drift-flux model, the cross-flow model, the film model and the heat transfer model. Assessment of TCAPE has been performed against data from BWR full-size fine-mesh bundle tests (BFBT), which consisted of two major parts: the void distribution measurement and the critical power measurement. Code and data comparison was made for void distributions with varying number of unheated rods in simulated actual fuel assembly. Prediction of steady-state critical power was compared with the measurement on full-scale bundle under a range of BWR operational conditions. Although the cross-sectional averaged void fraction was underestimated when it became lower, the accuracy was obtained that the averaged ratio 0.910 and its standard deviation 0.076. The prediction of steady-state critical power agreed well with the data in the range of BWR operations, where the prediction accuracy was obtained that the averaged ratio 0.997 and its standard deviation 0.043. These results demonstrated that TCAPE is well capable to predict two-phase flow distribution and liquid film dryout phenomena occurring in BWR rod bundles. Part of NUPEC BFBT database will be made available for an international benchmark exercise. The code assessment shall be continued against the OECD/NRC benchmark based on BFBT database. (author)

Molybdenum-99-producing 37-element fuel bundle neutronically and thermal-hydraulically equivalent to a standard CANDU fuel bundle

Energy Technology Data Exchange (ETDEWEB)

Nichita, E., E-mail: Eleodor.Nichita@uoit.ca; Haroon, J., E-mail: Jawad.Haroon@uoit.ca

2016-10-15

Highlights: • A 37-element fuel bundle modified for {sup 99}Mo production in CANDU reactors is presented. • The modified bundle is neutronically and thermal-hydraulically equivalent to the standard bundle. • The modified bundle satisfies all safety criteria satisfied by the standard bundle. - Abstract: {sup 99m}Tc, the most commonly used radioisotope in diagnostic nuclear medicine, results from the radioactive decay of {sup 99}Mo which is currently being produced at various research reactors around the globe. In this study, the potential use of CANDU power reactors for the production of {sup 99}Mo is investigated. A modified 37-element fuel bundle, suitable for the production of {sup 99}Mo in existing CANDU-type reactors is proposed. The new bundle is specifically designed to be neutronically and thermal-hydraulically equivalent to the standard 37-element CANDU fuel bundle in normal, steady-state operation and, at the same time, be able to produce significant quantities of {sup 99}Mo when irradiated in a CANDU reactor. The proposed bundle design uses fuel pins consisting of a depleted-uranium centre surrounded by a thin layer of low-enriched uranium. The new molybdenum-producing bundle is analyzed using the lattice transport code DRAGON and the diffusion code DONJON. The proposed design is shown to produce 4081 six-day Curies of {sup 99}Mo activity per bundle when irradiated in the peak-power channel of a CANDU core, while maintaining the necessary reactivity and power rating limits. The calculated {sup 99}Mo yield corresponds to approximately one third of the world weekly demand. A production rate of ∼3 bundles per week can meet the global demand of {sup 99}Mo.

Studies of quench propagation in a superconducting window frame magnet

International Nuclear Information System (INIS)

Allinger, J.; Carroll, A.; Danby, G.; DeVito, B.; Jackson, J.; Leonhardt, M.; Prodell, A.; Stoehr, R.

1981-01-01

During the testing of a meter long, superconducting window frame magnet, information from many spontaneously generated quenches have been recorded by an on-line computer system. Nearly every layer in an eleven layer dipole had a voltage tap and for some layers this subdivided into two halves. This allowed us to study development of the quenches in some detail. Knowledge of the resistances throughout the magnet also allowed the temperature distributions in the superconducting windings to be determined. A qualitative picture of the quench was developed and quantitative values of quench propagation velocities were compared to heat transfer calculations

Quench Heater Experiments on the LHC Main Superconducting Magnets

OpenAIRE

Rodríguez-Mateos, F; Pugnat, P; Sanfilippo, S; Schmidt, R; Siemko, A; Sonnemann, F

2000-01-01

In case of a quench in one of the main dipoles and quadrupoles of CERN's Large Hadron Collider (LHC), the magnet has to be protected against excessive temperatures and high voltages. In order to uniformly distribute the stored magnetic energy in the coils, heater strips installed in the magnet are fired after quench detection. Tests of different quench heater configurations were performed on various 1 m long model and 15 m long prototype dipole magnets, as well as on a 3 m long prototype quad...

Study on quench effects in liquid scintillation counting during tritium measurements

International Nuclear Information System (INIS)

Ivana Jakonic; Jovana Nikolov; Natasa Todorovic; Miroslav Veskovic; Branislava Tenjovic

2014-01-01

Quench effects can cause a serious reduction in counting efficiency for a given sample/cocktail mixture in liquid scintillation counting (LSC) experiments. This paper presents a simple experiment performed in order to test the influence of quenching on the LSC efficiency of 3 H. The aim of this study was to investigate the behavior of several quench agents with different quench strengths (nitromethane, nitric acid, acetone, dimethyl-sulfoxide) added in different amounts to tritiated water in order to obtain standard sets for quench calibration curves. The OptiPhase HiSafe 2 and OptiPhase HiSafe 3 scintillation cocktails were used in this study in order to compare their quench resistance. Measurements were performed using a low-level LS counter (Wallac, Quantulus 1220). (author)

The button effect of CANFLEX bundle on the critical heat flux and critical channel power

Energy Technology Data Exchange (ETDEWEB)

Park, Joo Hwan; Jun, Jisu; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Dimmick, G R; Bullock, D E; Inch, W [Atomic Energy of Canada Limited, Ontario (Canada)

1998-12-31

A CANFLEX (CANdu FLEXible fuelling) 43-element bundle has developed for a CANDU-6 reactor as an alternative of 37-element fuel bundle. The design has two diameter elements (11.5 and 13.5 mm) to reduce maximum element power rating and buttons to enhance the critical heat flux (CHF), compared with the standard 37-element bundle. The freon CHF experiments have performed for two series of CANFLEX bundles with and without buttons with a modelling fluid as refrigerant R-134a and axial uniform heat flux condition. Evaluating the effects of buttons of CANFLEX bundle on CHF and Critical Channel Power (CCP) with the experimental results, it is shown that the buttons enhance CCP as well as CHF. All the CHF`s for both the CANFLEX bundles are occurred at the end of fuel channel with the high dryout quality conditions. The CHF enhancement ratio are increased with increase of dryout quality for all flow conditions and also with increase of mass flux only for high pressure conditions. It indicates that the button is a useful design for CANDU operating condition because most CHF flow conditions for CANDU fuel bundle are ranged to high dryout quality conditions. 5 refs., 11 figs. (Author)

The button effect of CANFLEX bundle on the critical heat flux and critical channel power

Energy Technology Data Exchange (ETDEWEB)

Park, Joo Hwan; Jun, Jisu; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Dimmick, G. R.; Bullock, D. E.; Inch, W. [Atomic Energy of Canada Limited, Ontario (Canada)

1997-12-31

A CANFLEX (CANdu FLEXible fuelling) 43-element bundle has developed for a CANDU-6 reactor as an alternative of 37-element fuel bundle. The design has two diameter elements (11.5 and 13.5 mm) to reduce maximum element power rating and buttons to enhance the critical heat flux (CHF), compared with the standard 37-element bundle. The freon CHF experiments have performed for two series of CANFLEX bundles with and without buttons with a modelling fluid as refrigerant R-134a and axial uniform heat flux condition. Evaluating the effects of buttons of CANFLEX bundle on CHF and Critical Channel Power (CCP) with the experimental results, it is shown that the buttons enhance CCP as well as CHF. All the CHF`s for both the CANFLEX bundles are occurred at the end of fuel channel with the high dryout quality conditions. The CHF enhancement ratio are increased with increase of dryout quality for all flow conditions and also with increase of mass flux only for high pressure conditions. It indicates that the button is a useful design for CANDU operating condition because most CHF flow conditions for CANDU fuel bundle are ranged to high dryout quality conditions. 5 refs., 11 figs. (Author)

Experimental heat transfer in tube bundle

International Nuclear Information System (INIS)

Khattab, M.; Mariy, A.; Habib, M.

1983-01-01

Previous work has looked for the problem of heat transfer with flow parallel to rod bundle either by treating each rod individually as a separate channel or by treating the bundle as one unit. The present work will consider the existence of both the central and corner rods simultaneously inside the cluster itself under the same working conditions. The test section is geometrically similar to the fuel assembly of the Egyptian Research Reactor-1. The hydro-thermal performance of bundle having 16 - stainless steel tubes arranged in square array of 1.5 pitch to diameter ratio is investigated. Surface temperature and pressure distributions are determined. Average heat transfer coefficient for both central and corner tubes are correlated. Also, pressure drop and friction factor correlations are predicted. The maximum experimental range of the measured parameters are determined in the nonboiling region at 1400 Reynolds number and 3.64 W/cm 2 . It is found that the average heat transfer coefficient of the central tube is higher than that of the corner tube by 27%. Comparison with the previous work shows satisfactory agreement particularly with the circular tubes correlation - Dittus et al. - at 104 Reynolds number

ORNL rod-bundle heat-transfer test data. Volume 3. Thermal-hydraulic test facility experimental data report for test 3.06.6B - transient film boiling in upflow

International Nuclear Information System (INIS)

Mullins, C.B.; Felde, D.K.; Sutton, A.G.; Gould, S.S.; Morris, D.G.; Robinson, J.J.

1982-05-01

Reduced instrument responses are presented for Thermal-Hyraulic Test Facility (THTF) Test 3.06.6B. This test was conducted by members of the Oak Ridge National Laboratory Pressurized-Water-Reactor (PWR) Blowdown Heat Transfer (BDHT) Separate-Effects Program on August 29, 1980. The objective of the program was to investigate heat transfer phenomena believed to occur in PWR's during accidents, including small and large break loss-of-coolant accidents. Test 3.06.6B was conducted to obtain transient film boiling data in rod bundle geometry under reactor accident-type conditions. The primary purpose of this report is to make the reduced instrument responses for THTF Test 3.06.6B available. Included in the report are uncertainties in the instrument responses, calculated mass flows, and calculated rod powers

Vibration of fuel bundles

International Nuclear Information System (INIS)

Chen, S.S.

1975-06-01

Several mathematical models have been proposed for calculating fuel rod responses in axial flows based on a single rod consideration. The spacing between fuel rods in liquid metal fast breeder reactors is small; hence fuel rods will interact with one another due to fluid coupling. The objective of this paper is to study the coupled vibration of fuel bundles. To account for the fluid coupling, a computer code, AMASS, is developed to calculate added mass coefficients for a group of circular cylinders based on the potential flow theory. The equations of motion for rod bundles are then derived including hydrodynamic forces, drag forces, fluid pressure, gravity effect, axial tension, and damping. Based on the equations, a method of analysis is presented to study the free and forced vibrations of rod bundles. Finally, the method is applied to a typical LMFBR fuel bundle consisting of seven rods

Polyfluorophore Labels on DNA: Dramatic Sequence Dependence of Quenching

Science.gov (United States)

Teo, Yin Nah; Wilson, James N.

2010-01-01

We describe studies carried out in the DNA context to test how a common fluorescence quencher, dabcyl, interacts with oligodeoxynu-cleoside fluorophores (ODFs)—a system of stacked, electronically interacting fluorophores built on a DNA scaffold. We tested twenty different tetrameric ODF sequences containing varied combinations and orderings of pyrene (Y), benzopyrene (B), perylene (E), dimethylaminostilbene (D), and spacer (S) monomers conjugated to the 3′ end of a DNA oligomer. Hybridization of this probe sequence to a dabcyl-labeled complementary strand resulted in strong quenching of fluorescence in 85% of the twenty ODF sequences. The high efficiency of quenching was also established by their large Stern–Volmer constants (KSV) of between 2.1 × 104 and 4.3 × 105M−1, measured with a free dabcyl quencher. Interestingly, quenching of ODFs displayed strong sequence dependence. This was particularly evident in anagrams of ODF sequences; for example, the sequence BYDS had a KSV that was approximately two orders of magnitude greater than that of BSDY, which has the same dye composition. Other anagrams, for example EDSY and ESYD, also displayed different responses upon quenching by dabcyl. Analysis of spectra showed that apparent excimer and exciplex emission bands were quenched with much greater efficiency compared to monomer emission bands by at least an order of magnitude. This suggests an important role played by delocalized excited states of the π stack of fluorophores in the amplified quenching of fluorescence. PMID:19780115

"To Whom It May Concern": A Study on the Use of Lexical Bundles in Email Writing Tasks in an English Proficiency Test

Science.gov (United States)

Li, Zhi; Volkov, Alex

2017-01-01

Lexical bundles are worthy of attention in both teaching and testing writing as they function as basic building blocks of discourse. This corpus-based study focuses on the rated writing responses to the email tasks in the Canadian English Language Proficiency Index Program® General test (CELPIP-General) and explores the extent to which lexical…

Experimental benchmark data for PWR rod bundle with spacer-grids

International Nuclear Information System (INIS)

Dominguez-Ontiveros, Elvis E.; Hassan, Yassin A.; Conner, Michael E.; Karoutas, Zeses

2012-01-01

In numerical simulations of fuel rod bundle flow fields, the unsteady Navier–Stokes equations have to be solved in order to determine the time (phase) dependent characteristics of the flow. In order to validate the simulations results, detailed comparison with experimental data must be done. Experiments investigating complex flows in rod bundles with spacer grids that have mixing devices (such as flow mixing vanes) have mostly been performed using single-point measurements. In order to obtain more details and insight on the discrepancies between experimental and numerical data as well as to obtain a global understanding of the causes of these discrepancies, comparisons of the distributions of complete phase-averaged velocity and turbulence fields for various locations near spacer-grids should be performed. The experimental technique Particle Image Velocimetry (PIV) is capable of providing such benchmark data. This paper describes an experimental database obtained using two-dimensional Time Resolved Particle Image Velocimetry (TR-PIV) measurements within a 5 × 5 PWR rod bundle with spacer-grids that have flow mixing vanes. One of the unique characteristic of this set-up is the use of the Matched Index of Refraction technique employed in this investigation to allow complete optical access to the rod bundle. This unique feature allows flow visualization and measurement within the bundle without rod obstruction. This approach also allows the use of high temporal and spatial non-intrusive dynamic measurement techniques namely TR-PIV to investigate the flow evolution below and immediately above the spacer. The experimental data presented in this paper includes explanation of the various cases tested such as test rig dimensions, measurement zones, the test equipment and the boundary conditions in order to provide appropriate data for comparison with Computational Fluid Dynamics (CFD) simulations. Turbulence parameters of the obtained data are presented in order to gain

Behavior of a nine-rod PWR bundle under power-cooling-mismatch conditions

International Nuclear Information System (INIS)

Gunnerson, F.S.; Sparks, D.T.

1979-01-01

An experiment to characterize the behavior of a nine-rod pressurized water reactor (PWR) fuel bundle operating during power-cooling-mismatch (PCM) conditions has been conducted in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory (INEL). The experiment, designated Test PCM-5, is part of a series of PCM experiments designed to evaluate light water reactor (LWR) fuel rod response under postulated accident conditions. Test PCM-5 was the first nine-rod bundle experiment in the PCM test series. The primary objectives and the results of the experiment are described

Principal noncommutative torus bundles

DEFF Research Database (Denmark)

Echterhoff, Siegfried; Nest, Ryszard; Oyono-Oyono, Herve

2008-01-01

of bivariant K-theory (denoted RKK-theory) due to Kasparov. Using earlier results of Echterhoff and Williams, we shall give a complete classification of principal non-commutative torus bundles up to equivariant Morita equivalence. We then study these bundles as topological fibrations (forgetting the group...

Implementation of Canflex bundle manufacture - from 'bench scale' to production

International Nuclear Information System (INIS)

Pant, A.

1999-01-01

Zircatec Precision Industries (ZPI) has been involved with the development of the 43 element Canflex bundle design since 1986. This development included several 'prototype' campaigns involving the manufacture of small quantities of test bundles using enriched fuel. Manufacturing and inspection methods for this fuel were developed at ZPI as the design progressed. The most recent campaign involved the production of 26 bundles of the final Canflex design for a demonstration irradiation in the Point Lepreau Generating Station. This presentation will explore issues pertaining to the introduction of a new product line from initial trial quantities to full production levels. The Canflex fuel experience and a brief review of development efforts will be used as an example. (author)

Assessing the Effect of Language Demand in Bundles of Math Word Problems

Science.gov (United States)

Banks, Kathleen; Jeddeeni, Ahmad; Walker, Cindy M.

2016-01-01

Differential bundle functioning (DBF) analyses were conducted to determine whether seventh and eighth grade second language learners (SLLs) had lower probabilities of answering bundles of math word problems correctly that had heavy language demands, when compared to non-SLLs of equal math proficiency. Math word problems on each of four test forms…

Early Results of Anatomic Double Bundle Anterior Cruciate Ligament Reconstruction

Directory of Open Access Journals (Sweden)

Demet Pepele

2014-03-01

Full Text Available Aim: The goal in anterior cruciate ligament reconstruction (ACLR is to restore the normal anatomic structure and function of the knee. In the significant proportion of patients after the traditional single-bundle ACLR, complaints of instability still continue. Anatomic double bundle ACLR may provide normal kinematics in knees, much closer to the natural anatomy. The aim of this study is to clinically assess the early outcomes of our anatomical double bundle ACLR. Material and Method: In our clinic between June 2009 and March 2010, performed the anatomic double bundle ACLR with autogenous hamstring grafts 20 patients were evaluated prospectively with Cincinnati, IKDC and Lysholm scores and in clinically for muscle strength and with Cybex II dynamometer. Results: The mean follow-up is 17.8 months (13-21 months. Patients%u2019 scores of Cincinnati, IKDC and Lysholm were respectively, preoperative 18.1, 39.3 and 39.8, while the post-op increased to 27.2, 76.3 and 86.3. In their last check, 17 percent of the patients according to IKDC scores (85% A (excellent and B (good group and 3 patients took place as C (adequate group. The power measurements of quadriceps and hamstring muscle groups of patients who underwent surgery showed no significant difference compared with the intact knees. Discussion: Double-bundle ACL reconstruction is a satisfactory method. There is a need comparative, long-term studies in large numbers in order to determine improving clinical outcome, preventing degeneration and restoring the knee biomechanics better.

Doubler system quench detection threshold

International Nuclear Information System (INIS)

Kuepke, K.; Kuchnir, M.; Martin, P.

1983-01-01

The experimental study leading to the determination of the sensitivity needed for protecting the Fermilab Doubler from damage during quenches is presented. The quench voltage thresholds involved were obtained from measurements made on Doubler cable of resistance x temperature and voltage x time during quenches under several currents and from data collected during operation of the Doubler Quench Protection System as implemented in the B-12 string of 20 magnets. At 4kA, a quench voltage threshold in excess of 5.OV will limit the peak Doubler cable temperature to 452K for quenches originating in the magnet coils whereas a threshold of 0.5V is required for quenches originating outside of coils

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

International Nuclear Information System (INIS)

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

1993-09-01

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

Cotangent bundle approach to noninertial frames

International Nuclear Information System (INIS)

DeFacio, B.; Retzloff, D.

1980-01-01

The most general possible noninertial acceleration in special relativity is formulated with differential forms in the cotangent bundle. We show that the Lie derivative plays the same role in the cotangent bundle that the covariant derivative plays in the tangent bundle. We also show that a cotangent bundle analog of Fermi--Walker transport can be based upon the, ''cotangent-geodesic'' equation, L/sub u/ω=0. This gives a generalization of the work by Kiehn on classical Hamiltonian mechanics to special relativity

Mass Transport Through Carbon Nanotube-Polystyrene Bundles

Science.gov (United States)

Lin, Rongzhou; Tran, Tuan

2016-05-01

Carbon nanotubes have been widely used as test channels to study nanofluidic transport, which has been found to have distinctive properties compared to transport of fluids in macroscopic channels. A long-standing challenge in the study of mass transport through carbon nanotubes (CNTs) is the determination of flow enhancement. Various experimental investigations have been conducted to measure the flow rate through CNTs, mainly based on either vertically aligned CNT membranes or individual CNTs. Here, we proposed an alternative approach that can be used to quantify the mass transport through CNTs. This is a simple method relying on the use of carbon nanotube-polystyrene bundles, which are made of CNTs pulled out from a vertically aligned CNT array and glued together by polystyrene. We experimentally showed by using fluorescent tagging that the composite bundles allowed measureable and selective mass transport through CNTs. This type of composite bundle may be useful in various CNT research areas as they are simple to fabricate, less likely to form macroscopic cracks, and offer a high density of CNT pores while maintaining the aligned morphology of CNTs.

QUENCH: A software package for the determination of quenching curves in Liquid Scintillation counting

International Nuclear Information System (INIS)

Cassette, Philippe

2016-01-01

In Liquid Scintillation Counting (LSC), the scintillating source is part of the measurement system and its detection efficiency varies with the scintillator used, the vial and the volume and the chemistry of the sample. The detection efficiency is generally determined using a quenching curve, describing, for a specific radionuclide, the relationship between a quenching index given by the counter and the detection efficiency. A quenched set of LS standard sources are prepared by adding a quenching agent and the quenching index and detection efficiency are determined for each source. Then a simple formula is fitted to the experimental points to define the quenching curve function. The paper describes a software package specifically devoted to the determination of quenching curves with uncertainties. The experimental measurements are described by their quenching index and detection efficiency with uncertainties on both quantities. Random Gaussian fluctuations of these experimental measurements are sampled and a polynomial or logarithmic function is fitted on each fluctuation by χ"2 minimization. This Monte Carlo procedure is repeated many times and eventually the arithmetic mean and the experimental standard deviation of each parameter are calculated, together with the covariances between these parameters. Using these parameters, the detection efficiency, corresponding to an arbitrary quenching index within the measured range, can be calculated. The associated uncertainty is calculated with the law of propagation of variances, including the covariance terms. - Highlights: • The program “QUENCH” is devoted to the interpolation of quenching curves in LSC. • Functions are fitted to experimental data with uncertainties in both quenching and efficiency. • The parameters of the fitting function and the associated covariance matrix are evaluated. • The detection efficiency and uncertainty corresponding to a given quenching index is calculated.

effects of various effects of various quenching media on quenching

African Journals Online (AJOL)

eobe

ABSTRACT. Evaluation of palm kernel oil, cotton seed oil and olive oil as quenching media of 0.509Wt%C medium carbon steel ... Quenching is an essential element in developing the .... machine, heat treatment furnace, Avery Denison Izod.

Acoustic analysis of sodium boiling stability tests using THORS bundle 6A

International Nuclear Information System (INIS)

Sheen, S.H.; Bobis, J.P.; Carey, W.M.

1977-01-01

Acoustic data from boiling stability tests on the THORS (Thermal-Hydraulic Out-of-Reactor Safety) facility are presented and discussed. The THORS sodium loop is a high temperature test facility that contains the bundle 6A, a full length stimulated fuel subassembly with nineteen electrically heated pins. Boiling stability tests on the THORS facility were designed to determine if a stable boiling region exists during the thermal hydraulic test at normal and off-normal conditions. Boiling was observed and the stable boiling region was determined. The acoustic data observed by three ANL sodium-immersible microphones have provided the following information: (1) the boiling signal is clearly observed and shows a correlation with the inlet flow fluctuations; (2) the signal level and the repetition rate of the boiling signal are directly related to the applied heat flux; (3) a typical boiling pulse consists of a high frequency signal due mainly to the bubble collapse and a low frequency (approximately 75 Hz) void oscillation; (4) a boiling pulse yields a frequency spectrum with significant amplitudes up to 80 KHz as compared with 4 KHz for background pulses; and (5) the frequency content of a boiling pulse can be mostly explained in terms of various resonance frequencies of the loop. The characterization of these data is pertinent to the design of sodium boiling detection systems

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.)

Prediction of interfacial area transport in a scaled 8×8 BWR rod bundle

Energy Technology Data Exchange (ETDEWEB)

Yang, X.; Schlegel, J.P.; Liu, Y.; Paranjape, S.; Hibiki, T.; Ishii, M. [School of Nuclear Engineering, Purdue University, 400 Central Dr., West Lafayette, IN 47907-2017 (United States); Bajorek, S.; Ireland, A. [U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001 (United States)

2016-12-15

In the two-fluid model, it is important to give an accurate prediction for the interfacial area concentration. In order to achieve this goal, the interfacial area transport equation has been developed. This study focuses on the benchmark of IATE performance in a rod bundle geometry. A set of interfacial area concentration source and sink term models are proposed for a rod bundle geometry based on the confined channel IATE model. This model was selected as a basis because of the relative similarity of the two geometries. Benchmarking of the new model with interfacial area concentration data in an 8×8 rod bundle test section which has been scaled from an actual BWR fuel bundle is performed. The model shows good agreement in bubbly and cap-bubbly flows, which are similar in many types of geometries, while it shows some discrepancy in churn-turbulent flow regime. This discrepancy may be due to the geometrical differences between the actual rod bundle test facility and the facility used to collect the data which benchmarked the original source and sink models.

CFD simulation of flow and heat transfer in Canadian SCWR bundles

International Nuclear Information System (INIS)

Podila, K.; Rao, Y.F.

2014-01-01

Within the Generation-IV (Gen-IV) International Forum, Atomic Energy of Canada Limited (AECL) is leading the effort in developing a conceptual design for the Canadian supercritical water-cooled reactor (SCWR). AECL proposed a new fuel bundle design with two rings of fuel elements placed between central flow tube and the pressure tube. In line with the scope of the conceptual design, the objective of the present CFD work is to aid in developing a bundle heat transfer correlation for the Canadian SCWR fuel bundle design. This paper presents results from an ongoing effort in determining the conditions favorable for possible occurrence of heat transfer deterioration (HTD) in the supercritical bundle flows. In the current investigation, a bare-rod bundle geometry was tested for the proposed fuel bundle design at 23.5, 25 and 28 MPa using STAR-CCM+ CFD code. Taking advantage of the design symmetry of the fuel bundle, only 1/32 of the computational domain was simulated. The SST k-ω turbulence model along with y + <1 was used in the simulations. For lower mass flow simulations, the increase of inlet temperature and operational pressure was found effective in reducing the occurrence of HTD. For higher mass flow simulations, normal heat transfer behaviour was observed except for the lower pressure range (23.5MPa). Ultimately, the goal of this study is to aid the development of a criterion for the onset of HTD in the proposed SCWR bundles, which is planned in the next phase of the project. (author)

ORNL rod-bundle heat-transfer test data. Volume 7. Thermal-Hydraulic Test Facility experimental data report for test series 3.07.9 - steady-state film boiling in upflow

International Nuclear Information System (INIS)

Mullins, C.B.; Felde, D.K.; Sutton, A.G.; Gould, S.S.; Morris, D.G.; Robinson, J.J.

1982-05-01

Thermal-Hydraulic Test Facility (THTF) test series 3.07.9 was conducted by members of the Oak Ridge National Laboratory Pressurized-Water Reactor (ORNL-PWR) Blowdown Heat Transfer (BDHT) Separate-Effects Program on September 11, September 18, and October 1, 1980. The objective of the program is to investigate heat transfer phenomena believed to occur in PWRs during accidents, including small- and large-break loss-of-coolant accidents. Test series 3.07.9 was designed to provide steady-state film boiling data in rod bundle geometry under reactor accident-type conditions. This report presents the reduced instrument responses for THTF test series 3.07.9. Also included are uncertainties in the instrument responses, calculated mass flows, and calculated rod powers

High-resolution imaging of retinal nerve fiber bundles in glaucoma using adaptive optics scanning laser ophthalmoscopy.

Science.gov (United States)

Takayama, Kohei; Ooto, Sotaro; Hangai, Masanori; Ueda-Arakawa, Naoko; Yoshida, Sachiko; Akagi, Tadamichi; Ikeda, Hanako Ohashi; Nonaka, Atsushi; Hanebuchi, Masaaki; Inoue, Takashi; Yoshimura, Nagahisa

2013-05-01

To detect pathologic changes in retinal nerve fiber bundles in glaucomatous eyes seen on images obtained by adaptive optics (AO) scanning laser ophthalmoscopy (AO SLO). Prospective cross-sectional study. Twenty-eight eyes of 28 patients with open-angle glaucoma and 21 normal eyes of 21 volunteer subjects underwent a full ophthalmologic examination, visual field testing using a Humphrey Field Analyzer, fundus photography, red-free SLO imaging, spectral-domain optical coherence tomography, and imaging with an original prototype AO SLO system. The AO SLO images showed many hyperreflective bundles suggesting nerve fiber bundles. In glaucomatous eyes, the nerve fiber bundles were narrower than in normal eyes, and the nerve fiber layer thickness was correlated with the nerve fiber bundle widths on AO SLO (P fiber layer defect area on fundus photography, the nerve fiber bundles on AO SLO were narrower compared with those in normal eyes (P optic disc, the nerve fiber bundle width was significantly lower, even in areas without nerve fiber layer defect, in eyes with glaucomatous eyes compared with normal eyes (P = .026). The mean deviations of each cluster in visual field testing were correlated with the corresponding nerve fiber bundle widths (P = .017). AO SLO images showed reduced nerve fiber bundle widths both in clinically normal and abnormal areas of glaucomatous eyes, and these abnormalities were associated with visual field defects, suggesting that AO SLO may be useful for detecting early nerve fiber bundle abnormalities associated with loss of visual function. Copyright © 2013 Elsevier Inc. All rights reserved.

Nanotube bundle oscillators: Carbon and boron nitride nanostructures

International Nuclear Information System (INIS)

Thamwattana, Ngamta; Hill, James M.

2009-01-01

In this paper, we investigate the oscillation of a fullerene that is moving within the centre of a bundle of nanotubes. In particular, certain fullerene-nanotube bundle oscillators, namely C 60 -carbon nanotube bundle, C 60 -boron nitride nanotube bundle, B 36 N 36 -carbon nanotube bundle and B 36 N 36 -boron nitride nanotube bundle are studied using the Lennard-Jones potential and the continuum approach which assumes a uniform distribution of atoms on the surface of each molecule. We address issues regarding the maximal suction energies of the fullerenes which lead to the generation of the maximum oscillation frequency. Since bundles are also found to comprise double-walled nanotubes, this paper also examines the oscillation of a fullerene inside a double-walled nanotube bundle. Our results show that the frequencies obtained for the oscillation within double-walled nanotube bundles are slightly higher compared to those of single-walled nanotube bundle oscillators. Our primary purpose here is to extend a number of established results for carbon to the boron nitride nanostructures.

Aerosol retention in the flooded steam generator bundle during SGTR

International Nuclear Information System (INIS)

Lind, Terttaliisa; Dehbi, Abdel; Guentay, Salih

2011-01-01

Research highlights: → High retention of aerosol particles in a steam generator bundle flooded with water. → Increasing particle inertia, i.e., particle size and velocity, increases retention. → Much higher retention of aerosol particles in the steam generator bundle flooded with water than in a dry bundle. → Much higher retention of aerosol particles in the steam generator bundle than in a bare pool. → Bare pool models have to be adapted to be applicable for flooded bundles. - Abstract: A steam generator tube rupture in a pressurized water reactor may cause accidental release of radioactive particles into the environment. Its specific significance is in its potential to bypass the containment thereby providing a direct pathway of the radioactivity from the primary circuit to the environment. Under certain severe accident scenarios, the steam generator bundle may be flooded with water. In addition, some severe accident management procedures are designed to minimize the release of radioactivity into the environment by flooding the defective steam generator secondary side with water when the steam generator has dried out. To extend our understanding of the particle retention phenomena in the flooded steam generator bundle, tests were conducted in the ARTIST and ARTIST II programs to determine the effect of different parameters on particle retention. The effects of particle type (spherical or agglomerate), particle size, gas mass flow rate, and the break submergence on particle retention were investigated. Results can be summarized as follows: increasing particle inertia was found to increase retention in the flooded bundle. Particle shape, i.e., agglomerate or spherical structure, did not affect retention significantly. Even with a very low submergence, 0.3 m above the tube break, significant aerosol retention took place underlining the importance of the jet-bundle interactions close to the tube break. Droplets were entrained from the water surface with

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

International Nuclear Information System (INIS)

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

1991-01-01

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

Submersion Quenching of Undercooled Liquid Metals in an Electrostatic Levitator

Science.gov (United States)

SanSoucie, Michael P.; Rogers, Jan R.

2016-01-01

The NASA Marshall Space Flight Center (MSFC) electrostatic levitation (ESL) laboratory has a long history of providing materials research and thermophysical property data. The laboratory has recently added a new capability, a rapid quench system. This system allows samples to be dropped into a quench vessel that can be filled with a low melting point material, such as a gallium or indium alloy. Thereby allowing rapid quenching of undercooled liquid metals and alloys. This is the first submersion quench system inside an electrostatic levitator. The system has been tested successfully with samples of zirconium, iron-cobalt alloys, titanium-zirconium-nickel alloys, and silicon-cobalt alloys. This rapid quench system will allow materials science studies of undercooled materials and new materials development, including studies of metastable phases and transient microstructures. In this presentation, the system is described and some initial results are presented.

Holomorphic bundles over elliptic manifolds

International Nuclear Information System (INIS)

Morgan, J.W.

2000-01-01

In this lecture we shall examine holomorphic bundles over compact elliptically fibered manifolds. We shall examine constructions of such bundles as well as (duality) relations between such bundles and other geometric objects, namely K3-surfaces and del Pezzo surfaces. We shall be dealing throughout with holomorphic principal bundles with structure group GC where G is a compact, simple (usually simply connected) Lie group and GC is the associated complex simple algebraic group. Of course, in the special case G = SU(n) and hence GC = SLn(C), we are considering holomorphic vector bundles with trivial determinant. In the other cases of classical groups, G SO(n) or G = Sympl(2n) we are considering holomorphic vector bundles with trivial determinant equipped with a non-degenerate symmetric, or skew symmetric pairing. In addition to these classical cases there are the finite number of exceptional groups. Amazingly enough, motivated by questions in physics, much interest centres around the group E8 and its subgroups. For these applications it does not suffice to consider only the classical groups. Thus, while often first doing the case of SU(n) or more generally of the classical groups, we shall extend our discussions to the general semi-simple group. Also, we shall spend a good deal of time considering elliptically fibered manifolds of the simplest type, namely, elliptic curves

Post-irradiation examination of overheated fuel bundles

International Nuclear Information System (INIS)

Sears, D.F.; Primeau, M.F.; Leach, D.A.

1995-01-01

Post-irradiation examinations (PIE) were conducted on prototype 43-element CANDU fuel bundles that overheated during test irradiations in the NRU reactor. PIE revealed that the bundles remained physically intact, but on several elements the Zr-4 sheath collapsed into axial gaps between the pellet stack and end caps, between adjacent pellets within the stacks, and into missing pellet chips and cracks. Helium pressurization tests showed that none of the collapsed elements leaked. Hydride blisters were discovered on a few elements, but the source of the hydrogen was not linked to a breach of the cladding or end caps. These defects were attributed to primary hydriding. Microstructural changes in the fuel and cladding indicate that the cladding-was briefly exposed to temperatures in the range 600-800 o C and pressures above 11.2 MPa. The results show that Zr-4 cladding behaves in a highly ductile manner during such transient, high-temperature and high-pressure excursions. (author)

Post-irradiation examination of overheated fuel bundles

International Nuclear Information System (INIS)

Sears, D.F.; Primeau, M.F.; Leach, D.A.

1997-08-01

Post-irradiation examinations (PIE) were conducted on prototype 43-element CANDU fuel bundles that overheated during test irradiations in the NRU reactor. PIE revealed that the bundles remained physically intact, but on several elements the Zr-4 sheath collapsed into axial gaps between the pellet stack and end caps, between adjacent pellets within the stacks, and into missing pellet chips and cracks. Helium pressurization tests showed that none of the collapsed elements leaked. Hydride blisters were discovered on a few elements, but the source of the hydrogen was.not linked to a breach of the cladding or end caps. These defects were attributed to primary hydriding. Microstructural changes in the fuel and cladding indicate that the cladding was briefly exposed to temperatures in the range 600-800 o C and pressures above 11.2MPa. The results show that Zr-4 cladding behaves in a highly ductile manner during such transient, high-temperature and high-pressure excursions. (author)

SCADOP: Phenomenological modeling of dryout in nuclear fuel rod bundles

Energy Technology Data Exchange (ETDEWEB)

Dasgupta, Arnab, E-mail: arnie@barc.gov.in; Chandraker, D.K., E-mail: dineshkc@barc.gov.in; Vijayan, P.K., E-mail: vijayanp@barc.gov.in

2015-11-15

Highlights: • Phenomenological model for annular flow dryout is presented. • The model evaluates initial entrained fraction using a new methodology. • The history effect in annular flow is predicted and validated. • Rod bundle dryout is predicted using subchannel methodology. • Model is validated against experimental dryout data in tubes and rod bundles. - Abstract: Analysis and prediction of dryout is of important consequence to safety of nuclear fuel clusters of boiling water type of reactors. Traditionally, experimental correlations are used for dryout predictions. Since these correlations are based on operating parameters and do not aim to model the underlying phenomena, there has been a proliferation of the correlations, each catering to some specific bundle geometry under a specific set of operating conditions. Moreover, such experiments are extremely costly. In general, changes in tested bundle geometry for improvement in thermal-hydraulic performance would require re-experimentation. Understanding and modeling the basic processes leading to dryout in flow boiling thus has great incentive. Such a model has the ability to predict dryout in any rod bundle geometry, unlike the operating parameter based correlation approach. Thus more informed experiments can be carried out. A good model can, reduce the number of experiments required during the iterations in bundle design. In this paper, a phenomenological model as indicated above is presented. The model incorporates a new methodology to estimate the Initial Entrained Fraction (IEF), i.e., entrained fraction at the onset of annular flow. The incorporation of this new methodology is important since IEF is often assumed ad-hoc and sometimes also used as a parameter to tune the model predictions to experimental data. It is highlighted that IEF may be low under certain conditions against the general perception of a high IEF due to influence of churn flow. It is shown that the same phenomenological model is

Thermohydraulic tests of 3x3-rod bundle maquette

International Nuclear Information System (INIS)

Ladeira, L.C.D.

1986-10-01

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

Post CHF heat transfer and quenching

International Nuclear Information System (INIS)

Nelson, R.A.; Condie, K.G.

1980-01-01

This paper describes quantitatively new mechanisms in the post-CHF regime which provide understanding and predictive capability for several current two-phase forced convective heat transfer problems. These mechanisms are important in predicting rod temperature turnaround and quenching during the reflood phase of either a hypothetical loss-of-coolant accident (LOCA) or the FLECHT and Semiscale experiments. The mechanisms are also important to the blowdown phase of a LOCA or the recent Loss-of-Fluid Test (LOFT) experiments L2-2 and L2-3, which were 200% cold leg break transients. These LOFT experiments experienced total core quenching in the early part of the blowdown phase at high (1000 psia) pressures. The mechanisms are also important to certain pressurized water reactor (PWR) operational transients where the reactor may operate in the post-CHF regime for short periods of time. Accurate prediction of the post-CHF heat transfer including core quench during these transients is of prime importance to limit maximum cladding temperatures and prevent cladding deformation

Lexical bundles in an advanced INTOCSU writing class and engineering texts: A functional analysis

Science.gov (United States)

Alquraishi, Mohammed Abdulrahman

The purpose of this study is to investigate the functions of lexical bundles in two corpora: a corpus of engineering academic texts and a corpus of IEP advanced writing class texts. This study is concerned with the nature of formulaic language in Pathway IEPs and engineering texts, and whether those types of texts show similar or distinctive formulaic functions. Moreover, the study looked into lexical bundles found in an engineering 1.26 million-word corpus and an ESL 65000-word corpus using a concordancing program. The study then analyzed the functions of those lexical bundles and compared them statistically using chi-square tests. Additionally, the results of this investigation showed 236 unique frequent lexical bundles in the engineering corpus and 37 bundles in the pathway corpus. Also, the study identified several differences between the density and functions of lexical bundles in the two corpora. These differences were evident in the distribution of functions of lexical bundles and the minimal overlap of lexical bundles found in the two corpora. The results of this study call for more attention to formulaic language at ESP and EAP programs.

Experiments on the quench behavior of fuel rods

International Nuclear Information System (INIS)

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

1995-01-01

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

Experiments on the quench behavior of fuel rods

Energy Technology Data Exchange (ETDEWEB)

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

1995-08-01

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

Muon bundles from the Universe

Directory of Open Access Journals (Sweden)

Kankiewicz P.

2018-01-01

Full Text Available Recently the CERN ALICE experiment, in its dedicated cosmic ray run, observed muon bundles of very high multiplicities, thereby confirming similar findings from the LEP era at CERN (in the CosmoLEP project. Significant evidence for anisotropy of arrival directions of the observed high multiplicity muonic bundles is found. Estimated directionality suggests their possible extragalactic provenance. We argue that muonic bundles of highest multiplicity are produced by strangelets, hypothetical stable lumps of strange quark matter infiltrating our Universe.

Infinitesimal bundles and projective relativity

International Nuclear Information System (INIS)

Evans, G.T.

1973-01-01

An intrinsic and global presentation of five-dimensional relativity theory is developed, in which special coordinate conditions are replaced by conditions of Lie invariance. The notion of an infinitesimal bundle is introduced, and the theory of connexions on principal bundles is extended to infinitesimal bundles. Global aspects of projective relativity are studied: it is shown that projective relativity can describe almost any space-time. In particular, it is not necessary to assume that the electromagnetic field have a global potential. (author)

Influence of partial blockage of a BWR bundle on heat transfer, cladding temperature, and quenching during bottom flooding or top spraying under simulated LOCA conditions

International Nuclear Information System (INIS)

Brand, B.; Gaul, H.P.; Sarkar, J.

1982-01-01

In a test facility with two parallel boiling water reactor fuel assemblies, experiments were carried out with top spray and bottom flooding, simulating loss-of-coolant accident (LOCA) conditions. The flow area restriction, caused by the ballooning of fuel rod cladding within one of the bundles, was provided by blockage plates, which had reductions of 37% in one case and in a second series 70% of the flow area. Test parameters were system pressure (1, 5, and 10 bars), spray (0.68 and 1.02 m 3 /h) and flooding rates (1.5,2, and 3.3 cm/s), power input (520 and 614 kW), and the initial cladding temperature (600 and 800 0 C at midplane) of the heaters. The test results showed no significant variations from those without blockage, except in the blocked region. An enhancement of heat transfer was observed in a close region downstream from the blockage in cases such as bottom flooding and top spray tests. The results will serve the purpose of code verification for reactor LOCA analysis

Magnet Quench 101

OpenAIRE

Bottura, L.

2014-01-01

This paper gives a broad summary of the physical phenomena associated with the quench of a superconducting magnet. This paper gives a broad summary of the physical phenomena associated with the quench of a superconducting magnet.

Experimental study of mixing in a square array rod bundle with grid spacer

International Nuclear Information System (INIS)

Zong Guifang; Cai Zuti; Zhang Demei

1989-01-01

This paper describes the experimental study of mixing in a full scale 15x15 square array rod bundle fuel assembly with 10 mm diameter and 13.3 mm pitch. The experiment was carried out in an open water loop, K 2 CrO 4 was used as tracer. Each subchannel was sampled at the open bundle outlet. Titration, spectrophotometry and fibreoptic methods were used to measure the concentration. The Reynolds numbers ranged from 2.12x10 4 to 4.37x10 4 . For the turbulent mixing of the bare rod bundle, the results of this study agreed with the formulas recommended by other authors. Both flow visualisation studies and the quantitative analysis indicated that flow scattering caused by the grid has a little effect on the mixing. The cause has been examined in this paper. (orig.)

Fiber bundles in non-relativistic quantum mechanics

International Nuclear Information System (INIS)

Moylan, P.

1979-11-01

The problem of describing a quantum-mechanical system with symmetry by a fiber bundle is considered. The quantization of a fiber bundle is introduced. Fiber bundles for the Kepler problem and the rotator are constructed. The fiber bundle concept provides a new model for a physical system: it provides a model for an elementary particle with extension having integral values of spin. 5 figures

Quenching experiments on niobium

International Nuclear Information System (INIS)

Schwirtlich, I.A.; Schultz, H.; Max-Planck-Institut fuer Metallforschung, Stuttgart

1980-01-01

High-purity niobium wire specimens have been quenched in superfluid helium from near the melting point in order to obtain information on vacancies in this material. The quenched-in resistivity Δsub(pQ) for a quench from 2600 K was very small (approximately 0.3 x 10 -12 Ω m) and near the limit of detection. It is assumed that large quenching losses are responsible for the small quenched-in resistance. From the experimental cooling curve estimates have been made for the formation and migration enthalpies (Hsub(1V)sup(F), Hsub(1V)sup(M)), where Hsub(1V)sup(M)+Hsub(1V)sup(F)=Qsub(1V)sup(SD)=3.62 ev. For Ssub(1V)sup(F), the formation entropy, two different values were assumed. (author)

Accelerator Magnet Quench Heater Technology and Quality Control Tests for the LHC High Luminosity Upgrade

CERN Document Server

AUTHOR|(CDS)2132435; Seifert, Thomas

The High Luminosity upgrade of the Large Hadron Collider (HL-LHC) foresees the installation of new superconducting Nb$_{3}$Sn magnets. For the protection of these magnets, quench heaters are placed on the magnet coils. The quench heater circuits are chemically etched from a stainless steel foil that is glued onto a flexible Polyimide film, using flexible printed circuit production technology. Approximately 500 quench heaters with a total length of about 3000 m are needed for the HL-LHC magnets. In order to keep the heater circuit electrical resistance in acceptable limits, an approximately 10 µm-thick Cu coating is applied onto the steel foil. The quality of this Cu coating has been found critical in the quench heater production. The work described in this thesis focuses on the characterisation of Cu coatings produced by electrolytic deposition, sputtering and electron beam evaporation. The quality of the Cu coatings from different manufacturers has been assessed for instance by ambient temperature electrica...

Measurement and CFD calculation of spacer loss coefficient for a tight-lattice fuel bundle

International Nuclear Information System (INIS)

In, Wang Kee; Shin, Chang Hwan; Kwack, Young Kyun; Lee, Chi Young

2015-01-01

Highlights: • Experiment and CFD analysis evaluated the pressure drop in a spacer grid. • The measurement and CFD errors for the spacer loss coefficient were estimated. • The spacer loss coefficient for the dual-cooled annular fuel bundle was determined. • The CFD prediction agrees with the measured spacer loss coefficient within 8%. - Abstract: An experiment and computational fluid dynamics (CFD) analysis were performed to evaluate the pressure drop in a spacer grid for a dual-cooled annular fuel (DCAF) bundle. The DCAF bundle for the Korean optimum power reactor (OPR1000) is a 12 × 12 tight-lattice rod array with a pitch-to-diameter ratio of 1.08 owing to a larger outer diameter of the annular fuel rod. An experiment was conducted to measure the pressure drop in spacer grid for the DCAF bundle. The test bundle is a full-size 12 × 12 rod bundle with 11 spacer grid. The test condition covers a Reynolds number range of 2 × 10 4 –2 × 10 5 by changing the temperature and flow rate of water. A CFD analysis was also performed to predict the pressure drop through a spacer grid using the full-size and partial bundle models. The pressure drop and loss coefficient of a spacer grid were predicted and compared with the experimental results. The CFD predictions of spacer pressure drop and loss coefficient agree with the measured values within 8%. The spacer loss coefficient for the DCAF bundle is estimated to be approximately 1.50 at a nominal operating condition of OPR1000, i.e., Re = 4 × 10 5

Twisted Vector Bundles on Pointed Nodal Curves

Indian Academy of Sciences (India)

Abstract. Motivated by the quest for a good compactification of the moduli space of -bundles on a nodal curve we establish a striking relationship between Abramovich's and Vistoli's twisted bundles and Gieseker vector bundles.

LHC magnet quench protection system

Science.gov (United States)

Coull, L.; Hagedorn, D.; Remondino, V.; Rodriguez-Mateos, F.

1994-07-01

The quench protection system for the superconducting magnets of the CERN Large Hadron Collider (LHC) is described. The system is based on the so called 'cold diode' concept. In a group of series connected magnets if one magnet quenches then the magnetic energy of all the magnets will be dissipated in the quenched magnet so destroying it. This is avoided by by-passing the quenched magnet and then rapidly de-exciting the unquenched magnets. For the LHC machine it is foreseen to use silicon diodes situated inside the cryostat as by-pass elements - so called 'cold diodes'. The diodes are exposed to some 50 kGray of radiation during a 10 year operation life-time. The high energy density of the LHC magnets (500 kJ/m) coupled with the relatively slow propagation speed of a 'natural' quench (10 to 20 m/s) can lead to excessive heating of the zone where the quench started and to high internal voltages. It is therefore necessary to detect quickly the incipient quench and fire strip heaters which spread the quench out more quickly over a large volume of the magnet. After a quench the magnet chain must be de-excited rapidly to avoid spreading the quench to other magnets and over-heating the by-pass diode. This is done by switching high-power energy-dump resistors in series with the magnets. The LHC main ring magnet will be divided into 16 electrically separated units which has important advantages.

LHC magnet quench protection system

International Nuclear Information System (INIS)

Coull, L.; Hagedorn, D.; Remondino, V.; Rodriguez-Mateos, F.

1994-01-01

The quench protection system for the superconducting magnets of the CERN Large Hadron Collider (LHC) is described. The system is based on the so called ''cold diode'' concept. In a group of series connected magnets if one magnet quenches then the magnetic energy of all the magnets will be dissipated in the quenched magnet so destroying it. This is avoided by by-passing the quenched magnet and then rapidly de-exciting the unquenched magnets. For the LHC machine it is foreseen to use silicon diodes situated inside the cryostat as by-pass elements--so called ''cold diodes''. The diodes are exposed to some 50 kGray of radiation during a 10 year operation life-time. The high energy density of the LHC magnets (500 kJ/m) coupled with the relatively slow propagation speed of a ''natural'' quench (10 to 20 m/s) can lead to excessive heating of the zone where the quench started and to high internal voltages. It is therefore necessary to detect quickly the incipient quench and fire strip heaters which spread the quench out more quickly over a large volume of the magnet. After a quench the magnet chain must be de-excited rapidly to avoid spreading the quench to other magnets and over-heating the by-pass diode. This is done by switching high-power energy-dump resistors in series with the magnets. The LHC main ring magnet will be divided into 16 electrically separated units which has important advantages

QUENCH: A software package for the determination of quenching curves in Liquid Scintillation counting.

Science.gov (United States)

Cassette, Philippe

2016-03-01

In Liquid Scintillation Counting (LSC), the scintillating source is part of the measurement system and its detection efficiency varies with the scintillator used, the vial and the volume and the chemistry of the sample. The detection efficiency is generally determined using a quenching curve, describing, for a specific radionuclide, the relationship between a quenching index given by the counter and the detection efficiency. A quenched set of LS standard sources are prepared by adding a quenching agent and the quenching index and detection efficiency are determined for each source. Then a simple formula is fitted to the experimental points to define the quenching curve function. The paper describes a software package specifically devoted to the determination of quenching curves with uncertainties. The experimental measurements are described by their quenching index and detection efficiency with uncertainties on both quantities. Random Gaussian fluctuations of these experimental measurements are sampled and a polynomial or logarithmic function is fitted on each fluctuation by χ(2) minimization. This Monte Carlo procedure is repeated many times and eventually the arithmetic mean and the experimental standard deviation of each parameter are calculated, together with the covariances between these parameters. Using these parameters, the detection efficiency, corresponding to an arbitrary quenching index within the measured range, can be calculated. The associated uncertainty is calculated with the law of propagation of variances, including the covariance terms. Copyright © 2015 Elsevier Ltd. All rights reserved.

Research reactor fuel bundle design review by means of hydrodynamic testing; Ensayos hidrodinamicos para verificacion de diseno de un elemento combustible para reactores de investigacion

Energy Technology Data Exchange (ETDEWEB)

Pastorini, A; Belinco, C [Comision Nacional de Energia Atomica, San Martin (Argentina). Centro Atomico Constituyentes

1998-12-31

During the design steps of a fuel bundle for a nuclear reactor, some vibration tests are usually necessary to verify the prototype dynamical response characteristics and the structural integrity. To perform these tests, the known hydrodynamic loop facilities are used to evaluate the vibrational response of the bundle under the different flow conditions that may appear in the reactor. This paper describes the tests performed on a 19 plate fuel bundle prototype designed for a low power research reactor. The tests were done in order to know the dynamical characteristics of the plates and also of the whole bundle under different flow rate conditions. The paper includes a description of the test facilities and the results obtained during the dynamical characterization tests and some preliminary comments about the tests under flowing water are also presented. (author) 4 refs., 12 figs., 4 tabs. [Espanol] Durante el diseno de un elemento combustible para un reactor nuclear se requiere de la realizacion de ensayos con el objeto de verificar el comportamiento de ese diseno y permitir, de ser necesario, la introduccion de modificaciones al mismo. Para verificar las caracteristicas de respuesta dinamica e integridad estructural, se realizan ensayos de vibraciones que incluyen someter al prototipo a condiciones de circulacion del fluido similares a las que soportara durante la operacion del reactor. Estos ensayos se realizan en facilidades de ensayos conocidas como circuitos hidrodinamicos, que permiten no solo someter el prototipo al flujo de fluido, sino tambien obtener una adecuada caracterizacion de la respuesta del mismo a traves del luso de sensores de distinto tipo. En este trabajo se describen los ensayos realizados sobre un prototipo de elemento combustible de 19 placas destinado a un reactor de investigacion multiproposito de baja potencia. Los ensayos tuvieron como objetivo conocer la respuesta dinamica de las placas individuales y del elemento combustible en su

Adjustment of pipe flow explicit friction factor equations for application to tube bundles

International Nuclear Information System (INIS)

Wiltz, Christopher L.; Bowen, Mike D.; Von Olnhausen, Wayne A.

2005-01-01

Full text of publication follows: The accurate determination of single phase friction losses or friction pressure drop in tube bundles is essential in the thermal-hydraulic analyses of components such as nuclear fuel assemblies, heat exchangers and steam generators. Such friction losses are normally calculated using a friction factor, f, along with the experimental observation that the friction pressure drop in a pipe is proportional to the dynamic pressure (1/2 ρV 2 ) of the flow: ΔP = 1/2 ρV 2 (fL/D). In this equation L is the pipe or tube bundle length and D is the hydraulic diameter of the pipe or tube bundle. The friction factor is normally calculated using one of a number of explicit friction factor equations. A significant amount of work has been accomplished in developing explicit friction factor equations. These explicit equations range from approximations, which were developed for ease of numerical evaluation, to those which are mathematically complex but yield very good fits to the test data. These explicit friction factor equations are based on a large experimental data base, nearly all of which comes from pipe flow geometry information, and have been historically applied to tube bundles. This paper presents an adjustment method which may be applied to various explicit friction factor equations developed for pipe flow to accurately predict the friction factor for tube bundles. The characteristic of the adjustment is based on experimental friction pressure loss data obtained by Framatome ANP through flow testing of a nuclear fuel assembly (tube bundle) at its Richland Test Facility (RTF). Through adjustment of previously developed explicit friction factor equations for pipe flow, the vast amount of historical development and experimentation in the area of single phase pipe flow friction loss may be incorporated into the evaluation of single phase friction losses within tube bundles. Comparisons of the application of one or more of the previously

Preliminary report: NIF laser bundle review

International Nuclear Information System (INIS)

Tietbohl, G.L.; Larson, D.W.; Erlandson, A.C.

1995-01-01

As requested in the guidance memo 1 , this committe determined whether there are compelling reasons to recommend a change from the NIF CDR baseline laser. The baseline bundle design based on a tradeoff between cost and technical risk, which is replicated four times to create the required 192 beams. The baseline amplifier design uses bottom loading 1x4 slab and flashlamp cassettes for amplifier maintenance and large vacuum enclosures (2.5m high x 7m wide in cross-section for each of the two spatial filters in each of the four bundles. The laser beams are arranged in two laser bays configured in a u-shape around the target area. The entire bundle review effort was performed in a very short time (six weeks) and with limited resources (15 personnel part-time). This should be compared to the effort that produced the CDR design (12 months, 50 to 100 personnel). This committee considered three alternate bundle configurations (2x2, 4x2, and 4x4 bundles), and evaluated each bundle against the baseline design using the seven requested issues in the guidance memo: Cost; schedule; performance risk; maintainability/operability; hardware failure cost exposure; activation; and design flexibility. The issues were reviewed to identify differences between each alternate bundle configuration and the baseline

A phenomenological model of the thermal-hydraulics of convective boiling during the quenching of hot rod bundles: Part 2, Assessment of the model with steady-state and transient post-CHF data

International Nuclear Information System (INIS)

Unal, C.; Nelson, R.

1991-01-01

After completing the thermal-hydraulic model developed in a companion paper, we performed assessment calculations of the model using steady-state and transient post-critical heat flux (CHF) data. This paper discusses the results of those calculations. The hot-patch model, in conjunction with the other thermal-hydraulic models, was capable of modeling the Winfrith post-CHF hot-patch experiments. The hot-patch model kept the wall temperatures at the specified levels in the hot-patch regions and did not allow any quench-front propagation from either the bottom or the top of the test section. Among the four Winfrith runs selected to assess the hot-patch model, the average deviation in hot-patch power predictions was 15.4%, indicating reasonable predictions of the amount of energy transferred to the fluid by the hot patch. The interfacial heat-transfer model tended to slightly under-predict the vapor temperatures. The maximum difference between calculated and measured vapor superheats was 20%, with a 10% difference for the remainder of the runs considered. The wall-to-fluid heat transfer was predicted reasonably well, and the predicted wall superheats were in reasonable agreement with measured data with a maximum relative error of less than 13%. The effects of pressure, test section power, and flow rate on the axial variation of tube wall temperature are predicted reasonably well for a large range of operating parameters. A comparison of the predicted and measured local wall. The thermal-hydraulic model in TRAC/PF1-MOD2 was used to predict the axial variation of void fraction as measured in Winfrith post-CHF tests. The predictions for reflood calculations were reasonable. The model correctly predicted the trends in void fraction as a result of the effect of pressure and power, with the effect of pressure being more apparent than that of power. 13 refs

Warps, grids and curvature in triple vector bundles

Science.gov (United States)

Flari, Magdalini K.; Mackenzie, Kirill

2018-06-01

A triple vector bundle is a cube of vector bundle structures which commute in the (strict) categorical sense. A grid in a triple vector bundle is a collection of sections of each bundle structure with certain linearity properties. A grid provides two routes around each face of the triple vector bundle, and six routes from the base manifold to the total manifold; the warps measure the lack of commutativity of these routes. In this paper we first prove that the sum of the warps in a triple vector bundle is zero. The proof we give is intrinsic and, we believe, clearer than the proof using decompositions given earlier by one of us. We apply this result to the triple tangent bundle T^3M of a manifold and deduce (as earlier) the Jacobi identity. We further apply the result to the triple vector bundle T^2A for a vector bundle A using a connection in A to define a grid in T^2A . In this case the curvature emerges from the warp theorem.

Investigation of velocity distribution in an inner subchannel of wire wrapped fuel pin bundle of sodium-cooled fast reactor

International Nuclear Information System (INIS)

Nishimura, Masahiro; Kamide, Hideki; Ohshima, Hiroyuki; Kobayashi, Jun; Sato, Hiroyuki

2011-01-01

A sodium cooled fast reactor is designed to attain a high burn-up of core fuel in commercialized fast reactor cycle systems. In high burn-up fuel subassemblies, deformation of fuel pin due to the swelling and thermal bowing may decrease local flow velocity via change of flow area in the subassembly and influence the heat removal capability. Therefore, it is important to obtain the detail of flow velocity distribution in a wire wrapped pin bundle. In this study, water experiments were carried out to investigate the detailed velocity distribution in a subchannel of nominal pin geometry as the first step. These basic data are not only useful for understanding of pin bundle thermal hydraulics but also a code validation. A wire-wrapped 3-pin bundle water model was applied to investigate the detailed velocity distribution in the subchannel which is surrounded by 3 pins with wrapping wire. The test section consists of an irregular hexagonal acrylic duct tube and three pins made of fluorinated resin pins which has nearly the same refractive index with that of water and a high light transmission rate. This enables to visualize the central subchannel through the pins. The velocity distribution in the central subchannel with the wrapping wire was measured by PIV (Particle Image Velocimetry) through a side wall of the duct tube. Typical flow velocity conditions in the pin bundle were 0.36m/s (Re=2,700) and 1.6m/s (Re=13,500). Influence of the wrapping wire on the velocity distributions in vertical and horizontal directions was confirmed. A clockwise swirl flow around the wire was found in subchannel. Significant differences were not recognized between the two cases of Re=2,700 and 13,500 concerning flow patterns. (author)

Quench detection/protection of an HTS coil by AE signals

International Nuclear Information System (INIS)

Yoneda, M.; Nanato, N.; Aoki, D.; Kato, T.; Murase, S.

2011-01-01

A quench detection/protection system based on measuring AE signals was developed. The system was tested for a Bi2223 coil. Temperature rise after a quench occurrence was restrained at very low value. The normal zone propagation velocities in high T c superconductors are slow at high operation temperature and therefore local and excessive temperature rise generates at quench occurrence, and then the superconductors are degraded or burned. Therefore it is essential to detect the temperature rise in high T c superconducting (HTS) coils as soon as possible and protect them. The authors have presented a quench detection method for HTS coils by time-frequency visualization of AE signals and have shown its usefulness for a HTS coil with height and outer diameter of 40-50 mm. In this paper, the authors present a quench detection/protection system based on superior method in quench detection time to the previous method and show its usefulness for a larger HTS coil (height and outer diameter: 160-190 mm) than the previous coil.

46 CFR 160.066-13 - Technical tests.

Science.gov (United States)

2010-10-01

... 46 Shipping 6 2010-10-01 2010-10-01 false Technical tests. 160.066-13 Section 160.066-13 Shipping....066-13 Technical tests. (a) The following conditions apply to technical tests as described in this... signals are protected by sealed packaging, then the conditioning for the technical tests must be conducted...

Bundling ecosystem services in Denmark

DEFF Research Database (Denmark)

Turner, Katrine Grace; Odgaard, Mette Vestergaard; Bøcher, Peder Klith

2014-01-01

We made a spatial analysis of 11 ecosystem services at a 10 km × 10 km grid scale covering most of Denmark. Our objective was to describe their spatial distribution and interactions and also to analyze whether they formed specific bundle types on a regional scale in the Danish cultural landscape....... We found clustered distribution patterns of ecosystem services across the country. There was a significant tendency for trade-offs between on the one hand cultural and regulating services and on the other provisioning services, and we also found the potential of regulating and cultural services...... to form synergies. We identified six distinct ecosystem service bundle types, indicating multiple interactions at a landscape level. The bundle types showed specialized areas of agricultural production, high provision of cultural services at the coasts, multifunctional mixed-use bundle types around urban...

Post-test calculation and uncertainty analysis of the experiment QUENCH-07 with the system code ATHLET-CD

International Nuclear Information System (INIS)

Austregesilo, Henrique; Bals, Christine; Trambauer, Klaus

2007-01-01

In the frame of developmental assessment and code validation, a post-test calculation of the test QUENCH-07 was performed with ATHLET-CD. The system code ATHLET-CD is being developed for best-estimate simulation of accidents with core degradation and for evaluation of accident management procedures. It applies the detailed models of the thermal-hydraulic code ATHLET in an efficient coupling with dedicated models for core degradation and fission products behaviour. The first step of the work was the simulation of the test QUENCH-07 applying the modelling options recommended in the code User's Manual (reference calculation). The global results of this calculation showed a good agreement with the measured data. This calculation was complemented by a sensitivity analysis in order to investigate the influence of a combined variation of code input parameters on the simulation of the main phenomena observed experimentally. Results of this sensitivity analysis indicate that the main experimental measurements lay within the uncertainty range of the corresponding calculated values. Among the main contributors to the uncertainty of code results are the heat transfer coefficient due to forced convection to superheated steam-argon mixture, the thermal conductivity of the shroud isolation and the external heater rod resistance. Uncertainties on modelling of B 4 C oxidation do not affect significantly the total calculated hydrogen release rates

Element bow profiles from new and irradiated CANDU fuel bundles

International Nuclear Information System (INIS)

Dennier, D.; Manzer, A.M.; Ryz, M.A.

1996-01-01

Improved methods of measuring element profiles on new CANDU fuel bundles were developed at the Sheridan Park Engineering Laboratory, and have now been applied in the hot cells at Whiteshell Laboratories. For the first time, the outer element profiles have been compared between new, out-reactor tested, and irradiated fuel elements. The comparison shows that irradiated element deformation is similar to that observed on elements in out-reactor tested bundles. In addition to the restraints applied to the element via appendages, the element profile appears to be strongly influenced by gravity and the end loads applied by local deformation of the endplate. Irradiation creep in the direction of gravity also tends to be a dominant factor. (author)

Cherenkov radiation effects on counting efficiency in extremely quenched liquid scintillation samples

International Nuclear Information System (INIS)

Grau Carles, A.; Grau Malonda, A.; Rodriguez Barquero, L.

1993-01-01

The CIEMAT/NIST tracer method has successfully standardized nuclides with diverse quench values and decay schemes in liquid scintillation counting. However, the counting efficiency is computed inaccurately for extremely quenched samples. This article shows that when samples are extremely quenched, the counting efficiency in high-energy beta-ray nuclides depends principally on the Cherenkov effect. A new technique is described for quench determination, which makes the measurement of counting efficiency possible when scintillation counting approaches zero. A new efficiency computation model for pure beta-ray nuclides is also described. The results of the model are tested experimentally for 89 Sr, 90 Y, 36 Cl and 204 Tl nuclides with independence of the quench level. (orig.)

Increasing the Useful Life of Quench Reliefs with Inconel Bellows

Energy Technology Data Exchange (ETDEWEB)

Soyars, W. M. [Fermilab

1999-01-01

Reliable quench relief valves are an important part of superconducting magnet systems. Fermilab developed bellows-actuated cryogenic quench reliefs which have been in use since the early l 980's. The original design uses a stainless steel bellows. A high frequency, low amplitude vibration during relieving events has resulted in fatigue failures in the original design. To take advantage of the improved resistance to fatigue of Inconel, a nickel-chromium alloy, reliefs using Inconel 625 bellows were made. Design, development, and testing of the new version reliefs will be discussed. Tests show that relief valve lifetimes using Inconel bellows are more than five times greater than when using the original stainless steel bellows. Inconel bellows show great promise in increasing the lifetime of quench relief valves, and thus the reliability of accelerator cryogenic systems.

Bundle Security Protocol for ION

Science.gov (United States)

Burleigh, Scott C.; Birrane, Edward J.; Krupiarz, Christopher

2011-01-01

This software implements bundle authentication, conforming to the Delay-Tolerant Networking (DTN) Internet Draft on Bundle Security Protocol (BSP), for the Interplanetary Overlay Network (ION) implementation of DTN. This is the only implementation of BSP that is integrated with ION.

Connections on discrete fibre bundles

International Nuclear Information System (INIS)

Manton, N.S.; Cambridge Univ.

1987-01-01

A new approach to gauge fields on a discrete space-time is proposed, in which the fundamental object is a discrete version of a principal fibre bundle. If the bundle is twisted, the gauge fields are topologically non-trivial automatically. (orig.)

Fission product release measured during fuel damage tests at the Power Burst Facility

International Nuclear Information System (INIS)

Osetek, D.J.; Hartwell, J.K.; Vinjamuri, K.; Cronenberg, A.W.

1985-01-01

Results are presented of fission product release behavior observed during four severe fuel damage tests on bundles of UO 2 fuel rods. Transient temperatures up to fuel melting were obtained in the tests that included both rapid quench and slow cooldown, low and high (36 GWd/t) burnup fuel and the addition of Ag-In-Cd control rods. Release fractions of major fission product species and release rates of noble gas species are reported. Significant differences in release behavior are discussed between heatup and cooldown periods, low and high burnup fuel and long- and short-lived fission products. Explanations are offered for the probable reasons for the observed differences and recommendations for further studies are given

Thermal analysis methods for LMFBR wire wrapped bundles

International Nuclear Information System (INIS)

Todreas, N.E.

1976-11-01

A note is presented which was written to stimulate an awareness and discussion of the fundamental differences in the formulation of certain existing analysis codes for LMFBR wire wrap bundles. The contention of the note is that for those array types where data exists (one wire per pin, equal start angles), the ENERGY method results for coolant temperature under forced convection conditions provide benchmarks of reliability equal to the results of codes COBRA and TH1-3D

Deformation quantization of principal fibre bundles

International Nuclear Information System (INIS)

Weiss, S.

2007-01-01

Deformation quantization is an algebraic but still geometrical way to define noncommutative spacetimes. In order to investigate corresponding gauge theories on such spaces, the geometrical formulation in terms of principal fibre bundles yields the appropriate framework. In this talk I will explain what should be understood by a deformation quantization of principal fibre bundles and how associated vector bundles arise in this context. (author)

Development of inspection equipment for fuel bundles of CANDU-PHWR using R981 underwater radiation tolerant camera

Energy Technology Data Exchange (ETDEWEB)

Koo, Dae-Seo; Cho, Moon-Sung; Jo, Chang-Keun; Jun, Ji-Su; Jung, Jong Yeob; Park, Kwang-June; Suk, Ho-Chun

2005-03-15

The inspection equipment of fuel bundles was developed, which could perform visual inspection and dimensional measurement on fuel bundles of CANDU-PHWR, to evaluate, analyze the defective behavior of fuel bundles and inner surface of pressure tubes of inherent two-phase flow over 24kg/s in CANDU-6. The R981 radiation tolerant camera system with pan and tilt function was ordered and manufactured, which was waterproof, shielding radiation in underwater 10m in depth. The performance test, of the system ,due to camera-object distance was carried out in air/underwater atmosphere. The results of performance test of R981 radiation tolerant camera system are good. The inspection equipment of fuel bundles using R981 radiation tolerant camera system and underwater-radiation tolerant LVDT sensor(D5/200AW) was fabricated, which could perform visual inspection and dimensional measurement on fuel bundles of CANDU-PHWR with measurement accuracy 10{mu}m. This equipment will be utilizable integrity evaluation of fuel bundles which are irradiated in pressure tube of CANDU-PHWR.

Deformations of the generalised Picard bundle

International Nuclear Information System (INIS)

Biswas, I.; Brambila-Paz, L.; Newstead, P.E.

2004-08-01

Let X be a nonsingular algebraic curve of genus g ≥ 3, and let Mξ denote the moduli space of stable vector bundles of rank n ≥ 2 and degree d with fixed determinant ξ over X such that n and d are coprime. We assume that if g = 3 then n ≥ 4 and if g = 4 then n ≥ 3, and suppose further that n 0 , d 0 are integers such that n 0 ≥ 1 and nd 0 + n 0 d > nn 0 (2g - 2). Let E be a semistable vector bundle over X of rank n 0 and degree d 0 . The generalised Picard bundle W ξ (E) is by definition the vector bundle over M ξ defined by the direct image p M ξ *(U ξ x p X * E) where U ξ is a universal vector bundle over X x M ξ . We obtain an inversion formula allowing us to recover E from W ξ (E) and show that the space of infinitesimal deformations of W ξ (E) is isomorphic to H 1 (X, End(E)). This construction gives a locally complete family of vector bundles over M ξ parametrised by the moduli space M(n 0 ,d 0 ) of stable bundles of rank n 0 and degree d 0 over X. If (n 0 ,d 0 ) = 1 and W ξ (E) is stable for all E is an element of M(n 0 ,d 0 ), the construction determines an isomorphism from M(n 0 ,d 0 ) to a connected component M 0 of a moduli space of stable sheaves over M ξ . This applies in particular when n 0 = 1, in which case M 0 is isomorphic to the Jacobian J of X as a polarised variety. The paper as a whole is a generalisation of results of Kempf and Mukai on Picard bundles over J, and is also related to a paper of Tyurin on the geometry of moduli of vector bundles. (author)

New, Coupling Loss Induced, Quench Protection System for Superconducting Accelerator Magnets

CERN Document Server

Ravaioli, E; Giloux, C; Kirby, G; ten Kate, H H J; Verweij, A P

2014-01-01

Email Print Request Permissions Save to Project A new and promising method for the protection of superconducting high-field magnets is developed and tested on the so-called MQXC quadrupole magnet at the CERN magnet test facility. The method relies on a capacitive discharge system inducing, during a few periods, an oscillation of the transport current in the superconducting cable of the coil. The corresponding fast change of the local magnetic field introduces a high coupling-current loss, which, in turn, causes a fast quench of a large fraction of the coil due to enhanced temperature. Results of measured discharges at various levels of transport current are presented and compared to discharges by quenching the coils using conventional quench heaters and an energy extraction system. The hot-spot temperature in the quenching coil is deduced from the coil voltage and current. The results are compared to simulations carried out using a lumped-element dynamic electro-thermal model of the so-called MQX...

Stability of Picard bundle over moduli space of stable vector bundles ...

Indian Academy of Sciences (India)

Springer Verlag Heidelberg #4 2048 1996 Dec 15 10:16:45

Since the morphism ϕ is given by the universal property of the moduli space, the pullback of the universal bundle E on X × M to X × P by the map idX × ϕ is isomorphic (up to a twist by a line bundle coming from P) to ˜E. In other words, there is an integer k such that. 0 −→ (idX × ϕ)∗E −→ W ⊠ OP (k) −→ Ox×P (k + 1) −→ 0.

Fermilab R and D test facility for SSC magnets

International Nuclear Information System (INIS)

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

1989-01-01

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

A novel microbond bundle pullout technique to evaluate the ...

Indian Academy of Sciences (India)

... pullout technique to evaluate the interfacial properties of fibre-reinforced plastic composites ... https://www.ias.ac.in/article/fulltext/boms/040/04/0737-0744. Keywords. Microbond bundle pullout test; carbon/epoxy; fibre-reinforced composites; ...

Fluorescence quenching of uric acid solubilized in bicontinuous microemulsion by nitrobenzene

Directory of Open Access Journals (Sweden)

Maurice O. Iwunze

2013-02-01

Full Text Available Abstract: Uric Acid is known to be practically insoluble in aqueous and alcoholic media. However, it exhibits a reasonable solubility in a Bicontinuous Microemulsion system – a 15-fold or more increase in solubility in this system compared to its solubility in water. The bicontinuous microemulsion is made up of three components –Dodecane-Surfactant-water. Uric acid solubilized in this system is quenched by nitrobenzene. The obtained fluorescence data do not obey the Stern-Volmer equation when plotted accordingly. Therefore, the modified Stern-Volmer equation was used to analyze the data. It was observed that only one third (1/3 of uric acid is accessible to quenching in this medium and the reaction is diffusion-limited. The Stern-Volmer quenching constant, KSV, was calculated to be 130 M-1 and the fluorescence lifetime, 0, the quantum yield,, and the bimolecular quenching rate constant, kq, were calculated as 10.6 nanoseconds, 0.06 and 1.231010 M-1s-1, respectively.

CLIQ – Coupling-Loss Induced Quench System for Protecting Superconducting Magnets

CERN Multimedia

Ravaioli, E; Kirby, G; ten Kate, H H J; Verweij, A P

2014-01-01

The recently developed Coupling-Loss-Induced Quench (CLIQ) protection system is a new method for initiating a fast and voluminous transition to the normal state for protecting high energy density superconducting magnets. Upon quench detection, CLIQ is triggered to generate an oscillating current in the magnet coil by means of a capacitive discharge. This in turn introduces a high coupling loss in the superconductor which provokes a quick transition to the normal state of the coil windings. The system is now implemented for the protection of a two meter long superconducting quadrupole magnet and characterized in the CERN magnet test facility. Various CLIQ configurations with different current injection points are tested and the results compared to similar transients lately measured with a not optimized configuration. Test results convincingly show that the newly tested design allows for a more global quench initiation and thus a faster discharge of the magnet energy. Moreover, the performance of CLIQ for reduc...

Study of fuel bundle geometry on inter subchannel flow in a 19 pin wire wrapped bundle

International Nuclear Information System (INIS)

Naveen Raj, M.; Velusamy, D.K.

2015-01-01

In typical sodium cooled fast reactor (SFR) fuel pin bundle, gap between the pins is maintained by helically wound wire wrap around each pin. The presence of wire induces large inter-subchannel transverse flow, eventually promoting mixing and heat transfer. The magnitude of the transverse flow is highly dependent on the various pin-bundle dimensions. Appropriate modeling of these transverse flows in subchannel codes is necessary to predict realistic temperature distribution in pin bundle. Hence, detailed parametric study of transverse flow on pin-bundle geometric parameters has been conducted. The parameters taken for the present study are pin diameter, wire diameter, helical wire pitch and edge gap. Towards this 3-D computational fluid dynamic analysis on a structured mesh of 19 pin bundle is carried out using k-epsilon turbulence model. Periodic oscillations along the primacy flow direction were found in subchannel transverse flow and peripheral pin clad temperatures with periodicity over one pitch length. Based on parametric studies, correlations for transverse flow in central subchannels are proposed. (author)

The Role of Quench-back in the Passive Quench Protection of Long Solenoids with Coil Sub-division

International Nuclear Information System (INIS)

Green, Michael A.; Guo, XingLong; Wang, Li; Pan, Heng; Wu, Hong

2009-01-01

This paper describes how a passive quench protection system can be applied to long superconducting solenoid magnets. When a solenoid coil is long compared to its thickness, the magnet quench process will be dominated by the time needed for uench propagation along the magnet length. Quench-back will permit a long magnet to quench more rapidly in a passive way. Quenchback from a conductive (low resistivity) mandrel is essential for spreading the quench along the length of a magnet. The andrel must be inductively coupled to the magnet circuit that is being quenched. Current induced in the mandrel by di/dt in the magnet produces heat in the mandrel, which in turn causes the superconducting coil wound on the mandrel to quench. Sub-divisions often employed to reduce the voltages to ground within the coil. This paper explores when it is possible for quench-back to be employed for passive quench protection. The role of sub-division of the coil is discussed for long magnets.

Quench Protection Studies of 11T Nb$_3$Sn Dipole Models for LHC Upgrades

Energy Technology Data Exchange (ETDEWEB)

Zlobin, Alexander [Fermilab; Chlachidze, Guram [Fermilab; Nobrega, Alfred [Fermilab; Novitski, Igor [Fermilab; Karppinen, Mikko [CERN

2014-07-01

CERN and FNAL are developing 11 T Nb3Sn dipole magnets for the LHC collimation system upgrade. Due to the large stored energy, protection of these magnets during a quench is a challenging problem. This paper reports the results of experimental studies of key quench protection parameters including longitudinal and radial quench propagation in the coil, coil heating due to a quench, and energy extraction and quench-back effect. The studies were performed using a 1 m long 11 T Nb3Sn dipole coil tested in a magnetic mirror configuration.

Characterization of oil based nanofluid for quench medium

Science.gov (United States)

Mahiswara, E. P.; Harjanto, S.; Putra, W. N.; Ramahdita, G.; Yahya, S. S.; Kresnodrianto

2018-01-01

The choice of quench medium depends on the hardenability of the metal alloy, the thickness of the component, and the geometry of the component. Some of these will determine the cooling rate required to obtain the desired microstructure and material properties. Improper quench media will cause the material to become brittle, suffers from geometric distortion, or having a high undesirable residual stresses in the components. In heat treatment industries, oil and water are frequently used as the quench media. Recently, nanofluid as a quench medium has also been studied using several different fluids as the solvent. Examples of frequently used solvents include polymers, vegetable oils, and mineral oil. In this research, laboratory-grade carbon powder were used as nanoparticle. Oil was used as the fluid base in this research as the main observation focus. To obtain nanoscale carbon particles, planetary ball mill was used to ground laboratory grade carbon powder to decrease the particle size. This method was used to lower the cost for nanoparticle synthesis. Milling speed and duration were set at 500 rpm and 15 hours. Field Emission Scanning Electron Microscope (FE-SEM), and Energy Dispersive X-Ray (EDX) measurement were carried out to determine the particle size, material identification, particle morphology, and surface change of samples. The carbon nanoparticle content in nanofluid quench mediums for this research were varied at 0.1%, 0.2%, 0.3%, 0.4, and 0.5 % volume. Furthermore, these mediums were used to quench JIS S45C or AISI 1045 carbon steel samples which annealed at 1000°C. Hardness testing and metallography observation were then conducted to further examine the effect of different quench medium in steel samples.

Two-phase flow and cross-mixing measurements in a rod bundle

International Nuclear Information System (INIS)

Yloenen, A.; Prasser, H.-M.

2011-01-01

The wire-mesh sensor technique has been used for the first time to study two-phase flow and liquid mixing in a rod bundle. A dedicated test facility (SUBFLOW) was constructed at Paul Scherrer Institut (PSI) in a co-operation with the Swiss Federal Institute of Technology (ETH Zurich). Simultaneous injection of salt water as tracer and air bubbles can be used to quantify the enhancement of liquid mixing in two-phase flow when the results are compared with the single-phase mixing experiment with the same test parameters. The second aspect in the current experiments is the two-phase flow in bundle geometry. (author)

Cecil gives in-bundle access for inspection and lancing [steam generators

International Nuclear Information System (INIS)

Trovato, S.A.; Ruggieri, S.K.

1989-01-01

Cecil (Consolidated Edison Combined Inspection and Lancing System) is a robotic device which makes it possible to take inspection and sludge lancing equipment deep inside steam generator tube bundles. Cecil is teleoperated to perform tube bundle inspections, sludge sampling and sludge lancing. The first field test of Cecil at Indian Point 2 reactor, successfully demonstrated its capability for high quality inspection, and its potential for improved sludge removal, both with reduced personnel radiation exposure. (U.K.)

Effects of quenching and partial quenching on QCD penguin matrix elements

NARCIS (Netherlands)

Golterman, Maarten; Pallante, Elisabetta

2002-01-01

We point out that chiral transformation properties of penguin operators change in the transition from unquenched to (partially) quenched QCD. The way in which this affects the lattice determination of weak matrix elements can be understood in the framework of (partially) quenched chiral perturbation

In-pool damaged fuel bundle recovery

International Nuclear Information System (INIS)

Piascik, T.G.; Patenaude, R.S.

1988-01-01

While preparing to rerack the Oyster Creek Nuclear Generating Station, GPU Nuclear had need to move a damaged fuel bundle. This bundle had no upper tie plate and could not be moved in the normal manner. GPU Nuclear formed a small, dedicated project team to disassemble, package and move this damaged bundle. The team was composed of key personnel from GPU Nuclear Fuels Projects, OCNGS Operations and Proto-Power / Bisco, a specialty contractor who has fuel bundle reconstitution and rod consolidation experience, remote tooling, underwater video systems and experienced technicians. Proven tooling, clear procedures and a simple approach were important, but the key element was the spirit of teamwork and leadership exhibited by the people involved

Boiling on a tube bundle: heat transfer, pressure drop and flow patterns

International Nuclear Information System (INIS)

Royen Van, E.

2011-11-01

The complexity of two-phase flow boiling on a tube bundle presents many challenges to the understanding of the physical phenomena taking place. It is important to quantify these numerous heat flow mechanisms in order to better describe the performance of tube bundles as a function of the operational conditions. In the present study, the bundle boiling facility at the Laboratory of Heat and Mass Transfer (LTCM) was modified to obtain high-speed videos to characterise the two-phase regimes and some bubble dynamics of the boiling process. It was then used to measure heat transfer on single tubes and in bundle boiling conditions. Pressure drop measurements were also made during adiabatic and diabatic bundle conditions. New enhanced boiling tubes from Wolverine Tube Inc. (Turbo-B5) and the Wieland-Werke AG (Gewa-B5) were investigated using R134a and R236fa as test fluids. The tests were carried out at saturation temperatures T sat of 5 °C and 15 °C, mass flow rates from 4 to 35 kg/m 2 s and heat fluxes from 15 to 70 kW/m 2 , typical of actual operating conditions. The flow pattern investigation was conducted using visual observations from a borescope inserted in the middle of the bundle. Measurements of the light attenuation of a laser beam through the intertube two-phase flow and local pressure fluctuations with piezo-electric pressure transducers were also taken to further help in characterising the complex flow. Pressure drop measurements and data reduction procedures were revised and used to develop new, improved frictional pressure drop prediction methods for adiabatic and diabatic two-phase conditions. The physical phenomena governing the enhanced tube evaporation process and their effects on the performance of tube bundles were investigated and insight gained. A new method based on a theoretical analysis of thin film evaporation was used to propose a new correlating parameter. A large new database of local heat transfer coefficients were obtained and then

In-pile post-DNB behavior of a nine-rod PWR-type fuel bundle

International Nuclear Information System (INIS)

Gunnerson, F.S.; MacDonald, P.E.

1980-01-01

The results of an in-pile power-cooling-mismatch (PCM) test designed to investigate the behavior of a nine-rod, PWR-type fuel bundle under intermittent and sustained periods of high temperature film boiling operation are presented. Primary emphasis is placed on the DNB and post-DNB events including rod-to-rod interactions, return to nucleate boiling (RNB), and fuel rod failure. A comparison of the DNB behavior of the individual bundle rods with single-rod data obtained from previous PCM tests is also made

Anatomic Double-bundle ACL Reconstruction

NARCIS (Netherlands)

Schreiber, Verena M.; van Eck, Carola F.; Fu, Freddie H.

2010-01-01

Rupture of the anterior cruciate ligament (ACL) is one of the most frequent forms of knee trauma. The traditional surgical treatment for ACL rupture is single-bundle reconstruction. However, during the past few years there has been a shift in interest toward double-bundle reconstruction to closely

Output commitment through product bundling : Experimental evidence

NARCIS (Netherlands)

Hinloopen, Jeroen; Mueller, Wieland; Normann, Hans-Theo

We analyze the impact of product bundling in experimental markets. One firm has monopoly power in a first market but competes with another firm la Cournot in a second market. We compare treatments where the multi-product firm (i) always bundles, (ii) never bundles, and (iii) chooses whether to

ACM Bundles on Del Pezzo surfaces

Directory of Open Access Journals (Sweden)

Joan Pons-Llopis

2009-11-01

Full Text Available ACM rank 1 bundles on del Pezzo surfaces are classified in terms of the rational normal curves that they contain. A complete list of ACM line bundles is provided. Moreover, for any del Pezzo surface X of degree less or equal than six and for any n ≥ 2 we construct a family of dimension ≥ n − 1 of non-isomorphic simple ACM bundles of rank n on X.

Bundling and mergers in energy markets

International Nuclear Information System (INIS)

Granier, Laurent; Podesta, Marion

2010-01-01

Does bundling trigger mergers in energy industries? We observe mergers between firms belonging to various energy markets, for instance between gas and electricity providers. These mergers enable firms to bundle. We consider two horizontally differentiated markets. In this framework, we show that bundling strategies in energy markets create incentives to form multi-market firms in order to supply bi-energy packages. Moreover, we find that this type of merger is detrimental to social welfare. (author)

The possibility of tribopair lifetime extending by welding of quenched and tempered stainless steel with quenched and tempered carbon steel

Directory of Open Access Journals (Sweden)

V. Marušić

2015-04-01

Full Text Available In the conditions of tribocorrosion wear, extending of parts lifetime could be achieved by using stainless steel,which is hardened to sufficiently high hardness. In the tribosystem bolt/ bushing shell/link plate of the bucket elevator transporter conveyor machine, the previously quenched and tempered martensitic stainless steel for bolts is hardened at ≈47 HRC and welded with the quenched and tempered high yield carbon steel for bolts. Additional material, based on Cr-Ni-Mo (18/8/6 is used. The microstructure and hardness of welded samples are tested. On the tensile tester, resistance of the welded joint is tested with a simulated experiment. Dimensional control of worn tribosystem elements was performed after six months of service.

Hot compressive deformation behavior of the as-quenched A357 aluminum alloy

International Nuclear Information System (INIS)

Yang, X.W.; Lai, Z.H.; Zhu, J.C.; Liu, Y.; He, D.

2012-01-01

Highlights: ► We create a thermal history curve which was applied to carry out compression tests. ► We make an analysis of deformation performance for as-quenched A357 alloy. ► We create a constitutive equation which has good accuracy. - Abstract: The objective of the present work was to establish an accurate thermal-stress mathematical model of the quenching operation for A357 (Al–7Si–0.6Mg) alloy and to investigate the deformation behavior of this alloy. Isothermal compression tests of as-quenched A357 alloy were performed in the temperature range of 350–500 °C and at the strain rate range of 0.001–1 s −1 . Experimental results show that the flow stress of as-quenched A357 alloy decreases with the increase of temperature and the decrease of strain rate. Based on the hyperbolic sine equation, a constitutive equation is a relation between 0.2 pct yield stress and deformation conditions (strain rate and deformation temperature) was established. The corresponding hot deformation activation energy (Q) for as-quenched A357 alloy is 252.095 kJ/mol. Under the different small strains (≤0.01), the constitutive equation parameters of as-quenched A357 alloy were calculated. Values of flow stress calculated by constitutive equation were in a very good agreement with experimental results. Therefore, it can be used as an accurate thermal-stress model to solve the problems of quench distortion of parts.

Quench propagation and quench detection in the TF system of JT-60SA

International Nuclear Information System (INIS)

Lacroix, Benoit; Duchateau, Jean-Luc; Meuris, Chantal; Ciazynski, Daniel; Nicollet, Sylvie; Zani, Louis; Polli, Gian-Mario

2013-01-01

Highlights: • The JT-60SA primary quench detection system will be based on voltage measurements. • The early quench propagation was studied in the JT-60SA TF conductor. • The impact of the conductor jacket on the hot spot criterion was quantified. • The detection parameters were investigated for different quench initiations. -- Abstract: In the framework of the JT-60SA project, France and Italy will provide to JAEA 18 Toroidal Field (TF) coils including NbTi cable-in-conduit conductors. During the tokamak operation, these coils could experience a quench, an incidental event corresponding to the irreversible transition from superconducting state to normal resistive state. Starting from a localized disturbance, the normal zone propagates along the conductor and dissipates a large energy due to Joule heating, which can cause irreversible damages. The detection has to be fast enough (a few seconds) to trigger the current discharge, so as to dump the stored magnetic energy into an external resistor. The JT-60SA primary quench detection system will be based on voltage measurements, which are the most rapid technology. The features of the detection system must be adjusted so as to detect the most probable quenches, while avoiding inopportune fast safety discharges. This requires a reliable simulation of the early quench propagation, performed in this study with the Gandalf code. The conductor temperature reached during the current discharge must be kept under a maximal value, according to the hot spot criterion. In the present study, a hot spot criterion temperature of 150 K was taken into account and the role of each conductor component (strands, helium and conduit) was analyzed. The detection parameters were then investigated for different hypotheses regarding the quench initiation

Computational imaging through a fiber-optic bundle

Science.gov (United States)

Lodhi, Muhammad A.; Dumas, John Paul; Pierce, Mark C.; Bajwa, Waheed U.

2017-05-01

Compressive sensing (CS) has proven to be a viable method for reconstructing high-resolution signals using low-resolution measurements. Integrating CS principles into an optical system allows for higher-resolution imaging using lower-resolution sensor arrays. In contrast to prior works on CS-based imaging, our focus in this paper is on imaging through fiber-optic bundles, in which manufacturing constraints limit individual fiber spacing to around 2 μm. This limitation essentially renders fiber-optic bundles as low-resolution sensors with relatively few resolvable points per unit area. These fiber bundles are often used in minimally invasive medical instruments for viewing tissue at macro and microscopic levels. While the compact nature and flexibility of fiber bundles allow for excellent tissue access in-vivo, imaging through fiber bundles does not provide the fine details of tissue features that is demanded in some medical situations. Our hypothesis is that adapting existing CS principles to fiber bundle-based optical systems will overcome the resolution limitation inherent in fiber-bundle imaging. In a previous paper we examined the practical challenges involved in implementing a highly parallel version of the single-pixel camera while focusing on synthetic objects. This paper extends the same architecture for fiber-bundle imaging under incoherent illumination and addresses some practical issues associated with imaging physical objects. Additionally, we model the optical non-idealities in the system to get lower modelling errors.

A comparison of core degradation phenomena in the CORA, QUENCH, Phébus SFD and Phébus FP experiments

Energy Technology Data Exchange (ETDEWEB)

Haste, T., E-mail: tim.haste@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire, IRSN, BP 3, F-13115 St. Paul-lez-Durance Cedex (France); Steinbrück, M., E-mail: martin.steinbrueck@kit.edu [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Barrachin, M., E-mail: marc.barrachin@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire, IRSN, BP 3, F-13115 St. Paul-lez-Durance Cedex (France); Luze, O. de, E-mail: olivier.de-luze@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire, IRSN, BP 3, F-13115 St. Paul-lez-Durance Cedex (France); Grosse, M., E-mail: mirco.grosse@kit.edu [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Stuckert, J., E-mail: juri.stuckert@kit.edu [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany)

2015-03-15

Highlights: • The results of the experiments CORA, QUENCH and Phébus SFD/FP are summarised. • All phenomena expected up to melt movement to the lower head are shown consistently. • Separate-effect tests performed at KIT and IRSN aid improve their modelling. • Data from the integral tests help independent validation of new and improved models. • The improved codes will help reduce uncertainties in safety-critical areas for core degradation. - Abstract: Over the past 20 years, integral fuel bundle experiments performed at IRSN Cadarache, France (Phébus-SFD and Phébus FP – fission heated) and at Karlsruhe Institute of Technology, Germany (CORA and QUENCH – electrically heated), accompanied by separate-effect tests, have provided a wealth of detailed information on core degradation phenomena that occur under severe accident conditions, relevant to such safety issues as in-vessel retention of the core, recovery of the core by water reflood, hydrogen generation and fission product release. These data form an important basis for development and validation of severe accident analysis codes such as ASTEC (IRSN/GRS, EC) and MELCOR (USNRC/SNL, USA) that are used to assess the safety of current and future reactor designs, so helping to reduce the uncertainty associated with such code predictions. Following the recent end of the Phébus FP project, it is appropriate now to compare the core degradation phenomena observed in these four major experimental series, indicating the main conclusions that have been drawn. This covers subjects such as early phase degradation up to loss of rod-like geometry (all the series), late phase degradation and the link between fission product release and core degradation (Phébus FP), oxidation phenomena (all the series), reflood behaviour (CORA and QUENCH), as well as particular topics such as the effects of control rod material and fuel burn-up on core degradation. It also outlines the separate-effects experiments performed to

MODULAR BUNDLE ADJUSTMENT FOR PHOTOGRAMMETRIC COMPUTATIONS

Directory of Open Access Journals (Sweden)

N. Börlin

2018-05-01

Full Text Available In this paper we investigate how the residuals in bundle adjustment can be split into a composition of simple functions. According to the chain rule, the Jacobian (linearisation of the residual can be formed as a product of the Jacobians of the individual steps. When implemented, this enables a modularisation of the computation of the bundle adjustment residuals and Jacobians where each component has limited responsibility. This enables simple replacement of components to e.g. implement different projection or rotation models by exchanging a module. The technique has previously been used to implement bundle adjustment in the open-source package DBAT (Börlin and Grussenmeyer, 2013 based on the Photogrammetric and Computer Vision interpretations of Brown (1971 lens distortion model. In this paper, we applied the technique to investigate how affine distortions can be used to model the projection of a tilt-shift lens. Two extended distortion models were implemented to test the hypothesis that the ordering of the affine and lens distortion steps can be changed to reduce the size of the residuals of a tilt-shift lens calibration. Results on synthetic data confirm that the ordering of the affine and lens distortion steps matter and is detectable by DBAT. However, when applied to a real camera calibration data set of a tilt-shift lens, no difference between the extended models was seen. This suggests that the tested hypothesis is false and that other effects need to be modelled to better explain the projection. The relatively low implementation effort that was needed to generate the models suggest that the technique can be used to investigate other novel projection models in photogrammetry, including modelling changes in the 3D geometry to better understand the tilt-shift lens.

Modular Bundle Adjustment for Photogrammetric Computations

Science.gov (United States)

Börlin, N.; Murtiyoso, A.; Grussenmeyer, P.; Menna, F.; Nocerino, E.

2018-05-01

In this paper we investigate how the residuals in bundle adjustment can be split into a composition of simple functions. According to the chain rule, the Jacobian (linearisation) of the residual can be formed as a product of the Jacobians of the individual steps. When implemented, this enables a modularisation of the computation of the bundle adjustment residuals and Jacobians where each component has limited responsibility. This enables simple replacement of components to e.g. implement different projection or rotation models by exchanging a module. The technique has previously been used to implement bundle adjustment in the open-source package DBAT (Börlin and Grussenmeyer, 2013) based on the Photogrammetric and Computer Vision interpretations of Brown (1971) lens distortion model. In this paper, we applied the technique to investigate how affine distortions can be used to model the projection of a tilt-shift lens. Two extended distortion models were implemented to test the hypothesis that the ordering of the affine and lens distortion steps can be changed to reduce the size of the residuals of a tilt-shift lens calibration. Results on synthetic data confirm that the ordering of the affine and lens distortion steps matter and is detectable by DBAT. However, when applied to a real camera calibration data set of a tilt-shift lens, no difference between the extended models was seen. This suggests that the tested hypothesis is false and that other effects need to be modelled to better explain the projection. The relatively low implementation effort that was needed to generate the models suggest that the technique can be used to investigate other novel projection models in photogrammetry, including modelling changes in the 3D geometry to better understand the tilt-shift lens.

Standard-model bundles

CERN Document Server

Donagi, Ron; Pantev, Tony; Waldram, Dan; Donagi, Ron; Ovrut, Burt; Pantev, Tony; Waldram, Dan

2002-01-01

We describe a family of genus one fibered Calabi-Yau threefolds with fundamental group ${\\mathbb Z}/2$. On each Calabi-Yau $Z$ in the family we exhibit a positive dimensional family of Mumford stable bundles whose symmetry group is the Standard Model group $SU(3)\\times SU(2)\\times U(1)$ and which have $c_{3} = 6$. We also show that for each bundle $V$ in our family, $c_{2}(Z) - c_{2}(V)$ is the class of an effective curve on $Z$. These conditions ensure that $Z$ and $V$ can be used for a phenomenologically relevant compactification of Heterotic M-theory.

Reduction of symplectic principal R-bundles

International Nuclear Information System (INIS)

Lacirasella, Ignazio; Marrero, Juan Carlos; Padrón, Edith

2012-01-01

We describe a reduction process for symplectic principal R-bundles in the presence of a momentum map. These types of structures play an important role in the geometric formulation of non-autonomous Hamiltonian systems. We apply this procedure to the standard symplectic principal R-bundle associated with a fibration π:M→R. Moreover, we show a reduction process for non-autonomous Hamiltonian systems on symplectic principal R-bundles. We apply these reduction processes to several examples. (paper)

Bundles of C*-categories and duality

OpenAIRE

Vasselli, Ezio

2005-01-01

We introduce the notions of multiplier C*-category and continuous bundle of C*-categories, as the categorical analogues of the corresponding C*-algebraic notions. Every symmetric tensor C*-category with conjugates is a continuous bundle of C*-categories, with base space the spectrum of the C*-algebra associated with the identity object. We classify tensor C*-categories with fibre the dual of a compact Lie group in terms of suitable principal bundles. This also provides a classification for ce...

Evaluating big deal journal bundles.

Science.gov (United States)

Bergstrom, Theodore C; Courant, Paul N; McAfee, R Preston; Williams, Michael A

2014-07-01

Large commercial publishers sell bundled online subscriptions to their entire list of academic journals at prices significantly lower than the sum of their á la carte prices. Bundle prices differ drastically between institutions, but they are not publicly posted. The data that we have collected enable us to compare the bundle prices charged by commercial publishers with those of nonprofit societies and to examine the types of price discrimination practiced by commercial and nonprofit journal publishers. This information is of interest to economists who study monopolist pricing, librarians interested in making efficient use of library budgets, and scholars who are interested in the availability of the work that they publish.

Dimensional measurement of fresh fuel bundle for CANDU reactor

International Nuclear Information System (INIS)

Jo, Chang Keun; Cho, Moon Sung; Suk, Ho Chun; Koo, Dae Seo; Jun, Ji Su; Jung, Jong Yeob

2005-01-01

This report describes the results of the dimensional measurement of fresh fuel bundles for the CANDU reactor in order to estimate the integrity of fuel bundle in two-phase flow in the CANDU-6 fuel channel. The dimensional measurements of fuel bundles are performed by using the 'CANDU Fuel In-Bay Inspection and Dimensional Measurement System', which was developed by this project. The dimensional measurements are done from February 2004 to March 2004 in the CANDU fuel storage of KNFC for the 36 fresh fuel bundles, which are produced by KNFC and are waiting for the delivery to the Wolsong-3 plant. The detail items of dimensional measurements are included fuel rod and bearing pad profiles of the outer ring in fuel bundle, diameter of fuel bundle, bowing of fuel bundle, fuel rod length, and surface profile of end plate profile. The measurement data will be compared with those of the post-irradiated bundles cooled in Wolsong-3 NPP spent fuel pool by using the same bundles and In-Bay Measurement System. So, this analysis of data will be applied for the evaluation of fuel bundle integrity in two-phase flow of the CANDU-6 fuel channel

Sasakian and Parabolic Higgs Bundles

Science.gov (United States)

Biswas, Indranil; Mj, Mahan

2018-03-01

Let M be a quasi-regular compact connected Sasakian manifold, and let N = M/ S 1 be the base projective variety. We establish an equivalence between the class of Sasakian G-Higgs bundles over M and the class of parabolic (or equivalently, ramified) G-Higgs bundles over the base N.

Fuel bundle movement due to reverse flow

Energy Technology Data Exchange (ETDEWEB)

Wahba, N N; Akalin, O [Ontario Hydro, Toronto, ON (Canada)

1996-12-31

When a break occurs in the inlet feeder or inlet header, the rapid depressurization will cause the channel flow to reverse forcing the string of bundles to accelerate and impact with upstream shield plug. A model has been developed to predict the bundle motion due to the channel flow reversal. The model accounts for various forces acting on the bundle. A series of five reverse flow, bundle acceleration experiments have been conducted simulating a break in the inlet feeder of a CANDU fuel channel. The model has been validated against the experiments. The predicted impact velocities are in good agreement with the measured values. It is demonstrated that the model may be successfully used in predicting bundle relocation timing following a large LOCA (loss of coolant). (author). 7 refs., 3 tabs., 11 figs.

Bundles over Quantum RealWeighted Projective Spaces

Directory of Open Access Journals (Sweden)

Tomasz Brzeziński

2012-09-01

Full Text Available The algebraic approach to bundles in non-commutative geometry and the definition of quantum real weighted projective spaces are reviewed. Principal U(1-bundles over quantum real weighted projective spaces are constructed. As the spaces in question fall into two separate classes, the negative or odd class that generalises quantum real projective planes and the positive or even class that generalises the quantum disc, so do the constructed principal bundles. In the negative case the principal bundle is proven to be non-trivial and associated projective modules are described. In the positive case the principal bundles turn out to be trivial, and so all the associated modules are free. It is also shown that the circle (coactions on the quantum Seifert manifold that define quantum real weighted projective spaces are almost free.

Temperature Profiles During Quenches in LHC Superconducting Dipole Magnets Protected by Quench Heaters

OpenAIRE

Maroussov, V; Sanfilippo, S; Siemko, A

1999-01-01

The efficiency of the magnet protection by quench heaters was studied using a novel method which derives the temperature profile in a superconducting magnet during a quench from measured voltage signals. In several Large Hadron Collider single aperture dipole models, temperature profiles and temperature gradients in the magnet coil have been evaluated in the case of protection by different sets of quench heaters and different powering and protection parameters. The influence of the insulation...

Quench detection and behaviour in case of quench in the ITER magnet systems

International Nuclear Information System (INIS)

Coatanea-Gouachet, M.

2012-02-01

The quench of one of the ITER magnet system is an irreversible transition from superconducting to normal resistive state, of a conductor. This normal zone propagates along the cable in conduit conductor dissipating a large power. The detection has to be fast enough to dump out the magnetic energy and avoid irreversible damage of the systems. The primary quench detection in ITER is based on voltage detection, which is the most rapid detection. The very magnetically disturbed environment during the plasma scenario makes the voltage detection particularly difficult, inducing large inductive components in the coils and voltage compensations have to be designed to discriminate the resistive voltage associated with the quench. A conceptual design of the quench detection based on voltage measurements is proposed for the three majors magnet systems of ITER. For this, a clear methodology was developed. It includes the classical hot spot criterion, the quench propagation study using the commercial code Gandalf and the careful estimation of the inductive disturbances by developing the TrapsAV code. Specific solutions have been proposed for the compensation in the three ITER magnet systems and for the quench detection parameters, which are the voltage threshold (in the range of 0.1 V - 0.55 V) and the holding time (in the range of 1-1.4 s). The selected values, in particular the holding time, are sufficiently high to ensure the reliability of the system and avoid fast safety discharges not induced by a quench, which is a classical problem. (author)

Deciphering jet quenching with JEWEL

CERN Multimedia

CERN. Geneva

2018-01-01

In heavy ion collisions jets arising from the fragmentation of hard quarks and gluons experience strong modifications due to final state re-scattering. This so-called jet quenching is related to the emergence of collectivity and equilibration in QCD. I will give an introduction to jet quenching and its modeling in JEWEL, a Monte Carlo implementation of a dynamical model for jet quenching. I will then discuss examples highlighting how JEWEL can be used to elucidate the physical mechanisms relevant for jet quenching.  

Characterization of water based nanofluid for quench medium

Science.gov (United States)

Kresnodrianto; Harjanto, S.; Putra, W. N.; Ramahdita, G.; Yahya, S. S.; Mahiswara, E. P.

2018-04-01

Quenching has been a valuable method in steel hardening method especially in industrial scale. The hardenability of the metal alloys, the thickness of the component, and the geometry is some factors that can affect the choice of quench medium. Improper quench media can cause the material to become too brittle, suffers some geometric distortion, and undesirable residual stress that will cause some effect on the mechanical property and fracture mechanism of a component. Recently, nanofluid as a quench medium has been used for better quenching performance and has been studied using several different fluids and nanoparticles. Some of frequently used solvents include polymers, vegetable oils, and mineral oil, and nanoparticles frequently used include CuO, ZnO, and Alumina. In this research, laboratory-grade carbon powder were used as nanoparticle. Water was used as the fluid base in this research as the main observation focus. Carbon particles were obtain using a top-down method, whereas planetary ball mill was used to ground laboratory grade carbon powder to decrease the particle size. Milling speed and duration were set at 500 rpm and 15 hours. Field Emission Scanning Electron Microscope (FE-SEM), and Energy Dispersive X-Ray (EDX) measurement were carried out to determine the particle size, material identification, particle morphology, and surface change of samples. Nanofluid was created by mixing percentage of carbon nanoparticles with water using ultrasonic vibration for 280s. The carbon nanoparticle content in nanofluid quench mediums for this research were varied at 0.1%, 0.2%, 0.3%, 0.4, and 0.5 % volume. Furthermore, these mediums were used to quench JIS S45C or AISI 1045 carbon steel samples which austenized at 1000°C. Hardness testing and metallography observation were then conducted to further check the effect of different quench medium in steel samples. Preliminary characterizations showed that carbon particles dimension after milling was still in sub

The Preliminary Study for Numerical Computation of 37 Rod Bundle in CANDU Reactor

International Nuclear Information System (INIS)

Jeon, Yu Mi; Park, Joo Hwan

2010-09-01

A typical CANDU 6 fuel bundle consists of 37 fuel rods supported by two endplates and separated by spacer pads at various locations. In addition, the bearing pads are brazed to each outer fuel rod with the aim of reducing the contact area between the fuel bundle and the pressure tube. Although the recent progress of CFD methods has provided opportunities for computing the thermal-hydraulic phenomena inside of a fuel channel, it is yet impossible to reflect numerical computations on the detailed shape of rod bundle due to challenges with computing mesh and memory capacity. Hence, the previous studies conducted a numerical computation for smooth channels without considering spacers and bearing pads. But, it is well known that these components are an important factor to predict the pressure drop and heat transfer rate in a channel. In this study, the new computational method is proposed to solve complex geometry such as a fuel rod bundle. Before applying a solution to the problem of the 37 rod bundle, the validity and the accuracy of the method are tested by applying the method to simple geometry. The split channel method has been proposed with the aim of computing the fully shaped CANDU fuel channel with detailed components. The validity was tested by applying the method to the single channel problem. The average temperature have similar values for the considered two methods, while the local temperature shows a slight difference by the effect of conduction heat transfer in the solid region of a rod. Based on the present result, the calculation for the fully shaped 37-rod bundle is scheduled for future work

Process evaluation of a cluster-randomised trial testing a pressure ulcer prevention care bundle: a mixed-methods study.

Science.gov (United States)

Roberts, Shelley; McInnes, Elizabeth; Bucknall, Tracey; Wallis, Marianne; Banks, Merrilyn; Chaboyer, Wendy

2017-02-13

As pressure ulcers contribute to significant patient burden and increased health care costs, their prevention is a clinical priority. Our team developed and tested a complex intervention, a pressure ulcer prevention care bundle promoting patient participation in care, in a cluster-randomised trial. The UK Medical Research Council recommends process evaluation of complex interventions to provide insight into why they work or fail and how they might be improved. This study aimed to evaluate processes underpinning implementation of the intervention and explore end-users' perceptions of it, in order to give a deeper understanding of its effects. A pre-specified, mixed-methods process evaluation was conducted as an adjunct to the main trial, guided by a framework for process evaluation of cluster-randomised trials. Data was collected across eight Australian hospitals but mainly focused on the four intervention hospitals. Quantitative and qualitative data were collected across the evaluation domains: recruitment, reach, intervention delivery and response to intervention, at both cluster and individual patient level. Quantitative data were analysed using descriptive and inferential statistics. Qualitative data were analysed using thematic analysis. In the context of the main trial, which found a 42% reduction in risk of pressure ulcer with the intervention that was not significant after adjusting for clustering and covariates, this process evaluation provides important insights. Recruitment and reach among clusters and individuals was high, indicating that patients, nurses and hospitals are willing to engage with a pressure ulcer prevention care bundle. Of 799 intervention patients in the trial, 96.7% received the intervention, which took under 10 min to deliver. Patients and nurses accepted the care bundle, recognising benefits to it and describing how it enabled participation in pressure ulcer prevention (PUP) care. This process evaluation found no major failures

Fuel temperature characteristics of the 37-element and CANFLEX fuel bundle

International Nuclear Information System (INIS)

Bae, Jun Ho; Rho, Gyu Hong; Park, Joo Hwan

2009-10-01

This report describes the fuel temperature characteristics of CANFLEX fuel bundles and 37-element fuel bundles for a different burnup of fuel. The program was consisted for seeking the fuel temperature of fuel bundles of CANFLEX fuel bundles and 37-element fuel bundles by using the method in NUCIRC. Fuel temperature has an increasing pattern with the burnup of fuel for CANFLEX fuel bundles and 37-element fuel bundles. For all the case of burnup, the fuel temperature of CANFLEX fuel bundles has a lower value than that of 37-element fuel bundles. Especially, for the high power channel, the CANFLEX fuel bundles show a lower fuel temperature as much as about 75 degree, and the core averaged fuel temperature has a lower fuel temperature of about 50 degree than that of 37-element fuel bundles. The lower fuel temperature of CANFLEX fuel bundles is expected to enhance the safety by reducing the fuel temperature coefficient. Finally, for each burnup of CANFLEX fuel bundles and 37-element fuel bundles, the equation was present for predicting the fuel temperature of a bundle in terms of a coolant temperature and bundle power

ANTERIOR CRUCIATE LIGAMENT RECONSTRUCTION USING THE DOUBLE-BUNDLE TECHNIQUE - EVALUATION IN THE BIOMECHANICS LABORATORY.

Science.gov (United States)

D'Elia, Caio Oliveira; Bitar, Alexandre Carneiro; Castropil, Wagner; Garofo, Antônio Guilherme Padovani; Cantuária, Anita Lopes; Orselli, Maria Isabel Veras; Luques, Isabela Ugo; Duarte, Marcos

2011-01-01

The objective of this study was to describe the methodology of knee rotation analysis using biomechanics laboratory instruments and to present the preliminary results from a comparative study on patients who underwent anterior cruciate ligament (ACL) reconstruction using the double-bundle technique. The protocol currently used in our laboratory was described. Three-dimensional kinematic analysis was performed and knee rotation amplitude was measured on eight normal patients (control group) and 12 patients who were operated using the double-bundle technique, by means of three tasks in the biomechanics laboratory. No significant differences between operated and non-operated sides were shown in relation to the mean amplitudes of gait, gait with change in direction or gait with change in direction when going down stairs (p > 0.13). The preliminary results did not show any difference in the double-bundle ACL reconstruction technique in relation to the contralateral side and the control group.

Parameter optimization for steel quenching by C02-laser irradiation

International Nuclear Information System (INIS)

Moryashchev, S.F.; Kislitsyn, A.A.; Kosyrev, F.K.

1984-01-01

The dependence of average absorption factor on maximal temperature of the article surface during quenching by CO 2 -laser irradiation was determined empirically. The calculations of depth of a hardening zone and process productivity in 40 Kh, 4Kh13 steels and Armco-iron with regard to this dependence were conducted

Episodic payments (bundling): PART I.

Science.gov (United States)

Jacofsky, D J

2017-10-01

Episodic, or bundled payments, is a concept now familiar to most in the healthcare arena, but the models are often misunderstood. Under a traditional fee-for-service model, each provider bills separately for their services which creates financial incentives to maximise volumes. Under a bundled payment, a single entity, often referred to as a convener (maybe the hospital, the physician group, or a third party) assumes the risk through a payer contract for all services provided within a defined episode of care, and receives a single (bundled) payment for all services provided for that episode. The time frame around the intervention is variable, but defined in advance, as are included and excluded costs. Timing of the actual payment in a bundle may either be before the episode occurs (prospective payment model), or after the end of the episode through a reconciliation (retrospective payment model). In either case, the defined costs over the defined time frame are borne by the convener. Cite this article: Bone Joint J 2017;99-B:1280-5. ©2017 The British Editorial Society of Bone & Joint Surgery.

EFFECT OF CONTROLLED QUENCHING ON THE AGING OF 2024 ALUMINUM ALLOY CONTAINING BORON

Directory of Open Access Journals (Sweden)

N. Khatami